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DOE OSTI2022

Efficient Multi-Scale, Thermo-Hydro-Mechanical Numerical Model for Simulation of Long-Term Stability of Rooms for Nuclear Waste Disposal

This project demonstrated the feasibility of developing a numerical modeling tool that provides an efficient and unique approach to analyze the geomechanical performance of a deep geological repository (DGR) for high-level nuclear waste. The numerical modeling tool is physics-based and models the thermo-hydro-mechanical (THM) processes during the required time (up to one million years) for a DGR. The unique capability of the developed code is explicit representation of fracturing and a discrete fracture network (DFN) in the model that includes the entire repository layout for simulated time of one million years. Therefore, it can potentially address one of the challenges of analysis of a DGR, which is efficient numerical simulation of a multi-scale, physics-based models.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Actinide Containment Safety Case in a TRU/HLW Repository in Salt - 16369

The safety case for the permanent isolation of actinides in a salt nuclear repository is discussed. The need for actinide and brine chemistry research is driven by the regulatory process that requires an accounting of low-probability 'worst case' scenarios to show that the release criteria are met under these unexpected conditions. An overview of the WIPP approach to establish the actinide source term, which is defined by the oxidation state distribution, solubility and tendency to form colloidal species, is presented for the key actinides. This approach is then discussed in the context of extending the salt repository concept to defense high level waste (HLW) and possibly spent fuel (SF) to identify key needs and gaps. It is an overall conclusion of this review that the WIPP serves as a good template for the HLW/SF case and the safety case can be readily extended to this repository concept with proper repository design and supporting experimental results. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Technical Feasibility of Direct Disposal of Electrorefiner Salt Waste

The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of this EBR-II used fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy decided to treat the EBR-II sodium-bonded used fuel in an electrorefiner (ER), which produces a metallic waste, mostly from the cladding. The salt remaining in the ER contains most of the actinides and fission products. Two baseline waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams are being investigated that may reduce the complexity. For example, performance assessments show that both mined repositories in salt and deep boreholes in basement crystalline rock can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Hence the focus of a direct disposal option, as described herein, is now on the feasibility of packaging the ER salt waste in the near term such that it can be transported to a repository in the future without repackaging. A vessel for direct disposal of ER salt waste has been previously proposed, designed, and a prototype manufactured based on desirable features for use in the hot cell. The reported analysis focused on the feasibility of transporting this proposed vessel and whether any issues would suggest that a smaller or larger size is more appropriate. Specifically, three issues are addressed (1) shielding necessary to reduce doses to acceptable levels; (2) the criticality potential and the ease which it can be shown to be inconsequential, and (3) temperatures of the containers in relation to acceptable cask limits. The generally positive results demonstrate that direct disposal of ER in the proposed packaging is feasible without the need to secure funding to modify the facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Should WIPP Resume Unfiltered Discharge of Underground Ventilation?

On February 14, 2014 an incident in Panel 7, Room 7 of the Waste Isolation Pilot Plant (WIPP) underground repository resulted in the release of americium and plutonium (mostly {sup 241}Am) into the environment and contaminated portions of the repository primarily along the ventilation path from the location of the incident. The WIPP uses a ventilation system to ensure that underground air is circulated throughout the repository and to ensure that conditions are safe for the workers. Since the air in the repository exits to the surface through an exhaust shaft, this shaft is the sole potential pathway for airborne radioactivity release from the repository. During normal operations, exhaust air is released to the environment unfiltered. However, since the radiation release event, the WIPP ventilation system has been, and will likely remain in what is called 'filtration mode', (i.e., exhaust air is routed through a HEPA filter system). Redirection of the ventilation system through the HEPA (High Efficiency Particulate Air) filter system was necessary at the peak of the radiation release event to protect aboveground workers at the site and the public in the surrounding areas. However, evaluation of more than two years of underground ventilation sampling data indicates that residual radioactivity levels in the underground no longer warrant HEPA filtration in order to meet either worker or environmental protection criteria. WIPP should consider resumption of the unfiltered discharge of underground ventilation to the environment. Recent sampling data indicate that there have been no detectable releases that would have resulted in a hypothetical dose to a worker or a member of the public in excess of the 0.1 mSv (10 mrem) per year limit for periodic confirmatory sampling as required for operating repositories by the Environmental Protection Agency (40 CFR 191)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Updated Declaration of Legacy Waste Based on Re-evaluated Records - 18283

Kerntechnische Entsorgung Karlsruhe (KTE) has the assignment to decommission and dispose of the nuclear installations on the Karlsruhe site. The entire inventory of these installations as well as the existing waste materials stored in the interim storage of the central waste management section (formerly central decontamination department / HDB) is planned to be sent to the repository Konrad commissioned for the acceptance of low level waste and intermediate level waste. Therefore it is treated according to the respective acceptance criteria. KTE has approximately 45.000 m{sup 3} of legacy waste, which was produced prior to the first issue of these acceptance criteria [1] and hence has to be re-qualified. Legacy waste generally has a radiological declaration, which is not sufficient to meet the criteria applied by the regulators and independent experts today. In order to achieve a modern declaration, additional information such as acquisition records, operational reports, calculated activation, old and new analyses and burn-up calculations as well as knowledge of the physical and chemical boundary conditions of defined material streams is needed. At KTE the waste was treated in batches according to accepted methods documented in the database system KADABRA. For each batch, records regarding the waste type, activity and mass are available for reevaluation. Each batch record consists of information on the raw waste treated in the batch, on the batch itself and on the resulting interim products or waste products. On the basis of this existing or additionally acquired knowledge a declaration concept matching the existing information and the requirements of the repository can be derived and discussed with the independent expert and the regulator. (author)

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DOE OSTI2016

Retrieval of High Level Waste for the Cigeo Project - A Practical Retrieval Test Story - 16085

Since the passing of the June 28, 2006 Act, Andra has incorporated in its step-wise design development of the French high level waste (HLW) and intermediate level long lived waste (IL-LLW) deep geological repository (DGR, aka Cigeo) various requirements linked to 'reversibility' at different stages of the repository life cycle (from its construction phase to its final closure). Reversibility is based on a number of practical dispositions including the progressivity of the development and deployment of the facility, the flexibility of the nature of operations being carried out, the adaptability of the facility to future modifications and the possible retrieval of waste. The capacity to retrieve waste from Cigeo, also called 'retrievability', bears specific requirements. On a shorter term, one must demonstrate that the waste container retrieval operations are technically safe and feasible over at least 100 years. The current paper starts with two acceptations: retrievability and reversibility. It details how those two notions are defined and inter-linked. The 'Retrievability Scale' as proposed by OECD/NEA (and adopted by Andra) is also reminded. The needs for demonstration of retrieval operations are also exposed for the two categories of waste containers disposed of in Cigeo (IL-LLW and HLW) within the context of their respective emplacement configurations (in disposal horizontal vaults and micro-tunnels) at different stages of the repository life cycle. The technical part of the paper is then focused on a practical case story: a technological test campaign (still running) of HLW container retrieval out of a steel cased disposal micro-tunnel. On a test bench built on surface to simulate (create) the environmental conditions that may prevail inside a HLW micro-tunnel at time of retrieval, it was possible to check whether the mechanical means developed for the purpose were adequate and sufficient. The test case story is exposed, c/w the problems and challenges encountered (high temperature impairing the retrieval robot's proper functioning, presence of corrosion products...). The trouble-shooting solutions implemented and the results are commented. This article concludes with a critical analysis of the methods and equipment used, the results obtained and provides some perspectives of improvement. The positive outcomes of this technical test campaign pave the way for planning the in situ retrieval tests to be programmed in the 'industrial pilot phase' of Cigeo (to be implemented over the ten first years of operations, including the active and inactive test period and the first years of disposal activities): the industrial pilot phase is presently scheduled around 2026-2036 (provided the Cigeo construction is authorized in 2021, following the licensing request filing scheduled in 2018). The issues of IL-LLW containers retrieval (for which specific solutions are also studied and technical demonstrations engaged or planned) are also briefly presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Qualification of Pre-existing Containers According to the Acceptance Criteria of the Repository Konrad - 17204

As well as working in the field of waste management, Kerntechnische Entsorgung Karlsruhe GmbH (KBE) is, for historical reasons, one of Germany's three constructors of repository containers for the repository Konrad. The focus of this container production is merely on the side of requalification of the existing legacy containers, which are partly already loaded and grouted. To a smaller extend, KTE also produces new containers for its own use because the containers exactly match the waste properties present on the Karlsruhe site. There are three major container types in the process of requalification: - The Type-IV-Drum-Steel-Container FSC-IV-D was successfully licensed in 2014 and regularly produced thereafter. At the same time, the pre-existing containers of this type are re-qualified on the basis of this license by proof of the compliance with this license. - The Type-IV-Product-Container PSC-I-D is identical to the FSC-VI-D in terms of steel construction but also contains a liner of concrete for shielding and protection of the steel construction. This container is designed for large components and demolition waste. After a successful drop test, it is currently in the advanced process of licensing. - Type I cylindrical concrete containers are simple single drum shieldings which were produced in large numbers in the 1970's and 1980's. KTE owns 7,000 of those containers, thus making requalification and licensing cheaper than retreatment of the waste after demolition of the containers. (authors)

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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↗
DOE OSTI2016

Evaluation on radiation protection aspect and radiological risk at Mukim Belanja repository

Asian Rare Earth (ARE) is a locally incorporated company that operated a mineral processing operation to extract rare earth element. ARE has received much attention from the public since the beginning of their operation until the work of decommissioning and decontamination of the plant. Due to the existence of Naturally Occurring Radioactive Material (NORM) in the residue, the decommissioning and disposal was done by the company in collaboration with the Perak State Government and the Atomic Energy Licensing Board (AELB). The main objective of this study is to review the level of compliance of the existing Radiation Protection Regulations enforced by AELB particularly in the achievement of allowed exposure dose limit. The next objective was to study the impact of the construction of the Mukim Belanja Repository to workers and public. This study was conducted by analyzing documents that were issued and conducting the area monitoring using a Geiger Muller detector (GM) and Sodium Iodide (NaI(Tl)) survey meters. The measurements were made at 5 cm and 1 m from the ground surface at 27 measurement stations. The external doses measured were within the background levels of the surrounding area. The annual effective dose using the highest reading at 5 cm and 1 m from ground surface by GM detector was calculated to be 1.36 mSv/year and 1.21 mSv/year respectively. Whereas the annual effective dose using the highest reading at 5 cm and 1 m from ground surface by using NaI(Tl) detector was calculated to be 3.31 mSv/year and 2.83 mSv/year respectively. The calculated cancer risks from the study showed that the risk is small compared with the risks derived from natural radiation based on global annual radiation dose to humans. This study therefore indicated that the repository is able to constrain the dose exposure from the disposed NORM waste. The study also revealed that the construction of the repository has complied with all the rules and regulations subjected to it. The exposed dose received by the radiation and the public workers during the construction of the repository were below the annual limit i.e. 20 mSv/year and 1mSv/year respectively.

61 RADIATION PROTECTION AND DOSIMETRY↗
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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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 OSTI2014

Performance Assessment Modeling and Sensitivity Analyses of Generic Disposal System Concepts

The Used Fuel Disposition Campaign (UFDC) of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) is conducting research and development (R&D) on generic deep geologic disposal systems (i.e., repositories) for high-activity nuclear wastes (i.e., used nuclear fuel (UNF) and high-level radioactive waste (HLW)) that exist today or that could be generated in future fuel cycles. This report describes specific activities in FY2014 toward the development of an enhanced generic disposal system modeling and analysis capability that utilizes high performance computing (HPC) environments to simulate important multi-physics phenomena and couplings associated with the potential behavior of a geologic repository for UNF and HLW.

Sevougian, S. David↗
DOE OSTI2016

Low- and Intermediate-Active Waste Management in the Federal Republic of Germany - 16567

Waste with negligible heat generation (low- and intermediate-active waste) generated in Germany is to be emplaced in the Konrad repository. As the start of operation of the Konrad repository is expected in 2022, regulations have been developed that allow a longer storage period while keeping the same high level of safety. The guidelines for the storage of radioactive waste with negligible heat-generation have become part of the regulatory framework and its enforcement is thoroughly guided by the responsible Federal Ministry. Radioactive waste generated during the operation or the decommissioning of nuclear power plants and nuclear fuel cycle facilities has to be stored by the operator until the Konrad repository is available. In contrast, radioactive waste generated from other industrial, medical and research applications including spent nuclear sources, initially has to be delivered to a Land collecting facility and has to be registered, treated and stored there. The Land collecting facilities will deliver the resulting waste packages to the disposal facility as of 2022. While the operation of the Land Collecting Facilities is a task of the responsible Laender authorities, the operation of the repository Konrad is a federal task. The interaction of the involved different parties will be described. This paper will focus on the management of low- and intermediate-active waste. It will give an insight to the expected amounts and different kinds of waste streams. The different concepts for collecting and qualification procedures for waste packages depending on the origin of the waste will be presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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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