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The Saltstone facilities at the DOE Savannah River Site (SRS) stabilize and dispose of low-level radioactive salt solution originating from liquid waste storage tanks at the site. The Saltstone Production Facility (SPF) receives treated salt solution and mixes the aqueous waste with dry cement, blast furnace slag, and fly ash to form a grout slurry which is mechanically pumped into concrete disposal cells that compose the Saltstone Disposal Facility (SDF). The solidified grout is termed “saltstone”. Cementitious materials play a prominent role in the design and long-term performance of the SDF. The saltstone grout exhibits low permeability and diffusivity, and thus represents a physical barrier to waste release. The waste form is also reducing, which creates a chemical barrier to waste release for certain key radionuclides, notably Tc-99. Similarly, the concrete shell of an SDF disposal unit (SDU) represents an additional physical and chemical barrier to radionuclide release to the environment. Together the waste form and the SDU compose a robust containment structure at the time of facility closure. However, the physical and chemical state of cementitious materials will evolve over time through a variety of phenomena, leading to degraded barrier performance over Performance Assessment (PA) timescales of thousands to …

A 3-D STOMP model has been developed for the Portsmouth On-Site Waste Disposal Facility (OSWDF) at Site D as outlined in Appendix K of FBP 2013. This model projects the flow and transport of the following radionuclides to various points of assessments: Tc-99, U-234, U-235, U-236, U-238, Am-241, Np-237, Pu-238, Pu-239, Pu-240, Th-228, and Th-230. The model includes the radioactive decay of these parents, but does not include the associated daughter ingrowth because the STOMP model does not have the capability to model daughter ingrowth. The Savannah River National Laboratory (SRNL) provides herein a recommended method to account for daughter ingrowth in association with the Portsmouth OSWDF Performance Assessment (PA) modeling.

This is the final report of Phase 2 of the Secure Power Systems Professional project, a 3 phase project. DOE will post to their website upon release.

The Cementitious Barriers Partnership (CBP) Project is a multi-disciplinary, multi-institutional collaboration supported by the U.S. Department of Energy (US DOE) Office of Tank Waste and Nuclear Materials Management. The CBP program has developed a set of integrated tools (based on state-of-the-art models and leaching test methods) that help improve understanding and predictions of the long-term structural, hydraulic and chemical performance of cementitious barriers used in nuclear applications. Tools selected for and developed under this program have been used to evaluate and predict the behavior of cementitious barriers used in near-surface engineered waste disposal systems for periods of performance up to 100 years and longer for operating facilities and longer than 1000 years for waste disposal. The CBP Software Toolbox has produced tangible benefits to the DOE Performance Assessment (PA) community. A review of prior DOE PAs has provided a list of potential opportunities for improving cementitious barrier performance predictions through the use of the CBP software tools. These opportunities include: 1) impact of atmospheric exposure to concrete and grout before closure, such as accelerated slag and Tc-99 oxidation, 2) prediction of changes in Kd/mobility as a function of time that result from changing pH and redox conditions, 3) concrete degradation …

This report summarizes the approach that ORNL developed for screening a sample set of small coal stations for possible repowering with SMRs; the methodology employed, including spatial modeling; and initial results for these sample plants. The objective in conducting this type of siting evaluation is to demonstrate the capability to characterize specific sample coal plant sites to identify any particular issues associated with repowering existing coal stations with SMRs using OR-SAGE; it is not intended to be a definitive assessment per se as to the absolute suitability of any particular site.

This report summarizes the approach that ORNL developed for screening a sample set of US Department of Defense (DOD) military base sites and DOE sites for possible powering with an SMR; the methodology employed, including spatial modeling; and initial results for several sample sites. The objective in conducting this type of siting evaluation is demonstrate the capability to characterize specific DOD and DOE sites to identify any particular issues associated with powering the sites with an SMR using OR-SAGE; it is not intended to be a definitive assessment per se as to the absolute suitability of any particular site.

Because Engineered Trench #3 (ET#3) will be placed in the location previously designated for Slit Trench #12 (ST#12), Solid Waste Management (SWM) requested that the Savannah River National Laboratory (SRNL) determine if the ST#12 limits could be employed as surrogate disposal limits for ET#3 operations. SRNL documented in this Unreviewed Disposal Question Evaluation (UDQE) that the use of ST#12 limits as surrogates for the new ET#3 disposal unit will provide reasonable assurance that Department of Energy (DOE) 435.1 performance objectives and measures (USDOE, 1999) will be protected. Therefore new ET#3 inventory limits as determined by a Special Analysis (SA) are not required.

Model Evaluation Well ER-5-5 was drilled for the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office in support of Nevada Environmental Management Operations at the Nevada National Security Site (formerly known as the Nevada Test Site). The well was drilled in July and August 2012 as part of a model evaluation well program in the Frenchman Flat area of Nye County, Nevada. The primary purpose of the well was to provide detailed geologic, hydrogeologic, chemical, and radiological data that can be used to test and build confidence in the applicability of the Frenchman Flat Corrective Action Unit flow and transport models for their intended purpose. In particular, this well was designed to obtain data to evaluate the uncertainty in model forecasts of contaminant migration from the upgradient underground nuclear test MILK SHAKE, conducted in Emplacement Hole U-5k in 1968, which were considered to be uncertain due to the unknown extent of a basalt lava-flow aquifer present in this area. Well ER-5-5 is expected to provide information to refine the Phase II Frenchman Flat hydrostratigraphic framework model, if necessary, as well as to support future groundwater flow and transport modeling. The 31.1-centimeter (cm) diameter hole was drilled to …

Model Evaluation Well ER-11-2 was drilled for the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office in support of Nevada Environmental Management Operations at the Nevada National Security Site (formerly known as the Nevada Test Site). The well was drilled in August 2012 as part of a model evaluation program in the Frenchman Flat area of Nye County, Nevada. The primary purpose of the well was to provide detailed geologic, hydrogeologic, chemical, and radionuclide data that can be used to test and build confidence in the applicability of the Frenchman Flat Corrective Action Unit flow and transport models for their intended purpose. In particular, this well was designed to provide data to evaluate the uncertainty in model forecasts of contaminant migration from the upgradient underground nuclear test PIN STRIPE, conducted in borehole U-11b in 1966. Well ER-11-2 will provide information that can be used to refine the Phase II Frenchman Flat hydrostratigraphic framework model if necessary, as well as to support future groundwater flow and transport modeling. The main 31.1-centimeter (cm) hole was drilled to a total depth of 399.6 meters (m). A completion casing string was not set in Well ER-11-2. However, a piezometer string was …

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The opportunity for recovery of lost energy in the

The US Department of Energy sponsors the Advanced

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Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

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