A TEMPERATURE DEPENDENT ENDF/B-VI.8 ACE LIBRARY FOR UO2, THO2, ZIRC4, SS AISI-348, H2O, B4C AND AG-IN-CD

Transcription

1 2009 International Nuclear Atlantic Conference - INAC 2009 Rio de Janeiro,RJ, Brazil, September27 to October 2, 2009 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: A TEMPERATURE DEPENDENT ENDF/B-VI.8 ACE LIBRARY FOR UO2, THO2, ZIRC4, SS AISI-348, H2O, B4C AND AG-IN-CD Hugo M. Dalle, Mario Bianchini and Paulo Cezar Gomes Centro de Desenvolvimento da Tecnologia Nuclear (CDTN / CNEN - MG) Av. Presidente Antônio Carlos, 6.627, Campus da UFMG Pampulha Belo Horizonte, MG dallehm@cdtn.br ABSTRACT Most MCNP standard neutron ACE libraries are processed at room temperature, 293,6 K. The temperature enters into the processing of the evaluation of a data file through the Doppler broadening of cross-sections. The nuclear fuel burnup usually takes place at reactor core temperatures much higher than room temperature, consequently, Monte Carlo burnup calculations should not only use the best cross-sections evaluations available but also use evaluations that are at temperatures approximating the temperatures of the application. In order to face the scarcity of temperature dependent MCNP cross-sections data to most isotopes, CDTN is developing an in-house temperature dependent neutron library for those nuclides commonly necessary in the systems simulated for the ongoing projects demanding Monte Carlo burnup. This paper describes the data processing of the ENDF/B-VI, release 8, using the NJOY99 code, towards provides this temperature dependent ACE library. Up to now fifty one elements and isotopes of the materials uranium oxide, thorium oxide, zircaloy-4, stainless steel AISI-348, light water, boron carbide and the silver-indium-cadmium alloy were processed at temperatures range from 293,6 K to 1200 K. Some benchmarks for thorium cycle described in the OECD/NEA International Handbook of Evaluated Criticality Safety Benchmark Experiments were simulated using MCNP5 and the data set of this in-house library and the results usually agree with those obtained for the.60c standard MCNP neutron library for room temperature. 1. INTRODUCTION This paper reports the processing of a temperature dependent neutron pointwise crosssections library to be used by MCNP [1] transport code. The source of the basic nuclear data was the ENDF/B-VI, release 8, [2] and NJOY [3] was the processing code. This multitemperature ACE library has being used for simulations of some on going projects at CDTN/CNEN-MG, which demand materials cross-sections processed at higher temperatures than the default 293 Kelvin commonly distributed with MCNP. The library was evaluated through the MCNP simulation of benchmark experiments of critical configurations containing some thorium-based fuel [4]. However, since the benchmark experiments were carried out at room temperature, just the simulations results using the set of cross-sections processed at 293 Kelvin can be compared with the benchmark experimental values as well as the results calculated using the MCNP default.60c neutron library. 2. THE EVALUATED NUCLEAR DATA FILES As already mentioned the source of the evaluated nuclear data files was only the ENDF/B-VI. Release 8 was the most recent version available at CDTN when the library was processed. The nuclides chosen to be processed in this first version of the CDTN multi-temperature ACE

2 library are those usually existing in materials used for control rods (B4C and Ag-In-Cd alloy), for nuclear fuels (UO2 and ThO2), for cladding (Zircaloy-4 and stainless steel) and moderator (H2O). Adding more materials is foreseen as well as the processing of the newest data from ENDF/B-VII. Table 1 presents the isotopes and temperatures in which the basic data were processed. Table 1 Processed Isotopes and temperatures Isotope Temperatures (K) H , 323.6, 373.6, 423.6, 473.6, 523.6, 573.6, C-natural 293.6, 400, 500, 600, 700, 800, 900, 1200 B , 400, 500, 600, 700 B , 400, 500, 600, 700 O , 400, 500, 600, 700, 800, 900, 1200 O , 400, 500, 600, 700, 800, 900, 1200 Al , 400, 500, 600, 700 P , 400, 500, 600, 700, 800, 900, 1200 Fe , 400, 500, 600, 700, 800, 900, 1200 Fe , 400, 500, 600, 700, 800, 900, 1200 Fe , 400, 500, 600, 700, 800, 900, 1200 Fe , 400, 500, 600, 700, 800, 900, 1200 S-natural 293.6, 400, 500, 600, 700, 800, 900, 1200 Si-natural 293.6, 400, 500, 600, 700, 800, 900, 1200 Zr-natural 293.6, 400, 500, 600, 700, 800, 900, 1200 Mo-natural 293.6, 400, 500, 600, 700, 800, 900, 1200 Cd-natural 293.6, 400, 500, 600, 700 In-natural 293.6, 400, 500, 600, 700 Cr , 400, 500, 600, 700, 800, 900, 1200 Cr , 400, 500, 600, 700, 800, 900, 1200 Cr , 400, 500, 600, 700, 800, 900, 1200 Cr , 400, 500, 600, 700, 800, 900, 1200 Mn , 400, 500, 600, 700, 800, 900, 1200 Ni , 400, 500, 600, 700, 800, 900, 1200 Ni , 400, 500, 600, 700, 800, 900, 1200 Ni , 400, 500, 600, 700, 800, 900, 1200 Ni , 400, 500, 600, 700, 800, 900, 1200 Ni , 400, 500, 600, 700, 800, 900, 1200 Co , 400, 500, 600, 700, 800, 900, 1200 Cu , 400, 500, 600, 700, 800, 900, 1200 Cu , 400, 500, 600, 700, 800, 900, 1200 Nb , 400, 500, 600, 700, 800, 900, 1200 Ag , 400, 500, 600, 700 Ag , 400, 500, 600, 700 Sn , 400, 500, 600, 700, 800, 900, 1200 Sn , 400, 500, 600, 700, 800, 900, 1200 Sn , 400, 500, 600, 700, 800, 900, 1200

3 Isotope Temperatures (K) Sn , 400, 500, 600, 700, 800, 900, 1200 Sn , 400, 500, 600, 700, 800, 900, 1200 Sn , 400, 500, 600, 700, 800, 900, 1200 Sn , 400, 500, 600, 700, 800, 900, 1200 Sn , 400, 500, 600, 700, 800, 900, 1200 Sn , 400, 500, 600, 700, 800, 900, 1200 Sn , 400, 500, 600, 700, 800, 900, 1200 Ta , 400, 500, 600, 700, 800, 900, 1200 Th , 400, 500, 600, 700, 800, 900, 1200 U , 400, 500, 600, 700, 800, 900, 1200 U , 400, 500, 600, 700, 800, 900, 1200 U , 400, 500, 600, 700, 800, 900, 1200 U , 400, 500, 600, 700, 800, 900, 1200 Pu , 400, 500, 600, 700, 800, 900, PROCESSING THE ENDF/B-VI.8 TO ACE FORMAT The ENDF/B-VI.8 was processed to the ACE format using the code NJOY99, update 112. Figure 1 shows the NJOY processing sequence. Two calculations are carried out to each processing sequence of the isotope [5]. The first, the PENDF calculation, runs the modules RECONR, BROADR, HEATR, GASPR, PURR and THERMR to write the continuous library on ENDF format. Thus, the ACER module writes the continuous data to the ACE format which can be used to the MCNP family of Monte Carlo transport codes. In addition, the errors checking and verifications procedures are also run. The main processing parameters to each NJOY module are: RECONR Reconstruction tolerance: 0.1% Resonance integral check tolerance: 0.3% Reconstruction temperature: 0 K BROADR Thinning tolerance: 0.1% Integral criterion tolerance: 0.3% Maximum energy: 2.0 MeV Temperatures: See table 1 Bootstrap=0 Restart=0 HEATR MT=444 e MT=443 PURR Number of probability bins: 20 Number of resonance ladders: 100

4 THERMR Number of angles bins: 12 Tolerance: 0.1% Maximum energy: 4.0 ev Scattering laws: S(α,β) for hydrogen bound in water and free gas model for all other isotopes. ACER Type of ACE file: 1 ZAID suffix:.01c to.08c Detailed photon calculation, no thinning. Figure 1 NJOY processing sequence

5 After the processing of the ENDF, the new ACE files were added to the MCNP5 libraries. Furthermore, the MCNP xsdir file was properly modified to include the new data. Some inhouse test cases were ran to test the compatibility between the code and the library and the simulations were completed successfully. In addition, a few criticality benchmark experiments containing thorium fuel were simulated. Next section presents the results of these benchmark calculations. 4. LIBRARY VERIFICATION Three benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments [4] were selected for simulation. They are formed by 14 experiments prepared for testing and validation of calculation codes and data for applications in the thorium fuel cycle. All benchmarks have uranium and thorium in the fuel. However, they are very different in terms of geometry, enrichment, materials and neutron energy spectra. Complete information about these benchmark experiments can be found on reference [4]. The 14 integral experiments were simulated using the.60c standard MCNP library and the CDTN library. Tables 2, 3 and 4 present the k eff results of these calculations and the experimental values, as well as the calculated to experimental ratios (C/E) and calculation to calculation ratio (CDTN/.60c). Considering the results of the Light Water Breeder Reactor benchmark, table 2, the differences in calculated k eff for the two libraries (column CDTN/.60c) are within 0.5%, while the C/E CDTN column has 7 of 8 values within 0.5% difference and C/E.60c column has 6 of 8 values in this range. Table 2 Results for U233-COMP-THERM-001 benchmark Case k eff Experimental ± ± ± ± ± ± ± ± LWBR SB CORE EXPERIMENTS k eff.60c K eff CDTN C/E.60c C/E CDTN CDTN/.60c ± ± ± ± ± ± ± ± ± ± ± ± ± ± ± ±

6 Observing the KBR benchmarks, tables 3 and 4, one can notice a larger dispersion among the C/E values going from 0.1% up to almost 5%. However, the differences in calculation to calculation ratio (CDTN/.60c) are mostly within 0.5%, which show that both libraries give similar results. In summary, these preliminary tests indicate that CDTN library and.60c standard MCNP library give similar results at room temperature. Even the well known large discrepancy of C/E values for KBR-21 experiment is reproduced. At higher temperatures there are not, so far, any public available benchmarks for thorium fuel to test the CDTN library. However, some simulation results, only for uranium fuel, at a higher temperature can be found in [6]. Table 3 Results for IEU-COMP-INTER-001 benchmark K-INFINITY MEASUREMENTS WITH ENRICHED URANIUM MIXED WITH THORIUM AND POLYETHYLENE (KBR-18, KBR-19, KBR-20, AND KBR-21 ASSEMBLIES) Case k eff experimental k eff.60c k eff CDTN C/E.60c C/E CDTN CDTN/.60c KBR ± ± ± KBR ± ± ± KBR ± ± ± KBR ± ± ± Table 4 Results for HEU-MET-FAST-068 benchmark HIGHLY ENRICHED URANIUM, THORIUM, AND POLYETHYLENE ASSEMBLIES (KBR-22 AND KBR-23) Case k eff Experimental k eff.60c k eff CDTN C/E.60c C/E CDTN CDTN/.60c KBR ± ± ± KBR ± ± ±

7 5. CONCLUSIONS NJOY99 code was used to process ENDF/B-VI, release 8, in order to produce a multitemperature continuous energy MCNP library. Up to now only a limited set of materials, B4C, Ag-In-Cd alloy, UO2, ThO2, Zircaloy-4, AISI-348 stainless steel and H2O were processed in a temperature range going from to 1200 Kelvin. Adding more materials is in progress as well as the processing of the newest data from ENDF/B-VII. Preliminary tests, using well known thorium integral benchmarks, indicate that this multi-temperature continuous energy CDTN library and the.60c standard MCNP library give similar results at room temperature. Verification in other temperatures is still pending due to the lack of available public benchmarks for thorium fuel at high temperatures. REFERENCES 1. X-5 Monte Carlo Team, MCNP A General Monte Carlo N-Particle Transport Code, Version 5, Los Alamos National Laboratory, V. Mclane, (editor), ENDF-6 Formats Manuals, IAEA-NDS-76 Rev. 6, April R. E. Macfarlane, D. W. MUIR, NJOY-99.0: Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B data, Los Alamos National Laboratory, PSR-480, NEA Nuclear Energy Agency, International Handbook of Evaluated Criticality Safety Benchmark Experiments, September 2006 edition. 5. D. L. Aldama, A. Trkov, ADS-Lib/V1.0 - A test library for Accelerator Driven Systems, Vienna, Austria, IAEA-INDC(NDS)-0474, August H. M. Dalle, Monte Carlo Burnup Simulation of the Takahama-3 Benchmark Experiment, Proceedings of the 2009 International Nuclear Atlantic Conference, INAC2009, Brazil (2009).