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DOE OSTI2022

A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing

Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. Thirteen additional tests would be performed in the CABRI reactor over the next decade on fuel rods with higher burnups [1]. Additionally, a larger testing program at the NSRR reactor in Japan with high burnup fuel would uncover a similar trend of pre-DNB ruptures in high burnup test rods at lower-than-expected peak enthalpies [2]. Numerous out of pile testing programs involving a variety of innovative mechanical testing techniques have been employed in an attempt to better quantify the failure thresholds of corroded zirconium alloy cladding in these rapid heating and loading conditions [3][4][5]. While previously interim guidance had been issued, in June of 2020 the NRC officially published updated regulatory guidance to account of these pre-DNB failures in Regulatory Guide 1.236 [6]. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Waste Management Improvement Initiatives at Atomic Energy of Canada Limited - 13091

Atomic Energy of Canada Limited's (AECL) Chalk River Laboratories (CRL) has been in operation for over 60 years. Radioactive, mixed, hazardous and non-hazardous wastes have been and continue to be generated at CRL as a result of research and development, radioisotope production, reactor operation and facility decommissioning activities. AECL has implemented several improvement initiatives at CRL to simplify the interface between waste generators and waste receivers: - Introduction of trained Waste Officers representing their facilities or activities at CRL; - Establishment of a Waste Management Customer Support Service as a Single-Point of Contact to provide guidance to waste generators for all waste management processes; and - Implementation of a streamlined approach for waste identification with emphasis on early identification of waste types and potential disposition paths. As a result of implementing these improvement initiatives, improvements in waste management and waste transfer efficiencies have been realized at CRL. These included: 1) waste generators contacting the Customer Support Service for information or guidance instead of various waste receivers; 2) more clear and consistent guidance provided to waste generators for waste management through the Customer Support Service; 3) more consistent and correct waste information provided to waste receivers through Waste Officers, resulting in reduced time and resources required for waste management (i.e., overall cost); 4) improved waste minimization and segregation approaches, as identified by in-house Waste Officers; and 5) enhanced communication between waste generators and waste management groups. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

New Developments in the Technology Readiness Assessment Process in US DOE-EM - 13247

A Technology Readiness Assessment (TRA) is a systematic, metric-based process and accompanying report that evaluates the maturity of the technologies used in systems; it is designed to measure technology maturity using the Technology Readiness Level (TRL) scale pioneered by the National Aeronautics and Space Administration (NASA) in the 1980's. More recently, DoD has adopted and provided systematic guidance for performing TRAs and determining TRLs. In 2007 the GAO recommended that the DOE adopt the NASA/DoD methodology for evaluating technology maturity. Earlier, in 2006-2007, DOE-EM had conducted pilot TRAs on a number of projects at Hanford and Savannah River. In March 2008, DOE-EM issued a process guide, which established TRAs as an integral part of DOE-EM's Project Management Critical Decision Process. Since the development of its detailed TRA guidance in 2008, DOE-EM has continued to accumulate experience in the conduct of TRAs and the process for evaluating technology maturity. DOE has developed guidance on TRAs applicable department-wide. DOE-EM's experience with the TRA process, the evaluations that led to recently developed proposed revisions to the DOE-EM TRA/TMP Guide; the content of the proposed changes that incorporate the above lessons learned and insights are described. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

In Situ Disposal to Optimize Decommissioning Waste Management - 17455

Cavendish Fluor Partnership (CFP) is teaming with the UK government's Nuclear Decommissioning Authority (NDA) to explore opportunities for onsite disposal of structures and wastes from decommissioning of twelve Magnox nuclear sites. CFP is piloting implementation of draft guidance issued by the Environmental Agency (EA), Scottish Environmental Protection Agency (SEPA) and Natural Resources Wales (NRW) designed to optimise management of wastes generated in closing nuclear sites positioned across Scotland, Wales and England. Where it can be demonstrated to be the best overall solution for people and the environment, on-site / in situ disposal as an option for optimising waste management could save millions of pounds sterling in the overall costs for site closure as opposed to previous plans that required retrieval, packaging, transportation and offsite disposal of all radioactive wastes generated during site decommissioning. The UK government's Department for Business, Energy and Industrial Strategy (BEIS), NDA, and UK regulators are consulting with stakeholders on a proposal to amend legislation and enable better, more proportionate regulation of nuclear sites in the final stages of decommissioning and clean-up. Practical application of existing UK legislation typically means removing most, if not all, radioactive waste and residual contamination from the nuclear licensed site for disposal or management elsewhere, irrespective of any wider health and safety, social, environmental or economic considerations. Proposed changes to legislation would give greater weight to the environment agencies' guidance and provide for a more sustainable approach where optimization could include on-site/in situ disposal solutions and early opportunities to reuse land with appropriate regulatory oversight. Recent characterization activities at two reactor sites have identified areas of sub-grade low level contamination in structures and soil around used fuel ponds. CFP, a partnership of UK-based Cavendish Nuclear Limited and US-based Fluor Corporation, has initiated preliminary site modelling that indicates much of this material, including significant volumes of sub-grade structures would be conducive for in situ disposal under the new proposed guidance. Based on Fluor experience with similar decommissioning projects at US Department of Energy sites, CFP will look to leverage in situ disposal to optimise decommissioning waste management and consequently optimise overall facility decommissioning strategies. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Comparison of U.S. Special Permit/Exemption/Authorization Processes for Licensing Type B Radioactive Material Packages - 17513

In the United States, there are three regulators (agencies) associated with licensing Type B radioactive material packages for shipping. They are the U.S. Department of Energy (DOE), the United States Nuclear Regulatory Commission (NRC), and the United States Department of Transportation (DOT). Each regulator has requirements and guidance which implement United States Code of Federal Regulations 10 CFR 71 - Packaging and Transportation of Radioactive Material and 49 CFR 171 through 180 - Hazardous Materials Regulations. Periodically, select packages may not clearly fit within the requirements or guidance. In such cases, the planned shipper can petition the applicable agency for a special permit/exemption/authorization approving the shipment with the specified deviations from the requirements or guidance. If equivalent safety is successfully demonstrated to the applicable regulator, such authorization may be obtained. This paper compares the: 1) DOE exemption process; 2) the NRC special package authorization process; and, 3) the DOT special permit process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Andra's strategy and approach for operational safety of French underground facility Cigeo - 15127

According to the French Act on radioactive waste of 28 June 2006, Andra shall design a reversible repository. In response to this demand, Andra is developing the industrial project namely 'Cigeo', a reversible geological disposal facility for HLW and ILW located in Meuse/Haute-Marne. The previous safety cases of 2005, 2009 and 2012 give a sound basis to assess Cigeo's safety both for the operational and post-closure periods. Regarding operational safety, in this new stage of the project, the challenging issues for the preparation of the safety case are the following: - To adapt the various regulatory frameworks (nuclear and non-nuclear) to the underground facility - To design the underground facility and handling equipment's to reduce at first risks in the underground facility as such radioactive particles release, gas explosion, fire and co-activity and limit their consequence. - To establish waste acceptance criteria to be fulfilled by waste producers in line with operational safety and design options. The objective of the paper is to present the experience of Andra in operation safety of the underground facility of Cigeo. Operational safety for Cigeo is assessed as much as possible according to the existing French regulations and guidance for nuclear facilities. Indeed, compared to existing nuclear facilities, a geological repository is a specific case due to the existence of underground facilities with particularities such as: - ventilation management in a large scale underground facility, - fire risk management. As an input for the industrial design of Cigeo for operation, Andra defined a fire guidance and a specific reference safety guidance, based on the adaptation of the French nuclear regulatory texts for and when feasible to mines and tunnels regulatory texts and on its own experience gained through safety operational reports since 2005 and their reviews by the Safety Authority. - underground co-activity of nuclear operation with construction works. According to the common 'defense-in depth' principle for nuclear facilities, these guidelines for the industrial design of Cigeo for operation consider the successive levels of defense of: - Preventing the risk and its spreading by design and construction; - Detecting incidents and early fighting by design and operation dispositions (detection and control systems, survey procedure); - Controlling incidents and limiting of their consequences by safeguarding systems, specific procedures. At that stage, design options for Cigeo constitute an adequate response to the operational safety functions and requirements: - Two levels of containment are set up: the first one is ensured by the waste package, the second one is ensured from the surface facility to the disposal vaults by the transfer cask during the transfer of waste packages in the disposal vaults by the disposal vault itself and a nuclear ventilation with HEPA filter; - Favorable options for fire risk prevention and mitigation: transport system minimizing heat load (cable car for the transfer from the surface, electric transfer engines), twin access tunnels to ease fire-fighting and evacuation...; - Permanent complete physical separation of nuclear zone and construction zone (including ventilation systems). In practice, operation safety assessment of Cigeo, designed following the aforesaid reference safety guidelines, follows a common risk analysis (internal aggressions such as auxiliary losses, chemical and toxic risks, fall or collision, fire, human factors and external aggressions such as earthquake, air traffic, flooding) with the following steps: - Identification of the major risks and the design basis scenarios that have to be considered for dimensioning of the installations and the safety devices; - On the basis of these different situations, those who will have to be quantified are identified; one or more representative and envelop situations are chosen in order to evaluate robustness of the installation. Quantitative evaluation of these situations permits to: - identify the consequences on man, environment and installation, and in particular to gain confidence in the maintaining of the safety functions; - identify or define the technic and/or organisational dispositions classified as 'important elements for protection' (IEP); - Situations beyond dimensioning permit to verify the robustness of the installation. They permit, on one hand to identify possible complementary material or organisational dispositions; on the other hand, to define the 'hard core' i.e. structures, systems and components that permit to maintain the confinement function of radioactive substance. (authors)

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DOE OSTI2015

US EPA superfund radiation risk assessment update: new, revised, and upcoming tools - 15350

The U.S. EPA Superfund remedial program is finishing up a significant number of revisions to its guidance for the risk assessment process at radioactively contaminated Superfund sites. The six Preliminary Remediation Goal (PRG) and Dose Compliance Concentration (DCC) internet based calculators for risk and dose assessment at Superfund sites are being revised to reflect better science, revisions to existing exposure scenarios and new scenarios, and changes to match up more closely with the EPA chemical regional screening level calculator. A comprehensive set of revisions to the PRG calculator will be finished in November 2014. A revised version of the 1999 guidance document that provides an overview for the Superfund risk assessment process at radioactively contaminated sites, 'Radiation Risk Assessment At CERCLA Sites: Q and A' was issued on June 13, 2014 reflects Superfund recommended guidance and other technical documents issued over the past 15 years. Also on June 13, 2014 EPA issued a new tool-kit comprised of 22 fact sheets that was developed to help the general public understand more about EPA's risk assessment process used at radioactively contaminated Superfund sites. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Revision of the Branch Technical Position on Concentration Averaging and Encapsulation - 16315

On February 25, 2015, the U.S. Nuclear Regulatory Commission (NRC) issued Revision 1 of the Branch Technical Position on Concentration Averaging and Encapsulation (CA BTP), culminating a six-year process. This revision provides updated guidance on the interpretation of paragraph 61.55(a)(8) of Title 10 of the Code of Federal Regulations, 'Determination of concentrations in wastes', as it applies to the classification of a variety of different types and forms of low-level radioactive waste (LLW) as Class A, B, or C waste. The waste class determines the minimum safety measures to be applied in order to provide reasonable assurance of safe disposal of the waste. The CA BTP is used by thousands of LLW generators, by the waste processing industry, by all operating LLW disposal sites in the U.S. and by Agreement States. The previous version of the CA BTP, published in 1995, was issued before the NRC adopted its risk-informed and performance-based regulatory policy. The revised CA BTP has been informed by that policy. The revised CA BTP also contains new guidance on acceptable methods that can be used to perform concentration averaging of LLW for the purpose of determining its waste class for disposal. This paper provides background information on why the CA BTP was issued and subsequently updated. This paper summarizes the new guidance found in the revised CA BTP. This paper also explains some of the key issues that required NRC's resolution, and the reasoning that underpins the major decisions reflected in Revision 1 of the CA BTP. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Endpoint Evaluation for 200-PW-1 Operable Unit Soil Vapor Extraction, Hanford Site, Washington - 16478

On the Central Plateau of the Hanford Site in semi-arid southeast Washington State, the 216-Z-9 Trench, 216-Z-1A Tile Field, and 216-Z-18 Crib were used from 1955 to 1973 for soil column disposal of liquid wastes containing carbon tetrachloride. The geological setting underneath these disposal areas consists of an approximately 70 m thick vadose zone with a nominally 5 m thick low-permeability layer at a depth of 30 m. Infiltration of the liquid waste through the vadose zone to the groundwater was driven by the liquid volume introduced during disposal. Carbon tetrachloride readily volatilises in the vadose zone, making soil vapor extraction (SVE) the preferred remediation technology. Between 1992 and 2012, SVE recovered 80,107 kg of carbon tetrachloride from the vadose zone below the three waste sites. Of this total recovery, 93 percent was removed during the first 5 years of operations. The last 7 percent was recovered in the last 16 years of operation. In 2011, a final cleanup level of 100 ppmv for carbon tetrachloride in soil vapor was established in the record of decision for the 200-PW-1 Operable Unit, which includes the three carbon tetrachloride waste sites. The final cleanup decision specified that soil vapor concentrations and the cleanup level would be further refined and assessed in the future to ensure groundwater protection. A subsequent treatability test was conducted and the results were used to further refine the definition of vadose zone source conditions that are protective of groundwater. Based on this treatability test, predictions indicate that current and future vadose zone conditions will meet groundwater protection objectives. SVE operations were not performed in 2013, 2014, and 2015, and soil vapor concentrations in all SVE wells (and all but two monitoring probes installed using direct-push technology) remained below the 100 ppmv cleanup level during this time. SVE endpoint guidance was prepared in 2013 as a joint effort between Pacific Northwest National Laboratory, U.S. Army Corps of Engineers, and U.S. Environmental Protection Agency. This SVE guidance was used to develop a path forward for the 200-PW-1 Operable Unit SVE remedy. The guidance asks the following questions: 1. Are data collected to date adequate to support a well-defined conceptual site model? 2. Have remediation goals (cleanup levels) been defined? 3. Are environmental pathways and risk understood well enough to support site closeout? 4. Will remaining contamination in the vadose zone cause groundwater cleanup levels to be exceeded? If the answer to the first three questions is 'Yes', and the answer to the fourth question is 'No', then the site is ready for closure. At this time, the site has compiled appropriate information to answer these questions to support a remedy decision to terminate SVE operations. The U.S. Environmental Protection Agency recently provided concurrence that termination of SVE operations within the 200-PW-1 Operable Unit is appropriate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Cost Estimates for NPP Decommissioning Projects and Insights from Cost Analysis - 18118

The technical ability to decommission NPPs shut down after normal operation has been well demonstrated. However, concerns remain about the ability to accurately calculate and demonstrate the validity of decommissioning cost estimates, and to control costs during decommissioning. This paper describes recent international guidance on quality cost estimates for nuclear decommissioning projects and explores the potential added value that could be attained through more systematic analysis and reporting of decommissioning project costs. The OECD Nuclear Energy Agency (NEA) and the International Atomic Energy Agency (IAEA) have long been concerned about the need to promote greater transparency in decommissioning costing and have collaborated in a number of initiatives in pursuit of this objective. Together with the European Commission, they published 'The International Structure for Decommissioning Costing (ISDC)' in 2012. The ISDC provides a useful common reporting format for decommissioning costing. The NEA has also published guidance on the international peer reviews of decommissioning cost studies (2014) and on the practice of decommissioning cost estimation (2015). In September 2017, the NEA and IAEA published detailed guidance on addressing uncertainties in decommissioning cost estimates (the 'Addressing Uncertainties' report). This joint report sets out an approach to treating uncertainties reflecting current good practices in cost estimating. Specifically, it describes how uncertainties in decommissioning cost estimation can be addressed using standardized methods of estimating uncertainty and risk analysis. Its recommendations aim at enabling a consistent and comprehensive treatment of risk and uncertainty in the preparation of decommissioning cost estimates. Together these publications establish a good basis for producing comprehensive and high-quality cost estimates for nuclear decommissioning projects and for enhancing confidence in the estimates through improving transparency concerning the underpinning data and calculations, and more explicit representation of the uncertainties that may impact on project costs. There is increasing attention on understanding variations between cost estimates and apparent escalation of decommissioning costs. This arises both from the need to assure the adequacy of financing arrangements for future decommissioning projects and the ambition to improve the performance and ensure value-for-money in the delivery of decommissioning projects and services. A systematic analysis of actual decommissioning project costs in relation to the cost estimates would be expected to yield valuable insights about key cost drivers in actual decommissioning projects, and the extent to which current cost estimates accurately capture and reflect these. Aligning the reporting of decommissioning project costs with the underpinning cost estimates for the projects would also facilitate the identification of causes of major deviations between projected and actual costs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Radiological Characterization for Decommissioning: International Good Practice and Experience - 18396

The 'Task Group on Radiological Characterization for Decommissioning (TGRCD)' functions under the umbrella of the Nuclear Energy Agency' (NEA) 'Working Party on Decommissioning and Dismantling (WPDD),' which is currently reporting to NEA Radioactive Waste Management Committee (RWMC). The TGRCD task group, which is composed of international experts, has been working since 2014 on a commitment to provide a high-level guidance for characterization practitioners working in nuclear decommissioning. The work has focused on facility characterization from a waste and materials perspective. Radiological characterization is a key activity enabling planning and implementation of decommissioning of nuclear facilities. Effective characterization allows the extent, location and nature of contamination to be determined providing crucial information to facilitate facility dismantling, the management of material and arising waste, protection of workers, the public, the environment and associated cost estimation. The report is written for characterization practitioners who carry out the tactical planning, preparation, optimization and implementation of characterization to support the decommissioning of nuclear facilities and the management of associated materials and waste. It compiles recent experience of NEA member countries in radiological characterization, including the experience of international experts and practitioners; the learning from international case studies and international conferences; and national approaches to the application of international and national regulations, standards and guidance documents. Using this comprehensive approach and evaluation studies, the report identifies relevant good practices and sets out set out practical guidance covering all stages of the characterization process as related to decommissioning. This paper provides a summary of the work conducted and the main findings from the detailed report. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Challenges to Evaluating Ecological Risks in Performance Assessments - 18424

Performance assessments (PAs), whether probabilistic or deterministic, have largely focused on evaluating human health effects of radionuclides. In the past, it was asserted that protecting human health would automatically protect the environment. This blanket assertion has been questioned, so tools and methodologies to directly estimate non-human biota dose have been developed and made available since the early 2000's. For example, the Biota Dose Assessment Committee (BDAC) guidance led to the development of the RESRAD-BIOTA (RESRAD stands for Residual Radioactive materials) software to implement this guidance. The BDAC was a US Department of Energy (DOE) led working group with members from academia and consultancies in addition to DOE technical staff. Parallel efforts have been in place in Europe with the development of the ERICA tool (Environmental Risk from Ionising Contaminants: Assessment and Management). More recently, interpretation of the requirements of US Department of Energy (DOE) Order O458.1 have led some waste management facilities to include ecological risks in addition to human health effects in their PAs. We have recently developed methodologies to assess ecological risks within probabilistic PAs (PPAs) for two Federal facilities. One of the challenges in applying the non-human biota assessment models to PPAs is that ecological risk assessments following US Environmental Protection Agency (EPA) or DOE guidance start with protective screening assessments before moving into more site-specific ecological risk assessments. Conservative assumptions bias the screening assessments and greatly overestimate the risks posed by contaminants. However, although the screens are biased they can be useful in ruling out model spatial domain areas or types of biota from requiring further assessment. In some cases, the results of the screen could entirely rule out ecological risks from further assessment. For a more refined and unbiased estimate of dose to non-human biota, one would typically conduct site-specific studies to better define exposure through the food web or adverse effects based on eco-toxicity studies of environmental, and generally weathered, contaminant mixtures. However, for PPAs that consider future site conditions after the loss of institutional control, such site-specific studies are not possible. Rather, site-specific future ecological risk models use probabilistic inputs that result in a distribution of population doses or exposures and calculate the distribution of possible adverse effects to ecological populations. This approach considers both radionuclide and chemical inventory. For both Federal facilities, the model spatial domains include riparian and aquatic environments located downstream of the waste management site. These riparian and aquatic environments represent linear exposure elements for some wildlife populations and require special consideration of exposures to individuals and to populations along such spatial domains. At a minimum, wildlife foraging should be focused along the axis of such water bodies. Another challenge is how to incorporate ecologically relevant spatial scales into models that have contaminant transport as the primary focus. Riparian and aquatic environments also have threatened and endangered (T and E) species, which necessitate greater levels of protection given their rarity. A final challenge for conducting ecological risk assessments is how to evaluate T and E species in the future. This might be addressed by evaluating potential for impacts on individuals of feeding guilds with species that are currently threatened or endangered. We also have to consider how ecological populations might change as a result of climate change. It is a more difficult problem to predict what species might become rare in the future, and how to best assess future populations of threatened or endangered species. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

EM Program Expectations on Implementation of DOE O 458: Radiation Protection of Public/Environment - 19553

DOE Order 458.1, Radiation Protection of the Public and Environment, was issued in 2011. DOE sites were given 18 months to implement. The revised Order promulgates requirements for property clearance actions and independent verification. DOE Order 458.1 allows for preapproved authorized limits for the clearance of property. However, when property with the potential to contain residual radioactive material does not meet preapproved authorized limits, an authorized limits package must be developed. The Order contains a range of requirements for demonstrating with reasonable assurance that dose constraints are met for any actual or likely future use of the property to be released. EM has developed a draft guidance document that provides the elements of an authorized limits package. These elements are used to review all authorized limits packages that are submitted to EM. Expected elements include property description and process knowledge documentation, current and likely future use, the conceptual site model, dose modeling methodologies, specific authorized limits, dose- or concentration-based constraints, collective dose estimate to the potentially exposed population, stakeholder communications, description of the As Low As Reasonably Achievable (ALARA) analysis, implementation plan, evaluation of difficult to assess surfaces, and independent verification plan. The draft guidance document also provides further explanation of the desired elements of an independent verification (IV) plan. An IV plan should describe the planning phase, implementation phase, assessment phase, and decision phase of an activity. The planning phase includes scope and cost, concurrent with cleanup contractor procurement, and review and approval of site radiological survey clearance program and IV program plans. The implementation phase assures sufficient collection of data and adjustment can be made as necessary. During the assessment phase, data and reports are reviewed, and coordination between the clean-up contractor and IV personnel occurs. A decision of whether or not the clearance actions need to be altered occurs during the decision phase. IV certification that the clearance action is complete also occurs during the decision phase. Issuance of the final guidance document is meant to improve the consultation process and minimize iterative interactions with EM. Standardizing and formalizing property clearance and IV actions also helps assure that cleanup actions are transparent and sustainable for future scrutiny. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

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

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

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Effects of adaptive automation on operator-system performance - 103

Adaptive automation (AA) is the dynamic, real-time change in the degree of automation in response to situational changes. AA systems are being developed in many industries, including for nuclear power plant applications. AA is a promising means of mitigating human performance issues that often arise in highly automated systems. The purpose of this Nuclear Regulatory Commission study was to characterize important AA design features, define the effects of AA on performance, examine the human factors engineering (HFE) guidance available to design and evaluate AA systems, and identify research and development needs. In general, we found that AA improves task performance and situation awareness (SA) while managing workload. Although the research is limited, AA also supports operator recognition of automation failure and recovery. Research on human-system interfaces has identified principles showing promise for supporting SA, including failure detection, and limiting workload. Research further suggests that AA systems are both more effective, and less disruptive and distracting, when they follow rules of interaction similar to those used by human crewmembers. Our general conclusion about the HFE guidance available to designers and reviewers is that it is limited and has not kept pace with automation technology. We discuss several research and development needs that will help support a better understanding of where and how AA should be used and support the development of more comprehensive guidance for AA safety reviews. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Computer Security for I and C Systems at Nuclear Facilities - 116

Cyber-attacks on Instrumentation and Control (I and C) systems may jeopardize the safety and security of nuclear facilities by contributing to sabotage or aiding in the unauthorized removal of nuclear material. The effects of cyber-attacks on I and C systems important to safety to safety may result in a wide range of consequences, such as a temporary loss of process control or unacceptable radiological consequences. Public awareness of cyber-attacks that affect I and C systems may also undermine confidence in the safety and security of nuclear facilities. The International Atomic Energy Agency (IAEA) recognizes the need for the guidance on the protection of I and C systems that provide safety, security, or auxiliary functions at nuclear facilities from cyber-attacks. To meet this need, The IAEA has developed a new Nuclear Security Series publication with the provisional title Computer Security of Instrumentation and Control Systems at Nuclear Facilities which builds upon and expands the guidance published in Nuclear Security Series No. 17 Computer Security at Nuclear Facilities (2011) . This publication will be formally released in 2017 as Technical Guidance and will be the second international consensus publication dedicated to Computer Security at Nuclear Facilities. It was developed in parallel with the new IAEA Nuclear Safety publications providing recommendations on I and C systems at research reactors (RR) and nuclear power plants (NPP) . The purpose of this paper is to provide an overview of this publication and its application. The paper will also discuss its relation to other IAEA Nuclear Security Series (NSS) publications on computer security. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Electromagnetic compatibility qualification of power electronics for I and C power applications - 152

Electromagnetic Compatibility (EMC) is a growing field in nuclear equipment qualification. Previously confined to I and C equipment residing in control and cable spreading rooms, EMC qualification has expanded to include the inverters and battery chargers which often power safety-related I and C equipment, as well as adjustable speed drives and material handling equipment which could potentially affect sensitive safety-related I and C systems. Because much of the data collected to form the current EMC limits and guidance was collected from low-power equipment in relatively low EMI environments, the same guidance, applied to power electronic equipment, produces overly conservative estimations of the performance needed for EMC. Excessive modification in an attempt to bring the equipment into compliance with EMC limits may not be necessary. This paper provides a brief summary of current applicable EMC guidance and demonstrates a road map for the qualification of larger, more powerful devices which takes into account the entire EMC process, including design, testing, mapping of the installed location, and justification of EMC failures with the end goal being successful, timely, and cost-effective implementation of upgrades and compatibility with existing plant infrastructure (in the intended installation environment), even if equipment exceeds traditional operating envelopes. (authors)

42 ENGINEERING↗

DOE OSTI2017

Development of NUREG-0700, rev 3 - 229

The U.S. Nuclear Regulatory Commission (NRC) evaluates the human factors engineering (HFE) of nuclear power plant design and operations to protect public health and safety. The HFE safety reviews encompass both the design process and its products. Detailed guidance for these reviews is provided in two key documents: the HFE Program Review Model (NUREG-0711) and the Human-System Interface Design Review Guidelines (NUREG-0700). The NRC is committed to the periodic update and improvement of the guidance to ensure that they remain state-of-the-art design evaluation tools. In this paper, we will describe the updates made to NUREG-0700 that resulted in Revision 3 of the guidance. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗