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At least 235 records · Page 13
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DOE OSTI2015

Effect of low-level radioactive waste leachate on antioxidant depletion in HDPE geo-membranes - 15135

Low-level radioactive waste (LLW) and mixed waste (MW) disposal facilities in the US are required to have a service life in excess of 1000 yr. Understanding the rate of degradation of high-density polyethylene (HDPE) geo-membranes (GMs) used in barriers of LLW and MW facilities is necessary to assess their service life. HDPE GMs undergo degradation in three stages: antioxidant depletion (Stage I), induction time (stage II), and polymer degradation (Stage III). For LLW and MW facilities, these degradation mechanisms must be assessed with the GM in contact with leachates representative of LLW or MW. This paper focuses on antioxidant depletion in LLW leachate. HDPE GM coupons (2-mm thick) were immersed in radioactive synthetic leachate (RSL) with chemistry representative of leachate in LLW disposal facilities operated by the US Department of Energy's environmental restoration programs. Depletion was assessed at four temperatures (25, 50, 70, and 90 deg.C). Comparative tests were conducted with non-radioactive synthetic leachate (NSL), which has the same chemistry as RSL except radionuclides are excluded. Control tests were conducted with deionized (DI) water. Specimens were removed periodically and tested to determine their mechanical and chemical properties. Antioxidant depletion was measured by both standard and high-pressure oxidative induction time (OIT) tests. The rate of antioxidant depletion in RSL is statistically no different than depletion in NSL. Extrapolating the laboratory data at elevated temperatures to a typical LLW liner at 15 deg.C via Arrhenius modeling showed that the time for antioxidant depletion for a GM in a composite liner is approximately 650 yr when exposed to LLW leachate. Actual service life of the GM will be longer. (authors)

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DOE OSTI2016

A Study on Vitrification for LLW Radioactive Wastes - 16235

Ash and ion exchange resin wastes to be vitrified were considered among various radioactive wastes in the Japanese nuclear industry. The appropriate glass composition for the waste is calculated using the computer program. The glass composition is systematically studied using a ternary diagram, considering viscosity, electrical conductivity, and durability. The optimized glass composition based on calculated parameters of vitrification process and the convenience of mixing wastes is selected for crucible experiment. Since differences between calculated values and measured values may occur upon evaluation of glass formulations using glass formulation computer simulation program, the calculated glass formulations should be verified through experiments on a laboratory scale for more meticulous development of glass formulations. For the follow-up development of glass formulations, several glass formulations are planned to be developed. The candidate glasses for treatment of ash and ion exchange resin were melted and fabricated using reagents, respectively. The chemicals were weighed out to produce the desired batch compositions and mixed thoroughly and then put into clay crucibles for melting. Next, the crucibles were introduced into an electric furnace at 1,150 deg. C in the ambient atmosphere and pressure with occasional stirring by a quartz rod. A hold time of 1 hour was introduced to promote homogenization. Finally, the melted material was poured on a graphite plate to be rapidly cooled. The viscosities of all fabricated glasses were in the desired range. After rapid cooling, the micro-structure of the glasses was analyzed using the SEM. They exhibited amorphous structure with good homogeneity. (authors)

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DOE OSTI2021

Assessing the Potential for Inadvertent Human Intrusion at the Area 3 and Area 5 Radioactive Waste Management Sites on the Nevada National Security Site, Nye County, Nevada

This paper recommends an approach to inadvertent human intrusion (IHI) at the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) on the Nevada National Security Site (NNSS). IHI analysis uses the consequences of an individual inadvertently contacting buried waste to set waste concentration limits for near-surface disposal of low-level radioactive waste (LLW). Regulatory agencies are increasingly applying risk-informed decision-making to LLW waste management (NRC 2006). Risk-informed decision-making combines scientific risk assessment with stakeholder values and perceptions to determine a level of acceptable risk. Risk considers not only the consequences of an event, but also its probability of occurring.

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DOE OSTI2017

A Study on Korea Experience in Regulating the Transport of Low and Intermediate-level Radioactive Wastes - 17182

In the Republic of Korea, the transport of Low and Intermediate-Level Radioactive Wastes (LILW) from nuclear power plants (NPP), nuclear research reactor, and radioactive isotope storage facility to Gyeongju LILW disposal facility has increased since the operation license of facility was approved in December 2014. As a result, there has been an increase in public anxiety and concerns over damage or social disorder with LILW transport. So the necessity of strengthened safety confirmation on LILW has been constantly raised. In August 2016, the Enforcement Regulations of the Nuclear Safety Act were amended in relation to the reporting and inspection of LILW transport. The first section of this paper describes the domestic transport regulations including the recently revised contents. The second section shows that the status of domestic and inspection of LILW transport were introduced and transportation regulatory experience was shared in Korea. (authors)

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DOE OSTI2020

A Dose Assessment Model for Radioactive Waste Exposed by Gully Erosion at West Valley - 20513

The Western New York Nuclear Service Center (WNYNSC), located approximately 48 km south of Buffalo, New York, is the site of a former nuclear fuel reprocessing and radioactive waste disposal facility. Spent nuclear fuel was processed there from 1966 to 1972, leaving behind radioactive and chemical wastes in two disposal areas and a waste tank farm. Site operations also resulted in releases of radioactivity to site soils, groundwater, and to surface waters draining the site. The New York State Energy Research and Development Authority (NYSERDA) and the U.S. Department of Energy (DOE) are collaborating in a process of decision making for decommissioning those facilities remaining at the WNYNSC following the completion of Phase 1 decommissioning. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) computer model to assist the agencies in this process. The PPA Model includes a contaminant transport component focusing on the movement of contaminants within and among environmental media including groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The model also includes evaluation of potential exposure and health effects for a Resident Farmer exposure scenario, where the Resident Farmer represents a critical group, described as that group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances. The West Valley Site is located in the glaciated Allegheny Plateau region of western New York State. The waste reprocessing and disposal areas were constructed on a relatively fat area of plateau dissected by drainages of Buttermilk Creek, including Erdman Brook, Franks Creek, and Quarry Creek. An important aspect of the contaminant transport component of the PPA model is consideration of erosive processes such as slumping of the stream slopes and the advance of gullies from these streams. These erosion processes remove material from the plateau, growing the size of the creek valleys and making them wider and deeper. Of particular interest for the impact of erosion is radioactive waste contained in the Nuclear Regulatory Commission (NRC)-Licensed Disposal Area (NDA), the New York State-Licensed Disposal Area (SDA), and residual radiological inventory in the underground storage tanks at the Waste Tank Farm (WTF). The PPA Model is organized around geographically-defined facilities which were constructed upon the plateau, including the NDA, SDA, and WTF. Ongoing stream erosion processes will potentially transfer radioactive waste and residual inventory from these facilities to the ground surface on adjacent hillslope areas where erosion has breached the facility. Hence, it is important to evaluate the consequences of potential exposures to a Resident Farmer on the hillslopes below a breached facility. Two interrelated aspects of the dose assessment model related to hillslope exposure are discussed: 1) representation of the physical processes related to transport of radionuclides from facilities onto the hillslopes, and from the hillslopes into adjoining creeks where contaminated material migrates downstream with surface water and sediment, and; 2) adaptation of the activities associated with the Resident Farmer scenario to assess potential exposures to contamination in the hillslope areas. This discussion will cover the conceptual basis of the hillslope exposure and transport models, and also implementation in the PPA computer model. (authors)

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DOE OSTI2019

Unreviewed Disposal Question Evaluation: Disposal of the Perma-Fix Environmental Services Macroencapsulated Sealed Sources at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation assesses whether the Perma-Fix Environmental Services Macroencapsulated Sealed Sources is suitable for shallow land burial at the Area 5 Radioactive Waste Management Site on the Nevada National Security Site. Disposal of the waste stream meets all performance objectives in DOE M 435.1-1, “Radioactive Waste Management Manual,” Chapter IV, Section P. The Perma-Fix Environmental Services Macroencapsulated Sealed Sources waste stream is recommended for acceptance without conditions.

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DOE OSTI2018

Environmental Effects of Cavitation Technology for Radioactive Waste Management

Results from studies of the removal from storage and solidification of low- and intermediate-level liquid radioactive waste in inorganic binders (cementation) are presented. Cavitation technology for recycling radiochemical wastes from nuclear power plant decommissioning increases the extraction throughput and strength of the cement compound.

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DOE OSTI2019

Nevada National Security Site 2018 Data Report: Groundwater Monitoring Program, Area 5 Radioactive Waste Management Site

This report is a compilation of the groundwater and waste cell leachate sampling results from the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada National Security Site, Nye County, Nevada. Groundwater samples from the aquifer immediately below the Area 5 RWMS have been collected and analyzed and static water levels have been measured in this aquifer since 1993. These groundwater data are evaluated for evidence of effects on the aquifer related to the Area 5 RWMS. Leachate from one lined mixed waste cell at the Area 5 RWMS (Cell 18) has been sampled since the cell opened in 2011. These leachate data are analyzed for hazardous contaminants to determine appropriate leachate handling and disposal. A second lined mixed waste cell at the Area 5 RWMS (Cell 25) began receiving waste in August 2018, but no leachate samples were collected or analyzed during 2018. This report includes data from 2014 through 2018 but focuses on the 2018 data. During 2018, groundwater samples were collected and static water levels were measured at three wells surrounding the Area 5 RWMS. Groundwater samples were collected at wells UE5PW-1, UE5PW-2, and UE5PW-3 on March 6 and August 14, 2018. Static water levels were measured at each of these wells on March 5, June 4, August 13, and October 25, 2018. Groundwater monitoring at the Area 5 RWMS complies with Title 40 Code of Federal Regulations (CFR) Part 264.97, General ground-water monitoring requirements , and 40 CFR 264.98, Detection monitoring program. Groundwater samples were analyzed for indicators of contamination (pH, specific conductance [SC], total organic carbon, total organic halides, and tritium) and, beginning in 2017, toxicity characteristic metals (arsenic, barium, cadmium, chromium, lead, mercury, selenium, and silver). General water chemistry parameters (calcium, iron, magnesium, manganese, potassium, sodium, sulfate, chloride, fluoride, bicarbonate, and silicate) were also measured. Results from samples collected in 2018 are within the limits for each parameter established by the Nevada Division of Environmental Protection in Resource Conservation and Recovery Act (RCRA) Permit NEV HW0101, Revision 6. Data from the shallow aquifer indicate there has been no measurable impact to the uppermost aquifer from the Area 5 RWMS, and there were no significant changes in measured groundwater parameters compared to previous years. Leachate from lined mixed waste Cell 18 drains into a sump and is pumped into an above ground tank. Samples are collected from the tank when the leachate volume approaches the 3,000-gallon tank capacity. During 2018, leachate samples were collected on March 1, 2018 and August 1, 2018. All leachate analysis results are below the regulatory levels for toxicity characteristic contaminants (40 CFR 261.24) and below the investigation levels for polychlorinated biphenyls (PCBs), tritium, pH, and SC specified in RCRA Permit NEV HW0101, Revision 6. No quantifiable PCB levels were detected in any leachate sample. Based on the leachate analysis results, RCRA Permit NEV HW0101 allowed all leachate collected from Cell 18 to be pumped from the collection tank and used for dust control at Cell 18.

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DOE OSTI2016

Determination of Erosion/Corrosion Rates in Hanford Tank Farms Radioactive Waste Transfer System Pipelines - 16395

The twenty-eight double-shell underground radioactive waste storage tanks at the U.S. Department of Energy's Hanford Site near Richland, WA are interconnected by the Waste Transfer System network of buried steel encased pipelines and pipe jumpers in below-grade pits. The pipeline material is stainless steel or carbon steel in 51 mm to 152 mm (2 in. to 6 in.) sizes. The pipelines carry slurries ranging up to 20 volume percent solids and supernatants with less than one volume percent solids at velocities necessary to prevent settling. The pipelines, installed between 1976 and 2011, were originally intended to last until the 2028 completion of the double-shell tank storage mission. The mission has been subsequently extended. In 2010 the Tank Operating Contractor began a systematic evaluation of the Waste Transfer System pipeline conditions applying guidelines from API 579-1/ASME FFS-1 (2007), Fitness-For- Service. Between 2010 and 2014 Fitness-for-Service examinations of the Waste Transfer System pipeline materials, sizes, and components were completed. In parallel, waste throughput histories were prepared allowing side-by-side pipeline wall thinning rate comparisons between carbon and stainless steel, slurries and supernatants and throughput volumes. The work showed that for transfer volumes up to 6.1 E+05 m{sup 3} (161 million gallons), the highest throughput of any pipeline segment examined, there has been no detectable wall thinning in either stainless or carbon steel pipeline material regardless of waste fluid characteristics or throughput. The paper describes the field and laboratory evaluation methods used for the Fitness-for-Service examinations, the results of the examinations, and the data reduction methodologies used to support Hanford Waste Transfer System pipeline wall thinning conclusions. (authors)

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DOE OSTI2018

Development of Hot-Isostatic Pressing for Treatment of UK ILW Radioactive Wastes - 18280

Hot-Isostatic Pressing (HIPing) has been proposed as an improved method for the treatment of radioactive wastes in the UK. Sludges and slurries housed on the Sellafield site are a prime candidate for HIP treatment due to their large volumes and issues associated with the use of the baseline cementation strategy. Options have been developed by GeoRoc Ltd to co-process these wastes, including a detailed conceptual plant design. This paper will describe research and development aimed at demonstrating the viability of the concept presented in this design. This paper summarizes two studies performed to address the knowledge gaps identified for treatment of these wastes using HIP technology. The first study related to the poorly characterized nature of the wastes, which requires any treatment option to be flexible in terms of the chemical and physical properties of the waste feed. The second study involved demonstrating that HIPing can be effectively utilized at a scale applicable to treat a waste inventory estimated to be in excess of 5,000 m{sup 3}. The first section of this paper presents a summary of research carried out to determine where variation in the waste stream composition may affect the processing and wasteform produced when HIPing Magnox sludges. This study aimed to qualify where restrictions on the process envelope may exist due to poorly defined waste-stream characteristics. A baseline wasteform was selected with the aim of achieving significant waste volume reduction, while producing a physically and chemically stable wasteform. A glass-bonded magnesium-titanate/magnesium-silicate wasteform was selected as the preferred option for the Magnox sludge wastes. A series of studies investigated the densification and phase formation of waste-forms by altering key variables in the Magnox sludge simulant. To test the extremes of the processable waste loading, samples were produced and characterized at 100 wt% waste-loading for Magnox sludge, Site-Ion Exchange Plant (SIXEP) sand/clinoptilolite and a 50:50 mix of the two waste-streams. Further studies involved a) altering the waste loading to additive mass ratio (to simulate changes in the waste loading/sludge solids content), b) varying the particle size distribution from the datum to include only small or large particulates in the sludge feed; and c) introducing up to 20 wt% metallic Magnox alloy, organic species or carbonate materials (to determine how their inclusion may affect the processing and densification of the materials). A U{sub 3}O{sub 8} loaded sample was also produced using the baseline Magnox sludge formulation. All the tests performed were shown to produce monolithic, well densified waste-forms with substantial volume reductions. Minimal variations from the baseline were noted but the waste-forms produced with these deviations was determined to be of comparable quality to of the baseline product. U{sub 3}O{sub 8} was shown to behave in a similar fashion to the ZrO{sub 2} surrogate used to simulate it in the process envelope studies and was encapsulated in the magnesium titanate/magnesium silicate matrix. This work demonstrates that HIP technology can successfully produce waste-forms across the broad process envelope expected in the Magnox sludge inventory. It is also possible to produce monolithic waste-forms at 100 wt% waste loading from Magnox sludges, SIXEP sand clinoptilolite slurries or a co-processed mixture of the two. The second part of this paper summarizes a commercially relevant demonstration of HIP technology for the treatment of sludge and slurry wastes housed at the Sellafield site (UK). This study culminated in production of the largest HIPed wasteform successfully produced to date. This study was performed in a series of stages. Investigations into formulation and properties of waste-forms were performed at a lab scale to evaluate suitable wasteform systems and identify suitable combinations of additives and waste loadings. Scale up demonstrations were then performed in stages with custom-designed HIP canisters. Scale-up and characterization of key wasteform properties were successfully performed with initial canister volumes of 50 ml, 400 ml and 8 L. Suitable front-end unit operations to produce a granular, free-flowing calcined simulant were demonstrated at an engineering scale. The demonstration was successful in processing a HIP canister with an initial starting volume in excess of 100 L. A 75 % volume reduction from the raw sludge was achieved at scale while demonstrating that HIP is applicable at suitably large scale for immobilization of ILW waste. (authors)

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DOE OSTI2019

Unreviewed Disposal Question Evaluation: Disposal of the Oak Ridge National Laboratory Contaminated Debris at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Oak Ridge National Laboratory (ORNL) Contaminated Debris (ORNLDEBRIS001, Revision 11 [ORNL 2019a]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the waste stream meets all performance objectives in U.S. Department of Energy (DOE) Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 2011). The ORNL Contaminated Debris waste stream is recommended for acceptance with conditions on the total activity disposed of tritium ( 3 H), niobium-94 ( 94 Nb), technetium-99 ( 99 Tc), cesium-137 ( 137 Cs), iridium-192 ( 192 mIr), radium-226 ( 226 Ra), uranium-234 ( 234 U), and 238 U. The total activity disposed shall not exceed the following limits: 3 H: 2E12 Bq, 94 Nb: 2E9 Bq, 99 Tc: 8E10 Bq, 137 Cs: 1E11 Bq, 192 mIr: 1E9 Bq, 226 Ra: 3E11 Bq, 234 U: 2E9 Bq, 238 U: 5E9 Bq.

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DOE OSTI2019

Unreviewed Disposal Question Evaluation: Disposal of the Idaho National Laboratory Repackaged Uranium Metal and Oxide with Polychlorinated Biphenyls at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation assesses whether the Idaho National Laboratory (INL) Repackaged Uranium Metal and Oxide with polychlorinated biphenyls (PCBs) is suitable for shallow land burial at the Area 5 Radioactive Waste Management Site on the Nevada National Security Site. Disposal of the INL Repackaged Uranium Metal and Oxide with PCBs waste stream meets all the performance objectives contained in DOE M 435.1-1, “Radioactive Waste Management Manual,” Chapter IV, Section P. The INL Repackaged Uranium Metal and Oxide with PCBs waste stream is recommended for acceptance without conditions.

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DOE OSTI2018

FY 2017 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada

The U.S. Department of Energy Standard Disposal Authorization Statement and Tank Closure Documentation (DOE-STD-5002-2017) requires preparation of an Annual Summary Report comparing low-level radioactive waste disposal facilities operations with the Disposal Authorization Statement (DAS) and its supporting technical documents including the performance assessment (PA) and the composite analysis (CA). The fiscal year (FY) 2017 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) was prepared by reviewing planned and discovered changes, waste receipts, monitoring results, research and development (R&D) results, and the status of DAS conditions and issues to assess the validity of the DASs, PAs, and CAs.

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DOE OSTI2015

Repository planning for spent nuclear fuel and high-level radioactive waste: insights from U.S.. and international programs - 15638

For decades, the United States and other nations with nuclear programs have been planning for permanent disposition of spent nuclear fuel and high-level waste, to address the increasingly difficult problem of storing these materials for the long term. These plans are complicated by issues ranging from technological factors to regulatory approaches and public concerns about health and environmental protection. In its 2012 report, the Blue Ribbon Commission (BRC) on America's Nuclear Future identified a number of recommendations for creating a safe, long-term solution to the nation's problem of managing and disposing of its spent fuel and high-level radioactive waste. These include a recommendation that the U.S. Environmental Protection Agency (EPA) revise its standards for a high-level waste repository. The EPA is currently reviewing and synthesizing literature relevant to such standards, including: (1) existing national and international standards and guidelines for repositories; and (2) benefits and barriers to such standards, including limitations related to current technologies, as well as siting, licensing, and construction processes. This paper highlights findings of this literature review and synthesis. More than 90 countries, from Albania to Vietnam and Yemen, were identified as potential sources of relevant information based on existing nuclear programs or consideration of such programs to address future energy needs. Pursuant to the BRC recommendations, the U.S. Department of Energy (DOE) has compiled extensive information resources relevant to spent fuel and radioactive waste repositories in the Centralized Used Fuel Resource and Information Exchange (CURIE) database developed by Oak Ridge National Laboratory. Further valuable sources of information include annual proceedings of recent Waste Management Symposia, which provide insights into recent trends and progress across various national and international programs. Recent literature from a number of countries, including the United States and Canada, and the United Kingdom, Finland, Sweden, France, Germany, and other countries in Western Europe, as well as other regions including Australia, indicates that consent-based siting has emerged as a dominant theme. In addition to effective siting processes, key program needs identified in the literature include regulatory standards and performance context, as well as reliable technologies and engineered systems, such as for deep drilling and assured safe containment of these radioactive materials extending into the very long term. (authors)

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DOE OSTI2015

On the possibility of reprocessing spent nuclear fuel and radioactive waste by plasma methods

The concept of plasma separation of spent nuclear fuel and radioactive waste is presented. An approach that is based on using an accelerating potential to overcome the energy and angular spread of plasma ions at the separation region inlet and utilizing a potential well to separate spatially the ions of different masses is proposed. It is demonstrated that such separation may be performed at distances of about 1 m with electrical potentials of about 1 kV and a magnetic field of about 1 kG. The estimates of energy consumption and performance of the plasma separation method are presented. These estimates illustrate its potential for technological application. The results of development and construction of an experimental setup for testing the method of plasma separation are presented.

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DOE OSTI2018

Unreviewed Disposal Question Evaluation: Disposal of the Idaho National Laboratory Repackaged Uranium Metal and Oxide at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation assesses whether the Idaho National Laboratory (INL) Repackaged Uranium Metal and Oxide is suitable for shallow land burial at the Area 5 Radioactive Waste Management Site on the Nevada National Security Site. Disposal of the INL Repackaged Uranium Metal and Oxide meets all performance objectives in DOE M 435.1-1, “Radioactive Waste Management Manual,” Chapter IV, Section P. The INL Repackaged Uranium Metal and Oxide waste stream is recommended for acceptance without conditions.

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DOE OSTI2016

Nuclide Release Pathways in and Around a Trench Repository for Radioactive Waste - 16014

An effective model for a safety assessment of a conceptual trench repository system, in which low-and intermediate level radioactive waste (LILW) is disposed of has been developed. The repository system is assumed to be located in the surface area of a mountain hill near the seashore. An easy-to-use computer program based on this model, developed as a GoldSim template, has been extended based on the previous one in consideration of the rather possible complex pathways in and around the trench for its performance assessment, which is able to evaluate a nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural, and manmade events, and scenarios that can occur after a failure of waste drums with associated uncertainty. To demonstrate the nuclide release and transport behavior through these various possible pathways in the near- and far-fields of the repository system under certain scenarios, some illustrative evaluations and a simple sensitivity investigation with C-14 for the concrete barriers of the trench were made. (authors)

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DOE OSTI2014

2013 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analyses

The Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (National Security Technologies, LLC 2007a) requires an annual review to assess the adequacy of the performance assessments (PAs) and composite analyses (CAs), with the results submitted to the U.S. Department of Energy (DOE) Office of Environmental Management. The Disposal Authorization Statements for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) also require that such reviews be made and that secondary or minor unresolved issues be tracked and addressed as part of the maintenance plan (DOE 1999a, 2000). The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office performed an annual review of the Area 3 and Area 5 RWMS PAs and CAs for fiscal year (FY) 2013. This annual summary report presents data and conclusions from the FY 2013 review, and determines the adequacy of the PAs and CAs. Operational factors (e.g., waste forms and containers, facility design, and waste receipts), closure plans, monitoring results, and research and development (R&D) activities were reviewed to determine the adequacy of the PAs. Likewise, the environmental restoration activities at the Nevada National Security Site (NNSS) relevant to the sources of residual radioactive material that are considered in the CAs, the land-use planning, and the results of the environmental monitoring and R&D activities were reviewed to determine the adequacy of the CAs. Important developments in FY 2013 include the following: • Development of a new Area 5 RWMS closure inventory estimate based on disposals through. FY 2013 • Evaluation of new or revised waste streams by special analysis. • Development of version 4.115 of the Area 5 RWMS GoldSim PA/CA model.

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