VERIFICATION OF MONTE CARLO CALCULATIONS OF THE NEUTRON FLUX IN THE CAROUSEL CHANNELS OF THE TRIGA MARK II REACTOR, LJUBLJANA

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1 International Conference Nuclear Energy for New Europe 2002 Kranjska Gora, Slovenia, September 9-12, VERIFATION OF MONTE CARLO CALCULATIONS OF THE NEUTRON FLUX IN THE CAROUSEL CHANNELS OF THE TRIGA MARK II REACTOR, LJUBLJANA Radojko Jaćimović Jožef Stefan Institute Department of Environmental Sciences Jamova 39, SI-1000 Ljubljana, Slovenia Marko Maučec 1 Kernfysisch Versneller Instituut Nuclear Geophysics Division Zernikelaan 25, 9474 AA Groningen, The Netherlands Maucec@kvi.nl Andrej Trkov 1 International Atomic Energy Agency Department of Nuclear Sciences and Applications Wagramerstrasse 5, P.O. Box 100, A-1400 Vienna, Austria A.Trkov@iaea.org ABSTRACT In this work experimental verification of Monte Carlo neutron flux calculations in the carousel facility (CF) of the 250 kw TRIGA Mark II reactor at the Jožef Stefan Institute is presented. Simulations were carried out using the Monte Carlo radiation-transport code, MCNP4B. The objective of the work was to model and verify experimentally the azimuthal variation of neutron flux in the CF for core No. 176, set up in April Au activities of Al-Au(0.1%) disks irradiated in 11 s of the CF covering 180 around the perimeter of the core were measured. The comparison between MCNP calculation and measurement shows relatively good agreement and demonstrates the overall accuracy with which the detailed spectral characteristics can be predicted by calculations. 1 INTRODUCTION MCNP4B [1] is the Los Alamos National Laboratory developed general-purpose Monte Carlo radiation-transport code. It facilitates independent or coupled neutron, photon and electron transport calculations. The code treats an arbitrary three-dimensional configuration of materials and geometries and provides a versatile description of the source, a rich collection of variance reduction techniques [2], a flexible tally structure and an extensive collection of cross-section data. 1 On leave: Reactor Physics Division, Jožef Stefan Institute, Jamova 39, 1000 Ljubljana, Slovenia

2 The Monte Carlo model of the TRIGA Mark II reactor in Ljubljana was originally developed to reproduce the benchmark experiment, carried out initially with the fresh [3,4] and recently with fuel subject to burn-up [5]. The model considers all the important details concerning the reactor core, graphite reflector, thermal and thermal column, irradiation s and biological shielding. The TRIGA reactor core model has a cylindrical, though non-periodic configuration with 91 locations. Fuel and control elements are arranged in six concentric rings. The core is surrounded by an aluminium-lined graphite reflector containing an embedded carousel irradiation facility. The objective of the present work was to model and verify experimentally the azimuthal variation of neutron flux in the CF for core No. 176, set up in April For the set of Monte Carlo calculations, the MCNP model was extended to implement the changes recently introduced in the general outline of the core. The triangle formed by s E10, E11 and D8 (called the triangular and denoted T in Figure 1), used as an experimental irradiation facility (empty at the time of measurements), and a new fast pneumatic transfer irradiation (aluminium tube; FPTS in Figure 1) at location F22 were added. Furthermore, core No. 176 differs from the previously analysed core No. 169 (March 2000) in one additional new fuel element inserted at location E18. Activities of Al-Au(0.1%) disks irradiated in 11 s of the CF covering 180 around the perimeter were measured. The comparison between calculation and measurement shows relatively good agreement and demonstrates that the detailed spectral characteristics can be predicted by Monte Carlo calculations with acceptable accuracy. 2 METHODS 2.1 Monte Carlo computational aspects In principle, the MCNP calculations were carried out in two consecutive stages. Initially, a point source was defined in each fuel element and the standard MCNP4B criticality calculation option KCODE was used for the simulation of the fission source distribution over the reactor core. The first batch of calculations was performed in order to achieve a stable and converging fission source distribution. For this purpose 2000 neutron histories were simulated per cycle, with 1000 non-contributing and 4000 actively contributing cycles [1,6]. In the final set of criticality calculations a total number of active cycles was used. The effective multiplication factors k eff of the modelled core No. 176 was found to be ± In subsequent calculations, the converging fission source distribution was used to set-up the surface source, distributed on the aluminium lining of the reactor core. The surface source was then used in forward transport calculations through the graphite reflector. This approach, although more advanced and sophisticated than straightforward criticality calculations, was required to obtain statistically reliable results in a reasonable amount of time for the relatively small (cylindrical volume of ~ 5 cm in diameter and 3 cm height) sample positions in the irradiation s. For the estimation of neutron flux in the CF (see Figure 1), the standard MCNP volumeaveraged track length estimator was implemented. The flux was calculated in a 640-group energy structure, ranging from 10-4 ev to 20 MeV [7]. Parameter f (thermal to epithermal ratio) determined by simulation is given in Table 1 for the 11 carousel s. 2.2 Experimental details Al-Au(0.1%) disks were inserted in 11 s covering 180º around the perimeter (from -30 over -40 to -10). The Al-Au(0.1%) disks of 6 mm in diameter and 0.2 mm

3 thickness were pressed from IRMM-530 wire of 1 mm diameter. The disks were fixed about 5 mm from the bottom of the container to minimise the influence of the vertical neutron flux gradient. Samples were irradiated for two hours in a thermal neutron flux of about cm -2 s -1. After irradiation all samples were measured on the same day, at the same distance and on the same HPGe detector (Ortec, USA, with 40 % relative efficiency), connected to a Canberra S100 multi analyser. For peak area evaluation, the HYPERMET-PC [8,9] program was used. The counting statistics of the net peak area for the measured radionuclide 198 Au (411.8 kev) was kept at about 0.3% to allow for detection of small variations of the neutron flux distribution. Logarithmic Pulse Safety 40 F14 F15 F16 F17 F18 F13 E12 E13 F19 E14 E15 F12 F20 F11 T D9 D10 D11 E16 C D12 F21 E9 C7 E17 F10 D7 C6 C8 D13 F22 E8 C5 B4 C9 E18 FPTS F9 D6 B3 B5 D14 F23 F8 E7 C4 CC C10 E19 NS D5 P B2 B6 S D15 F24 E6 C3 B1 C11 E20 PT F7 D4 C2 C12 D16 F25 E5 C1 E21 F6 D3 D17 F26 E4 D2 D1 D18 E22 F5 E3 R E23 F27 F4 E2 E1 E24 F28 F3 F29 F2 F1 F Graphite Carousel Linear Start Fuel elements 12 % U-235 Control rods FPTS PT Fast pneumatic transfer system Pneumatic transport tube NS Neutron source CC Central Irradiation s T Triangular Figure 1: Ground plan of the TRIGA Mark II reactor with irradiation s

4 RESULTS AND DISCUSSION The sensitivity of the gold foils is not uniform in energy, but highly peaked around the 5 ev gold resonance. The reaction rate energy distribution (i.e. the product of the spectrum and the gold capture cross section) is shown in Figure 2. The peaked nature of the distribution makes comparison between measurements and calculations more difficult, because the most significant contribution to the overall reaction rate comes from a narrow energy band. With the Monte Carlo technique it is rather difficult to achieve sufficiently low statistical uncertainty to make comparison meaningful. Figure 2: Reaction rate distribution as a function of energy for gold irradiated in the carousel of the TRIGA reactor The Monte Carlo results are compared to experimental ones in Figure 3. The values are arbitrarily normalised to those from 40. The measured values vary rather smoothly around the perimeter. Similarly, the calculated total flux also varies smoothly and shows the same general trends as the measurements. Differences arise due to the non-uniform sensitivity of gold to neutrons of different energies. When the calculated spectrum is convoluted with the gold cross section to calculate the reaction rate (labelled React.Rate in Figure 3), significant scattering in the calculated values is observed. This is attributed to the statistical uncertainty in the calculated flux in the energy bins which contribute most significantly to the reaction rate.

5 norm sp. activ. Au-198 1,05 1,00 0,95 0,90 0,85 0,80 0,75 Measured React.Rate MCNP Au-197 Activation in the CF, , Carusel s Figure 3: Comparison of experimental results with MCNP results for thermal neutron flux for 11 s of the CF of the TRIGA Mark II reactor Table 1: Comparison of experimental results for reaction rate with those of Monte Carlo calculations in 11 carousel s at the TRIGA reactor. Parameter f was obtained only by simulation Channel Experimental Total flux (MCNP) Reaction rate (MCNP) Spectral ratio (MCNP) Spectral ratio (Measured) ± 0.81* * Experimentally determined by Cd ratio multi-monitor method. The following set of monitors was used: Al-Au(0.1%), Al-In(0.099%), Al-Lu(0.1%), Fe(99.9%), Zn(99.99%) and Zr(99.8%). The thermal to epithermal spectrum ratio was also measured in 40 and is comparable to the calculated value. In this case the calculation is much less sensitive to the statistical uncertainty and hence easier to compare with the measurement. In fact, excellent agreement is observed as evident from Table 1.

6 CONCLUSIONS Due to problems in reducing (or smoothing out) the statistical uncertainty in the calculated spectra over the energy region of the gold resonance, the calculated data are considered preliminary. Further work is in progress to: 1. reduce the statistical uncertainty by increasing the number of histories in the Monte Carlo calculation, 2. refine the technique to smooth the statistical fluctuations by fitting an analytic function to the calculated spectrum. The calculated spectral ratio is much less sensitive to the statistical uncertainty and agrees much better with the measurement. ACKNOWLEDGMENTS The authors would like to thank the Ministry of Education, Science and Sport of the Republic of Slovenia for financial support our Project group P REFERENCES [1] J. F. Briesmeister, In: MCNP, (Ed.), A General Monte Carlo N-Particle Transport Code, Ver. 4B, LA-12625, LANL, NM, [2] E. Booth, A sample problem for Variance Reduction in MCNP, LA MS, LANL, NM, [3] R. Jeraj, B. Glumac, M. Maučec, "Monte Carlo simulation of the TRIGA Mark II benchmark experiment", Nucl. Technol., 120, 3, 1997, pp [4] R. Jeraj, M. Ravnik, "TRIGA Mark II Benchmark Critical Experiment", IEU-COMP- THERM-003, International Handbook of Evaluated Criticality Safety Benchmark Experiments, In: NEA/NSC/DOC(95)03/III, (Ed.), September [5] R. Jeraj, T. Žagar, M. Ravnik, "Monte Carlo simulation of the TRIGA Mark II benchmark experiment with burned fuel", Nucl. Technol., 137, 3, 2002, pp [6] M. Maučec, "Monte Carlo calculations of neutron and photon transport in complex geometries" (in Slovene), Dissertation, Faculty of mathematics and physics, University of Ljubljana, January [7] D. E. Cullen, The ENFB Pre-processing Codes Pre-Pro-96. IAEA-NDS-39, [8] B. Fazekas, G. Molnár, T. Belgya, L. Dabolczi, A. Simonits, "Introducing HYPERMET- PC for automatic analysis of complex gamma-ray spectra", J. Radioanal. Nucl. Chem., 215, 2, 1997, pp [9] HYPERMET-PC V5.0, User s Manual, Institute of Isotopes, Budapest, Hungary, 1997.