Idaho National Laboratory’s (INL) Role in Radioisotope Power Systems (RPS)
Overview of INL's support of Radioisotope Power Systems. Given to Mount Science and Energy Museum Association
07 ISOTOPE AND RADIATION SOURCES↗
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Results for “Radioisotope Power Systems”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
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Radioisotope Power System Cask Trailer Chiller Low Heat Load Study
This report documents the study of the Radioisotope Thermoelectric Generator Transport System (RTGTS) chillers capability to transport a Radioisotope Power System (RPS) with low heat loads in the cask. Tests of the RTGTS chillers included setting the glycol/water (coolant) temperature to a control set point of 40° and 60°F while applying heat loads ranging from 2500 to 0 W heat load. The tests performed showed that the chiller systems could handle the entire range of heat loads. However, a noise/vibration started to occur when the set point was 60°F with low heat loads. Three recommended changes to the RTGTS chillers are recommended with varying implementation windows. The first is an operational change that can be implemented immediately. The RTGTS chillers should have a set point of 40°F when transporting an RPS less than 1500 W. The second is a short-term change (within the next two years) augmenting RPS heat by utilizing one RTGTS onboard heater (2500 W) while transporting an RPS with less than 1500 W which will also require a change to the Instrument and Data Acquisition System (IDAS) to allow monitoring and possible control of power to the onboard heaters. However, if the heater does fail during transport, no emergency actions or immediate repairs will be required due to the chillers’ ability to handle low heat loads. Running a 2500 W heater when transporting low heat loads aids in prolonging the life of the chillers. The remaining recommended long-term change is to replace the current chillers with a new portable or permanently installed chiller system in the next five to ten years. This time frame takes advantage of the proven reliability and durability of the current chillers while the new chillers are phased in and proven to the RPS program.
42 ENGINEERING↗
A Comparison of the Radiation Shielding Requirements of Select Radioisotopes for Potential Use in Radioisotope Power Systems
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07 ISOTOPE AND RADIATION SOURCES↗
The Radioisotope Power System Dose Estimation Tool (RPS-DET) User Manual, Version 2.0
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Smith, Michael [ORNL]↗
NASA Dragonfly Mission MMRTG
The Radioisotope Power Systems (RPS) group at the Idaho National Laboratory (INL) is supporting NASA’s New Frontier class mission, Dragonfly, to explore the Saturn moon Titan. Titan has been observed by both Voyager spacecrafts in 1979 and 1980 and during over 100 close flyby encounters by the Cassini orbiter and subsequent landing of the Huygens probe which landed on Titan in 2005. The probe lasted about 2.5 hours without an RPS due to the thick atmosphere and cold climate but gave scientists valuable insight into the geology of Titan. The surface is composed of vast sand dunes and liquid methane with an atmosphere composed primarily of nitrogen that is 60% more dense as the Earth’s with 1/7th of Earth’s gravity. These conditions allow for exploration using a dual-quadcopter. Due to the extreme temperatures (-290°F or -180°C), changes to the Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) are required. This generator, designated flight unit 4 or MMRTG-F4, is being built by a team lead by INL and includes the prime contractor Aerojet Rocketdyne and their supplier Teledyne Energy System. This poster outlines the mission, the unique challenges and updates the build progress.
42 ENGINEERING↗
Revival and Technology Transfer of 90SrTiO3 Production for Heat Source Applications
Demand for radioisotope power systems (RPS) appears to be on the rise, and it is unlikely that this new demand can be met by plutonium-238. As a result, new heat source designs are under development. One such design is Z1. Z1 is a strontium-90 demonstration heat source developed by Zeno Power Systems in collaboration with the University of Dayton Research Institute and the Pacific Northwest National Laboratory. SrTiO3 was chosen as the fuel form for Z1. SrTiO3 was used to produce dozens of terrestrial RPS decades ago. While this indicates the technology is proven, the skill and experience necessary to effectively produce SrTiO3 has been lost. Recapturing SrTiO3 production technology therefore became necessary. Development started with non-radioactive surrogate experiments using natural strontium. This work showed that the old Oak Ridge flowsheets for SrTiO3 production are quite robust, but some critical pieces of the technology are left out of the old documentation. Specifically, particle sizes of the reagents are critical to producing a phase pure SrTiO3. If the reagent particles are not the right size, then a significant Sr3Ti2O7 phase impurity is observed. This phase impurity was proven to be the result of an incomplete reaction between the strontium and titanium precursors. This technology was transferred to Pacific Northwest National Laboratory where the technology was adapted for use in a hot cell. With the adaptations complete, this technology was used to produce the fuel for Z1 late last year. This represents the first new heat source design produced in the U.S. in over 40 years.
90Sr, strontium titanate, SrTi03, RTG, Radioisotop↗
Metallic Coating of Cerium Oxide Microspheres
The ability to remove heat is paramount to nuclear fuel performance and longevity. Retaining fission product and separating fuel from reactor coolant and the environment is also necessary to prevent radiological contamination. Conventional nuclear fuel for commercial light water reactors and radioisotope power systems (RPS) is composed of oxide powders pressed into a pellet (cm-scale) and then sealed into a metal cladding to confine the fuel. What typical fuels lack is a method to surround each particle of nuclear fuel in metal, thus providing a more intimate protection layer for accident tolerance and boosting the thermal extraction from the fuel element. In such a way, metal-coated fuel particles increase heat extraction efficiency over clad-pellet designs while increasing the accident tolerance of the fuel. Metal oxide microspheres have wide-ranging applications, including the realm of fuels for nuclear reactors and RPS. Microspheres of uranium oxide/uranium carbide, mixed uranium/plutonium oxides, transuranics, and thorium fuels have been extensively studied. Pacific Northwest National Laboratory has also demonstrated the production of 238 PuO 2 microspheres for RPS applications. Metal-coated oxide microsphere fuels may also be attractive for other applications such as nuclear thermal rockets, future nuclear reactor designs, and catalysts.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Neptunium mononitride as a target material for Pu-238 production
Deep space exploration requires specialized sources for both thermal and power applications. Radioactive decay heat of plutonium-238 (238Pu) provides these sources in the form of radioisotope thermoelectric generators (RTGs). The 238 Pu is produced via neutron capture reaction involving neptunium-237 ( 237 Np) target material. Continual optimization of 237 Np target materials and evaluation of potential alternative targets for production of 238 Pu RTGs are advantageous for meeting ongoing space power system resource requirements. Current production of 238 Pu for RTGs for the United States space program utilizes neptunium dioxide ( 237 NpO 2 ) targets; however, the use of neptunium mononitride ( 237 NpN) presents an opportunity to increase the mass of 237 Np per target compared to the dioxide form, as well as increase the thermal conductivity of the target. To assess the viability of a 237 NpN target material, the material chemistry must be thoroughly evaluated, including synthesis methods and dissolution and reprocessing schemes. This review presents a summary of synthesis pathways for 237 NpN based on published literature on actinide mononitrides. Specific literature on 237 NpN is limited, necessitating evaluation of other actinide systems to gather parallels. This suggests a need for additional experimental studies on 237 NpN. A particular limitation in the existing literature is a lack of information on the differences in material characteristics, such as morphology, particle size, and trace chemical impurities, as a function of synthesis method. These parameters may affect subsequent reactor performance or dissolution of irradiated targets. The evaluation of existing literature is presented with a focus on the efficacy of 237 NpN targets for 238 Pu production.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Radiation Modeling of Z1: Validation of a Novel Radioisotope System
This paper presents the development of a radiation model for a novel strontium-90 (Sr-90) radioisotope heat source developed by Zeno Power Systems (Zeno), which demonstrates a groundbreaking fuel and shielding design that significantly reduces weight compared to traditional concepts. A Monte Carlo N-Particle (MCNP) model has been created to assess the effectiveness of this fuel and shielding design, however validation of the MCNP model is desired. Zeno has developed a prototype device (Z1) to aid in this model validation effort. This project is a collaborative effort between Zeno, the University of Dayton Research Institute (UDRI) and Pacific Northwest National Laboratory (PNNL), where the Z1 prototype was constructed and characterized.
RTG↗
A virtual Frisch-grid geometry-based CZT gamma detector for in-field radioisotope identification
Here, we present a Virtual Frisch-Grid geometry-based CZT gamma detector developed for identifying different radioisotopes over an energy range from a few keV up to 2 MeV, and useful for efficient characterization of CZT crystals. The detector is built with a 3 x 3 matrix of CZT crystals, each measuring approximately 6 mm x 6 mm x 15 mm. The charge generated within the sensor’s active volume is read out via an anode connected directly to the AVG3_Dev integrated circuit. A current signal induced by charge drift is collected on side pads of the crystals, enabling reconstruction of a 3D interaction position. This paper discusses the design, development, and performance of the standalone, mobile detector system, which integrates the AVG3_Dev readout IC developed at Brookhaven National Laboratory, a high-speed FPGA-based with per-channel digital signal processing, and embedded system capabilities. The device is compact, battery-powered, and supports wireless data streaming, making it suitable for field operations for radioisotope identification.
47 OTHER INSTRUMENTATION↗
Alternative Power and Propulsion for Space Systems
What are the main sources of power and propulsion for space systems and what are their advantages, disadvantages, how do they work, and what are the politics behind them? The following topics will be analyzed solar panels, batteries, radioisotope thermoelectric generator (RTG), advanced Stirling radioscopic generator (ASRG), reactors, nuclear thermal rockets (NTP), thermionic experiment with conversion (TOPAZ), electrothermal plasma thruster (VASIMR), ion propulsion, and finally nuclear pulse propulsion. The production of Pu-238 is extremely controversial but a vital source of fuel for many of these space systems. Soral and Batteries are the most common and the least controversial forms of power used in space systems today. However, is ineffective for deep space exploration as they only function for up to four years without needing to be recharged. This is where nuclear systems come into play. Nuclear systems such as an RTG will “live” up to 88 years. For example, Voyager 1 is powered by an RTG It has been traveling since 1977 and has left out solar system in 2012 and has yet to run out of power. However, this type of technology (nuclear) has many hurtles it needs to jump through before advancements in can go much further.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Nuclear Safety [Vol. 37, No. 1, January-March 1996]
Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 A Decision Support System for Maintenance Management of a Boiling-Water Reactor Power Plant, J. H. Shen, A. Ray, and S. Levine; ACCIDENT ANALYSIS: 12 On Prediction of the Ignition Potential of Uranium Metal and Hydride, M. Epstein, W. Luangdilok, M. G. Plys, and H. K. Fauske; 26 An Overview of the Primary Parameters and Methods for Determining Condensation Heat Transfer to Containment Structures, J. Green and K. Almenas; DESIGN FEATURES: 49 Modem Tornado Design of Nuclear and Other Potentially Hazardous Facilities, J. D. Stevenson and Y. Zaho; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 73 1994 Accident Sequence Precursor Program Results, R. J. Belles, J. W. Cletcher, D. A. Coplnger, B. W. Dolan, J. W. Minarick, and P. D. O'Reilly; ANNOUNCEMENTS: 93 American Institute of Chemical Engineers (AlChE) Spring 1997 National Meeting; 94 European Safety and Reliability Association International Conference on Safety and Reliability ESREL ’97; 95 Criticality Safety Challenges in the Next Decade; 96 21st International Symposium on the Scientific Basis for Nuclear Waste Management; 84 The Authors; 88 Letter to the Editor; 90 Indexes to Nuclear Safety, Volume 36.
05 NUCLEAR FUELS↗
Nuclear Safety [Vol. 37, No. 1, January-March 1996]
Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 A Decision Support System for Maintenance Management of a Boiling-Water Reactor Power Plant, J. H. Shen, A. Ray, and S. Levine; ACCIDENT ANALYSIS: 12 On Prediction of the Ignition Potential of Uranium Metal and Hydride, M. Epstein, W. Luangdilok, M. G. Plys, and H. K. Fauske; 26 An Overview of the Primary Parameters and Methods for Determining Condensation Heat Transfer to Containment Structures, J. Green and K. Almenas; DESIGN FEATURES: 49 Modem Tornado Design of Nuclear and Other Potentially Hazardous Facilities, J. D. Stevenson and Y. Zaho; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 73 1994 Accident Sequence Precursor Program Results, R. J. Belles, J. W. Cletcher, D. A. Copinger, B. W. Dolan, J. W. Minarick, and P. D. O'Reilly; ANNOUNCEMENTS: 93 American Institute of Chemical Engineers (AlChE) Spring 1997 National Meeting; 94 European Safety and Reliability Association International Conference on Safety and Reliability ESREL ’97; 95 Criticality Safety Challenges in the Next Decade; 96 21st International Symposium on the Scientific Basis for Nuclear Waste Management; 84 The Authors; 88 Letter to the Editor; 90 Indexes to Nuclear Safety, Volume 36.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
General-Purpose Heat Source Radioisotope Thermoelectric Generator Flight Unit 5 Refurbished (GPHS-RTG F5R) Electrically Heated Thermal Vacuum Testing
Idaho National Laboratory (INL) has been tasked with the evaluation and refurbishment of the General Purpose Heat Source (GPHS) Radioisotope Thermoelectric Generator (RTG) Flight Unit 5 (GPHS-RTG F5R or F5R) that was defueled in 2005. This paper describes the testing of F5R in the thermal vacuum chamber and evaluation of the data demonstrating the generator meets updated Cassini-Huygens requirements. The generator produced 260 We of power at the specified 4100 Wth heat source inventory. This power exceeded the power requirement for 250 We , confirming system integrity and demonstrating that the generator is capable of potential NASA missions.
30 - DIRECT ENERGY CONVERSION↗
GPHS-RTG F5R Electrically Heated Thermal Vacuum Testing
Idaho National Laboratory (INL) has been tasked with the evaluation and refurbishment of the General Purpose Heat Source (GPHS) Radioisotope Thermoelectric Generator (RTG) Flight Unit 5 (GPHS-RTG F5R or F5R) that was defueled in 2005. This paper describes the testing of F5R in the thermal vacuum chamber and evaluation of the data demonstrating the generator meets updated Cassini-Huygens requirements. The generator produced 260 We of power at the specified 4100 Wth heat source inventory. This power exceeded the power requirement for 250 We , confirming system integrity and demonstrating that the generator is capable of potential NASA missions.
30 - DIRECT ENERGY CONVERSION↗
Testing of Helium Cooled Metal Molybdenum Disk Target System
NorthStar Medical Radioisotopes LLC is planning to produce the important medical radioisotope molybdenum-99 (Mo-99) through a photonuclear reaction on molybdenum-100 (Mo-100). In this approach, multiple thin disks of enriched molybdenum metal will be bombarded with a 40-MeV electron beam. Because enriched Mo-100 is expensive, we intend to use as much beam power as possible to achieve maximum production yield and minimize the size of the target. This requirement leads to very high beam power density (heat deposition in the target), which sets challenging requirements for cooling. Together with scientists at Los Alamos National Laboratory, a team at Argonne National Laboratory has developed and demonstrated a cooling approach using pressurized helium, which allows for efficient heat removal. One of the challenges in this approach is the management of the heat load on the target window. The target window separates the high-pressure helium atmosphere inside the target from the vacuum in the beamline, so it is constantly under stress from differential pressure. Also, the window is cooled only by the helium gas flow from one side, making the window design challenging. High heat deposition in the target disks also imposes a strict requirement on performance of the helium cooling system and thickness of the target disks. The target disks are produced from metal powder via a press-and-sinter process. The resulting disks do not possess as high a tensile strength as solid molybdenum and might not survive the vibration from the high-velocity helium coolant and the high thermal stress from beam heating.
42 ENGINEERING↗
An Evaluation and Qualification of U.S.-Based Research Reactors for Irradiation Capabilities Supporting Advanced Nuclear Systems
Irradiation experiments are a prerequisite for evaluating nuclear reactor system designs, analyzing the performance of these systems, and obtaining licenses. Likewise, irradiation facilities are necessary for producing the radioisotopes used in industrial and medical applications. Recent developments in modeling and simulation capabilities and advancements in computational resources have further enabled the design of irradiation experiments for evaluating radiation-induced phenomena and determining nuclear fuel, material, and system design and safety criteria pertaining to both normal and accident scenarios. These computational tools and models require comprehensive experimental datasets acquired under prototypic radiation conditions—for exploring material and system performance under the uniquely harsh environments found in nuclear reactors—to enable verification and validation for qualification and licensing purposes. However, qualification of irradiation experimental facilities, primarily research and test reactors (RTRs), necessitates that their performance be evaluated based on the irradiation environment (e.g. flux, power, testing capabilities) using an appropriate scoring matrix. Although many university campus RTRs are available for research and development (R&D) activities and initiatives, this study focuses on evaluating and qualifying the irradiation facilities (mostly RTRs) within the United States that are suitable for advanced nuclear fuel, material, and system irradiation experiments aimed at establishing operational-performance limits and informing component and fuel designs so as to improve operational efficiencies and mitigate proliferation vulnerabilities, as well as for radioisotope production aimed at multipurpose applications. As a result, the findings of the present study support the acceleration of nuclear fuel and material qualifications, thus hastening new and advanced nuclear energy system demonstrations and radioisotope production efforts by using extended R&D.
irradiation experiment↗
Simulations Supporting the Development of Northstar's Indirect Beam Parameters Monitoring System
NorthStar Medical Radioisotopes, LLC is planning to produce the important medical radioisotope molybdenum-99 (Mo-99), the parent of technetium-99m (Tc-99m), through photonuclear reactions in molybdenum-100 (Mo-100). In this approach, a target comprising multiple thin disks of enriched molybdenum metal is bombarded with a 40-MeV electron beam. Electrons impinged on the molybdenum target produce bremsstrahlung X-rays that cause the nuclear reaction. Because enriched Mo-100 is expensive, there is a desire to utilize as much beam power as possible to achieve maximum production yield and minimize the size of the target. This requirement leads to very high beam power density (and heat deposition in the target), which creates challenging requirements for the cooling of the target. The critical part of the target is the target window. It separates the high-pressure helium cooled target from the vacuum beamline and the subject of structural and thermal stress. The temperature of the target window is proportional to the energy density deposited by the beam, so it is critical to maintain the desired beam profile on the target window. The feasibility of indirectly monitoring the maximum energy density of the beam on the beam window through beam losses at the main collimator (Collimator) before the production target was verified. A model of the NorthStar beam transport line was constructed for this purpose using MAD-X and Tao/Bmad codes. Beam optics were computed for the standard operational scenario, followed by an investigation involving approximately 400 cases with parameter variations in the last tuning quadrupoles. This was done to assess the correlation between losses in the collimator and the peak energy density on the target. We developed a model to explore the potential application of Optical transition radiation (OTR) for controlling beam parameters in the NorthStar beam delivery system. This model was based on a generic formula derived from the fundamental solution of the inhomogeneous wave equation of the vector potential, and allowed us to consider various surfaces, even those with irregular or random features, using numerical integration. We applied the model to OTR generated by relativistic electrons impacting an Inconel® 718 beam window. We examined cases with different levels of the window’s surface roughness, ranging from 0.5 to 3.0 microns of root square mean (RMS) deviation. The results of the OTR simulations provided distributions of OTR photons that can be used to study the limitations of optical systems for controlling beam parameters.
43 PARTICLE ACCELERATORS↗