Delayed Gamma Ray Modeling Around Irradiated JSI TRIGA Fuel Element by R2S Method

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1 Delayed Gamma Ray Modeling Around Irradiated JSI TRIGA Fuel Element by R2S Method ABSTRACT Klemen Ambrožič Jožef Stefan Institute Jamova cesta Ljubljana, Slovenia Luka Snoj Jožef Stefan Institute Jamova cesta Ljubljana, Slovenia The JSI TRIGA reactor has several sample irradiation facilities with well characterized neutron fields by both simulations and measurements data. Because of this, JSI TRIGA has become a reference center for neutron irradiation of detector for ATLAS experiment (CERN). However, γ-ray characterization of irradiation facilities is yet to be performed. Monte Carlo γ-ray characterizations up to date have accounted for prompt γ-rays, neglecting delayed γ contribution. Measurements suggest that delayed γ-rays may contribute to up to 30 % of total γ flux, and is the only contribution of γ-rays after reactor shutdown. Initial steps in accounting for delayed γ-rays using R2S method have been performed and described in this paper. A Rigorous two-step method (R2S) described here utilizes Monte Carlo particle transport code MCNP 6.1 for particle transport part of the problem, and FISPACT-II for neutron activation and delayed γ-ray generation, and custom Python scripts, joining the two codes. An example on the use of code is presented, in terms of evaluation on utilization of activated nuclear fuel as a viable γ-ray source in sample irradiation. For this example, fresh nuclear fuel is considered and a silicon pipe sample modeled in the computational model. Two nuclear fuel activation time regimes were simulated, short (several hours) and long term (in terms of 1 MWd and 10 MWd fuel burn-up). 1 INTRODUCTION The Jožef Stefan Institute (IJS) TRIGA reactor is a 250 kw pool type reactor featuring numerous irradiation facilities with different characteristics in terms of size and neutron field properties [1]. Due to good neutron characterization, both computationally [1] and with measurements [2], the reactor has become a reference facility for neutron irradiation of ATLAS experiment (CERN) detectors [3]. Currently γ characterization of the reactor is also being performed. Prompt γ characterization is being performed using the Monte Carlo MCNP 6.1 code [4] with ENDF VII.0 [5] nuclear data libraries [6]. However these results are valid only for short irradiations on moderate to high power, as the measurements with ionization chamber indicate, that delayed γ-rays due to neutron activation contribute up to 30 % of total γ-ray flux in steady state operation [7, 8, 9], and is the only source of γ rays after reactor shutdown. Thorough γ-ray characterization in terms of computational analysis and measurements in an operational reactor and after reactor shutdown, for both prompt and delayed 306.1

2 306.2 γ-rays is planned in the near future. Initial steps in modeling the delayed γ-ray contributions from activated nuclear fuel, due to the fact that it can be a viable γ-ray source and is a main contributor of delayed γ-rays, have been taken, and methods and results described in this paper. A rigorous two-step method (R2S) using Monte Carlo particle transport code MCNP 6.1 [4] for particle transport part of the problem, and FISPACT-II [10] for neutron activation and delayed γ-ray generation part with custom in-house developed Python scripts, combining the two codes. 2 RIGOROUS TWO-STEP METHOD (R2S) An R2S method [11] in reactor physics is a method of coupling particle transport capabilities with neutron activation analysis and delayed γ-ray generation. The basic idea is to divide the problem geometry into small voxels and calculate neutron spectra and total neutron flux values in each of them, or even subdivide the voxel to conform with geometry (cell-under-voxel approach) [12] by the means of a Monte Carlo neutron transport. Voxel neutron flux and spectrum, along with voxel material composition and mass is used in neutron activation analysis calculation, generating time dependent isotope inventory and delayed γ ray spectra and intensities respectively. The latter results are used in a Monte Carlo photon transport as time dependent γ sources of delayed γ in a respective voxel for delayed photon field, spectrum, kerma and dose particle transport calculations. 3 COMPUTATIONAL SETUP SIDE TOP 106 cm Tangential irr. port 655 cm 370 cm Concrete Door plug 264 cm Heavy concrete door Heavy concrete Graphite 198 cm 91.4 cm Core ~490 cm Lead Alu. housing Boral 274 cm 366 cm Polyethylene Radial piercing irr. port Thermal column door Plug Boral Radial beam irr. port Thermal irr.port Reactor tank Steel shield Lead Polyethylene 274 cm Thermalizing column 244 cm Figure 1: Schematic of the JSI TRIGA Mark II reactor, side (left) and top (right) view. JSI TRIGA is a pool type reactor, with maximum steady state power of 250 kw (Fig. 1). The fuel elements are arranged in an annular configuration into rings A-F (Fig. 2). A detailed JSI TRIGA reactor MCNP model, based on the criticality benchmark model [13], which has been thoroughly validated by multiple experiments [14, 15, 16], was used for total neutron flux and spectra distribution inside the fissionable material part of the fuel element (U-ZrH fuel meat) of several standard JSI TRIGA fuel elements (12 wt% uranium, enrichment 20 %). Neutron activation for fresh fuel meat (U-ZrH) only was calculated, as it is the main contributor of delayed γ-rays. Kerma in sample was calculated using voxel volume averaged energy deposition estimators, and the dose was calculated using ICRP-21 [17] flux to dose conversion factors for photons. Fuel isotopic

3 306.3 compositions, as well as major isotopes, contributing to the dose on the sample were also calculated. Each of the core ring fuel element representatives fuel meat was divided in 10 axial 5 radial F14 F15 F16 F17 F18 F13 E12 E13 E14 F19 F12 E15 D9 D10 F20 TriC D11 E16 F11 C6 C7 D12 F21 E9 E17 F10 C8 D7 D13 F22 E8 C5 B4 C9 E18 F9 D6 B3 B5 D14 F23 E7 C4 A1 C10 E19 F8 D5 B2 B6 D15 F24 E6 C3 B1 C11 E20 F7 D4 D16 C2 C12 F25 E5 C1 E21 D3 F6 D17 F26 E4 D2 D18 D1 E22 F5 E3 E23 F27 F4 E2 E1 E24 F28 F3 F29 F2 F1 F30 20% LEU fuel element Empty/irradiation in-core position Representative fuel element Neutron source Compensating control rod Shim control rod Regulating control rod Pulse control rod 72.5 cm Fuel element Fuel meat segmentation Upper pin along element length Upper grate Graphite reflector Fuel meat Zr rod Figure 2: JSI TRIGA core with typical fuel element arrangement and denoted representative fuel elements from each core ring. Figure 3: JSI TRIGA fuel element and fuel meat division for R2S calculation cm Molybdenum disc 3.76 cm Lower pin 3.81 cm Lower grate Fuel meat segment division over radius and angle 5 azimuthal voxels, and a 100-group neutron spectra and total flux values for steady state operation were calculated in each of them. An irradiation plan in terms of reactor power and duration for the whole core, along with voxel calculated neutron spectra, fluxes and material compositions was used to construct neutron activation code FISPACT-II input files for each voxel respectively. After the calculations for all the voxels were performed, a Python script was used, to merge the output calculations data accordingly and construct MCNP γ sources for each stage of neutron irradiation or shutdown. A flow-chart of the procedure is depicted in Fig. 4. Two irradiation time regimes were simulated: short-term irradiation, where reactor operated for a few hours, and long-term irradiation, where a typical 40 h/w operation at full reactor power was simulated until a burn-up of 1 MWd or 10 MWd was achieved. As the gamma irradiation in the JSI TRIGA reactor is performed mostly to study radiation hardness of Si based semiconductors, the main motivation of this work is to use the R2S method to calculate kerma and dose on Si. For this purpose a cylindrical pipe silicon sample was modeled around the activated fuel element, with kerma and dose distributions calculated in it (Fig. 5).

4 306.4 START Neutron transport input file Neutron transport ź Tot. neutron flux Tot. Preparation of ź Neutron group flux in flux inputs files for predetermined group activation eval. structure Group neutron flux Activities, isotopic compositions, γ spectra isotopic contributions to activity and dose, uncertainties Merging time dependent results in order KEY: Text files Neutron flux and spectrum in voxels Construction of bulk γ-ray source from activation data Monte Carlo calc. Voxel data division Neutron spectrum energy group conversion, tot. flux Batch procedure for serial input calculations γ-ray transport Doserate sample value and error Input file for neutron energy group conversion Batch procedure for multiple input evaluation Time dependent activity, isotopic composition and γ spectrum. Doserate range evaluaiton Voxel data merging, activities, uncertainties, isotopic contributions to dose Use of script END FISPACT-II calc. Time dependent activities, dose-rates, uncertainties, isotopic contributions to dose Figure 4: Flow chart of the R2S calculation. Neutron flux and spectra preparation Activation time coursr Doserate calculations, merging voxel into bulk source Corect temporal sequence, result preparation Silicon sample TRIGA fuel 40 Figure 5: Irradiation of the silicon sample with irradiated fuel element. 4 RESULTS In this section, an overview on the results for both short-term and long-term irradiations using the above mentioned R2S approach on will be presented. Activities of fuel elements, dose and kerma rates on the silicon sample are given in this section. Fuel elements inside the JSI TRIGA reactor are arranged in concentric rings. A sigle element from each ring was selected for calculations (Fig. 2). 4.1 Short activation A short fuel activation of several hours is considered in this case. This corresponds to day to day operations, where reactor is operational for several hours at full reactor power. Fresh nuclear fuel prior to irradiation is considered. Kerma-rate and dose-rate results are given in terms of kerma and dose-rate range on the sample, due to their distribution in the silicon sample, with maximal values being in the most inner part at center hight of the sample, and lowest values at the sample outer edges. All of the results have a computational uncertainty below 1 %. A [Bq] B2 1h B2 8h 10 7 B2 2h B2 10h B2 4h B2 20h 10 6 (a) B2 fuel element activity. A [Bq] E5 1h E5 8h 10 7 E5 2h E5 10h E5 4h E5 20h 10 6 (b) E5 fuel element activity.

5 306.5 Dose-rate H*10 [Sv/h] 10 3 B2 1h B2 2h B2 4h B2 8h B2 10h B2 20h Dose-rate H*10 [Sv/h] E5 1h E5 2h E5 4h E5 8h E5 10h E5 20h (c) B2 fuel element dose-rate to sample. (d) E5 fuel element dose-rate to sample. Kerma-rate [Gy/h] B2 1h B2 2h B2 4h B2 8h B2 10h B2 20h Kerma-rate [Gy/h] E5 1h E5 2h E5 4h E5 8h E5 10h E5 20h (e) B2 fuel element kerma-rate to sample. (f) E5 fuel element kerma-rate to sample. Figure 6: Two representative fuel elements (B2 and E5), with different activation times. X-axis denotes time after the irradiation. P(E) B2 1 h B2 20 h 0.00 E [MeV] (a) B2 fuel element γ-ray spectra after 1 h and 20 h of operation at full reactor power. Rel. contribution to contact dose [%] I 138 Cs 93 Sr 142 La 91 Rb 89 Rb 144 La 90 Rb 135 I 92 Sr 101 Mo 89 Kr 140 Cs 95 Y (b) Isotope contributions to the contact dose in B2 element, after 20 h of operation at full reactor power. Figure 7 Spectra of γ rays for each voxel element are calculated during the neutron activation calculation procedure, to enable us to produce sources of delayed γ-rays respectively. An integral spectrum can also be calculated (Fig. 7a). In addition, isotope contribution to contact dose is also calculated, which gives insight on radiation hazards due to neutron activation of components in a nuclear facility (Fig. 7b).

6 Long activation For long term activations, a 40 h/w operation at full power is considered, until a burnup of 1 MWd or 10 MWd is achieved, starting with fresh fuel. Fuel element activity, kerma-rate and dose-rate to the silicon sample after reactor shutdown are calculated. γ-ray spectra emitted by activated voxels have again been summed in terms of fuel element γ-ray spectra (Fig. 9a) as a whole, and isotopic contribution to contact dose have been calculated (Fig. 9b). A[Bg] B2, 1 MWd 10 9 C3, 1 MWd 10 8 D4, 1 MWd E5, 1 MWd A[Bg] B2, 10 MWd 0 C3, 10 MWd 10 9 D4, 10 MWd E5, 10 MWd Dose-rate H*10 [Sv/h] (a) 1 MWd fuel elements activity. B2, 1 MWd C3, 1 MWd D4, 1 MWd E5, 1 MWd 1 Dose-rate H*10 [Sv/h] (b) 10 MWd fuel elements activity. B2, 10 MWd C3, 10 MWd D4, 10 MWd E5, 10 MWd 1 Kerma rate[gy/h] (c) 1 MWd fuel elements dose-rate to sample. B2, 1 MWd C3, 1 MWd D4, 1 MWd E5, 1 MWd 1 (d) 10 MWd fuel elements dose-rate to sample. Kerma rate[gy/h] B2, 10 MWd C3, 10 MWd D4, 10 MWd E5, 10 MWd 1 (e) 1 MWd fuel elements kerma-rate to sample. (f) 10 MWd fuel elements kerma-rate to sample. Figure 8: Results for representative fuel elements with burn-up of 1 MWd and 10 MWd.

7 306.7 P(E) MWd 10 MWd 0 E [MeV] (a) Emitted γ-ray spectra from B2 element after 1 MWd and 10 MWd burn-up. Rel. contribution to contact dose [%] La140 I135 Np239 I133 Nb97 I133 Nb97m Zr95 Nb95 Sr91 Ce143 Xe135 (b) Isotopic contribution to contact dose on B2 element after 10 MWd burn-up. t col Figure Results comment γ-decaying isotopes in the nuclear fuel are mainly produced in nuclear fission, and their distribution is proportional to fission rates and therefore to power distribution. The differences in the γ-activities of fuel elements irradiated for same periods of time are therefore attributed to power distribution which has been calculated per fuel element [15]. The ranges of kerma-rates and dose-rates can be attributed to the axial neutron flux distributions, which has also been evaluated both computationally and by measurements [2]. For longer irradiation times, or until a desired burn-up is achieved, we can clearly observe the effect of increased neutron flux distribution in inner elements. The γ-decaying isotope production rate is again proportional to neutron flux, and with shorter irradiation time, fewer of the so produced isotopes have decayed up to the irradiation end. This affects the isotopes, which reach saturation in several hours, which can are visible in the Fig. 8 with differences between fuel elements for same burn-up. With the increased burn-up, long-lived isotopes are produced linearly with the increased irradiation time and burn-up, therefore the 10 increase in activities and sample dose-rates and kerma-rates at 10 MWd burn-up, compared to the 1 MWd. 5 SUMMARY The R2S method presented here aims to provide preliminary analysis on contribution of delayed γ-rays to photon kerma and dose-rates in the reactor irradiation facilities. Initial measurement of delayed γ-ray dose in the vicinity of several fuel elements show agreement with the calculations. The codes and scripts developed for this problem specific calculation procedure give valuable insight on developing a generalized method for R2S calculations, with possible geometry under voxel approach implemented [12], while current method performs division, conforming to model geometry. The model segmentation, neutron flux and spectrum calculations, as well as neutron activation and delayed γ-ray source generation would be performed automatically. The code would be used for delayed γ-field characterization of the JSI TRIGA reactor, which would increase utilization for γ-ray radiation hardness studies. R2S methods are applied almost exclusively to fusion devices, where little to no utilization on fission reactors. In the following months, the method will be applied to the whole JSI TRIGA reactor core, along with the development of the above mentioned generalized R2S code and with γ field measurements in the JSI TRIGA reactor during reactor operation and after reactor shutdown. The

8 306.8 measurements will serve to establish measurement procedures of γ measurements in mixed γ-neutron fields and to validate the computational results. The current and a generalized method will be used in the future to evaluate component activation inside nuclear power plants, as well as for accurate classification of nuclear waste after nuclear facility s decommissioning, therefore increasing long-term storage facility utilization. REFERENCES [1] Luka Snoj, Gašper Žerovnik, and Andrej Trkov. Computational analysis of irradiation facilities at the JSI TRIGA reactor. Applied Radiation and Isotopes, 70(3): , [2] Gašper Žerovnik, Tanja Kaiba, Vladimir Radulovic, Anže Jazbec, Sebastjan Rupnik, Loc Barbot, Damien Fourmentel, and Luka Snoj. Validation of the neutron and gamma fields in the JSI TRIGA reactor using in-core fission and ionization chambers. Applied Radiation and Isotopes, 96:27 35, [3] Luka Snoj and Borut Smodiš. 45 years of TRIGA Mark II in Slovenia. In Proceedings of the International Conference Nuclear Energy for New Europe, September [4] T Goorley, M James, T Booth, F Brown, J Bull, LJ Cox, J Durkee, J Elson, M Fensin, RA Forster, et al. Initial MCNP6 release overview. Nuclear Technology, 180(3): , [5] M.B. Chadwick, P. Obloinsk, M. Herman, et al. ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology. Nuclear Data Sheets, 107(12): , Evaluated Nuclear Data File ENDF/B-VII.0. [6] Klemen Ambrožič, Luka Snoj, and Gašper Žerovnik. Computational analysis of the dose rates at JSI TRIGA reactor irradiation facilities. To be published this year, [7] Vladimir Radulović, Damien Fourmentel, Loïc Barbot, Jean-Franois Villard, Tanja Kaiba, Žerovnik Gašper, and Luka Snoj. Measurements of miniature ionization chamber currents in the JSI TRIGA mark II reactor demonstrate the importance of the delayed contribution to the photon field in nuclear reactors. Nuclear Instruments and Methods in Physics Research Section A:Accelerators, Spectrometers, Detectors and Associated Equipment, 804: , [8] Damien Fourmentel, Vladimir Radulović, Loïc Barbot, Jean Francis Villard, Gašper Žerovnik, Luka Snoj, Mikolaj Tarchalski, Krysztof Pytel, and Fadhel Malouch. Delayed gamma measurements in different nuclear research reactors bringing out the importance of the delayed contribution in gamma flux calculations. In Advancements in Nuclear Instrumentation Measurement Methods and their Applications, ANIMMA, Lisbon, April Instituto de Plasmas e Fusão Nuclear. [9] Vladimir Radulovi, Damien Fourmentel, Loc Barbot, Jean-Franois Villard, Tanja Kaiba, erovnik Gaper, and Luka Snoj. Measurements of miniature ionization chamber currents in the {JSI} {TRIGA} mark {II} reactor demonstrate the importance of the delayed contribution to the photon field in nuclear reactors. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 804: , 2015.

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