D. López, C. Töre
SEA Shielding Engineering and Analysis S.L, Spain
Abstract. Burnup credit has taken a great importance during the last years. Hence, a special effort for improving the reliability of current burnup codes and cross-section libraries has been necessary in order to simulate as accurately as possible the behavior of the irradiated fuel in nuclear power plants and/or storage and transport facilities.
The aim of the work performed at SEA is to compare the new code MONTEBURNS, developed in Los Alamos Laboratory, with ORIGEN-S for the WWER fuel rod isotopics calculation in an infinite array. MONTEBURNS links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2.1. The results obtained show that MONTEBURNS may become in the future a very important tool for isotopic studies, though its accuracy depends on the quality of the cross-sections used.
OBJECTIVES
Due to the importance, which the burnup credit has taken during the last years, it is necessary to evaluate and validate new codes, which simulate accurately the behaviour of the irradiated fuel.
The aim of the work performed at SEA is to compare the new code MONTEBURNS with ORIGEN-S for the WWER fuel rod isotopics calculation in an infinite array, considering:
• Uniform burnup
• Three different burnup values: 10, 30 and 40 MWd/kgU
• Cooling time: 1 year (for 10 and 40 MWd/kgU) and 1 and 5 years (for 30 MWd/kgU) Finally, the influence of the isotopic evolution on the criticality of the storage system has also been considered.
MONTEBURNS CODE
MONTEBURNS code has been developed in Los Alamos Laboratory by David Poston and Holly Trellue. Its principal feature is the capability of linking the Monte Carlo transport code MCNP4B with the radioactive decay and burnup code ORIGEN2.1. MONTEBURNS transfers one-group cross-section and flux values from MCNP to ORIGEN, and then transfers the resulting material compositions (after irradiation and/or decay) from ORIGEN back to MCNP in a repeated cyclic fashion. This operation is undertaken only with those isotopes, which are considered as important for the neutronic problem by the user. For the rest of the isotopes the default ORIGEN cross-section libraries are used by MONTEBURNS.
INITIAL DATA (from the CB3 benchmark)
The WWER design is hexagonal. The next two figures correspond to the horizontal and vertical cross-sections, respectively, taken from the MCNP plot function.
FIG. 1.
• Fuel specifications
Fuel enrichment: 3.6% wt. 235U Fuel cell pitch: 1.22 cm
Fuel radius: 0.38 cm
Pellet density (effective): 10.193 g/cm3 Active fuel length: 244 cm
• Cladding specifications Inner radius: 0.38 cm
Outer radius: 0.455 cm
Material: 1% wt. Nb, 98.97% wt. Zr, 0.03% wt. Hf Density: 6.0881 g/cm3
• Moderator Water at 0.72 g/cm3 Soluble boron: 500 ppm
• Operating history data
Specific power: 4.143 MW/assembly (average value)
Since there are 126 rods/assembly, the specific power for one fuel rod is: 0.0329 MW Number of cycles: 1, 3, 4 for the 10, 30, 40 MWd/kgU burnup cases respectively Cycle duration
- Uptime: 300 days
- Downtime (between cycles): 65 days Cooling time
- 1 year for 10 and 40 MWd/kgU burnup cases - 1 and 5 years for 30 MWd/kgU case
HYPOTHESES
MONTEBURNS combines the codes MCNP and ORIGEN. The geometry is built by MCNP and ORIGEN undertakes the burnup simulation. Since MCNP is a three-dimensional code, the WWER fuel rod is modeled in 3-D, but using reflective boundaries to avoid any leakage in any direction in order to approximate ORIGEN-S conditions for an infinite array.
The density of the moderator (light water) is supposed to be constant and the fuel composition is also uniform, radially and axially.
The ENDF/B-VI library is used for most of the isotopes, though the ENDF/B-V is employed for some fission products. Finally, the Watt fission spectrum is considered.
METHODOLOGY
Three input files are usually necessary to define a simulation with MONTEBURNS:
− The MCNP input file defines the geometry of the system and provides the detailed cross-section libraries to ORIGEN for those isotopes considered as important by the user
− By means of the MONTEBURNS input file the power of the system, the isotopes whose results are required and the default ORIGEN library (for the not important isotopes) are determined
− The MONTEBURNS feed file defines the number of times which the ORIGEN libraries are updated by MCNP
The concentration of the next isotopes (actinides + fission products) is required:
235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 95Mo, 99Tc, 101Ru,
103Rh,109Ag, 133Cs,143Nd, 145Nd,147Sm,149Sm,150Sm,151Sm,152Sm,153Eu,155Gd RESULTS
The next figures represent the ratio MONTEBURNS/ORIGEN-S for the isotopics calculation for the different values of burnup and cooling time. The ORIGEN-S values are taken from the CB3 benchmark specification.
FIG. 2. 10 MWd/kgU and 1 year cooling time.
0 0,5 1 1,5 2 2,5
U-235 U-236 U-238 Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-243 Mo-95 Tc-99 Ru-101 Rh-103 Ag-109 Cs-133 Nd-143 Nd-145 Sm-147 Sm-149 Sm-150 Sm-151 Sm-152 Eu-153 Gd-155
MONTEBURNS/ORIGEN-S
FIG. 3. 30 MWd/kgU and 1 year cooling time.
FIG. 4. 30 MWd/kgU and 5 years cooling time.
FIG. 5. 40 MWd/kgU and 1 year cooling time.
0 0,5 1 1,5 2 2,5
U-235 U-236 U-238 Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-243 Mo-95 Tc-99 Ru-101 Rh-103 Ag-109 Cs-133 Nd-143 Nd-145 Sm-147 Sm-149 Sm-150 Sm-151 Sm-152 Eu-153 Gd-155
MONTEBURNS/ORIGEN-S
0 0,5 1 1,5 2 2,5
U-235 U-236 U-238 Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-243 Mo-95 Tc-99 Ru-101 Rh-103 Ag-109 Cs-133 Nd-143 Nd-145 Sm-147 Sm-149 Sm-150 Sm-151 Sm-152 Eu-153 Gd-155
MONTEBURNS/ORIGEN-S
0 0,5 1 1,5 2 2,5
U-235 U-236 U-238 Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-243 Mo-95 Tc-99 Ru-101 Rh-103 Ag-109 Cs-133 Nd-143 Nd-145 Sm-147 Sm-149 Sm-150 Sm-151 Sm-152 Eu-153 Gd-155
MONTEBURNS/ORIGEN-S
0.85 0.9 0.95 1 1.05 1.1 1.15 1.2
U-235 U-236 U-238 Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-243
MONTEBURNS/ORIGEN-S
FIG. 6. Actinides.
0.95 1 1.05 1.1
Mo-95 Tc-99 Ru-101 Rh-103 Ag-109 Cs-133 Nd-143 Nd-145
MONTEBURNS/ORIGEN-S
FIG. 7. Fission Products.
It is clear, looking at the figures above, that there are important discrepancies for some isotopes: 243Am,155Gd, 149Sm,151Sm and 152Sm. In the next two figures (one for the actinides and a second one for the fission products except for the samarium family and 155Gd) the ratio MONTEBURNS/ORIGEN-S for 10 MWd/kgU and 1 year cooling time (the rest of the cases are similar to this one) is represented in a more detailed scale, in order to appreciate with more clarity the differences between the results of the two codes.
Influence on the multiplication factor
The different isotopic compositions calculated with MONTEBURNS and ORIGEN-S are introduced into the MCNP input which models the WWER fuel rod in cold conditions, in order to calculate the keff of the system in all the burnup and cooling time cases. For both calculations the same libraries (ENDF/B-VI and ENDF/B-V) are used. Hence, the deviations observed are only produced by the differences in the concentrations. These are the results:
FIG. 8.
CONCLUSIONS
An isotopic calculation has been undertaken for the WWER fuel rod in order to compare the results provided by two different codes, ORIGEN-S and MONTEBURNS, and for evaluating the capabilities of prediction, which the latter has in burnup calculations.
In most of the isotopic calculations a relative error lower than 10% is achieved between MONTEBURNS and ORIGEN-S results, except for the next nuclides: 243Am, 155Gd, 149Sm,
151Sm and 152Sm. For the Sm family and 155 Gd there are some undefined resonance regions in the ENDF-B-V cross-section libraries, which may be the reason of the big discrepancies. In the 243Am case, the relative error is about 20%, because there are insufficient resonance shielding calculations for 242 Pu and its progeny (243Am and 244Cm).
In the criticality calculations, the differences in the isotopic compositions calculated with MONTEBURNS and ORIGEN-S produce a constant overestimation of the keff calculated with the MONTEBURNS composition with respect to the ORIGEN-S one. The average value of this difference is about 3%, which is due to the overestimation of the 235U and 239 Pu isotopics by MONTEBURNS. This discrepancy may be important in spent fuel storage or transport facility.
REFERENCES
[1] POSTON D.I. TRELLUE, H.R., User’s manual Version 2.0 for Monteburns, Version 1.0. UR-99-4999. Sept 1999.
[2] MARKOVA, L., Calculational Burnup Credit Benchmark Proposal, Sept. 1996.
[3] BRIESMEISTER, J.F., “MCNP-A General Monte Carlo N-Particle Transport Code System” LA-12625-M, Version 4-B. Los Alamos National Laboratory. March 1997.
Burnup (cooling time) 10(1) 30(1) 30(5) 40(1) k-eff (monteburns) 1.255 1.102 1.086 1.028 k-eff (origen-s) 1.216 1.068 1.050 1.006 deviation (%) 3.2 3.2 3.4 2.2
0,9 1 1,1 1,2 1,3
0 1 2 3 4 5
BURNUP (cooling time) [(MWd/kgU (years)]
keff
kmonteburns korigens