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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Port Hope Area Initiative Project Update - 19146

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL), on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. With funding of $1.28 billion CAD it is currently one of the largest environmental remediation projects in Canada. The safety of workers, the public and the environment while delivering the PHAI on time and within budget is CNL's number one priority. The PHAI is being carried out as two projects - the Port Hope Project and the Port Granby Project. The Port Hope Project is the larger of the two and includes the cleanup of approximately 1.2 million cubic metres of historic waste from various locations within the community. The Port Granby Project will relocate approximately 450,000 cubic metres of historic low-level radioactive waste and marginally contaminated soils, located at an existing waste management facility on the shoreline of Lake Ontario, to a new engineered aboveground mound being built about a kilometre north of the current site. This paper will provide an update on both projects with an emphasis on successes to date and challenges yet be faced. (authors)

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DOE OSTI2013

Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the second shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)

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DOE OSTI2017

Preliminary Concept of Operations for the Spent Fuel Management System--WM2017

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the U.S. Department of Energy s Office of Nuclear Energy is tasked with identifying, planning, and conducting activities to lay the groundwork for developing interim storage and transportation capabilities in support of an integrated waste management system. The system will provide interim storage for commercial spent nuclear fuel (SNF) from reactor sites and deliver it to a repository. The system will also include multiple subsystems, potentially including; one or more interim storage facilities (ISF); one or more repositories; facilities to package and/or repackage SNF; and transportation systems. The project team is analyzing options for an integrated waste management system. To support analysis, the project team has developed a Concept of Operations document that describes both the potential integrated system and inter-dependencies between system components. The goal of this work is to aid systems analysts in the development of consistent models across the project, which involves multiple investigators. The Concept of Operations document will be updated periodically as new developments emerge. At a high level, SNF is expected to travel from reactors to a repository. SNF is first unloaded from reactors and placed in spent fuel pools for wet storage at utility sites. After the SNF has cooled enough to satisfy loading limits, it is placed in a container at reactor sites for storage and/or transportation. After transportation requirements are met, the SNF is transported to an ISF to store the SNF until a repository is developed or directly to a repository if available. While the high level operation of the system is straightforward, analysts must evaluate numerous alternative options. Alternative options include the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if any), repackaging technology, the types of containers used, repository design, component sizing, and timing of events. These alternative options arise due to technological, economic, or policy considerations. As new developments regularly emerge, the operational concepts will be periodically updated. This paper gives an overview of the different potential alternatives identified in the Concept of Operations document at a conceptual level.

Cumberland, Riley M↗
DOE OSTI2015

A cradle to grave decontamination process - 15266

As the rate of decommissioning of nuclear facilities is accelerating there is an ever increasing need for technologies which will decrease dose to operators, reduce overall cost and allow minimization of waste volumes. A new fully integrated decontamination and waste management process is described herein. This method approaches surface decontamination in an integrated fashion in that it deals with the full cycle of the process including the final disposition of the secondary wastes generated during the cleaning cycle. This approach also simplifies decontamination in that it addresses issues observed in previous chemical decontamination applications such as generation of orphan wastes and the requirements for multi-step reagent additions over repeated cycles. The waste management aspect of this system is key to providing a fully integrated process and deals with the issues related to grout formulations in the presence of high waste loadings. (authors)

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DOE OSTI2013

Safety aspects of dry spent fuel storage and spent fuel management

As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. After the events of Fukushima, the advantages of passively and inherently safe dry storage systems have become more obvious. TUV and BAM, who work as independent experts for the competent authorities, present the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields. All safety relevant issues like safe enclosure, shielding, removal of the decay heat or behavior of cask and building under accident conditions are checked and validated with state-of-the-art methods and computer codes before the license approval. It is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

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DOE OSTI2019

Transport and Treatment of Chapelcross Gas Ducts for Clearance and Recycling - 19662

In the summer of 2018, Cyclife, the EDF group's specialist radioactive waste management and decommissioning company, transported sixteen gas ducts and associated equipment from the shutdown Chapelcross nuclear power plant to treatment facilities in the UK and Sweden. The client requested a full removal, treatment and disposal service for approximately 820 te of low level metallic waste comprising Upper Gas Duct Assemblies (including bellows), Upper Gas Duct elbows, Filter Pots and Convoluted pipework. The complex scope involved heavy lifting, transport by road and sea, processing and recycling on a large scale, thereby necessitating involvement of a wide range of suppliers at different tiers all managed by Cyclife. The purpose of the entire project was to safely handle and treat all of the components for clearance and recycling into the public domain for the manufacture of new steel products. Most of the treatment took place in Cyclife's nuclear licensed metal treatment facility in Sweden. Cyclife has a long and unique experience of arranging safe transport of large contaminated components on road and sea. The metal treatment facility in Sweden can, without special arrangements, receive components up to 30 meters long with a weight of up to 400 tonnes for decontamination, size reduction and melting. The actual treatment of the components consisted of the following steps: a) Separation of a gas duct from the gantry, b) size reduction, radiological verification measurements and clearance of the gantry, c) stepwise segmentation of the gas duct and elbow including removal of asbestos, d) decontamination, e) melting, f) radiological assessment and clearance of the metal ingots, g) management of the secondary waste for return to UK. The result of the project, upon completed treatment of all the material, demonstrates a high recycling rate of the material. 5-10% of the original mass of material will have to disposed as radioactive waste. This number include all residues, untreatable material (asbestos, mineral wool, filters etc.) and two melt batches with a too high activity content for clearance. The paper takes the reader through the entire project from the initial planning to the clearance of the metal and the return of the secondary waste to UK for disposal. It will also present and discuss pros and cons for alternative management approaches in several perspectives. (authors)

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DOE OSTI2012

The Role of the Environmental Management Site Specific Advisory Board under the Federal Advisory Committee Act, With Emphasis on the Northern New Mexico Citizens' Advisory Board - 12482

The Environmental Management Site Specific Advisory Board (EM SSAB) is comprised of eight Citizens' Advisory Boards, chartered under the Federal Advisory Committee Act (FACA) by the Assistant Secretary for Environmental Management for the U. S. Department of Energy (DOE). The purpose of the EM SSAB is to provide the DOE with recommendations regarding Environmental Management issues from legacy waste produced at major sites across the DOE Complex. The Northern New Mexico Citizens' Advisory Board (NNMCAB) is the site specific advisory board to the DOE on issues of environmental monitoring, remediation, waste management and long-term stewardship at Los Alamos National Laboratory (LANL). The other boards are located at: Hanford, Idaho, Nevada, Paducah, Portsmouth, Oak Ridge and Savannah River. Using broad citizen outreach and input, the SSAB provides an official mechanism for the citizenry at each of these sites to monitor current activities affecting the region and also to have input into the prioritization of future activities. 'The mission of the EM SSAB is to more directly involve stakeholders in EM Planning and decision-making processes for the nuclear weapons complex cleanup. DOE has various means of involving the public in its planning and decision-making processes; the EM SSAB is only one component of EM's public participation program, and is not intended to be an exclusive means of public participation. It is the policy of DOE and EM to conduct it programs in an open and responsive manner, thereby encouraging and providing the opportunity for public participation in its planning and decision-making processes. EM SSAB members are appointed to a two year term and may serve up to three terms. During this time, members are able to hear many presentation from subject matter experts, attend several site tours at their site and across the DOE complex and are able to express their concerns and give input to the prioritization of clean up at each site. It is an excellent education and it is hoped that these members will go out to their communities, friends and family and discuss these issues in an informed manner. Many times our citizens rely only on newspaper articles or the Internet, which are sometimes replete with fear and misinformation. The more stakeholders are able to be truly informed, the more accurate the discourse will be at each of the DOE sites. (author)

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DOE OSTI2024

Presentation on the INL Remote-Handled Low-Level Waste Disposal Facility FY-2023 Annual Summary Report

The abstract below is from the report INL/RPT-24-76103. The powerpoint presentation contains information taken from the report. This Fiscal Year (FY) 2023 annual summary report (ASR) documents the continued adequacy of the performance assessment (PA), the composite analysis (CA) and associated operating disposal-authorization statement (ODAS) technical basis documents for the Remote Handled (RH) Low-Level Waste (LLW) Disposal Facility at Idaho National Laboratory. Annual review of the adequacy of the PA and CA for the RHLLW Disposal Facility ensures that conclusions of the analyses remain valid in accordance with requirements of Department of Energy (DOE) Order 435.1, “Radioactive Waste Management.” In FY 2023, no significant operational changes or other activities occurred that would cause deviation from the assumptions in the PA and CA pertaining to disposal geometry, verification of waste characteristics, tracking disposal inventories against total limits, facility closure design, or institutional controls. Sixteen total waste-canister shipments were received at the RHLLW Disposal Facility, and 15 total waste canisters were emplaced.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2015

Draft Infrastructure Framework for a Generic Repository Licensing Organization

This document is milestone M4FT-15SN0826011: Draft Summary of generic repository licensing infrastructure, the result of efforts under work Package FT-15SN082601 DOE Managed HLW and SNF Research: Establish a generic organization framework for records management, IT infrastructure, quality assurance, and systems engineering - Sandia National Laboratories (SNL). The objective of this effort was to develop this draft document identifying and summarizing the principle elements of a generic repository licensing infrastructure organizational framework. The FY16 follow on work package is FT-16SN050302- Establish organizational framework to meet regulator expectations - SNL, having the consistent objective to 'Develop a draft document identifying and summarizing the principle elements of a generic repository licensing infrastructure and organizational framework.' This document describes the framework for a generic hypothetical organization with the responsibility to site, characterize, design, license construct and operate a repository for the disposal of high-level radioactive waste (HLW) and spent nuclear fuel ( F) managed by the Department of Energy. organizational framework suggested is intended to be agnostic of the facility location and disposal medium. It is assumed that the facility is to be licensed under U.S. Nuclear Regulatory Commission (NRC) regulation with DOE as the applicant (licensee). Detailed organizational elements reflect a workforce composition and practices that facilitate compliance with NRC expectations and NQA-1 quality standards.

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DOE OSTI2017

Challenges in the UK Generic Design Assessment Process - 17062

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

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DOE OSTI2019

Improving Operational Efficiency, Compliance and Community Perception with Real-time Data - 19260

The nuclear and medical radiation industry has always faced a significant burden in effectively balancing exposure and productivity. Maintenance operations and development of new techniques require significant staffing resources to monitor and review personnel exposures and changes in operational radiation levels. Currently utilised monitoring technology and review techniques remain unchanged since the 1950's, relying upon mostly analogue instrumentation and paper reporting. After utilising isolated-area monitoring during a reactor decommissioning project it became apparent that current systems created false assurance as the units continually failed without warning due to 'dirty' power. These instrument failures could have easily led to an unnecessary exposure while operators continued to assume they were being safely monitored. This paper is a philosophical discussion on how current shortcomings could be addressed, security increased, and exposure risks minimised, by utilising real-time monitoring and automated reporting. The community and media expectations of reported data and nuclear monitoring systems continue to be coloured by sensational technology seen in movies and emerging technology available through smart homes and smart cities initiatives. For the general public to be more engaged with real radiation levels and effects, the nuclear industry needs to more effectively communicate safe levels. Organisations can have regain control of exposure levels discussions through the controlled release of background and expected exposure level data. An increase in real world monitoring data communicated in a controlled and more effective way will aid in the general public accepting, and being educated about, the minimal increase above natural background levels of the nuclear industry and other radiation facilities. This is particularly as Australia is in the process of implementing its Nuclear Waste Facility, with community acceptance the number one inhibitor of being able to move forward. Storage, mobile assets, and processing all present unique management, security, exposure, and telemetry challenges for a facility operations manager. Active monitoring and automated review of the monitoring data can alert operators to any unexpected events before any resultant increases in radiation levels cause an unplanned exposure to operators or the environment. Integration of radiation monitoring data with global positioning telemetry enables better organisational awareness of mobile asset location and early detection of any potential shifting and reduction in shielding integrity of sources and encapsulated waste during transport. Additionally, the real-time monitoring can be set-up to enhance an organisations' or nation's security capabilities. For instance, the real time monitor will send an automated alert in the case of an unexpected movement of a source through a monitored facility or for an unscheduled movement of a monitored vehicle. There are further benefits to be gained in control, ALARA principles and effluent management by increasing real-time monitoring capabilities through the stages of the waste cycle. Increasing operational effectiveness while delivering operational cost savings sounds like an attractive proposition. With increased access to operational area radiation data comes drastically improved real-time management of exposures for new, and experimental practices. Increased live data can be utilised to identify optimum practices for reducing operational and environmental exposures or releases. This increased operational data effectively comes with the added benefit of a reduction in the required health physics personnel resources for general operations. By reducing exposure and the detachment of required health physics resources from routine operations, there is greater availability for these resources to support maintenance operations and project-based work. The review of this increased data will create internal capability to identify low-- and high-risk activities, resulting in the ability to reduce the number of highly skilled workers in an industry where the availability of this talent is declining. During my time as the Radiation Safety and Nuclear Security Manager at a research facility, I implemented a real-time monitoring system. The system identified a secondary risk that had previously been unidentified and posed a higher exposure and security risk than the radiation practice being monitored. What stops industry from implementing technology that encourages agile work practices, while minimising security risks and personnel exposure? (authors)

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DOE OSTI2015

Draft Operating Procedures for a Generic Repository Licensing Organization

This document is milestone M4FT-15SN0826021: Draft Summary of principle operating procedures for a generic repository licensing organization, the result of efforts under work package FT-15SN082602 DOE Managed HLW and SNF Research: Generic Operating Procedures; e.g., Quality Assurance — Sandia National Laboratories (SNL). The objective of this effort was to develop this draft document identifying and summarizing the principle operating procedures for a generic repository licensing organization. The FY16 follow on work package is FT-165N050303: Develop and Implement Operating Procedures — SNL, having a consistent objective to develop a draft document identifying and summarizing the principle operating procedures for a generic repository licensing organization. This document identifies and describes the principle operating procedures for a hypothetical and generic organization with the responsibility to site, characterize, design, license, construct and operate a repository for the disposal of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) managed by the Department of Energy. These principle operating procedures reflect practices that facilitate compliance with U.S. Nuclear Regulatory Commission (NRC) expectations and NQA-1 quality standards.

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DOE OSTI1989

Nuclear Safety [Vol. 30, No. 1, January-March 1989]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 International Conference on Thermal Reactor Safety: NUCSAFE 88 by A. P. Malinauskas and J. G. Pruett; ACCIDENT ANALYSIS: 18 Chernobyl Today: State of Research by A. Yu. Gagarinskii; CONTROL AND INSTRUMENTATION: 23 Problems Associated with Routine In-Plant Radioactive Effluent Monitoring Systems at U.S. Light-Water Reactors by Barry G. Wahlig, David M. Walker, Mahmoud R. Ghavi, and John M. Palms; DESIGN FEATURES: 34 Twentieth DOE/NRC Nuclear Air-Cleaning Conference by R. R. Bellamy, D. W. Moeller, and M. W. First, 46 Design Options for the Core Support Structure for Liquid-Metal-Cooled Reactor Plants by J. P. Burelbach, W. J. Kann, J. G. Saiveau, and R. W. Seidensticker; ENVIRONMENTAL EFFECTS: 53 The Radiological Consequences of the Chernobyl Accident by L. V. Konstantinov and A. J. González; WASTE AND SPENT FUEL MANAGEMENT: 70 Activities Related to Waste Management Compiled by E. G. Silver; OPERATING EXPERIENCES: 80 Fuel Performance Annual Report for 1986 by W. J. Bailey, 103 Safety Considerations Related to 12-Hour Shift Schedules by S. Gould, 105 Reactor Shutdown Experience Compiled by J. W. Cletcher, 108 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 130 General Administrative Activities Compiled by E. G. Silver, 140 Reports, Standards, and Safety Guides by D. S. Queener, 145 Status of Power-Reactor Licensing Activities Compiled by E. G. Silver, 153 Proposed Rule Changes as of Sept. 30, 1988; ANNOUNCEMENTS: 69 Northwestern University Short Course on Radiation Safety, 79 International Seminar on Fission Product Transport Processes in Reactor Accidents, 129 Harvard School of Public Health Offers Short Courses, 152 MIT Offers Summer Short Courses on Reactor Safety and Nodal Analysis, 152 Seminar on Use of PCs in Probabilistic Safety Analysis, 157 The Authors, 160 Indexes to Nuclear Safety, Volume 29.

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DOE OSTI2014

Investigation of Electrochemical Recovery of Zirconium from Spent Nuclear Fuels

This project uses both modeling and experimental studies to design optimal electrochemical technology methods for recovery of zirconium from used nuclear fuel rods for more effective waste management. The objectives are to provide a means of efficiently separating zirconium into metallic high-level waste forms and to support development of a process for decontamination of zircaloy hulls to enable their disposal as low- and intermediate-level waste. Modeling work includes extension of a 3D model previously developed by Seoul National University for uranium electrorefining by adding the ability to predict zirconium behavior. Experimental validation activities include tests for recovery of zirconium from molten salt solutions and aqueous tests using surrogate materials. *This is a summary of the FY 2013 progress for I-NERI project # 2010-001-K provided to the I-NERI office.

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DOE OSTI2011

SELECTIVE REMOVAL OF STRONTIUM AND CESIUM FROM SIMULATED WASTE SOLUTION WITH TITANATE ION-EXCHANGERS IN A FILTER CARTRIDGE CONFIGURATIONS-12092

Experimental results for the selective removal of strontium and cesium from simulated waste solutions with monosodium titanate (MST) and crystalline silicotitanate (CST) laden filter cartridges are presented. In these proof-of-principle tests, effective uptake of both Sr-85 and Cs-137 were observed using ion-exchangers in this filter cartridge configuration. At low salt simulant conditions, the instantaneous decontamination factor (D{sub F}) for Sr-85 with MST impregnated filter membrane cartridges measured 26, representing 96% Sr-85 removal efficiency. On the other hand, the Sr-85 instantaneous D{sub F} with co-sintered active MST cartridges measured 40 or 98% Sr-85 removal efficiency. Strontium-85 removal with the MST impregnated membrane cartridges and CST impregnated membrane cartridges, placed in series arrangement, produced an instantaneous decontamination factor of 41 compared to an instantaneous decontamination factor of 368 for strontium-85 with co-sintered active MST cartridges and co-sintered active CST cartridges placed in series. Overall, polyethylene co-sintered active titanates cartridges performed as well as titanate impregnated filter membrane cartridges in the uptake of strontium. At low ionic strength conditions, there was a significant uptake of Cs-137 with co-sintered CST cartridges. Tests results with CST impregnated membrane cartridges for Cs-137 decontamination are currently being re-evaluated. Based on these preliminary findings we conclude that incorporating MST and CST sorbents into membranes represent a promising method for the semi-continuous removal of radioisotopes of strontium and cesium from nuclear waste solutions.

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DOE OSTI2017

Potential Cost Implications of an Interim Storage Facility for Commercial SNF+ - 17132

The question of whether or not consolidated interim storage of commercial spent nuclear fuel (SNF) should be part of the federal waste management system as an intermediate step before permanent disposal has been debated for more than four decades. This paper summarizes an evaluation of the cost implications of incorporating a consolidated interim storage facility (ISF) into the waste management system (WMS). In this study, the order-of-magnitude estimates of total system costs were calculated and tabulated. The analyzed scenarios involve shipment of SNF from reactors in dual-purpose storage and transportation canisters (DPCs) currently being used by utilities for dry storage at reactor sites. A number of pertinent conclusions can be drawn from this evaluation, including: - Delay in repository availability increases total system costs. Any delay in opening a repository increases total system costs, regardless of whether the system has an ISF or not. This is due to the increased cost associated with an extended duration for storage, whether at multiple independent spent fuel storage installations (ISFSIs) or at an ISF, until waste can be disposed of.- There is a potential total system lifecycle cost avoidance in realistic scenarios with an ISF when compared to scenarios with no ISF. However, most of the cost avoidance occurs several decades after the ISF is opened. The total WMS cost differential over the long term is mainly attributed to the reduced operational costs of storing the fuel in a consolidated facility rather than at individual reactor sites. - Earlier establishment of an ISF allows for more avoidance of post-shutdown at-reactor storage costs for any repository opening date. An ISF allows earlier acceptance of fuel from reactors, which reduces at-reactor costs from a total system perspective. - Transportation costs have little impact on a WMS with or without an ISF. These impacts range from 3-11% of the total cost in all scenarios. Therefore, transporting the fuel twice does not appear to be a significant cost concern driver relative to other system costs. In conclusion, designing, licensing, and constructing an ISF will require a large near-term financial investment. However, reduced annual operating and maintenance costs for an integrated waste management system with an ISF, compared to a system without an ISF, could eventually result in system-wide (integrated) cost avoidance due to reductions in at-reactor storage costs in the long term. The largest cost avoidances would not occur for several decades, so assumptions about inflation, escalation, and discount rates will have a significant effect on potential economic impacts of an ISF. It should be noted that other benefits associated with the investment of an ISF-such as earlier removal of SNF from shutdown reactor sites, earlier prospects for site re-utilization, and additional overall integrated waste management system flexibility-were not addressed in this study. (authors)

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DOE OSTI2013

The Management of the Radioactive Waste Generated by Cernavoda NPP, Romania, an Example of International Cooperation - 13449

The design criteria and constraints for the development of the management strategy for radioactive waste generated from operating and decommissioning of CANDU Nuclear Units from Cernavoda NPP in Romania, present many specific aspects. The main characteristics of CANDU type waste are its high concentrations of tritium and radiocarbon. Also, the existing management strategy for radioactive waste at Cernavoda NPP provides no treatment or conditioning for radioactive waste disposal. These characteristics embodied a challenging effort, in order to select a proper strategy for radioactive waste management at present, when Romania is an EU member and a signatory country of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. The helping of advanced countries in radioactive waste management, directly or into the frame of the international organizations, like IAEA, become solve the aforementioned challenges at adequate level. (authors)

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