Publications

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In the summer of 2020, the National Aeronautics and Space Administration (NASA) plans to launch a spacecraft as part of the Mars 2020 mission. One option for the rover on the proposed spacecraft uses a Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) to provide continuous electrical and thermal power for the mission. An alternative option being considered is a set of solar panels for electrical power with up to 80 Light-Weight Radioisotope Heater Units (LWRHUs) for local component heating. Both the MMRTG and the LWRHUs use radioactive plutonium dioxide. NASA is preparing an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act. The EIS will include information on the risks of mission accidents to the general public and on-site workers at the launch complex. This Nuclear Risk Assessment (NRA) addresses the responses of the MMRTG or LWRHU options to potential accident and abort conditions during the launch opportunity for the Mars 2020 mission and the associated consequences. This information provides the technical basis for the radiological risks of both options for the EIS.

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With renewed interest in disposal of heat-generating waste in bedded or domal salt formations, scoping analyses were conducted to estimate rates of waste package vertical movement. Vertical movement is found to result from thermal expansion, from upward creep or heave of the near-field salt, and from downward buoyant forces on the waste package. A two-pronged analysis approach was used, with thermal-mechanical creep modeling, and coupled thermal-viscous flow modeling. The thermal-mechanical approach used well-studied salt constitutive models, while the thermal-viscous approach represented the salt as a highly viscous fluid. The Sierra suite of coupled simulation codes was used for both approaches. The waste package in all simulations was a right-circular cylinder with the density of steel, in horizontal orientation. A time-decaying heat generation function was used to represent commercial spent fuel with typical burnup and 50-year age. Results from the thermal-mechanical base case showed approximately 27 cm initial uplift of the package, followed by gradual relaxation closely following the calculated temperature history. A similar displacement history was obtained with the package density set equal to that of salt. The slight difference in these runs is attributable to buoyant displacement (sinking) and is on the order of 1 mm in 2,000 years. Without heat generation the displacement stabilizes at a fraction of millimeter after a few hundred years. Results from thermal-viscous model were similar, except that the rate of sinking was constant after cooldown, at approximately 0.15 mm per 1,000 yr. In summary, all calculations showed vertical movement on the order of 1 mm or less in 2,000 yr, including calculations using well-established constitutive models for temperature-dependent salt deformation. Based on this finding, displacement of waste packages in a salt repository is not a significant repository performance issue.

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Coupled thermal-mechanical, three-dimensional, finite-element analyses were used to evaluate generic design concepts for a repository in salt, for spent nuclear fuel and high-level waste. This work used heat generation by spent nuclear fuel (SNF) typical of that presently stored at reactor sites in the U.S. For waste packages containing 4-PWR SNF assemblies, the results show peak temperatures within previously identified ranges acceptable for salt media. Peak temperatures and maximum backfill consolidation occur at the package-salt interface. Significant consolidation of the backfill, and closure of the mined opening, is projected to continue after peak temperatures are realized. For larger 21-PWR SNF packages, the peak temperature could approach 450°C locally or lower, depending on the aging history of the fuel. This ongoing study suggests the feasibility of a SNF management strategy using decay storage and larger (e.g., 21-PWR) waste packages.

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Thermal analyses of disposal strategies in a generic salt repository with high-level nuclear waste (HLW) have been completed. These studies were undertaken primarily to examine details of temperature distribution as a function of time for disposal concepts of wastes resulting from the recycling of spent nuclear fuel from a light water reactor. These analyses confirm that a conceptual salt repository for HLW appears feasible and worthy of more detailed evaluation. The analyses examined the temporal temperature distribution near a HLW package, as well as the far-field thermal response due to its transient heat pulse. The sensitivity of temperature distribution to several variations of primary features (e.g. the waste emplacement rate, waste configuration, etc.) was also determined. The principal observations of the study include the following. The temperatures involved ensure sufficient time for waste emplacement within a panel and adequate time to mine adjacent panels without adverse consequences. The modeled concept of a single level repository is workable. Thermal loading is the primary driver of repository-wide (far-field relative to the waste canister) heat effects. Decay storage, decreasing the loading of the waste package and changing the waste configuration are viable methods for reducing the peak waste and salt temperatures. The results of the thermal analyses show that with application of informed heat management strategies, thermal front migration rates are slow enough that a feasible design of the repository can be implemented. Peak temperatures within the waste package can be controlled with modest engineering considerations.

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