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Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor
Annals of Nuclear Energy xxx (2016) xxx–xxx
Contents lists available at ScienceDirect
Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor Kevin J. Schillo a,⇑, Akansha Kumar b, Kurt E. Harris c, Yayu M. Hew d, Steven D. Howe e a
Department of Mechanical & Aerospace Engineering, University of Alabama in Huntsville, Huntsville, AL 35899, United States Center for Space Nuclear Research, Idaho National Laboratory, Idaho Falls, ID 83401, United States c Department of Mechanical & Aerospace Engineering, Utah State University, Logan, UT 84322, United States d Department of Aeronautics and Astronautics Engineering, Stanford University, Stanford, CA 94305, United States e Talos Power LLC, Idaho Falls, ID 83402, United States b
a r t i c l e
i n f o
Article history: Received 31 March 2016 Received in revised form 13 May 2016 Accepted 16 May 2016 Available online xxxx Keywords: Supercritical Neutronics Thermal hydraulics CO2 Mars
a b s t r a c t A fission reactor utilizing low-enriched uranium cermet fuel and supercritical carbon dioxide as coolant was designed to provide electrical power for a manned Mars base. The reactor was designed to generate 1.67 MWth for a fifteen year operational lifetime, with an electric output of 333 kWe. The core has alternating rows of fuel elements and a breeder blanket, with nineteen coolant channels in the fuel element and a single coolant channel in the breeder blanket. The design uses 15% isotopic enriched U-235 based cermet fuel, and uranium dioxide fuelled blankets. S-CO2 is used as a coolant, which converts the heat generated by the reactor to electricity using a closed Brayton cycle. Cermet fuel is used in the form of hexagonal shaped elements with 19 coolant channels and a zirconium hydride neutron moderator. The reactor uses B4C based control drums for control and safety. ZrC is used as thermal insulator, and ensures that the ZrH moderator does not reach an unacceptably high temperature. The coolant channels have a cladding of tungsten to prevent the release of fission gas from the fuel into the coolant. Beryllium reflectors are used to moderate and reflect neutrons back into the active core. The active core has a bull’s eye configuration, in which there are alternate fuel and blanket circular rows. Nuclear reactor modeling, neutronics, and depletion analysis were done using MCNP6. The neutronics analysis found the maximum peaking factor that would occur in the core. This was used to determine the greatest amount of thermal power that the core’s fuel elements and breeder blanket would experience. This provided the basis for the thermal hydraulics, which sought to determine the maximum inlet and outlet temperatures of the S-CO2 that could be obtained while also keeping all of the reactor materials within an acceptable temperature range. Maximizing these temperatures would provide the highest performance for a power conversion system. Finding the coolant conditions that kept this section of the core below the maximum permissible temperature would ensure that the rest of the core would also remain within an acceptable temperature limit. Simulations were conducted at the different power levels the fuel element and breeder blanket would generate throughout the reactor’s fifteen-year lifecycle. The thermal hydraulics which was done using COMSOL Multiphysics. This research presents a very viable reactor design that uses materials currently tested on other types of reactors. Ó 2016 Elsevier Ltd. All rights reserved.
1. Introduction In 2009, NASA identified a surface nuclear power reactor as a mission enabling technology for manned Mars expeditions by supplying the power needed for in-situ resource utilization (ISRU). Such a power system offers continuous power as well as lower
⇑ Corresponding author. E-mail address: [email protected] (K.J. Schillo).
mass and volume than an equivalent solar power system (Drake, 2009). The reactor designed uses a low-enriched uranium (LEU) cermet fuel developed at the Center for Space Nuclear Research (CSNR) O’Brien et al., 2012. A Brayton cycle was selected for the power conversion system due to its simplicity, low mass and historic use in spacecraft systems (Mason, 2001). Supercritical carbon dioxide (S-CO2) was selected as the reactor coolant due to the abundance of carbon dioxide in the Martian atmosphere. ISRU equipment that will also be needed for the Mars base will be
http://dx.doi.org/10.1016/j.anucene.2016.05.035 0306-4549/Ó 2016 Elsevier Ltd. All rights reserved.
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
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K.J. Schillo et al. / Annals of Nuclear Energy xxx (2016) xxx–xxx
Nomenclature CSNR ISRU LEU S-CO2
Center for Space Nuclear Research in-situ resource utilization low enriched uranium supercritical carbon dioxide
UO2 W ZrC ZrH
capable of extracting the CO2 from the atmosphere and bring it to supercritical conditions. This capability to produce reactor coolant on Mars is crucial for the long-term survival and prosperity of a colony (Figs. 1 and 7). This paper is part of a larger study into the design of a fission power reactor intended to provide a total of 1 MWe for a manned Mars base for fifteen years. The study analyzed the neutronics and thermal hydraulics of the core, as well as the power distribution and power conversion systems. The focus of this paper is the neutronics an thermal hydraulics analysis of the core. MCNP6 was used for the nuclear reactor modeling, neutronics, shielding, and depletion analyses. COMSOL Multiphysics is used to simulate the most extreme thermal power that would be subjected to a fuel element and breeder channel in the reactor core that has been designed. Maximum values for the inlet and outlet temperatures of the S-CO2 for both the fuel element and breeder channel were obtained while also ensuring that the materials in the core did not exceed their own maximum permissible temperatures. The thermal power generated by the fuel element and breeder channel varies as a function of time, and numerous simulations were conducted at different phases of the reactor’s operational lifetime. The maximum values for the S-CO2 inlet and outlet temperatures were obtained at these different thermal power levels. This provides an overview of how the flow conditions of the S-CO2 must be carefully managed as the reactor operates. 2. Reactor core design The design of the power conversion system found that the system mass would be optimized by using three separate nuclear reactors, with each generating 333 kWe. The mass of the power conversion system would also be minimized by operating at 20% efficiency. This necessitated that each of the three reactor cores generated 1.67 MWth. The reactor core has a bullseye configuration, in which there are alternating layers of fuel and breeder blanket hexagonal configurations. The cermet fuel used in the core is comprised of tungsten and uranium-thorium dioxide, with the uranium having an enrichment of 15%. The W-(U-Th)O2 cermet offers good thermal conduc-
uranium dioxide tungsten zirconium carbide zirconium hydride
tivity, a high melting point, resistance to creep deformations at high temperatures, and good radiation self-shielding. The fuel element has nineteen coolant channels Tungsten cladding is used in the coolant channels of the fuel element in order to prevent the dissipation of gas from the cermet fuel into the coolant. Zirconium hydride (ZrH) was selected as the neutron moderator for both the fuel element and the breeder blanket. Zirconium carbide (ZrC) is used as thermal insulator, and ensures that the ZrH moderator does not become as hot as the rest of the fuel element materials. The breeder blanket consists of natural uranium dioxide with a single coolant channel and is surrounded by additional ZrH. Fig. 2 illustrates this. B4C control drums are used in the core for control and safety. Beryllium reflectors are used to moderate and reflect neutrons back into the active core. Fig. 3 provides a radial overview of the core design: The specifications of the core are provided in the following table (see Table 1). For this study, the reactor must provide continuous 1.67 MWth for fifteen years of operation. It is therefore vital that the reactor remains critical for this entire duration of time. The reactor’s keff for this fifteen years of operation is shown in Fig. 4. From this figure, it can clearly be seen that the reactor will remain critical throughout the entire fifteen year lifecycle. Throughout the reactor’s operational lifetime, the U-235 in the fuel elements is gradually depleted as it undergoes fission reactions, causing the fuel elements to generate less power over an extended period of time. While this is happening, the U-238 in the breeder blanket is converted to Pu-239 from neutron capture. This causes the breeder blanket to generate more power as the reactor operates over time. Fig. 5 shows the change in the power fraction generated by the fuel element and the breeder blanket throughout fifteen years of the reactor’s operation: At the beginning of the reactor’s operation, the fuel element generates 83% of the thermal power and the breeder blanket generates 17%. After fifteen years, the fuel element will generate 76% and the breeder blanket will generate 24%. The main limiting factor for the reactor was the use of ZrH. While ZrH is a very effective neutron moderator, it can only be permitted to reach a maximum temperature of 923 K before it
Fig. 1. Cermet fuel element.
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
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K.J. Schillo et al. / Annals of Nuclear Energy xxx (2016) xxx–xxx 1.12 1.1
keff
1.08 1.06 1.04 1.02 1 0.98 0.0
2.0
4.0
6.0
8.0
10.0
12.0
14.0
Time (years) Fig. 4. Keff throughout the reactor’s fifteen years of operation.
0.90 0.80
Fig. 2. Breeder blanket.
Power Fraction
0.70 0.60
Fuel
Blanket
0.50 0.40 0.30 0.20 0.10 0.00 0.0
2.0
4.0
6.0
8.0
10.0
12.0
14.0
Time (years) Fig. 5. Fraction of power generated by the fuel elements and breeder blanket throughout the reactor’s lifecycle.
2
Fig. 3. Radial overview of core.
Table 1 Core Specifications. Parameter
Value
Core diameter Core length Number of fuel elements Number of breeder blankets Number of coolant channels in fuel element Number of coolant channels in breeder blanket Fuel element coolant channel radius Breeder blanket coolant channel radius Pitch of fuel hex Pitch of blanket hex Annular outer radius of breeder blanket UO2
80 cm 134 cm 337 384 19 1 0.17 cm 0.27 cm 1.90 cm 2.81 cm 0.80 cm
dissociates (Ponomarev-Stepnoi et al., 2007). This is the lowest operating temperature of all the materials used in the core. The loss of hydrogen in the ZrH leads to corresponding loss of reactivity in the reactor. This is shown in Fig. 6. Even a small amount of hydrogen loss in the ZrH leads to an unacceptable amount of reactivity loss in the core, with a ten percent loss leading to a reactivity loss of about $1.80. This makes it crucial that the ZrH is kept below 923 K in order to prevent any hydrogen loss. Ensuring that this is achieved was one of the main objectives of the thermal hydraulics analysis of the core, which is provided in the following section.
Reactivity Loss ($)
1.8 1.6 1.4 1.2 1 0.8 0.6 0.4 0.2 0 2
3
4
5 6 7 Hydrogen Loss (%)
8
9
10
Fig. 6. Corresponding reactivity loss as hydrogen diffuses from ZrH.
3. Thermal hydraulics analys of core ZrC was used as a thermal insulator to prevent the ZrH from reaching 923 K while also permitting the rest of the core to achieve higher temperatures. S-CO2 was selected as the reactor’s coolant due to the abundance of carbon dioxide in the Martian atmosphere. In order for carbon dioxide to be supercritical, it must be kept at a minimum temperature and pressure of 304.25 K and 7.38 MPa, respectively (Thermophysical Properties of Fluid Systems, 2011). The flow velocity of the S-CO2 must also be restricted to a maximum value of 3 m/s (Upgraded Calculator for CO2 Pipeline Systems, 2009). The principle objective of the thermal hydraulics analysis was to determine the maximum inlet and outlet temperatures of the
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
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K.J. Schillo et al. / Annals of Nuclear Energy xxx (2016) xxx–xxx
Fig. 7. COMSOL mesh of fuel element (left) and Breeder Blanket (right).
S-CO2 for the fuel element and the breeder channel while keeping the S-CO2 in supercritical conditions and keeping the ZrH below 923 K. The reason for this is because maximizing the inlet and outlet temperatures would also help to minimize the mass of the power conversion system, due to increased efficiency. COMSOL Multiphysics was used for all of the thermal hydraulics simulations. S-CO2 data was obtained from the NIST Standard Reference Database and then incorporated into the COMSOL model using tables and a piecewise cubic interpolation function. The pressure of the S-CO2 was set at 7.4 MPa and the outlet velocity was set to 3 m/s. The neutronics analysis found a maximum peaking factor of 1.6 in the core design. The COMSOL model simulated a single fuel element and breeder channel with this peaking factor. This represents the maximum thermal power generated in the core, and provides a conservative value for the thermal hydraulics analysis. As such, if this simulation provided satisfactory results, then the rest of the core will also have acceptable temperature ranges. To save computational power, the fuel element and breeder channel were partitioned along their two planes of symmetry. This allowed for only one quarter of each of these reactor components to be simulated while providing the same results that would be obtained if the entire component had been simulated. The meshes for the partitioned fuel element and breeder channel are provided in the following figure: As discussed in the neutronics section of this paper, the fuel elements provide 83% of the thermal power and the breeder blankets provide 17% of the thermal power at the beginning of the reactor’s operation. This provided the basis for the first thermal hydraulics simulations of the fuel element and breeder blanket. The variable that was examined and manipulated in these simulations was the inlet temperature of the S-CO2. The inlet temperature of the S-CO2 was the same for each of the coolant channels. Multiple simulations were conducted to determine the maximum value the inlet temperature could have while keeping the ZrH from reaching a maximum value of 923 K. After numerous were conducted, the maximum inlet temperature was found to be 753 K. The outlet temperature of the S-CO2 was found to be 909.7 K. The maximum temperatures for each of the different materials in the fuel element are presented in the following table:
Table 2 Maximum temperatures of fuel element materials at beginning of reactor operational lifetime. Material
Maximum temperature (K)
W LEU ZrC ZrH
926.3 926.4 926.4 922.4
The maximum temperature of the ZrH is shown to be kept just below 923 K, ensuring that dissociation does not occur. The temperature distribution of the fuel element is presented in the following figure: The maximum temperatures for the different materials reported in Table 2 occur at a height of about 120 cm. Heat that is generated by the fuel element is being conducted into the surrounding breeder blanket. Due to the heat that is dumped into the breeder blanket and the lower total coolant channel volume, the inlet temperature of the S-CO2 for the breeder blanket must be much lower than it is for the fuel element. After numerous simulations were conducted, the maximum inlet and outlet temperatures were found to be 578 K and 794.6 K, respectively. The maximum temperatures for each of the different materials in the breeder blanket are presented in the following table (see Table 3). As with the fuel element, the maximum temperature of the ZrH was kept below 923 K with the inlet and outlet temperatures of the S-CO2. The temperature distribution of the breeder blanket is presented in the following figure: With the bullseye configuration of the fuel elements and breeder blanket depicted in Fig. 3, each of the breeder blankets is in physical contact with four fuel elements and two other breeder blankets. This causes a large amount of heat to be conducted from the fuel elements into the breeder blanket, the effect of which can be seen in Fig. 9. The highest temperatures to occur along the wall of the breeder blanket. The highest material temperatures also occurred at about 120 cm from the bottom of the breeder blanket, corresponding to about the same location of the highest temperatures in the fuel element. Additional simulations of both the fuel element and breeder blanket with thermal power levels that the two components would experience at various times during the reactor’s fifteen years of operation. One simulated power level was after fifteen years of reactor operation, in which the fuel element and breeder blanket generate 76% and 24% of the thermal power, respectively. Other simulations were conducted for the two reactor components with thermal power levels between those experienced at the beginning and end of the reactor’s lifecycle. As with the first series of simulations, the principle objective was to determine the maximum inlet temperature that the S-CO2 could have while also keeping the ZrH below 923 K.
Table 3 Maximum temperatures of breeder blanket materials at beginning of reactor operational lifetime. Material
Maximum temperature (K)
UO2 ZrH
876.2 922.2
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
K.J. Schillo et al. / Annals of Nuclear Energy xxx (2016) xxx–xxx
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Fig. 8. Temperature distribution for exterior (left) and interior (right) of fuel element at beginning of reactor operational lifetime.
Fig. 9. Temperature distribution for exterior (left) and interior (right) of breeder blanket at beginning of reactor operational lifetime.
Fig. 10. S-CO2 inlet and outlet temperatures in fuel element as a function of thermal power.
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
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K.J. Schillo et al. / Annals of Nuclear Energy xxx (2016) xxx–xxx
Fig. 11. S-CO2 inlet and outlet temperatures in breeder blanket as a function of thermal power.
The temperature distributions followed the same general patterns seen in Figs. 8 and 9 at each of the different thermal power levels. Figs. 10 and 11 show the changes in the inlet and outlet temperatures for both the fuel element and the breeder blanket throughout the reactor’s lifecycle. As the reactor operates, the U-235 in the fuel element is depleted, causing a decrease in the amount of thermal power the fuel element generates. This in turn allows for the maximum temperature of the S-CO2 to reach higher values while also retaining nearly the same outlet temperature. The total increase in the inlet temperature for the fuel element from the beginning to the end of the reactor’s lifecycle is about 14 K. For the breeder blanket, both the inlet and outlet temperatures must decrease as the reactor operates over an extended period of time. This can be attributed to the larger amount of thermal power that the breeder blanket generates as the U-238 transmutes to Pu-239. The maximum inlet temperature and outlet temperatures decreased by about 98 K and 26 K throughout the reactor’s entire fifteen year lifecycle, respectively.
The inlet and outlet temperatures of the S-CO2 for both the fuel element and the breeder channel must change over time as a result of the fuel element generating less thermal power and the breeder channel generating more power as the U-235 is consumed and Pu-239 is produced. The ZrH remained below 923 K for all of the simulations that were conducted, ensuring that it does not dissociate when the reactor is operating. The pressure drop in both the fuel element and the breeder channel was low enough to ensure that the carbon dioxide remains in the supercritical regime throughout the reactor’s operational lifetime. The continued development of technologies such as this is paramount for the future manned exploration of Mars.
4. Conclusion
References
This paper presents a very viable reactor design that uses materials currently tested on other types of reactors. The reactor has been shown to be remain critical for 15 years of operation while generating continuous 1.67 MWth. Using a Brayton cycle power conversion system with 20% efficiency, the reactor will be capable of generating 333 kWe. Three such reactors would be capable of supplying 1 MWe for a manned Mars base over a fifteen year time frame, which would be more than sufficient power for both the habitat and ISRU operations. S-CO2 is a particularly attractive coolant to be used for such a reactor due to the abundance of carbon dioxide in the Martian atmosphere.
Drake, B., 2009. Human Exploration of Mars Design Reference Architecture 5.0 Addendum. NASA, Houston, TX. O’Brien, R., Jerred, N., Howe, S., Samborsky, R., Brasuell, D., Zillmer, A., 2012. Recent research activities at the center for space nuclear research in support of the development of nuclear thermal rocket propulsion. Nuclear and Emerging Technologies for Space. Mason, L.S., 2001. A Comparison of Brayton and Stirling Space Nuclear Power Systems for Power Levels from 1 Kilowatt to 10 Megawatts. NASA Glenn Research Center, Cleveland, Ohio. Ponomarev-Stepnoi, N., Bubelev, V., Glushkov, E., Garin, V., Kukharkin, N., Kompaniets, G., Nosov, V., Chunyaev, E., 2007. Estimation of the hydrogen emission from a hydride moderator by measuring the reactivity and using mathematical statistics. At. Energ. 102 (2), 87–93. Thermophysical Properties of Fluid Systems, NIST, 2011. [Online]. Available:
Acknowledgments The authors of this paper wish to thank all of the staff and personnel at the CSNR and the Idaho National Laboratory for providing their summer fellowship opportunity, and for their ceaseless guidance and support throughout the course of this work.
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
Contents lists available at ScienceDirect
Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor Kevin J. Schillo a,⇑, Akansha Kumar b, Kurt E. Harris c, Yayu M. Hew d, Steven D. Howe e a
Department of Mechanical & Aerospace Engineering, University of Alabama in Huntsville, Huntsville, AL 35899, United States Center for Space Nuclear Research, Idaho National Laboratory, Idaho Falls, ID 83401, United States c Department of Mechanical & Aerospace Engineering, Utah State University, Logan, UT 84322, United States d Department of Aeronautics and Astronautics Engineering, Stanford University, Stanford, CA 94305, United States e Talos Power LLC, Idaho Falls, ID 83402, United States b
a r t i c l e
i n f o
Article history: Received 31 March 2016 Received in revised form 13 May 2016 Accepted 16 May 2016 Available online xxxx Keywords: Supercritical Neutronics Thermal hydraulics CO2 Mars
a b s t r a c t A fission reactor utilizing low-enriched uranium cermet fuel and supercritical carbon dioxide as coolant was designed to provide electrical power for a manned Mars base. The reactor was designed to generate 1.67 MWth for a fifteen year operational lifetime, with an electric output of 333 kWe. The core has alternating rows of fuel elements and a breeder blanket, with nineteen coolant channels in the fuel element and a single coolant channel in the breeder blanket. The design uses 15% isotopic enriched U-235 based cermet fuel, and uranium dioxide fuelled blankets. S-CO2 is used as a coolant, which converts the heat generated by the reactor to electricity using a closed Brayton cycle. Cermet fuel is used in the form of hexagonal shaped elements with 19 coolant channels and a zirconium hydride neutron moderator. The reactor uses B4C based control drums for control and safety. ZrC is used as thermal insulator, and ensures that the ZrH moderator does not reach an unacceptably high temperature. The coolant channels have a cladding of tungsten to prevent the release of fission gas from the fuel into the coolant. Beryllium reflectors are used to moderate and reflect neutrons back into the active core. The active core has a bull’s eye configuration, in which there are alternate fuel and blanket circular rows. Nuclear reactor modeling, neutronics, and depletion analysis were done using MCNP6. The neutronics analysis found the maximum peaking factor that would occur in the core. This was used to determine the greatest amount of thermal power that the core’s fuel elements and breeder blanket would experience. This provided the basis for the thermal hydraulics, which sought to determine the maximum inlet and outlet temperatures of the S-CO2 that could be obtained while also keeping all of the reactor materials within an acceptable temperature range. Maximizing these temperatures would provide the highest performance for a power conversion system. Finding the coolant conditions that kept this section of the core below the maximum permissible temperature would ensure that the rest of the core would also remain within an acceptable temperature limit. Simulations were conducted at the different power levels the fuel element and breeder blanket would generate throughout the reactor’s fifteen-year lifecycle. The thermal hydraulics which was done using COMSOL Multiphysics. This research presents a very viable reactor design that uses materials currently tested on other types of reactors. Ó 2016 Elsevier Ltd. All rights reserved.
1. Introduction In 2009, NASA identified a surface nuclear power reactor as a mission enabling technology for manned Mars expeditions by supplying the power needed for in-situ resource utilization (ISRU). Such a power system offers continuous power as well as lower
⇑ Corresponding author. E-mail address: [email protected] (K.J. Schillo).
mass and volume than an equivalent solar power system (Drake, 2009). The reactor designed uses a low-enriched uranium (LEU) cermet fuel developed at the Center for Space Nuclear Research (CSNR) O’Brien et al., 2012. A Brayton cycle was selected for the power conversion system due to its simplicity, low mass and historic use in spacecraft systems (Mason, 2001). Supercritical carbon dioxide (S-CO2) was selected as the reactor coolant due to the abundance of carbon dioxide in the Martian atmosphere. ISRU equipment that will also be needed for the Mars base will be
http://dx.doi.org/10.1016/j.anucene.2016.05.035 0306-4549/Ó 2016 Elsevier Ltd. All rights reserved.
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
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K.J. Schillo et al. / Annals of Nuclear Energy xxx (2016) xxx–xxx
Nomenclature CSNR ISRU LEU S-CO2
Center for Space Nuclear Research in-situ resource utilization low enriched uranium supercritical carbon dioxide
UO2 W ZrC ZrH
capable of extracting the CO2 from the atmosphere and bring it to supercritical conditions. This capability to produce reactor coolant on Mars is crucial for the long-term survival and prosperity of a colony (Figs. 1 and 7). This paper is part of a larger study into the design of a fission power reactor intended to provide a total of 1 MWe for a manned Mars base for fifteen years. The study analyzed the neutronics and thermal hydraulics of the core, as well as the power distribution and power conversion systems. The focus of this paper is the neutronics an thermal hydraulics analysis of the core. MCNP6 was used for the nuclear reactor modeling, neutronics, shielding, and depletion analyses. COMSOL Multiphysics is used to simulate the most extreme thermal power that would be subjected to a fuel element and breeder channel in the reactor core that has been designed. Maximum values for the inlet and outlet temperatures of the S-CO2 for both the fuel element and breeder channel were obtained while also ensuring that the materials in the core did not exceed their own maximum permissible temperatures. The thermal power generated by the fuel element and breeder channel varies as a function of time, and numerous simulations were conducted at different phases of the reactor’s operational lifetime. The maximum values for the S-CO2 inlet and outlet temperatures were obtained at these different thermal power levels. This provides an overview of how the flow conditions of the S-CO2 must be carefully managed as the reactor operates. 2. Reactor core design The design of the power conversion system found that the system mass would be optimized by using three separate nuclear reactors, with each generating 333 kWe. The mass of the power conversion system would also be minimized by operating at 20% efficiency. This necessitated that each of the three reactor cores generated 1.67 MWth. The reactor core has a bullseye configuration, in which there are alternating layers of fuel and breeder blanket hexagonal configurations. The cermet fuel used in the core is comprised of tungsten and uranium-thorium dioxide, with the uranium having an enrichment of 15%. The W-(U-Th)O2 cermet offers good thermal conduc-
uranium dioxide tungsten zirconium carbide zirconium hydride
tivity, a high melting point, resistance to creep deformations at high temperatures, and good radiation self-shielding. The fuel element has nineteen coolant channels Tungsten cladding is used in the coolant channels of the fuel element in order to prevent the dissipation of gas from the cermet fuel into the coolant. Zirconium hydride (ZrH) was selected as the neutron moderator for both the fuel element and the breeder blanket. Zirconium carbide (ZrC) is used as thermal insulator, and ensures that the ZrH moderator does not become as hot as the rest of the fuel element materials. The breeder blanket consists of natural uranium dioxide with a single coolant channel and is surrounded by additional ZrH. Fig. 2 illustrates this. B4C control drums are used in the core for control and safety. Beryllium reflectors are used to moderate and reflect neutrons back into the active core. Fig. 3 provides a radial overview of the core design: The specifications of the core are provided in the following table (see Table 1). For this study, the reactor must provide continuous 1.67 MWth for fifteen years of operation. It is therefore vital that the reactor remains critical for this entire duration of time. The reactor’s keff for this fifteen years of operation is shown in Fig. 4. From this figure, it can clearly be seen that the reactor will remain critical throughout the entire fifteen year lifecycle. Throughout the reactor’s operational lifetime, the U-235 in the fuel elements is gradually depleted as it undergoes fission reactions, causing the fuel elements to generate less power over an extended period of time. While this is happening, the U-238 in the breeder blanket is converted to Pu-239 from neutron capture. This causes the breeder blanket to generate more power as the reactor operates over time. Fig. 5 shows the change in the power fraction generated by the fuel element and the breeder blanket throughout fifteen years of the reactor’s operation: At the beginning of the reactor’s operation, the fuel element generates 83% of the thermal power and the breeder blanket generates 17%. After fifteen years, the fuel element will generate 76% and the breeder blanket will generate 24%. The main limiting factor for the reactor was the use of ZrH. While ZrH is a very effective neutron moderator, it can only be permitted to reach a maximum temperature of 923 K before it
Fig. 1. Cermet fuel element.
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
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K.J. Schillo et al. / Annals of Nuclear Energy xxx (2016) xxx–xxx 1.12 1.1
keff
1.08 1.06 1.04 1.02 1 0.98 0.0
2.0
4.0
6.0
8.0
10.0
12.0
14.0
Time (years) Fig. 4. Keff throughout the reactor’s fifteen years of operation.
0.90 0.80
Fig. 2. Breeder blanket.
Power Fraction
0.70 0.60
Fuel
Blanket
0.50 0.40 0.30 0.20 0.10 0.00 0.0
2.0
4.0
6.0
8.0
10.0
12.0
14.0
Time (years) Fig. 5. Fraction of power generated by the fuel elements and breeder blanket throughout the reactor’s lifecycle.
2
Fig. 3. Radial overview of core.
Table 1 Core Specifications. Parameter
Value
Core diameter Core length Number of fuel elements Number of breeder blankets Number of coolant channels in fuel element Number of coolant channels in breeder blanket Fuel element coolant channel radius Breeder blanket coolant channel radius Pitch of fuel hex Pitch of blanket hex Annular outer radius of breeder blanket UO2
80 cm 134 cm 337 384 19 1 0.17 cm 0.27 cm 1.90 cm 2.81 cm 0.80 cm
dissociates (Ponomarev-Stepnoi et al., 2007). This is the lowest operating temperature of all the materials used in the core. The loss of hydrogen in the ZrH leads to corresponding loss of reactivity in the reactor. This is shown in Fig. 6. Even a small amount of hydrogen loss in the ZrH leads to an unacceptable amount of reactivity loss in the core, with a ten percent loss leading to a reactivity loss of about $1.80. This makes it crucial that the ZrH is kept below 923 K in order to prevent any hydrogen loss. Ensuring that this is achieved was one of the main objectives of the thermal hydraulics analysis of the core, which is provided in the following section.
Reactivity Loss ($)
1.8 1.6 1.4 1.2 1 0.8 0.6 0.4 0.2 0 2
3
4
5 6 7 Hydrogen Loss (%)
8
9
10
Fig. 6. Corresponding reactivity loss as hydrogen diffuses from ZrH.
3. Thermal hydraulics analys of core ZrC was used as a thermal insulator to prevent the ZrH from reaching 923 K while also permitting the rest of the core to achieve higher temperatures. S-CO2 was selected as the reactor’s coolant due to the abundance of carbon dioxide in the Martian atmosphere. In order for carbon dioxide to be supercritical, it must be kept at a minimum temperature and pressure of 304.25 K and 7.38 MPa, respectively (Thermophysical Properties of Fluid Systems, 2011). The flow velocity of the S-CO2 must also be restricted to a maximum value of 3 m/s (Upgraded Calculator for CO2 Pipeline Systems, 2009). The principle objective of the thermal hydraulics analysis was to determine the maximum inlet and outlet temperatures of the
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
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Fig. 7. COMSOL mesh of fuel element (left) and Breeder Blanket (right).
S-CO2 for the fuel element and the breeder channel while keeping the S-CO2 in supercritical conditions and keeping the ZrH below 923 K. The reason for this is because maximizing the inlet and outlet temperatures would also help to minimize the mass of the power conversion system, due to increased efficiency. COMSOL Multiphysics was used for all of the thermal hydraulics simulations. S-CO2 data was obtained from the NIST Standard Reference Database and then incorporated into the COMSOL model using tables and a piecewise cubic interpolation function. The pressure of the S-CO2 was set at 7.4 MPa and the outlet velocity was set to 3 m/s. The neutronics analysis found a maximum peaking factor of 1.6 in the core design. The COMSOL model simulated a single fuel element and breeder channel with this peaking factor. This represents the maximum thermal power generated in the core, and provides a conservative value for the thermal hydraulics analysis. As such, if this simulation provided satisfactory results, then the rest of the core will also have acceptable temperature ranges. To save computational power, the fuel element and breeder channel were partitioned along their two planes of symmetry. This allowed for only one quarter of each of these reactor components to be simulated while providing the same results that would be obtained if the entire component had been simulated. The meshes for the partitioned fuel element and breeder channel are provided in the following figure: As discussed in the neutronics section of this paper, the fuel elements provide 83% of the thermal power and the breeder blankets provide 17% of the thermal power at the beginning of the reactor’s operation. This provided the basis for the first thermal hydraulics simulations of the fuel element and breeder blanket. The variable that was examined and manipulated in these simulations was the inlet temperature of the S-CO2. The inlet temperature of the S-CO2 was the same for each of the coolant channels. Multiple simulations were conducted to determine the maximum value the inlet temperature could have while keeping the ZrH from reaching a maximum value of 923 K. After numerous were conducted, the maximum inlet temperature was found to be 753 K. The outlet temperature of the S-CO2 was found to be 909.7 K. The maximum temperatures for each of the different materials in the fuel element are presented in the following table:
Table 2 Maximum temperatures of fuel element materials at beginning of reactor operational lifetime. Material
Maximum temperature (K)
W LEU ZrC ZrH
926.3 926.4 926.4 922.4
The maximum temperature of the ZrH is shown to be kept just below 923 K, ensuring that dissociation does not occur. The temperature distribution of the fuel element is presented in the following figure: The maximum temperatures for the different materials reported in Table 2 occur at a height of about 120 cm. Heat that is generated by the fuel element is being conducted into the surrounding breeder blanket. Due to the heat that is dumped into the breeder blanket and the lower total coolant channel volume, the inlet temperature of the S-CO2 for the breeder blanket must be much lower than it is for the fuel element. After numerous simulations were conducted, the maximum inlet and outlet temperatures were found to be 578 K and 794.6 K, respectively. The maximum temperatures for each of the different materials in the breeder blanket are presented in the following table (see Table 3). As with the fuel element, the maximum temperature of the ZrH was kept below 923 K with the inlet and outlet temperatures of the S-CO2. The temperature distribution of the breeder blanket is presented in the following figure: With the bullseye configuration of the fuel elements and breeder blanket depicted in Fig. 3, each of the breeder blankets is in physical contact with four fuel elements and two other breeder blankets. This causes a large amount of heat to be conducted from the fuel elements into the breeder blanket, the effect of which can be seen in Fig. 9. The highest temperatures to occur along the wall of the breeder blanket. The highest material temperatures also occurred at about 120 cm from the bottom of the breeder blanket, corresponding to about the same location of the highest temperatures in the fuel element. Additional simulations of both the fuel element and breeder blanket with thermal power levels that the two components would experience at various times during the reactor’s fifteen years of operation. One simulated power level was after fifteen years of reactor operation, in which the fuel element and breeder blanket generate 76% and 24% of the thermal power, respectively. Other simulations were conducted for the two reactor components with thermal power levels between those experienced at the beginning and end of the reactor’s lifecycle. As with the first series of simulations, the principle objective was to determine the maximum inlet temperature that the S-CO2 could have while also keeping the ZrH below 923 K.
Table 3 Maximum temperatures of breeder blanket materials at beginning of reactor operational lifetime. Material
Maximum temperature (K)
UO2 ZrH
876.2 922.2
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
K.J. Schillo et al. / Annals of Nuclear Energy xxx (2016) xxx–xxx
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Fig. 8. Temperature distribution for exterior (left) and interior (right) of fuel element at beginning of reactor operational lifetime.
Fig. 9. Temperature distribution for exterior (left) and interior (right) of breeder blanket at beginning of reactor operational lifetime.
Fig. 10. S-CO2 inlet and outlet temperatures in fuel element as a function of thermal power.
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035
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Fig. 11. S-CO2 inlet and outlet temperatures in breeder blanket as a function of thermal power.
The temperature distributions followed the same general patterns seen in Figs. 8 and 9 at each of the different thermal power levels. Figs. 10 and 11 show the changes in the inlet and outlet temperatures for both the fuel element and the breeder blanket throughout the reactor’s lifecycle. As the reactor operates, the U-235 in the fuel element is depleted, causing a decrease in the amount of thermal power the fuel element generates. This in turn allows for the maximum temperature of the S-CO2 to reach higher values while also retaining nearly the same outlet temperature. The total increase in the inlet temperature for the fuel element from the beginning to the end of the reactor’s lifecycle is about 14 K. For the breeder blanket, both the inlet and outlet temperatures must decrease as the reactor operates over an extended period of time. This can be attributed to the larger amount of thermal power that the breeder blanket generates as the U-238 transmutes to Pu-239. The maximum inlet temperature and outlet temperatures decreased by about 98 K and 26 K throughout the reactor’s entire fifteen year lifecycle, respectively.
The inlet and outlet temperatures of the S-CO2 for both the fuel element and the breeder channel must change over time as a result of the fuel element generating less thermal power and the breeder channel generating more power as the U-235 is consumed and Pu-239 is produced. The ZrH remained below 923 K for all of the simulations that were conducted, ensuring that it does not dissociate when the reactor is operating. The pressure drop in both the fuel element and the breeder channel was low enough to ensure that the carbon dioxide remains in the supercritical regime throughout the reactor’s operational lifetime. The continued development of technologies such as this is paramount for the future manned exploration of Mars.
4. Conclusion
References
This paper presents a very viable reactor design that uses materials currently tested on other types of reactors. The reactor has been shown to be remain critical for 15 years of operation while generating continuous 1.67 MWth. Using a Brayton cycle power conversion system with 20% efficiency, the reactor will be capable of generating 333 kWe. Three such reactors would be capable of supplying 1 MWe for a manned Mars base over a fifteen year time frame, which would be more than sufficient power for both the habitat and ISRU operations. S-CO2 is a particularly attractive coolant to be used for such a reactor due to the abundance of carbon dioxide in the Martian atmosphere.
Drake, B., 2009. Human Exploration of Mars Design Reference Architecture 5.0 Addendum. NASA, Houston, TX. O’Brien, R., Jerred, N., Howe, S., Samborsky, R., Brasuell, D., Zillmer, A., 2012. Recent research activities at the center for space nuclear research in support of the development of nuclear thermal rocket propulsion. Nuclear and Emerging Technologies for Space. Mason, L.S., 2001. A Comparison of Brayton and Stirling Space Nuclear Power Systems for Power Levels from 1 Kilowatt to 10 Megawatts. NASA Glenn Research Center, Cleveland, Ohio. Ponomarev-Stepnoi, N., Bubelev, V., Glushkov, E., Garin, V., Kukharkin, N., Kompaniets, G., Nosov, V., Chunyaev, E., 2007. Estimation of the hydrogen emission from a hydride moderator by measuring the reactivity and using mathematical statistics. At. Energ. 102 (2), 87–93. Thermophysical Properties of Fluid Systems, NIST, 2011. [Online]. Available:
Acknowledgments The authors of this paper wish to thank all of the staff and personnel at the CSNR and the Idaho National Laboratory for providing their summer fellowship opportunity, and for their ceaseless guidance and support throughout the course of this work.
Please cite this article in press as: Schillo, K.J., et al. Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor. Ann. Nucl. Energy (2016), http://dx.doi.org/10.1016/j.anucene.2016.05.035