SEARCH · Engineering Papers

Results for “Fuel Reprocessing”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 451 records · Page 25

All sources NASA NTRS DOE OSTI

DOE OSTI2015

Use of natural resource damages settlement funds for watershed protection at the Fernald Preserve, Harrison, Ohio - 15309

The Fernald, Ohio, Site, also called the Fernald Preserve is a former uranium processing plant that underwent extensive remediation pursuant to CERCLA. The Fernald site is managed by the US DOE Office of Legacy Management. US DOE and the Ohio EPA settled a long-standing CERCLA natural resource damage claim against the site in November 2008. As part of the settlement, US DOE paid $13.75 million to compensate for natural resource injury. These settlement funds were appointed for restoration, replacement, or acquisition of equivalent natural resources at or near the Fernald Preserve. Following finalization of a Funds Utilization Plan in February 2010, the Fernald Natural Resource Trustees (composed of US DOE, Ohio EPA, and the US Department of the Interior and known collectively as the Trustees) used a portion of the settlement funds to initiate the Paddys Run Conservation Project. This effort is a means to secure development rights and conservation easements within the Paddys Run watershed. Paddys Run is a third-order stream located on the western side of the Fernald Preserve. The Trustees partnered with Three Valley Conservation Trust to obtain assistance with land acquisition and conservation easement management. Since Fall 2012, more than 1,093 hectares (2,700 acres) have been permanently protected via easement or purchase. The Trustees are also partnering with MetroParks of Butler County, Ohio, to establish a nature preserve on acquired land. The Salamander Run Preserve is situated on 51 hectares (125 acres) of forest and former agricultural land approximately 11 kilometers (7 miles) north of the Fernald site. The land was purchased by Three Valley Conservation Trust and is undergoing ecological restoration using settlement funds. Upon completion, the property will be donated to MetroParks of Butler County, Ohio.(authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Contribution Slides for External IAEA Conference Radiation Chemistry Review

Contribution slides for an invited talk (presented by Professor Mohamad Al-Sheikhly, University of Maryland ) at the International Atomic Energy Agency (IAEA)- Virtual Workshop on Radiation Technology for Industry and Environment. The title of the review talk is "Future Trends in Radiation Technology Applications in Advancing Science and Engineering".

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2023

Fault Detection in the Solvent Extraction Process with Non-traditional Sensors

Solvent extraction is used to separate metals or other complexes into two different immiscible liquids and is an essential component of the PUREX Process. Improvements to the solvent extraction process can directly contribute to an organization’s ability to ensure the purity and recovery of special nuclear materials from spent nuclear fuel. Such improvements can benefit nuclear reprocessing efforts, increase fuel reutilization, and limit the concentration of actinides in nuclear waste repositories.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2024

Investigating the Impacts of Direct Dissolution Conditions on the Radiolytic Longevity of Butyramide Extractants

Removing the nitric acid (HNO3) dissolution step in used nuclear fuel (UNF) reprocessing would reduce the volume of radioactive waste streams generated, thereby, improving process efficiency. A promising strategy for this is the direct dissolution of UNF that has been pretreated by voloxidation into an organic solvent composed of specialized extractants and diluent. However, removal of the aqueous HNO3 phase from the envisioned reprocessing system has the potential to drastically change the suite of radiation-induced processes occurring, and thus, alter the longevity of proposed reagents. Furthermore, the impacts of fission product and transuranic metal ion complexation on the aforementioned radiation-induced processes is poorly understood, and yet can cause significant changes in radiolytic longevity. To bridge these knowledge gaps and support the continued development of direct dissolution strategies, we present an investigation into the impacts of direct dissolution conditions on the gamma radiation-induced degradation of N,N-di-(2-ethylhexyl) butyramide (DEHBA) and N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) ligands—candidate replacements for tributyl phosphate—in pre-equilibrated n-dodecane solvent in the presence and absence of envisioned loading amounts of uranium.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2024

Investigating the impacts of used nuclear fuel direct dissolution on the radiolytic longevity of solvent and butyramide extractants

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2025

All-LiCl Flowsheet Development

Annual report on the LiCl Pyro project. The objectives of this project are to begin work to determine the feasibility of transitioning to LiCl-only electrorefining for reprocessing nuclear fuel.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Human capital needs - teaching, training and coordination for nuclear fuel cycle

Human capital is the accumulation of competencies, knowledge, social and creativity skills and personality attributes, which are necessary to perform work so as to produce economic value. In the frame of the nuclear fuel cycle, this is of paramount importance that the right human capital exists and in Europe this is fostered by a series of integrated or directed projects. The teaching, training and coordination will be discussed in the frame of University curricula with examples from several programs, like e.g. the Master of Nuclear Engineering at Chalmers University, Sweden and two FP7 EURATOM Projects: CINCH - a project for cooperation in nuclear chemistry - and ASGARD - a research project on advanced or novel nuclear fuels and their reprocessing issues for generation IV reactors. The integration of the university curricula in the market needs but also the anchoring in the research and future fuel cycles will be also discussed, with examples from the ASGARD project. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Dismantling of the 904 Cell at the HAO/Sud Facility - 13466

La Hague facility, in France, is the spent fuel recycling plant wherein a part of the fuel coming from some of the French, German, Belgian, Swiss, Dutch and Japanese nuclear reactors is reprocessed before being recycled in order to separate certain radioactive elements. The facility has been successively handled by the CEA (1962-1978), Cogema (1978-2006), and AREVA NC (since 2006). La Hague facility is composed of 3 production units: The UP2-400 production unit started to be operated in 1966 for the reprocessing of UNGG metal fuel. In 1976, following the dropout of the graphite-gas technology by EDF, an HAO workshop to reprocess the fuel from the light water reactors is affiliated and then stopped in 2003. - UP2-400 is partially stopped in 2002 and then definitely the 1 January 2004 and is being dismantled - UP2-800, with the same capacity than UP3, started to be operated in 1994 and is still in operation. And UP3 - UP3 was implemented in 1990 with an annual reprocessing capacity of 800 tons of fuel and is still in operation The combined licensed capacity of UP2-800 and UP3 is 1,700 tons of used fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

ORNL experience and perspectives related to processing of thorium and 233 U for nuclear fuel

Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233 U for recycle in thermal reactors, fast reactors, or externally driven systems. Here, lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233 U.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Non-Aqueous Electrochemical Fluorination of Used Nuclear Fuel as an Advanced Separation Process

Development of efficient and environmentally benign methods to reprocess used nuclear fuel (UNF) will enable technologies for a nuclear renaissance. Electrochemical and fluorination methods for reprocessing UNF have been proposed, but combinations of electrochemical and fluorination methods have not been investigated. Electrochemical fluorination may reduce waste volumes compared to the main large-scale aqueous methods for processing used nuclear fuel. A non-aqueous electrochemical fluorination reprocessing method has been developed and demonstrated that enables gas phase uranium recovery while allowing for potential control of the reaction using a single process. Thermodynamic modeling showed a galvanic reaction between UNF and a fluorinating agent, such as NF 3 , in a molten fluoride electrolyte was possible and could selectively fluorinate U to UF 6 . To verify the model results, a reactor system for electrochemical fluorination was constructed along with vessels for product collection and analysis. Initial trials of the electrochemical reaction were performed and characterized by several electrochemical methods. The electrochemical oxidation of U using NF 3 as an oxidizing agent was demonstrated and UF 6 as a reaction product was detected.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Neutronics and computational fluid dynamics analyses for a natural circulation loop of molten fuel salt

A conceptual reactor system of molten fuel salt loop designed for Mo-99 production was analyzed by coupling TRITON neutronics and Fluent computational fluid dynamics (CFD) simulations. The loop system operated in 1 MW thermal power features natural circulation and potential extraction of non-soluble fission products (FP) in fuel salt. The behavior of natural circulation was studied and the power and temperature distributions in equilibrium were obtained. A long-term fuel depletion analysis of the system considering fuel circulation and reprocessing was then performed using an approximate stepwise sequence of simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Waste Management Approach for a Potential Future U.S. Closed Fuel Cycle - 16233

The U.S. Department of Energy, Office of Nuclear Energy has chartered an effort to develop technologies to enable safe and cost effective recycle of commercial used nuclear fuel (UNF) in the U.S. Part of this effort includes the evaluation of exiting waste management technologies for effective treatment of wastes in the context of current U.S. regulations and development of waste forms and processes with significant cost and/or performance benefits over those existing. This study summarizes the results of these ongoing efforts with a focus on the highly radioactive primary waste streams. The primary streams considered and the recommended waste forms include: •Tritium separated from either a low volume gas stream or a high volume water stream. The recommended waste form is low-water cement in high integrity containers. •Iodine-129 separated from off-gas streams in aqueous processing. There are a range of potentially suitable waste forms. As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals. •Carbon-14 separated from LWR fuel treatment off-gases and immobilized as a CaCO3 in a cement waste form. •Krypton-85 separated from LWR and SFR fuel treatment off-gases and stored as a compressed gas. •An aqueous reprocessing high-level waste (HLW) raffinate waste which is immobilized by the vitrification process in one of three forms: a single phase borosilicate glass, a borosilicate based glass ceramic, or a multi-phased titanate ceramic [e.g., synthetic rock (Synroc)]. •An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel that is either included in the borosilicate HLW glass or is immobilized in the form of a metal alloy in the case of glass ceramics or titanate ceramics. •Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware that are washed and super-compacted for disposal or as an alternative Zr purification and reuse (or disposal as low-level waste, LLW) by reactive gas separations. •Electrochemical process salt HLW which is immobilized in a glass bonded Sodalite waste form known as the ceramic waste form (CWF). •Electrochemical process UDS and SS cladding hulls which are melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported.

Fuel Cycle↗

DOE OSTI2018

Plutonium: Advancing our Understanding to Support Sustainable Nuclear Fuel Cycles

With Global energy needs increasing, real energy solutions to meet demands now, are needed. Fossil fuels are not an ideal candidate to meet these needs because of their negative impact on the environment. Renewables such as wind and solar have huge potential, but still need major technological advancements (particularly in the area of battery storage) before they can effectively meet growing world needs. The best option for meeting large energy needs without a large carbon footprint is nuclear energy. Of course, nuclear energy can face a fair amount of opposition and concern. However, through modern engineering and science many of these concerns can now be addressed. Many safety concerns can be met by engineering advancements, but perhaps the biggest area of concern is what to do with the used nuclear fuel after it is removed from the reactor. Currently the United States (and several other countries) utilize an open fuel cycle, meaning fuel is only used once and then discarded. It should be noted that fuel coming out of a reactor has utilized approximately 1% of the total energy that could be produced by the uranium in the fuel rod. The answer here is to close the fuel cycle and recycle the nuclear materials. By reprocessing used nuclear fuel, all the U can be repurposed without requiring disposal. The various fission products can be removed and either discarded (hugely reduced waste volume) or more reasonably, utilized in specialty reactors to make more energy or needed research/medical isotopes. While reprocessing technology is currently advanced enough to meet energy needs, completing research to improve and better understand these techniques is still needed. Better understanding behavior of fission products is one area of important research. Despite it being discovered over 75 years ago, plutonium is still an exciting element to study because of the complex solution chemistry it exhibits. In aqueous solutions Pu can exist simultaneously in multiple oxidation states, including 3+, 4+, and 6+. It also readily forms a variety of metal-ligand complexes depending on solution pH and available ligands. Understanding of the behavior of Pu in solution remains an important area of research today, with relevance to developing sustainable nuclear fuel cycles, minimizing its impact on the environment, and detecting and preventing the spread of nuclear weapons technology.

Lines, Amanda M.↗

DOE OSTI2018

Regulating Operational Safety at a Future Geological Disposal Facility in the UK - 18038

The United Kingdom (UK) has generated commercial nuclear energy for over 60 years; generating a significant and varied inventory of higher activity radioactive waste (HAW). In accordance with international practice, UK Government policy for long-term management of HAW is disposal in a geological disposal facility (GDF). The nuclear materials comprising HAW include high level waste, intermediate level waste and some low level waste, not suitable for disposal to the Low Level Waste Repository. In addition, other nuclear materials may be reclassified as waste in the future and will require to be managed as HAW for disposal to a GDF; these include spent fuel, uranium and plutonium. The operator of a GDF will be required to obtain a nuclear site licence from the Office for Nuclear Regulation (ONR), as the nuclear safety and security regulator, in order to construct and operate the facility. In the period prior to submission of a licence application, ONR is working to develop its regulatory expectations and provide advice on its regulatory role with regards to a future GDF. ONR has developed regulatory guidance on operational safety, benchmarked against the relevant IAEA Safety Standards and the Western European Nuclear Regulators Association (WENRA) Safety Reference Levels for Disposal. This guidance draws upon ONR's current expectations for nuclear sites; including operational reactors, fuel manufacture and reprocessing facilities, and reactors undergoing decommissioning; and incorporates relevant good practice drawn from international programmes for geological disposal and related industries (eg mining). In addition to the technical assessment guidance, ONR has developed guidance on the licensing process specific to a GDF; providing clarity for example on when a licence will be required. A review of the standard licence conditions applied to all nuclear sites concluded no significant amendment would be required for application to a GDF. Through its voluntary agreement with Radioactive Waste Management Limited (RWM), the developer and future operator of a GDF in the UK, ONR exerts regulatory influence on the design and organisational development of the future licensee. As part of this scrutiny, ONR has assessed RWM's generic Disposal System Safety Case. Subjecting the generic design and safety case to regulatory review provides opportunity for early engagement to ensure the appropriate standards will be achieved. ONR will only grant the necessary regulatory permissions if RWM adequately demonstrates the GDF will be constructed, operated and closed safely. Future challenges faced by ONR with respect to a GDF include building and maintaining regulatory competence and confidence, and engaging effectively with the public as the UK begins its search for a suitable site to provide appropriate assurance that a GDF will be operated safely and securely. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and {sup 238}U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with {sup 238}U alone should be reprocessed in the usual way.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Mu*STAR ADSR Fuel Conversion Facility Evaluation and Cost Analysis

The purpose of the reported effort is the development of a conceptual process for conversion of used nuclear fuel (UNF) from light water reactors (LWRs) to a fluoride salt form to facilitate its use for power generation in a Mu*STAR Accelerator-driven Subcritical Reactor (ADSR). This document evaluates potential processes for converting UNF into a fluoride salt, evaluates the costs/benefits of using enriched lithium in the carrier salt, and analyzes comparable costs of similar facilities to make a rough order of magnitude estimate for the cost to design, construct, and commission a Mu*STAR ADSR fuel conversion facility. This document also makes a high-level comparison of the Mu*STAR ADSR fuel cycle with other important fuel cycles. The results of the cost analysis show that the Mu*STAR ADSR fuel conversion facility will cost more than $100M, possibly up to around $1B, to design and construct. The cost will depend on the size of the hot cell required, which will depend on the processing options chosen for converting the UNF to a molten salt. Operations and maintenance of the facility will be on the order of $10M per year. The design, fabrication, and installation of the remote process equipment to support the fuel conversion will be an additional cost of more than $50M. Cost surveys of Department of Energy hot cell designs and previous Department of Energy hot cell construction suggest that the cost variance may be large, and the risk of escalating costs may be high in the regulation-mandated atmosphere of nuclear facility design and construction. The molten salt for the Mu*STAR ADSR contains all of the nonvolatile isotopes in the UNF, and there is never a separation of the plutonium from the fission products. This makes this conversion process much more proliferation resistant than the reprocessing options used for other closed fuel cycles. When normalized to the amount of electricity produced, the Mu*STAR ADSR fuel cycle’s expected cost might be less than any reprocessing fuel cycle and possibly less than the once-through cycle now used. Simulations show that once the conversion to fluoride is made, that up to seven times as much energy can be extracted from the UNF than was produced by the LWR while making it [Bowman, 2010]. Using once-only, on-site conversion to fluoride means no mining, no enrichment, no manufacturing of fuel rods, and no transportation costs. Typically, the UNF was produced over a 40-year period in an LWR; it can potentially be used for 280 years in Mu*STAR, producing the same power as the LWR did, ultimately requiring disposal of approximately the same volume of material as the LWR generated. Mu*STAR burns most of the higher actinides, which could reduce both the cost of disposal and significantly reduce the longevity of its radiotoxicity. Mu*STAR ADSR’s ability to demonstrate unique and valuable peripheral benefits are additional arguments for this technology, including (1) turning UNF into a valuable commodity (thus, its disposal will not be such an immediate problem, so continued operation of existing LWRs will be more acceptable); (2) burning fissile or fertile fuel including excess weapons-grade plutonium in subcritical mode; and (3) providing high-temperature process heat to convert natural gas to synthetic diesel fuel, water to hydrogen, or for other process-heat applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Recycle of Am-241 obtained from long term stored plutonium for use in radioisotope power systems - 15035

Pu-238 has been used as a power source for spacecraft since the early days of space exploration. It has proven to be an effective source of power where the use of solar generated power is impractical. Its application for use in spacecraft power systems was developed by the USA and Russia who have spent many $ 100s of millions on its development and production. However, the original facilities for the production of Pu-238 are no longer in use and existing worldwide stockpiles have reduced to levels sufficient for only one or two further missions. Production of Pu-238 requires considerable facilities including a nuclear reactor and reprocessing plants that are very expensive to build and operate. The European Space Agency (ESA) has assessed the options for post-launch power generation in future European space missions and has made the decision to pursue the use of Am-241 as an alternative isotope to power future European Radioisotope Power Systems (RPS). What makes Am-241 attractive for the European market is its 100% isotopically pure production from the decay of Pu-241 in separated civil plutonium stockpiles. Historically, Europe has relied on collaborations with the USA or Russia to access these nuclear power sources. During 2009, the European Space Agency (ESA) funded a project to examine the cost and practicality of establishing a European source of material suitable for Radioisotope Power Systems (RPS). The study concluded that a small number of radioisotopes could be used in RPSs. On further review two leading candidates, Am-241 and Pu-238, were selected for further study. Whilst Pu-238 was recognized as the isotope of choice for use in RPSs, the costs of setting up a European based production facility was prohibitively expensive. Although Am-241 has a number of limitations compared to Pu-238, the much reduced production costs make Am-241 an attractive option for European based production of RPSs. Despite its lower power density of ∼ 0.11 Wth/g, Am-241 is viewed as a potential alternative to Pu-238 because of its availability within the nuclear fuel cycle. During the reprocessing of spent nuclear fuel plutonium is separated as a chemically pure PuO{sub 2} product, containing around 3-11% w/w of the Pu-241 isotope, and placed into storage. During storage the Pu-241 isotope beta decays to Am-241 with a half-life of 14.4 years. {sub 94}{sup 241}Pu → (β 14.4 y) {sub 95}{sup 241}Am. This process has resulted in many 100 kgs of americium growing into the civil plutonium stockpile in the UK; material that is of value in generating heat sources and is able to be recycled. The key benefit of Am-241 as a potential European RPS source isotope is the availability of feed material along with the facilities to process it. It is expected that aged civil PuO{sub 2} will contain 25-50 g Am-241 per kg PuO{sub 2} depending on the source. A conceptual flowsheet using facilities for separating americium from plutonium is expected to produce on average 8.4 kg Am/year. ESA are now following their published road-map of studies to develop flight ready RTGs and RHUs. This has included the development and testing of the flowsheet to be used for production of Am-241. The project has focused on underpinning the conceptual flowsheet, using aged plutonium in NNL's PuMA laboratory. This has underpinned the design of a plant to produce a steady supply of Am-241. The scope of the overall project is 'store to store' - the starting point is aged plutonium currently housed in existing Sellafield stores and the end point is the storage of separated plutonium and americium powders on the Sellafield site. The full scale process is planned to be housed within the NNL's Central Laboratory in an existing facility designed for plutonium active operations. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

The Environmental Protection Agency's Safety Standards for Disposal of Spent Nuclear Fuel: Potential Path Forward in Response to the Report of the Blue Ribbon Commission on America's Nuclear Future - 13388

Following the decision to withdraw the Yucca Mountain license application, the Department of Energy created a Blue Ribbon Commission (BRC) on America's Nuclear Future, tasked with recommending a national strategy to manage the back end of the nuclear fuel cycle. The BRC issued its final report in January 2012, with recommendations covering transportation, storage and disposal of spent nuclear fuel (SNF); potential reprocessing; and supporting institutional measures. The BRC recommendations on disposal of SNF and high-level waste (HLW) are relevant to the U.S. Environmental Protection Agency (EPA), which shares regulatory responsibility with the Nuclear Regulatory Commission (NRC): EPA issues 'generally applicable' performance standards for disposal repositories, which are then implemented in licensing. For disposal, the BRC endorses developing one or more geological repositories, with siting based on an approach that is adaptive, staged and consent-based. The BRC recommends that EPA and NRC work cooperatively to issue generic disposal standards-applying equally to all sites-early in any siting process. EPA previously issued generic disposal standards that apply to all sites other than Yucca Mountain. However, the BRC concluded that the existing regulations should be revisited and revised. The BRC proposes a number of general principles to guide the development of future regulations. EPA continues to review the BRC report and to assess the implications for Agency action, including potential regulatory issues and considerations if EPA develops new or revised generic disposal standards. This review also involves preparatory activities to define potential process and public engagement approaches. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗