Energy response for high-energy neutrons of multi-functional electronic personal dosemeter

Energy response for high-energy neutrons of multi-functional electronic personal dosemeter T. Nunomiya 1, T. Ishikura 1, O. Ueda 1, N. Tsujimura 2,, M. Sasaki 2,, T. Nakamura 1,2 1 Fuji Electric Systems Co., Ltd., Fuji-machi 1, Hino-city, Tokyo, 191-8502 Japan E-mail : nunomiya-tomoya@fesys.co.jp 2 CYRIC, Tohoku University, Aramaki, Aoba, Sendai, 980-8579 Japan Abstract. We have developed a small, light, and multi-functional electronic personal dosimeter (EPD) using four semiconductor radiation detectors for dose monitoring of workers at nuclear power plants and high-energy accelerator facilities. This dosimeter is about 103 x 55 x 15 mm 3 in size and approximately 110 g in weight, and capable of measuring gamma-ray and neutron doses in Hp(10), and beta-ray dose in Hp(0.07), simultaneously. The neutron dose equivalent is obtained by using two-types of silicon semiconductors; a slow-neutron sensor (below 1 MeV) and a fast-neutron sensor (above 1 MeV). In this work, the response for high-energy neutrons above 15 MeV is experimentally investigated using a quasi-monoenergetic neutron field with the EPD, NRY21, of Fuji Electric Systems Co., Ltd, which is gamma-ray and neutron dosemeter and it was confirmed to have good sensitivity for high-energy neutrons as well as the response for 15 MeV neutrons. 1. Introduction Recently, high-energy accelerators are in increasing use for various purposes such as nuclear physics, medical science, and engineering research. Proton and heavy-ion accelerators of high beam-intensities are also being constructed for intense neutron sources and radioactive beam sources. Since high-energy neutrons above 10 MeV can be generated in these facilities, it becomes very important to monitor the neutron dose equivalents of workers in such facilities by using a real-time personal neutron dosemeter which has a good sensitivity to such high-energy neutrons. A real-time personal dosemeter that could estimate the neutron dose equivalent over a wide energy range from thermal to a few dozens of MeV was first developed [1, 2] by our group using two silicon neutron sensors, which is now on sale by Fuji Electric Systems Co., Ltd. Very recently, similar real-time personal neutron dosemeters using silicon sensors have been developed by PTB (Physikalisch-Technische Bundesanstalt in Germany) [3], and SAPHYMO [4]. The performance tests of Fuji-Electric dosemeter developed by our group in actual neutron fields with various energy spectra have been performed in the six fields at four facilities of CYRIC (Cyclotron and Radioisotope Center) of Tohoku University, CERN, KEK (High Energy Accelerator Research Organization) and PNC (Power Reactor and Nuclear Fuel Development Corporation). The obtained results are in good agreement with the ambient dose equivalent H*(10) within a factor of 2 [1, 2]. Present address Japan Nuclear Cycle Development Institute, Muramatsu 4-33, Tokai-mura, Naka-gun, Ibaraki, 319-11 Japan Central Research Institute of Electric Power Industries, Iwato-kita 2-11-1, Komae-city, Tokyo, 201-8511 Japan 1

However, the sensitivity of this dosemeter to neutrons of energy above 20 MeV has not been clarified yet, therefore, in this work, we measured the energy responses of quasi-monoenergetic 31.5 and 64.5 MeV neutrons produced through the Li(p, n) reaction from a thin Li target bombarded by 35 and 70 MeV protons at the CYRIC. 2. Experiment 2-1. Dosimeter description The dosemeter, NRY21, uses two silicon semiconductor detectors of a fast neutron sensor and a thermal neutron sensor. Each sensor has a 10 x 10 mm 2 p-type low-resistivity silicon wafer on which an amorphous silicon was deposited, as shown in Fig. 1. Low-resistivity silicon is used to reduce the gamma-ray sensitivity by reducing the thickness of the deposition layer. For the fast neutron sensor, a polyethylene radiator of about 0.8-mm thickness is inserted in the direction of neutron incidence, and this sensor acts as a recoil proton detector. The slow neutron sensor consists of amorphous silicon on which natural boron is doped in order to detect low-energy neutrons by using the 10 B(n, α)li reaction. Neutron dose equivalent can be determined by summing up the counts above the discrimination level, which is about 900 kev. This discrimination level is determined to discriminate the gamma-ray event pulses from 137 Cs source at 1 Sv/h in free air [1]. The fast neutron sensor is used to measure neutrons with energies above 1 MeV and slow neutron sensor with energies below 1 MeV. But for neutrons of energies higher than 10 MeV, neutrons can produce charged particles of protons and alpha particles through the direct silicon reactions such as Si(n, p) and Si(n, α) reactions, then the dosemeter may have high sensitivity with increasing the neutron energy. 2-2. Experimental geometry Fig. 2 shows the experimental geometry at CYRIC. Quasi-monoenergetic neutrons are generated at a 1.88 or 6.03 mm-thick natural Li target bombarded by high-energy protons accelerated up to 35 MeV or 70 MeV with the AVF cyclotron, respectively. Source neutrons emitted in the forward direction reach an experimental room through a 7 x 7 x 30 cm 3, 12 x 12 x 30 cm 3, 17 x 17 x 30 cm 3, 21 x 19.5 x 30 cm 3 and 30 x 25 x 30 cm 3 tapered concrete collimator surrounded by concrete blocks of 100-cm thickness and through the aperture (100-cm wide x 50-cm high) in a concrete wall of 283-cm thickness between the 5 th target room and the TOF room [5]. The protons penetrating through the target were bent down toward a beam dump by a clearing magnet. A copper plate of 0.5-mm thickness was set at the entrance of the primary collimator as a neutral proton beam stopper [6]. The dosemeter was put on the acrylic phantom of 40 x 40 x 15 cm 3 placed on the beam axis at 1140-cm downstream from the Li target. Proton beam intensity was measured during the experiment by a Faraday cup in the beam dump. The irradiation time was decided in order to obtain the neutron fluence beyond 10 7 cm -2, which was about 7 and 16 hour for 31.5 and 64.5 MeV quasi-monoenergetic neutrons, respectively. 2

(a) Fast neutron sensor (b) Slow neutron sensor FIG. 1. Cross-sectional view of the fast and slow neutron sensors. FIG. 2. Experimental geometry at CYRIC of Tohoku University. 3. Analysis The spectra of source neutrons above 6 MeV were measured by the time of flight (TOF) method with the 5.08-cm diam by 5.08-cm long BC501A organic liquid scintillation detector [5]. The results are shown in Fig. 3. The peak energy of quasi-monoenergetic neutrons is 31.5 and 64.5 MeV, respectively for 35 and 70 MeV proton, and continuous tail component extends down to about 6 MeV, which comes from several break-up reactions. Neutron fluence of the monoenergetic peak can be obtained from the measured neutron energy spectrum, which is from 28 to 35 MeV and from 58 to 70 MeV for 31.5 and 64.5 MeV quasi-monoenergetic neutrons, respectively. The low energy component below 6 MeV is estimated by the extrapolation under an assumption that the energy spectrum keeps constant below 6 MeV (see Fig. 3). 3

The dosemeter can measure the neutron dose equivalent by adding the dose equivalent from thermal to 1 MeV neutron, H s, and the dose equivalent above 1 MeV neutron, H f, which are measured by the slow and fast neutron sensors, respectively. The total dose equivalent, H t, is expressed as follows, H t = H f + H s = ac f + bc s, (1) where C f and C s are the counts of fast and slow neutron sensor above gamma-ray cut off levels (900 kev). The conversion coefficients a and b are given by [1], a = b = thermal H p (10, E) φ(e)de, R (E) φ(e)de thermal f H p (10, E) φ(e)de, R (E) φ(e)de s (2) (3) where R f (E) and R s (E) are the respective energy responses of fast and slow neutron sensors and Hp(10,E) is the fluence-to-personal dose equivalent conversion factor given in ICRP 74 [7]. Since the thus-obtained quasi-monoenergetic neutrons have a continuous tail component in the spectra, it is necessary to subtract the contribution from this neutron component in order to get the energy response to monoenergetic peak neutrons of 31.5 and 64.5 MeV, R 31.5 and R 64.5, respectively. The subtraction method is given as, R 64.5 = H t 64.5 - A - B - C, A = R 1.0 F a, B = R 5.0 F b, C = R 31.5 F c, R 31.5 = H t 31.5 - A - B, (4) where F a, F b and F c are neutron fluences below 1 MeV, 1 to 27 MeV and 27 to 58 MeV, respectively, 31.5 H t and H 64.5 t are total neutron doses of 31.5 and 64.5 MeV quasi-monoenergetic neutrons obtained by Eq. (1). R 1.0 and R 5.0 is the energy response of monoenergetic neutrons of 1 MeV and 5 MeV, respectively, obtained at the Fast Neutron Laboratory (FNL) of Tohoku University [1,2]. Fig.4 shows the schematic diagram of this subtraction method. 4. Results and discussions The energy responses for high-energy neutrons of 31.5 MeV and 64.5 MeV, R 31.5 and R 64.5, are obtained by using Eq.(4) and are tabulated in Table 2 with the neutron fluence and parameters of A, B and C. Total 31.5 neutron dose equivalents obtained by the dosemeter, H t and H 64.5 t, are 64.7 and 280 msv for 31.5 and 64.5 MeV quasi-monoenergetic neutrons, respectively. The energy responses, R 31.5 and R 64.5, are then 2.67 and 2.93 nsv cm 2, respectively, which is a little larger than 2.08 nsv cm 2 for 15 MeV.. 4

FIG. 3. Neutron energy spectra at CYRIC of Tohoku University obtained by Hagiwara et al. [5]. A B C 31.5 MeV 64.5 MeV 6 MeV 1 MeV 27 MeV 58 MeV FIG. 4. Schematic diagram of energy response estimation for 31.5 and 64.5 MeV peak components of quasi-monoenergetic neutrons. Neutron energy [MeV] Table 2 Energy responses to 31.5 and 64.5 MeV quasi-monoenergetic neutrons. H t Peak Response A B C fluence (R 31.5, R 64.5 ) [msv] [msv] [msv] [msv] [n cm -2 ] [nsv cm 2 ] *31.5 ** 6.47E+01 6.30E-02 1.87E+01-1.72E+07 2.67 64.5 2.80E+02 6.66E-02 3.35E+01 1.19E+02 4.37E+07 2.93 * Peak energy of quasi-monoenergetic neutrons ** Read as 6.47 x 10 1 5

Since the neutron sensor consists of amorphous-silicon semiconductor and thin plastic radiator, it implies that in addition to the H(n, p) reaction in the polyethylene radiator, the Si(n,α) and Si(n, p) reactions increasingly contribute to the response in this energy region. From this result, this dosemeter has good sensitivity for high-energy neutrons from 31.5 MeV to 64.5 MeV. The energy responses, R 31.5 and R 64.5, given by Eq. (4) are shown in Fig. 5 as a function of neutron energy together with the energy responses from 8 kev to 15 MeV. The values for neutron energies between 8 kev to 15 MeV have been obtained by using monoenergetic neutron beams at FNL in the previous works [1, 2], and those for 31.5 and 64.5 MeV were obtained here in this work. The Hp(10) values are adjusted to be equal to the experimental values at 2 MeV, because this dosemeter was calibrated by using 252 Cf having an average neutron energy of 2.13 MeV. In Fig. 5, it can be seen that this dosemeter have rather lower response for neutron energy from about a few hundred kev to 1 MeV by comparing to Hp(10) values. Since the fast neutron sensor of this dosemeter uses the recoil protons from polyethylene radiator, the pulse heights of these recoil protons are lower than the discrimination level of 900 kev in this energy region. For slow neutron sensor, it uses the 10 B(n, α)li reaction and the reaction cross section becomes smaller with increasing the neutron energy. It is very important to improve this poor response for several hundred kev neutrons. The neutron detection efficiency above 15 MeV, on the other hand, increases up to 64.5 MeV because of Si(n, p) and Si(n, α) reactions. With increasing the neutron energy, the pulses from the protons and alpha particles produced by these reactions become larger enough to over the discrimination level. From this result, it is confirmed that this dosemeter, NRY21, have a good sensitivity for high-energy neutrons above 15 MeV. FIG. 5. Comparison with energy response of the Fuji-Electric EPD, NRY21, as a function of neutron energy and personal dose equivalent conversion factor Hp(10) described in ICRP 74. 6

5. Conclusion The energy response for high-energy neutrons of 31.5 and 64.5 MeV with the EPD, NRY21, of Fuji Electric Systems Co., Ltd, which is gamma-ray and neutron dosemeter, are obtained using a quasi-monoenergetic neutrons of Li(p, n) reaction, Ep= 35 and 70 MeV at the Cyclotron and Radioisotope Center (CYRIC) of Tohoku University. In this energy region, where the Si(n, p) and Si(n,α) reactions increase, this dosemeter has a sensitivity of 2.67 and 2.93 nsv cm -2 for 31.5 MeV and 64.5 MeV, respectively, which is a little larger than 2.08 nsv cm -2 for 15 MeV. Acknowledgement The authors wish to thank the operating staffs of the CYRIC for operating the cyclotron and thank the staffs of Prof. Baba Laboratory for their great help during the experiment. Reference 1. M. Sasaki, T. Nakamura, N. Tsujimura, O. Ueda, T. Suzuki, Development and characterization of real-time personal dosemeter with two silicons, Nucl. Instr. and Meth. A. 418 465-475 (1998). 2. N. Tujimura, Characteristic evaluation and standard calibration of Si-type real-time personal dosemeter, Master s thesis of Tohoku University, Feb. 1993 (in Japanese). 3. M. L. Bhadra, W. Wendt, M. Weierganz, The electronic neutron/photon dosemeter PTB DOS-2002, Proceedings of 9 th Neutron Dosimetry Symposium, Delft University Of Technology, The Netherlands, 28 Sep 3 Oct 2003, (to be published in Radiat. Protec. Dosim.). 4. T. Lahaye, Q. Chau, S. Menard, M. N. Moyo, T. B. Milsztajn and A. Rannou. Numerical and experimental results of the personal neutron dosemeter SAPHYDOSE-N, Proceedings of 9 th Neutron Dosimetry Symposium, Delft University Of Technology, The Netherlands, 28 Sep 3 Oct 2003, (to be published in Radiat. Protec. Dosim.). 5. M. Hagiwara, A study of secondary particle production and activation induced by particles of tens of MeV, Master s thesis of Tohoku University (in Japanese), March 2003. 6. Y. Nakane, Y. Sakamoto, Measurement of absorbed dose distributions in a plastic phantom irradiated by 40- and 65-MeV quasi-monoenergetic neutrons, Nucl. Instr. and Meth. A. 459 552-564 (2001). 7. International Commission on Radiological Protection, Conversion Coefficients for use in Radiological Protection against External Radiation, ICRP Publication 74, (1995). 7