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DOE OSTI2017

Stakeholder Involvement in the Revegetation of the 92-Acre Area - 17152

The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) works closely with stakeholders that have an interest in the revegetation of a closed landfill on the Nevada National Security Site (NNSS). Located at the Radioactive Waste Management Complex on the NNSS, the 92-Acre Area was operational from 1961 to 2010. In 2009, the Nevada Division of Environmental Protection (NDEP) and NNSA/NFO agreed on a closure path consisting of a vegetative cover. Closure activities were initially completed in 2012. Although revegetation was initially promising, subsequent monitoring revealed that a viable plant community had not been established. Subsequently, sections of the site were chosen as test plot locations to determine the best path forward. Despite various seeding methods, mulching techniques, and irrigation frequencies, the test plots failed to produce viable plant communities in 2013 and 2014. NNSA/NFO had been in communication with one stakeholder, NDEP, during the revegetation process. NDEP agreed that after three failed revegetation attempts, a change in strategy was necessary. The Consolidated Group of Tribes and Organizations (CGTO), a coalition of tribal members representing seventeen culturally affiliated tribes and one organization, agreed to provide guidance using traditional ecological knowledge for re-vegetating the 92-Acre Area. A Tribal Revegetation Committee (TRC) was formed and tasked with developing guidance from a unique tribal perspective. After visiting the 92-Acre Area and receiving background information, the TRC held several meetings and deliberated on how to address the revegetation challenges. The TRC submitted recommendations to NNSA/NFO identifying approaches that incorporate tribal perspectives in re-vegetating the site. The Nevada Site-Specific Advisory Board (NSSAB), which is made up of community members who represent Nevada stakeholders, was also asked for input regarding the 92-Acre Area. After a public meeting in May 2016, the NSSAB submitted recommendations, which included a suggestion to consider input from the CGTO. (authors)

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DOE OSTI2017

EPRI Report: Review of Geostatistical Approaches to Characterization of Subsurface Contamination - 17442

The key metric in the final release of a nuclear facility from radiological controls following plant decommissioning is the potential long-term radiation exposure to future users of the site. This is estimated from modeling of the exposure that could conceivably result from residual contamination remaining on the site in buildings, soil and groundwater after the completion of decontamination, dismantlement and remediation activities. A key input to such modelling is characterization data obtained from surveys and sampling. To date, the regulatory guidance for conducting the final status survey required to show compliance with regulator approved site release limits (i.e., radionuclide concentrations or contamination levels) has been limited to surface contamination. EPRI has completed a project that provides guidance for use of geo-statistics in the development of a site final status survey that will address the impact of dose to a future user of the site from subsurface contamination. (authors)

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DOE OSTI2017

EPRI Research Results: The EPRI Implementation Guide for the NRC BTP on Concentration Averaging and Encapsulation - 17488

The Nuclear Regulatory Commission published Revision 1 of the Branch Technical Position (BTP) on 'Concentration Averaging and Encapsulation' (BTP 2015.) PRI developed an Implementation Guide for applying BTP 2015 to nuclear power plant specific waste streams. This Implementation Guide has been vetted and supported by an EPRI organized working group consisting of industry participants (including utilities subject matter experts, disposal sites, NRC and Agreement State representatives and industry experts) will ensure compliant and consistent application during waste generation, packaging, processing and disposal activities. The Implementation Guide provides for a comparison to previous BTP guidance and provides clarification and common understanding of the key concepts contained within the revised BTP. These include averaging constraints, hot spots, waste streams, waste types, discrete items, blending, solidification, encapsulation and alternative approaches. EPRI published the Implementation Guidance for the Nuclear Regulatory Commission Branch Technical Position on 'Concentration Averaging and Encapsulation, Revision 1' (EPRI Report 3002008189) in the September 2016 and it was made available to the public. (authors)

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DOE OSTI2015

Assessment of defense-related uranium mines in the United States and other US DOE office of legacy management domestic and international uranium initiatives - 15335

US DOE's Office of Legacy Management (LM) has taken several initiatives to improve domestic uranium mining and milling operations in the United States and to contribute to the management of legacy uranium mining and milling sites around the world. Initiatives include completing the 2014 Report to Congress on Defense-Related Uranium Mines, managing the Title X Uranium and Thorium Program with the Office of Environmental Management, and assisting IAEA with uranium issues in member state countries. The 2014 Report to Congress identified abandoned uranium mines (AUMs) in the United States that provided uranium ore to the US Atomic Energy Commission (AEC) between 1947 and 1970. The report identified AUM locations, reclamation/remediation status, radiological risks and other hazards, costs for reclamation and remediation, and potential prioritization schemes for mine cleanup. To develop reclamation and remediation cleanup cost estimates and assess risks, AUMs were grouped into six categories by tons of uranium ore produced, ranging from Small (less than 91 metric tons [100 tons]) to Very Large (greater than 453,900 metric tons [500,000 tons]). Radiological risk was calculated using the RESRAD computer code for five exposure scenarios: offsite resident, onsite resident, occasional visitor, recreational visitor, and mine reclamation worker. Results indicate that 69 percent of the AUMs are in Colorado and Utah and 23 percent are in Arizona, Wyoming, and New Mexico; and that the majority of the AUMs (68 percent) were considered Small or Small/Medium production mines, producing less than 908 metric tons (1,000 tons) of uranium ore. Although New Mexico has fewer AUMs (247) than the other states in the Colorado Plateau region, AUMs in that state (primarily those in the Grants Mineral Belt) produced 45 percent of the 79.5 million tons of AEC-purchased uranium ore. About 50 percent of the AUMs are on public land managed by the US Bureau of Land Management; 435 AUMs are on the Navajo Nation. Risk estimates for the onsite resident scenario (plausible on tribal and non-federal land) could result in an incremental cancer risk greater than 10{sup -4}. Radon inhalation was the dominant contributor to radiological risk for the five exposure scenarios evaluated. Different state and federal agencies are conducting cleanup of some AUMs under various remedial and legal authorities. Activities constituting mine 'reclamation' may reduce radiological risks to humans to acceptable levels for many mines on federal public lands if the occasional visitor and recreational visitor exposure scenarios are assumed. Addressing physical hazards (e.g., open shafts) at mines where conditions could cause serious injuries is typically a priority of public land management agencies. At locations where mine-related groundwater contamination occurs, it is a significant contributor to cleanup costs. However, most Small and Small/Medium AUMs were likely developed above the water table, and some wet mines are located in areas where groundwater has high, naturally occurring levels of the same constituents (including radioactive elements) that are typically associated with uranium mines. LM is also administering, with the US DOE Office of Environmental Management, the Title X program, which was developed by Congress to reimburse private companies (i.e., those that held active operating licenses prior to 1978) for a portion of their reclamation costs for mill sites that processed ore sold to AEC. LM recently updated the program guidance document and established internal protocols for annually auditing companies. LM is assisting IAEA on two different initiatives: assisting developing countries with training on the basic steps to take to perform reclamation of uranium mines, and developing technical guidance and case studies related to safety and environmental assessments and post-closure management of uranium legacy sites as part of the IAEA Regulatory Supervision of Legacy Sites initiative. (authors)

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DOE OSTI2018

Development of a Technical Standard for Radiological Characterization of DAW - 18062

Throughout its 70+-year history, the U.S. Department of Energy (DOE) and its predecessor agencies have performed a wide mission involving nuclear materials and weapons development, production, testing, and modernization. While the initial mission components relied heavily upon reactor physics and chemistry, the environmental programs have increasingly relied upon development of more accurate and effective measurement methodologies to determine the radiological and chemical properties of the unique waste streams that were generated during the weapons production and testing mission. Over time, the DOE laboratories and former production sites have developed a wide range of site-specific processes and procedures for conducting radiological characterization of dry active waste (DAW) that is managed as low-level radioactive waste. This has resulted in a Complex-wide variability among sites with respect to the methodologies and instrumentation used to characterize individual waste streams. The DAW-type waste was selected for development of a characterization standard since it is a simple, voluminous, and repetitive waste stream with a substantial knowledge base, established experience history at multiple site locations, and the potential for realization of programmatic benefits though standardization. The Energy Facility Contractors Group (EFCOG) Waste Management Working Group (WMWG) has proposed the development of a new DOE Technical Standard for radiological characterization of DAW. The purpose of the proposed new standard is to provide a technical resource that integrates and enhances the utilization of current site-specific methodologies and processes. A Technical Basis Document (TBD) has been prepared to support the development and utilization of the new standard by providing detailed information for use by site personnel when implementing the guidance provided in the standard. It also provides the rationale for and details concerning the recommended technical approaches for radiological characterization of representative DAW-type waste streams. The TBD will be included as an appendix to the proposed standard. Adoption of a standardized approach to DAW characterization will take advantage of multiple applications for a technology that is both mature and stable. The DAW characterization process is repetitive and lends itself to implementation at a wide range of operating locations - resulting in economies of scale, compliance with regulatory requirements, and consistent implementation of procedures which support design, operations, maintenance, and management of facilities, activities and programs. The purpose of the proposed new Technical Standard for DAW Radiological Characterization is to integrate and improve the utilization of current site-specific methodologies and processes and will include a comprehensive evaluation and comparison of current radiological characterization methodologies to determine which ones are most reliable and effective in producing relevant and valid data. In view of the long history and participation by several agencies in the regulation of nuclear programs, existing Voluntary Consensus Standards have also been reviewed for relevancy and applicability to the new proposed standard. The proposed new Standard will be developed in concert with existing scientific and regulatory guidance to produce a document that incorporates the state of current knowledge and site practices with a graded approach to applying those methodologies that are applicable and effective for characterizing a wide range of DAW-type wastes - including, but not limited to, debris, low specific activity (LSA) material, and surface contaminated objects (SCO). Under the leadership of the National Laboratory Subgroup of the EFCOG/WMWG), an EFCOG team will develop the proposed technical standard. This process will incorporate the information gathered during preparation of the TBD, which will serve as a technical resource to support the technical approach for selection of DAW characterization methodologies. Preparation of the TBD included a detailed survey of historical site practices and examination of the characterization approaches and measurement methodologies currently being applied to DAW-type wastes. Future activities include the development and submittal of a Project Justification Statement that will be submitted to the DOE Technical Standards Program for review and comment. The EFCOG/WMWG Team will respond to reviewer comments, as required. Following final concurrence by affected organizations through the DOE Technical Standards Coordination process, the EFCOG/WMWG Team will prepare the draft technical standard and respond to comments generated during RevCom review and final format review by the DOE Technical Standards Management Program. (authors)

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DOE OSTI2018

Knowledge Management in Nuclear Liabilities Regulation in the UK - 18155

Within the UK's nuclear safety regulator (Office for Nuclear Regulation; ONR) nuclear liabilities regulation addresses the safe management of radioactive waste, decommissioning and land quality through a facility's lifetime. Higher activity radioactive waste (HAW) could be stored on site for up to 150 years prior to disposal, Magnox decommissioning sites are planned to be in Care and Maintenance for up to 70 years and Sellafield decommissioning is not planned to be complete until after 2100. The long timescales over which decommissioning occurs and radioactive waste is managed is a challenge to maintaining a suitable level of knowledge to effectively and efficiently regulate nuclear liabilities activities. There is a danger that implicit knowledge accumulated over decades could be lost as people retire. Effective knowledge capture and transfer are key elements in maintaining a workforce with the capacity and capability to deliver high levels of nuclear safety, given the demographics of the UK nuclear industry. Maintenance of recorded information over long timescales with changing technologies is necessary to allow safe storage and disposal of waste and intelligent response to incidents. As the safety regulator, ONR's ability to influence improvement in nuclear safety relies on our inspectors having access to high quality knowledge. The ONR Strategy 2016-2020 [1,2] commits to continue to recruit, retain and develop staff so that we have a core workforce with the capacity and capability to deliver high quality, enabling nuclear regulation. Over the next 30 years, the UK plans to develop a geological disposal facility and UK EDF power stations will transition into decommissioning. The ONR Knowledge Management Strategy sets out the activities planned to maintain specialist knowledge to support staff development and build organisational resilience. Guidance on waste management is given in 'The management of higher activity radioactive waste on nuclear licensed sites' (referred to as the Joint Guidance [3]) published by ONR and the UK environmental regulators. It includes our expectations for managing information and records relating to radioactive waste, and introduces the concept of Radioactive Waste Management Cases (RWMCs). Our expectation is that knowledge and records must both be managed and transferred in such a way that future licensees are properly equipped to ensure that waste continues to be managed in a manner that prevents or minimises hazards to human health and the environment. RWMCs are an effective vehicle for demonstrating how the key elements of long-term safety and environmental performance will be delivered over the life cycle of the waste. They give a history of how the waste has been managed, capturing information and change over time. This paper will look at the challenges to managing knowledge in the UK nuclear industry, focusing on the safe management of radioactive waste, decommissioning and land quality. It will look at the problems faced by both licensees and ONR, drawing out common challenges. The paper will outline the approach taken within ONR to capture and communicate waste management and decommissioning knowledge and experience. Finally, it will look at how ONR regulates and influences how knowledge and records are dealt with in the industry to ensure safe nuclear liabilities management in the long term. (authors)

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DOE OSTI2018

International Standards Development and Its Role in Transitioning D and D Technologies to Deployment - 18315

As the need to perform deactivation and decommissioning (D and D) of the most challenging nuclear facilities worldwide is growing, facilitating the integration of state-of-the-art, cutting edge technologies will be critically important to safely, effectively, and efficiently complete this mission. However, end users in the field face significant obstacles to integrating new technologies into cleanup initiatives. By leveraging the development of international standards and testing protocols, the risks associated with implementing new technologies can be mitigated and further support the need to update existing directives and guidance handbooks to account for the tremendous recent advancements in technology development. The American Society for Testing and Materials (ASTM) International's E10.03 Subcommittee on Radiological Protection for D and D of Nuclear Facilities and Components has begun the development of standard specifications and uniform testing protocols designed to support the nuclear industry's testing and evaluation requirements for D and D technologies. The vision of the E10.03 Subcommittee is to promulgate operationally relevant, uniform testing protocols that can be leveraged by the various technology innovation and development programs across the nuclear environmental management community, and support decision makers and end users with common references in the selection and employment of those standards and associated technologies. These standards could also serve as the basis for updating existing directives and guidance documents such as the various DOE handbooks to better account for the role of cutting edge technology, and facilitate / expedite lab to market efforts that result in improved returns on investment (ROIs) for technology development programs. (authors)

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DOE OSTI2018

Managing Resin Wastes - Crossing the Digital Divide - 18326

The most important factor undercutting optimal waste management in the nuclear industry is developing a clear picture of the radioactivity content of the waste and its impact on waste disposal cost. For almost 35 years since the publishing of 10CFR61, waste characterization has relied on building its basis from samples collected from the final waste product as it is drawn from service or as it is packaged for disposal. This practice of sampling the product after it was collected continues to be followed to this day; 35 years after the publication of 10CFR61. This process, as it is followed, relies on minimum information and plant knowledge to formulate the characterization. While 10CFR61 and additional guidance documents promote the approach of using anticipated and statistically based scaling factors in addition to using data collected over time and a finite grouping of waste streams encompassing plant wastes, this guidance is often ignored in favor of using the earlier fallback options in the case of a limited sample base. The fallback allows the use of a final sample drawn from the waste when other data was not sufficiently available to be aligned to develop meaningful statistics. The determination of representativeness is obviated by the single sample resulting in a new waste stream being declared with every package. The sample used without statistical corroboration often is not representative. It would be unexpected if that sample were actually representative of anything more than itself. This paper discusses viable options for developing the basis for characterization through direct process monitoring to follow the accumulation of activity at the point of generation relying on hundreds of measurements being collected on a daily and weekly basis that can provide a much more reliable result along with significant cost savings. Direct monitoring of the accumulation of activity in process streams generating solid radwaste, including demineralizers and filter streams, is a more precise approach. Nearly all of these streams are monitored by plant chemistry on a regular schedule to maintain water quality and track fuel performance. The discussion in this paper will focus on resin bed tracking and how process knowledge drawn from routinely collected chemistry data can be brought together to form a consistent and precise solid waste radioactivity inventory. Some of the specific points covered in the paper include application and expansion of the fission product release computer program 3R-STAT, the importance of the spent fuel pool in radioactivity generation and its influence on the overall source term, and results from sample measurements collected over several years to corroborate using direct process measurements for the primary characterization basis. (authors)

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DOE OSTI2018

To the 'Safety related?' Question: What is Related to Safety? - 18430

Most, if not all, national programs for radioactive waste management (RWM) pledge their overall commitment to safety or - in the case of radioactive waste disposal - to long-term safety. Therefore, it may be somewhat surprising to find that the term 'safety' is hardly defined in these programs. The same holds for some of the core international guidance literature on the deep geological repository (DGR) 'safety case' concept. With respect to stakeholder concern over the safety of geological disposal, it seems, however, advisable to seek common ground in the understanding of the idea of 'safety'. In this paper, we review and discuss the most relevant international guidance literature's provision (or non-provision) of definitions of 'safety' in the context of radioactive waste disposal. Against this background, we develop a tentative, practical, glossary-style definition of 'safety' in the DGR context. As a seed of discussion, it may help to eventually expose possible mismatches in the base assumptions of safety experts and other stakeholders. (authors)

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DOE OSTI2018

Aging Management Considerations for Dry Cask Storage Canisters - 18472

Dry cask storage systems (DCSSs) are used to store used nuclear fuel (UNF) at independent spent fuel storage installations (ISFSIs) prior to final geologic storage or reprocessing. As storage durations increase beyond the twenty-year period commonly implemented for licensing the storage sites, it is important to consider aging related degradation issues for DCSS components. A large portion of the DCSS designs used throughout the world employ a ventilated concrete overpack with a welded stainless steel canister that confines the UNF. This paper explores key issues related to aging management for these canister components, including: (1) canister environment, (2) inspection limitations, and (3) aging management program guidance and requirements. While the austenitic stainless steel used for the canisters are generally resistant to most forms of environmental attack, it is possible for conditions to arise that lead to pitting and even stress corrosion cracking of the canister systems. A large number of studies related to canister degradation have been performed by industry groups and government researchers, both in the US and worldwide. These advances in the collective knowledge provide insight into risk factors for DCSS. Such information is beneficial to improve ISFSI siting decisions and ensure successful relicensing of existing ISFSIs. This paper highlights relevant research and also discusses applicability to degradation of other equipment and containers used to process and isolate UNF and other nuclear waste streams from the public and the environment. Inspection of the canister surface is an important tool to investigate and evaluate the potential for degradation. Existing DCSS designs typically do not offer ready access to the canister surface in order to perform the inspection. Further complicating inspection efforts are the surface conditions present on the sheltered canisters after extended storage times, including atmospheric debris such as dust and pollen. Given ongoing development efforts, inspection capabilities and associated limitations should be assessed at multiple points in time over the course of an aging management program to incorporate new advances in the field. This paper explores the limitations of currently available capabilities in inspection and delivery, and provides insight into the gaps remaining to be filled to permit an effective inspection program. Recommendations and requirements related to programs for aging management of canisters have been published or are in progress from a variety of groups, including industry groups, regulatory authorities, and Codes and Standards. An effective aging management program should consider the technical bases of these various inputs and maintain compliance with requirements. Given multiple sources, decisions will be required when considering conflicting guidance, and this paper will include discussion of the relative merits of the various approaches. (author)

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DOE OSTI2018

DOE/NNSA Efforts to Facilitate the Disposal of Risk Significant Sealed Sources - 18590

Commercial disposal of high-activity Class B and C disused sources has been particularly challenging in the U.S. For many years, a lack of disposal facility access, a shortage of certified of Type B transportation containers, and outdated regulatory guidance had effectively relegated a large number of risk-significant sources to indefinite storage. Progress has been made in all of these efforts. The DOE/NNSA Office of Radiological Security (ORS) and the Conference of Radiation Control Program Directors (CRCPD) Source Collection and Threat Reduction (SCATR) piloted the commercial disposal of a high-activity Cs- 137 source. To facilitate implementation of the NRC's revised waste classification guidance on sealed sources. Disposal of the 560 Ci blood irradiator at the U.S. Ecology facility in Richmond has helped to establish the availability of disposal access for similar sources under the 'alternative approach' provisions of the NRC's 2015 Concentration Averaging and Encapsulation Branch Technical Position (CA BTP). However, while these developments are significant for making disposal access available, the available information suggests that commercial sealed source users continue to keep disused sources in storage. This may be due to remaining uncertainty regarding disposal access, or due to the high cost of commercial disposal. (authors)

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DOE OSTI2019

Radon Considerations at Environmental Remediation Sites - 19469

Radon and how it is regulated poses many challenges to the U.S. Army Corps of Engineers (USACE) in executing the Formerly Utilized Sites Remedial Action Program (FUSRAP). These challenges include:, terminology; radon measurement methods and limitations; state certification and testing requirements; limited and potentially conflicting federal guidance; background considerations; and limited experience regarding application of guidance; among others. These challenges and others are discussed in detail. The FUSRAP Maywood Superfund Site (FMSS) is presented as a case study demonstrating these challenges and considerations to mitigate or address the challenges. The USACE, U.S. Environmental Protection Agency (USEPA), and the New Jersey Department of Environmental Protection (NJDEP) (NJDEP) have spent several years working on and continue to work through the radon issues at FMSS. The FMSS experience, information, and solutions presented in this paper will be helpful to other environmental sites with radium and thorium contamination. (authors)

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DOE OSTI2019

Designing a Site Closure Strategy for Uranium in Groundwater and Surface Water - 19534

A site closure strategy was designed and mapped for residual uranium present in groundwater and surface water at the US DOE Office of Legacy Management (LM) Monticello Mill Tailings Site (MMTS) Operable Unit (OU) III, which is regulated under CERCLA. The MMTS is a former mill site that was used to process uranium and vanadium ore in southeastern Utah until 1960. OU III includes surface water in Montezuma Creek and groundwater present in a shallow alluvial aquifer along the creek valley for several kilometers downgradient of the former mill site. The current remedy consists of an operating pump-and-treat system, monitored natural attenuation (MNA), and institutional controls (ICs). Performance evaluations indicate that a prolonged restoration period is likely, as well as an asymptotic trend in uranium concentrations that still exceed the remediation goal. Multiple strategies and scenarios were evaluated before selecting a preferred site closure strategy. Recommendations were developed to support advancing the site towards closure over the next few years and guide the process of regulatory interactions and decision making at the MMTS OU III. Site closure is an important milestone that reflects successful remedial progress at LM sites. Post-closure site management would be maintained to ensure conditions are protective of human health and the environment. The site closure strategy was considered in the context of site remedial progress. The primary sources of groundwater contamination have been removed, including mill tailings and contaminated soils, sediment, and debris. Five-year reviews have regularly concluded that the remedy is protective of human health and environment and that ICs are effective in preventing exposure to contaminated groundwater. The affected aquifer is low-yielding and has no history of domestic use. In 2004, the final remedy selected in the Record of Decision (ROD) was MNA with ICs. However, a pump-and-treat system was installed and continues to operate as a contingency remedy in the area with the highest residual uranium concentrations. The need for a contingency pump-and-treat system was triggered by a departure of the long-term monitoring data from model projections that were used as the basis for remedy performance metrics specified in the ROD. In 2015, the pump-and-treat system was expanded. Average uranium concentrations have declined locally over time. However, monitoring data and column studies indicate that the system effectiveness will eventually become ineffective in reducing uranium concentrations, likely due to the reequilibration of uranium present in the solid matrix in various states (i.e., sorbed, precipitated). Currently, there are no agreed-upon metrics for discontinuing the contingency pump-and-treat system. Three closure strategies effectively applied at other CERCLA sites were considered at MMTS OU III: (1) restoring natural attenuation conditions and conducting MNA; (2) applying alternate concentration limits (ACLs); and (3) waiving a remedial goal based on technical impracticability (TI). Each closure strategy was evaluated in the following manner: (1) identifying the criteria and requirements associated with current regulations, policy, and guidance, (2) assessing the ability to meet those criteria based on the current site knowledge and the conceptual site model, and (3) identifying data gaps that need to be filled to further assess the closure strategy. Following this evaluation, a preferred closure strategy was selected: to transition to a remedy based on MNA and maintenance of ICs, without active remediation. Several scenarios for transitioning to the preferred closure strategy were identified. Each scenario consisted of 'if/then logic' associated with a larger decision diagram. The decision diagram considered the current conceptual site model (CSM) and potential outcomes of studies that will be conducted to fill data gaps, as well as the performance of the existing treatment system and long-term monitoring data. Two scenarios for transitioning to the preferred closure strategy were highlighted for future planning and were described in detail. For both scenarios, the preferred closure strategy includes demonstrating criteria for MNA acceptance, pump-and-treat termination, long-term monitoring, and eventual site delisting. Recommended actions over the next few years were also developed, including updating the CSM, conducting numerical modeling to predict remedial performance and time frames, conducting geochemical analyses to assess uranium fate and transport, and summarizing the lines of evidence and basis for MNA. This detailed example of the development of a site-specific closure strategy provides the thought process and decision logic used to assess uncertainty using an if/then framework; identify common elements and recommended actions to advance the site towards closure; and navigate the inherent uncertainty in the timing of future events, outcome of regulatory discussions, and remedial progress over time that is common to many complex US DOE facilities. In addition, the national regulations, policy, guidance, and state of practice regarding commonly used site closure strategies was summarized for consideration, including MNA, ACLs, and TI. (authors)

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DOE OSTI2020

The Need for International Consensus and Acceptance of Free Released Material - 20177

World Nuclear Association (WNA) believes that the waste hierarchy is best implemented when a good compromise exists between disposal and reuse/recycling methods of valuable materials and waste generated by the nuclear industry. The rapid normalization of recycling in both domestic and industrial worlds strengthens WNA's position, as new facilities worldwide emerge providing treatment for clearance. As a foundation for reuse and recycling, the International Atomic Energy Agency (IAEA) provided international guidance on clearance (Safety Guide RSG 1.7) [1], defining associated threshold values approved by all member states. Nevertheless, despite this consensus, a lack of alignment of national regulations has occurred as countries have departed for this international guidance, leading to confusion, even between neighborhood countries. As a consequence, this discrepancy between national rules, despite industrial and technical international consensus based on radiological and environmental considerations appeared as a brake, weakening public acceptance as well as nuclear industry credibility. As a matter of fact, what is allowed in one side of a border may be forbidden on the other side. Conversely, an international consensus on clearance would promote: - Sustainability improvements, including preservation of natural and rare resources resulting from optimized recycling and reuse practices; - A broader acceptance of free released materials; - taking advantage of the available technologies worldwide in materials treatment aiming for clearance; - Increased sharing and use of methods, know-how and operators, leading to better efficiency, innovation, improvement as well as technical and financial mastery. (authors)

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DOE OSTI2020

Opportunities to Implement Solutions to Achieve Remedial Action Objectives in Consideration of Stakeholder Interests - 20365

A case study in the implementation of a soils excavation and removal project conducted at a Formerly Utilized Sites Remedial Action Program (FUSRAP) Maywood Superfund Site (FMSS) vicinity property; an active commercial business, in a densely populated area, with significant operational, technical and, logistical constraints that were expected to limit areas that could be remediated. The United States Army Corps of Engineers-led (USACE) team successfully navigated complex overlapping stakeholder interests to achieve a more efficient and complete removal of contaminated soils and debris while minimizing unnecessary excess costs to the Government. This paper centers on FUSRAP activities at the FMSS vicinity property located at 149-151 Maywood Avenue, Maywood, Bergen County, New Jersey. For most of its FUSRAP history, this ∼109,000 square meter (27-acre) property housed a now-demolished ∼26,000 square meter (6.5-acre) warehouse operated by Sears Logistics Services, a unit of the Sears Roebuck Company (Figure 2). Sears ended its property lease and vacated the warehouse in December 2016. Given that long history and for the purposes of this paper, the property will be referred to as the 'Sears property' or simply 'the property.' The Sears property is approximately 109,000 square meters (27-acres) and is currently zoned for commercial use. The property is bound to the north and northwest by 100 West Hunter Avenue (the Stepan Company), to the northeast by 205 Maywood Avenue, to the east by Maywood Avenue, to the south by 23 West Howcroft Road, and to the west by businesses on NJ Route 17 and the NJ Route 17 roadway. Until December 2016, the property housed a warehouse and distribution center operated by Sears Logistical Services. On-site structures were demolished by the property owner in 2017. The warehouse covered the north section of the property along the Stepan Company property line. A railroad spur from the adjacent property now known as the Maywood Interim Storage Site (MISS) ended at the northeast corner of the warehouse. Wetlands were located east of the warehouse; the rest of the property was covered with paved parking lots and grassed areas. During remediation at the property, opportunities frequently arose where the project team was able to coordinate effectively with stakeholders to sequence remediation activities to facilitate removal of otherwise inaccessible soils and identify opportunities for materials reuse when supported by residual radiological and chemical levels. Some specific opportunities include: - remediating individual truck loading dock bays to maintain tenant operations; - working along a busy highway and in a utility corridor containing a 30-inch high pressure gas main; - remediating Lodi Brook and associated wetlands requiring bypass pumping and other diversionary structures to maintain local community stormwater drainage; - protecting during remediation and supporting access to facilitate tenant/owner maintenance of critical fire protection and water supply system service lines buried in contaminated soils and necessary for safe warehouse operations; and - coordinating with stakeholders to ensure non-radiological contaminants of concern for the site (unaffiliated with FUSRAP) were addressed by the responsible party in a manner that maximized benefit to all parties while supporting an efficient overall site remediation program. Once the property was vacant and the warehouse structure and radiologically non-impacted above-grade structures were demolished, the project identified an opportunity to significantly reduce the volume of waste associated with the foundation of the former warehouse, a foundation suspected of being partially constructed in radiologically contaminated soils. The project developed and implemented supplemental radiological verification survey and sampling strategies based on radiological cross-contamination risk potential with process flow-charts and screening-level based decision points; a program that classified saw-cut sections of concrete based on observed residual surface radioactivity conditions using a combination of gamma sensitive sodium-iodide scans and beta sensitive Geiger-Muller direct measurements. The proposed approach was reviewed with stakeholders with feedback, including consideration of applicable State of New Jersey Site Remediation Program criteria and guidance before implementation. Once classified, additional sampling was performed at frequencies driven by screening. The sampling methods, stockpile sampling frequencies, criteria and, approach to data evaluation were developed in consideration of existing site cleanup standards, regional background ranges, State of New Jersey radiological remediation program guidance, and specific stakeholder input With the property remediation and survey efforts nearly completed, it is relevant to examine retrospectively the objectives, assumptions and limitations in the remedial design (i.e., what was planned) versus what was able to be accomplished through effective teaming. Benefits to other environmental remediation site programs include better understanding of how to work effectively with stakeholders to achieve win-win outcomes, and approaches to site remediation, waste minimization and materials reuse that were successful in their overall outcomes. (authors)

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

Development of an update to ISA S67.04 and RP 67.04: 'Setpoints for Nuclear Safety-Related Instrumentation for Nuclear Power Plants' - 146

The purpose of this paper is to provide an overview of the update to International Society of Automation (ISA) Standard ISA S67.04 - 2006, 'Setpoints for Nuclear Safety-Related Instrumentation for Nuclear Power Plants', and ISA RP 67.04-2000, 'Methodologies for Determination of Setpoints for Nuclear Safety-Related Instrumentation'. Significant advances have occurred on the utility, Nuclear Steam Supply System (NSSS) vendor, and regulatory side, related to improved techniques for analysis and monitoring. These and other changes, including those related to Technical Specifications and surveillance interval extensions, are being included in the latest update to this primary standard for setpoint control in nuclear plants in the U.S. and around the world. In 2016, the ISA S67.04 Committee agreed to proceed with a revision to this Standard and Recommended Practice, with major updates in the following areas: 1. Definitions - Updates for improved definitions based on TSTF-493, Rev 4, RIS-2006-17 and NRC updated Regulatory Guide (RG) 1.105 2. 95/95 - Updates based on improved guidance on the development of component input data and analysis techniques for 95% probability/95% confidence calculations 3. TSTF-493, RIS 2006-17 updates for Technical Specification input and to support changes 4. NRC input from updates in the Regulatory Guide 1.105 and Branch Technical Position BTP-7-12 revision process. 5. Improved guidance from the Electric Power Research Institute (EPRI) in References 7, 8 and 9. These standards are being updated with a new version anticipated to be issued as a draft in 2017 and submitted for ISA approval in 2019. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Implementation of wireless technologies in nuclear power plants' electromagnetic environment using cognitive radio system - 153

Wireless technologies have increased efficiency in many industrial settings across the world by enabling a mobile workforce and improving communications. Unfortunately, the nuclear power industry has been slow to adopt wireless technology as a result of safety, security, and reliability concerns expressed by regulators and others. A Research and Development (R and D) effort is currently being conducted to address two significant technical concerns for the introduction of wireless technology in nuclear power plants, which are electromagnetic compatibility (EMC) and wireless coexistence. Most equipment in nuclear power plants has never been tested for vulnerability to wireless transmission. As such, the impact of modern wireless devices on nuclear safety and plant reliability is not understood. Existing guidance from the Nuclear Regulatory Commission (NRC) and Electric Power Research Institute (EPRI) rely on the use of exclusion zones to protect plant equipment, which can result in exclusion zones of up to eight feet for some tablet devices rendering them ineffective for the mobile workforce. Another unknown is how different wireless devices will operate in close proximity to one another in the harsh electromagnetic environment of a nuclear power plant. These issues are being addressed through development of a cognitive radio system that has the ability to generate and output multiple wireless signals such as Wi-Fi, Bluetooth, and cellular communications at varying power levels and frequencies. The cognitive radio system is a modular system that can test nuclear plant equipment located in training areas, simulators, and the actual plant environment. In addition to using the system for testing plant equipment to wireless signals and EMC standards, the system can also monitor the radio spectrum for usage in other plant focuses such as cyber security. The R and D work will provide the objective guidance for the nuclear industry in the two areas mentioned and will develop a cognitive radio system for in-plant EMC and coexistence testing. The system will assist in the seamless use of wireless technology in nuclear power plants while greatly reducing the risks wireless signals may introduce. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Software Quality Metrics for Geant4: An Initial Assessment - Paper 122

In the context of critical applications, such as shielding and radiation protection, ensuring the quality of simulation software they depend on is of utmost importance. The assessment of simulation software quality is crucial both to determine its adoption in experimental applications and to guarantee reproducibility of outcome over time. Geant4 is a simulation system used in a wide range of experimental applications, some of which, such as radiation protection and biomedical studies, may involve critical use cases and would especially profit from an objective assessment of the software quality. In this study, we present the first results from an ongoing analysis of Geant4 code based on established software metrics from the perspective of software engineering. The analysis evaluates the current status of the code to quantify its characteristics with respect to documented quality standards; further appraisals concern evolutions over a series of release distributions. The quantitative assessment of the software is initially focused on a subset of Geant4 packages, which play a key role in scientific applications and are representative of different software development processes. We describe the chosen software metrics that map the entities of a software system to numeric values through mathematical definitions, quantifying software attributes ranging from functionality to maintainability. To measure such metrics, we exploited a set of software metrics tools, i.e. programs that implement a set of software metrics definitions. They allow software engineers to assess a software system by extracting the required attributes from the software and providing the corresponding metric values. To correlate software metrics values in a well-defined way, we used software quality standard models, such as the ISO/IEC 9126 standard, which aggregate numerical values to aid quality analysis and assessment. With this work we aim to create a baseline for evaluating correlations between software quality embedded in the Geant4 development process and simulation observables produced by Geant4-based applications. Furthermore, we intend to highlight what metrics are most effective at evaluating radiation transport software quality. We are confident that the result will provide constructive guidance both to improve key software tools, such as Geant4, and to estimate their contribution in risk analysis concerning sensitive applications. The remainder of this paper is structured as follows. We first discuss the approach adopted to fulfil our assessment. Then, we present the setup of our experiments and the way we collected data. We provide an initial interpretation of a subset of data resulting from measurements made on the selected Geant4 packages. As the analysis is still ongoing at the time of submitting this abstract, some results will be documented in the final paper and in the conference presentation. With this work we aim to create a proper Geant4 data set for evaluating its software characteristics and identifying both strengths and weaknesses. In this initial study, we considered the geometry and processes packages, which play a major role in most simulation applications. The data analysis is in progress at the time of writing; its complexity is related to the differences among software metrics tools, to the large amount of data to be evaluated and to the pioneering character of this project. We are confident that the results will provide valuable indications for the Geant4 user community and constructive guidance to improve key Geant4 packages. Other Monte Carlo codes could profit from the methodology developed in this pilot project for similar assessments of their software quality. At the end of this study, we intend to create a baseline for evaluating correlations between software quality embedded in the Geant4 development process and simulation observables produced by Geant4-based applications. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗