Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “CLOSURES”

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.

At least 1,279 records · Page 71

Mine Flooding History of a Regional Below-Drainage Coalfield Dominated by Barrier Leakage (1970–2014)

A 44-year record of water level fluctuations in a series of adjacent closed underground mines documents the history of closure and mine flooding in the Fairmont Coalfield, one of the oldest coal mining districts in the Pittsburgh coal basin, West Virginia, USA. As closures proceeded and mines began to flood, US environmental regulations were first enacted mandating mine water control and treatment, rendering uncontrolled surface discharges unacceptable. The purpose of this study is to present this flooding history and to identify critical events that determined how mine pools evolved in this case. Also examined is the strategy developed to control and treat water from these mines. Flooding is visualized using both water level hydrographs and mine flooding maps with the latter constructed assuming mine water hydraulic continuity between one or more mines. The earliest flooding formed small pools within near-surface mines closed prior to 1962 yet still pumped following closure to minimize leaking into adjacent still-active workings. These subpools gradually enlarged and merged as more closures occurred and the need for protective pumping was removed, forming what is today referred to as the unconfined Fairmont Pool. Later, deeper mines, separated by intact updip barriers from the Fairmont Pool, were closed and flooded more gradually, supplied in large part by leakage from the Fairmont Pool. By 1985, all mines except 2 had closed and by 1994 all had fully flooded, with the Fairmont Pool interconnected to deeper single mine pools via barrier leakage. As protective pumping ceased, the Fairmont Pool rose to a water level 3 m higher than surface drainage elevation and in 1997 discharged from an undermined section of Buffalo Creek near the Monongahela River. The principal mine operator in the basin then designed a pumping system to transfer water from the Fairmont Pool to their existing treatment facilities to the north, thus terminating the discharge. It may be concluded that the progress of mine flooding was influenced by mining history and design, by the timing of closures, by barrier leakage conditions, and by geologic structure. A key element in how flooding proceeded was the presence of a series of intact barriers separating deep from shallow mines. The shallow mines closed and flooded early, but then lost sufficient water by barrier leakage into the deeper mines to delay the completion of flooding until after the deep mines had all closed and flooded as well. Intensive mine water control has continued from the 1997 breakout to the present. The final water control scheme was likely unanticipated and serendipitous; future district-wide mining efforts should be advised to consider in advance closeout strategies to control mine water postmining.

Donovan, Joseph J.↗

Assessing Gaseous Hydrogen Assisted Fatigue Crack Growth Susceptibility of Pipeline Steel Weld Fusion Zones and Heat Affected Zones

The objective of this work was twofold: (1) measure reliable fatigue crack growth relationships for X65 steel and its girth weld in high-pressure hydrogen gas to enable structural integrity assessments of hydrogen pipelines, and (2) evaluate the hydrogen accelerated fatigue crack growth susceptibility of the weld fusion zone and heat-affected zone relative to the base metal. Fatigue crack growth relationships (da/dN versus ΔK) were measured for girth welded X65 pipeline steel in high pressure hydrogen gas (21 MPa) and in air. Hydrogen assisted fatigue crack growth was observed for the base metal (BM), fusion zone (FZ), and heat-affected zone (HAZ), and was manifested through crack growth rates reaching nearly an order of magnitude acceleration over rates in air. Addditionally, at higher ΔK values, crack growth rates of BM, FZ, and HAZ were coincident; however, at lower ΔK, the fatigue crack growth relationships exhibited some divergence with the fusion zone having the highest crack growth rates. These relative fatigue crack growth rates in the BM, FZ, and HAZ were provisional, however, since both crack closure and residual stress contributed to the crack-tip driving force in specimens extracted from the HAZ. Despite the relatively high applied R-ratio (R = 0.5), crack closure was detected in the heat affected zone tests, in contrast to the absence of crack closure in the base metal tests. Crack closure corrections were performed using the adjusted compliance ratio method and the effect of residual stress on K max was determined by the crack-compliance method. Crack-tip driving forces that account for closure and residual stress effects were quantified as a weighted function of ΔK and K max (i.e., K norm ), and the resulting da/dN versus K norm relationships showed that the HAZ exhibited higher hydrogen accelerated fatigue crack growth rates than the BM at lower K norm values.

36 MATERIALS SCIENCE↗

TANK 18 AND 19-F TIER 1A EQUIPMENT FILL MOCK UP TEST SUMMARY

The United States Department of Energy (US DOE) has determined that Tanks 18-F and 19-F have met the F-Tank Farm (FTF) General Closure Plan Requirements and are ready to be permanently closed. The high-level waste (HLW) tanks have been isolated from FTF facilities. To complete operational closure they will be filled with grout for the purpose of: (1) physically stabilizing the tanks, (2) limiting/eliminating vertical pathways to residual waste, (3) discouraging future intrusion, and (4) providing an alkaline, chemical reducing environment within the closure boundary to control speciation and solubility of select radionuclides. Bulk waste removal and heel removal equipment remain in Tanks 18-F and 19-F. This equipment includes the Advance Design Mixer Pump (ADMP), transfer pumps, transfer jets, standard slurry mixer pumps, equipment-support masts, sampling masts, dip tube assemblies and robotic crawlers. The present Tank 18 and 19-F closure strategy is to grout the equipment in place and eliminate vertical pathways by filling voids in the equipment to vertical fast pathways and water infiltration. The mock-up tests described in this report were intended to address placement issues identified for grouting the equipment that will be left in Tank 18-F and Tank 19-F. The Tank 18-F and 19-F closure strategy document states that one of the Performance Assessment (PA) requirements for a closed tank is that equipment remaining in the tank be filled to the extent practical and that vertical flow paths 1 inch and larger be grouted. The specific objectives of the Tier 1A equipment grout mock-up testing include: (1) Identifying the most limiting equipment configurations with respect to internal void space filling; (2) Specifying and constructing initial test geometries and forms that represent scaled boundary conditions; (3) Identifying a target grout rheology for evaluation in the scaled mock-up configurations; (4) Scaling-up production of a grout mix with the target rheology (16 second flow cone value) from 0.25 cubic feet to 4.3 cubic feet. (Ten 0.43 cubic batches were produced because full-scale equipment was not available for the Tier 1A test.); (5) Demonstrating continuous gravity filling of the ADMP mock up test form; (6) Demonstrating continuous gravity filling of 1 inch and 2 inch schedule 40 pipe; and (7) Demonstrating filling of 1 inch and 2 inch schedule 40 pipe from the bottom up by discharging through a tube inserted into the pipes. The Tier 1A mock-up test focused on the ADMP and pipes at least one inch in diameter. The ADMP which is located in center riser of Tank 18-F is a concern because the column for this long-shaft (55 ft) pump is unique and modification to the pump prior to placing it in service limited the flow path options for filling by creating a single flow path for filling and venting the ADMP support column. The large size, vertical orientation, and complicated flow path in the ADMP warrants a detailed description of this piece of ancillary equipment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Center for Gyrokinetic/MHD Hybrid Simulation of Energetic Particle Physics in Toroidal Plasmas (Final Report)

At Colorado University-Boulder the primary task is to extend our gyrokinetic Particle-in-Cell simulation of tokamak micro-turbulence and transport to the area of energetic particle physics. We have implemented a gyrokinetic ion/massless fluid electron hybrid model in the global δƒ-PIC code GEM, and benchmarked the code with analytic results on the thermal ion radiative damping rate of Toroidal Alfvén Eigenmodes (TAE) and with mode frequency and spatial structure from eigenmode analysis. We also performed nonlinear simulations of both a single-n mode (n is the toroidal mode number) and multiple-n modes, and in the case of single-n, benchmarked the code on the saturation amplitude vs. particle collision rate with analytical theory. Most simulations use the f method for both ions species, but we have explored the full-ƒ method for energetic particles in cases where the burst amplitude of the excited instabilities is large as to cause significant re-distribution or loss of the energetic particles. We used the hybrid model to study the stability of high-n TAEs in ITER. Our simulations show that the most unstable modes in ITER lie in the rage of 10 < n < 20. Thermal ion pressure effect and alpha particles non-perturbative effect are important in determining the mode radial location and stability threshold. The thermal ion Landau damping rate and radiative damping rate from the simulations are compared with analytical estimates. The thermal ion Landau damping is the dominant damping mechanism. Plasma elongation has a strong stabilizing effect on the alpha driven TAEs. The central alpha particle pressure threshold for the most unstable n=15 mode is about β α (0) = 0.7% for the fully shaped ITER equilibrium. We also carried nonlinear simulations of the most unstable n = 15 mode and found that the saturation amplitude for the nominal ITER discharge is too low to cause large redistribution or loss of alpha particles. To include kinetic electron effects in the hybrid model we have studied a kinetic electron closure scheme for the fluid electron model. The most important element of the closure scheme is a complete Ohm’s law for the parallel electric field E || , derived by combining the quasi-neutrality condition, the Ampere’s equation and the v || moment of the gyrokinetic equations. A discretization method for the closure scheme is studied in detail for a three-dimensional shear-less slab plasma. It is found that for long-wavelength shear Alfvén waves the kinetic closure scheme is both more accurate and robust than the previous GEM algorithm using the split-scheme, whereas for the ion-gradient-driven instability the previous algorithm is more efficient. This kinetic electron closure scheme will be implemented in GEM in the future. We have studied the beam driven Reverse Shear Alfven Eigenmodes (RSAE) observed in DIII-D discharge 142111. For this purpose a new scheme for obtaining the electric potential is implemented, i.e., by solving the gyrokinetic moment (GKM) equation, which is essentially the equation for $∂\phi/∂t$ used in GEM’s split-weight scheme, and then integrating in time. Due to charge-neutrality the ExB motions of the equilibrium densities of all species cancel each other and do not cause charge separation if there is no finite Larmor radius effect. The advantage of solving the GKM equation is that this lowest-order cancellation can be made explicit. The GKM approach is found to be more accurate and robust. GEM simulations have reproduced many features of RSAE seen in the experiment, such as frequency chirping and the chirping range. It has been reported by other simulation codes that the shearing direction of the mode structure in the poloidal plane disagrees with observation. We found that the mode structure, including the shearing in the poloidal plane, is in general sensitive to the beam distribution. Using the same beam density profile as in other codes but with a slowing-down distribution in velocity, GEM simulations reproduce the mode shearing direction seen in the experiment. We have carried out extensive nonlinear simulations of RSAE.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Monitoring Potential Transport of Radioactive Contaminants in Shallow Ephemeral Channels: FY 2012

The US Department of Energy (DOE) National Nuclear Security Administration (NNSA), Nevada Site Office (NSO), Environmental Management’s Soils Activity has authorized the Desert Research Institute (DRI) to conduct field assessments of potential sediment transport of contaminated soil from Corrective Action Unit (CAU) 550, Smoky Contamination Area (CA), during precipitation runoff events. CAU 550 includes Corrective Action Sites (CASs) 08-23-03, 08-23-04, 08-23-06, and 08-23-07; these CASs are associated with tests designated Ceres, Smoky, Oberon, and Titania, respectively. Field measurements at the T-4 Atmospheric Test Site, CAU 370, suggest that radioactive material may have migrated along a shallow ephemeral drainage that traverses the site (NNSA/NSO, 2009). (It is not entirely clear how contaminated soils got into their present location at the T-4 Site, but flow to the channel has been redirected and the contamination does not appear to be migrating at present.) Although DRI initially looked at the CAU 370 site, given that it could not be confirmed that migration of contamination into the channel was natural, an alternate study site was selected at CAU 550. Aerial surveys in selected portions of the Nevada National Security Site (NNSS) also suggest that radioactivity may be migrating along ephemeral channels in Areas 3, 8, 11, 18, and 25 (Colton, 1999). Figure 1 shows the results of a low-elevation aerial survey (Colton, 1999) in Area 8. The numbered markers in Figure 1 identify ground zero for three safety experiments conducted in 1958 [Oberon (number 1), Ceres (number 2), and Titania (number 4)] and a weapons effects test conducted in 1964, Mudpack (number 3). This survey suggests contaminants may be migrating down the ephemeral channels that traverse CAU 550. Note particularly the lobe of higher concentration extending southeastward at the south end of the high concentration area marked as number 3 in Figure 1. CAU 550 in Area 8 of the NNSS was selected for the study because the aerial survey indicates that a channel mapped on the United States Geological Survey topographic map of the area traverses the south end of the area of surface contamination; this channel lies south of the point marked number 3 in Figure 1, and anecdotal information indicates that sediment has been deposited on the road bordering the southeast boundary of the CAU from an adjacent channel (Traynor, J, personal communication, 2011). Because contamination is particularly close to the boundary of CAU 550, Smoky CA, it is important to know if contaminants are moving, what meteorological conditions result in movement of contaminated soils, and what particle size fractions associated with contamination are involved. Closure plans are being developed for the CAUs on the NNSS. The closure plans may include post-closure monitoring for possible release of radioactive contaminants. Determining the potential for transport of contaminated soils under ambient climatic conditions will facilitate an appropriate closure design and post-closure monitoring program.

54 ENVIRONMENTAL SCIENCES↗

Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analyses (2012 Annual Summary Report)

The Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (National Security Technologies, LLC 2007a) requires an annual review to assess the adequacy of the performance assessments (PAs) and composite analyses (CAs), with the results submitted to the U.S. Department of Energy (DOE) Office of Environmental Management. The Disposal Authorization Statements for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) also require that such reviews be made and that secondary or minor unresolved issues be tracked and addressed as part of the maintenance plan (DOE 1999a, 2000). The U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office performed an annual review of the Area 3 and Area 5 RWMS PAs and CAs for fiscal year (FY) 2012. This annual summary report presents data and conclusions from the FY 2012 review, and determines the adequacy of the PAs and CAs. Operational factors (e.g., waste forms and containers, facility design, and waste receipts), closure plans, monitoring results, and research and development (RD) activities were reviewed to determine the adequacy of the PAs. Likewise, the environmental restoration activities at the Nevada National Security Site (NNSS) relevant to the sources of residual radioactive material that are considered in the CAs, the land-use planning, and the results of the environmental monitoring and RD activities were reviewed to determine the adequacy of the CAs. Important developments in FY 2012 include the following: Release of a special analysis for the Area 3 RWMS assessing the continuing validity of the PA and CA; Development of a new Area 5 RWMS closure inventory estimate based on disposals through FY 2012; Evaluation of new or revised waste streams by special analysis; and Development of version 4.114 of the Area 5 RWMS GoldSim PA model. The Area 3 RWMS has been in inactive status since July 1, 2006, with the last shipment received in April 2006. The FY 2012 review of operations, facility design, closure plans, monitoring results, and RD results for the Area 3 RWMS indicates no changes that would impact PA validity. A special analysis using the Area 3 RWMS v2.102 GoldSim PA model was prepared to update the PA results for the Area 3 RWMS in FY 2012. The special analysis concludes that all performance objectives can be met and the Area 3 RWMS PA remains valid. There is no need to the revise the Area 3 RWMS PA. Review of Area 5 RWMS operations, design, closure plans, monitoring results, and RD activities indicates no significant changes other than an increase in the inventory disposed. The FY 2012 PA results, generated with the Area 5 RWMS v4.114 GoldSim PA model, indicate that there continues to be a reasonable expectation of meeting all performance objectives. The results and conclusions of the Area 5 RWMS PA are judged valid, and there is no need to the revise the PA. A review of changes potentially impacting the CAs indicates that no significant changes occurred in FY 2012. The continuing adequacy of the CAs was evaluated with the new models, and no significant changes that would alter CA results or conclusions were found. The revision of the Area 3 RWMS CA, which will include the Underground Test Area source term (Corrective Action Unit [CAU] 97), is scheduled for FY 2024, following the completion of the Yucca Flat CAU 97 Corrective Action Decision Document/Corrective Action Plan in FY 2016. Inclusion of the Frenchman Flat CAU 98 results in the Area 5 RWMS CA is scheduled for FY 2016, pending the completion of the CAU 98 closure report in FY 2015. Near-term RD efforts will focus on continuing development of the Area 3 and Area 5 RWMS GoldSim PA/CA and inventory models.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

E-Area Vault Concrete Material Property And Vault Durability/Degradation Projection Recommendations

Subsequent to the 2008 E-Area Low-Level Waste Facility (ELLWF) Performance Assessment (PA) (WSRC 2008), two additional E-Area vault concrete property testing programs have been conducted (Dixon and Phifer 2010 and SIMCO 2011a) and two additional E-Area vault concrete durability modeling projections have been made (Langton 2009 and SIMCO 2012). All the information/data from these reports has been evaluated and consolidated herein by the Savannah River National Laboratory (SRNL) at the request of Solid Waste Management (SWM) to produce E-Area vault concrete hydraulic and physical property data and vault durability/degradation projection recommendations that are adequately justified for use within associated Special Analyses (SAs) and future PA updates. The Low Activity Waste (LAW) and Intermediate Level (IL) Vaults structural degradation predictions produced by Carey 2006 and Peregoy 2006, respectively, which were used as the basis for the 2008 ELLWF PA, remain valid based upon the results of the E-Area vault concrete durability simulations reported by Langton 2009 and those reported by SIMCO 2012. Therefore revised structural degradation predictions are not required so long as the mean thickness of the closure cap overlying the vaults is no greater than that assumed within Carey 2006 and Peregoy 2006. For the LAW Vault structural degradation prediction (Carey 2006), the mean thickness of the overlying closure cap was taken as nine feet. For the IL Vault structural degradation prediction (Peregoy 2006), the mean thickness of the overlying closure cap was taken as eight feet. The mean closure cap thicknesses as described here for both E-Area Vaults will be included as a key input and assumption (I&A) in the next revision to the closure plan for the ELLWF (Phifer et al. 2009). In addition, it has been identified as new input to the PA model to be assessed in the ongoing update to the new PA Information UDQE (Flach 2013). Once the UDQE is approved, the SWM Key I&A database will be updated with this new information.

36 MATERIALS SCIENCE↗

2016 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analyses

The Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (National Security Technologies, LLC 2007a) requires an annual review to assess the adequacy of the performance assessments (PAs) and composite analyses (CAs), with the results submitted to the U.S. Department of Energy (DOE) Office of Environmental Management. The Disposal Authorization Statements for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) also require that such reviews be made and that secondary or minor unresolved issues be tracked and addressed as part of the maintenance plan.The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) performed an annual review of the Area 3 and Area 5 RWMS PAs and CAs for federal fiscal year (FY) 2016. This annual summary report presents results and conclusions of the FY 2016 review, and determines the adequacy of the PAs and CAs. Operational factors (e.g., waste forms and containers, facility design, and waste receipts), closure plans, monitoring results, and research and development (R&D) activities were reviewed to determine the adequacy of the PAs. Likewise, the environmental restoration activities at the Nevada National Security Site (NNSS) relevant to the sources of residual radioactive material that are considered in the CAs, the land-use planning, and the results of the environmental monitoring and R&D activities were reviewed to determine the adequacy of the CAs. Important developments in FY 2016 include the following: Development of a new Area 5 RWMS closure inventory estimate based on disposals through FY 2016, Evaluation of new or revised waste streams by special analysis, and Development and acceptance of version 4.200 of the Area 5 RWMS GoldSim PA/CA model. The Area 3 RWMS has been in inactive status since July 1, 2006, with the last shipment received in April 2006. The FY 2016 review of operations, facility design, closure plans, monitoring results, and R&D results for the Area 3 RWMS indicates no changes that would impact PA validity. The conclusion of the annual review is that all performance objectives can be met and the Area 3 RWMS PA remains valid. There is no need to revise the Area 3 RWMS PA. Review of Area 5 RWMS operations, design, closure plans, monitoring results, and R&D activities indicates that no significant changes have occurred. The FY 2016 PA results, generated with the Area 5 RWMS v4.200 GoldSim PA model, indicate that there continues to be a reasonable expectation of meeting all performance objectives. All results remain less than 50% of their performance objectives. The results and conclusions of the Area 5 RWMS PA are judged valid, and there is no need to the revise the PA. A review of changes potentially impacting the CAs indicates that no significant changes occurred in FY 2016. A new Corrective Action Unit (CAU), CAU 576, with corrective action sites near the Area 3 RWMS and Area 5 RWMS was identified in FY 2016. The sites are currently under investigation and not expected to be important sources for the CA. The continuing adequacy of the CAs was evaluated with the new models, and no significant changes that would alter the CAs’ results or conclusions were found. The revision of the Area 3 RWMS CA, which will include the Yucca Flat Underground Test Area (CAU 97) source, is scheduled for FY 2024, following the completion of the Corrective Action Decision Document/Corrective Action Plan. Inclusion of the Frenchman Flat Underground Test Area (CAU 98) source in the Area 5 RWMS CA is planned pending the completion of the CAU 98 Closure Report in FY 2017. Near-term R&D efforts will focus on continuing development of the PA, CA, and inventory models for the Area 3 and Area 5 RWMSs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Machine Learning Assisted Safety Modeling and Analysis of Advanced Reactors

With the advances in computational power and numerical methods, analysts can now rely on first-principle simulations to predict ultra-fine details in a variety of applications. Advances in machine learning (ML) have produced algorithms that can now learn high-level abstractions via hierarchical models. This project aims to leverage advances in ML techniques and the available high-resolution simulation data to develop a novel modeling and simulation (M\&S) methodology for reactor safety analysis. While application-agnostic ML techniques are available, complex physics constraints need to be incorporated into ML techniques to build ML-based closures for computationally efficient predictive simulations. This project intends to develop a physics-guided data-driven multi-scale methodology for M\&S of advanced reactors. The project focuses on thermal fluid (T/F) phenomena, which play major roles in advanced reactor safety. Specifically, we propose a data-driven coarse-mesh turbulence model based on local flow features for the transient analysis of thermal mixing and stratification in a sodium-cooled fast reactor (SFR). The model has a coarse-mesh setup to ensure computational efficiency, while it is trained by fine-mesh computational fluid dynamics (CFD) data with Reynolds-averaged Navier-Stokes (RANS) turbulence model to ensure accuracy. Three different neural networks are developed and tested for loss-of-flow transients in the hot pool of SFR, i.e. the densely connected convolutional neural network (DCNN), long-short-term-memory network based on proper orthogonal decomposition (POD-LSTM), and the DCNN informed by LSTM (DCNN-LSTM). The performances of these three neural networks are evaluated based on baseline models. The DCNN-LSTM model has been chosen for further hyperparameter optimization. Furthermore, based on a simplified two-dimensional case, uncertainty quantification (UQ) of the developed ML-based closure are investigated with three methods, i.e. Monte Carlo dropout, deep ensemble, and Bayesian neural network. The developed ML-based turbulent viscosity closure relation based on deep ensemble is then integrated into the system analysis module SAM and serves as a term in the conservation equations. Such a SAM-ML based procedure guarantees that the obtained results are consistent with the physical constraints of the thermal-fluid system. The SAM-ML simulation on the same loss-of-flow transient showed comparable accuracy with the CFD simulation but with a much coarser mesh setup. Last but not least, the ML-based closure improvement with the support of higher-fidelity data from large eddy simulation (LES) is discussed. As a first step towards this direction, a baseline LES simulation is performed to obtain comparable data with RANS results. Based on the early results, future investigation on further improving the ML-based closure is discussed. We believe the developed approach that combines scientific machine learning with nuclear system analysis code can benefit the advanced reactor community as more accurate safety analyses will better characterize reactor safety margins and reduce licensing efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Composite Analysis for Low-Level Waste Disposal in the Hanford Site Central Plateau (FY 2020) (Rev. 2)

This document provides an updated Hanford Site composite analysis (CA). A Hanford Site CA was prepared and issued in 1998 (PNNL-11800) with an addendum provided in 2002 (PNNL-11800 Addendum 1). The CA was approved in 2002 (Frei, 2002) and has been maintained (DOE/RL-2000-29 and subsequent revisions) to support low-level waste disposal performance assessments (PAs) and disposal authorizations for facilities at the Hanford Site, including the following: Continued operation of the Environmental Restoration Disposal Facility (ERDF) and the 200 East and 200 West Low-Level Burial Grounds; Construction of the Integrated Disposal Facility (IDF); Forthcoming closure of tank residual waste systems such as Waste Management Area (WMA) C. The CA maintenance program resulted in a determination in 2015 (DOE/RL-2015-66) that the Hanford Site CA needed an update for the following reasons: While the initial Hanford Site CA has been maintained since 2001, the accumulation of basis changes reported in the annual summary reports over the succeeding 14 years merit evaluation in an updated analysis; The U.S. Department of Energy (DOE) Headquarters requested in a memorandum in 2015 (Gilbertson and Marcinowski, 2015) that “as soon as the relevant PAs are complete, the CA will be revised to account for all of the new information.” This updated Hanford Site CA accounts for the following new information: 1. Inclusion of a detailed Hanford Site baseline disposition that projects remedial activities through site closure. There have been significant changes through decision making in the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) process that were not available when the original CA was produced; 2. Inclusion of an updated inventory basis, new modeling capabilities, and new decisions reached in the associated record of decision (ROD) that was provided by issuance of DOE/EIS-0391 in fiscal year (FY) 2013. Development of a Hanford Site groundwater model from the baseline provided in a technical transfer of models for the Final Tank Closure and Waste Management Environmental Impact Statement (DOE/EIS-0391) commenced in FY 2014 and resulted in the plateau-to-river (P2R) groundwater flow model that is used in this updated Hanford Site CA; 3. Accounting for pump and treat systems, which were not evaluated in the initial CA. Pump and treat systems have had significant impact on groundwater flow system behavior, contaminant transport, and contaminant removal from Hanford Site groundwater. This process is accounted for in the use of the P2R model for this updated Hanford Site CA; 4. Inclusion of water-level data collected since the initial CA was completed. Data collected as the unconfined aquifer water levels continue to recede since the cessation of large liquid discharges in the late 1990s have led to marked improvement in understanding of the flow system for future conditions, particularly regarding northward flow potential in the critical Gable Gap area. The calibration in the latest version of the P2R model accounts for this information; 5. Inclusion of results from the revised ERDF PA completed in 2013 (WCH-5209) that account for updated inventory and expansion of the ERDF facility to about twice the size that was evaluated in the original CA; 6. Use of updated geoframeworks to provide the structural basis for numerical fate and transport models in the groundwater pathway of this updated CA. The geologic basis for groundwater models has continued to improve with additional data collection and interpretation with the creation and maintenance of the Hanford South Geoframework and the Central Plateau Vadose Zone Geoframework tools; 7. Incorporation of updated tank residual inventory estimates. Tank residual inventory estimates have improved with the incorporation of tank retrieval inventory data for those tanks that have completed retrieval. The CA inventory data package includes this updated information; 8. Incorporation of WMA C PA results. Two additional PAs for tank farm closure decisions are in preparation during the period required to prepare an updated Hanford Site CA: WMA C (FY 2016) and WMA A-AX (in preparation). The WMA C PA results are incorporated into the updated Hanford Site CA, and its grouted residuals model is used as the basis for a release model to account for the other tank farm systems modeling in this updated CA; 9. An update to the IDF PA was submitted in FY 2017 and has been reviewed and approved. The results of the IDF PA are incorporated into this updated CA; 10. Updated risk assessment scenarios. The risk assessment scenarios currently in use for Hanford Site CERCLA and Resource Conservation and Recovery Act of 1976 (RCRA) analyses differ from those evaluated in the initial Hanford Site CA. The representative person exposure scenario evaluated in this updated Hanford Site CA is consistent with recent PAs and CERCLA and RCRA analyses as the Hanford Site. This updated Hanford Site CA provides the following: A comparison of the updated Hanford Site CA all-pathway dose results with the performance measures during the compliance period, which is assumed to begin with site closure in calendar 2070 with the last scheduled disposal action and continue for 1,000 years postclosure (to calendar year 3069); A comparison of Hanford Site CA all-pathway dose results with the performance measures during the postcompliance period to address potential peaks beyond the compliance period. This is accomplished by evaluating dose in the period for 9,000 years following the compliance period (i.e., from calendar year 3070 to 12070).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Continued Development and Advanced Testing of DPC Filler Cements (on FY22 R&D and Demonstration Activities) (Progress Report)

Commercial generation of energy by nuclear power plants in the United States (U.S.) has produced thousands of metric tons of spent nuclear fuel (SNF), the disposal of which is the responsibility of the U.S. Department of Energy (DOE). Utilities typically utilize the practice of storing this SNF in dual-purpose canisters (DPCs). DPCs were designed, licensed, and loaded to meet Nuclear Regulatory Commission (NRC) requirements that preclude the possibility of a criticality event during SNF storage and transport, but were not designed or loaded to preclude the possibility of a criticality event during the regulated post-closure period following disposal, which could be up to 1,000,000 years (Price, 2019). There are several options being investigated that could facilitate the disposal of SNF stored in DPCs in a geologic repository (Hardin et al., 2015; SNL 2020b; SNL 2021b). These include: (1) repackage the SNF into canisters that are designed to prevent criticality during the regulated post-closure period following disposal, but with an increased disposal cost estimated at approximately $\$$20B in United States dollars (USD) (Freeze et al., 2019); (2) analysis of the probability and consequences of criticality from the direct disposal of DPCs during a 1,000,000-year post-closure period in several geologic disposal media (Price, 2019); and (3) filling the void space of a DPC with a material before its disposal that significantly limits the potential for criticality over the post-closure regulatory period. This report further investigates the third option, filling DPC already containing SNF with a material to limit the potential for criticality over the post-closure regulatory period.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 114: Area 25 EMAD Facility Nevada National Security Site, Nevada

This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the actions needed to achieve closure for Corrective Action Unit (CAU) 114, Area 25 EMAD Facility, identified in the Federal Facility Agreement and Consent Order (FFACO). CAU 114 comprises the following corrective action sites (CASs) located in Area 25 of the Nevada National Security Site: • 25-41-03, EMAD Facility (Building 3900) • 25-99-23, Manned Control Car (MCC) and Engine Installation Vehicle (EIV) • 25-33-05, Building 3901, Engine Transport System Maintenance Building (Train Shed) This plan provides the methodology for field activities needed to gather the necessary information for closing CAU 114. There is sufficient information and process knowledge from historical documentation and investigations of similar sites regarding the expected nature and extent of potential contaminants to recommend closure of CAU 114 using the SAFER process. Additional information will be obtained by conducting a field investigation before selecting the appropriate corrective actions for CAU 114. It is anticipated that the results of the field investigation and implementation of corrective actions will support a defensible recommendation that no further corrective action is necessary. The purpose of the corrective action investigation will be to document and verify the adequacy of existing information; to affirm the decision for either clean closure, closure in place, or no further action; and to provide sufficient data to implement the corrective action. The actual corrective action selected will be based on characterization activities implemented under this SAFER Plan. If specific conditions or findings fall outside the bounds of the conceptual site model, such as an unanticipated release, the Nevada Division of Environmental Protection (NDEP) will be consulted to determine the path forward before proceeding. Upon completion of SAFER activities, a closure report (CR) will be prepared and submitted to NDEP for review and approval. The schedule for completion of the CR will be established in consultation with NDEP.

54 ENVIRONMENTAL SCIENCES↗

E-Area Low-Level Waste Facility Inadvertent Human Intruder Limits and Doses in Support of the PA2022

This report documents the inadvertent human intruder (IHI) analysis for the E-Area Low-Level Waste Facility (ELLWF) at the Savannah River Site (SRS), near Aiken, South Carolina. This analysis supports the revised ELLWF Performance Assessment (PA), complying with the Department of Energy standard for operation of low-level waste disposal facilities (USDOE, 2017). The ELLWF is an operating waste disposal facility and is scheduled to continue accepting waste to 2065. One task of the revised PA is to establish waste inventory limits for the various disposal units at ELLWF. This is done by modeling future contaminant release and transport through applicable pathways to human receptors, comparing predicted doses per disposed curie with applicable performance measures, to obtain inventory limits which will assure that doses to receptors do not exceed performance measures. This report documents results of modeling future doses to one class of receptor, the inadvertent human intruder. It is assumed that after site closure, public knowledge of the site is lost, and IHIs will engage in activities on the ELLWF that will disrupt the closure cap, causing dose to the IHI. Following USDOE (2017), six different stylized exposure scenarios are considered, simulating activities by an IHI which could result in a radiological dose. The six scenarios are: • Acute – Basement Construction: IHI constructs a basement and encounters waste during excavation which is inadvertently mixed with clean soil and diluted. • Acute – Well Drilling: IHI drills a water well through waste and is exposed to drill cuttings mixed with clean soil that are brought to the surface. • Acute – Discovery: IHI begins constructing a basement but stops when encountering the riprap in the final closure cap and is exposed to photon radiation from unexcavated material residing in the undisturbed waste zone. • Chronic – Agriculture: Resident IHI is exposed to waste that was excavated for basement construction and mixed with native soil in the intruder’s vegetable garden. • Chronic – Post-Drilling: Resident IHI is exposed to waste from drill cuttings mixed with native soil and scattered in the garden area. • Chronic – Residential: Resident IHI is exposed to external radiation while in home located above waste with shielding provided by the concrete basement floor and any soil or engineered material remaining between the basement and waste. Dose calculations are performed using the SRNL Dose Toolkit (Aleman, 2023), following the approach of Smith et al (2019). Calculations are performed separately for 27 of the 33 disposal units (DUs) at ELLWF and are radionuclide specific. The results of the IHI analysis include: • Dose Factors: mrem per disposed curie (acute) and mrem/yr per disposed curie (chronic) for each parent radionuclide, for each DU. • Inventory Limits: in curies, for each parent radionuclide, for each DU. • Estimated Dose to IHI: mrem (acute) and mrem/yr (chronic), for each DU, given its projected closure inventory without inventory biases applied. Most DU-specific IHI inventory limits are in the range of 10 3 to 10 7 curies per nuclide. The lowest inventory limits are associated with gamma-emitters such as Sn-126, Ra-226, Th-232, and Cm-248. Radionuclides with short half-lives such as Pu-241, and nuclides which are pure beta emitters or which decay by electron capture, such as Ni-59 and Ni-63, have the highest limits. For the 27 evaluated DUs, predicted IHI doses are shown in Table ES-1. The maximum acute dose is 1.18 mrem, at ST23, much less than the DOE performance measure of 500 mrem (USDOE, 2017). The highest chronic dose is 37.2 mrem/yr at ST02, below the DOE performance measure of 100 mrem/yr. Also shown are estimated inventory sums of fractions (SOFs) at closure in 2065, for groundwater (GW) and IHI pathways. For each DU, the inventory is constrained by the GW pathway. For most DUs, the IHI SOFs are approximately 1000 times lower than the GW SOF values, and the IHI pathway does not drive risk for any disposal unit.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

On the modelling of non-reactive and reactive turbulent combustor flows

A study of non-reactive and reactive axisymmetric combustor flows with and without swirl is presented. Closure of the Reynolds equations is achieved by three models: kappa-epsilon, algebraic stress and Reynolds stress closure. Performance of two locally nonequilibrium and one equilibrium algebraic stress models is analyzed assuming four pressure strain models. A comparison is also made of the performance of a high and a low Reynolds number model for combustor flow calculations using Reynolds stress closures. Effects of diffusion and pressure-strain models on these closures are also investigated. Two models for the scalar transport are presented. One employs the second-moment closure which solves the transport equations for the scalar fluxes, while the other solves the algebraic equations for the scalar fluxes. In addition, two cases of non-premixed and one case of premixed combustion are considered. Fast- and finite-rate chemistry models are applied to non-premixed combustion. Both show promise for application in gas turbine combustors. However, finite rate chemistry models need to be examined to establish a suitable coupling of the heat release effects on turbulence field and rate constants.

Nikjooy, Mohammad↗

Near-wall modelling of compressible turbulent flows

Work was carried out to extend the near-wall models formulated for the incompressible Reynolds stress equations to compressible flows. The idea of splitting the compressible dissipation function into a solenoidal part that is not sensitive to changes of compressibility indicators and a compressible part that is directly affected by these changes is adopted. This means that all models involving the dissipation rate could be expressed in terms of the solenoidal dissipation rate and an equation governing its transport could be formulated to close the set of compressible Reynolds stress equations. The near-wall modelling of the dissipation rate equation is investigated and its behavior near a wall is studied in detail using k-epsilon closure. It is found that all existing modelled equations give the wrong behavior for the dissipation rate near a wall. Improvements are suggested and the resultant behavior is found to be in good agreement with near-wall data. Furthermore, the present modified k-epsilon closure is used too calculate a flat plate boundary layer and the results are compared with four existing k-epsilon closures. These comparisons show that all closures tested give essentially the same flow properties, except in a region very close to the wall. In this region, the present k-epsilon closure calculations are in better agreement with measurements and direct simulation data; in particular, the behavior of the dissipation rate.

So, Ronald M. C.↗

Analytical and Preliminary Simulation Study of a Pilot's Ability to Control the Terminal Phase of a Rendezvous with Simple Optical Devices and a Timer

One method of controlling the terminal phase of a space rendezvous between two vehicles is first to correct the flight path of the controlled vehicle so that a constant line of sight is established between the vehicles. This correction is accomplished by thrusting normal to the sight line in a direction to arrest the angular motion of this line. Once this collision course has been established, the second step is to control the closure rate for a safe approach along the line of sight. Adequate control of this maneuver requires range and closure-rate information. A combined analytical and preliminary simulation study was conducted to determine the ability of a human pilot to control the rendezvous by this method using visual sightings made during the initial collision-course control to obtain the range and closure rate. The analytical phase of the study reviewed the geometric relations between the vehicles and formed the basis for techniques to transform the angular sightings into range and closure rate. A preliminary simulation was then made to investigate the accuracy of these techniques. The simulation consisted of an analog computer, an oscilloscope to represent the view a pilot would have-with a stabilized sight, and a timer. Results indicate that pilots, using an optical sight and a timer, can successfully arrest the angular motion of the line of sight between two rendezvous vehicles and obtain relative range and closure rate with sufficient accuracy to perform the final braking maneuver successfully.

Lineberry, Edgar C., Jr.↗

Identifying and Closing Medical Capability Gaps for Human Spaceflight Missions Beyond Low Earth Orbit

BACKGROUND: Features of human space missions beyond low Earth orbit such as increasing distance from Earth, lack of real-time communication, and limited or no evacuation or resupply capability are expected to drive an increase in medical risk and require crews to operate in an increasingly autonomous fashion. A diverse set of stakeholders at NASA are seeking to fund the development of concepts of operations, medical requirements, and medical capabilities for such missions. However, a systematic approach to identification of current medical capability gaps and a strategic framework to gap closure is needed. OVERVIEW: The Exploration Medical Integrated Product Team (XMIPT) has developed a list of nine high-level medical capability gaps and associated activities required for gap closure. The list was derived based on inputs from subject matter experts at NASA including flight surgeons, other clinical providers, as well as operational and research communities regarding medical capabilities required to support human missions to the Moon and the Mars surface. Responses were reviewed and distilled to identify common themes across capabilities. To ensure alignment with established human system risks, the gap list was further refined based on inputs from NASA’s Human System Risk Board. Relevant medical gap closure activities outside of those funded by the XMIPT were identified through solicitation of inputs from Elements of NASA’s Human Research Program (particularly Exploration Medical Capability), the broader medical operations community, and other stakeholders. This medical capability gap list is reviewed and updated regularly as new information becomes available or new stakeholders are identified. DISCUSSION: The medical capability gap list has matured to include a large group of NASA stakeholders and development activities. This has enabled articulation of priorities to funding entities and programmatic stakeholders, while serving as an accessible resource summarizing gap closure activities, relevant programmatic infusion points, and opportunities for collaboration between stakeholders. This presentation will provide an overview of the nine NASA medical capability gaps and their associated gap closure activities.

Moriah Thompson↗

NRC Consultation and Monitoring at the Savannah River Site: Focusing Reviews of Two Different Disposal Actions - 12181

Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA) requires the U.S. Department of Energy (DOE) to consult with the U.S. Nuclear Regulatory Commission (NRC) for certain non-high level waste determinations. The NDAA also requires NRC to monitor DOE's disposal actions related to those determinations. In Fiscal Year 2011, the NRC staff reviewed DOE performance assessments for tank closure at the F-Tank Farm (FTF) Facility and salt waste disposal at the Saltstone Disposal Facility (SDF) at the Savannah River Site (SRS) as part of consultation and monitoring, respectively. Differences in inventories, waste forms, and key barriers led to different areas of focus in the NRC reviews of these two activities at the SRS. Because of the key role of chemically reducing grouts in both applications, the evaluation of chemical barriers was significant to both reviews. However, radionuclide solubility in precipitated metal oxides is expected to play a significant role in FTF performance whereas release of several key radionuclides from the SDF is controlled by sorption or precipitation within the cementitious wasteform itself. Similarly, both reviews included an evaluation of physical barriers to flow, but differences in the physical configurations of the waste led to differences in the reviews. For example, NRC's review of the FTF focused on the modeled degradation of carbon steel tank liners while the staff's review of the SDF performance included a detailed evaluation of the physical degradation of the saltstone wasteform and infiltration-limiting closure cap. Because of the long time periods considered (i.e., tens of thousands of years), the NRC reviews of both facilities included detailed evaluation of the engineered chemical and physical barriers. The NRC staff reviews of residual waste disposal in the FTF and salt waste disposal in the SDF focused on physical barriers to flow and chemical barriers to radionuclide release from the waste. Because the waste inventory and concentration at both sites is sufficient to generate unacceptable doses to an off-site member of the public or inadvertent intruder in the absence of engineered barriers, the NRC staff review focused on the engineering features DOE plans to put in place to limit radionuclide release. At the FTF, DOE expects that peak doses are delayed beyond a 10,000 year performance period by a combination of (1) the flow-limiting effect of the steel tank liner and (2) chemical conditions created by the stabilizing grout overlying the waste that limit the solubility of key radionuclides for tens of thousands of years. At the SDF, DOE expects that flow will be significantly limited by water shedding along the closure cap lower drainage layer and that radionuclide release will be further limited by radionuclide precipitation or sorption within the high pH, chemically reducing conditions created within the saltstone waste form. Because the performance of both facilities depends on the performance of engineered barriers for thousands of years, the reviews included a detailed evaluation of the expected long-term behavior of these barriers. As previously discussed, NRC staff reviews of DOE waste determinations during consultation are designed to evaluate the three NDAA criteria, whereas the review of an updated PA during monitoring only addresses whether the NRC staff has reasonable assurance that the planned disposal action will meet the performance objectives of 10 CFR Part 61. The NRC staff review of the Waste Determination for the FTF did not include conclusions about whether the planned disposal of residual waste at the FTF would meet the NDAA criteria because of the substantial uncertainties in the degree of waste removal DOE would achieve and other technical uncertainties. The main product of the NRC staff review of the planned FTF disposal action is the recommendation that DOE should conduct waste release experiments to increase support for key modeling assumptions related to: (1) the evolution of pH and Eh in the grouted tank system over time; (2) identification of HRR association with solid phases comprising the residual wastes; and (3) expected solubility of HRRs under a range of environmental or service conditions that the residual wastes in the contaminated zone are expected to be exposed to over time. Implementation of this recommendation is deemed crucial for NRC staff to have reasonable assurance that the performance objectives in 10 CFR Part 61, Subpart C can be met. Given the risk-significance of Tank 18 to the overall PA and the short timeline for closure of this tank, the NRC staff recommended that DOE should initiate discussions with NRC staff regarding implementation of this recommendation for Tank 18 as soon as practical. The NRC staff also recommended that experiments to address this recommendation should be conducted prior to final closure of Tank 18. Results of the Tank 18 residual waste experiments, if conducted, will be evaluated by NRC staff to determine the need for additional data collection, experiments, and modeling for Tank 18, as well as other FTF tanks. Additional information regarding the NRC staff's recommendations in this area, including details on the suggested implementation of other recommendations will be provided in the NRC staff's plan for monitoring the FTF later in FY 2012, after DOE makes a final decision on the waste determination. The NRC staff's review of waste disposal at the SDF is ongoing. When complete, the SDF TER will indicate whether the NRC staff continues to have reasonable assurance that waste disposal at the SDF will meet the performance objectives of 10 CFR Part 61 (NDAA Criterion 3). The TER also will include risk insights that will form the basis of the NRC staff's revised monitoring plan for the SDF. The NRC staff will publish an updated monitoring plan for the SDF later in FY 2012. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗