GB5 A LINKING CODE BETWEEN MCNP5 AND ORIGEN2.1 DEN/UFMG VERSION: INCLUDING RADIOACTIVITY HAZARD.

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1 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2011) Rio de Janeiro, RJ, Brazil, May 8-12, 2011, on CD-ROM, Latin American Section (LAS) / American Nuclear Society (ANS) ISBN GB5 A LINKING CODE BETWEEN MCNP5 AND ORIGEN2.1 DEN/UFMG VERSION: INCLUDING RADIOACTIVITY HAZARD. Renan Cunha, Claubia Pereira, Daniel Campolina, Maria Auxiliadora Fortini Veloso Departamento de Engenharia Nuclear Escola de Engenharia Universidade Federal de Minas Gerais Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPq Av. Antônio Carlos, Campus Pampulha Belo Horizonte, MG, Brazil renancunha@ufmg.br; claubia@nuclear.ufmg.br; campolina@ufmg.br; dora@nuclear.ufmg.br André Cavatoni Departamento de Ciência da Computação Instituto de Ciências Exatas Universidade Federal de Minas Gerais Av. Antônio Carlos, Campus Pampulha Belo Horizonte, MG, Brazil cavatoni@gmail.com ABSTRACT The GB code has been developed at the Departamento de Engenharia Nuclear/UFMG to couple the Monte Carlo transport code MCNP (Los Alamos National Laboratory) with the depletion and burnup code ORIGEN2.1(Oak Ridge National Laboratory). A previous version of GB code described the behavior during the burn up to only 25 isotopes. In the new version of GB presented on this paper, named GB5, the goal was to increase the isotope amount considered and to include radiotoxicity and radioactivity in the results. While simulating 75 time steps at 800kw of a Heat Pipe Power System model, we have found that GB5 generates very similar results compared to The small difference encountered with the neutron flux parameter is explained by the form that recoverable energy per fission is calculated in GB5. Key Words: ORIGEN, MCNP,, Heat Pipe Power System. 1. INTRODUCTION The use of computational codes made possible to simulate and evaluate complex scenarios in the area of nuclear engineering. Nowadays, we can find a lot of codes, each one developed for a different purpose. Two of the most important codes used in the area are MCNP [1] and ORIGEN [6]. While MCNP performs particle transport calculations using Monte Carlo techniques, ORIGEN determines isotopic composition balance of a nuclear core submitted to a neutron flux over time. The isolated use of each one of these codes allows us to create and evaluate appropriate scenarios for nuclear core observation, aiming to discover physical parameters such as reaction rates or k eff, or to observe fuel evolution over time, depending on the chosen code. To achieve more complex scenarios and more realistic results, it is often necessary to use linking codes.

2 Renan Cunha et al. Linking codes [3,4,7,11] are normally developed in order to combine two or more simulation codes within the same model, in order to improve reality fidelity. The first version of GB code [2] was developed so we could benefit from both MCNP and ORIGEN codes in a complementary fashion, reaching more precise results. More specifically, the first version of GB was developed to interface the MCNP4C and ORIGEN2.1 nuclear codes, in order to add the native feature of isotopic composition evolution (burnup) calculation of ORIGEN to the MCNP code. This paper presents GB5, an improvement of the previous GB code. On GB5, among other modifications, the MCNP4C code interface was substituted by MCNP5, allowing an increase in the number of considered isotopes and consequently improving the fidelity of the transport model. Furthermore, it includes radiotoxicity and radioactivity information in the results. To compare the results the HEAT PIPE POWER SYSTEM case was simulated using GB5 and Section 2 revises the calculations necessary to find some physical parameters needed as input for both MCNP or ORIGEN. Section 3 describes the computational workflow between MCNP and ORIGEN enabled by the GB code. It highlights the main differences between the preliminary and the new version of GB too. Section 4 presents the system used to evaluated the results and finally the section 5 compares results of GB (previous version) and GB5 with the results obtained with [7], used as reference model. This paper aims to find the standard deviation between the results of neutron parameters within the order of the standard deviations found in results. Section 6 presents the conclusions of this work. 2. METHODOLOGY MCNP tally results have information of the average number of neutrons per fission ( ), cell neutronic fluence ( ) and recoverable energy per fission (Q), normalized by fission neutron. Moreover, results inform the effective multiplication factor (k eff ), reaction rates of the isotopes in a cell, reactions (n,γ), (n,2n), (n,3n), (n, fission) for actinides and reactions (n,γ), (n,2n), (n,α), (n, p) for fission products. Energy integrated reaction rates are calculated by MCNP for all isotopes that have microscopic cross sections with transport information. This calculation is made according to equation 1: rate n ( E) Rm ( E) de (1) where n(e)= energy dependent neutron fluence in the cell (1/cm 2 ), R m (E)= response function of material m to a given reaction in function of energy (continuous microscopic cross section). To obtain one group microscopic cross sections from reaction rates, GB performs the division of the reaction rates by the neutron fluence in the cell the considered isotope is located (equation 2): 2/14

3 GB5 A linking code between MCNP5 and ORIGEN2.1 DEN/UFMG version: Including Radioactivity Hazard i rate n (2) where σ i= one group microscopic cross section, to material i in cell (barn). Neutron flux must be normalized to thermal power of the system. The recovery energy per fission is obtained from MCNP through equation 3: Q ave z i 1 z ( Q... n ) i 1 i n n i f i f i i (.. n ) (3) where Q ave= recoverable average energy per fission in (MeV/fission) Q i = recoverable energy per fission to isotope i (MeV/fission), σ i f = microscopic cross section to isotope i (barn), n i = atomic density of isotope i in cell (atom/barn.cm) z= number of isotopes in the cell. Solution of equation 3, calculated by MCNP, informs the averaged recoverable energy per prompt fission neutron (Q i ). To estimate the gamma capture heating, GB5 uses the normalization factor 1.111, as proposed in version [7]. From Q ave and v, it is possible to convert the fluence per system neutron fission ( ) in the power system normalized neutron flux (equation 4): Power.10 W / MW. v 6 n 13 (4) (1, J / MeV ). keff. Qave where Φ n= neutron flux in the cell (neutron/cm 2.s), Power= thermal power system supplied in the GB card POWER (MW), v = total fission neutrons produced in cell, k eff = effective multiplication factor of the system, =fluence in cell. 3/14

4 Renan Cunha et al. 3. COMPUTATIONAL WORKFLOW Figure 1 shows a general view of how GB5 works. First, GB5 generates a new input to MCNP based on the decisions of the user. After that, MCNP is run, and data like recoverable energy per prompt fission neutron, reaction rates and k eff, are extracted from the output by GB5. GB5 uses these data to calculate flux and cross sections, necessary to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. Then, GB5 eliminates all isotopes that don t match MCNP conditions, that is, isotopes that are not available in the continuous energy MCNP cross sections library (ENDF/VI). To update the input file of MCNP with the new isotopic composition it is necessary to compensate the lost of some isotopes (generally fission products). This compensation is made by the normalization to 100% for the isotopes (generated in ORIGEN) that are recognized by MCNP. On the previous version of GB it was used 24 isotopes to generate the new MCNP input. That was due to an intrinsic limitation of MCNP4C. In GB5, the amount of fission products considered increased from 24 up to 97. That was possible because there was no such limitation with MCNP5. adopted, in average, 12 actinides and 57 non-actinides. To develop the new version of GB, beyond using MCPN5, it was necessary to rewrite some procedures so that inputs and outputs were correctly read and written by MCNP5 and ORIGEN. Figure 1. Computational Workflow 4/14

5 GB5 A linking code between MCNP5 and ORIGEN2.1 DEN/UFMG version: Including Radioactivity Hazard 4. HEAT PIPE POWER SYSTEM In this work, we used the Heat Pipe Power System (HPS) [12, 13] as a study case to compare the results obtained using GB5 and This system is a low-mass and small space fission power system. The HPS draws on 40 years of United States and international experience to enable a system that can be developed in less than five years at a cost of less than 100 millions dollars. Total HPS mass is lighter than 600 kg at 5 kwe and less than 2000 kg at 50 kwe, assuming that thermoelectric power conversion is used [13]. Moreover, it has no pumped coolant loops and does not require a pressure vessel with hermetic seals. The HPS was designed to remain subcritical during all credible launch accidents without using in-core shutdown rods. This passive subcriticality results from the high radial reflector worth and the use of resonance absorbers in the core. The system also passively removes decay heat and is virtually nonradioactive at launch (no plutonium in the system). As another advantage, HPS has no single-point failures and is capable of delivering rated power, even if several modules and/or heatpipes fail. The HPS was designed such that the fuel can be stored and transported separately from the system until shortly before launch. This capability reduces storage and transportation costs significantly. A schematic of the HPS is shown in Figure 2. The fuel pins are bonded structurally and thermally to a central heatpipe, which transfers heat to an ex-core power conversion system. The heatpipe also provides structural support for the fuel pins. Modules are independent during normal operation. If a heatpipe fails, some thermal bonding between modules is desirable to reduce peak temperatures. In our experiments, we have used uranium nitride (UN) as the fuel of the reactor although uranium dioxide (U02) could be used too. Figure 2. Schematic of HPS Showing Fuel Pins and Radial Reflector [12]. The HPS primary heatpipes operate at a temperature of 1300 K and transfer heat to secondary heatpipes operating at 1275 K. Heat generated in the fuel is transferred to the module heatpipe, which transfers heat to the secondary heatpipes, with the unction located on the surface of the 5/14

6 Renan Cunha et al. shield. In the thermoelectric option, heat from the secondary heatpipes is transferred to thermoelectric converters that are bonded to the heatpipe surface. The 1275 K converter hot-side temperature is adequate for many types of power conversion, although higher or lower temperatures could be used. If needed, the HPS heatpipes are capable of long-term operation at temperatures 1500 K. In the tested scenarios, neutron spectrum is fast, as expected by a non moderated reactor. 5. GENERAL RESULTS To verify the approximation to the case considered for benchmarking (simulation using version 2.0.6), we compared GB and GB5 results. Simulation consisted in the analysis of the following parameters: effective multiplication factor, neutronic flux and isotopic composition of the fuel Uranium Nitride (UN). While the execution of GB, and its correspondent case, used ENDF.50C libraries, the execution of GB5, and its correspondent simulation, used ENDF.69C. Figure 3 presents a third order polynomial regression of the effective multiplication factor encountered by GB and GB5 in comparison with. It is seen in GB results that the separation of the curves increases during burnup. In GB5 the separation keeps inside the standard deviation associated to MCNP results. As the considered system was the same for both simulations, except for the isotopic composition evolution, the difference encountered between GB and GB5 is caused by the number of fission products considered in the transport model (MCNP). Regarding the isotopic composition GB adopted 11 actinides and 13 non-actinides, while adopted, in average, 12 actinides and 57 non-actinides. GB5 used 19 actinides and 78 non-actinides for the transport model. 6/14

7 k effective k effective GB5 A linking code between MCNP5 and ORIGEN2.1 DEN/UFMG version: Including Radioactivity Hazard 1,035 1,03 1,025 GB POL.REG.(GB) POL.REG.() 1,02 1,015 1,01 1, ,035 1,03 GB5 POL.REG.() POL.REG.(GB5) 1,025 1,02 1,015 1,01 1,005 1 Figure 3. Effective multiplication factor over time The standard deviations between the results are presented in Table 1. In GB, the standard deviation is higher than standard deviations associated with k effective on MCNP results (~100 pcm). This divergence wasn t encountered in GB5 results. Burnup (days) Table 1. Standard deviation in k eff results GB x MCNP (Intrinsic) GB5 x MCNP (Intrinsic) % 0.114% 0.022% 0.114% % 0.118% 0.079% 0.109% % 0.113% 0.063% 0.082% % 0.124% 0.066% 0.093% % 0.109% 0.082% 0.113% Curves of neutron flux calculated by GB and GB5 compared with its correspondent simutation are shown in Figure 4. 7/14

8 flux (n/cm^2.s) flux (n/cm^2.s) Renan Cunha et al. 9,78E+13 9,73E+13 9,68E+13 9,63E+13 GB POL.REG.(GB) POL.REG.() 9,58E+13 9,53E+13 9,48E+13 9,43E+13 9,38E+13 9,33E+13 9,66E+13 9,61E+13 9,56E+13 9,51E+13 GB5 POL.REG.() POL.REG (GB5) 9,46E+13 9,41E+13 9,36E+13 9,31E+13 9,26E Figure 4. Neutron flux over time Neutronic flux is proportional to the inverse of the averaged recoverable energy per fission Q avg (equation 4). The disagreement encountered between GB5 and is a consequence of the different values considered for Q avg in the codes. As it can be seen from Table 2, if the programs used the same methodology to calculate Q avg, the difference would be inside acceptable values considered for this study (~100pcm). Table 2. Neutron flux variation caused by difference in Q avg Burnup (days) Averaged recoverable energy per fission Q (MeV/fission) Neutron Flux Difference GB5 GB GB5 x GB x due difference in Q % 0.415% 0.242% % 0.654% 0.244% % 0.872% 0.246% % 1.004% 0.250% % 0.980% 0.253% 8/14

9 GB5 A linking code between MCNP5 and ORIGEN2.1 DEN/UFMG version: Including Radioactivity Hazard The following figures present the variation on composition of some isotopes. Not all isotopes present in the model were shown here, but only the main actinides. In Figure 5 is shown the U 235 consumption along burnup days for GB and GB5. We can see a very similar curve when compared with results. The difference between GB, GB5 and results are plotted in the graphs in terms of percentages. Figure 4. U E+04 composition versus time 7.45E % -1.20% -1.00% -0.80% -0.60% -0.40% -0.20% 0.00% %DIFF. GB U235 (grams) 7.40E E E E E E % -1.20% -1.00% -0.80% -0.60% -0.40% -0.20% 0.00% 7.50E E+04 %DIFF. GB5 U235 (grams) 7.40E E E E E E Figure 5. U 235 composition over time. As can be seen from Figure 6, the amount of U 238 drop more in the GB and GB5 than in model. U238 (grams) -8.00% -7.00% -6.00% -5.00% -4.00% -3.00% -2.00% -1.00% 0.00% 1.88E %DIFF. GB 9/14

10 Renan Cunha et al. U238 (grams) -8.00% -7.00% -6.00% -5.00% -4.00% -3.00% -2.00% -1.00% 0.00% 1.88E %DIFF. GB5 Figure 6. U 238 composition over time. Variation on the composition of Pu 239 is presented in Figure 7. The difference found in both cases remained below 0.27% for GB and below 0.10% for GB5 over all the burnup. Pu239 (grams) -0.30% -0.20% -0.10% 0.00% 0.10% 0.20% 0.30% 8.00E E E E E E E+00 Figure 6. Pu 239 composition versus time. %DIFF. GB 1.00E E+00 Figure 7. Pu 240 composition versus time % -0.02% 0.00% 0.02% 0.04% 0.06% 0.08% 0.10% 8.00E E E+00 %DIFF. GB5 Pu239 (grams) 5.00E E E E E E Figure 7. U 238 composition over time. 10/14

11 GB5 A linking code between MCNP5 and ORIGEN2.1 DEN/UFMG version: Including Radioactivity Hazard Figures 8 to 13 present information about radiotoxicity and radioactivity for actinides and daughters, and fission products during the cycles obtained by GB5. Figure 8. Radioactive Ingestion Hazard (Actinides and Daughters) Figure 9. Radioactive Inhalation Hazard (Actinides and Daughters) Figure 10. Radioactivity (Actinides and Daughters) 11/14

12 Renan Cunha et al. Figure 11. Radioactive Ingestion Hazard (Fission Products) Figure 12. Radioactive Inhalation Hazard (Fission Products) Figure 13. Radioactivity (Fission Products) 12/14

13 GB5 A linking code between MCNP5 and ORIGEN2.1 DEN/UFMG version: Including Radioactivity Hazard 7. CONCLUSIONS This work presented a new version of GB linking code. In the new version, the number of isotopes considered in the transport model was increased and the expected results were reached. Furthermore, information about radiotoxicity and radioactivity during the cycles is presented. To validate the new version, we compared results of GB and GB5 with results from version The difference found in the k effective curve remained inside the standard deviation intrinsic to MCNP results. The neutron flux values found with the new version were more similar to the ones found with. If the same calculation for the averaged recoverable energy per fission (Q avg ) were considered, results would fall inside the expected deviation values. However, comparisons of the isotopic evolution showed that the accordance in the transport model reached by the improvements in the version 5 of GB was not followed by an accordance in the isotopic composition evolution. The reason is that GB5 uses ORIGEN2.1 and uses Cinder90 libraries to generate reaction rates. The next step of this work is to verify the new code considering different neutron spectrum. ACKNOWLEDGMENTS The authors would like to thank FAPEMIG, CNPq and CAPES, which have directly or indirectly contributed to the accomplishment of this study. REFERENCES [1] BREISMEISTER, J. F. MCNPTM - A General Monte Carlo n-particle Transport Code, Los Alamos National Laboratory, LA M, (1993). [2] CAMPOLINA D.; PEREIRA C. et al; GB a preliminary linking code between MCNP4C and ORIGEN2.1 DEN/UFMG version. International Nuclear Atlantic Conference INAC (Octo. 2009). [3] CETNAR J. et al. Simulation of Nuclide Transmutation with Monte Carlo Continuous Energy Burnup Code (MCB1C). Proc. Accelerator Application and ADTTA, Nuclear Application in the New Millenium, Reno, USA, (2001). [4] CETNAR J. et al. MCB: a continuous energy Monte Carlo burnup simulation code, in actinide and fission product partitioning and transmutation. EUR EN, OECD/NEA 523, (1999). [5] CROFF A. G. A User s Manual for ORIGEN2 Computer Code. Oak Ridge National Laboratory, report ORNL/TM-7175, (Jul. 1980). [6] CROFF A. G. ORIGEN2: A Versatile Computer Code for Calculating the Nuclide Compositions and Characteristics of Nuclear Materials, Nuclear Technology, vol.62, p.335, (Sept.1983). [7] DENISE B. PELOWITZ. User s Manual, Version Los Alamos National Laboratory report, LA-CP , (Apr. 2008). [8] DEHART M.D.; PETRIE L.M. A Radioisotope depletion method using Monte Carlo Transport with variance reduction and error propagation. Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN /14

14 Renan Cunha et al. [9] POSTON D. L.; TRELLUE H.R.. User s Manual, Version 2.0 for MONTEBURNS Version 1.0. LA-UR , (Sept. 1999). [10] TALAMO A.; JI W.; CETNAR J.; GUDOWSKI W. Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn. Annals of Nuclear Energy 33, p , (Aug. 2006). [11] TRELLUE H.R.; POSTON D.I.; Monteburns: A Monte Carlo Burnup Code for Accelerator Applications, Los Alamos National Laboratory, LA-UR , (Dec. 2000). [12] HOUTS G. M.; POSTON I. D.; RANKEN A. W. Heatpipe Space Power and Propulsion Systems. Los Alamos National Laboratory, MS K551. Staif-96 Meeting, Albuquerque, NM, Jan.7-11,1996. [13] CAPELL B.; BETER M. Engineering design aspects of the heatpipe power system. AIP Conf. Space technology and applications international forum ; DOI: / , Volume 420, p , Jan. 15, /14