A new neutron monitor for pulsed fields at high-energy accelerators

A new neutron monitor for pulsed fields at high-energy accelerators Marlies Luszik-Bhadra *, Eike Hohmann Physikalisch-Technische Bundesanstalt, Bundesallee 100, D-38116, Braunschweig, Germany. Abstract. This paper describes a neutron monitor which consists of a spherical neutron moderator, 12 in diameter, with a thermal neutron detector which detects delayed β-radiation from silver activation. The thermal neutron detector is based on four silicon diodes at the centre of the moderator, two of which are covered by silver foils, 250 µm thick, and two by tin foils, 360 µm thick. The latter two diodes are used to subtract the photon contribution. The system registers chiefly counts of β-rays from the 109 Ag activation with a 25 s half-life. The performance of the neutron monitor in neutron and photon calibration fields is described by measurements and calculations. KEYWORDS: silver activation; neutron monitor; pulsed radiation; high-energy neutrons. 1. Introduction Recently, it has been shown that short-term pulse losses at the high-energy accelerator HERA at DESY can produce neutron dose increments in passive dosemeters (such as TLDs in a polyethylene cylinder) of about 1 msv, which are underestimated by conventional, active Anderson-Braun counters by several orders of magnitude (1) due to dead-time problems with these counters. In such cases, active devices based on the principle of measuring delayed activation products can improve the situation. Measurements on delayed nuclei resulting from reactions with 12 C have been proposed by Leuschner [1] and Klett [2]. However, the corresponding endothermic reactions have high-energy thresholds of about 13 MeV, which can underestimate an appreciable part of the dose contribution. We have built a new monitor which is able to measure neutrons in pulsed fields in the energy range from thermal up to about 100 MeV. It is based on 4 silicon diodes in the centre of a 12 polyethylene sphere. Two of the diodes are covered by converter foils, 250 µm thick and made of natural silver. Once thermalized, the neutrons can undergo nuclear capture by the silver foil, producing 108 Ag and 110 Ag daughter nuclei, which decay with half-lives of 2.4 min and 25 s via β-radiation with maximum energies of 1.7 MeV and 2.9 MeV, respectively. This β-radiation can be well detected by the silicon diodes, but they have also a high sensitivity to the accompanying photon radiation. Therefore, the other two silicon diodes are covered by tin layers, 360 µm thick - a material with similar absorption properties for photons, but with a negligible thermal neutron capture cross section and their signals are used to subtract the photon contribution. The new monitor has a more compact and symmetric converter/detector set-up than earlier neutron monitors based on silver- and tin-wrapped G-M tubes [3], and this reduces the uncertainty of the subtraction method. This paper presents the set-up of the device, the results of measurements in several quasi monoenergetic neutron fields (thermal, 144 kev, 250 kev, 565 kev, 1.2 MeV, 2.5 MeV, 5.0 MeV, 8.0 MeV, 14.8 MeV), neutron fields produced by radionuclide sources ( 252 Cf(bare), 252 Cf(D 2 O mod.), 241 Am-Be), and in several photon fields ( 137 Cs, 60 Co, 6-7 MeV high-energy photons). In order to estimate the response at higher neutron energies, calculations using the MCNPX code were performed up to 1 GeV. * Presenting author, E-mail: marlies.luszik-bhadra@ptb.de 1

2. Set-up of the monitor The monitor consists of a 12 polyethylene sphere with a thermal neutron counter in its centre. A sketch of the thermal neutron detector is shown in Fig. 1. It consists of a cylindrical aluminium capsule (diameter: 40 mm, length: 45 mm, wall thickness: 1 mm) which contains four silicon diodes (each 8.5 mm x 10.6 mm x 0.48 mm), two of which are covered on both sides either by silver foils, 250 µm thick, or tin foils, 360 µm thick. The silver foil thickness corresponds to the approximate range of the mean energy beta particles in silver. Plastic holders allow the diodes to be fixed in the centre of the moderator sphere. Behind the diode/converter arrangement, four preamplifiers are placed which provide a first amplification of the signals. Cables feed the signals of the preamplifiers as well as the diode bias (80 V for a full depletion thickness of 0.48 mm) and the preamplifier voltage (+/-12 V) through an aluminium tube (diameter: 12 mm, wall thickness: 1 mm) to the electronics outside the sphere, i.e. main amplifiers, a multi-channel analysing system and a personal computer. Figure 1: A cut through the thermal neutron detector. 3. Nuclear reactions and detection The polyethylene sphere, 30 cm in diameter, is used to thermalize neutrons. Once thermalized, the neutrons can undergo nuclear capture by the silver nuclei according to the following main reactions [4,5]: 109 Ag + n γ + 110 Ag 110 Cd +β - (E max = 2.9 MeV) (thermal neutron cross section 90.5 b, half-life of 110 Ag is 25 s) 107 Ag + n γ + 108 Ag 108 Cd +β - (E max = 1.7 MeV) (thermal neutron cross section 38.6 b, half-life of 108 Ag is 144 s) 109 Ag and 107 Ag contribute with similar isotopic abundance (48% and 52%, respectively) to natural silver. In addition to delayed beta counts, the diodes covered by silver register counts from prompt gamma rays of the two reactions shown above. Both sets of diodes Ag-covered and Sn-covered detect photons which are present in the external field, but also photons which are produced by the thermal neutron capture reactions 12 C(n, nγ) 12 C and H(n, γ)d within the polyethylene material of the spherical moderator (chiefly 2.2 MeV). 2

In order to allow an accurate subtraction of the gamma counts, a tin foil with the same gamma attenuation as the silver foil is used and diodes are arranged in a compact and symmetrical way (see Fig. 1). 4. Pulse height spectra The sensitivity and performance of the monitor depends on an accurate calibration of all four diodes and the number of events which are counted in a selected pulse height interval. Figure 2 shows a typical pulse height spectrum as measured with neutrons using a bare 252 Cf source. The number of counts per µsv is given in terms of ambient dose equivalent H*(10) [6]. The diode covered by silver showed count numbers higher by about a factor of two than the diode covered by tin. The shaded pulse height region between 0.662 MeV and 1.0 MeV was used for neutron monitoring in order to have no sensitivity to low-energy photons. Figure 3 shows pulse height spectra as measured with one Sn-covered diode and with photon radionuclide sources available at PTB ( 137 Cs, 60 Co) and also with 6-7 MeV photon radiation produced at the PTB accelerator by the nuclear reaction 19 F(p,αγ) [7]. Using the selected pulse height interval, the dose equivalent sensitivity for 60 Co and 6-7 MeV photon radiation is one order of magnitude higher than the neutron sensitivity (as indicated by the number of counts in the shaded regions of Figures 2 and 3), whereas the sensitivity for 137 Cs photons is almost zero. The well-defined Compton edge produced by irradiation with 137 Cs photons has been used for the energy calibration of pulse height spectra for all four detectors. In this case, Sn-covered and Ag-covered diodes show pulse height spectra which do not differ within their statistical uncertainties. Figure 2: Pulse height spectra measured with the Ag- and Sn-covered diodes with a bare 252 Cf source (see text). 10 1 H*(10) response / µsv -1 10 0 10-1 10-2 10-3 252 Cf(bare) Ag Sn 0 200 400 600 800 1000 1200 Pulse height / kev 3

Figure 3: Pulse height spectra measured with one Sn-covered diode for 137 Cs, 60 Co and 6-7 MeV photon irradiation (see text). 10 2 137 Cs H*(10) response / µsv -1 10 1 10 0 10-1 60 Co 6-7 MeV photons 10-2 0 200 400 600 800 1000 1200 Pulse height / kev 5. Time dependencies In order to more accurately investigate the delayed and prompt response of the detector system as a function of time, an irradiation had been performed with neutrons from a 252 Cf source and the number of counts with pulse height corresponding to an energy loss between 662 kev and 1 MeV per 10 s was registered. Figure 4 shows the number of counts as measured in the Ag-covered diodes subtracted by the number of counts as measured in the Sn-covered diodes in an equilibrium situation (after about 20 minutes of radiation) and the following decay. A decay curve consisting of two half-lives (25 s and 144 s) can be well fitted to the data. The fraction of the different reactions contributing to the saturation count rate can be deduced from the decay. Since the relative abundance of the silver isotopes is comparable, the sensitivity of the reaction with shorter half-lives is dominant due to the higher cross section (higher by factor 3), but also due to the higher maximum energy of beta radiation (2.9 MeV as compared to 1.7 MeV) and the selected pulse height interval for particle registration. The results of the measurement shown in Figure 4 give a higher intensity of the short half-life reaction by a factor of eight. This is considerably higher than the value of five reported for the monitor with silver- and tin-wrapped G-M tubes [3]. In addition, a prompt response from the Ag(n,γ)Ag reaction is undetectable within the statistical uncertainty. This is much lower than the value reported for the monitor with G-M tubes where a prompt photon contribution of 20% had been observed. In pulsed fields with high dose rates in a short pulse, this photon component may be erased due to dead-time effects. Both effects enhancement of counts of the short half-life reaction and lower prompt gamma contribution as compared to the monitor with G- M tubes are chiefly a result of the high thresholds set for pulse height registration. The number of counts per dose equivalent related to the different reactions and measured in the Agand Sn-covered diodes is given in Table 1. Subtracting counts of the Sn-covered diodes from those of the Ag-covered diodes yields a neutron response of (9.0±0.4) counts per µsv for the bare 252 Cf source. 4

Figure 4: The number of counts with statistical uncertainty (one standard deviation) in intervals of 10 s as measured after irradiation with a 252 Cf source (see text). The total decay (black line) divides into the fraction caused by 110 Ag (dotted line) and by 108 Ag decay (dashed line). 1000 Counts per 10 s 100 10 source on source off 1-300 -200-100 0 100 200 300 400 500 t / s Table 1: The number of counts per µsv in terms of H*(10) for Ag-covered and Sn-covered diodes as measured with a bare 252 Cf source. Ag-covered / µsv -1 Sn-covered / µsv -1 Beta counts from 109 Ag activation 7.8 ± 0.3 0 Beta counts from 107 Ag activation 1.0 ± 0.3 0 Gamma counts from Ag(n, γ)ag reactions 0.2 ± 0.3 0 Gamma counts from 12 C(n, nγ) 12 C and H(n, γ)d 5.6 ± 0.3 5.6 ± 0.2 reactions in PE and 252 Cf γ-rays Total 14.6 ± 0.3 5.6 ± 0.2 6. Evaluated response for various calibration fields Measurements have been performed with high-energy photons ( 60 Co and 6-7 MeV [7]) in a thermal neutron field available at the GKSS in Hamburg, Germany [8,9], in mono-energetic neutron reference fields as produced at the PTB accelerator in the neutron energy range from 144 kev to 14.8 MeV [8] and in neutron fields with broad spectral distributions using the radionuclide sources 252 Cf(bare), 252 Cf(D 2 O moderated) and 241 Am-Be [10]. The contributions of scattered neutrons have been subtracted using shadow cone (or shadow block) measurements. Table 2 shows the number of counts per dose equivalent as registered in the Ag- and Sn-covered diodes. In terms of the dose equivalent, the sensitivity of the Ag-covered diode to high-energy photons is roughly one order of magnitude higher than the sensitivity to neutrons. Although the detector in the centre of the monitor has been set up in a compact way and foils providing similar attenuation for photons were used, the response can change if the radiation comes from different directions. In order to examine this effect, the monitor has been irradiated with high-energy photons impinging onto the 5

device from different directions. The deviation in the number of counts as registered in the Ag-covered and Sn-covered diodes is in most cases in the order of the statistical uncertainty and in the order of 0.5 to 1% (see the neutron response for photon radiations in Table 2). This makes in-field calibrations as proposed earlier for a monitor with G-M tubes [3] unnecessary. In the case of neutron radiation sources, the highest contributions of photons were found in the thermal neutron field and in the 252 Cf(D 2 O moderated) neutron field. Obviously in these cases, photons generated by thermal neutron activation via 12 C(n, nγ) 12 C and H(n, γ)d within the materials of the moderator sphere contribute to a higher extent. The measured neutron response varies between 5.2 and 11.4 counts µsv -1. The response in the intermediate neutron energy region and for energies higher than 14.8 MeV is determined by calculations (see chapter 8). Table 2: The number of counts per µsv in terms of H*(10) for Ag-covered and Sn-covered diodes and neutron response (difference of counts registered in Ag-covered and Sn-covered diodes) as measured in various calibration fields. The statistical uncertainty (one standard deviation) is indicated in all cases. Sn-covered Neutron response Radiation source Ag-covered / µsv -1 / µsv -1 / µsv -1 60 Co, 0 112.3 ± 0.3 113.5 ± 0.3-1.2 ± 0.4 60 Co, 90 right 87.5 ± 0.3 87.2 ± 0.3 0.4 ± 0.4 60 Co, 90 left 82.6 ± 0.3 81.4 ± 0.3 1.2 ± 0.4 60 Co, 90 top 100.5 ± 0.3 101.3 ± 0.3-0.8 ± 0.4 60 Co, 90 bottom 103.8 ± 0.3 102.5 ± 0.3 1.2 ± 0.4 6-7 MeV gamma, 0 195.6 ± 1.0 196.7 ± 1.0-1.1 ± 1.4 6-7 MeV gamma, 90 right 181.3 ± 1.0 180.1 ± 1.0 1.2 ± 1.3 6-7 MeV gamma, 90 left 179.0 ± 1.0 177.3 ± 0.9 1.6 ± 1.3 6-7 MeV gamma, 90 top 181.0 ± 1.0 181.6 ± 1.0-0.7 ± 1.3 6-7 MeV gamma, 90 bottom 176.6 ± 0.9 175.2 ± 0.9 1.4 ± 1.3 Thermal neutrons 18.4 ± 0.3 12.5 ± 0.3 5.9 ± 0.5 144 kev neutrons 10.9 ± 0.2 4.3 ± 0.1 6.6 ± 0.3 250 kev neutrons 9.3 ± 0.2 3.5 ± 0.1 5.8 ± 0.2 565 kev neutrons 8.3 ± 0.1 2.5 ± 0.04 5.8 ± 0.1 1.2 MeV neutrons 10.1 ± 0.05 2.2 ± 0.02 7.9 ± 0.1 2.5 MeV neutrons 14.1 ± 0.04 2.7 ± 0.02 11.4 ± 0.1 5.0 MeV neutrons 14.2 ± 0.06 3.2 ± 0.03 11.0 ± 0.1 8.0 MeV neutrons 13.3 ± 0.05 3.4 ± 0.03 9.9 ± 0.1 14.8 MeV neutrons 7.5 ± 0.04 2.3 ± 0.02 5.2 ± 0.1 252 Cf(bare) 14.6 ± 0.3 5.6 ± 0.2 9.0 ± 0.4 252 Cf(D 2 O moderated) 25.0 ± 0.9 14.7 ± 0.51 10.3 ± 1.2 241 Am-Be 10.3 ± 0.3 4.1 ± 0.2 6.2 ± 0.4 7. Uncertainty in mixed n/γ fields The uncertainty for the neutron dose determination can increase, if, in addition to neutrons, highenergy photons contribute considerably to the dose equivalent in the external field. The number of counts as registered in the various detectors has been estimated for different contributions of 60 Co photon radiation and a neutron radiation field with a neutron response of 9 counts µsv -1. The standard 6

deviation caused by the count numbers calculated for different neutron dose equivalent values is shown in Fig. 5. The uncertainty for measuring 1 msv neutron dose equivalent approaches 20%, if 60 Co photons contribute the same dose equivalent. It can be appreciably higher if lower neutron doses need to be measured (see Figure 5). A neutron dose equivalent of 10 µsv is only measurable in a reliable way (uncertainty < 20%), if the contribution of high-energy photons is a factor of ten lower than that of neutrons. Figure 5: The relative statistical uncertainty (one standard deviation) for neutron dose equivalent, taking into account different contributions of 60 Co photon radiation and a neutron response of 9 counts µsv -1. 100 Relative statistical uncertainty / % 10 1 H γ /H n = 1.0 H γ /H n = 0.5 H γ /H n = 0.2 H γ /H n = 0.1 10 100 1000 Neutron dose equivalent / µsv 8. MCNPX calculations The response has been calculated by the Monte Carlo code MCNPX [11]. With this code, the number of 107,109 Ag(n,γ) reactions has been calculated in the silver layers. It is assumed that this number is proportional to the number of delayed electrons which are measured in the silicon diodes. The calculated response has been normalized to the measured one at 1 MeV. The measured and calculated data agree well up to 14.8 MeV. In case of high-energy neutrons (100 MeV), the response is a factor of three too low. This may be acceptable, since the dose equivalent contribution of high-energy neutrons is usually comparable or smaller than that of lower-energy neutrons in accelerator fields [12]. For comparison, also the response of the LB6411 neutron monitor is shown in Figure 6. The responses are similar for high-energy neutrons, but deviate for thermal and intermediate-energy neutrons, where the LB6411 shows lower values. 7

Figure 6: The response in terms of H*(10) as measured and calculated by MCNPX. The uncertainties indicated for the measured data correspond to one standard deviation and include statistical uncertainties and uncertainties of the reference values [see reference 13].For comparison, also the response of the LB6411 neutron monitor [14] is shown. 10 H*(10) response 1 Measured, This work LB6411 Calculated by MCNPX 0,1 10-8 10-6 10-4 10-2 10 0 10 2 E n / MeV 9. Conclusion The neutron monitor as proposed by us can measure the dose equivalent with a mean sensitivity of 9 counts/µsv. The detection of delayed radiation and the proposed subtraction procedure allow measurements in fields with pulsed and also with continuous radiation. The lower detection limit for neutrons depends on the presence of high-energy photons (10 µsv, if photons with energies higher than 662 kev contribute less than 10%). The photon subtraction is balanced in such a way that in-field calibrations as proposed earlier for a monitor with G-M tubes are unnecessary. A high pulse height threshold makes the device chiefly sensitive to reaction particles from the silver isotope 109 Ag with the shorter half-life of 25 s. Through the use of intelligent computer algorithms using count rate increments of count rates, we expect that an increasing dose level can be recognized within a few seconds. Acknowledgements The authors wish to thank the staff of the PTB calibration facilities and of the GKSS facility for their assistance in the measurements. 8

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