Outcomes from the Brine Availability Test in Salt (BATS) Heater Test at WIPP
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Mills, Melissa Marie [Sandia National Laboratories↗
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Performance Assessment Supporting the WIPP Replacement Panels Planned Change Request
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King, Seth John [Sandia National Laboratories (SNL↗
White Paper on Lithium-Ion Battery Safety for use in Safety Planning at the Waste Isolation Pilot Plant
The Waste Isolation Pilot Plant (WIPP) is an underground facility designed to safely dispose of radioactive waste. The WIPP uses many heavy vehicles to transport materials and equipment underground. Most of these vehicles are powered by traditional internal combustion engines (ICE) with diesel fuel. Recently, electric vehicles (EVs) powered with batteries have been used at the WIPP. EVs have very low operational and maintenance costs, not considering battery replacements, and they have zero emissions during operation. This absence of emissions makes them ideal for underground facilities with limited ventilation. Even if a facility has robust ventilation normally, ventilation systems can break down leading to restrictions in ICE powered operations. Figure 1 shows a rendering of the WIPP.
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Solidification of SRPPF Aqueous Recovery Liquid: Process Disruptions
Savannah River National Laboratory has identified a magnesium oxysulfate grout formulation to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). The formulation uses reactive, light burnt MgO, anhydrous MgSO 4 , and dead burnt MgO and provides good mixability, similar density to the original Portland Cement-based mix, and a leachate pH of 9.39, within the assumed Waste Isolation Pilot Plant (WIPP) brine pH range. The formulation will solidify the effluent using a solid first, lost paddle design, within a 55-gallon drum. A series of seven different process disruptions, or irregular processing during the expected ARS liquid effluent solidification process, were completed to assess the severity of the impact of the process disruption and how the formulation behaves under these conditions. Seven process disruptions were investigated including: No Mixing, Partial Mixing, Compacted/Stratified, Sealed, Overfilled, Underfilled, and Undermixed. The impact on solidification, leachate pH, Er (a Pu/Am surrogate) distribution, and density of the resulting grout were established. Overall, four of the process disruptions, No Mixing, Partial Mixing, Compacted/Stratified, and Overfilled, resulted in basic, unabsorbed liquid that would require additional solidification to be acceptable for WIPP. Most solids generated in this testing met the expected WIPP leachate pH requirements. Solids from the Compacted/Stratified process disruption test, however, produced a leachate pH slightly above the assumed range in the WIPP performance assessment. The distribution of Er for most of the tests was relatively consistent. The No Mixing process disruption test, however, resulted in non-homogeneous Er distribution, indicating that radioactive material would not be evenly distributed throughout this solid’s matrix. The Overfilled test is the only mix with a density slightly below the acceptable value. The Sealed and Overmixed tests produced a solid closest to the standard mix with no unabsorbed liquid, expected leachate pH levels, consistent Er concentrations, and expected density, indicating these process disruptions do not have a negative impact on the final grout form. Based on the other process disruptions, it is critical to have premixed dry materials and a well-defined mixing process to produce a solid, homogeneous magnesium oxysulfate cement, with no unabsorbed liquid. These process disruption tests show that the lack of adequate mixing is fundamental to creating an improperly solidified material that may require remediation.
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Improbability of Post-Closure Criticality in Array of Transuranic Waste Containers after Compaction by Salt Creep at Waste Isolation Pilot Plant
Based on the rationale presented, post-closure nuclear criticality is improbable when room closure compacts containers disposing transuranic (TRU) waste emplaced at the Waste Isolation Pilot Plant (WIPP), an operating repository in bedded salt in southeastern New Mexico. As described in the original WIPP certification, a qualitative estimate of the probability of post-closure criticality in TRU waste produced from atomic energy defense activities has been low either because remote-handled TRU waste canisters are neutronically isolated by the bedded salt or because the low fissile mass in an array of contact-handled TRU waste drums cannot be compacted sufficiently by room closure from salt creep. These situations are still valid for the majority of TRU waste that is disposed at WIPP without any disposal constraints, as updated herein. This report also qualitatively evaluates the probability of criticality after disposal of TRU waste in pipe overpack containers (POC) where every POC in a shipment may have the maximum 200 fissile gram equivalent of 239Pu content. The probability of criticality for a disposal room filled with POCs is estimated during four representative phases of repository evolution: (1) a large salt block falls onto POCs in the first 20 years, (2) salt creep closes a disposal room in the first 1000 years without brine seepage and subsequent gas generation, which permits maximum compaction, (3) some brine seepage occurs into the closed room, which initiates consumption of the fiberboard (cellulose) impact absorber in the POC in the second 1000 years, and (4) full brine inundation of a room and consumption of all fiberboard thereafter. Salt-block fall in the first phase does not greatly disrupt three tiers of POCs. The compacted spacing of POCs in the later three repository conditions is calculated through high-fidelity, geomechanical modeling. Criticality evaluation of compacted 200-g 239 Pu spheres at the compacted spacing shows that neither 12-inch nor 6-inch POCs are critical after the first 1000 years, the second 1000 years, or thereafter as the sea of reflector material changes to represent the three repository conditions. Specifically, fiberboard and iron isolates 239 Pu while dry, and brine reduces the reactivity when a room is partially and fully inundated. Because POC behavior bounds behavior of other standard TRU waste containers, post-closure criticality caused by room closure compacting containers is omitted in the performance assessments for the 2019 and 2026 WIPP compliance re-certification applications to the US Environmental Protection Agency.
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RH-TRU Waste Streams: HFDA Crucibles Create New Complex Waste Stream
This project focuses on the removal of the Hot Fuel Dissolution Apparatus (HFDA) crucibles from the Hot Fuel Examination Facility’s (HFEF) hot cells to create space for upcoming projects. The project team has created a crucible and pyro-salt non-debris waste stream that can be disposed of at Waste Isolation Pilot Plant (WIPP). Chemical Compatibility Examinations (CCEs) must be performed on candidate waste by reviewing the types of chemicals used in the experiments, information found in INL Process Knowledge Summary Reports (PKSRs), and Environmental Protection Agency guides on chemical compatibility. The CCEs identify which chemicals can be safely combined to ensure adverse reactions do not occur in the waste packages. The CCEs are then used to update the PKSR, Basis of Knowledge, and Acceptable Knowledge documents and demonstrate compliance with the WIPP Waste Acceptance Criteria. Once documents are finalized, they are reviewed and approved by the WIPP Central Characterization Project (CCP) to ensure proper waste stream characterization. After the documents are approved, the crucibles can be packaged and shipped to the Idaho Nuclear Technical and Engineering Center for storage awaiting certification. CCP will certify the waste using the approved documents and a series of confirmatory testing before being shipped and disposed in WIPP’s deep geological repository. Currently, the CCE is in progress and the PKSRs are being updated.
Basis of Knowledge↗
Long-term sinking of nuclear waste canisters in salt formations by low-stress creep at high temperature
Abstract Rock salt has a self-sealing capacity, low permeability, and high thermal conductivity, making it a potential host for heat-generating nuclear waste. The feasibility of nuclear waste disposal within salt formations has been investigated mostly for small-sized canisters. Geologic disposal of larger-sized canisters originally designed for spent fuel storage and transportation has lately been examined as a cost-effective alternative. This raises questions about their long-term vertical movement due to their weight and high decay heat. Low-stress creep governs this movement; however, most salt constitutive models do not incorporate it. In this paper, the Norton and the WIPP creep models are compared with the Lux/Wolters/Lerche (LWL) model and a simpler model that combines linear and Norton creep laws (named combined creep model). The LWL and combined creep models consider pressure solution creep, though all incorporate dislocation creep. The models are first applied to creep tests under various stress levels. The LWL and the combined creep models results fit the experimental data well in both high and low stress ranges, whereas the Norton and WIPP models results only fit in higher-stress ranges. The different models are further applied for analyzing long-term canister movement. A sinking rate of $$-4.4\times 10^{-7}$$ - 4.4 × 10 - 7 mm/year was predicted using the Norton and WIPP models versus $$-2.1\times 10^{-2}$$ - 2.1 × 10 - 2 mm/year and $$-3.1\times 10^{-2}$$ - 3.1 × 10 - 2 mm/year using the LWL and the combined creep models, respectively. This comparative study confirms that creep models calibrated exclusively against high-deviatoric stress data might result in an inaccurate estimation of waste packages sinking rate in salt formations.
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Post-closure Nuclear Criticality Safety Evaluations for Disposition of Criticality Control Overpacks at the Waste Isolation Pilot Plant
The Waste Isolation Pilot Plant (WIPP) is a geological repository in southern New Mexico that provides for disposal of transuranic (TRU) wastes from atomic energy defense activities. The Sandia National Laboratories (Sandia) Report, Consideration of Nuclear Criticality When Disposing of Transuranic Waste at the Waste Isolation Pilot Plant, addresses nuclear criticality safety based on the projected inventory characteristics for the initial compliance certification application of WIPP in 1996. As the inventory, waste forms, and disposal package designs change, revised or new analyses are necessary to demonstrate acceptability for these configurations within the WIPP safety basis and compliance with 10,000-year post-closure standards of the US Environmental Protection Agency (EPA). Saylor and Scaglione evaluated criticality control overpacks (CCOs) in 2017 based on conservative assumptions for post-closure repository structural conditions with resulting effects on containers and container spacing, The Saylor and Scaglione evaluation of CCOs addressed a single waste configuration that represents the Surplus Plutonium Disposition Program’s dilute and dispose waste form and composition. This initial CCO study demonstrated that 50 grams of boron carbide (B 4 C) per CCO is sufficient to ensure post-closure criticality safety based on a well-mixed waste composition, and Oak Ridge National Laboratory (ORNL) subsequently determined that this amount of B 4 C does not require constraints on moisture or plastic present as moderator. The Saylor and Scaglione analysis conservatively assumes repository room closure that eliminates all space between fissile gram equivalent (FGE) 239 Pu masses. The close-packed array was selected based on limited availability of repository salt creep modeling results at that time. In 2019, Brickner provided additional evaluations for pipe overpack containers (POCs), building on the conservative basis provided by Saylor and Scaglione. Brickner’s 2019 analysis made use of new geomechanical data for post-closure spacing that rely on advances in repository modeling as documented in the work by Reedlunn and Bean. This current CCO evaluation for generic waste materials expands on earlier work performed at ORNL and includes evaluation of CCOs across a much broader range of possible waste compositions and geometries. This evaluation is intended to provide input for the required feature, event and process (FEP) screening to determine if post-closure criticality must be included as an event in the 10,000-year regulatory evaluation. As such, the approach to modeling post-closure criticality presented in this report has been coordinated with the Sandia team responsible for FEP screening. The resulting analysis supports disposition of fissile materials in the CCO containing up to 380 FGE 239 Pu and expands conditions acceptable for disposal of fissile material in CCOs. This evaluation builds on the methodology of Saylor and Scaglione and Brickner, using the most recently available geomechanical data for CCO spacing under salt creep compaction scenarios provided by Reedlunn and Bean. The broad range of fissile material configurations analyzed in this report are intended to account for configurations that may occur during the post-closure disposal time period, and it also includes waste configurations that are not physically possible to support analysis of conditions that influence neutron fluence.
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Solubility of An(OH) 3 am and An(OH) 3 cr in the presence of Citrate and EDTA
The solubility of An(III), which pertains to Pu(III), Am(III), and Cm(III), in brine is an important contributor in WIPP (Waste Isolation Pilot Plant) performance assessment (PA) models to calculate the potential release of transuranic elements in the near-field environment of the WIPP repository. Americium, although present in much smaller quantities than plutonium (~ 336 kg on year 2033 is potentially WIPP-bound primarily as Am-241), has a significant impact on release due to its higher activity during the earlier times of repository history (~26% initially, decreasing to 17% and ~0% at 1000 and 10,000 years after emplacement) (SOTERM, 2019). Lanthanide analogues such as Nd 3+ possess physical and chemical characteristics that allow them to be used to examine the chemical behavior of trivalent actinides (Lucchini, et al., 2007).
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Waste Control Specialists Technical Review Team Report
1. Modeling of Drum 68660 and the WCS Drums: A calibrated model of Los Alamos National Laboratory (LANL) Drum 68660 indicates that pressurization of the drum by restriction of drum venting could have led to the thermal runaway reaction at the Waste Isolation Pilot Plant (WIPP) in 2014, supporting the hypothesis that the contents of the drum were not fundamentally different from the overall remediated nitrate salt (RNS) waste stream. 2. Nitric Acid Reactions and Aging of the Waste: The stability of the WCS RNS waste with respect to autocatalytic thermal runaway due to nitric acid chemistry should increase over time as the nitric acid is depleted from chemical reactions in the drum. However, this cannot be confirmed because of the lack of long-term experimental data on the impact of aging on the reactivity of RNS waste. The calibrated model indicates that the waste drums retain most of their reactive components, metal nitrate salts and sWheat Scoop ® , even after eight years. 3. Pressure Effects and Runaway Reactions: The calibrated model indicates that if the drums and standard waste boxes (SWBs) are not allowed to pressurize during transport or storage and they are stored at historical WIPP repository temperatures, then an autocatalytic thermal runaway event is not expected to occur. Since the cause of Drum 68660 pressurization is not known, decision makers are not able to tailor mitigating strategies to the initiating event. Consequently, mitigating strategies must consider all feasible internal and external events that could lead to the pressurization of the drums. 4. Temperature Effects and Runaway Reactions: Temperature also has a strong influence on the stability of the waste. The calibrated model indicates that cooling the drums or SWBs can reduce the possibility of autocatalytic thermal runaway reactions. Simulations suggest that for a drum with a plugged vent, the probability of thermal runaway may be reduced or eliminated if wastes are maintained at or below 43 °F through active cooling measures from the initiation of transport until emplacement at WIPP.
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Annual Transuranic Waste Inventory Report – 2021 (Data Cutoff Date 12/31/2020)
The purpose of this Annual Transuranic Waste Inventory Report (ATWIR) – 2021 is to document the inventory estimate of transuranic (TRU) waste reported by the TRU waste generator sites as of December 31, 2020. This report also notes major changes to the inventory used for the ATWIR-2020 that had a data cutoff date of December 31, 2019. This updated inventory information is available to the U.S. Department of Energy (DOE) TRU waste complex, Waste Isolation Pilot Plant (WIPP) stakeholders, and regulators. The TRU waste inventory information is used for strategic planning, and supports the DOE Carlsbad Field Office (CBFO) input into documents (e.g., WIPP Documented Safety Analysis and National Environmental Policy Act analyses), planned changes, and other design changes as needed for the WIPP project.
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Annual Transuranic Waste Inventory Report – 2022 (Data Cutoff Date 12/31/2021)
The purpose of this Annual Transuranic Waste Inventory Report (ATWIR) – 2022 is to document the inventory estimate of transuranic (TRU) waste reported by the TRU waste generator sites as of December 31, 2021. This report also notes major changes to the inventory since the ATWIR-2021, which had a data cutoff date of December 31, 2020. This updated inventory information is available to the U.S. Department of Energy (DOE) TRU waste complex, Waste Isolation Pilot Plant (WIPP) stakeholders, and regulators. The TRU waste inventory information is used for strategic planning, and supports the DOE Carlsbad Field Office (CBFO) input into documents (e.g., WIPP Documented Safety Analysis and National Environmental Policy Act evaluations), performance assessments, planned changes, and other design changes as needed for the WIPP project.
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Development of Magnesium Oxysulfate Formulation for SRPPF Aqueous Recovery Liquid Solidification
The liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery Processes will be solidified into a stable form that is acceptable by Waste Isolation Pilot Plant (WIPP) for disposal. The current Aqueous Recovery flow sheet proposes to solidify the liquid effluent using a grout formula that was developed and tested for the former Waste Solidification Building process. This Portland cement based mixture results in a high pH (~13) leachate from the solidified waste form which is not acceptable to WIPP in the large quantities expected from production at SRPPF. Various cementitious materials were previously evaluated as alternative grout formulations to Portland cement and a MgO-based mix was identified as a promising alternative. A magnesium oxysulfate (MOS) cement formulation comprised of reactive magnesium oxide (MgO), anhydrous magnesium sulfate (MgSO 4 ), and sand, as a non-reactive heat sink provided good mixability, similar density to the original Portland-cement based mix, and a leachate pH of 9.4, within the assumed WIPP brine pH range. However, the MOS formulation exhibited an appreciable amount of heat generation, which resulted in premature setting of a large-scale test.
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DECOVALEX-2023: Task E Final Report
This is the Task E final report for DECOVALEX-2023. Task E is focused on understanding thermal, two-phase hydrological, and mechanical (TH2M) processes, especially related to predicting brine migration in the excavation damaged zone around a heated excavation in salt. Salt is attractive as a disposal medium for radioactive waste because it is self-healing and is essentially impermeable and essentially non-porous in the far field (away from excavations). Investigation of the short-term (days to years) near-field (centimeters to tens of meters) behavior of salt is important for radioactive waste disposal because this early period strongly controls the amount of brine in a salt repository. Brine leads to corrosion of waste forms and waste packages, and possible dissolution of radionuclides with brine transport being a potential transport vector to the accessible environment. The main test case used in Task E is the ongoing Brine Availability Test in Salt (BATS) heater test located underground at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, USA. The Task was divided into a series of Steps. Step 0 was an introduction to processes in salt, that included matching historical unheated brine inflow data from boreholes at WIPP and matching temperature observations during BATS heater test 1a. Step 1 included validation of models against a thermo-poroelastic analytical solution relevant to heated boreholes in salt, and two-phase flow around an excavation in salt. Step 2 required all the individual components covered in steps 0 and 1 to come together to match observed brine inflow behavior during the BATS 1a heater test. There were a range of approaches from the teams, from mechanistic to prescriptive. Given the uncertainties in the problem, some teams used one- or two-dimensional models of the processes, while other teams included more geometrical complexity in three-dimensional models. The key learning points from Task E have been: • Heat conduction through salt typically requires non-linear thermal conductivity (as a function of temperature), but most models do a good job matching observations, given appropriate adjustments to the applied power and some thermocouple locations. • Thermal pressurization requires coupled thermal-hydrological-mechanical (THM) responses that consider the thermal expansion of the fluid and solid phases. • Initialization of two-phase flow models around a borehole or excavation in salt are more realistically represented as “wetting up”, rather than “drying down” (i.e., the initial state after excavation is mostly dry, rather than mostly wet). • The BATS 1a heater test includes a significant release of brine after the end of heating, which requires a large increase in permeability to recreate. Task E has been a great learning experience for all the teams involved, and feedback from the modeling teams has led to changes in the design of follow-on BATS experiments, which are now ongoing underground at WIPP. There was a balance throughout the task between freedom to model phenomena how each team saw fit, and prescriptiveness in problem design to bring the modeling teams closer together to allow attribution of smaller differences between models to different modeling choices. The modeling approaches seem to go through two phases: an early phase of discovery or testing, and a later phase of refinement and improvement. In future modeling efforts, different field data could be used (e.g., BATS 2) and more time should be included in the processes for teams to make multiple model refinement or even significant changes to their conceptual model or setup, based on lessons learned from the modeling exercise.
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Microbial Influence on the Mobility of +3 Actinides from a Salt-Based Nuclear Waste Repository
Biologically enhanced transport of radionuclides is one of several processes that can affect the performance of a nuclear waste repository. In this work, several microbial isolates from the Waste Isolation Pilot Plant (WIPP) were tested for their influence on the concentration of neodymium, as an analog for +3 actinides, in simple sodium chloride solutions and in anoxic WIPP brines. Batch sorption experiments were carried out over a period of 4–5 weeks. In many cases, the effect on neodymium in solution was immediate and extensive and assumed to be due to surface complexation. However, over time, the continued loss of Nd from the solution was more likely due to biologically induced precipitation and/or mineralization and possible entrapment in extracellular polymeric substances. The results showed no correlation between organism type and the extent of its influence on neodymium in solution. However, a correlation was observed between different test matrices (simple NaCl versus high-magnesium brine versus high-NaCl brine). Further experiments were conducted to test these matrix effects, and the results showed a significant effect of magnesium concentration on the ability of microorganisms to remove Nd from solution. Possible mechanisms include cation competition and the alteration of cell surface structures. This suggests that the aqueous chemistry of the WIPP environs could play a larger role in the final disposition of +3 actinides than the microbiology.
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Nevada National Security Site V3XA Spherical Explosive Confinement Vessel Awareness Package
This report summarizes the plans and activities that will facilitate shipment of the V3XA experimental spheres (a.k.a., Nevada Spheres) from the Nevada National Security Site (NNSS) to the Idaho National Laboratory’s (INL) Advanced Mixed Waste Treatment Project (AMWTP). The spheres require shipment to INL to be segmented, characterized, and repackaged for ultimate disposal at the Waste Isolation Pilot Plant (WIPP). The NNSS does not currently have the facilities or safety basis to perform these activities. The INL has the facilities, the INL M&O contractor owns an 8-120B cask, and AMWTP has a WIPP Waste Acceptance Criteria approved characterization program. The NNSS has shipped other legacy transuranic waste to INL for WIPP certified characterization in the past.
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Biomass and salt-dependent effects of Bacillus spores on radionuclide migration from the Waste Isolation Pilot Plant
Spores of a Bacillus sp., isolated from radioactive waste, were tested for their ability to influence the fate and transport of neodymium (Nd3+) under high salt conditions expected at the Waste Isolation Pilot Plant (WIPP) nuclear waste repository. Spores were suspended in neodymium-spiked saline solutions up to 4 M NaCl, and concentrations of Nd and the complexing agent dipicolinic acid (DPA), a component of spores, were monitored along with optical densities and spore numbers. Overall, results support neodymium bioassociation that is dependent upon biomass, with more apparent adsorption occurring at higher spore concentrations. However, probable spore lysis in 2 and 4 M NaCl solutions and possible germination at 0.15 M NaCl appear to drive the release of DPA and subsequent return of Nd to solution. The implications of this work for the WIPP will depend on actual biomass levels and the ionic strength of the repository brines.
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Linear and Nonlinear Solvers for Simulating Multiphase Flow within Large-Scale Engineered Subsurface Systems
Simulation of multiphase flow in the subsurface is well-known to be computationally challenging. While there have been many studies that have explored approaches to overcoming these challenges, they often utilize relatively simple case studies. In this paper, we focus on the unique numerical challenges posed by modeling large-scale engineered subsurface systems, characterized by discrete features embedded in a heterogeneous natural subsurface setting. The man-made features such as shafts, tunnels, and barriers often cause multiple challenges in modeling the domain for multiphase porous media flow. This flow scenario can have a wide range of applications such as nuclear waste repositories, enhanced recovery of a petroleum reservoir, geothermal engineering, and carbon sequestration. An example of these severe numerical challenges is the case of performance assessment (PA) for Waste Isolation Pilot Plant (WIPP), the only operating deep geological repository in the US, which simulates extreme material properties of bedded salt rock formation and extreme contrast due to open excavation next to the formation. The models have extremes not only of permeability and porosity but also of the constitutive models needed for multiphase flow; additionally, they have process models like salt creep closure reducing porosity over time, fracturing in clay and anhydrite interbeds of the bedded salt, gas generation from the waste materials, and unintentional human borehole intrusions in some scenarios. Numerical simulations require the solution of coupled systems of nonlinear PDEs; in our work, we use the open-source simulator PFLOTRAN which is based on Finite Volume discretization. The solution of the nonlinear equations requires use of the Newton-Raphson iteration at each time step, which entails the solution of the linearized Jacobian system at each iteration. The effects of all the processes (i.e., large number of unknowns, highly nonlinear constitutive relations, large contrasts in material properties in short distances) lead to an ill-conditioned Jacobian matrix that severely challenges traditional linear solver, i.e., stabilized biconjugate gradient with block Jacobi incomplete LU preconditioner (BCGS-ILU) leading to non-convergence for traditional Newton-Raphson nonlinear solver causing unacceptably long computation time for each model. This paper presents linear solvers such as constrained pressure residual (CPR) two-stage preconditioner with alternate-block-factorization (ABF) and quasi- implicit pressure and explicit saturation (QIMPES) decouplers and flexible generalized residual solver (FGMRES). The new general-purpose nonlinear solver, Newton trust-region dogleg Cauchy (NTRDC), is also introduced to resolve extreme nonlinearities in the models. We demonstrate the effectiveness of each method relative to the default BCGS-Newton solver. The two best cases had nearly 50 times speed-up and achieved completion of a simulation in 14 hours that never completed due to non-convergence with the default solver. We also investigate the strong scalability of each method and discuss some of the deficiencies found for Block Jacobi preconditioner using parallel domain decomposition, and node packing effects of modern processor architecture.
Preconditioner, Nonlinear, Porous media, Multiphas↗