Safe radioisotope thermoelectric generators and heat sources for space applications

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Abstract

Several isotopes are examined as alternatives to 238Pu that is traditionally used in radioisotope thermoelectric generators (RTGs) and heating units (RHUs). The radioisotopes discussed include 241Am, 208Po, 210Po, and 90Sr. The aim of this study is to facilitate the design of an RTG with a minimal radiation dose rate and mass including any required shielding. Applications of interest are primarily space and planetary exploration. In order to evaluate the properties of the alternative radioisotopes a Monte Carlo model was developed to examine the radiation protection aspect of the study. The thermodynamics of the power generation process is examined and possible materials for the housing and encapsulation of the radioisotopes are proposed. In this study we also present a historical review of radioisotope thermoelectric generators (RTGs) and the thermoelectric conversion mechanism in order to provide a direct comparison with the performance of our proposed alternative isotope systems.

Introduction

Radioisotope heating units (RHUs) and radioisotope thermoelectric generators (RTGs) have been successfully employed on a number of space missions and extensively used in terrestrial applications. Russian built ‘Beta-M’ RTGs fuelled with Strontium-90 were deployed in unmanned lighthouses, coastal beacons and remote weather and environment monitoring stations and had a typical power output of around 230 Watts electrical (We) [1]. The original network of automatic weather stations (AWSs) in Antarctica was powered by RTGs; however, by the early 1990s safety concerns led to the removal of these power sources [2]. Today, RTG devices have become safer in design due to advances in material fabrication techniques.

Efficient power production has always been one of the challenges of the exploration of space and the solar system. The challenge is even greater at increasing distance from the Sun or beneath planetary surfaces, where solar light intensity levels and extreme temperatures could preclude the use of solar power and chemical power generation systems. Additional power burdens on the budgets available for certain missions may be imposed by systems that have specific operating temperatures and temperature control requirements. Radioisotope power sources are capable of providing both thermal control and electrical power. Reducing the mass and maximising the efficiency of radioisotope power sources will have a positive impact on the overall mass budgets of science payloads and the range of mission scenarios for a specific project.

To date, 238Pu has been the most commonly used RTG and RHU isotope for space applications. Plutonium-238 decays primarily by alpha emission, where the energy of the alpha particle is ∼5 MeV. Plutonium-238 can also decay by spontaneous fission with a very low probability [3]. The absorption of the alpha particles (and any fission products) will produce heat that can be exploited to generate electricity by means of a power conversion system, in the form of thermoelectric junctions or other power conversion cycles (see Section 1.3). Current concerns over the limited supply and the cost of producing 238Pu [4] has increased the need to explore alternative isotopes for these applications. The radioisotopes that are presented as a possible solution by this paper are 241Am, 208Po and 210Po. The use of 90Sr is also discussed briefly.

Under the systems nuclear auxiliary power (SNAP) program, a range of RTGs and small reactors were developed in the United States for space and military use. The RTGs developed were successfully flown on several satellite and space exploration missions [5]. The efficiencies of some of the early devices were between 4% and 5% [6]. This was due to limitations in the power conversion process and losses within the system. Latter devices such as the multi hundred watt (MHW) power sources flown on the voyager missions and the general purpose heat source RTGs (GPHS-RTG) as flown on board Galileo and Cassini missions have efficiencies of the order of 6.6% [6].

SNAP-19 devices were successfully used on the Nimbus meteorological satellites [reference], Pioneer 10 and 11 [5] and were also included in the Viking 1 and 2 mars lander missions. The SNAP-19 devices were loaded with 238PuO2–Mo fuel cermets with a total activity ranging from 34.4 to 80 kCi [6] depending on the mission and application.

The SNAP-27 devices that were landed on the moon during the Apollo 12, 14, 15, 16 and 17 missions were designed to provided a beginning of life (BOL) output power of 63.5 We (electrical power in Watts) and a specific power density of 3.2 We /kg [6]. The total activity of the SNAP-27 fuel source was 44.5 kCi and consisted of microspheres of 238PuO2 ceramic [6]. These devices had a power conversion system with a 5% efficiency based on PbSnTe junctions. The output voltage of the devices was between 14 and 16 V D.C. SNAP-27 devices were used as power sources for lunar surface science experiments with life spans ranging from 4 to 8 years (Fig. 1).

The failure of the Apollo 13 mission enabled the testing of the integrity of the SNAP-27 design during the re-entry of the lunar module into the Earth’s atmosphere. The lunar module splashed down in the Tonga Trench, which is in the Pacific Ocean. No measurements of contamination have been made to date and it is assumed that the SNAP-27 fuel capsule is intact at a depth of 6.5 km below the surface of the ocean.

Even though the integrity of the SNAP design during re-entry is not in question, the additional mass of the aeroshell and cask used by this system reduced the overall system power density which could be improved today by the adoption of alternative encapsulation techniques.

Following the SNAP programmes, the US developed the multi-hundred watt (MHW) RTG systems to provide power to the Lincoln experimental satellites LES 8 and 9 and Voyager 1 and 2 spacecraft. These devices were fuelled by 24 pressed spheres of 238PuO2, each with an activity of 3.2 kCi. Each sphere was encapsulated within a cladding of iridium alloy and housed individually within a filament wound carbon–carbon impact shell. All 24 assembled spheres were housed within a cylindrical POCO graphite aeroshell for re-entry protection [6]. The thermoelectric conversion was performed using SiGe junctions [8] with an efficiency of 6.7% [6]. The MHW RTGs had a beginning of mission power level of 2.4 kWth [8] or 160.8 We. The overall beginning of mission system power density was 4.2 W/kg [8].

The general-purpose heat source radioisotope thermoelectric generator (GPHS-RTG) is a power source that features an integrated modular heat source design [9] (see Fig. 2). Originally designed for the Galileo spacecraft, the GPHS-RTG was successfully used for the NASA Cassini mission and more recently, the New Horizons Kuiper belt mission. The GPHS-RTG was originally built by the US Department of Energy (DOE) at the Mound Laboratory in Miamisburg, Ohio. As a result of increased security requirements and costs the DOE closed the Mound site. The responsibility for assembling RTGs was transferred to a new space battery facility at the Idaho National Laboratory Materials and Fuels Complex (MFC). This was done prior to the 2006 New Horizons mission.

A general purpose heat source (GPHS) module is a composite carbon body that houses a total of four fuel pellets and as a whole acts as an aero-impact shell. The isotope fuel for the GPHS-RTG is in the form of plutonium dioxide (238PuO2) at approximately 80% density. The fuel is pressed into pellets with an approximate length and diameter of 27.6 mm [9]. Each pellet has approximately 0.55 mm of iridium alloy (DOP-26) cladding [9] that is used to maintain the structural integrity of the pellet both under normal operating conditions and under impact. The iridium cladding also prevents the interaction of the source alpha particles with materials with low atomic masses, which could produce neutrons via α-n reactions. The internal structure of the GPHS module consists of two composite impact shells covered by a carbon-bonded carbon sleeve. Each internal impact shell contains two fuel pellets separated by a floating membrane [9]. The thermal power output of a single fuel pellet is approximately 62.5 Wth [9]. GPHS modules can be stacked together and thermally coupled. The required electrical output power levels are achieved through the appropriate selection of a number of GPHS modules incorporated in a RTG system. The number of modules required is directly related to the power conversion efficiency of the system. The current GPHS-RTG systems have a power conversion efficiency of 6.5% to 7% and an overall system power density of 5.2 We kg−1 [6] if a stack of 18 GPHS modules is used [9]. The thermoelectric junctions used for power conversion by the GPHS-RTG are SiGe type junctions.

The 238Pu fuel that is used within the GPHS-RTG systems is currently supplied to the US Department of Energy (DOE) by Russia in the form of 238PuO2 powder with an activity of between 12.6 and 15.1 Ci g−1. The radioisotope is manufactured by the proton irradiation of 237Np or via the neutron irradiation of 237Np in a high flux reactor. The latter is the more commonly used method and results in the production of 238Np (half-life of 2.117 days), which decays via beta emission into 238Pu. Current concerns over the supply of 238Pu [4] have prompted the DOE to investigate the feasibility of establishing a production line in the US in order to meet the future needs of NASA [10].

RTG systems employ thermoelectric power generators, which produce an electric potential by exploiting the Seebeck effect. The Seebeck effect is observed when a temperature gradient exists across the junction of two different metals or semiconductors [10]. The magnitude of the thermoelectric electro-motive force (EMF) across each junction is typically of the order of μV to several mV and is dependant upon junction material selection. An estimate for each junction potential with hot and cold side temperatures (TH and Tc respectively) can be made using Eq. (1). The Seebeck coefficient αpn can be defined as the thermoelectric voltage generated across a junction of thermoelectric materials p and n subjected to a junction temperature difference (TH  TC) and is usually stated in units of μV °C−1. For a junction of SiGe, this has been measured to be on average 135.4 μV °C−1 [11].Vjunction=TCTHαpn·dT.Eq. (1) Seebeck voltage for a thermoelectric junction p–n with Seebeck coefficient αp–n.

Multiple junctions are electrically connected in series to provide the required power. Thermoelectric generator module junctions are typically assembled such that the cold side of the semiconductor element is soldered to a metal cap and the hot side connection is made by compressively loading the element against the hot junction connecting-shoe. Multiple structures can be assembled in the same way [12]. This method of assembly leads to the requirement for the modules to be installed within the RTG system under mechanical compression [12].

RTG systems can exploit conduction, convection or radiative processes to transfer heat to the conversion units. Certain systems may utilise a combination of these methods. One of the main advantages of using radiative transfer in place of conduction or convection is the uniformity of the temperature of the hot junction [12]. The fundamental disadvantage of heat transfer via thermal radiation is the high temperature requirements placed on the heat source. High operating temperature requirements have a direct influence on the materials used within the system and hence the overall system mass.

Alternative power conversion technologies are currently being studied, including: thermionic [13], thermophotovoltaic [14], Brayton and Stirling power conversion systems [15]. The use of Stirling electric generators has recently been of great interest to several research institutes including the NASA Glenn Research Center, Cleveland, Ohio [16]. The main interest behind this research lies with the desire to improve on the power conversion efficiency of current thermoelectric based radioisotope power sources [17], which today have efficiencies of the order of 6.5–7% [6]. The power conversion efficiencies of modern Stirling generators is placed between 22% and 32% [16].

Despite their greater conversion efficiencies, mechanical systems such as Stirling generators have a much greater mass than thermoelectric systems. It is suggested that the greater efficiency allows for a reduction in fuel mass when compared to current GPHS-RTG systems [18]. However, this fuel saving is unlikely to translate into a net system mass saving. Concerns over the operational lifetimes of mechanical systems operating in a high radiation environment and the need to the balance the mass associated with such a unit is addressed by having a dual-device deployment strategy. This strategy has called into question the economics of mechanical power conversion. For these reasons, more research into the overall benefits of using these systems in place of the thermoelectric based systems is required. Research into alternative solid-state power converters and hybrids of these technologies will prove to be vital to the improvement of conversion efficiency and the minimisation of system mass.

Section snippets

Reduced neutron and gamma ray radiation emissions

The careful selection and design of the radioisotope fuel source will make it possible to derive the most mass efficient shielding configuration. The selection of an isotope for a specific application is dependent on several requirements. These requirements include; mission duration, electrical power levels, whether the device is to be used for heating in addition to the generation of electricity, proximity to sensitive instrumentation and biological systems, and the overall mission mass budget.

Radioisotope fuel cermets and encapsulation

Fuel cermets have been examined as a means of encasing fissile fuels for use in reactor systems such as in nuclear thermal rocket propulsion [34]. Cermet (ceramic–metallic) encapsulation is the formation of a material matrix composed of an intimate mixture of metallic carrier material and a ceramic compound such as a radioisotope in its oxide form. A cermet matrix may alternatively be formed where the ceramic materials are used as a carrier material for a metallic material. The matrix density

System radiation environment modelling and shielding criteria

Our initial study investigated the theoretical substitution of the GPHS 238Pu fuel pellets with the candidate isotopes listed in Section 5. The study explored the effect of isotope selection on the system mass, thermal power output and radiation dose. The candidate isotopes studied were 241Am in a 241AmO2 ceramic pellet form, 241AmO2 in tungsten matrices with varying tungsten volume fractions and 90Sr in a 90SrO ceramic form. The study also investigated the effects upon radiation dose by the

A laboratory breadboard model for low mass, reduced radiation RTGs

The overall RTG system power density is crucial for space and planetary science applications. A greater electrical power density ultimately translates into greater instrumentation or propellant mass for any given mission. For this reason, the systems and shielding configurations outlined in this study are designed to both minimise the radiation flux and to maximise the electrical power density.

Having selected 241Am in the form of americium dioxide as the fuel for a long-lived laboratory scale

Conclusions and future work

Through the comparison of isotopes made in Section 2, we have identified three isotopes that can deliver suitable power densities for space missions while reducing the radiation doses delivered by such sources. The factor that drives selection from these isotopes is the mission architecture. 208Po and 210Po are best suited to missions requiring short durations of thermal and/or electrical power before decaying into stable isotopes of lead. Such missions include melt-penetration of icy planetary

Acknowledgements

Engineering and Physical Sciences Research Council (EPSRC) for funding the project (EP/D030277/1). Mike Evans and Tim Stevenson – University of Leicester. Darryl Butt – Boise State University.

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