VALIDATION OF THE MONTEBURNS CODE FOR CRITICALITY CALCULATIONS OF TRIGA REACTORS. Hugo Moura Dalle* and Robert Jeraj**. *Centro de ...
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VALIDATION OF THE MONTEBURNS CODE FOR CRITICALITY CALCULATIONS OF TRIGA REACTORS
Hugo Moura Dalle* and Robert Jeraj**
*Centro de Desenvolvimento da Tecnologia Nuclear – CDTN/CNEN Caixa Postal 941 – Cidade Universitária – Pampulha 30123-970 – Belo Horizonte/MG - Brazil **Jozef Stefan Institute Jamova 39, 1000 Ljubljana Ljubljana/Slovenia
ABSTRACT Use of Monte Carlo methods in burnup calculations of nuclear fuel has become practical due to increased speed of computers. Monteburns is an automated computational tool that links the Monte Carlo code MCNP with the burnup and decay code ORIGEN2.1. This code system was used to simulate a criticality benchmark experiment with burned fuel on a TRIGA Mark II research reactor. Two core configurations were simulated and keff values calculated. The comparison between the calculated and experimental values shows good agreement, which indicates that the MCNP/Monteburns/ORIGEN2.1 system gives reliable results for neutronic simulations of TRIGA reactors. Keywords: TRIGA research reactor, Monte Carlo simulation, burned fuel. I. INTRODUCTION Monte Carlo simulation of the TRIGA Mark II Criticality Benchmark Experiment with Burned Fuel [1] with Monteburns [2] code system is presented in this paper. The criticality benchmark experiment was performed in 1998 at the Jozef Stefan Institute (JSI) TRIGA reactor for two core loading patterns. These two core configurations were named core 133.1 and core 134.1 and had an average fuel element burnup equal to 1.22 MWd (2.8% U235 burned) and 1.15 MWd (2.6% U235 burned) for cores 133.1 and 134.1, respectively. The fresh fuel experimental criticality data for both core configurations were also available from benchmark experiments with fresh fuel performed in 1991 [3, 7]. The criticality experiments for JSI TRIGA reactor are well known and can be considered a valuable benchmark experiment for validation of neutronic codes and calculation methods for TRIGA reactors. The two cores were simulated with Monteburns and values of effective neutron multiplication factor - keff - were obtained.
Figure 1. View of the JSI Core 133.1.
A very detailed description of the experimental facility, including geometric and material data for the JSI TRIGA, measuring methods and uncertainties can be found in references [1, 3, 7]. For this reason only the schematic diagrams of core 133.1 and core 134.1 are showed in Fig. 1 and Fig. 2.
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Figure 2. View of the JSI Core 134.1.
II. MONTEBURNS DESCRIPTION Due to the increased speed of computers, use of Monte Carlo methods for burnup calculations is becoming more practical. Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP [4] with the radioactive burnup and decay code ORIGEN2.1 [5]. The main function of Monteburns is to transfer one group crosssection and neutron flux values from MCNP to ORIGEN2.1 and then transfer the resulting material compositions after irradiation and/or decay from ORIGEN2.1 back to MCNP. The number of these iterations is defined by the user. Monteburns has two parts. A Perl script file called Monteburns.pl and the Fortran77 program Monteb. Besides these two files, in order to run properly, the system must have a working version of MCNP4, ORIGEN2.1 and Perl. The Perl software is usually distributed with Unix systems and it can be obtained freely for most operational systems at www.perl.com or www.activestate.com. MCNP and ORIGEN2.1. are distributed by Radiation Safety Information Computational Center (RSICC) from Oak Ridge and the codes are neither free of charge nor free of a previous approval by the United States government. The RSICC release of Monteburns does not support Linux systems. However, recompiling the Fortran source file with an appropriate Fortran compiler for Linux makes Monteburns possible to run on this system. Monteburns can also work in Linux clusters with small modifications in the Perl script, since multiprocessing version of MCNP is available [6]. As MCNP strongly determines the running time for Monteburns to use such a system of parallel processing is recommended. Figure 3 is a diagram showing the interaction between the codes, the data flow and input files. Monteburns is completely automated meaning that with just one simple line command a script is followed. In this script (Monteburns.pl) MCNP, ORIGEN and Monteb are ran as many times as necessary, which is determined by the user.
The data are also automatically exchanged among the codes and they are used in order to prepare automatically the new input files for the different burnup calculation steps. The MCNP code calculates one-group cross-sections and fluxes that are used in burnup and/or decay calculations by ORIGEN. Then, ORIGEN performs the burnup and/or decay calculations getting the isotopic compositions of materials that will feed MCNP in another burnup cycle. The information passage and data flow between the codes is performed by Monteb as well as the input files preparation. There are 2 input files that must be provided by the user and a third one that is optional. One of them is a MCNP input file (named INPUT in the diagram below) and it is the MCNP model that describes the system. The second one is the Monteburns input file itself and it is a very simple input (INPUT.INP). The optional file (INPUT.FEED) is the feed input file and it allows to add or remove different amounts of materials to the system during each burn step, to change the power released by the system and to shuffle the materials from one zone to another zone of the system. The MBXS.INP file is not a specific user supplying file, but a general file containing a list of MCNP cross-sections identifiers.
MCNP Initial material compositions, cross sections and fluxes
INPUT INPUT.INP INPUT.FEED
ORIGEN2
MONTEBURNS
New material compositions
MBXS.INP
MCNP Cross sections and fluxes
Figure 3. Diagram of the Monteburns/MCNP/ORIGEN2.1 Interaction.
The system is able to supply quantitative information about the keff in function of the irradiation time (burnup), mass of practically all fission products and transuranic elements produced and/or removed from the system, radionuclides activity, decay heat, ingestion and inhalation radiotoxicity. All the standard information supplied by ORIGEN and MCNP like criticality, neutron spectrum, activity, etc, can be obtained with small modifications of the system.
III. CRITICALITY CALCULATIONS The calculations were performed with the kcode option in MCNP. Water temperature was assumed 23
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TABLE 1. Measured and Calculated (Monteburns) keff for JSI TRIGA Cores 133 and 134 with Fresh and Burned Fuel Core
Burnup (MWd)
keff experim.
keff ± σ calculated
keff(calc.) – keff(exp.)
133
0
1.00277± 0.00015
1.00337 ± 0.00036
0.00060
133.1
1.22
0.9842 ± 0.00160
0.98836 ± 0.00035
0.00416
134
0
1.0202 ± 0.00200
1.02297 ± 0.00035
0.00277
134.1
1.15
1.00460± 0.00015
1.00761 ± 0.00035
0.00301
1.005
1
0.995 Experiment Simulation
keff
Celsius. No corrections of the temperature were applied to the MCNP cross-section data. 6000 neutrons per cycle and 1000 cycles, skipping the 100 first cycles were used. The material cross-sections from ENDF/B-VI continuos-energy library were used always when possible. The ENDF/B-V library was used for those fission products not available in the previous library. An average burnup for all fuel elements was assumed, 1.22 MWd - 2.8% U235 burned - for core 133.1 and 1.15 MWd - 2.6% U235 burned – for core 134.1. No shuffling of the fuel elements in the cores was considered. Studies of the influence of homogeneous core burnup calculations (consider that all fuel elements in the core has the same average burnup) and heterogeneous core burnup calculations (consider that each fuel element has its own burnup) showed that this effect is minimal and can be neglected [8]. The fresh cores were irradiated at constant and continuous power of 250 kW until the requested burnup was achieved. The whole burnup of the cores was divided in two burnup steps – the results are more accurate if long periods of irradiation are broken up in smaller steps. After that, a 10 day decay time was used to discard the fission products poisoning. The library Thermal.lib was used in the ORIGEN2.1 calculations of the isotopic compositions. Sensitivity studies of the effect of some nuclides over the results of keff calculation [8] have showed that is enough to consider only the influence of the fission products Xe135, Sm149, Sm151, Pu239, Nd143, U236, Pm147, Rh103, Xe131, Cs133, Tc99, Nd145 and Pu240 in criticality calculations. In the present simulation only these fission products were considered important regarding the effect on keff.
0.99
0.985
0.98 0
IV. RESULTS The running time for the complete Monteburns calculation, in which MCNP (almost all of the execution time is spent in MCNP) was ran six times, was approximately 66 hours on a Pentium III, 1000 MHz, for each core simulated. The results of Monteburns simulation for fresh and burned cores of the JSI TRIGA are showed in Table 1. This results are also presented in graphical view on Figs. 4 and 5. The differences between experimental and calculated keff results are bellow 500 pcm, that is the error of the MCNP model (due to inaccuracies on materials composition specification and geometrical simplifications). One can note from Fig. 5 that the difference between the measured and calculated keff is approximately constant, around 300 pcm, for the core 134 in the considered burnup range. On the other hand, Fig. 4 shows that for the core 133 this discrepancy is very low for the fresh core and increases for the burned core. This can reflect the larger error during measurements with the reactor operating far from the criticality.
0.61
1.22
Burnup (MWd)
Figure 4. Measured and Calculated (Monteburns) keff for Fresh and Burned Fuel – Core 133.
Besides keff other parameters were calculated by Monteburns. These parameters like the Xenon poisoning and the activity, mass, heatload and radiotoxicities for several fission products and actinides produced during the irradiation of the fuel are related with the isotopic compositions of the burned fuel. However, there is not any experimental data available for comparison. For this reason these results will not be presented in this paper which is oriented towards validation against experimental data. Nevertheless, since the keff is being correctly calculated by the Monteburns system one can assume that the other parameters are being correctly simulated since that a wrong calculation of the isotopic compositions could affect keff.
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REFERENCES 1.023
[1] Persic A., Ravnik M., Zagar T., TRIGA Mark II Criticality Benchmark Experiment with Burned Fuel, Nuclear Technology, vol. 132, p 325-337, 2000
1.019
1.015 keff
Experiment Simulation 1.011
[2] Potton D.I. and Trellue H. R., User’s Manual, Version 2.0 for Monteburns Version 1.0, LA-UR-99-4999 (September 1999). [3] Mele I., Ravnik M., Trkov A., TRIGA Mark II Benchmark Experiment, Part I: Steady-State Operation, Nuclear Technology, vol. 105, p 37-52, 1994
1.007
1.003 0
0.575
1.15
Burnup (MWd)
Figure 5. Measured and Calculated (Monteburns) keff for Fresh and Burned Fuel – Core 134.
V. CONCLUSIONS A criticality benchmark experiment with burned fuel performed at the TRIGA Mark II research reactor at Jozef Stefan Institute, Ljubljana, Slovenia was used to validate the Monteburns code for criticality calculations of TRIGA research reactors. The calculated values of keff agree well with the measured values. The results for both conditions, fresh and burned fuel for the two cores simulated are within the estimated uncertainty of the MCNP model. The discrepancy between the measured and calculated keff is approximately constant for the core 134 in the considered burnup range. For the core 133 this difference is very low for the fresh fuel and increases for the burned fuel. This probably reflects a larger error during measurements with the reactor operating far from the criticality. The good results obtained for the keff of burned cores indicate that the isotopic compositions for the burned fuel is being reliably calculated. Since other parameters calculated by Monteburns, like the Xenon poisoning and the activity, mass, heatload and radiotoxicities for several fission products and actinides produced during the irradiation of the fuel are related with the isotopic compositions of the burned fuel this can be an indirect indication that those parameters are being properly calculated as well. All in all, it can be concluded that MCNP/Monteburns/ORIGEN2.1 system gives reliable results for neutronic simulation of TRIGA fuel in both conditions, fresh and burned.
[4] Briesmeister J. F., MCNP – A General Monte Carlo N-Particle Transport Code, Version 4B, LA-12625-M, 1997. [5] Croff A. G., A User’s Manual for the ORIGEN2 Computer Code, ORNL/TM-7175, 1980. [6] Dalle H. M. and Muniz F. J., Computers Cluster for Parallel Processing with the MCNP4B Code: Linux + PVM + MCNP4B, XIII ENFIR, INAC 2002, Rio de Janeiro, Brazil, 2002. [7] Jeraj R. and Ravnik M., TRIGA Mark II reactor: U(20) – zirconium hydride fuel rods in water with graphite reflector, International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95) 03/III, Paris, Nuclear Energy Agency, 1999. [8] Jeraj R., Zagar T., Ravnik M., Monte Carlo Simulation of the TRIGA Mark II Benchmark Experiment with Burned Fuel, Nuclear Technology, 137(3), 2002.
