Separation of fission products from high-level waste salt systems by partial crystallization: CsCl-NaCl-LiCl-KCl study
Not Available
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
SEARCH · Engineering Papers
Results for “waste management system”
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.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Nevada National Security Site E-Tunnel Wastewater Disposal System Quarterly Monitoring Report, Water Pollution Control Permit NEV 96021 (First Quarter Calendar Year 2020)
This report includes the results of monthly inspections and monitoring for January, February, and March 2020. Monthly inspections and monitoring were conducted on January 15, February 12, and March 17, 2020. Inspections include a visual assessment of the berms, ponding, water levels relative to the overflow pipes, tunneling or burrowing into the berms, soil conditions on and around the berms, erosion or sloughing of the slopes of the berms, seeps at the toe of the berms, deep-rooting botanicals on the berms, and fencing. Monthly monitoring includes measurements of the instantaneous flow rate, hydrogen ion activity (pH), and specific conductance of the E-Tunnel Wastewater Disposal System discharge.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Update on ASME AG-1 and Revision of the DOE Nuclear Air Cleaning Handbook - 20536
The American Society of Mechanical Engineers AG-1 Code on Nuclear Air and Gas Treatment is updated every two years. This paper provides an overview of issues being discussed within AG-1 committee meetings and revisions that will be included in the next version of the code. Additional information is provided on issues being discussed ranging from media velocity limits, control of airflow using actual versus and standard cubic feet per minute. Navigation and identification of applicable high-efficiency particulate air (HEPA) filter code sections can be confusing for those unfamiliar with AG-1. A general reorganization of filtration code sections clarifies geometry, flow, and performance envelopes for nuclear grade HEPA filters. The initial step consists of establishment of Section FN covering qualification of conventional, ceramic, metal, and high strength fibrous glass media. Future filter sections are currently expected to be characterized by geometry and flow direction (axial vs. radial). The Department of Energy's (DOE) Nuclear Air Cleaning Handbook (NACH) is also undergoing revision. An update on the status of that effort, along with issues and changes being considered. The original charter of the AG-1 Code was specific to commercial nuclear power plants, and to safety-related air cleaning systems. This has evolved over the years to include normal non-safety ventilation systems and process gas systems. Corollary guidance for DOE facilities has been available in several editions of the Nuclear Air Cleaning Handbook, starting with ORNL-NSIC 65. The latest edition of the Handbook, 2003, is undergoing revision and modifications. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Battery Energy Storage System (BESS) End-of-Performance and Decommissioning Considerations [Slides]
This presentation provides a comprehensive overview of end-of-performance and decommissioning considerations for large-scale Battery Energy Storage Systems (BESS). It outlines expected system lifespans, midterm assessment needs, and pathways for extending operational life through augmentation or repowering. The presentation details regulatory requirements that govern decommissioning plans, cost estimates, financial assurance mechanisms, and performance obligations across multiple jurisdictions. It further examines end-of-life equipment management, including recycling, waste handling, transportation, and environmental compliance. Designed to support Malawi's electricity-sector institutions, the presentation highlights how planning for decommissioning and environmental stewardship can be integrated early in project development to ensure safe, financially accountable, and environmentally responsible BESS system retirement.
24 POWER TRANSMISSION AND DISTRIBUTION↗
Application of Process Chemical Modeling to Optimize Radioactive Waste Disposal at the Savannah River Site – 24242
The Liquid Waste Program (LWP) managed by Savannah River Mission Completion (SRMC) is responsible for the treatment and disposal of waste at the Savannah River Site (SRS). Radioactive waste at SRS is stored and processed at four key facilities – each with their respective functions to store, blend, grout, or vitrify waste. The tank farm, where waste is stored, consists primarily of legacy waste with new material incoming from the Accelerated Basin De-inventory program (ABD), which is managed by Savannah River Nuclear Solutions (SRNS). System planning is done by SRNS and SRMC to optimize ABD and LWP operations, respectively.
Georgiou, Andreas↗
Evaluation of H-Canyon Ventilation Exhaust Tunnel Inspection at the Savannah River Site - 20052
The Department of Energy (DOE) H-Canyon Facility at the Savannah River Site (SRS) is operated by Savannah River Nuclear Solutions (SRNS). The Facility Structural Integrity Program [1] provides a process to inspect, evaluate, and document the conditions of passive safety related structural systems and components, their degradation mechanisms and their impact on the safety envelope. The Structural Integrity Program performs periodic visual inspections to confirm the systems and components can perform their safety function and recommend corrective actions before the function is compromised. The H-Canyon Exhaust Tunnel is part of the H-Canyon Ventilation System and is periodically inspected under the Structural Integrity Program. The Tunnel performs a Safety Class passive design feature function that is available 100% of the time. The underground reinforced concrete Tunnel directs Canyon process air from the Canyon to the Sand Filter System. The radiological and chemical airborne activity coupled with the physical configuration of the tunnel precludes a manned entry into the tunnel to perform the inspection. Unlike active systems that can be subjected to physical testing to confirm capability, qualification calculations, including modeling, is performed to assess passive design feature capabilities relative to operational, accident and natural phenomena hazards events. Input for these calculations come from knowledge of field conditions, design drawings and conservative assumptions. The knowledge of field conditions comes from concrete and soil testing activities as well as inspection data. The use of cameras on crawlers to obtain inspection data and how that data is used in qualification calculations is the focus of this discussion. Remote Tunnel inspections have been performed using cameras on a stick, referred to as pole cameras, prior to Calendar Year 2003 (CY2003). Since CY2003, six camera equipped crawlers have been used. Crawlers have evolved and improved by incorporating lessons learned from the previous inspections. The latest 2019 crawler successfully traveled the entire length of the tunnel between the Canyon and Sand Filter System and captured images that were previously unseen, thus providing key visual data for the Structural Integrity Program report [2]. The inspection data serves as 1) input to qualification calculations, 2) validates current conditions remain bounded by qualification calculations, and 3) documents a record of change over time. This data is vital for the effective evaluation of the tunnel structure and aids with predicting the life cycle service life safety envelope to protect the public, environment, and facility worker. The goal of the 2019 deployment was to use a more robust vehicle to ensure travel over obstacles, perform the inspection with higher resolution cameras, and have the ability to elevate the cameras to view surfaces previously obscured by abandon-in-place ducts in the tunnel. The inspection information will serve to strengthen the input basis for qualification calculations, versus using excessive conservative calculation assumptions. This 2019 inspection was compared with previous crawler visual evidence to identify change over time to project remaining service life and confirm field conditions remain bounded by qualification calculations. The results of the 2019 Exhaust Tunnel inspection continue to improve and provide a firm basis to make sound engineering judgements regarding tunnel capability and approaches in projecting remaining service life with an aging system and component. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Scenario development for safety assessment in deep geologic disposal of high-level radioactive waste and spent nuclear fuel: A review
Radiation and radioactive substances result in the production of radioactive wastes which require safe management and disposal to avoid risks to human health and the environment. To ensure permanent safe disposal, the performance of a deep geological repository for radioactive waste is assessed against internationally agreed risk-based standards. Assessing postclosure safety of the future system's evolution includes screening of features, events, and processes (FEPs) relevant to the situation, their subsequent development into scenarios, and finally the development and execution of safety assessment (SA) models. Global FEP catalogs describe important natural and man-made repository system features and identify events and processes that may affect these features into the future. By combining FEPs, many of which are uncertain, different possible future system evolution scenarios are derived. Repository licensing should consider both the reference or “base” evolution as well as alternative futures that may lead to radiation release, pollution, or exposures. Scenarios are used to derive and consider both base and alternative evolutions, often through production of scenario-specific SA models and the recombination of their results into an assessment of the risk of harm. Furthermore, while the FEP-based scenario development process outlined here has evolved somewhat since its development in the 1980s, the fundamental ideas remain unchanged. A spectrum of common approaches is given here (e.g., bottom–up vs. top–down scenario development, probabilistic vs. bounding handling of uncertainty), related to how individual numerical models for possible futures are converted into a determination as to whether the system is safe (i.e., how aleatoric uncertainty and scenarios are integrated through bounding or Monte Carlo approaches).
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Cesium Removal from SY-101 Tank Waste Using Crystalline Silicotitanate
The newly operational Tank Side Cesium Removal (TSCR) system removes radioactive cesium-137 ( 137 Cs) and solids from tank waste supernate in the 200 east area of the Hanford site. Efforts to expand the removal capabilities to the 200 west area are underway by a system anticipated to be called the West Area Risk Management (WARM) system. Laboratory-scale ion exchange processing using expected WARM unit operations were conducted to contribute toward Washington River Protection Solutions (WRPS) establishing accurate process flowsheets for the individual feed campaigns planned for the west area supernate pretreatment. This report describes the small-scale ion exchange testing with 8.0 L of filtered supernate from tank 241-SY-101 (referred to as SY-101) at 16 °C (62 °F) to demonstrate processing conditions that would be prototypic of what the WARM system may experience. One of the waste acceptance criteria (WAC) for the Waste Treatment Plant (WTP) Low-Activity Waste Facility is that the waste must contain less than 3.18×10 -5 Ci 137 Cs per mole of Na. For the SY-101 tank waste to meet this criterion, only 0.44% of the influent 137 Cs concentration may be delivered to the WTP; this requires a Cs decontamination factor of 227. Testing with SY-101 matched current TSCR prototypic operations where a lead-lag configuration is used until the lag column reached the WAC limit. Neither the lead nor lag columns reached the WAC, so a polish column was never utilized during this test. Feed was processed at 1.9 bed volumes (BVs) per hour; the flowrate, in terms of contact time with the crystalline silicotitate (CST) bed, matched the current flowrate at TSCR (an expected flowrate to be used at WARM). The lead column only reached 0.09% Cs breakthrough after processing ~1363 BVs of feed. Cesium breakthrough from the lag column was not observed during the entire processing. Table S.1 and Figure S.1 summarize the observed column performance and relevant Cs loading characteristics. Table S.1 and Figure S.1 also summarize the measured SY-101 Cs load performance.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Packaging and Transportation Regulatory Inconsistencies and Challenges - 20485
This paper presents regulatory inconsistencies within and between various regulatory agencies. These inconsistencies relate to the packaging and transportation of radioactive materials typically encountered during decommissioning including mixed wastes that are required to be compliant with more than one regulatory agency during shipment. At least one possible solution is presented for consideration. The following four examples are presented: - Packaging discrepancies for radioactive material meeting the DOT definition of Radioactive LSA-I that also contain lead may be compliant when shipped unpackaged per Department of Transportation (DOT) rules in 173.427 paragraph C. However, since they contain lead, they are required to be shipped in a package per the Environmental Protection Agency (EPA) Hazardous Waste Manifest (HWM) instructions. - The difference between the units of measure used in 49 CFR Part 172.101 Appendix A and 40 CFR, Part 355, Appendices A and B. - The requirement for outer packaging for radioactive gauges meeting all other requirements of 49 CFR 173.424 and 173.425. - E-Manifesting EPA system difficult for waste materials containing both radioactive class 7 and PCBs. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Progress in demonstrating the performance of Engineered Barriers for the disposal of HLW and SNF: lab studies and demonstration experiments - 20293
Disposal of high-level radioactive waste (HLW) and spent nuclear fuels (SNF) in a geological disposal facility requires a fundamental understanding of the degradation processes of materials used in its engineered barrier system (EBS) over very long timescales. Beyond the wasteform itself (i.e. the borosilicate glass for HLW or the fuel matrix and cladding for SNF), materials often employed in a disposal concept include metallic containers (typically copper and/or carbon steel) and clay-based buffer materials (typically bentonite). In this paper we explore key developments on the long-term evolution of materials employed for the immobilisation, isolation and containment of HLW and SNF in a geological disposal facility, including nuclear glasses, candidate container materials and clay buffers. Our discussion is based on the latest findings from both experimental studies and in-situ demonstration experiments. We present relevant findings from the long-term evolution of HLW glasses (UK composition), metallic materials for the containment system (envisaged in Switzerland, Sweden, Canada and other countries), and bentonite clays (envisaged in many national disposal programmes). (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
TCCR Operational Summary and Optimization for Tank 9 Processing - 20314
Savannah River Remediation (SRR) manages and operates the liquid waste facilities at Savannah River Site (SRS) for the Department of Energy (DOE). Stored liquid waste is a complex mixture of insoluble solids (sludge) and soluble salts in an alkaline solution. SRR has deployed the Tank Closure Cesium Removal (TCCR) system, a tank-side ion exchange process, to remove radioactive cesium from salt waste and enable onsite disposal of the resulting decontaminated salt solution as low-level waste at the Saltstone facilities. The TCCR system consists of two prefilters, four ion exchange (IX) columns, one resin trap, and a ventilation system. The IX process uses a form of inorganic crystalline silicotitanate (CST), which has a high affinity for cesium and other alkali metals, strontium, and actinides. This process is currently deployed utilizing salt feed from Tank 10, with future plans to dissolve solid salt in Tank 9 and transfer the salt solution to Tank 10 for processing through TCCR. The feed for TCCR must be created from salt-cake in Tank 10 through a dissolution process. Once enough salt has been dissolved, a qualification process is entered. This process characterizes the feed and ensures the cesium loading on the columns will not cause boiling of waste within the columns during or after processing. Once the batch has been qualified, salt waste is fed to the TCCR system through a transfer pump in the center of the tank. The waste is filtered through a set of two shielded, dead-end prefilters that prevent solids buildup in the columns. The filtered salt solution then travels to the shielded IX columns, which can be operated individually or in series, where the cesium is sorbed on the CST media. The decontaminated salt solution (DSS) then travels through a resin trap and out of the module to Tank 11. TCCR has successfully processed approximately 795,000 L of Tank 10H radioactive salt waste over two batches to date. There has not yet been a system induced shutdown. The prefilters performed as expected with only minor degradation in recovery of differential pressure after a backflush sequence. The time between backflushes decreased as each batch reached the end of processing. The hydraulics in the IXCs mostly performed as expected at all flow rates, except for one IXC that will be further investigated during Batch 3 processing. The TCCR system has shown some opportunities for more efficient processing during the length of the demonstration so far. For future processing of material from Tank 9H through Tank 10H and the TCCR unit, TCCR 1A will implement changes to the prefilters and the IXCs. The prefilters will have an increased surface area and a new filter media in an effort to increase time between filter swaps and improve backwashing cleaning capability. The IXCs will have a reduced diameter to allow for increased heat transfer out of the column and increased loading of Cs-137. Additionally, a new form of CST with an increased kinetic performance is being investigated for use during TCCR 1A operation. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
A comparative analysis of numerical approaches for the description of gas flow in clay-based repository systems: From a laboratory to a large-scale gas injection test
There is nowadays a consensus among many countries that geological disposal is a favourable solution for the long-term management. Although different host formations and different barrier systems are under consideration around the world, clay-based materials form an important component for waste isolation in most national programmes. Hence, a good comprehension of the effect of gas flow on the hydro-mechanical behaviour of clay-based soils is essential, both at laboratory and field scale. Task B under the international cooperative project DECOVALEX-2023 has recently shown that, after some enhancement, models can be employed to reproduce laboratory scale tests, even with different sample geometries 37 . However, further work is required to understand whether they can be applied to simulate a large-scale experiment. Up-scaling of models for the advective transport of gas through clay-based low permeable material presents a number of problems related to the difficulty in obtaining consistent hydrogeological parameters and constitutive relationships at both laboratory and field scale. Based on a unique dataset from a large-scale gas injection test (Lasgit) performed at the Äspö Hard Rock Laboratory (Sweden), Task B within DECOVALEX-2023 has explored the refinement of these numerical strategies applied to the simulation of gas flow. Work performed within the task reveals that codes do not need to be substantially modified from the laboratory models to reproduce full-scale tests: indeed, model parameters calibrated and validated at laboratory scale have been applied to predict field scale gas flow at Lasgit, including peak gas pressure and injected cumulative gas volume. By means of (1) the introduction of interfaces between blocks to reflect the experimental configuration and the (2) adjustment of some parameters (e.g., higher permeability), the updated models are able to represent most of the key features observed in the experimental data, even at a large scale.
Tamayo-Mas, E↗
DEVELOPMENT OF AN INTEGRATED SECURITY AND SAFETY MONITORING SYSTEM FOR SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE TRANSPORTATION
This paper provides an overview of the progress to date, and discussion of the path forward, related to designing, fabricating, and testing an integrated security and safety monitoring system (ISSMS) for railcars used to transport spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in the United States. The system will comply with the US Department of Energy’s (DOE) Order 460.2B “Departmental Materials Transportation Management” and the Association of American Railroads’ (AAR) standard S-2043 “Performance Specification for Trains Used to Carry High-level Radioactive Material” [1] developed specifically for railcars used to transport high-level radioactive material (HLRM). DOE is in the process of developing and testing an ISSMS that will satisfy both DOE requirements and AAR standards. DOE has already developed railcar designs for transportation of HLRM. In 2024, DOE’s Atlas railcar project completed the design, fabrication and testing of three railcar types: transportation cask-carrying, buffer, and security escort, resulting in AAR conditional approval to operate on freight rail networks in North America. DOE decided to combine the required security and safety systems into one system, and this combined system is the subject of the current effort. DOE is developing and proposes to implement the ISSMS for these railcars as part of the build out of the railcar fleet. DOE began planning for development of the ISSMS in February 2020. The project is divided into seven phases starting with conceptual design and continuing through production design, as shown in Figure 1. An earlier version of the system was tested as part of the Atlas railcar consist demonstration test run in 2023. This paper describes the design features and system testing using the Atlas project railcars, and laboratory testing completed to date. The paper will also describe the activities planned to support the DOE project to ship the High Burn-Up Research Cask (HBRC) in 2027.
Schultze, Michael [ORNL] (ORCID:0000000283205671)↗
2018 Long-Term Hydrologic Monitoring Program Report for Rio Blanco, Colorado, Site
This report presents the monitoring data collected by the U.S. Department of Energy (DOE) Office of Legacy Management (LM) at the Rio Blanco, Colorado, Site (Figure 1). The Rio Blanco site was the location of an underground nuclear test during which three nuclear devices were detonated nearly simultaneously in a single borehole in 1973. The test resulted in residual radionuclide contamination at the detonation depths of 5840, 6230, and 6690 feet (ft) (DOE 2015). Monitoring includes the collection of samples from groundwater wells and surface water locations near the site to assess for any potential impacts that may be attributed to the Rio Blanco nuclear test. This report summarizes the laboratory analytical results obtained from the sampling event conducted in 2018. This annual report and previous reports are available on the LM public website at https://www.lm.doe.gov/rio_blanco/Sites.aspx. Data collected during this and previous monitoring events are available on the Geospatial Environmental Mapping System (GEMS) website at https://gems.lm.doe.gov/#site=RBL.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
The Need for a Maximum Intensity Accident in ANS 8-Series Standards
Y-12 is currently designing, installing, and qualifying a CAAS system in one of its nuclear facilities. The CAAS is considered a safety-significant system.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Vadose and Saturated Zone Flow and Transport Model Package Report for the Active Trenches of the Low-Level Burial Grounds, Hanford Site, Washington
This model package report (MPR) documents the development of the integrated vadose and saturated zone flow and transport model developed for the performance assessment of Trenches 31 and 34 in the low-level burial grounds (LLBGs). This modeling capability is intended for use in addressing the analysis requirements outlined in DOE O 435.1, Chg 1, Radioactive Waste Management1. The overall objective of the modeling effort is to provide a basis for making informed disposal decisions pertinent to Trenches 31 and 34. The purpose of the MPR is to document the development of the three-dimensional numerical vadose zone and saturated flow and transport model test case and evaluate its adequacy to support the LLBG performance assessment. The purpose is not to present results for DOE 435.1 decision making. The use of the model to perform base case and sensitivity analysis for the LLBG performance assessment, including inputs and results, is documented separately in subsequent environmental calculation files. This report discusses the development and translation of the conceptual model for flow and contaminant transport into the LLBG performance assessment three-dimensional numerical flow and transport model evaluated using the Subsurface Transport Over Multiple Phases (STOMP©) simulator. The development of representative geologic framework is described along with the implementation of waste release models used to represent contaminant release from waste disposed in the trenches. The report also provides the technical basis for specific model parameters and boundary conditions, along with description of modeling assumptions. This MPR includes certain calculations that are necessary to demonstrate the soundness of the model. Results provided by the model include vadose zone and saturated zone flow fields, and estimates of the possible future concentration in groundwater of technetium-99 and iodine-129 as example test cases. As an evaluation of the test cases, the model estimates of current vadose conditions are compared to available and analogous field and laboratory data, and compared to the results from the original performance assessment model (WHC-EP-0645, Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds 2 ). The features, events, and processes applicable to vadose zone and saturated zone flow and transport model are identified, and representative initial estimates for various parameters are documented. Note that the parameter estimates presented in this MPR are for illustration purposes, and may or may not reflect values selected for eventual performance assessment. Several key topical discussions (i.e., basis for recharge estimates, basis for vadose zone modeling, basis for saturated zone model development, and calibration) are included, which serve as the groundwork for confidence building for the groundwater pathway modeling and results. Numerical simulation results based on an example test case are included to illustrate the use of the combined saturated-unsaturated model in performance assessment calculations. Sensitivity and uncertainty results are not included in this MPR. Those results will be included in future environmental calculation files. The inclusion of the trapezoidal trench geometry and construction details in the finite difference grid introduces some gross simplifications regarding the trench and liner systems. The model test case presented in this MPR includes the assumption that the trench liner system does not affect flow through or around the trenches after the liner system is assumed to fail. This MPR also includes results of other test cases that involve alternate assumptions about how to incorporate the hydraulic effects of the trenches into the vadose zone of the model. These alternate cases do not attempt to account for the presence of the liner system after its assumed failure either. None of these cases is considered to be the base case at this time. Analysis and alternate cases that attempt to account for the presence of the liner system in greater detail, and its effect on flow through and around it, are to be documented in subsequent environmental calculation files. Depending on the results of those analyses and alternate cases, the eventual base case may involve more detailed inclusion of the effects of the liner system hydraulics.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Panel Session 45: Savannah River National Laboratory Advancing the EM Mission Through Technology
This panel focused on Savannah River National Laboratory (SRNL) activities and addressed: 1) Initial development work that SRNL performed to develop the technologies and processes, as well as design the equipment for the DOE EM HLW tank waste treatment and soil and groundwater mission, 2) Implemented technology and process improvements to the EM systems, and 3) Current technologies being evaluated for future implementation to improve throughput and reduce cost/schedule. Panelists with presentations: Advancing the EM Mission: SRNL Technology Approach (Connie Herman); Nuclear Materials Management (William Bates); Adjust, Adapt, Attest, and Adjudicate (Bruce Wiersma); SRNL Research and Development for Technological Advancements in the Liquid Waste Mission at SRS (Frank Pennebaker); EM TD Applies Solutions at SRS (Andrew Fellinger); Remediation of Complex Groundwater Plumes: Three Decades of Cleanup Success (Carol Eddy-Dilek)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Simulants for Testing the Blend Can Loading System
Eleven simulants have been provided to the Savannah River National Laboratory (SRNL) Mechanical Systems & Custom Equipment Development (MSCED) organization to further test the capabilities of the Blend Can Loading System (BCLS). The basis for the materials used in these simulants are based on SRNL-STI-2020-00503, “Simulant Recommendation for the Blend Can Loading System” (Ref. i). The 11 different simulants provided to MSCED are summarized in Table ES-1. This table provides a basis of why they were selected and what properties were specifically targeted, given the composition of the procured materials.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗