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DOE OSTI2019

Current Status of Decommissioning Plans and Chemical Decontamination of Metal Scraps in Chinshan NPP - 19304

The decommissioning of nuclear facilities has become a significantly attractive and worldwide topic due to the end of lifetime of many nuclear power plants within next coming years. In Taiwan, Chinshan nuclear power plant owned by the Taiwan Power Company will be the first candidate to be decommissioned according to the Nuclear Reactor Facilities Regulation Act. The Chinshan nuclear power plant located in New Taipei City has two units of 636 MWe boiling water reactors (BWRs) for the generation of 9 billion kWh of electricity per year. The date of cessation of operation of Unit No. 1 and Unit No. 2 will be December in 2018 and July in 2019, respectively. Based on the Regulations for the Review and Approval of Applications for a Decommissioning Permit of Nuclear Reactor Facilities, Taiwan Power Company have submitted a decommissioning plan of Chinshan nuclear power plant to the authority, Atomic Energy Council (AEC). After the review process, the decommissioning plan of Chinshan nuclear power plant has been permitted by AEC in 2017. Decontamination is generally utilized during the period of decommissioning of nuclear facility in order to minimize the radiation exposure of personnel and the cost of management of radioactive wastes. The successful decontamination also probably provides the benefits of lowering the level of radioactive wastes, facilitating potential releases, permitting the reuse of metals and facilitating the radiation principle of As Low As Reasonably Achievable (ALARA) management. Decontamination aims at the removal of radioactive materials from the contaminated surfaces by several technologies such as washing, heating, chemical reaction and mechanical process. The selection of proper decontamination technologies are depended on the requirement of the target materials and the purpose of the decontamination. Chemical decontamination has been proven to be an effective in reducing the radioactivity for metallic materials by using chemical reagents in contact with the contaminated surfaces. During the operation of Chinshan nuclear power plant, chemical decontamination was performed for Unit No.1 and No.2 in 2000 to decontaminate the Reactor Recirculation System (RRS) and Reactor Water Clean-up Systems (RWCU) by using LOMI-NP-LOMI process. Based on the previous experience of chemical decontamination, the dismantled metallic piping is provided as sample by Chinshan nuclear power plant for the evaluation of the feasibility of chemical decontamination. The provided metallic sample is composed of carbon steel with dose rate around 0.2 mSv/hr. In this report, a multistep decontamination process using strongly oxidizing solution and organic acid solution is performed to remove the contaminated oxide layer from the metal surface, and the resulting contact dose rate and decontamination factor (DF) are also shown to discuss the performance in relation to the process control factors. In this paper, the recent progress in decommissioning programs and preliminary results of the decontamination experiment are described to provide usable information regarding the future decommissioning for Chinshan nuclear power plant. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Texas authorizes disposal of large quantities of depleted uranium at wcs - 15626

The U.S. Nuclear Regulatory Commission (NRC) has been actively engaged in a rulemaking process that would revise Title 10 of the Code of Federal Regulations (CFR), Part 61, Licensing Requirements for Land Disposal of Radioactive Waste, to govern the disposal of large quantities of Depleted Uranium (DU) and Blended Waste (referred to as Unique Waste Streams). While the NRC plans to propose a Site Specific Analysis (SSA) rulemaking in early 2015, they have already issued guidance to Agreement States which identified general disposal criteria that could be used should a licensee propose to dispose of Unique Waste Streams before their rulemaking was completed. On August 5, 2013, Waste Control Specialists LLC (WCS) submitted a major license amendment application for review and approval to the Texas Commission on Environmental Quality (TCEQ). This major license amendment request proposed, among other things, authorization to dispose of large quantities of DU at its Federal Facility Waste Disposal Facility (FWF). Pursuant to Title 30 of the Texas Administrative Code (TAC), Rule 336.709, Technical and Environmental Analyses, disposal of waste must be sufficient to meet the radiological dose criteria for a minimum period of performance of 1,000 years after closure or the period where peak dose occurs, whichever is longer. These regulations are more stringent than those of any other state that hosts a facility authorized to dispose of Low-Level Radioactive Waste (LLW) in the U.S. On August 28, 2014, the TCEQ approved the major amendment request authorizing disposal of large quantities of DU. During the licensing review process, significant technical issues were successfully resolved ensuring that large quantities of DU could be safely isolated from the biosphere for long time periods after site closure. Many important lessons were learned during the licensing proceedings that have potential significant impacts to the licensed community. Namely that a modern, well designed disposal facility that is properly sited geologically and in an arid environment is ideal for ensuring that long-lived alpha emitting radionuclides are sufficiently isolated to protect public health for long into the future. The DOE is currently finalizing plans to complete a Supplemental Environmental Analysis (EA) to support shipments of DU for disposal at WCS' FWF as required under the National Environmental Policy Act of 1970. The DOE may begin shipments of DU for disposal to WCS FWF that was licensed and constructed to support disposal of waste that is the responsibility of the federal government as stipulated in the Low-Level Radioactive Waste Policy Amendments Act of 1985. This paper will present the regulatory history of events that initiated the NRC's rulemaking for Unique Waste Streams and the technical analysis that were conducted in support of WCS' major amendment that has been approved by the TCEQ. A description of the geological characteristics and engineering designs used to support the Performance Assessment will be included. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Geochemical data package for performance assessment calculations related to the Savannah River Site

The Savannah River Site (SRS) disposes of low-level radioactive waste (LLW) and stabilizes high-level radioactive waste (HLW) tanks in the subsurface environment. Calculations used to establish the radiological limits of these facilities are referred to as Performance Assessments (PAs), Special Analyses (SAs), and Composite Analyses (CAs). The purpose of this document is to revise the existing geochemical data package used for these calculations (Kaplan, 2016). This work builds on earlier compilations of geochemical data, referred to a geochemical data packages (Kaplan, 2007; Kaplan, 2010; Kaplan, 2016; McDowell-Boyer et al., 2000). This work is being conducted as part of the on-going maintenance program of the SRS PA programs that periodically updates calculations and data packages when new information becomes available.

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

Decommissioning of the Chernobyl NPP: Problems, Progress, Prospects - 18485

Decommissioning of the Chernobyl NPP provides a special case due to a number of peculiarities: it is prescheduled decommissioning, the facilities undergone a major accident with off-site contamination, no previous experience of decommissioning of RBMK-type nuclear power units, unavailability of proper spent fuel and radioactive waste management infrastructure, special types of waste. The most challenging issue is the final decommissioning of the Unit 4 destroyed by the major accident on 26 April 1986. Neither of four RBMK-1000 nuclear power units was in operation till the end of the design-based lifetime, Units 1 and 3 being in operation for 19 years each, Units 2 and 4 even less. The last operating unit, Unit 3, was permanently shutdown on December 15 2000, however the main obstacle to complete a complex technological and radiological survey was spent fuel left in reactor pools and in the core due to lack of capacity in the existing interim spent fuel storage of wet type (ISF-1). So removal of spent fuel from the core and reactor pools was considered as one of the priority tasks. While construction of ISF-2 (of dry type) was delayed due to different reasons, safety reassessment and modernization of ISF-1 enabled capacity increase and transfer of all non-damaged fuel assemblies into ISF-1. Thus Chernobyl NPP Units were fully vacated from spent fuel. Since that Chernobyl NPP Units are considered as the radioactive waste facility. Dismantlement of Units 1,2 turbine hall equipment is on the way and some progress is achieved in technical solution for fragmentation of long-size equipment. All dismantled equipment after decontamination and characterization is either released from regulatory control or disposed as solid radioactive waste. The details are discussed in the paper. Eventually the comprehensive technological and radiological survey of Chernobyl NPP Units had been completed and detailed decommissioning programme had been endorsed. The decommissioning strategy chosen for the Chernobyl NPP is based on the deferred dismantlement approach, so final dismantlement of all equipment and constructions is supposed after some period of decay storage. However, the issue of the final state of the site and its further use is still open. The main reason for it is the lack of clear picture as for the destroyed Unit 4, as technical solutions concerning its highly damaged and contaminated constructions and so called fuel containing masses (FCM) are not available yet. It could be reasonably predicted also that the internal part of the so called exclusion zone will not be released into non-restricted use in near future due to characteristics of its contamination. Possible implications of these and other factors on further decommissioning activities and some approaches to find solutions are discussed in the paper. (authors)

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

Challenges in the UK Generic Design Assessment Process - 17062

In the UK, Generic Design Assessment (GDA) is a voluntary step-wise joint assessment process undertaken by the Office for Nuclear Regulation (ONR) and the Environment Agency (EA) of a generic reactor design. The process focuses purely on the reactor design and does not consider any site specific issues. The key objective of GDA is for the regulators to consider the adequacy of claims, arguments and evidence (CAE) put forward by a requesting party (RP) to demonstrate the safety of its design in the UK context. This includes all CAE relating to management of spent fuel and radioactive waste. The CAE are generally applied to a reference design put forward by the RP that is substantially complete in terms of the overall concepts and functions of the major systems, having already been deployed elsewhere. GDA assessment encompasses a broad range of topic areas which include fuel and core design, management of spent fuel, radioactive waste management and decommissioning. Other technical disciplines also have valid interests in particular aspects of the management of spent fuel. This paper summarises the UK GDA process and the types of issues raised; in particular aspects of the proposed operation of liquid waste management systems and Spent Fuel Interim Storage are presented and the instructive resolution of the RP's original limited design scope to allow the regulators to carry out a meaningful assessment. Some examples are presented and useful lessons learned for future GDAs are discussed. (authors)

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

Future German R and D on HLW Disposal - 16519

In 2009, after the General elections, the new German Government showed a relatively positive attitude towards a prolonged use of nuclear power. Power plant shutdown was planned for the year 2032. Nuclear power was considered as a sort of transition energy technology until the energy supply would take place by the use of renewable energy sources. This was in line with the legal framework and the political strategy of the energy turnaround. Then, 17 nuclear power plants (NPP) were operational. The Fukushima incident heralded the sound of drastic changes in the German energy policy: Besides an intermediate stop of operation of reactors until a so-called stress test was accomplished, the general political decision was to phase-out nuclear earlier than planned and to implement faster the decisions regarding the energy transition policy. The shut-down of the last German nuclear power plant will occur in 2022. Accordingly, all the necessary legal changes were made. A consensus about the appropriate waste management strategy for heat generating waste (spent nuclear fuel (SNF) and vitrified waste (HLW)) was still pending at this time. Public and political discussions on this subject continued as in the years before. In July 2013, after intensive discussions, most of the parties represented in the German parliament agreed to eventually tackle the challenge of finding a repository by a consensual, transparent and science-based approach. This was to be codified in a law. In July 2013, the Site Selection Act passed the parliament. This law should pave the way for a new holistic process that addresses social, political, and scientific aspects of siting and the site selection process. The basics for this process should be provided by the 'Commission Storage of Radioactive Waste Materials' which was established according this law. Against this background discussions showed that scientific-technical R and D activities should be mirrored accordingly and R and D has to deal both with the future challenges of radioactive waste HLW disposal and related aspects. Considering this the Federal Ministry for Economic Affairs and Energy (BMWi), which is responsible for non-site specific R and D, created the new Research Concept 'Research for the management of radioactive waste', being fundamental and the frame for non-site specific R and D projects funded by BMWi. Key changes - mirroring the new requirements and different aspects - touch the issues: host rock research, i.e. all potential host rocks available in Germany are equally addressed and there is no prioritization, consideration of consequences of prolonged storage at NPP sites, alternative waste management strategies compared to direct disposal in a mined repository, and involvement of socio-technical aspects.This paper intends to highlight the key changes, its consequences, and will give a brief overview on first initiated R and D activities. (authors)

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

Packaging for Transport of the ILW-LL of EDF First Generation Power Plant Dismantling - 16517

EDF has a first generation decommissioning plan involving different types of reactors: Heavy Water Reactor, Natural Uranium and Gas Cooled Reactor, Fast Neutron Reactor and Chooz-A the first Pressurized Water Reactor (PWR) of the French decommissioning plan. The dismantling of these first generation reactors produces various types of radioactive waste, the most radioactive of them being the intermediate-level long-lived waste (ILW LL). After cutting and/or extraction from the reactors, these ILW LLs have to be transported to the EDF Conditioning and Storage Facility for Activated Waste ICEDA where they will be temporarily stored until a final depository solution is built. To transport these irradiated and contaminated metallic wastes, EDF decided to give ROBATEL Industries the responsibility of designing a new B- type package and obtaining the associated approval from the French Nuclear Safety Authority (ASN). A lot of issues had to be overcome to comply with the nuclear transport regulations and answer the ASN and its technical support IRSN's questions. Among them, was the difficulty to have a single type of package to transport unblocked waste coming from reactors of different technologies, sites in different stages of decommissioning and using different methods of waste production, conditioning and control. In the end, EDF had to apply for a specific approval for each dismantling site. These difficulties, some other technical issues encountered, and the decision to get first an approval for the transport of Chooz A's ILW LLs, are described here. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Scaling Input Distributions for Probabilistic Models - 19472

It is important to address scaling issues when building models to support risk-based decisions for radioactive waste disposal sites. Currently, probabilistic risk assessment and probabilistic performance assessment models are run for multiple realizations where each realization uses one value from each input distribution over the entire modeled time. The input distributions represent the current state of knowledge in the mean of the input variable over the spatial region of interest and over the time period of the model run. This paper focuses on the time factor in input distribution scaling because it presents more serious conceptual and distributional challenges than the spatial factor. One of the challenges in temporal scaling is the effect of non-linearity on propagation of distribution uncertainty through to uncertainty in results. Radioactive waste disposal models tend to have both linear and non-linear components in their models. Dose assessment calculations are typically represented by linear models, while contaminant transport models have linear and non-linear components. In principle, temporal scaling of input parameter distributions for a linear model is straightforward, but scaling of input distributions for non-linear models is not. This paper uses examples to investigate the effects of input distribution temporal scaling on simple linear and non-linear models. Different scaling scenarios were investigated by comparing three analysis tools: analytical mathematical representations, statistical software programming, and GoldSim Monte Carlo simulation software. R statistical software is an open source statistical programming language often used for statistical analyses and graphical representation of data. GoldSim is a robust probabilistic dynamic modeling platform used in performance assessment modeling of radioactive waste sites. A simple linear model, a quadratic model and a multiplicative model were the cases explored for each of these tools. For the three analysis tools considered and with each of the scaling cases, 1,000 realizations were implemented from distributions of the random variables of interest. This is a sufficient number of realizations for simple models to capture the model space. An annual temporal resolution of a 100-year simulation period was chosen for all models and analysis tools. The effects of temporal scaling were demonstrated by investigating two different options for time scaling. Option 1 was set up so that the input distributions represented the expected value on an annual time scale, and a new value from that distribution was chosen every year (at every time step which equaled a year) for 100 years. In Option 2, the input distributions were scaled to represent the expected mean value of the input variable over the 100 years of simulation, and one value from that distribution was chosen for the entire elapsed time. Results show agreement across the three analysis tools. As expected, the linear models have identical results for the two different scaling Options, and the non-linear models diverge in the results for different time scale implementation. The implications of this work point to the need for careful consideration of input distribution temporal scaling. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

The Development of a Small-Scale Project Model for Waste Management at the Port Hope Conversion Facility - 19126

Cameco Corporation's (Cameco) Port Hope Conversion Facility (PHCF) was previously owned by the federal Crown Corporation, Eldorado Nuclear, which held a significant inventory of legacy waste material at the time Cameco was formed in 1988. Over time, Cameco's ongoing operations contributed to an increase in the volume of low-level radioactive waste managed onsite. Since 2011, Cameco has completed annual waste management campaigns with the goal of accelerating the recycle/disposal of accumulated legacy radioactive wastes stored at its Ontario operations. With disposal options for low-level radioactively contaminated wastes in Canada limited due to cost considerations and no commercially accessible landfills licensed for radioactive material. In recent years, Cameco has started to ship marginally contaminated wastes (<500 ppm uranium) to the United States for disposal in an appropriate landfill owned by US Ecology in Idaho. Numerous successful campaigns totaling over 1,000 successful shipments, have been completed to transport drummed and/or bagged material by rail in gondola cars to the disposal facility in Idaho since 2011. All shipments have been completed in compliance with all environmental regulations, with a perfect compliance record maintained to date. To achieve this excellent compliance record, the campaigns require careful planning by both organizations, as well as deployment of personnel from the Idaho facility during the campaign for project management, characterization and/or radiation monitoring support. As the accumulated waste inventory has been reduced through these projects, the scope and duration of each campaign has also decreased. In the last few years, road transport directly to the Idaho facility has also been utilized. This option allows for the deployment of small projects that may be completed for a smaller inventory of a specific waste type and to manage newly-generated low-level radioactive wastes on an ongoing basis. These projects have ranged from a single standard van shipment to multiple trucks running continuous loops between the Ontario and Idaho or Ontario and an intermediary site and have been utilized for marginally contaminated wastes (<500 ppm uranium) as well as marginally contaminated wastes with additional hazard classification. Cameco and US Ecology personnel have collaborated to develop characterization and shipment procedures that enable these projects to be executed in full compliance with all Canadian and US regulations, without deployment of personnel from the Idaho facility, while maintaining adequate oversight by both organizations to ensure that the waste is safely and compliantly managed. These protocols have been implemented and optimized for the Port Hope Conversion Facility and will be rolled out to other Cameco facilities in Ontario through 2019. Cameco expects to utilize the small project model for the long term to ensure that eligible marginally contaminated wastes (<500 ppm uranium) are managed appropriately to ensure that once the legacy material is eliminated, it remains so. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Test for Reconditioning RA Waste with Simulated Bitumen and Concrete in a 1,2 MW Plasma Test Facility - 20092

The operation and maintenance of nuclear power plants, the non nuclear fuel cycle, etc generate low-level radioactive waste which, along with the historical radioactive waste from past nuclear activities, needs to be treated and stored, awaiting final disposal. Plasma technology offers a very effective way of treating this waste with a high volume reduction factor (VRF), free from organics, liquids and moisture, and meets without a doubt the acceptance criteria for safe storage and disposal. By means of a plasma beam of approximately 5000 deg. C, the inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes while the organic material is gasified, oxidized and purified in an off-gas cleaning system. First the paper describes the new full-scale Plasma Melting Facility (PMF) at the Kozloduy Nuclear Power Plant in Bulgaria which was taken in nuclear operation In May 2018. The plant has a capacity of 250 tons per year and the maximum contact dose rates of the incoming waste is 2 mSv/h. Different mixtures of radioactive waste packed in 200 l drums were successfully treated resulting in a glassy slag free from liquids and organic material with an important volume reduction factor (VRF). The Project was co-financed through a grant by Kozloduy International Decommissioning fund (KIDSF) administrated by the EBRD through Bulgarian national funding. Plasma is a suitable technology for treatment of problematic waste or even reconditioning waste so Belgoprocess was contracted to do plasma tests with simulated conditioned waste types. One can do tests on a laboratory scale on smaller samples and torch capacities of e.g. 50 kW but Belgoprocess wanted to do more realistic and reliable tests. So Belgoprocess contracted Phoenix Solutions Co who has a full-scope test facility equipped with a 1200 kW plasma torch for full-scope treatment of simulated conditioned waste. For a first confidential contract simulated 200 l (55 gallon) bitumen drums were treated. The drums contained different pucks of compacted waste such as rags, used filters, granulates, etc. The pucks were stacked in the 200 l drums and subsequently embedded with bitumen. A total of 6 drums were treated in the plasma facility. For a second contract simulated homogeneous 200 l (55 gallon) concrete drums with on the one hand concentrates and on the other hand spent resins were selected. A total of 6 drums with concrete and spent resins were treated and melted in the plasma testing facility. The paper describe the test facility, volume reduction factor (VRF) of different waste streams and most important parameters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

DECOVALEX-2019 (Task E Final Report)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. Later it was realized that these processes also play a critical role in other subsurface engineering activities, such as subsurface CO 2 storage, enhanced geothermal systems, and unconventional oil and gas production through hydraulic fracturing. Research teams from many countries (e.g., Canada, China, Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the United States) various institutions have participated in the DECOVALEX Project over the years, providing a wide range of perspectives and solutions to these complex problems. These institutions represent radioactive waste management organizations, national research institutes, regulatory agencies, universities, as well as industry and consulting groups. This document is the final report of Task E which was proposed and coordinated by Andra, the National Radioactive Waste Management Agency in France, presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. The purpose of Task E of the DECOVALEX-2019 project is to investigate upscaling THM modelling from small-scale experiments (some cubic meters) to full-scale experiments (some ten cubic meters) and finally to the scale of the waste repository (cubic kilometers). To achieve this aim, the data of two in-situ heating experiments performed by Andra (the French National Radioactive Waste Management Agency) in the Meuse/Haute-Marne Underground Research Laboratory (MHM URL) have formed the basis for the understanding of the THM behavior of the COx at different scales. The first experiment provided the reference values of the THM parameters by means of a calibration exercise and they were used for a blind prediction and an interpretative analysis of the second one.

58 GEOSCIENCES↗
DOE OSTI2020

The Importance of Inventory for Informing Waste Strategy - 20280

A waste inventory aims to present the best available information about the nature, quantity and location of radioactive wastes and materials at a facility, site or in a country. Quality inventory data is necessary for developing appropriate strategies and plans for managing radioactive wastes and materials, with high standards of protection for people and the environment. The inventory of wastes will change over time due to a wide range of technical, commercial or policy reasons, for example, if reactor lifetimes are extended, if decommissioning strategies change or if new technologies are introduced for managing waste. Inventory information can also improve as better waste characterisation data becomes available. At a facility or site level, good quality inventory data are needed to underpin decision making about the waste management options and to demonstrate compliance with the chosen treatment or disposal route. At a site and country level, knowledge of the inventory can drive the improvement to waste strategy; waste management asset and service planning; and implementation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Ceramic–Metal (Cermet) Composites: A Review of Key Properties and Synthesis Methods Focused on Nuclear Waste Immobilization

Here this paper reviews key properties, applications, and examples of ceramic-metal composites (cermets), and metal matrix composites (MMC) with emphasis on their applicability as waste forms for immobilizing nuclear waste. While the literature is mature for vitrified and cementitious radioactive waste forms, cermet materials have not received adequate attention as potential candidates for immobilizing nuclear waste. To promulgate this effort, this review connects cermet and MMC design, such as hardened tools, with the chemistry of radioactive waste streams. A discussion on certification and qualification standards for cermet waste forms, literature gaps, and “how-to” sections on cermet processing techniques. Key parameters discussed include thermal conductivity, chemical durability, and waste loading, as well as examples for metal-containing waste forms. As cermet waste forms gain momentum within the community, this review paper aims ensure the end-of-life nuclear fuel cycle is addressed from a materials and waste disposition perspective.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Using Digital and Web-Based Platforms to Enhance Nuclear Power Plant Knowledge Transfer - 20203

As the workforce ages, the loss of experienced personnel can result in a loss of important knowledge. Updating Electric Power Research Institute (EPRI) research and guidance with the latest technology developments and operating experience is an important method to transfer knowledge between the retiring and incoming workforce. Studies suggest that the incoming workforce is accustomed to using digital and/or internet-based platforms to find information and may have a higher comfort with these types of platforms than they have with static, written reports. Additionally, the ease and speed of updating and delivering content through digital formats may offer an effective method for knowledge transfer and combining information from multiple reports into a single access point. EPRI has recently produced an online Decommissioning Hub available to members and is working on replicating the effort in the area of radioactive waste management called the Radwaste Web References. EPRI has extensive guidance in both disciplines. The Decommissioning Hub uses a wiki format to provide a platform that allows for both the updating of that guidance resulting from new research and new operating experiences, and combining related topical information from multiple reports. It also allows users to access the information based on key words and topical searches, and can be a tool for users to input experience to share with others. The Decommissioning Hub is accessed via a web-based portal and includes topical pages; experience summaries; a question and answer capability; and a smart search function of all EPRI decommissioning reports. Use of a wiki format allows users to input new experiences or modify existing experience summaries to add additional detail. EPRI moderates all content. The experience data base has been initially populated with information available in published EPRI reports and unpublished information currently available within EPRI. Moving forward, population of the database will continue with information available within the open literature, new information developed from current decommissioning projects, and user inputs. The web site became active in late 2017, and a major update will be issued in early 2020. Larger scale updates are planned biannually, and smaller-scale updates will occur continuously. Capturing and applying lessons learned from industry experiences is a hallmark of the global nuclear industry. This is even more important in the decommissioning technology area since very few plant staff have relevant experience and thus there is a steep learning curve. Moreover, decommissioning involves a number of complex, non-routine and specialized tasks. The web-accessible and searchable experience database provides an easy-to-use resource for capturing and disseminating practical experiences from completed and ongoing decommissioning projects. The EPRI Radwaste Web References will build upon the philosophy and objectives of the Radwaste Desk References that were published in the 1990's. The purpose of the original Radwaste Desk References was to provide nuclear power plant professionals with a how-to manual for managing radioactive waste. The Radwaste Web References will adopt this original purpose and will be a wiki type web site that provide fundamental practical and theoretical information, operating experience (OE), and information about advanced technologies and methods. As much as practicable, features that allow for interaction between users to share OE, best practices, and benchmarking will be added. Where they exist and are appropriate, the various EPRI Guidelines documents on radwaste topics will be updated and incorporated into the Radwaste Web References. The first Radwaste Web Reference will be on Low Level Waste Characterization and will build upon the EPRI Low Level Waste Characterization Guidelines (EPRI Report TR-107201, 1996) and will be available in 2020. The next Radwaste Web Reference will be on the topic of liquid radioactive waste processing and management, tentatively set for unveiling in 2021. This paper summarizes the EPRI's previous experiences with web-based platforms and future plans in this area. (authors)

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

Life Extension of Nuclear Plant Facilities through Monitoring and Detailed Analysis - 16175

The UK has accumulated a substantial legacy of radioactive waste from various civil nuclear and defence programmes. Presently, however most of the UK radioactive waste is generated from nine power stations which generate approximately one sixth of the of the country's electricity. Irradiated fuel is removed from these reactors and safe management of the radioactive solid and liquid effluents is required. The safe management of these effluents is essential. Sellafield, in West Cumbria UK, contains two reprocessing plants, one designed to process nuclear fuel from Magnox reactors (CO{sub 2} cooled with Uranium bar fuel) and the other to process nuclear fuel from Advanced Gas cooled Reactors and Pressurized Water Reactors. The reprocessing stage enables extraction of fissile material for recycling. This leaves a volume of dilute but high-level waste. The high-level radioactive liquid waste is concentrated to reduce its volume and then incorporated into borosilicate glass, using a process called vitrification. The glass is then poured into stainless steel containers and placed in long term storage. This required concentration of HLW liquor is achieved using a suite of aging evaporators. All evaporators are of similar design constructed from corrosion resistant stainless steel having dimensions height ∼11 m, internal diameter 3.05 m and capable of holding 13.7 m{sup 3} of liquor. The basic principle of operation is to heat the dilute liquor under partial vacuum, at constant volume, until the target concentration is achieved. Heating of the liquor is achieved using steam heated coils within the evaporator, and steam in the jacket which surrounds the lower half of the evaporator vessel. The evaporator is then brought to atmospheric pressure and the concentrated liquor is ejected whilst hot. The evaporator is then cooled with water cooled coils and a water filled jacket. New dilute liquor is then added for a new batch. The process is then repeated. As the evaporators reach the end of their original design life accurate prediction of the current condition of nuclear plant facilities is essential to safely justify continued plant operation by ensuring structural integrity. This poses a significant challenge as many aging UK nuclear facilities were not inherently built with inspection in mind. The potential life limiting components of the HA evaporators at Sellafield are the surfaces that contact the product liquor at elevated temperatures i.e. the steam heated surfaces of the coils within the evaporator vessel and the vessel base and wall surrounded by the steam jacket. Each heating component is critical to the operational life of the evaporator - the removal from service of a coil heating component significantly reduces the operational flexibility and remnant life of this plant. Over the last decade a variety of cutting-edge expertise and methodologies have been utilised to accurately measure and then predict the current condition of the plant and also to predict its remnant life. This paper provides an overview of the development of several techniques deployed, namely non-destructive testing/inspection, in-house software development and finite element modelling. These methods have been validated by lab-scale or full-scale experiments along with historical plant data. These techniques have been evolved and refined for over a decade, providing vital insights into the physical behaviour of this complex reprocessing system. This enables not only accurate assessment of the current condition of the plant equipment but also allows scientifically underpinned predictions to be made of how it will behave into the future. This is critical for Sellafield as this knowledge feeds into their safety case to the Office of Nuclear Regulation to justify continued operation. It is concluded that this integrated approach provides a robust analysis tool which has driven change and optimisation of plant procedures ensuring continued safe operation whilst extending the life of the facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Guidance on Good Practice for Waste Management in Decommissioning - 19080

The World Nuclear Association (WNA) is an international organization that represents the global nuclear industry. Its mission is to promote a wider understanding of nuclear energy among key international influencers by producing authoritative information, developing common industry positions, and contributing to the energy debate. The Waste Management and Decommissioning Working Group of WNA monitors developments and shapes industry positions in radioactive waste management in nuclear decommissioning. The Working Group promotes accepted international principles and methods, including the sustainable management of natural resources, appropriate re-use and recycling of material, as well as safe and cost-efficient disposal of radioactive waste. In 2015, the WNA decided to identify international good practices in decommissioning and waste management and to appoint a group of international experts to produce a report bringing together this knowledge. The report is aimed at policy- and decision-makers, as well as nuclear professionals looking for guidance or facing decommissioning challenges related to waste management. The scope of the report covers commercial nuclear plants and nuclear research facilities, including legacy plants being decommissioned after normal operation. Although light water and graphite reactor types are mainly considered, the methods described can be applied to other nuclear facilities. The report is due to be published in February 2019. This paper summarizes its findings and recommendations. (authors)

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DOE OSTI2017

The Status of LILW Complex Disposal Facility Construction in Korea - 17123

In this paper, the experiences and directions of the construction of the LILW disposal facility in Korea are discussed. The LILW disposal facility is being constructed in Gyeongju-si, Korea. The disposal facility is officially called 'Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (WLDC)'. It can dispose of 800,000 drums (hereafter, 'drums' means '200 liter drum equivalents' unless specified otherwise) of radioactive waste in a site of 2,060,000 m{sup 2}. The first stage of the LILW facility construction was completed in June 2014, and approved for use by Nuclear Safety and Security Commission (NSSC) in December 2014. The first stage of the LILW facility is a type of underground silo with disposal capacity of 35,200 m{sup 3} (100,000 drums). The first stage of the LILW facility consists of a construction tunnel, an operation tunnel, an access shaft and six silos, etc. Silo for final disposal is located 80∼130 m below sea level, 25 m in diameter and 50 m in height. The capacity of each silo is approximately 16,700 drums. The second stage of the LILW facility is under construction to be completed by December 2019. Site construction is currently underway. The second stage of the LILW facility is a type of near-surface disposal with disposal capacity of 32,875 m{sup 3} (125,000 drums). This stage of the LILW disposal facility consists of an underground gallery, three movable cranes and twenty vaults, etc. Vault for final disposal is located 70 m above sea level, 20 m in width, 20 m length and 10 m in height. The capacity of each vault is approximately 6,250 drums. These stages of the LILW disposal facilities are together called LILW complex disposal facility. It's the world's first LILW disposal facility with two different types of disposal facilities in one site. For efficient operation, these disposal facilities share a reception and storage building, waste water treatment system, measurement and monitoring system, etc. After construction of the first two stages is completed, LILW disposal facilities will be incrementally expanded, and as the whole LILW disposal facilities are completed, all LILW in Korea will be managed safely in WLDC. (authors)

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DOE OSTI2016

Evaluation and Impacts of Mercury in the SRS Liquid Waste System - 16121

The Savannah River Site (SRS) Liquid Waste System (LWS) stores radioactive waste in 43 underground tanks. The radionuclides in the waste are removed through a series of separation processes and the low-level fraction is immobilized in a grout waste form while the high level fraction is disposed in a glass waste form. Mercury originated from decades of canyon processing (used as a catalyst for dissolving the aluminum cladding of reactor fuel) and is present throughout the LWS (∼60 metric tons). Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. Mercury is removed from the LWS in many ways including: 1) the Tank Farm evaporator condensates; 2) at the Defense Waste Processing Facility (DWPF) where it is steam-stripped and removed during the feed preparation processes; 3) it is removed at the Effluent Treatment Plant (ETP); and it is immobilized in the low-level, grout waste form. An integrated, system-wide evaluation and assessment of mercury behavior in the LWS is being performed. The recently completed Phase I evaluation focused on mercury inventory and speciation in the LWS; mercury holdup and chemical processing behavior; mercury impact Identification, including worker safety and equipment degradation; and mercury removal and disposal options. The mercury removal system at the DWPF has been unable to remove mercury successfully due to equipment and chemistry challenges since the 2008 timeframe. Phase I evaluations indicate that approximately 43 percent of the mercury in waste sent to the DWPF is now returning to the Tank Farm in the recycle stream and is concentrating to higher than expected levels (∼500 mg/l) around the 2H evaporator system. The chemical behavior of mercury is further complicated by the formation of higher than expected quantities of organomercury species, specifically methyl mercury, now present in the DWPF recycle evaporator system (2H Tank farm evaporator), which is used to volume-reduce the recycle stream. Phase II activities are building on the Phase I mercury activities, including additional sampling and characterization activities, a re-assessment of overall system knowledge, ranking and prioritizing critical gaps/information in mercury behavior across the flowsheet, assessing impacts of mercury removal and disposal options, and documenting an action plan for overall mercury management in the LWS. Phase II activities include two System Engineering Evaluations to 1) 'Reestablish Mercury removal Capability within DWPF', and 2) 'Determine the Best Alternative Mercury Removal Location with the LWS'. This paper provides the results of the Phase I and II mercury evaluations for the SRS LWS. (authors)

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