A Systems Study of a Stirling Convertor based Space Nuclear Power System

In October of 2002, The Power and Propulsion Office and The Risk Management Office of NASA Glenn Research Center in Cleveland, Ohio began developing the reliability, availability, and maintainability (RAM) engineering approach for the Space Nuclear Propulsion and Power Systems Project. The objective of the Space Nuclear Power and Propulsion Project is to provide safe and reliable propulsion and power systems for planetary missions. The safety of the crew, ground personnel, and the public has to be the highest priority of the RAM engineering approach for nuclear powered space systems. The project will require a top level reliability goal for substantial mission success in the range from 0.95 to 0.98. In addition, the probability of safe operation without loss of crew, vehicle, or danger to the public, cannot be less than 0.9999. The achievement of these operational goals will require the combined application of many RAM engineering techniques. These include: advanced reliability, availability, and maintainability analysis, probabilistic risk assessment that includes hardware, software, and human induced faults, accelerated life testing, parts stress analysis, and selective end to end sub-system testing. Design strategy must involve the selection of parts and materials specifically to withstand the stresses of prolonged operation in the space and planetary environments with a wide design margin. Interplanetary distances and resulting signal time delay drive the need for autonomous control of major system functions including redundancy management.

Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) could provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program is the Topaz II reactor. This space nuclear power system was built and flight qualified, though never tested in space, by the former Soviet Union.The NEPSTP is faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system will complicate many spacecraft qualification activities. However, the NEPSTP activities are further complicated because the reactor power system is a Russian design. Therefore, this program must deal not only with the unique flight qualification issues associated with space nuclear power, but also ...

<p indent=0mm>After the Fukushima nuclear accident in Japan, the safety requirements of nuclear energy systems are constantly increasing. A series of new nuclear energy systems are under development, such as very high-temperature gas-cooled reactor (VHTR), lead-cooled fast reactor (LFR), gas-cooled fast reactor (GFR) and so on. In order to achieve the goals of higher safety, these nuclear systems have put forward higher requirements for nuclear materials. For example, materials working in VHTR and GFR are supposed to withstand higher temperatures, and LFR requires materials to be more corrosion-resistant. Nuclear materials have to meet these requirements under normal operating conditions, as well as in incidental and accidental conditions. Therefore, it is necessary to optimize the existing material system and develop new high-performance materials to meet the future development needs of advanced nuclear power systems. Compared with traditional nuclear materials, carbide ceramics have better mechanical properties under elevated temperatures, stronger irradiation resistance and better comprehensive thermophysical properties. Up to now, great breakthroughs have been made in the research of carbide ceramic materials and some key materials are going to be used in the new-generation nuclear systems. For example, uranium carbide (UC) is gradually replacing traditional uranium dioxide (UO₂), and silicon carbide (SiC) is acting as the main fission product barrier layer of coated fuel particles in HTGR. Other new carbide materials such as zirconium carbide (ZrC) and MAX phase carbide materials are also becoming a new research hotspot in the field of nuclear materials. Compared with other ceramic materials, the fabrication especially its densification of carbide is generally more difficult. In addition, high purity and good neutron properties are also the basic requirements of carbide ceramic materials for nuclear energy. The vapor deposition process and powder metallurgy process have been used to fabricate carbide ceramics for different components in nuclear reactors. Some new preparation methods have been developed to fabricate fuels and structural materials. For example, gelation progress is applied to prepare UCO microsphere fuel, and nano infiltration and transient eutectic (NITE) process is used in the preparation of SiC_f/SiC composites, which is potential for nuclear claddings. The irradiation-induced degradation of nuclear ceramic materials in the reactor is unavoidable. In order to evaluate the performance changes of materials under irradiation conditions, different irradiation methods such as electron irradiation, ion irradiation, proton irradiation and neutron irradiation are used. Due to its unique microstructure and properties, carbide ceramics exhibit different properties from other nuclear materials under irradiation. The performance changes of main nuclear carbide materials after irradiation are summarized, including microstructure evolution, macroscopic irradiation-induced swelling, changes in mechanical and thermal properties, etc. In general, based on the current research progress, the article reviews the application of carbide ceramics in new nuclear power systems and elaborates their service scenarios. Basic properties, preparation methods and irradiation properties of carbide ceramic materials are introduced in detail. The future development directions and the processing techniques are proposed. In the future, carbide ceramic materials should be further developed, and carbide ceramic materials will find diverse applications in the new generation nuclear systems.

Heat rejection with mechanical pumped cooling loop for lunar surface nuclear reactor power system

[Background]With the in-depth exploration of space,the power supply requirements for energy supply are gradually increasing.Space nuclear reactor power supply stands out in the deep space exploration mission due to its advantages of passivity,long life and high reliability,hence heat pipe nuclear reactor has become the focus of research in the field of space nuclear reactor.[Purpose]This study aims to evaluate the progress and technology of heat pipe space nuclear reactor used as power supply.[Method]First of all,the conceptual design of heatpipe reactor in the early stage and its application in Kilopower was summarized.Then the details of heatpipe power system(HPS),the heatpipe-operated mars exploration reactor(HOMER),safe affordable fission engine(SAFE)and Kilopower were investigated with emphasis on the structure design,fuel selection,heatpipe arrangement and power design of each reactor.[Results]This survey provides ideas and references for future research and design of heatpipe space nuclear reactor power system.[Conclusions]The design of heatpipe reactor is highly flexible and innovative.The use of different alkali metal heat pipes and thermoelectric conversion methods has a direct impact on the system power and the quality of the entire system.The combination of heat pipe and various thermoelectric conversion methods can be attempted.

Unshielded nuclear reactors provide the lightest and most survivable long‐lived sources of electric power available to support military satellites. Restricting their use now, before a new generation of larger space reactors is tested and deployed by the US and USSR, could help prevent an arms race in space. Space nuclear power systems have been used by the United States and the Soviet Union since the 1960s. The Soviet Union has used orbiting nuclear reactors to power more than 30 radar ocean reconnaissance satellites (RORSATs). Two RORSATs have accidentally re‐entered and released their radioactivity into the environment, and a third, Cosmos 1900, narrowly avoided a similar fate. The United States is developing much more powerful space reactors, of which the SP‐100 is farthest along, primarily to power satellite components of the Strategic Defense Initiative (SDI). A working group associated with the Federation of American Scientists (FAS) and the Committee of Soviet Scientists for Peace and Against the Nuclear Threat (CSS) has been studying a proposed ban on orbiting reactors. A proposal by the FAS/CSS group that includes such a ban is attached in the appendix to the Overview. The first five papers in this section, all by members of the working group, summarize the technological and historical background to nuclear power in space and show that restrictions on orbiting reactors are verifiable. The final paper, by Rosen and Schnyer of NASA, surveys the civilian uses of nuclear power in space. The overview is a nontechnical introduction to the issues of space reactor arms control, including the proposed ban on orbiting reactors.

Performance expectations of closed-Brayton-cycle heat exchangers to be used in 100-kWe nuclear space power systems were forecast. Proposed cycle state points for a system supporting a mission to three of Jupiter's moons required effectiveness values for the heat-source exchanger, recuperator, and rejection exchanger (gas cooler) of 0.98, 0.95, and 0.97, respectively. Performance parameters such as number of thermal units Ntu, equivalent thermal conductance UA, and entropy generation numbers Ns varied from 11 to 19, 23 to 39 kW/K, and 0.019 to 0.023 for some standard heat exchanger configurations. Pressure-loss contributions to entropy generation were significant; the largest frictional contribution was 114% of the heat-transfer irreversibility. Using conventional recuperator designs, the 0.95 effectiveness proved difficult to achieve without exceeding other performance targets; a metallic, plate-fin counterflow solution called for 15% more mass and 33% higher pressure loss than the target values. Two types of gas coolers showed promise. Single-pass counterflow and multipass cross-counterflow arrangements both met the 0.97 effectiveness requirement. Potential reliability-related advantages of the cross-counterflow design were noted. Cycle modifications, enhanced heat-transfer techniques, and incorporation of advanced materials were suggested options to reduce system development risk. Carbon-carbon sheeting or foam proved an attractive option to improve overall performance.

NASA has completed a preliminary mission and systems study of nuclear electric propulsion (NEP) systems for “split‐sprint” human exploration and related robotic cargo missions to Mars. This paper describes the study, the mission architecture selected, the NEP system and technology development needs, proposed development schedules, and estimated development costs. Since current administration policy makers have delayed funding for key technology development activities that could make Mars exploration missions a reality in the near future, NASA will have time to evaluate various alternate mission options, and it appears prudent to ensure that Mars mission plans focus on astronaut and mission safety, while reducing costs to acceptable levels. The split‐sprint nuclear electric propulsion system offers trip times comparable to nuclear thermal propulsion (NTP) systems, while providing mission abort opportunities that are not possible with “reference” mission architectures. Thus, NEP systems offer short transit times for the astronauts, reducing the exposure of the crew to intergalactic cosmic radiation. The high specific impulse of the NEP system, which leads to very low propellant requirements, results in significantly lower “initial mass in low earth orbit” (IMLEO). Launch vehicle packaging studies show that the NEP system can be launched, assembled, and deployed, with about one less 240‐metric‐ton heavy lift launch vehicle (HLLV) per mission opportunity ‐ a very large cost savings! Technology development cost of the nuclear reactor for an NEP system would be shared with the proposed nuclear surface power systems, since nuclear systems will be required to provide substantial electrical power on the surface of Mars. The NEP development project plan proposed includes evolutionary technology development for nuclear electric propulsion systems that expands upon SP‐100 (Space Power ‐ 100 kw) technology that has been developed for lunar and Mars surface nuclear power, and small NEP systems for interplanetary probes. System upgrades are expected to evolve that will result in even shorter trip times, improved payload capabilities, and enhanced safety and reliability. Non‐nuclear technology development for the NEP system is estimated to cost about $721 M (1993 $). Nuclear technology development costs are not included in the costs in this report, since these costs will be incurred in the nuclear surface power development program. NEP Phase A/B studies are estimated to cost about $154 M. Flight system hardware development (Phase C/D) is estimated to cost about $2.8 B, and fabrication of flight hardware is estimated to be about $7.8 B for four mission opportunities in 2009, 2011, 2014, and 2016.

There is a need in the space science community for nuclear‐powered electric propulsion systems to enable high‐value, deep space and planetary exploration. Certain missions are driven by once‐in‐a‐lifetime or highly infrequent occurrences that require the near‐term development of a flight‐capable nuclear space power and electric propulsion system in order to take advantage of the scientific opportunity. The broader applicability of Brayton power systems to the commercial and military aircraft markets has provided fertile ground for the continued development and implementation of new technologies applicable to a closed Brayton cycle space Power Conversion Unit (PCU). One concept for effectively achieving a near‐term Brayton space power capability is based on the development work associated with the Integrated Power Unit (IPU). This unit embodies the state of the art in turbomachinery, generators, bearing systems and electric power management and distribution capability that can readily be evolved into a closed Brayton cycle PCU. This paper provides an overview of aircraft‐based Brayton power system technologies, their implementation into the IPU and one approach for leveraging this capability into a near‐term closed Brayton cycle space power conversion unit.

Nuclear transport power systems (NTPS) can provide solving such important science, commerce and defense tasks in space as radar surveillance, information affording, global ecological monitoring, defense of Earth from dangerous space objects, manufacturing in space, investigations of asteroids, comets and solar systems’ planets (Kuzin et al. 1993a, 1993b). The creation of NTPS for real space systems, however, must be based on proved NTPS effectiveness in comparison with other power and propulsion systems such as, nonnuclear electric‐rocket systems and so on. When the NTPS effectiveness is proved, the operation safety of such systems must be suited to the UN requirements for all stages of the life cycle in view of possible failures. A nuclear transport power module provides both a large amount of thermal and electrical power and a long acting time (about 6–7 years after completing the delivery task). For this reason such module is featured with the high power supplying‐mass delivery effectiveness and the co...

The Automatic Control System (ACS) of the space nuclear reactor power system TOPAZ II that generates electricity from nuclear heat using in‐core thermionic converters is considered. Sliding Mode Control Technique was applied to the reactor system controller design in order to provide the robust high accuracy following of a neutron (thermal) power reference profile in a start up regime and a payload electric power (current) reference profile following in an operation regime. Extensive simulations of the TOPAZ II reactor system with the designed sliding mode controllers showed improved accuracy and robustness of the reactor system performances in a start up regime and in an electric power supply regime as well.

The automatic control system (ACS) of the space nuclear reactor power system TOPAZ II that generates electricity from nuclear heat using in-core thermionic converters is considered. The sliding mode control technique was applied to the reactor system controller design in order to provide the robust high accuracy following of a neutron (thermal) power reference profile in a start up regime and a payload electric power (current) reference profile following in an operation regime. Extensive simulations of the TOPAZ II reactor system with the designed sliding mode controllers showed improved accuracy and robustness of the reactor system performances in a start up regime and in an electric power supply regime as well.

Nuclear power sources have enabled or enhanced some of the most challenging and exciting space missions yet conducted, including missions such as the Pioneer flights to Jupiter and Saturn; the Voyager flights to Jupiter, Saturn, Uranus, and Neptune; the Apollo lunar surface experiments; the Viking Lander studies of Mars; the Ulysses mission to study the polar regions of the sun; and the Galileo mission to orbit Jupiter. This article surveys the early missions and continues to the present Galileo and Ulysses missions and provides a summary of the power technology that was used in each mission. Current activities aimed at improving the specific power and efficiency of space nuclear power sources are summarized. In addition to radioisotope power sources, this article surveys the U.S. space nuclear reactor program focusing on the flight of the SNAP-lOA reactor and the technology developed in the SP-100 space nuclear reactor power system program. The general attributes of space nuclear power sources are described and possible future applications are identified.

Life-cycle emission of oxidic gases from power-generation systems

A nuclear space power system that consists of a solid metal nuclear reactor core with heat pipes carrying energy to a cylindrical thermoelectric converter surrounding each of the heat pipes with a heat pipe radiator surrounding the thermoelectric converter is the most simple and reliable space power system. This means no single point of failure since each heat pipe and cylindrical converter is a separate power system and if one fails it will not affect the others. The heat pipe array in the solid core is designed so that if an isolated heat pipe or even two adjacent heat pipes fail, the remaining heat pipes will still transport the core heat without undue overheating of the uranium nitride fuel. The primary emphasis in this paper is on simplicity, reliability and fabricability of such a space nuclear power source. The core and heat pipes are made of Niobium 1% Zirconium alloy (Nb1Zr), with rhenium lined fuel tubes, bonded together by hot isostatic pressure, (HIPing) and with sodium as the heat pipe working fluid, can be operated up to 1250K. The cylindrical thermoelectric converter is made by depositing the constituents of the converter around a Nb1%Zr tube and encasing it in a Nb 1% Zr alloy tube and HIPing the structure to get final bonding and to produce residual compressive stresses in all brittle materials in the converter. A radiator heat pipe filled with potassium that operates at 850K is bonded to the outside of the cylindrical converter for cooling. The solid core heat pipe and cylindrical converter are mated by welding during the final assembly. A solid core reactor with 150 heat pipes with a 0.650‐inch (1.65 cm) ID and a 30‐inch (76.2 cm) length with an output of 8 Watts per square inch as demonstrated by the SP100 PD2 cell tests will produce about 80 KW of electrical power. An advanced solid core reactor made with molybdenum 47% rhenium alloy, with lithium heat pipes and the PD2 theoretical output of 11 watts per square inch or advanced higher temperature converter to operate at 1350K could produce a greater output of approximately 100KW.