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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

'A Good Decision Is Based On Information, Not On Numbers' - A Process For National Strategic Decision Making - 17203

Waste management decision making is the process used by waste owners to determine and justify choices in across the lifecycle of a waste (from the manner of generation or retrieval to final disposition). In the UK, waste owners in the nuclear industry are obligated, by their environmental permits and authorisations, to demonstrate the application of robust and well-underpinned decision making processes for waste management. This is captured through the requirement for waste owners to demonstrate the identification and implementation of Best Available Technique (BAT). The National Waste Programme (NWP), led by LLW Repository Ltd, is a UK wide programme which leads the implementation of the UK National Strategy for the management of solid LLW from the nuclear industry [1] and which aims to deliver a self-sustaining culture for optimised management of Low Level Waste (LLW) in the UK. Recognising the requirement on waste owners, the NWP has - in conjunction with appointed contractors - led the development and publication of national strategic optioneering studies (so-called national strategic BAT studies) to support waste owners. Recognising the benefits that this approach to optioneering and strategic decision making could have for other waste producers, this paper provides an overview of the process used for the development of the suite of national strategic BAT assessments for the NWP. It details: the main phases of the process, the key features of each process phase, the benefits of the approach and key learning from its deployment in practice. Note: The legislative regime in England and Wales requires the identification and application of BAT. There is an equivalent regime in Scotland known as Best Practicable Environmental Option (BPEO) and Best Practicable Means (BPM). The requirements are comparable across the two legislative regimes and, for the purposes of simplicity, the term BAT will be utilised in this paper. The approach and outcomes of the national strategic optioneering process described are applicable across the UK. (authors)

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DOE OSTI2020

Dual Purpose Canister Reactivity and Groundwater Absorption Analyses

The current spent nuclear fuel (SNF) management strategy includes reliance on dry storage. Utilities are meeting their interim storage needs on an individual basis with use of large-capacity dry storage casks, with a current focus on meeting existing storage and transportation requirements, as disposal requirements are not currently available. These casks are commonly known as dual-purpose (i.e., storage and transportation) canisters (DPCs). However, a small percentage of single-purpose (storage only) systems is also being used to meet storage needs. These are included under the “DPC” heading. This report investigates the postclosure criticality safety aspects of DPCs. Placing large, heavy waste packages containing DPCs into a repository for direct disposal has not yet been implemented domestically or internationally. Therefore, direct disposal of DPCs represents new engineering and scientific challenges. Some of the engineering challenges that have already been addressed include handling and placement, use of ramps vs. shafts, use of hoists, use of transport equipment, and thermal management. Additionally, some studies have been conducted in the past regarding the feasibility of direct disposal from a criticality analysis perspective and have concluded that while possible, demonstrating subcriticality over the disposal time period is a challenge. The alternative to direct disposal of DPCs into a repository is to repackage the SNF into different canisters. The direct disposal of DPCs without cutting them open and repackaging is appealing because it could be more cost-effective, reduce the complexity of fuel management operations both in and outside reactor facilities, and result in less cumulative worker dose during interim storage and handling before eventual disposal in a deep geologic repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1981

Safety aspects of nuclear waste disposal in space

Safety issues involved in the disposal of nuclear wastes in space as a complement to mined geologic repositories are examined as part of an assessment of the feasibility of nuclear waste disposal in space. General safety guidelines for space disposal developed in the areas of radiation exposure and shielding, containment, accident environments, criticality, post-accident recovery, monitoring systems and isolation are presented for a nuclear waste disposal in space mission employing conventional space technology such as the Space Shuttle. The current reference concept under consideration by NASA and DOE is then examined in detail, with attention given to the waste source and mix, the waste form, waste processing and payload fabrication, shipping casks and ground transport vehicles, launch site operations and facilities, Shuttle-derived launch vehicle, orbit transfer vehicle, orbital operations and space destination, and the system safety aspects of the concept are discussed for each component. It is pointed out that future work remains in the development of an improved basis for the safety guidelines and the determination of the possible benefits and costs of the space disposal option for nuclear wastes.

Rice, E. E.↗
DOE OSTI2020

Optimization of the Post-Operational Phase on Two Belgian Multi-Unit Nuclear Power Plants: the Case of the Non-Fissile Irradiated Core Items - 20156

The current legal framework in Belgium foresees the progressive phase out of nuclear power between October 2022 (Doel 3) and December 2025 (Doel 2). Upon its definitive shutdown, each unit of the Tihange and Doel sites will enter a Post-Operational Phase (POP) and be prepared for its Decontamination and Decommissioning (D and D). Prior to obtaining the D and D license, the Operator Electrabel is legally required to remove any non-fissile irradiated core items stored in the deactivation pools. The non-fissile irradiated core items consist essentially of control rods, poison rods and source thimbles as well as thimble plugs and foreign materials irradiated during operation: - Their significant content in highly radiant radionuclides (up to 6 TBq of Co-60 per kg of irradiated material) renders all existing operational waste management processes inadequate due to insufficient biological shielding; - Their high concentrations in long-lived radionuclides call for their disposal in a geological repository for which no final design nor waste acceptance criteria are expected prior to 2050. Uncertainties in the Belgian energy supply and security, however, require the Operator to be prepared for a partial nuclear phase out, where one or more units would benefit from lifetime extension while the remaining units would undergo decommissioning. The present paper aims at presenting how Electrabel, in partnership with Tractebel, addressed this challenge by maximizing the use of synergies within the respective sites as well as between both sites themselves, all the while accounting for site specificities. The most recent results and state of progress of the project will be detailed and the first lessons learned will be shared. The project has been split in multiple tasks and phased as follows: - An inventory phase aimed at mapping the contents, origin, composition and history of the non-fissile irradiated core items; - A pre-characterization phase based on neutron activation models; - A waste sorting phase aimed at separating waste forms for which an evacuation route exists from those for which such route does not exist; - A feasibility phase aimed at exploring all possible scenarios for the management of non-fissile irradiated core items and identifying the optimal feasible solution for each site; - A preparation phase (currently ongoing), developing further the optimal solution and ensuring that back-up solutions are available for any foreseeable change of context (licensing issue, modification in the nuclear phase-out program, etc.) and initiating early contacts with potential subcontractors for segmentation works and cask manufacturers, as well as the Belgian regulatory body and waste management agency. This phase also foresees the investigation of destructive and non-destructive radiological measurements to support the detailed characterization of the waste forms; - A realization phase (future work). (authors)

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DOE OSTI2017

Deep Borehole Disposal Research in the United States - 17427

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) manages a research and development (R and D) program for the disposal of spent nuclear fuel and high-level radioactive waste in the United States. Deep boreholes are being evaluated for disposal of some smaller DOE-managed waste as an alternative to disposal in a mined geologic repository. The deep borehole disposal concept consists of drilling borehole(s) to a total depth of about 5,000 m (16,400 ft - or greater than three miles) into crystalline basement rock, placing waste packages in the lower emplacement zone portion of the borehole, and sealing and plugging the upper portion of the borehole. The required bottom-hole diameter of the borehole(s) depends on the waste package configuration for the reference concept, but ranges from 22 to 43 cm (8.5 to 17 inches). Additional research and development is desired in several important areas for further consideration of deep borehole disposal of radioactive waste. A Deep Borehole Field Test (DBFT) will be conducted by DOE to further assess the viability of deep borehole disposal as an alternative for the disposal of smaller DOE-managed waste forms. The DBFT will be used to evaluate three main areas: 1) evaluation of the capability for drilling and construction of deep, large-diameter boreholes; 2) downhole scientific analyses to assess hydrogeochemical conditions that control waste stability and containment; and 3) engineering activities to assess the viability and safety of waste package emplacement and retrieval in deep boreholes. The DBFT will include first drilling a characterization borehole, followed by an optional field test borehole, to a total depth of about 5,000 m (16,400 feet) into crystalline basement rock in a geologically stable continental location. The initial smaller diameter characterization borehole will be drilled and completed to facilitate downhole scientific testing and analysis. If site conditions are found to be suitable, DOE intends to drill a larger-diameter field test borehole to evaluate emplacement activities using surrogate waste packages. The surrogate waste packages will not contain any radioactive waste, nor will any radioactive waste be used in these tests. One of the advantages of deep borehole disposal is that it would give DOE the flexibility to consider options for disposal of smaller waste forms. Deep borehole disposal also potentially offers earlier disposal of some wastes than might be possible in a mined repository, and could reduce costs associated with projected treatments of some wastes. In order to accomplish the DBFT, DOE will require the support of multiple national laboratories, along with a site contractor to acquire a suitable test site and to manage the site and the required testing activities. DOE issued a competitive solicitation via a request for proposal (RFP) in July, 2015, which requested proposals to provide a test site and site management services. The contract was awarded in January, 2016, to a team led by Battelle Memorial Institute which included a proposed test site in Pierce County, North Dakota. After efforts to acquire both the initial test site in North Dakota and an alternative proposed site in Spink County, South Dakota were unsuccessful, activities were suspended. The experiences in Pierce County, ND and Spink County, SD highlighted the importance of public engagement and support for the DBFT, and that relevant levels of government and other public stakeholders should be involved from the beginning. Using the lessons learned in North and South Dakota, DOE issued a new RFP on August 22, 2016 which emphasized local, state, and tribal (if applicable) government engagement, as well as public and other stakeholder involvement ahead of proposal submittals and throughout the contract execution phases. The new RFP allowed for multiple initial awards and multiple phases of contract execution. In December, 2016, DOE announced the selection of four companies - AECOM, ENERCON, RE/SPEC, and TerranearPMC - to begin exploring the possibility of conducting a DBFT under the terms of the new RFP. Ultimately, only one site will be chosen for the field test. (authors)

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DOE OSTI2025

Effect of Citrate on the Solubility of Uranium (VI) in WIPP Brine

The Waste Isolation Pilot Plant (WIPP) is the only active deep geological repository in the United States for the disposal of defense-related transuranic (TRU) waste, located in the northern part of the Delaware Basin in southeastern New Mexico, approximately 26 miles east of Carlsbad. The repository is situated 2,150 feet (about 610 meters) underground within the Salado Formation, a thick layer of stable salt deposits. This unique geological formation provides a secure and long-term environment for isolating TRU waste, ensuring its safe containment for thousands of years.

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DOE OSTI2021

Repository-Scale Performance Assessment Incorporating Postclosure Criticality

A key objective of the United States Department of Energy’s (DOE) Office of Nuclear Energy’s Spent Fuel and Waste Science and Technology Campaign is to better understand the technical basis, risks, and uncertainty associated with the safe and secure disposition of spent nuclear fuel (SNF) and high-level radioactive waste. Commercial nuclear power generation in the United States has resulted in thousands of metric tons of SNF, the disposal of which is the responsibility of the DOE (Nuclear Waste Policy Act of 1982, as amended). Any repository licensed to dispose of SNF must meet requirements regarding the long-term performance of that repository. For an evaluation of the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration during the postclosure period. Of particular interest is the potential behavior of SNF in dual-purpose canisters (DPCs), which are currently licensed and being used to store and transport SNF but were not designed for permanent geologic disposal. A study has been initiated to examine the potential consequences, with respect to long-term repository performance, of criticality events that might occur during the postclosure period in a hypothetical repository containing DPCs. The first phase (a scoping phase) consisted of developing an approach to creating the modeling tools and techniques that may eventually be needed to either include or exclude criticality from a performance assessment (PA) as appropriate; this scoping phase is documented in Price et al. (2019a). In the second phase, that modeling approach was implemented and future work was identified, as documented in Price et al. (2019b). This report gives the results of a repository-scale PA examining the potential consequences of postclosure criticality, as well as the information, modeling tools, and techniques needed to incorporate the effects of postclosure criticality in the PA.

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DOE OSTI2026

Thermal loading effects on chalk hydromechanical behavior for nuclear waste disposal

Safe disposal of heat-generating nuclear waste depends on host rock stability under thermal, hydrological, and mechanical stresses. This study investigates the effect of thermal loading on mechanical behavior of the shallowly buried Ghareb formation chalk through triaxial and hydrostatic constant strain rate and creep tests at temperatures up to 100 ˚C and effective pressures up to 20.7 MPa. Experimental results show that thermal loading reduces the elastic moduli of chalk by 50–75%, and a transition occurs above 60 ˚C where creep rates increase rapidly. Water saturation nearly doubles the thermally induced strain compared to dry conditions and strongly decreases material rigidity. Thermal loading also leads to significant pore pressure increases under undrained conditions and leads to reductions in the apparent permeability during drained conditions. Laboratory experimental data were used to parameterize and develop a preliminary constitutive model for predicting future deformation during repository operations in the Ghareb. The strongly coupled effects – mechanical weakening, fluid pressure fluctuations, and permeability modification – demonstrate that elevated repository temperatures will have a pronounced effect on the near field Ghareb behavior during waste disposal operations. The findings indicate that the coupled interactions must be considered in predictive models and repository design to ensure long-term nuclear waste isolation and safety.

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DOE OSTI2018

DECOVALEX-2015: an international collaboration for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems

Here, an important part of the performance and safety assessment of disposal systems for radioactive waste and spent nuclear fuel in deep geological formations is to evaluate the impact of the coupled effects of mechanical deformation, fluid and gas flow through the repository, and thermal loading from the decaying waste. It was recognized early on in safety and performance assessments conducted for geologic disposal sites that to be able to conduct such an evaluation, there was a need to rigorously enhance the theoretical background and to develop models capable of simulating coupled thermo-hydro-mechanical (THM) processes. More recently, chemical processes have also been added to enable the study of fully coupled THMC processes in geosystems.

58 GEOSCIENCES↗
DOE OSTI2020

Deep Isolation-Development of the Safety Case for Disposal of Radioactive Wastes in Horizontal Boreholes - 20028

Deep Isolation has developed a safe, secure, and permanent geological disposal method for high-level waste, including spent nuclear fuel as well as sealed sources and other highly radioactive materials. The method leverages well established directional drilling technology to create horizontal repositories deep underground. The combination of great depth (1-3 km) and the ability to precisely position repositories in a horizontal orientation provides access to geologic strata that are inaccessible for typical mined and deep vertical borehole repositories. A number of potential direct and indirect safety benefits accrue including:1) greater depth below regional freshwater aquifers; 2) increased flexibility and more geologic options for siting; 3) access to deep formations with persistent and sustained reducing conditions. In addition there are a number of safety elements related specifically to the horizontal repository geometry including: 1) passive direction of thermally driven fluid and radionuclide movement away from the vertical access hole and toward the 'dead end' portion of the repository; 2) mitigation of seismic hazards by orienting repositories parallel to local and regional fault structures. In this paper we explore and discuss some of the key technologic, geologic and hydrologic elements that support the deep horizontal borehole safety case. The stalemate seen across the globe on the disposal of spent nuclear fuel and high-level waste can be broken. Deep Isolation offers a novel option for safe, secure, and permanent deep geological disposal of nuclear waste that can be developed as centralized repositories or adapted to smaller regional or site-specific repositories located near waste sources. (authors)

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DOE OSTI2018

International Collaboration Activities on Engineered Barrier Systems

International collaborations on nuclear waste disposal is an integral part of the Spent Fuel Waste Science and Technology (SFWST) campaign within the DOE Fuel Cycle and Technology (FCT) program. These engagements with international repository R&D programs provide key opportunities to participate in experiments with international partners on research investigations developing laboratory/field (underground research laboratories (URL) experiments) data of engineered barrier system (EBS) components (e.g., near-field) and characterization of transport phenomena in the host rock (e.g., far-field). The results of these field and laboratory experiments are used in the evaluation of coupled processes and the development of state-of-the-art simulation approaches to evaluate repository performance. Thermal heating from radionuclide decay in the waste canisters will increase temperature in the surrounding EBS driving chemical and transport processes in the near- and far-field domains of the repository. URL heater-tests for extended periods of times (e.g., years) provide key information and data of thermal effects on barrier responses to temperature and water saturation levels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

The Criticality Safety Studies of Joonhong Ahn

The recent passing of UC Berkeley professor Joonhong Ahn at the age of 57 has been a dear loss to the community of nuclear scientists at large. His contributions to studies on advanced fuel cycle development and the societal and ethical aspects of nuclear technology were known around the world and fostered a great reputation. He produced at least fifty peer-reviewed journal papers, authored or contributed to ten books, and organized a considerable number of international meetings. Ahn's studies led to his eminence in the area of nuclear waste management, where he focused on repository performance assessments along with their impact on society. These assessments included the development of analytical models for nuclide transport phenomena and evaluation of the possibility of criticality events imparting a dose to the biosphere. This paper will summarize his achievements in criticality safety along with developments before his passing.

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DOE OSTI2024

DECOVALEX-2023: Task B Final Report

In all repository concepts for the geological disposal of radioactive waste, an engineered barrier system (EBS) is used to encapsulate the waste canister, or, to act as borehole or gallery seals. These systems are often based on bentonite clays due to their low permeability and high swelling capacity enabling the closure of engineering voids. However, in all repository concepts gases will be generated through the corrosion of metallic materials (under anoxic conditions), the radioactive decay of waste and the radiolysis of water. Thus, understanding the processes and mechanisms controlling the advective movement of gas (as a discrete phase) in clay-based materials is a key aspect when assessing the impact of gas flow in a repository safety case.

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DOE OSTI2023

Computational Materials Design for Ceramic Nuclear Waste Forms Using Machine Learning, First-Principles Calculations, and Kinetics Rate Theory

Ceramic waste forms are designed to immobilize radionuclides for permanent disposal in geological repositories. One of the principal criteria for the effective incorporation of waste elements is their compatibility with the host material. In terms of performance under environmental conditions, the resistance of the waste forms to degradation over long periods of time is a critical concern when they are exposed to natural environments. Due to their unique crystallographic features and behavior in nature environment as exemplified by their natural analogues, ceramic waste forms are capable of incorporating problematic nuclear waste elements while showing promising chemical durability in aqueous environments. Recent studies of apatite- and hollandite-structured waste forms demonstrated an approach that can predict the compositions of ceramic waste forms and their long-term dissolution rate by a combination of computational techniques including machine learning, first-principles thermodynamics calculations, and modeling using kinetic rate equations based on critical laboratory experiments. By integrating the predictions of elemental incorporation and degradation kinetics in a holistic framework, the approach could be promising for the design of advanced ceramic waste forms with optimized incorporation capacity and environmental degradation performance. Such an approach could provide a path for accelerated ceramic waste form development and performance prediction for problematic nuclear waste elements.

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DOE OSTI2017

The Reactor Graphite Handling. Problems and Solutions - 17208

The problem of industrial uranium-graphite reactors and NPP units decommissioning is one of the urgent tasks that require efficient and cost-effective solutions. To date, only Russia has already shut down and subject to decommissioning more than 120 objects. World-wide, more than 260,000 ton of irradiated graphite have arisen from the commercial nuclear power operations, research reactors, and defense production reactors. Under the circumstances the problem to develop cost-effective methods of the irradiated graphite management has appeared. Alternative disposal options for the irradiated graphite were discussed for at least 25-30 years. To date, various methods of irradiated graphite reprocessing have been proposed. They are based on the application of different physical and chemical processes, most of which involve incineration of the entire mass of graphite. However, they all have disadvantages, and do not solve the whole complex of problems. The primary goal of the research presented here was to carry out a study aimed at the development of a more simple technology (as compared to the incineration of the entire mass of graphite) that would provide for the graphite purification and allow the possibility of its transfer to the less hazardous waste category followed by the disposal of it in a near-surface repository. To achieve it is proposed to consider two directions, each of them reducing the cost of graphite reprocessing and the waste volume disposed of in the repository. The first alternative is based on the principle of partitioning radionuclides of different nature from the graphite, thus allowing to transfer it into the LLW category and minimize the volume of secondary waste products. The second option to change the reprocessing technology is the idea to use reactor graphite as the starting material to produce stable final waste form compositions intended for the long-term storage of radioactive waste including transuranic elements. (authors)

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DOE OSTI2022

Incorporating Heterogeneity into 3D Geologic Models for the Alluvial Basin Reference Case

The unsaturated zone alluvium reference case is one of several geologic systems under consideration by the U.S. Department of Energy Office of Nuclear Energy for hosting repositories for spent nuclear fuel and associated waste (Sevougian et al., 2019). As noted by Mariner et al. (2018), the generic alluvial basin offers positive attributes that merit its consideration as a reference case by the Spent Fuel and Waste Science and Technology (SFWST) campaign. Hundreds of alluvial basins and sub-basins are scattered across the arid western United States (Figure 1-1). Precipitation and infiltration rates are relatively low with high evapotranspiration, resulting in deep regional water tables. Deep regional water tables allow for significant vertical separation between proposed repository depths and groundwater and thus longer transport paths from a hypothetical repository to an aquifer. Accumulations of alluvial sediments within these basins are typically on the order of hundreds of meters, and locally may exceed 1,000 m in thickness, as is the case for the Deming sub-basin in southern New Mexico. In alluvial basins, alluvium creates a thick geologic host medium that serves as the natural barrier system (NBS) in the conceptual model framework of a geologic disposal system and isolates the waste packages from receptors in the biosphere. Further, alluvial basin fill is typically comprised of stacked playa and lacustrine deposits along the basin axis (Perry et al., 2018). Characterized by low permeability, these layers protect the biosphere above the repository and the groundwater resources below the repository by limiting flow and transport of radionuclides.

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DOE OSTI2019

Postclosure Criticality Consequence Analysis - Scoping Phase

Commercial generation of energy via nuclear power plants in the United States (U.S.) has generated thousands of metric tons of spent nuclear fuel (SNF), the disposal of which is the responsibility of the U.S. Department of Energy (DOE) (Nuclear Waste Policy Act of 1982). Any repository licensed to dispose of the SNF must meet requirements regarding the long-term performance of the repository. In evaluating the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration. Of particular interest is the potential behavior of SNF in dual-purpose canisters (DPCs), which are currently being used to store the SNF but were not designed for permanent disposal. As part of a multiyear plan that is currently being developed for the DOE, a two-phase study has been initiated to examine the potential consequences, with respect to long-term repository performance, of criticality events that might occur during the postclosure period in a hypothetical repository containing DPCs. Phase I, a scoping phase, consists of generating an approach intended to be a starting point for the development of the modeling tools and techniques that may eventually be required either to exclude criticality from or include criticality in a performance assessment (PA) as appropriate. The Phase I approach will be used to guide the analyses and simulations done in Phase II to further the development of these modeling tools and techniques as well as the overall knowledge base. The purpose of this report is to document the approach created during Phase I. The study discussed herein focuses on the consequences of criticality in a DPC; it does not address the probability of occurrence of a criticality event. This approach examines two types of criticality events for SNF disposed of in a single type of DPC: a steady-state criticality and a transient criticality. The steady-state critical event is characterized by a relatively low constant power output over 10,000 years, while the transient critical event is characterized by a power spike that lasts on the order of seconds. Possible effects of the criticality are an increase in the radionuclide inventory; an increase in temperature; and a change in the chemistry inside the waste package, along with a change in radionuclide solubilities, fuel degradation rates, and steel corrosion rates. Additionally, for transient criticality the possibility of mechanical damage to the engineered and natural barriers also exists.

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DOE OSTI2021

Reconstruction of the Room D, B, G, and Q Closure Histories at the Waste Isolation Pilot Plant

Experimental measurements of room closure in salt repositories are valuable for understanding the evolution of the underground and for validating geomechanical models. Room closure was measured during a number of experiments at the Waste Isolation Pilot Plant (WIPP) during the 1980's and 1990's. Most rooms were excavated using a multi-pass mining sequence, where each pass necessarily destroyed some of the mining sequence closure measurement points. These destroyed points were promptly reinstalled to capture the closure after the mining pass. After the room was complete, the mining sequence closure measurement stations were supplemented with remotely read closure measurement stations. Although many aspects of these experiments were thoroughly documented, the digital copies of the closure data were inadvertently destroyed, the non-trivial process of zeroing and shifting the raw closure measurements after each mining pass was not precisely described, the various closure measurements within a given room were not directly compared on the same plot, and the measurements were collected for several years longer than previously reported. Consequently, the hand-written mining sequence closure measurements for Rooms D, B, G, and Q were located in the WIPP archives, digitized, and reanalyzed for this report. The process of reconstructing the mining sequence closure histories was documented in detail and the raw data can be found in the appendices. Within the mid-section of a given room, the reconstructed closure histories were largely consistent with other mining sequence and remotely read closure histories, which builds confidence in the experiments and suggests that plane strain is an appropriate modeling assumption. The reconstructed closure histories were also reasonably consistent with previously published results, except in one notable case: the reconstructed Room Q closure histories 30 days after excavation were about 45 % less than the corresponding closures reported in Munson's 1997 capstone paper.

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