A robust optimization approach for design of a General Purpose Heat Source

Design optimization and integration of nickel/Haynes-25 AMTEC cells into radioisotope power systems

A shielded storage rack has been installed as part of the Radioisotope Power Systems Facility (RPSF) at the U.S. Department of Energy’s (DOE) Hanford Site in Washington State. The RPSF is designed to replace an existing facility at DOE’s Mound Site near Dayton, Ohio, where General Purpose Heat Source (GPHS) modules are currently assembled and installed into Radioisotope Thermoelectric Generators (RTG). The overall design goal of the RPSF is to increase annual production throughput, while at the same time reducing annual radiation exposure to personnel. The shield rack design successfully achieved this goal for the Module Reduction and Monitoring Facility (MRMF), which processes and stores assembled GPHS modules, prior to their installation into RTGs. The shield rack design is simple and effective, with the result that background radiation levels within Hanford’s MRMF room are calculated at just over three percent of those typically experienced during operation of the existing MRMF at Mound, despite the fact that Hanford’s calculations assume five times the GPHS inventory of that assumed for Mound.

A series of previous OSC papers described: 1) a novel methodology for the coupled thermal, fluid flow, and electrical analysis of multitube AMTEC (Alkali Metal Thermal-to-Electric Conversion) cells; 2) the application of that methodology to determine the effect of numerous design variations on the cell’s performance, leading to selection and performance characterization of an OSC-recommended cell design; and 3) the design, analysis, and characterization of an OSC-generated power system design combining sixteen of the above AMTEC cells with two or three GPHS (General Purpose Heat Source) radioisotope heat source modules, and the applicability of those power systems to future space missions (e.g. Pluto Express and Europa Orbiter) under consideration by NASA. The OSC system design studies demonstrated the critical importance of the thermal insulation subsystem, and culminated in a design in which the eight AMTEC cells on each end of the heat source stack are embedded in Min-K fibrous insulation, and the Min-K and the GPHS modules are surrounded by graded-length Mo multifoil insulation. The present paper depicts the OSC-recommended AMTEC cell and generator designs, and identifies the need for an electrically heated (scaled-down but otherwise prototypic) test assembly for the experimental validation of the generator’s system performance predictions. It then describes the design of an OSC-recommended test assembly consisting of an electrical heater enclosed in a graphite box to simulate the radioisotope heat source, four series-connected prototypic AMTEC cells of the OSC-recommended configuration, and a prototypic hybrid insulation package consisting of Min-K and graded-length Mo multifoils. Finally, the paper describes and illustrates an OSC-recommended detailed fabrication sequence and procedure for the above cell and test assembly. That fabrication procedure is being implemented by AMPS, Inc. with the support of DOE’s Oak Ridge and Mound Laboratories, and the Air Force Phillips Laboratory (AFPL) will test the performance of the assembly over a range of input thermal powers and output voltages. The experimentally measured performance will be compared with the results of OSC analyses of the same insulated test assembly over the same range of operating parameters.

The General-Purpose Heat Source (GPHS) is a modular heat source that will supply energy for Radioisotope Thermoelectric Generators (RTGs) in space missions. The Safety Verification Tests (SVTs) are performed to assess the plutonia containment capability of heat source modules subjected to certain accident environments. This interim report described the GPHS module configuration, the test environment, and the response of the module components following simulated reentry and solid Earth impact. The specific test environment of these initial six tests results from failure of the booster rocket to place the spacecraft in a proper trajectory and subsequent reentry of the GPHS modules from Earth orbit. 36 figs.

The purpose of the project was to: Develop a finite element thermal model of the General Purpose Heat Source (GPHS) Module loaded with four fueled clads; Perform steady state analysis to validate the thermal model via energy balance; Perform transient thermal analysis to predict clad temperatures at various times; and, Transient model results will be used to inform module soak time needed to reach a desired clad temperature prior to impact.

Since the launch of the first radioisotope power system (RPS) on Transit 4A in 1961, numerous research and development activities have been performed centered on increasing the overall thermal to electrical conversion efficiency of the selected nuclear fueled power system. The latest U.S. fielded RPS (MMRTG - Multi-Mission Radioisotope Thermoelectric Generator) contains eight GPHS (General Purpose Heat Source) modules which nominally yield ∼2000W Th from the thirty-two 238PU5O 2 ceramic fuel pellets. MMRTG's PbTe/TAGS-85 based thermoelectric couples have a thermal to electrical conversion efficiency of ∼5.5% thus yielding ∼110W e at launch. This paper centers on the discussion of a conceptual idea that entails employing a second set of thermoelectrics on an MMRTG. These would be employed for possibly converting a portion of the excess “waste heat” (∼1750 W th ) into additional electrical mission power. First-order experiments and calculations employing bismuth telluride (Bi 2 Te 3 ) based thermoelectric modules being considered for a European RPS by the University of Leicester indicate that an improvement in the efficiency of an MMRTG could be achieved by integrating them with the PbTe/TAGS-85 thermoelectrics being utilized in the MMRTG in a “dual” or “cascaded” arrangement. This arrangement of two different integral thermoelectric materials suggests the intriguing possibility of the additional harvesting of a portion of the MMRTG's currently unutilized “waste heat”. This appears to be feasible since the cold side temperature of the PbTe/TAGS-85 in an MMRTG is ∼200°C which corresponds to typical hot-side operating temperatures of the Bi 2 Te 3 thermoelectric modules. It is recognized that extensive thermo-mechanical-electrical design, modeling, and analysis are required to fully investigate the cascaded thermoelectric concept. In addition, materials compatibility and assembly aspects will need to be fully addressed in the future. The present work indicates that system-level performance gains could be achieved via a “cascaded” or cMMRTG which could result in electrical power increases at BOM (Beginning of Mission) of up to ∼25%, and perhaps more significantly, gains in EODL (End of Design Life) power of up to ∼40% which could be utilized on a future space mission.

The United States Department of Energy is planning to develop the Advanced Stirling Radioisotope Generator (ASRG) for the National Aeronautics and Space Administration (NASA) for potential use on future space missions. The ASRG provides substantial efficiency and specific power improvements over radioisotope power systems of heritage designs. The ASRG would use General Purpose Heat Source modules as energy sources and the free-piston Advanced Stirling Convertor (ASC) to convert heat into electrical energy. Lockheed Martin Corporation of Valley Forge, Pennsylvania, is integrating the ASRG systems, and Sunpower, Inc., of Athens, Ohio, is designing and building the ASC. NASA Glenn Research Center of Cleveland, Ohio, manages the Sunpower contract and provides technology development in several areas for the ASC. One area is reliability assessment for the ASC heater head, a critical pressure vessel within which heat is converted into mechanical oscillation of a displacer piston. For high system efficiency, the ASC heater head operates at very high temperature (850 C) and therefore is fabricated from an advanced heat-resistant nickel-based superalloy Microcast MarM-247. Since use of MarM-247 in a thin-walled pressure vessel is atypical, much effort is required to assure that the system will operate reliably for its design life of 17 years. One life-limiting structural response for this application is creep; creep deformation is the accumulation of time-dependent inelastic strain under sustained loading over time. If allowed to progress, the deformation eventually results in creep rupture. Since creep material properties are not available in the open literature, a detailed creep life assessment of the ASC heater head effort is underway. This paper presents an overview of that creep life assessment approach, including the reliability-based creep criteria developed from coupon testing, and the associated heater head deterministic and probabilistic analyses. The approach also includes direct benchmark experimental creep assessment. This element provides high-fidelity creep testing of prototypical heater head test articles to investigate the relevant material issues and multiaxial stress state. Benchmark testing provides required data to evaluate the complex life assessment methodology and to validate that analysis. Results from current benchmark heater head tests and newly developed experimental methods are presented. In the concluding remarks, the test results are shown to compare favorably with the creep strain predictions and are the first experimental evidence for a robust ASC heater head creep life.

There is currently a renewed interest in the use of free-piston Stirling power convertors for space power applications. More specifically, the Stirling convertor is being developed to be part of the Stirling Radioisotope Power System to supply electric power to spacecraft for NASA deep space science missions. The current development effort involves the Department of Energy, Germantown, MD, the NASA Glenn Research Center, Cleveland, OH, and the Stirling Technology Company, Kennewick, WA. The Stirling convertor will absorb heat supplied from the decay of plutonium dioxide contained in the General Purpose Heat Source modules and convert it into electricity to power the spacecraft. For many years the potentials of the free-piston Stirling convertor have been publicized by its developers. Among these potentials were long life and high reliability. This paper presents an overview of the critical areas that enable long life of the free-piston Stirling power convertor, and present some of the techniques that have been used when long life has been achieved.

Impact testing of general purpose heat sources (GPHSs) and their component GPHS clads is done to benchmark extensive safety calculations quantifying launch safety. Impact testing is done in the Isotope Fuels Impact Tester (IFIT), a large-bore gas gun at Los Alamos National Laboratory. Efforts to conduct an impact test at the extreme low end of the temperature range for launch have highlighted uncertainties in determining the GPHS clad temperature during impact tests. In IFIT impact tests, the GPHS clad temperature is inferred from the temperature of the radiological confinement. Heating tests have been done in the IFIT to determine the fueled clad surface temperature as a function of the surface temperature of the tantalum radiological confinement can. Direct measurement of clad temperatures in the impact configuration are described and the effect of emissivity of the various components indicated. The analytical model used to predict clad temperatures is seen to work well at temperatures above 625°C. Appropriate values of emissivity for use in the model were measured in the experiment. Calculation of the experimental clad impact temperature using the ANSYS thermal transport model is necessary at clad temperatures below 625°C. ANSYS modeling indicates that the clad temperature in a recent low-temperature impact was outside the relevant range for launch safety modeling of GPHS clad behavior.

This paper presents a detailed description of Orbital Science Corporation's (OSC) design of a new support system for the General Purpose Heat Source (GPHS) using the aeroshell's existing indentations for interlocking the heat source modules. The housing and support system were recently fabricated and tested to the Jet Propulsion Laboratory (JPL) draft Advanced Radioisotope Power System (ARPS) test requirements at the the Department of Energy's (DoE) Mound facility in Miamisburg, Ohio, USA. The paper presents the test environments and results. This paper also illustrates some of the latest ARPS system design concepts using this specific support system and housing for future deep space missions.

The General-Purpose Heat Source (GPHS) provides power for space missions by transmitting the heat of /sup 238/PuO/sub 2/ decay to thermoelectric elements. Because of the inevitable return of certain missions, the heat source must be Designed and constructed to survive both re-entry and Earth impact. The Design Iteration Test (DIT) series is part of an ongoing test program. In the first Design Iteration Test (DIT-1), a full GPHS module ontaining four iridium-alloy capsules loaded with /sup 238/PuO/sub 2/ was impacted at 57 m/s and 930/sup 0/C. All four capsules survived and none was breached. The capsules used in DIT-1 were loaded and welded at Los Alamos. The second Design Iteration Test (DIT-2) also used a full GPHS module and was impacted at 58 m/s and 930/sup 0/C. The four iridium-alloy capsules used in this test were loaded and welded at the Savannah River Plant (SRP). Postimpact examination revealed that two capsules had survived and two capsules had breached; a small quantity (approx. = 50 ..mu..g) of /sup 238/PuO/sub 2/ was released from the breached capsules. Internal cracking similar to that observed in the DIT-1 capsules was evident in all four of the DIT-2 capsules. Postimpact analyses of the units are described with emphasis on weld structure and performance.

In a Stirling Radioisotope Power System (RPS), heat must be continuously removed from the General Purpose Heat Source (GPHS) modules to maintain the modules and surrounding insulation at acceptable temperatures. The Stirling convertor normally provides this cooling. If the Stirling convertor stops in the current system, the insulation is designed to spoil, preventing damage to the GPHS at the cost of an early termination of the mission. An alkali-metal Variable Conductance Heat Pipe (VCHP) can be used to passively allow multiple stops and restarts of the Stirling convertor. In a previous NASA SBIR Program, Advanced Cooling Technologies, Inc. (ACT) developed a series of sodium VCHPs as backup cooling systems for Stirling RPS. The operation of these VCHPs was demonstrated using Stirling heater head simulators and GPHS simulators. In the most recent effort, a sodium VCHP with a stainless steel envelope was designed, fabricated and tested at NASA Glenn Research Center (GRC) with a Stirling convertor for two concepts; one for the Advanced Stirling Radioisotope Generator (ASRG) back up cooling system and one for the Long-lived Venus Lander thermal management system. The VCHP is designed to activate and remove heat from the stopped convertor at a 19 degC temperature increase from the nominal vapor temperature. The 19 degC temperature increase from nominal is low enough to avoid risking standard ASRG operation and spoiling of the Multi-Layer Insulation (MLI). In addition, the same backup cooling system can be applied to the Stirling convertor used for the refrigeration system of the Long-lived Venus Lander. The VCHP will allow the refrigeration system to: 1) rest during transit at a lower temperature than nominal; 2) pre-cool the modules to an even lower temperature before the entry in Venus atmosphere; 3) work at nominal temperature on Venus surface; 4) briefly stop multiple times on the Venus surface to allow scientific measurements. This paper presents the experimental results from integrating the VCHP with an operating Stirling convertor and describes the methodology used to achieve their successful combined operation.

The Brazilian Nuclear Research Institute (IPEN) proposed a design for the disposal of Disused Sealed Radioactive Sources (DSRS) based on the IAEA Borehole Disposal of Sealed Radioactive Sources (BOSS) design that would allow the entirety of Brazil’s inventory of DSRS to be disposed in a single borehole. The proposed IPEN design allows for 170 waste packages (WPs) containing DSRS (such as Co-60 and Cs-137) to be stacked on top of each other inside the borehole. The primary objective of this work was to evaluate the thermal performance of a conservative approach to the IPEN proposal with the equivalent of two WPs and two different inside configurations using Co-60 as the radioactive heat source. The current WP configuration (heterogeneous) for the IPEN proposal has 60% of the WP volume being occupied by a nuclear radioactive heat source and the remaining 40% as vacant space. The second configuration (homogeneous) considered for this project was a homogeneous case where 100% of the WP volume was occupied by a nuclear radioactive heat source. The computational models for the thermal analyses of the WP configurations with the Co-60 heat source considered three different cooling mechanisms (conduction, radiation, and convection) and the effect of mesh size on the results from the thermal analysis. The results of the analyses yielded maximum temperatures inside the WPs for both of the WP configurations and various mesh sizes. The heterogeneous WP considered the cooling mechanisms of conduction, convection, and radiation. The temperature results from the heterogeneous WP analysis suggest that the model is cooled predominantly by conduction with effect of radiation and natural convection on cooling being negligible. From the thermal analysis comparing the two WP configurations, the results suggest that either WP configuration could be used for the design. The mesh sensitivity results verify the meshes used, and results obtained from the thermal analyses were close to being independent of mesh size. The results from the computational case and analytically-calculated case for the homogeneous WP in benchmarking were almost identical, which indicates that the computational approach used here was successfully verified by the analytical solution.

In a Stirling Radioisotope Power System (RPS), heat must be continuously removed from the General Purpose Heat Source (GPHS) modules to maintain the modules and surrounding insulation at acceptable temperatures. The Stirling convertor normally provides this cooling. If the Stirling convertor stops in the current system, the insulation is designed to spoil, preventing damage to the GPHS at the cost of an early termination of the mission. An alkali-metal variable conductance heat pipe (VCHP) can be used to passively allow multiple stops and restarts of the Stirling convertor by bypassing the heat during stops. In a previous NASA Small Business Innovation Research (SBIR) Program, Advanced Cooling Technologies, Inc. (ACT) developed a series of sodium VCHPs as backup cooling systems for the Stirling RPS. In 2012, one of these VCHPs was successfully tested at NASA Glenn Research Center with a Stirling convertor as an Advanced Stirling Radioisotope Generator (ASRG) backup cooling system. The prototype; however, was not optimized and did not reflect the final heat rejection path. ACT through further funding has developed a semioptimized prototype with the finalized heat path for testing at Glenn with a Stirling convertor. The semioptimized system features a two-phase radiator and is significantly smaller and lighter than the prior prototype to reflect a higher level of flight readiness. The VCHP is designed to activate and remove heat from the GPHS during stoppage with a small temperature increase from the nominal vapor temperature. This small temperature increase from nominal is low enough to avoid risking standard ASRG operation and spoiling of the multilayer insulation (MLI). The VCHP passively allows the Stirling convertor to be turned off multiple times during a mission with potentially unlimited off durations. Having the ability to turn the Stirling off allows for the Stirling to be reset and reduces vibrations on the platform during sensitive measurements or procedures. This paper presents the design of the VCHP and its test results with a Stirling convertor at Glenn. Tests were carried for multiple on and off cycles to demonstrate repeatability. The impacts associated with the addition of the VCHP to the system are also addressed in terms of mass and additional heat losses due to the presence of the VCHP.

For the past 25 years, unicouples made of Silicon Germanium alloys (Si0.8Ge0.2 and Si0.63Ge0.38) have been used in the majority of the spacecraft powered by Radioisotope Thermoelectric Generators (RTGs) to explore various planets of the solar system as well as the sun. These unicouples operate typically at hot and cold side temperatures of 1175–1273 K and 573 K, respectively, converting the heat produced in the General Purpose Heat Source (GPHS) modules to electricity at efficiencies <7%. Recently, Segmented Thermoelectric Unicouples (STUs) made of n-type Bi2Te3 and CoSb3-based alloys and p-type Bi2Te3 and CeFe4Sb12-based alloys have been developed and tested successfully with a demonstrated efficiency to date of ∼10% and the potential for achieving 14–15% when operated at hot and cold side temperatures of 973 K and 300 K, respectively. This paper compares the performance of SiGe unicouples and skutterudite STUs, at same hot side temperature of 973 K and cold side temperatures of 300, 573, and 673 K, and the same total length and cross sectional dimensions of the p-leg. The area of the n-leg and the lengths of the segments of various materials in the STUs are determined based on maximizing either the electrical power density or the conversion efficiency. Results showed that replacing SiGe with skutterudite STUs in future radioisotope power systems (or even in conjunction with a space nuclear reactor and space probes), could increase the thermal to electric conversion efficiency at a heat source temperature of 973 K by ∼40% to 198% for Tc=673 K and 300 K, respectively. This would reduce by more than half the mass of the 238PuO2 fuel needed and the radiator’s surface area for rejecting the waste heat.