Advanced Reactor Licensing
This is the licensing presentation for the Waste Management Symposia 2023. It covers licensing activities, LMP, and manufacturing licenses.
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Can we use Pictograms in a Long Term Communication System Related to a Radioactive Waste Repository? - 18301
Pictograms are often mentioned as a possible tool for sending messages about radioactive waste repositories in a far future, when the understand ability of our languages will have been lost. But they have also limits, related to the use of conventions that may fade over time. The goals of a signaling system for the long term may be (1) to alert (raise attention), (2) to inform of an underlying danger due to the existence of radioactive wastes, (3) to induce a favorable behavior (provoke or deter actions) in order not to disturb the performance of the repository. The quest for trans-cultural and trans-generational performance requires that some universal semiotic features be identified, shared by human beings of various cultures nowadays, and from far in the past to the present. Considering the typology of colors in various languages, the use of colors in ancient rock art and in children's drawings, a dedicated role may be assigned to each of the three colors, white, black and red/ocher, with a high potential for sustainable meaning: 1) red background = pay attention to this object because of the simple presence of the red color, that is to say the first color felt as color; 2) white inscription = absence of the object / action represented; 3) black inscription = presence of the object / action represented. The representation of a human body in movement inside a context is also understandable by people from various cultures. The combination of these 'semiotic primitives', used as relevant semantic features of the object and/or the experience to be represented, and 'topo-sensitive' criteria, used for pictogram production processes, would thus become a way to conceive a trans-cultural and trans-generational sign system for the signaling of radioactive danger. The choice of the stone could guarantee the durability of the support; the color could guarantee its visibility; and both could guarantee the transmission of the message. Combining these criteria could open the way to a long term communication system about radioactive waste repositories. (authors)
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Waste Information Management System with 2017-18 Waste Streams - 18302
The Waste Information Management System (WIMS) 2017-18 was updated to support the Department of Energy (DOE) accelerated cleanup program. The schedule compression required close coordination and a comprehensive review and prioritization of the barriers that impeded treatment and disposition of the waste streams at each site. Many issues related to waste treatment and disposal were potential critical path issues under the accelerated schedule. In order to facilitate accelerated cleanup initiatives, waste managers at DOE field sites and at DOE Headquarters in Washington, D.C., needed timely waste forecast and transportation information regarding the volumes and types of radioactive waste that would be generated by DOE sites over the next 35 years. Each local DOE site historically collected, organized, and displayed waste forecast information in separate and unique systems. In order for interested parties to understand and view the complete DOE complex-wide picture, the radioactive waste and shipment information of each DOE site needed to be entered into a common application. The WIMS application was therefore created to serve as a common application to improve stakeholder comprehension and improve DOE radioactive waste treatment and disposal planning and scheduling. WIMS allows identification of total forecasted waste volumes, material classes, disposition sites, choke points, technological or regulatory barriers to treatment and disposal, along with forecasted waste transportation information by rail, truck and intermodal shipments. The Applied Research Center (ARC) at Florida International University (FIU) in Miami, Florida, developed and deployed the web-based forecast and transportation system and is responsible for updating the radioactive waste forecast and transportation data on a regular basis to ensure the long-term viability and value of this system. (authors)
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Risk perception and social representation on radioactive waste: a case study with students of IPEN
Proper management of radioactive waste in teaching and research laboratories involves not only techniques and procedures for the treatment, storage and disposal of these materials but also the responsibility of all those who handle them. Responsibility begins with with methods that generate less waste and with the correct segregation of contaminated materials prior to disposal, In this context, the present study aimed to know, identify, understand and analyze the perception of risk and social representations about radioactive waste and its management by students and researchers of the Nuclear and Energy Research Institute. According to Renn (1992), two of the approaches to risk perception are cultural and sociological, which served as the basis to the present study. Within this analysis, the concept of risk is understood as a phenomenon and, therefore, the phenomena experienced by the individuals are related to their life histories, social relations, religious beliefs and academic formation, elements that interfere in the perception of risk. The method used to collect data was free evocation, with participants receiving the following instructions: 'Please write five words that pop up into your mind when you listen to the words radioactive waste and, sequentially, write a sentence using these words'. All participants received instructions of the voluntary character of the participation and of the confidentiality of the data. A total of 73 students and researchers were interviewed, including scientific initiation, masters degree, doctorate and postdoctoral researchers. 353 words were mentioned, of which 209 were different from each other. After analyzing the sentences, the words were inserted in 12 categories and later analyzed in the software INVOCADO, which provides a diagram with the possible structure of the social representations of the group studied. The analysis of the four-quadrants and their categories, as well as the mentioned words that compose them, are therefore consistent with the universe of the public analyzed. While it is not the totality of students dealing with radioactive materials, in general these individuals understand the importance of technology and that the adoption of safe practices ensures the successful development of research. Future developments: Analyze quantitatively and qualitatively the other questions of the questionnaire. Analyze the data and information obtained from the participant observation, which consisted of a follow-up of three students in their laboratory practices. Propose to the institute a manual of good practices that can be used by the students in their laboratory practices.
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Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453
The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non-time critical removal action for the In Situ Decommissioning (ISD) of the 105-C Disassembly Basin. ISD consisted of stabilization/isolation of remaining contaminated water, sediment, activated reactor equipment, and scrap metal by filling the DB with underwater non-structural grout to the appropriate (-4.877 meter) grade-level, thence with dry area non-structural grout to the final -10 centimeter level. The roof over the DB was preserved due to its potential historical significance and to prevent the infiltration of precipitation. Forced evaporation was the form of treatment implemented to remove the approximately 9.1 M liters of contaminated basin water. Using specially formulated grouts, irradiated materials and sediment were treated by solidification/isolation thus reducing their mobility, reducing radiation exposure and creating an engineered barrier thereby preventing access to the contaminants. Grouting provided a low permeability barrier to minimize any potential transport of contaminants to the aquifer. Efforts were made to preserve the historical significance of the Reactor in accordance with the National Historic Preservation Act. ISD provides a cost effective means to isolate and contain residual radioactivity from past nuclear operations allowing natural radioactive decay to reduce hazards to manageable levels. This method limits release of radiological contamination to the environment, minimizes radiation exposure to workers, prevents human/animal access to the hazardous substances, and allows for ongoing monitoring of the decommissioned facility. Field construction was initiated in August 2011; evaporator operations commenced January 2012 and ended July 2012 with over 9 M liters of water treated/removed. Over 8,525 cubic meters of grout were placed, completing in August 2012. The project completed with an excellent safety record, on schedule and under budget. (authors)
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Development of disposal container designs for disposal of high level waste and legacy spent fuel in the UK - 15471
Radioactive Waste Management Limited (RWM), a fully owned subsidiary of the Nuclear Decommissioning Authority (NDA), is responsible for the implementation of geological disposal in the UK. Three illustrative concepts of a geological disposal facility (GDF), corresponding to three generic geological environments - higher strength rock, lower strength sedimentary rock and evaporite rock - have been developed to demonstrate the viability of geological disposal of intermediate level waste, high level waste and spent fuel. It is envisaged that HLW which has been encapsulated in Waste Vitrification Plant (WVP) canisters, advanced gas-cooled reactor (AGR) SF which has been stripped of their spacers and packaged into slotted fuel cans, and PWR SF will be packaged into disposal containers for disposal. A set of schematic designs of these disposal containers was developed in the late 2000's. The design concept for the higher strength rock environment was based on the joint Posiva/SKB disposal container design, and the design concept for the lower strength sedimentary rock and evaporate rock illustrative environments was based on the disposal container concept from Switzerland's National Cooperative for the Disposal of Radioactive Waste (Nagra). The size, mass and dimensions of the disposal containers in the scheme designs vary significantly. The construction method and the lid closing method also vary between the disposal containers for the different waste contents. A design study has indicated that optimization of the number of HLW WVP canisters and the number of AGR slotted fuel cans in the disposal containers, standardizing construction method and closure arrangement across the containers for each geology type, providing handling features at both ends of the disposal container, and balancing of shielding around the containers, could bring significant advantage to the overall GDF programme by reducing the number of disposal containers, reducing GDF footprint, and reducing the extent of excavations, spoil generated and materials that are required for construction. A project to develop the scheme designs to conceptual level was carried in 2011 to 2013. In this project, the above changes were taken into account, technical underpinning of the designs was carried out and manufacturing method was considered. Two disposal container variants for different host rock environments have been developed: - Variant 1, for compatibility with the higher strength rock illustrative disposal system design. This is constructed from copper, with a cast iron insert and is based on the joint Posiva/SKB disposal container design. - Variant 2, for compatibility with the lower strength sedimentary rock and evaporite rock illustrative disposal system designs. This is a steel container based on the NAGRA disposal container design. Each variant has three internal configurations to accommodate 3 HLW vitrified canisters, 16 consolidated cans of AGR SF and 4 PWR spent fuel assemblies. These designs have been shown to bring significant benefit to the GDF designs, bringing a significant reduction to the operating programme and GDF programme cost, including reducing the number of disposal containers by 33% for HLW, 50% for AGR SF and reducing the size of the GDF footprint by 25%. (authors)
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Further Development of an Execution Strategy Analysis (ESA) Capability and Tool for the Management of Spent Nuclear Fuel (SNF) - 17146
An Execution Strategy Analysis (ESA) capability and tool are being developed to evaluate alternative execution strategies for the potential future deployment of a consolidated Interim Storage Facility as part of an integrated waste management system. Application of the ESA approach not only leverages but also goes beyond traditional project analysis tools. The ESA tool allows for ongoing performance assessment of the evolving project execution plans that take into account significant assumptions, risks, and uncertainties throughout the project lifecycle. Development and application of the ESA capability and tool continued in 2016, primarily focusing on enhancing its flexibility to model a range of scenarios regarding how an integrated system for managing spent nuclear fuel and high-level radioactive waste could be deployed. (authors)
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Addendum 2 to the Closure Report for CAU 577 Area 5 Chromium Containing Waste Disposal Cells, NNSS, Nevada
Corrective Action Unit (CAU) 577, “Area 5 Chromium Containing Waste Disposal Cells,” includes five low-level waste cells at the Nevada National Security Site Area 5 Radioactive Waste Management Site where buried waste received from Nuclear Fuel Services, Inc., was subsequently determined to contain chromium that exceeded the toxicity characteristic leaching procedure regulatory limit, which would require the waste to carry hazardous waste code D007. CAU 577 was created to satisfy the requirements of the Settlement Agreement (SA) executed between the Nevada Division of Environmental Protection (NDEP) and the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) on April 25, 2019 (NDEP 2019). The SA required that the chromium-containing waste received from NFS would be addressed following the closure process laid out in the Federal Facility Agreement and Consent Order (FFACO). The FFACO process ensures proper closure of the chromium-containing waste and documentation of that closure through FFACO-type documents. The following three Corrective Action Sites (CASs) were closed, and their closure was documented in the Closure Report for Corrective Action Unit 577: Area 5 Chromium Containing Waste Disposal Cells, Nevada National Security Site, Nevada, DOE/EMNV--0030, dated September 2021 (U.S. Department of Energy [DOE] Environmental Management [EM] Nevada Program 2021a): • CAS 05-21-02, Waste Disposal Cell 12 • CAS 05-21-03, Waste Disposal Cell 15 • CAS 05-21-04, Waste Disposal Cell 17 Closure of the following CAS was previously documented in the Addendum to the Closure Report for Corrective Action Unit 577: Area 5 Chromium Containing Waste Disposal Cells, Nevada National Security Site, Nevada, DOE/EMNV--0030-ADD, dated September 2022 (DOE EM Nevada Program 2022): • CAS 05-21-05, Waste Disposal Cell 20 This second addendum to the Closure Report documents the closure activities that have occurred for CAS 05-21-06, Waste Disposal Cell 21. This is the last CAS in CAU 577. The final waste shipment was placed in the waste disposal cell on November 21, 2022. Following this, closure activities began on November 28, 2022, and were conducted according to the Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) for CAU 577 (DOE EM Nevada Program 2021b). The following closure activities were performed: • Constructing an engineered evapotranspiration cover • Installing two subsidence monuments and vadose zone monitoring equipment • Seeding the cover with a mixture of native plant species • Installing four concrete monuments on the corners of the cover and placing two use restriction (UR) warning signs on each monument These activities fulfill applicable federal and state regulations for closure of CAS 05-21-06 and minimize potential future exposure pathways to buried waste. Completed closure activities are also consistent with closure of the nine historical Resource Conservation and Recovery Act (RCRA) units included in Section 10.2.2 of the RCRA Permit that governs hazardous waste management activities at the Nevada National Security Site (Permit NEV HW0101) (NDEP 2023). UR documentation for this CAS is included in Appendix B of this report. The post-closure plan is presented in detail in the CADD/CAP for CAU 577 (DOE EM Nevada Program 2021b), and the requirements are summarized in Section 5.2 of this document. In accordance with paragraph 5D of the SA, a request to incorporate the requirements for post-closure monitoring of CAU 577 was included with the permit application for RCRA Permit NEV HW0101 that was submitted in January 2022 (NNSA/NFO 2022). The request included the post-closure requirements for the three CASs that had been closed at the time of submittal of the application as well as requirements that would be implemented upon future approval of closure of the remaining two CASs. All CAU 577 post-closure monitoring requirements have been captured in the April 4, 2023, Revision 7 of the RCRA Permit (NDEP 2023). As the RCRA Permit NEV HW0101 has been revised since the submittal of the original CAU 577 Closure Report (DOE EM Nevada Program 2021a) and Addendum 1 (DOE EM Nevada Program 2022), the post-closure requirements in this Addendum 2 report do not align with the previous documents. Specific changes resulting from the issuance of Revision 7 of the RCRA Permit are discussed in Section 5.2 of this report. The requirements in this report are consistent with the current permit (NDEP 2023) and supersede all requirements listed in the CAU 577 Closure Report and Addendum 1. All CAU 577 post-closure requirements should be conducted in accordance with the version of the RCRA Permit that is current at the time of the activities being performed. The DOE EM Nevada Program is requesting a Notice of Completion from NDEP for closure of CAU 577. Although CAU 577 is not a legacy site, the FFACO process is being followed to ensure proper closure of the chromium-containing waste. Therefore, transfer of CAU 577 from Appendix III of the FFACO to Appendix IV, Closed Corrective Action Units, is requested, as all closure activities have been completed.
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Nuclear Material Management Database Development at the Nevada National Security Site
The Nuclear Material Management & Storage (NMM&S) program at the Nevada National Security Site (NNSS) have the responsibility for the life-cycle management of accountable nuclear materials (NM) in accordance with Department of Energy (DOE) Order 410.2, “Management of Nuclear Materials” and for ensuring workers are protected from exposure to nuclear materials stored outside of an authorized shipping package under DOE Manual 441.1-1. To ensure these responsibilities were addressed in an effective, concise, and secure manner the development of NMM&S databases was undertaken. Using the existing database server structure and data acquired from the official Material Control & Accountability (MC&A) accounting program development of two separate applications, one for O 410.2 inventory requirements (NMIA) and one for M441, were begun in September 2015. The databases were deployed onto the production server in January 2017. The inventory database allows for entering NM characterization data, production of O 410.2 reports, and the ability to retain history files. The M441 database screens NM to determine if it is subject to the manual's requirements, ability to pull in unclassified RadCon materials, and provide reports as required by NMM&S, NM owners, and headquarters programs. These two databases provide NMM&S the ability to better ensure safe and secure packaging, storage, stabilization, consolidation, or disposition of NM and achieve life-cycle management goals.
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Evaluation of Rapid for a UNF cask benchmark problem
To ensure criticality safety and maintain a high level of security of Used Nuclear Fuel (UNF), it is necessary to perform detailed Particle Transport simulations. The commonly used technique is the 3-D, continuous energy Monte Carlo method that is time-consuming and, for eigenvalue calculations, may have difficulties with the source convergence. Wenner and Haghighat and Dufek demonstrated that in criticality problems with high dominance ratio and/or under-sampling possibilities, the fission matrix (FM) methodology should be used in order to be able to obtain unbiased solution. The RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system is based on the Multi-stage Response-function Transport (MRT) methodology, and uses the FM formulation to determine eigenvalue, subcritical multiplication and fission density distribution of nuclear systems. This paper uses a benchmark problem similar to the GBC-32 Cask to evaluate the accuracy and performance of RAPID against the results of a reference MCNP solution. We performed a detailed sensitivity analysis on a single-assembly model in order to determine the MCNP eigenvalue parameters necessary to obtain a reliable reference solution. These results provided a guide for the selection of an appropriate set of parameters for the eigenvalue calculation for the whole cask. The paper is organized as follows: initially, the RAPID code system is briefly discussed; then, the benchmark problem is introduced. The methodology used to analyze results is proposed and some of the results of different cases are analyzed to identify the best set of eigenvalue parameters for a reference MCNP calculation. The selected case is then compared with the RAPID calculation for both a single assembly and a full cask model. In the end some conclusions and future work perspectives are drawn.
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Compliance Monitoring Implementation Plan for 40 CFR §191.14(b), Assurance Requirement (Rev. 7)
This Compliance Monitoring Implementation Plan outlines monitoring activities conducted by the U.S. Department of Energy (DOE) at the Waste Isolation Pilot Plant (WIPP) to demonstrate compliance with the U.S. Environmental Protection Agency (EPA) disposal regulations at Title 40 Code of Federal Regulations (CFR) Part 191, "Environmental Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes," Subparts B and C (EPA 1993); and the EPA criteria for certifying compliance at 40 CFR Part 194, "Criteria for the Certification and Recertification of the Waste Isolation Pilot Plant's Compliance with the Disposal Regulations," Certification Decision, Final Rule (EPA 1998a).
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International Collaboration Activities in Geologic Disposal R&D: Spent Fuel and High-Level Waste Disposition (FY25 Progress Report)
This report describes the FY25 status of international collaboration on geologic disposal research and development (R&D) in the Office of Spent Fuel and High-Level Waste Disposition (SFHLWD) in the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE). The mission of this office is to provide confidence in the safe long-term management of the nation’s spent nuclear fuel and high-level radioactive waste by reducing uncertainty and advancing technology for extended storage, transportation, and geologic disposal. R&D in geologic disposal is administered in SFHLWD’s Disposal R&D Campaign.
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The DISTINCTIVE University Consortium: Structural Integrity - 18272
The Engineering and Physical Sciences Research Council (EPSRC) sponsored DISTINCTIVE consortium (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) is developing technologies for civil infrastructure repair, in-situ subsurface waste immobilisation, and groundwater protection during construction and decommissioning. The consortium has contributed to the development of skilled cross-disciplinary civil engineers and scientists, that have the knowledge and experience required to develop engineering solutions tailored for application within radiologically contaminated sites. The Structural Integrity Theme focuses on challenges ranging from site-scale infrastructure preservation and restoration, through injectable ground barriers for risk mitigation, to the remote characterisation and handling of individual waste packages. The main aim of the theme is to develop novel engineering solutions, tailored for use on radiologically contaminated sites, for: ground protection; infrastructure characterisation; concrete restoration and waste characterisation. Technologies should minimise current, and future, radiation exposure of the workforce whilst providing economically viable engineering solutions. (authors)
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{sup 137}Cs sorption into bentonite from Cidadap-Tasikmalaya as buffer material for disposal demonstration plant facility at Serpong
According to co-location principle, near surface disposal type the disposal demonstration plant facility will be build at Serpong nuclear area. The facility also for anticipation of future needs to provide national facility for the servicing of radwaste management of non-nuclear power plant activity in Serpong Nuclear Area. It is needs to study the material of buffer and backfill for the safety of demonstration plant facility. A local bentonite rock from Cidadap-Tasikmalaya was used as the buffer materials. Objective of experiment is to find out the specific data of sorption characteristic of Cidadap bentonite as buffer material in a radwaste disposal system. Experiments were performed in batch method, where bentonite samples were contacted with CsCl solution labeled with Cs-137 in 100 ml/g liquid:solid ratio. Initial Cs concentration was 10{sup −8} M and to study the effects of ionic strength and Cs concentration in solution, 0.1 and 1.0 M NaCl also CsCl concentration ranging 10{sup −8} - 10{sup −4} M were added in solution. As the indicator of Cs saturated in bentonite samples, Kd value was applied. Affected parameters in the experiment were contact time, effects of ionic strength and concentration of CsCl. Results showed that sorption of Cs by bentonite reached constantly after 16 days contacted, and Kd value was 10.600 ml/g. Effect of CsCl concentration on Kd value may decreased in increased in CsCl concentration. Effect of ionic strength increased according to increased in concentration of background and would effect to Kd value due to competition of Na ions and Cs in solution interacts with bentonite. By obtaining the bentonite character data as buffer material, the results could be used as the basis for making of design and the basic of performance assessment the near surface disposal facility in terms of isolation capacity of radwaste later.
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Analysis of Tc and U Spent Ion Exchange Resins from Hanford 200 West P and T Facility - 19202
The 200 West Area Pump-and-Treat (P and T) facility at the Hanford Site is being used to remove radionuclides such as uranium (U) and technetium-99 (Tc-99) from the contaminated groundwater. To support evaluation of U and Tc ion exchange (IX) treatment operational conditions and performance with various influent conditions, knowledge of resin capacity and contaminant retention is needed. To meet these objectives, samples of the spent Dowex{sup R} 21K and Purolite{sup R} A530E IX resins from the 200 West P and T facility used for the treatment of U and Tc, respectively, have been obtained and analyzed for the target analytes and other retained contaminants. The objectives of this work included 1) determination of the effective U and Tc loading on the resins, 2) evaluation of redox and chemical speciation of U and Tc retained by the resins and impact of the feed composition on the retention of U by the resin, and 3) quantification of other contaminants and groundwater constituents retained on the resins. Speciation of U and Tc-99 on the resin was determined by a combination of X-ray absorption fine structure, nuclear magnetic resonance, X-ray photoelectron spectroscopy (XAS), and optical spectroscopic techniques. Nondestructive analytical techniques including gamma energy analysis and low-energy photon spectroscopy were incorporated to quantify radionuclides retained on the resin. Scanning electron microscopy and energy-dispersive X-ray spectroscopy were used to observe the IX beads and quantify chemical constituents on the surface and penetration through the beads. Destructive analysis of the resin was explored through wet oxidation to identify and quantify co-contaminants sorbed to the resins. Inductively coupled plasma-optical emission spectroscopy (ICP-OES) and inductively coupled plasma-mass spectrometry (ICP-MS) were used to analyze for metals, iodine, and specific isotopes retained on the resin, including Tc-99 and U-238. A single pertechnetate Tc-99 species was identified to be retained on the resin. For U, the obtained structural information was correlated with the laboratory findings using pristine resin as reference material to identify the nature of the U complexes and predict U retention by the resin as a function of variable influent conditions. Both resins demonstrated high loading with sulfate, silica, calcium, iron, zinc, and other groundwater constituents, suggesting that the competing anions, fouling, or a combination thereof significantly reduced their effective capacity for U and Tc-99. A significant amount of nonradioactive iodine-127 (I-127) and long-lived radioactive iodine-129 (I-129) isotope was also retained on the resins. Obtained quantitative information provided insights on the impact of water chemistry and the presence of common mineral constituents on the retention of contaminants of concern, including U, Tc-99, and I-129. (authors)
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Life-Cycle Baseline Customization for the Formerly Utilized Sites Remedial Action Program - 20304
One of the responsibilities of the federal government is to estimate sound and defensible life-cycle baseline costs for use in federal budget estimates and to meet federal financial reporting requirements. To accomplish this, the US Department of Energy (DOE) Office of Legacy Management (LM) and the US Army Corps of Engineers (USACE) have partnered together to ensure that the liabilities documented in each Formerly Utilized Sites Remedial Action Program (FUSRAP) site's life-cycle baseline are specifically tailored. FUSRAP was created in the mid-1970's to clean up radiological contamination resulting from the early development of nuclear weapons. DOE was responsible for FUSRAP until October 1997, when Congress transferred the administration and execution of FUSRAP site cleanups to USACE. By 1997, DOE had completed the cleanup of 25 of the 46 sites that were active within the program and had begun cleanup at 13 additional sites. USACE was assigned responsibility for the cleanup of the 21 remaining FUSRAP sites, and at 8 additional sites that had since been referred for cleanup. The LM mission for the FUSRAP sites is to perform long-term surveillance and maintenance (LTS and M). Currently, LM provides long-term stewardship for 34 completed FUSRAP sites. Another 20 sites are under active remediation by USACE. Within the last 5 years, USACE has completed the cleanup at five FUSRAP sites and the LTS and M responsibility has been transferred to LM. By 2029, USACE will compete remediation at eight additional sites and LTS and M for those sites will transfer to LM. Because responsibilities for the FUSRAP sites transfer between USACE and LM upon completion of remedial actions, both agencies maintain life-cycle baselines for different stages of the project and both must have a strong understanding of the needs and requirements for each site. This understanding ensures that the life-cycle baselines form a complete and accurate picture of what is required for the site and for the FUSRAP program. LM focuses on several things when customizing the life-cycle baseline estimates for the FUSRAP sites, including (1) Understanding the unique requirements for each site. By reviewing site-specific documents prepared by USACE, such as Feasibility Studies and Records of Decision, and partnering with USACE to gain additional insight about site conditions and requirements for stewardship, as well as potential risks, LM can better develop the life-cycle baselines. (2) Developing site-specific labor breakdowns. This ensures the required labor mix is baselined for specific activities by comparing the labor mix required to perform activities at (a) other LM-managed FUSRAP sites and (b) non-FUSRAP LM sites and (c) by USACE at active FUSRAP sites. (3) Taking a tiered approach to life-cycle baseline planning. Estimates for sites transferring to LM in the near term (5 years) are more definitive than for sites transferring in the out-year period. Remedial actions at the near-term sites are at or near completion, providing LM a strong understanding of the LTS and M requirements and remaining liabilities. This in turn allows for site-specific customization of the baseline. (4) Using a robust risk management approach to ensure that liabilities specific to each site are identified, evaluated by probability and severity, and documented in relation to the impact to cost or schedule. To ensure the most accuracy within all the baselines, the FUSRAP life-cycle baselines are updated as needed to support program, project, and contract management needs. (authors)
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Best Practice on Facility Characterisation from a Material and Waste End- State Perspective - 16041
Radiological characterisation plays a significant role in the process of decommissioning of shut-down nuclear facilities in order to ensure the protection of the environment and radiation safety. At all stages of a decommissioning programme or project, adequate radiological characterisation is of crucial importance, not least from a material and waste perspective. The radiological characterisation is a key element for planning, controlling and optimising decommissioning and dismantling activities. Experience has shown that data and information from the operation of a facility can - supplemented by recently collected and analysed data and information - be of crucial importance for decisions on waste management and for characterisation of radioactive waste. Once the dismantling has been done, some information may be hard, costly or even impossible to obtain later in the waste management process. This was the reason why the Working Party on Decommissioning and Dismantling (WPDD) of the OECD Nuclear Energy Agency (NEA) decided in late 2013 to extend the mandate of the Task Group on Radiological Characterisation and Decommissioning (TGRCD) for a second phase focusing on nuclear facility characterisation from a waste and material end-state perspective whereas the first phase focused on overall strategies of radiological characterisation. This paper gives an overview of the activities and findings within both phases up to now. (authors)
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Decommissioning concepts for the norwegian nuclear research facilities Halden and Kjeller - 15022
The Norwegian nuclear research facilities Halden and Kjeller are near the end of their current operational licenses. Thus, the Norwegian Ministry of Trade, Industry and Fisheries have decided that a concept evaluation study shall be made on the decommissioning alternatives for these facilities. This task was given to Ndcon (Westinghouse and Studsvik) with their broad experience in decommissioning planning and decades of nuclear expertise, together with DNV GL, that is certified in Norway for concept evaluation studies. The decommissioning alternatives that have been evaluated are strategy, end-state and waste management options. The decommissioning strategies are the IAEA defined immediate dismantling, deferred dismantling and entombment. The different end-states that have been evaluated are unrestricted use (green field), light industry (brown field) and other nuclear activity, in which the site will continue to have a nuclear license even after the decommissioning. The waste management options are direct disposal of materials, recycling of materials off-site and recycling on-site. The result of the cost estimation is that it is possible to identify viable alternatives for a chosen decommissioning strategy, end state and waste management options. The expected duration of the decommissioning program is 10 years for both sites, the dismantling and demolition period is expected to be able to perform within 4 years. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗