Experimental Verification of Radiation Dose in Mixed Neutron/Gamma ...

Experimental Verification of Radiation Dose in Mixed Neutron/Gamma Radiation Fields Luka Snoj, Marjeta Šentjurc, Boštjan Črnič, Matjaž Ravnik, Robert Jeraj “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected] ABSTRACT The TRIGA research reactor at Jozef Stefan Institute is used for irradiation of various samples. The Monte Carlo code for transport of neutrons and photons, MCNP, was used to calculate dose rates in irradiation channels in the operating TRIGA research reactor. Several measurements of dose rates in individual irradiation channels were performed with CaF2 and LiF TLDs. The calculated dose rates significantly differ from the measured ones especially for the neutron dose rate. The second experimental method used was tooth enamel dosimetry. Results indicate that human teeth are suitable for radiation dose assessment in mixed neutron/gamma radiation fields with dose rates of several Gy per second. 1

INTRODUCTION

Accurate knowledge about the radiation dose is of great importance in the TRIGA research reactor at the Jozef Stefan Institute (JSI), where various samples (particle detectors, materials for future fusion reactors, biological samples, etc.) are irradiated every day. The aim of our work was to develop computational and experimental tools for radiation dose assessment in mixed neutron and photon radiation fields. The Monte Carlo code for transport of neutrons and photons, MCNP, was used to calculate dose rates in irradiation channels in the operating TRIGA research reactor. Several measurements of dose rates in individual irradiation channels were performed with CaF2 and LiF thermoluminiscent detectors (TLD). The second experimental method used was tooth enamel dosimetry. It is based on electron paramagnetic resonance measurements (EPR) of free radicals induced by ionizing radiation. The latter method is especially convenient for population dosimetry. TLD and biodosimeters were irradiated at the JSI TRIGA research reactor. The irradiation was performed in three different irradiation channels, each featuring different neutron and photon spectra and a different neutron to photon dose ratio. 2

METHODS AND RESULTS

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Description of the TRIGA research reactor

TRIGA research reactor at JSI, Ljubljana, Slovenia is a light water reactor used for training, research and isotope production. Detailed description of the reactor can be found in [1]. The configuration of the reactor core and the position of the irradiation channels are schematically depicted in Figure 1. The highest flux of 2×1013 n/(cm2·s) is achieved in the central channel at the maximum power of 250 kW. 602.1

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Figure 1: Configuration of the reactor core no. 189, in which the irradiation took place. 2.2

Monte Carlo calculations

Monte Carlo code MCNP was used for calculation of neutron and photon transport [3]. An MCNP model of the TRIGA reactor is based on the benchmark model [1]. The verification and validation of the model for calculation of the multiplication factor, neutron flux and spectra has been performed and is thoroughly described in references [1] and [2]. An eigenvalue algorithm to determine keff (effective multiplication factor of a fissile system) was used for fission process simulation. Only prompt photons are produced from neutron Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 10-13, 2007

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collisions by this algorithm in MCNP, meaning that delayed gammas are neglected. Photon spectra and the dose are therefore expected to differ from true ones. The neutron and photon spectra were calculated for free air in various irradiation channels during the reactor operation and are presented in Figure 2 and Figure 3.

Figure 2: Neutron spectra in various irradiation channels during the reactor operation. We can see in Figure 2 that the neutron spectrum strongly depends on the irradiation channel. It is interesting to note that the thermal (E < 0.625 eV) to fast (E > 100 keV) neutron ratio is the largest in the irradiation channel 40 (IC 40) in the carrousel facility in the graphite reflector surrounding the core. This is mainly due to large amounts of graphite surrounding this irradiation channels and absence of neutron sources in the vicinity of the irradiation channel. Photon spectrum in IC 40 is much "softer" than in other irradiation channels. This is mainly due to relatively large distance between IC 40 and the reactor core, in contrast to other irradiation channels, which are directly in the reactor core.

Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 10-13, 2007

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Figure 3: Photon spectra in various irradiation channels during the reactor operation. Neutron and photon energy dose rates were calculated by using the F6, heating tally feature of the code, which calculates the kinetic energy released per unit mass (kerma) in the corresponding sample. The calculated dose rates for human teeth normalised per unit of reactor power are presented in Table 1. It is interesting to note that in human teeth the photon dose rate is four (CC) to eight (IC40) times larger than the neutron dose rate. Table 1: Calculated dose rates for human teeth normalized per 1 W of reactor power. In the last column is the calculated neutron (N) to photon (P) dose rate ratio. Irradiation channel IC 40 F19 CC TC

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Normalised Dose rate [Gy/(W·s)] neutron photon total 8.3E-06 6.6E-05 7.4E-05 5.2E-05 2.7E-04 3.2E-04 2.2E-04 8.4E-04 1.1E-03 1.1E-04 5.1E-04 6.2E-04

N/P 0.13 0.19 0.26 0.22

TLD measurements

Two types of TLDs, i.e. CaF2 and LiF, were used to separate neutron and photon components. The CaF2 TLD was used for photon dose measurement and LiF was used for total (neutron and photon) dose measurement. The TLD dosimeters were the same as those used for monitoring of occupational exposure. They were irradiated in the irradiation channels of the TRIGA reactors. As it is very difficult to model the background gamma rays (delayed gamma rays from fission and activation) products in the shutdown reactor with MCNP, we used TLDs mainly to measure the background gamma rays in the irradiation channels. The background gamma dose rate strongly depends on the time from the reactor shut down after operation. Therefore the background gamma dose rate was measured twice; one and three days after shut down. Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 10-13, 2007

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Table 2: Measured gamma background dose rates with CaF2 and LiF TLDs.. Days after Irradiation shutdown position CC 1 IC 40 CC 3 IC 40 F19

Dose rate [Gy/s] LiF to CaF2 LiF CaF2 ratio 8.87E-01 3.02E+00 3.40 2.78E-02 4.07E-02 1.46 1.99E-01 2.54E-01 1.28 9.02E-03 1.00E-02 1.11 7.85E-02 8.36E-02 1.06

It can be seen from Table 2 that gamma background dose in IC 40 is approximately three times smaller than in CC, which is expected as CC is in the centre of the reactor core and IC 40 outside the core. We can also observe that LiF TLDs give systematically larger values than CaF2. The discrepancies are larger at the first experiment one day after shutdown. In the operating reactor the dose rates are very high (several Gy/s at 10000 W), which makes the dose measurements with TLD very difficult and the results are not very reliable. Therefore only one measurement of dose rate was performed in the operating reactor, namely in the IC 40, which has the lowest dose rate. The irradiation of TLDs in the IC 40 was performed at reactor power of 500 W one day after the last shutdown. 2.4

Comparison of calculated and measured dose rates

In order to compare the calculated values of dose rate with the measured ones, we also calculated the dose rate in free air. Dose rate in free air can be obtained directly from the TLD measurements by dividing the measured value by 1.14. The results are presented in Table 3. Table 3: Calculated dose and measured dose rates in free air in IC 40. Dose rate [Gy/s] Dose rate [Gy/s] at P = 0 W at P = 500 W CaF2 LiF LiF/CaF2 CaF2 LiF LiF/CaF2 measured 1.90E+00 2.82E+00 1.49 5.11E-02 9.74E-01 19.06 photon total total/photon photon total total/photon calculated 3.29E-02 3.29E-02 1.00 6.57E-02 6.99E-02 1.06

We can observe from Table 3 that the measured gamma dose rate at 500 W is approximately two times higher than the gamma background, meaning that the gamma background one day after shutdown approximately equals to the dose rate from prompt gamma rays at 500 W. There is large discrepancy between the calculated and measured values of photon and total dose rate. If we take into account the background gamma dose rate, which corresponds to the prompt gamma dose rate at 500 W, the measured dose is still two times larger than the calculated one. The differences are even larger at total (neutron and photon) dose rate. There is also large discrepancy between the measured and calculated LiF and CaF2 or photon to total dose rate ratio. The results show that the response of LiF dosimeters to neutrons is not accurately known, which leads to large discrepancies between the calculated and measured dose rates. 2.5

EPR measurements

Electron paramagnetic resonance (EPR) is a physical method of determining accumulated doses [4]. The method is based on the measurement of radiation-induced radicals Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 10-13, 2007

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in hydroxyapatite. The concentration of radiation-induced radicals and hence the intensity of the EPR signal increase proportionally to the absorbed dose from less than 30mGy to more than 10 kGy [5]. Obtaining dose-response relationship is crucial for determining the delivered dose. In our experiment several teeth (human and animal) were irradiated in two irradiation channels, namely F 19 in the core and IC 40 in the reflector. These irradiation channels were chosen because they differ a lot in the neutron and photon spectra and neutron to photon dose ratio. Firstly the irradiation was performed when the reactor was completely shut down. At that time there were almost no neutrons in the reactor, only delayed gammas were present. Afterwards the irradiation of teeth was performed also in F19, TC and IC 40 at different reactor power levels. After the irradiation the teeth were sent to the Laboratory for EPR spectroscopy, where the EPR measurements were performed. The measured EPR spectra are presented in Figure 4. The height of the peak is proportional to the EPR signal intensity and to the assessed dose. It can be seen from Figure 4 that EPR signal intensity is very weak and difficult to distinguish at 3.6 Gy.

Figure 4: EPR spectra (rel. units) of irradiated teeth at various assessed dose. EPR signal intensity as a function of assessed dose in F 19 and IC 40 is presented in Figure 5. We can observe that the EPR signal intensity is linearly proportional to the assessed dose. The assessed dose was calculated with MCNP and the gamma background was taken into account by increasing the photon dose rate for the value calculated at 500 W. We can observe that the EPR signal intensity is approximately proportional to the assessed dose.


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Figure 5: EPR signal intensity as a function of assessed dose. The Y error bars represent the error in EPR signal measurement and the X error bars represent the systematic error in calculated dose due to normalization to reactor power and gamma background modeling 3

CONCLUSION

We have observed that the discrepancies between the calculated and measured dose are large. As the MCNP model of the TRIGA reactor has been verified and validated for neutron transport, further work should be focused on improving the accuracy of the neutron dose measurement and improving the model for the photon transport. At lower doses the EPR signal intensity is very weak and the relative error is very large, meaning that our methods are still not good enough for non-lethal dose measurements. Results show linear relation between the measured and delivered dose for both type of dosimeters. Currently the EPR dosimetry can be used as a tool for measuring high doses, such as ones met in or near critical fissile systems. In the future we will develop very accurate methodological approach to eliminate systematic uncertainties and verify the calculations by the use of different dosimeters appropriate for measuring high neutron and photon dose rates separately. Preliminary results show that LiF is 10 times more sensitive to neutrons than CaF2. This should be verified by other experiments. Further studies are needed to confirm this hypothesis. LiF can not be used without detailed knowledge of the response function.

ACKNOWLEDGMENTS The authors of this paper gratefully acknowledge prof. dr. Milan Petelin from the Ljubljana Clinic of stomatology for providing us a considerable amount of human teeth for our experiments. Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 10-13, 2007

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REFERENCES [1] R. Jeraj and M. Ravnik, "TRIGA Mark II benchmark Critical Experiments-Fresh Fuel, IEU_COMP-THERM-003", International Handbook of Evaluated Critical Safety Benchmark Experiments, Organization for Economic Cooperation and Development Nuclear Energy Agency Data Bank, 1999 [2] A. Trkov, L. Snoj, P. Rogan, R. Jačimovič, M .Ravnik, "On the use of computational methods for the optimization of research reactor utilization", PHYTRA1: International Conference on Physics and Technology of Reactors and Applications, Marrakech, Morocco, March 14-16, GMTR, 2007 [3] J.J. Briesmeister, "MCNP5- A General Monte Carlo N-Particle Transport Code, Version 5 Los Alamos National Laboratory", March, 2005 [4] M. Bhat, EPR tooth dosimetry as a tool for validation of retrospective doses: an end-user perspective, Applied Radiation an Isotopes, 62, 2005, pp. 155-161 [5] R. Gruen, Errors in dose assessment introduced by the use of the "linear part" of a saturating dose response curve. Radiat. Meas, 26, 1996, pp. 297-302.
