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At least 415 records · Page 23
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DOE OSTI2015

Guidance for decommissioning plants entering safe storage - 15280

Although a number of plants worldwide have been placed into safe storage after permanent shutdown, there is no generic guidance for establishment of safe storage conditions. To address this need, EPRI is conducting a project to document the best practices to reduce risk, maintenance and monitoring requirements during safe storage, as well as practices that optimize decommissioning following the safe storage period. This project additionally evaluates potential options for establishment of safe storage conditions, such as performance of chemical decontamination, completion of environmental remediation activities, early removal of highly activated components, dismantlement of non-power block structures and partial site release in advance of completion of decommissioning of the nuclear island. (authors)

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DOE OSTI2015

Solutions for management of defective fuel assemblies - 15318

Defective fuel management is a major challenge for nuclear operators when the time comes for finding a long-term solution for their used (or spent) fuel assemblies. This question of defective fuel management can be raised in several cases: - For optimizing the remaining operating costs of shutdown reactors, the spent fuel pools must be emptied as soon as possible; a solution needs therefore to be found for evacuating all the spent fuel assemblies, defective fuel assemblies and in particular defective fuel rods included; - Some utilities are today facing a saturation of their reactor spent fuel pool; evacuating the defective fuel assemblies and extracted defective fuel rods allows optimizing the space in the pool. Whereas the solutions for managing intact used fuel assemblies are well-known, a specific technical solution must be implemented for managing the defective spent fuel assemblies as well. The purpose of this paper is to describe the existing solutions for the management of all types of defective fuel assemblies. Interim Dry Storage Solution Interim dry storage of defective spent fuel assemblies is being implemented for decades in the USA. Specific operations are performed for preparing the defective fuel assemblies for interim dry storage and they are then stored with intact used fuel assemblies in the dry storage cask. The advantage of such a solution is its short-term cost effectiveness. The paper presents the existing dry interim storage solutions for defective used fuel rods, the way they are implemented, the associated operational experience. Long-term Dry Storage Solution A long-term dry storage solution for defective fuel assemblies is being developed, compliant with direct disposal requirements. The solution is based on the encapsulation of each defective fuel rod into a fuel rod capsule and an associated specific dry storage licensing approach: this is the best available technology regarding long term storage safety requirements. The paper presents this unique capsule technology, the transport and storage casks that can be used for transportation of defective fuel rods to the storage facility, and the associated transport licensing approach currently used in Europe. Treatment Solution. The most definitive way to get rid of defective fuel assemblies is their treatment (or reprocessing) in a treatment plant. There is then no need to assess the behaviour of these specific used fuel assemblies as for an interim storage facility or a final repository. The advantage of such a solution is its sustainable approach. The paper presents the existing transport and treatment solutions, the way they are implemented, the associated operational experience. (authors)

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DOE OSTI2015

Performance improvement and human resources transition management in D and D projects - 15440

Power generation and other nuclear facilities production is mainly characterized by stable working environment, with continuous efforts for improvement based on previous achievements, constant workload, foreseeable and established organizational hierarchy; and generation of cash. In contrary to normal production - when phasing down the facility for final shutdown, decommissioning and dismantling - one has to manage a new, highly dynamic working environment, with decreasing workload, new skills required; and consumption of cash. The key to successful transition is to recognize the need for establishing an explicit performance improvement program with an effective change management process in order to address this profound evolution as early as possible in the project. AREVA owns and operates large nuclear material processing facilities. This includes the responsibility for their end of life up to final Decommissioning and Dismantling (D and D). Faced with this transitional challenge, AREVA took example from well-known management methods (Lean Management, 6 Sigma, TPM...) successfully applied by the non-nuclear production industry with the corresponding philosophy of Operational Excellence. These methods were adapted to the specifics of nuclear decommissioning projects. Furthermore, specific computerized tools were developed to assist managers in implementing measures for re-deploying, re-training and re-qualifying personnel as well as re-organizing the work structure while maintaining safety and social consensus. Significant improvements in the dismantling works as well as in the management of generated waste were achieved early after start of the implementation of the performance improvement programs. (authors)

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DOE OSTI2015

Overview of the DOE nuclear fuels storage and transportation planning project - 15444

The US Department of Energy (DOE) Office of Nuclear Energy is conducting planning activities within the Nuclear Fuels Storage and Transportation Planning Project (NFST) to lay the groundwork for implementing interim storage, including associated transportation, per the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste, and to develop a foundation for a new nuclear waste management organization. The purpose of the NFST is to make progress on this important national issue within existing legislative and budgetary authorizations. The objective of the NFST is to develop and begin implementation of an integrated plan to (1) implement interim storage; (2) improve the overall integration of storage as a planned part of the waste management system; (3) prepare for the large-scale transportation of spent nuclear fuel and high-level radioactive waste, with an initial focus on removing SNF from the shutdown reactor sites; and (4) develop foundational information, resources, and capabilities needed to support the aforementioned objectives and future implementation decisions and actions. An over-arching goal is to develop options for decision-makers on the design of an integrated waste management system. The planning activities are divided into four major topical areas: (1) Consent-Based Siting, (2) Storage, (3) Transportation, and (4) Strategic Crosscuts. This paper provides an overview of the NFST and its activities. (authors)

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DOE OSTI2015

Issue focused stakeholder collaboration: reaching consensus on spent nuclear fuel transportation operational practices - 15465

This paper examines stakeholder and issue-focused approaches used by Department of Energy (DOE) programs to collaborate with key stakeholders on transportation practices to ship radioactive material, particularly used nuclear fuel (UNF) and transuranic waste (TRU). In accordance with the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste, the current effort by DOE to plan for the development of a future transportation system to ship UNF from commercial reactors, with an initial focus on shutdown reactors, is using approaches similar to those used in other successful shipping campaigns to collaborate with stakeholders about sustainable transportation practices. As part of a coordinated approach to stakeholder outreach and collaboration, DOE is currently developing a draft Transportation Planning Framework to provide a template for future shipments of UNF from commercial reactors and is engaged with states and tribes and other stakeholders in discussions on the institutional and operational components of that Framework. (authors)

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DOE OSTI2015

MP197HB; First licensed transportation package with canistered high burnup fuel after dry storage - 15580

AREVA TN has recently received the NRC 10 CFR Part 71 transportation license CoC 9302 Revision 7 for the MP197HB transportation package that allows transportation of canistered high burnup fuel assemblies. The Nuclear Horizontal Modular Storage (NUHOMS{sup R}) system canisters containing high burnup PWR and BWR fuel assemblies are now licensed for transportation as a payload in MP197HB after these canisters were in dry storage at the utility ISFSIs. Enhancements have been made to the MP197HB transportation package to improve thermal performance and also provide superior shielding capabilities. Design features of existing AREVA TN transportation packages such as radial fins, thermal shield, neutron resin and optimized gamma and neutron shielding were used to accommodate transportation of high burnup fuel assemblies with heat loads up to 32 kW. The transportation package is designed for exclusive use by rail, truck or marine transport. State-of-the-art analyses methodologies were used to demonstrate compliance with regulatory requirements of 10 CFR 71.55 for the high burnup fuel after dry storage at any ISFSI. Some of the issues addressed in the license application related to the transportation of high burnup fuel included potential degradation of fuel cladding, uncertainty associated with high burnup fuel cladding material properties, and larger radiation and thermal sources due to higher burnups. Bounding conservative assumptions were made where limited test data was available. AREVA TN's licensing approach did not rely on the cladding integrity of high burnup fuel cladding. Instead, the criticality, thermal and shielding analyses were carried out assuming high burnup fuel reconfiguration. With the approval of this transportation package for PWR and BWR canistered high burnup fuel from dry storage at ISFSIs, the utilities can now transport these canisters which are in dry storage at their sites. This is also a major step forward for the canistered fuel that is currently in dry storage at the shutdown plants because they can use this AREVA TN MP197HB transportation package to ship that fuel to any interim storage facility (ISF) or to a repository when they become available. (authors)

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DOE OSTI2015

Radiological characterization and Decommissioning of Research and Power Reactors - 15602

The production of the detailed activity and dose rate atlases (ADAs) is one of the indispensable measures in the decommissioning process of nuclear reactors which allows to determine the radiation field for optimal radiation protection as well as to quantify and characterize nuclear waste for disposal. High-detailed simulation models capable of performing neutron and radiation transport calculations at different levels of precision and details are applied for this purpose. Within the framework of the research project 'Decommissioning of Research and Power Reactors', funded by the German Federal Ministry of Education and Research (BMBF), and implemented by the Institute of Nuclear Engineering and Technology Transfer (NET) -RWTH Aachen, sophisticated simulation tools are developed. The enhances capability of these tools for the generation of the activity and dose rate distribution are tested and demonstrated on the shutdown German research reactor FRJ-2. To achieve the project goal different computational steps were required: 1. Development of a calculational CAD-based model for Monte Carlo neutron and gamma transport simulation using converter tool McCad 2. Design of the neutron source which reproduce the neutron distribution leaving the core 3. Enhancement of the simulation performance by using the hybrid deterministic Monte Carlo variance reduction methods 4. Calculation of N-flux distribution in the entire geometry 5. Development of couple codes between MCNP5 and the activation code ORIGEN2.2 This paper describes the main significant project milestones. (authors)

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DOE OSTI2016

Examining Supply Chain Resilience for the Inter-modal Shipment of Spent Nuclear Fuel and High Level Radioactive Materials - 16309

The U.S. Department of Energy (DOE) has a significant programmatic interest in the safe and secure routing and transportation of Spent Nuclear Fuel (SNF) and High Level Waste (HLW) in the United States, including shipments entering the country from locations outside U.S borders. In any shipment of SNF/HLW, there are multiple 'chains'; a jurisdictional chain as the material moves between jurisdictions (state, federal, tribal, administrative), a physical supply chain (which mode), as well as a custody chain (which stakeholder is in charge/possession) of the materials being transported. Given these interconnected networks, there lies vulnerabilities, whether in lack of communication between interested stakeholders or physical vulnerabilities such as interdiction. By identifying key links and nodes as well as administrative weaknesses, decisions can be made to 'harden' the physical network and improve communication between stakeholders. This paper examines the parallel chains of oversight and custody as well as the chain of stakeholder interests for the shipments of SNF/HLW and the potential impacts on systemic resiliency. Using the Crystal River shutdown location as well as a hypothetical international shipment brought into the United States, this paper illustrates the parallel chains and maps them out visually. (authors)

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DOE OSTI2017

Database dedicated to nuclear criticality events that occurred in fuel cycle facilities

Nuclear fuel cycle facilities can encounter events, resulting from human errors, failure of safety organization or technical failures. However, the application of the concept of defense in depth, especially the first two levels (prevention of abnormal situations and detection) requires to ensure safety in such undesirable situations. Furthermore, the periodic safety reviews of each nuclear fuel cycle facility involve the evaluation of the safety of these facilities with regard to the regulation evolution, the best safety practices, the facility evolution (modifications, aging...) and the global feedback. Part of this work is based on the analysis of the feedback and lessons learned from the events that occurred during the past period. This specific work indeed provides lessons about the operating conditions of the facility and highlights malfunctions that have occurred during operations, maintenance or restart after a shutdown. Taking into account this feedback, IRSN developed a database gathering the events which have affected the safety of French fuel cycle facilities (from conversion to waste storage, laboratories and experimental facilities, facilities under dismantling and decommissioning). Among all the recorded events, it is possible to extract in particular those impacting the criticality safety. In order to extend the analysis of these events, the IRSN criticality safety department developed a specific database dedicated to criticality risks. The 'criticality events' database is continuously updated on the basis of the declared events and their analysis by engineers from the criticality safety department. At this time, about 600 events are described and analyzed in this database. (author)

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DOE OSTI2018

Preliminary Planning of Radwaste Management for the Decommissioning of Chinshan NPP in Taiwan - 18201

The Chinshan NPP was the first commercially operated nuclear power plant in Taiwan since 1978. There are two BWR units in Chinshan NPP, which each unit has electrical output of 636 MWe. After 40 year's period of license, the final shutdown date of Unit No. 1 and Unit No. 2 will be in December, 2018 and in July, 2019, respectively. According to 'Nuclear Reactor Facilities Regulation Act' in Taiwan, decommissioning of nuclear reactor facilities should be implemented by the strategy of DECON (Decontamination followed by dismantling) within 25 years when the permission is approved by Atomic Energy Council (AEC). The decommissioning plan of Chinshan NPP has been permitted by AEC in the end of June, 2017. The implementation of Chinshan NPP decommissioning will be started while the Environmental Impact Assessment is approved by EPA. In this study, the Turbine Building in Chinshan NPP is selected as a case study for the preliminary planning of decommissioning. First, we thoroughly examine the inventory of all macro-components in the Turbine Building to estimate the waste amounts. Accordingly, the work breakdown structure (WBS) for dismantling of facilities is scheduled. Then, considering to minimize waste amounts, several suitable decontamination techniques, such as chemical and mechanical methods, which are developed by INER are considered to be adopted. As for the estimation of man hours of dismantling facilities, all events during the decommissioning tasks are classified into 16 work procedures and appropriative factors of corresponding team hours are gotten. In addition, the detailed methodology of the preliminary plan of decommissioning is also discussed. Finally, we hope the results of this planning can provide practical benefits to upcoming decommissioning work in Taiwan. (authors)

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DOE OSTI2018

Peculiarities of Transition from Operation to Decommissioning for the Chernobyl NPP - 18486

Usually the transition period is supposed to last around 5 years and to cover the eventual defueling of the reactor and the reactor cooling pool, completion of the characterization survey and preparation of all necessary documents to start the decommissioning activities in accordance with the decommissioning strategy chosen and the decommissioning plan approved. It was not the case for the Chernobyl NPP due to a number of reasons and obstacles, including delayed defueling due to unavailability of the storage capacities, lack of preparatory work due to prescheduled decommissioning, luck of experience of RBMK-type reactors decommissioning, influence of the destroyed Chernobyl-4 Unit and related works on site. All these factors not only delayed the completion of the characterization survey, but added complications to development of the decommissioning strategy and the decommissioning plan. De facto the issue of end state is still open as the end state for the destroyed Chernobyl-4 as well as the prospects of the exclusion zone transformation need further clarifications and justifications. As a result, the need to reconsider the standard concept of transition period was recognized already by the end of 2000, when Chernobyl-1 and Chernobyl-2 were out of operation for a few years. It was clear that the personnel had generally a bad perception of the prescheduled permanent shutdown decision and a low motivation to change and to learn. The first signs of organizational degradation were visible and the framework of the operational license did not fit well to prevent it. So it was decided to start the legal and organizational transformation into the decommissioning mode without waiting for polishing of the decommissioning strategy and decommissioning plan as in this case both documents seemed to need further adjustments anyway. The transition period (or the termination of operation as it called in the national nuclear safety regulations) was considered as the first phase of decommissioning to be covered by the decommissioning license. The detailed plan of activities for the transition period with proper safety justification documents was the main part of the relevant license application. Human resource management plan and adoption of the management system were important constituents as well. It was also considered reasonable to start some dismantling works already at this phase to avoid further delays and degradation, the works being carefully justified. Eventually such approach added to streamlining the decommissioning process and to avoid further organizational and technical degradation. Flexibility in transition from operation to decommissioning seems necessary in cases of prescheduled or post-accidental decommissioning but may be also beneficial for a scheduled decommissioning case. (authors)

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DOE OSTI2018

Replacement of Melter 2 at the Defense Waste Processing Facility (DWPF) - 18608

The original DWPF melter was replaced in 2002 with an identical unit, and radioactive operation of Melter 2 commenced in 2003. After nearly 14 years of service, a lower electrode failure of Melter 2 occurred in February 2017 which required its removal and replacement with Melter 3. The scope of the replacement outage involves several significant evolutions: - Shutdown of Melter 2 and placement into a safe state; - Remote external and internal inspection of Melter 2 to chronicle its condition prior to removal; - Disassembly and removal of Melter 2 from the process cell; - Placement of Melter 2 into a storage box and transport to a storage vault; - Preparation and initial testing of Melter 3 prior to transport to the DWPF; - Transport to the DWPF and installation of Melter 3 in the process cell; - Final startup testing and initiation of radioactive operation. Upon determination that repair of Melter 2 was not feasible, DWPF created an outage team to integrate schedules and resources needed to execute melter replacement as well as including maintenance scope for facility work that can only be performed when the melter is de-energized. The execution of this scope required integration with other mission critical outage activities already in progress, and planned to start, including initial facility tie-in scope for the new Salt Waste Processing Facility (SWPF). This unplanned addition of the melter replacement scope to the existing outage scope resulted in the longest facility outage period since the initiation of radioactive operations at the DWPF in 1996. Successful completion of this scope across all Liquid Waste facilities would require balance - for schedules, resources, priority, funding, etc. To overcome the significant challenges created by this additional scope, a WAR Room was established to ensure adequate integration and coordination of critical resources, and daily progress. The execution of this scope has required the integration of nearly two dozen work groups across multiple facilities at Savannah River Site (SRS), in addition to support from nearly a dozen vendors. The melter replacement outage was executed with a high degree of disciplined operation to ensure the facility was ready to resume glass pouring operation in time to support restart of the Liquid Waste program at the Savannah River Site (SRS). This paper will provide a brief overview of Melter 2 operational history. The execution of the removal and replacement operations will be described with a specific focus on lessons learned and opportunities where the outage execution diverged from the original Melter 1 replacement outage. It is expected to be of interest to vitrification facilities, as well as facilities, which require remote maintenance and replacement of large process equipment. (author)

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DOE OSTI2016

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the 80's at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Preliminary criticality calculations have been performed using both models (deterministic and stochastic). Results show the good agreement between different models. It should also be noted that deterministic solution could still be further improved by implementing super-homogenization factors and by increasing the transport calculation order.

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DOE OSTI2016

Convective Heat Transfer in Decay Heat Exchanger of Sodium-cooled Fast Reactor

Sodium-to-sodium heat exchangers such as the decay heat exchanger (DHX) are key components of the prototype generation [V sodium-cooled fast reactor (PGSFR). The DHX transfers the decay heat generated in the reactor core to the decay heat removal loop after a reactor shutdown. Shell-and-tube type with straight tube arrangement has been widely adopted for the sodium-to-sodium heat exchanger of the sodium-cooled fast reactor. Thus liquid metal heat transfer characteristics inside the tube and around the tube bundle have to be fully understood to design the sodium-to-sodium heat exchanger and evaluate its thermal performance accurately. Plenty of studies on the convective heat transfer of the liquid metals were conducted to present the Nusselt number (Nu) correlations with respect to the Peclet number (Pe). However, those correlations still deviated from each other considerably and few of correlations for low Pe range (less than 100) have been reported. Accordingly, it is difficult for a design code developer to choose a proper heat transfer coefficient correlation for the sodium-to-sodium heat exchanger particularly in a low flow rate range. Recently, a separate effect test for the DHX has been completed using the STELLA-1 facility at KAERI according to the STELLA program which aims at establishment of separate and integral test database (DB) for the design of the PGSFR. The heat transfer performance test of the DHX was successfully conducted under major operating conditions. The objective of this work is to suggest proper convective heat transfer correlations for tube- and shell-sides of the sodium-to-sodium heat exchanger based on the test DB obtained from STELLA-1. The Pe range of the analyzed experimental data is concentrated on less then 100. The adequacy of the suggested correlations is evaluated by applying to the DHX design code, SHXSA. Based on the DHX test DB of the STELLA-1, the proper convective heat transfer correlations for the tube- and shell-side of sodium-to-sodium heat exchangers were suggested. Two correlation combinations for each side were proposed and turned out to be reasonable through the evaluation using the SHXSA design code: Suggestion no. 1: - Lyon correlation (constant = 4) for tube-side; - Lubarsky-Kaufman correlation for shell-side. Suggestion no. 2: - Modified Lyon correlation (constant = 1.45) for tube-side; - Schad-modified correlation for shell-side The Pe ranges of the DHX test DB were 53 - 225 for the tube-side and 45 - 148 for the shell-side, but most of test conditions were concentrated on the Pe range of less than 100. Two correlation suggestions can be practically used for the design of sodium-to-sodium heat exchangers. However, by utilizing more test DB, it needs to be verified which one is closer to real phenomena. (authors)

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DOE OSTI2019

French Innovative Thermal Treatment Processes for The Management Radioactive Organic Liquid Waste - 19017

The nuclear industry as well as hospitals or some research laboratories produce Radioactive Organic Liquid Waste (ROLW) such as scintillation cocktails for medical imaging, labeled molecules for medical research or liquids from former activities. In order to be compatible with disposal facilities requirements, this waste must be inerted and stabilized in solid matrix. The route generally used is incineration in the Centraco facilities at the Marcoule site (Gard, France), ashes produced being cemented downstream. However, liquids whose chemical or radiological characteristics are not compatible with the incinerator operating rules (high level of {sup 14}C or halogen content, corrosive gas production etc.) cannot follow this path. The purpose of the MILOR project is then to propose alternative paths to treat all kinds of ROLW. Initiated in September 2017, this project has the technical objective of developing, over a four-year period, two pilot-scale mineralization processes. Two variants of complementary plasma technologies are evaluated: aerial plasma with the 'IDOHL' process and submerged plasma with the 'ELIPSE' process. Compared with other methods, thermal plasmas have unique characteristics such as rapid decomposition with high throughput, fast startup and shutdown, and high-energy delivery. The process called IDOHL consists of a process involving an induction plasma in which the liquids are directly introduced. It has been designed for organic liquids without mineral load and for treatment rates less than 0.5 l/h. This type of plasma allows the destruction of liquid organohalogens without any prior evaporation or treatment. Specific gas treatment systems have been developed to direct the gas produced to a dry or wet gas treatment system. One aim of the project is to subsequently install the process in a radioactive zone at Saclay DRF. The second called ELIPSE consists of a breakthrough process involving a submerged plasma at the heart of which organic liquids are also introduced. Current research has demonstrated that such a technique may enable the instantaneous and complete destruction of liquids with a wide variety of constituents, such as chlorine, fluorine, or phosphorus. The submersion solution offers many advantages: quenching and cleaning of combustion gases; filtering of the particles they contain; and cooling maintained for the entire process, which guarantees excellent corrosion control. An advantage of this type of design is that the gas treatment system can be reduced to a demister-condenser followed by a simple safety filter, thereby offering the additional advantage of an extremely compact treatment system. This design also allows the ELIPSE process to become by this way an embeddable process if required. The MILOR project, financed by the French State through investment program is a tripartite project bringing together Andra, in charge of radioactive waste management, the CEA in charge of leading scientific and technological researches to propose effective ways of treatment and the company A3i whose role is to ensure the subsequent industrial development. (authors)

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DOE OSTI2019

Environmental Remediation of Waste Management Areas at Chalk River Laboratories - 19130

The Government of Canada's nuclear legacy liabilities have resulted from ∼70 years of nuclear research and development (R and D). The majority of these liabilities are located at the CRL Site and consist of shutdown research buildings (including prototype and research reactors), plus a wide variety of buried and stored wastes, and contaminated lands. The Environmental Remediation Projects (ERP) branch are assessing the waste inventory in the Chalk River Laboratories Waste Management Areas in order to make informed and risk based decisions on remediating affected lands. This paper provides examples of contaminated sites at CRL that require remediation as well as examples of remediation activities already complete. The environmental remediation process followed at CRL and the development of the CRL site cleanup plan are outlined. Until suitable disposal facilities are available ERP will continue to characterize and assess contaminated sites, monitoring and managing the risks in place while preparing for full-scale remediation activities. (authors)

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DOE OSTI2019

Postclosure Safety Assessment for the Nuclear Power Demonstration Closure Project - 19158

Following 30 years in a safe shutdown state, Canadian Nuclear Laboratories (CNL) is proposing the final decommissioning of the Nuclear Power Demonstration Waste Facility (NPDWF). A Postclosure Safety Assessment (PostSA) has been prepared as part of this process. The present PostSA represents the second assessment, following an iterative approach. The PostSA considered Normal Evolution of the facility, as well as several Disruptive Events and other cases to show robustness and account for uncertainties. The results of the PostSA show that the NPDWF is a robust design that will contain and limit release of contaminants long into the future. (authors)

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DOE OSTI2019

Integrated Waste Treatment Unit Plan for Vessel and Canister Decontamination - 19283

The Integrated Waste Treatment Unit (IWTU) was designed to use a proprietary fluidized-bed steam reforming process to treat 3,400,000 L (898,184 gal) of radioactive sodium bearing waste (SBW). IWTU is currently in commissioning and startup phase prior to future processing of SBW. The current simulant SBW is introduced into the denitration mineralization reformer and transformed into a dry, granular product that is filtered and eventually transferred into a product canister. The intention of the facility is to process the SBW and store the product canisters in vaults to await shipping to a depository at a later date. The IWTU is a part of the Idaho Cleanup Project Core, located at the INL Site and operated by Fluor Idaho, LLC. During early operation and testing of the facility, it was observed that the original process equipment decontamination system did not possess the capability to recover from plant shutdowns that result from upset conditions or maintenance outages; thus, a need for a redesigned decontamination system was identified. This paper discusses current decontamination methods and their capabilities and areas where these methods fall short of decontaminating process equipment and product canisters. This paper discusses the design and purpose of the decontamination methods being tested and developed for implementation into the IWTU. (authors)

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