- Email: [email protected]
Spallation Source Modelling for an ADS Using the MCNPX and GEANT4 Packages for Sensitivity Analysis of Reactivity
Available online at www.sciencedirect.com
Nuclear Data Sheets 118 (2014) 510–512 www.elsevier.com/locate/nds
Spallation Source Modelling for an ADS Using the MCNPX and GEANT4 Packages for Sensitivity Analysis of Reactivity M.Q. Antolin,1, ∗ F. Marinho,2 D.A.P. Palma,3 and A.S. Martinez4 1
Centro Universitario Estadual da Zona Oeste - UEZO , Rio de Janeiro, 23070-200, Brazil Universidade Federal do Rio de Janeiro, Av. Aluizio da Silva Gomes, 50, Maca´e, 27930-560, Brazil 3 Comiss˜ ao Nacional de Energia Nuclear, Rua Gal Severiano, 90, Rio de Janeiro, 22290-901, Brazil 4 Instituto Alberto Luiz Coimbra de P´ os-Gradua¸c˜ ao e Pesquisa COPPE/UFRJ, Rio de Janeiro, 21941-972, Brazil 2
A simulation for the time evolution of the MYRRHA conceptual reactor was developed. The SERPENT code was used to simulate the nuclear fuel depletion and the spallation source which drives the system was simulated using both MCNPX and GEANT4 packages. The obtained results for the neutron energy spectrum from the spallation are coherent with each other and were used as input for the SERPENT code which simulated a constant power operation regime. The obtained results show that the criticality of the system is not sensitive to the spallation models employed and only relative small deviations with respect to the inverse kinetic model coming from the point kinetic equations proposed by Gandini were observed. I.
INTRODUCTION
A significant fraction of electric energy produced by nuclear reactors is from light water based reactors. The use of low-enriched fissile material in such reactors makes their commissioning and operation more attractive. However, these reactors use an open fuel cycle and generate a great deal of radioactive waste which includes transuranic elements (Am, Cm, Np, etc) with half-lives up to 107 years. Because of this, source-driven sub-critical reactors have been thought around the world as an efficient way for the transmuting of large volumes of waste, which helps reducing the volumes in geological storage [1]. In order to evaluate the feasibility and efficiency of an ADS reactor before its construction it is important to develop independent methods to ensure consistency on the results of such estimates. For that, Monte Carlo (MC) simulations can provide important information such as indicating advantages on the materials and geometries to be adopted in the commissioning phase. It can also allow one to validate new methods for calculation of important parameters such as the reactivity for both types of reactors (hybrid and thermal). In this work a set of specific MC codes was used to develop a complete simulation from the interactions of protons on the spallation source to the fuel depletion in the MYRRHA reactor. A brief description of the reactor is given in section II. Each simulation step is described in section III and results are presented in section IV. ∗
Corresponding author: [email protected]
http://dx.doi.org/10.1016/j.nds.2014.04.120 0090-3752/2014 Published by Elsevier B.V.
II.
REACTOR DESCRIPTION
The MYRRHA hybrid reactor project for high technology research purposes started in 1988 by the Belgian Nuclear Research Center (SCK·CEN). The main goal of MYRRHA is to demonstrate the ADS concept at reasonable power level, technological feasibility for transmuting of minor actinides and of long half-life fission products. MYRRHA is also meant for the development of lead alloys technologies for generation IV reactors [2]. The reactor is expected to start its operation in 2023 and is currently under licensing phase being one of the hybrid reactors with the highest rates of citation in the literature as it will serve the international research community. This reactor will be capable of operating in critical and sub-critical mode generating power in the 50-100 MWt range. The accelerator will inject protons at an energy of 600 MeV in the reactor core. III.
SIMULATION DESCRIPTION
The simulation developed in this work consisted of two parts: simulation of the spallation source; and the fuel depletion in the reactor. This section describes each of these parts. In order to simulate the interaction of the protons with the target source two software packages, MCNPX [3] and GEANT4 [4], were used to allow comparison between different interaction models and implementation codes. The target geometry consisted of a cylinder with 20 cm height, 7.2 cm diameter and its material being a 45%
Spallation Source Modelling . . .
M.Q. Antolin et al.
NUCLEAR DATA SHEETS
lead/55% bismuth alloy. The protons reached the target at a 600 MeV energy with same direction as the cylinder height. In order to provide a detailed picture of the inelastic scattering of protons with the target material an intranuclear cascade model was used followed by a set of deexcitation codes (precompound, break-up, etc) to deal with the excited nucleus at the end of the cascade. Three main sets of implementations were used in the simulation of the spallation source which differ mainly due to intranuclear cascade models. In the Bertini Cascade Model an incident projectile interacts with nucleons inside a nucleus in the material producing secondaries which may collide with other nucleons. After each interaction excitons are introduced within the nucleus and the final excited state of the nucleus is given by a sum of these states. The final state then decays via pre-equilibrium, evaporation, fragmentation and fission methods. An important feature of this model is that the nucleus is taken as a continuum medium [5]. Unlike the Bertini implementation the Binary Cascade Model describes the nucleus by displacing the nucleons in space according to nuclear density. On the other hand, primary and secondary particles interact with single nucleon as well. Particles propagation in nuclear field is obtained via a numerical method. After specific cascading conditions are no longer satisfied precompound and de-excitation treatment follow. A recent implementation dedicated to spallation source studies in GEANT4 is the so called INCL/ABLA model which is the combination of a C++ version of the intranuclear cascade Li`ege code with the evaporation ABLA code [6]. The Li`ege approach is such that the number of free parameters needed are comparatively reduced therefore its predictive power is a preserved feature. The SERPENT code was used to simulate the operation of the MYRRHA reactor during 3 hours using fresh fuel and burned fuel [7]. The power level considered in these simulations was a constant value of 80 MWt . This code is one of the few Monte Carlo tools which allows the simulation of a nuclear reactor with external source and multiplicative medium all together. It is a code for simulation of nuclear reactors having pre-defined elements being capable of generating constant multigroups, allowing analysis of the fuel cycle via burning calculation and uses the same libraries as MCNPX. The core of the MYRRHA reactor uses a hexagonal arrangement with 68 fuel elements, being the central position dedicated to the target for the spallation reaction. The geometry used on the core of the reactor is illustrated in Fig. 1, with its main geometric parameters listed in Table I. The composition of the materials used in the simulation is listed in Tables II, IV and III.
FIG. 1. Core geometry of the sub-critical reactor.
TABLE I. General parameters of the reactor. Parameter Value (cm) Fuel element pitch 14.685 Fuel rod pitch 0.9144 Fuel rod diameter 0.8048 Coating width 0.0523 Fuel active height 84.4108
TABLE II. Fuel rod coating (ρ = 7.5357 g/cm3 ). Element 54
Fe Fe 57 Fe 58 Fe 58 Ni 60 Ni 61 Ni 62 Ni 56
Concentration Element Concentration (atoms/barn cm) (atoms/barn cm) 64 4.08E+01 Ni 3.99E-02 50 6.41E+02 Cr 4.51E+00 52 1.48E+01 Cr 8.70E+01 53 1.97E+00 Cr 9.87E+00 54 2.93E+00 Cr 2.46E+00 55 1.13E+00 Mn 4.60E+00 40 4.91E-02 Zr 4.91E+00 1.57E-01
TABLE III. Control bar (ρ = 1, 5643 g/cm3 ). Element 12
C B 11 B 10
IV.
Concentration (atoms/barn cm) 4.07E+01 1.64E+02 6.62E+02
RESULTS
The products from the simulation of the spallation source served as input information to the SERPENT code to simulate the burnup in the reactor. Table V lists the average number of neutrons produced per colliding proton obtained for each of the interaction models used. Note the estimated values are similar even when comparing the results from both MCNPX and GEANT4 for the Bertini model. Fig. 2 shows the energy spectrum of the produced neutrons for each model. The MCNPX spectrum exhibits a sharper peak around 2 MeV while the 511
Spallation Source Modelling . . .
by Gandini [9]
TABLE IV. Reflector (ρ = 5, 3908 g/cm3 ). Element 23
Na 54 Fe 56 Fe 57 Fe 58 Fe 58 Ni 60 Ni 61 Ni
M.Q. Antolin et al.
NUCLEAR DATA SHEETS
Concentration Element Concentration (atoms/barn cm) (atoms/barn cm) 62 7.15E+01 Ni 1.06E-01 64 2.77E+01 Ni 2.71E-02 50 4.35E+02 Cr 3.06E+00 52 1.01E+01 Cr 5.91E+01 53 1.34E+00 Cr 6.70E+00 54 1.99E+00 Cr 1.67E+00 55 7.67E-01 Mn 3.12E+00 40 3.33E-02 Zr 3.33E+00
kef f (t) =
P0 P0 + + 1, ζ(1 − P0 ) ΛQ
(1)
where P0 is the normalized nuclear power, Q is the external neutron source, Λ is the mean neutron generation time and ζ is a subcritical index. Fig 3 shows the keff evaluated as a function of the operation time in hours for fresh fuel.
keff
frame BERTINI Entries
other models present a broader width. Similar calculations are performed in [8].
0 0 0
INCL/ABLA
0.988
MCNPX Theoretical 0.9875
TABLE V. Average number of produced neutrons and energy for each model and package used in the target simulation. Model Bertini(MCNPX) Bertini(GEANT4) Binary(GEANT4) INCL/ABLA(GEANT4)
Mean BINARY RMS
0.987
0.9865
N E(MeV) 12.34 ± 0.02 13.26 ± 0.02 15.37 ± 0.02 9.87 ± 0.02 13.82 ± 0.02 11.37 ± 0.03 14.08 ± 0.02 11.92 ± 0.03
0.986 0
0.5
1
1.5
2
2.5 3 Time (hours)
FIG. 3. Values for keff as a function of time. mcnpx bert
Intensity
BERTINI
Entries
129
Mean 1.722 2.67 RMS
1.807 3.286
V.
BINARY
0.16
MCNPX
0.12
A complete MC simulation for the MYRRHA reactor was developed. It allowed to estimate parameters related to the produced neutrons in its spallation source using different hadronic interaction models. Using the results of this part as input for the reactor simulation it was possible to compare the estimated values of keff with the expected values from the inverse kinetic model. The estimated values from the simulations are relatively close to theory. The maximum difference observed was ∼ 100 pcm per point but mainly due to the observed statistical fluctuations while on average the maximum difference observed was 60 pcm with the INCL/ABLA model. The GEANT4 implementation of the Bertini model was the calculation which gave the keff estimates closest to the theoretical values. An average deviation below 6 pcm was observed for this model in particular.
0.1 0.08 0.06 0.04 0.02 0 0
CONCLUSIONS
INCL/ABLA
0.14
2
4
6
8
10 12 Energy (MeV)
FIG. 2. Normalized energy spectrum of the neutrons produced in the spallation source simulation.
Given the information in Table V and Fig. 2 it was possible to estimate the keff of the reactor and compare the obtained values with the theoretical expectation provided from the inverse point kinetic equations proposed
[6] A. Boudard et al., Phys. Rev. C 66, 044615 (2002). [7] J. Lepp˜ anen, SERPENT USER’S MANUAL, VTT Technical Research Center of Finland (2012). [8] C. Bungau et al., 23rd Part. Acc. Proc., TU6PFP029 (2009). [9] M. Antolin, PhD Thesis, Universidade Federal do Rio de Janeiro, Brazil (2013).
[1] A. Gandini, M. Salvatores, J. Nucl. Sci. Tech. 6, 673 (2002). [2] H A. Abderrahim et al., Ener. Conv. Man. 63, 4 (2012). [3] D.B. Pelowitz et al., MCNPX USER’S MANUAL V 2.6.0, LA-UR-08-2216 (2008). [4] S. Agostinelli et al., Nucl. Instrum. Methods Phys. Res. A 506, 250 (2003). [5] H.W. Bertini, P. Guthrie, Nucl. Phys. A 169, 670 (1971).
512
Nuclear Data Sheets 118 (2014) 510–512 www.elsevier.com/locate/nds
Spallation Source Modelling for an ADS Using the MCNPX and GEANT4 Packages for Sensitivity Analysis of Reactivity M.Q. Antolin,1, ∗ F. Marinho,2 D.A.P. Palma,3 and A.S. Martinez4 1
Centro Universitario Estadual da Zona Oeste - UEZO , Rio de Janeiro, 23070-200, Brazil Universidade Federal do Rio de Janeiro, Av. Aluizio da Silva Gomes, 50, Maca´e, 27930-560, Brazil 3 Comiss˜ ao Nacional de Energia Nuclear, Rua Gal Severiano, 90, Rio de Janeiro, 22290-901, Brazil 4 Instituto Alberto Luiz Coimbra de P´ os-Gradua¸c˜ ao e Pesquisa COPPE/UFRJ, Rio de Janeiro, 21941-972, Brazil 2
A simulation for the time evolution of the MYRRHA conceptual reactor was developed. The SERPENT code was used to simulate the nuclear fuel depletion and the spallation source which drives the system was simulated using both MCNPX and GEANT4 packages. The obtained results for the neutron energy spectrum from the spallation are coherent with each other and were used as input for the SERPENT code which simulated a constant power operation regime. The obtained results show that the criticality of the system is not sensitive to the spallation models employed and only relative small deviations with respect to the inverse kinetic model coming from the point kinetic equations proposed by Gandini were observed. I.
INTRODUCTION
A significant fraction of electric energy produced by nuclear reactors is from light water based reactors. The use of low-enriched fissile material in such reactors makes their commissioning and operation more attractive. However, these reactors use an open fuel cycle and generate a great deal of radioactive waste which includes transuranic elements (Am, Cm, Np, etc) with half-lives up to 107 years. Because of this, source-driven sub-critical reactors have been thought around the world as an efficient way for the transmuting of large volumes of waste, which helps reducing the volumes in geological storage [1]. In order to evaluate the feasibility and efficiency of an ADS reactor before its construction it is important to develop independent methods to ensure consistency on the results of such estimates. For that, Monte Carlo (MC) simulations can provide important information such as indicating advantages on the materials and geometries to be adopted in the commissioning phase. It can also allow one to validate new methods for calculation of important parameters such as the reactivity for both types of reactors (hybrid and thermal). In this work a set of specific MC codes was used to develop a complete simulation from the interactions of protons on the spallation source to the fuel depletion in the MYRRHA reactor. A brief description of the reactor is given in section II. Each simulation step is described in section III and results are presented in section IV. ∗
Corresponding author: [email protected]
http://dx.doi.org/10.1016/j.nds.2014.04.120 0090-3752/2014 Published by Elsevier B.V.
II.
REACTOR DESCRIPTION
The MYRRHA hybrid reactor project for high technology research purposes started in 1988 by the Belgian Nuclear Research Center (SCK·CEN). The main goal of MYRRHA is to demonstrate the ADS concept at reasonable power level, technological feasibility for transmuting of minor actinides and of long half-life fission products. MYRRHA is also meant for the development of lead alloys technologies for generation IV reactors [2]. The reactor is expected to start its operation in 2023 and is currently under licensing phase being one of the hybrid reactors with the highest rates of citation in the literature as it will serve the international research community. This reactor will be capable of operating in critical and sub-critical mode generating power in the 50-100 MWt range. The accelerator will inject protons at an energy of 600 MeV in the reactor core. III.
SIMULATION DESCRIPTION
The simulation developed in this work consisted of two parts: simulation of the spallation source; and the fuel depletion in the reactor. This section describes each of these parts. In order to simulate the interaction of the protons with the target source two software packages, MCNPX [3] and GEANT4 [4], were used to allow comparison between different interaction models and implementation codes. The target geometry consisted of a cylinder with 20 cm height, 7.2 cm diameter and its material being a 45%
Spallation Source Modelling . . .
M.Q. Antolin et al.
NUCLEAR DATA SHEETS
lead/55% bismuth alloy. The protons reached the target at a 600 MeV energy with same direction as the cylinder height. In order to provide a detailed picture of the inelastic scattering of protons with the target material an intranuclear cascade model was used followed by a set of deexcitation codes (precompound, break-up, etc) to deal with the excited nucleus at the end of the cascade. Three main sets of implementations were used in the simulation of the spallation source which differ mainly due to intranuclear cascade models. In the Bertini Cascade Model an incident projectile interacts with nucleons inside a nucleus in the material producing secondaries which may collide with other nucleons. After each interaction excitons are introduced within the nucleus and the final excited state of the nucleus is given by a sum of these states. The final state then decays via pre-equilibrium, evaporation, fragmentation and fission methods. An important feature of this model is that the nucleus is taken as a continuum medium [5]. Unlike the Bertini implementation the Binary Cascade Model describes the nucleus by displacing the nucleons in space according to nuclear density. On the other hand, primary and secondary particles interact with single nucleon as well. Particles propagation in nuclear field is obtained via a numerical method. After specific cascading conditions are no longer satisfied precompound and de-excitation treatment follow. A recent implementation dedicated to spallation source studies in GEANT4 is the so called INCL/ABLA model which is the combination of a C++ version of the intranuclear cascade Li`ege code with the evaporation ABLA code [6]. The Li`ege approach is such that the number of free parameters needed are comparatively reduced therefore its predictive power is a preserved feature. The SERPENT code was used to simulate the operation of the MYRRHA reactor during 3 hours using fresh fuel and burned fuel [7]. The power level considered in these simulations was a constant value of 80 MWt . This code is one of the few Monte Carlo tools which allows the simulation of a nuclear reactor with external source and multiplicative medium all together. It is a code for simulation of nuclear reactors having pre-defined elements being capable of generating constant multigroups, allowing analysis of the fuel cycle via burning calculation and uses the same libraries as MCNPX. The core of the MYRRHA reactor uses a hexagonal arrangement with 68 fuel elements, being the central position dedicated to the target for the spallation reaction. The geometry used on the core of the reactor is illustrated in Fig. 1, with its main geometric parameters listed in Table I. The composition of the materials used in the simulation is listed in Tables II, IV and III.
FIG. 1. Core geometry of the sub-critical reactor.
TABLE I. General parameters of the reactor. Parameter Value (cm) Fuel element pitch 14.685 Fuel rod pitch 0.9144 Fuel rod diameter 0.8048 Coating width 0.0523 Fuel active height 84.4108
TABLE II. Fuel rod coating (ρ = 7.5357 g/cm3 ). Element 54
Fe Fe 57 Fe 58 Fe 58 Ni 60 Ni 61 Ni 62 Ni 56
Concentration Element Concentration (atoms/barn cm) (atoms/barn cm) 64 4.08E+01 Ni 3.99E-02 50 6.41E+02 Cr 4.51E+00 52 1.48E+01 Cr 8.70E+01 53 1.97E+00 Cr 9.87E+00 54 2.93E+00 Cr 2.46E+00 55 1.13E+00 Mn 4.60E+00 40 4.91E-02 Zr 4.91E+00 1.57E-01
TABLE III. Control bar (ρ = 1, 5643 g/cm3 ). Element 12
C B 11 B 10
IV.
Concentration (atoms/barn cm) 4.07E+01 1.64E+02 6.62E+02
RESULTS
The products from the simulation of the spallation source served as input information to the SERPENT code to simulate the burnup in the reactor. Table V lists the average number of neutrons produced per colliding proton obtained for each of the interaction models used. Note the estimated values are similar even when comparing the results from both MCNPX and GEANT4 for the Bertini model. Fig. 2 shows the energy spectrum of the produced neutrons for each model. The MCNPX spectrum exhibits a sharper peak around 2 MeV while the 511
Spallation Source Modelling . . .
by Gandini [9]
TABLE IV. Reflector (ρ = 5, 3908 g/cm3 ). Element 23
Na 54 Fe 56 Fe 57 Fe 58 Fe 58 Ni 60 Ni 61 Ni
M.Q. Antolin et al.
NUCLEAR DATA SHEETS
Concentration Element Concentration (atoms/barn cm) (atoms/barn cm) 62 7.15E+01 Ni 1.06E-01 64 2.77E+01 Ni 2.71E-02 50 4.35E+02 Cr 3.06E+00 52 1.01E+01 Cr 5.91E+01 53 1.34E+00 Cr 6.70E+00 54 1.99E+00 Cr 1.67E+00 55 7.67E-01 Mn 3.12E+00 40 3.33E-02 Zr 3.33E+00
kef f (t) =
P0 P0 + + 1, ζ(1 − P0 ) ΛQ
(1)
where P0 is the normalized nuclear power, Q is the external neutron source, Λ is the mean neutron generation time and ζ is a subcritical index. Fig 3 shows the keff evaluated as a function of the operation time in hours for fresh fuel.
keff
frame BERTINI Entries
other models present a broader width. Similar calculations are performed in [8].
0 0 0
INCL/ABLA
0.988
MCNPX Theoretical 0.9875
TABLE V. Average number of produced neutrons and energy for each model and package used in the target simulation. Model Bertini(MCNPX) Bertini(GEANT4) Binary(GEANT4) INCL/ABLA(GEANT4)
Mean BINARY RMS
0.987
0.9865
N E(MeV) 12.34 ± 0.02 13.26 ± 0.02 15.37 ± 0.02 9.87 ± 0.02 13.82 ± 0.02 11.37 ± 0.03 14.08 ± 0.02 11.92 ± 0.03
0.986 0
0.5
1
1.5
2
2.5 3 Time (hours)
FIG. 3. Values for keff as a function of time. mcnpx bert
Intensity
BERTINI
Entries
129
Mean 1.722 2.67 RMS
1.807 3.286
V.
BINARY
0.16
MCNPX
0.12
A complete MC simulation for the MYRRHA reactor was developed. It allowed to estimate parameters related to the produced neutrons in its spallation source using different hadronic interaction models. Using the results of this part as input for the reactor simulation it was possible to compare the estimated values of keff with the expected values from the inverse kinetic model. The estimated values from the simulations are relatively close to theory. The maximum difference observed was ∼ 100 pcm per point but mainly due to the observed statistical fluctuations while on average the maximum difference observed was 60 pcm with the INCL/ABLA model. The GEANT4 implementation of the Bertini model was the calculation which gave the keff estimates closest to the theoretical values. An average deviation below 6 pcm was observed for this model in particular.
0.1 0.08 0.06 0.04 0.02 0 0
CONCLUSIONS
INCL/ABLA
0.14
2
4
6
8
10 12 Energy (MeV)
FIG. 2. Normalized energy spectrum of the neutrons produced in the spallation source simulation.
Given the information in Table V and Fig. 2 it was possible to estimate the keff of the reactor and compare the obtained values with the theoretical expectation provided from the inverse point kinetic equations proposed
[6] A. Boudard et al., Phys. Rev. C 66, 044615 (2002). [7] J. Lepp˜ anen, SERPENT USER’S MANUAL, VTT Technical Research Center of Finland (2012). [8] C. Bungau et al., 23rd Part. Acc. Proc., TU6PFP029 (2009). [9] M. Antolin, PhD Thesis, Universidade Federal do Rio de Janeiro, Brazil (2013).
[1] A. Gandini, M. Salvatores, J. Nucl. Sci. Tech. 6, 673 (2002). [2] H A. Abderrahim et al., Ener. Conv. Man. 63, 4 (2012). [3] D.B. Pelowitz et al., MCNPX USER’S MANUAL V 2.6.0, LA-UR-08-2216 (2008). [4] S. Agostinelli et al., Nucl. Instrum. Methods Phys. Res. A 506, 250 (2003). [5] H.W. Bertini, P. Guthrie, Nucl. Phys. A 169, 670 (1971).
512