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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 OSTI2019

Ensuring Future Qualified Employees through Workforce Development Initiatives - 19183

UCOR, an AECOM-led partnership with Jacobs, manages the cleanup of the 2,200-acre East Tennessee Technology Park for its client, the U.S. Department of Energy (DOE), in Oak Ridge, Tennessee. The site was contaminated with radioactive, hazardous and industrial wastes generated by more than 40 years of national defense and energy missions. While successfully conducting its cleanup responsibilities, DoE's Oak Ridge Office of Environmental Management and UCOR are strongly committed to developing the current and future workforce through the application of multiple partnering opportunities and the use of innovative learning techniques that are adaptive to different generations of learners. The goal is to maintain a highly-skilled workforce for cleanup activities while establishing bridges for the development and placement of our future workforce. The training and professional development of the UCOR workforce is provided by multiple sources, including UCOR in-house training, trade unions, regional higher educational institutions and regional technical schools. In concert with these partners, UCOR is working to maintain and expand the existing talent base, enhance safety, improve processes, and resolve challenges through aggressive and well-coordinated initiatives designed to develop the future workforce. While UCOR's efforts are primarily focused on the cleanup of DOE facilities, its workforce development programs can also benefit the commercial nuclear power industry - both in ongoing plant operations and in decommissioning aging plants as they come to the end of their useful life. UCOR's Workforce Development Program is built on four distinct pillars: - Higher Education; - Community Colleges; - Apprenticeship Readiness Program; - K-12 STEM Program. (authors)

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2016

Synergetic Interactions between Uranium, Humic Acid, Silica Colloids and SRS Sediments at Variable pH - 16524

The Savannah River Site (SRS) was constructed during the 1950's and became one of the major producers of plutonium for the United States during the Cold War. During its production life, the F/H Area Seepage Basins received approximately 1.8 billion gallons of acidic waste containing radionuclides and dissolved heavy metals. This led to the creation of a highly contaminated groundwater plume with uranium (VI) as a key contaminant of concern within the plume. Now, as a designated hazardous waste management facility, the main activities taking place at the site are nuclear storage and remediation of contaminated soil and groundwater from radionuclides. Humic substances (HS) are major components of soil organic matter having the ability to influence migration behavior and fate of heavy metals and are being investigated for potential use in environmental remediation at SRS. Essentially, HS are polyfunctional organic macromolecules formed by the chemo-microbiological decomposition of biomass or dead organic matter. HS are able to interact with both metal ions and organic compounds based on solubility. Humic acid (HA) represents the fraction of HS soluble at pH greater than 3.5. HA carries a large number of functional groups, provides an important function in ion exchange and is a metal complexing ligand with a high complexation capacity. Silica is the term applied to solid forms with the stoichiometric composition of SiO{sub 2}. The colloidal silica used in this work is amorphous and non-porous in suspension; silica's net surface charge for pH values higher than 3 is negative, due to the isoelectric point being 1.5 - 2.5. Humic acid has been previously used in remediation techniques and silica is found naturally in the soil. This investigation studied the synergistic interactions between four key components: U(VI), humic acid (HA), colloidal silica and SRS sediment under varying pH conditions ranging from 3 to 8. Multi-component batch systems were constructed to effectively analyze the removal of U(VI). The batches were pH adjusted, left to equilibrate for a minimum of 48 hours on a shaker, and then centrifuged. A set of samples, filtered to remove any colloids in suspension, and a set of unfiltered samples were analyzed. For the samples at acidic pH, the uranyl cation is the dominant species; with increasing pH, the mononuclear and polynuclear hydrolyzed uranyl ions become dominant based on speciation modeling. In filtered samples, uranium removal is highest at low pH with the negatively charged moieties of HA readily binding to the uranyl cation. Percent removal of U(VI) steadily decreased with an increase in pH and then remained comparable at a neutral/alkaline pH with the change in uranyl speciation. The percentage of colloidal silica removal was observed to be approximately 80% for both sediment- and non-sediment-bearing samples. As the pH increased, the silica removal in the non-sediment set showed a slight decrease and was in the range of 70% - 75%. Uranium removal in the unfiltered samples showed similar trends as compared to the filtered samples; however, the removal of uranium in the unfiltered samples was slightly lower compared to the filtered samples. Colloidal silica removal in non-sediment-bearing batches showed a decline with increasing pH, from 90% to 50%; conversely, the silica removal remained constant for the sediment-bearing batches at approximately 60%. Colloidal silica does not seem to have a significant effect on the removal of U(VI) before filtration due to U(VI) bound to colloidal silica remaining in suspension. The presence of humic acid, silica colloids, and SRS sediment was shown to synergistically effect the removal of U(VI). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Panel Session 36: Panel: Young Professionals in Nuclear Science and Engineering an International Perspective

This panel focuses on the young professional views from around the world on radioactive waste management. An informal panel will enable these future leaders of our industry to share ideas and experiences and develop contacts within peer groups on a genuine international level. The informal and interactive panel format will stimulate valuable discussion among the participants and the audience. The panel will be led by young professionals who will actively encourage participation from all those attending. Mentoring will be discussed. This panel aims to encourage fresh thinking and provide an opportunity for an open and frank discussion of issues. Presentations proposed include: 1 - International Youth Nuclear Congress (Alexander Borovskis, AusYGN (Australia)); 2 - AECOM Early Career Professionals Technology Exchange (Natalie Young, Washington River Protection Solutions (USA)); 3 - Young Professionals in Nuclear Science and Engineering - An International Perspective (Shona Thompson, Canadian Nuclear Safety Commission (Canada))

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Destructive and Non-Destructive Characterization of various LLW/ILW at NRG - 19078

The Nuclear Research and consultancy Group is located at Petten, the Netherlands, and operates the High Flux Reactor (HFR). The HFR is used for isotope production for medical purposes and material research. During its exploitation several types of radioactive waste are produced, which partly have accumulated at the reactor site. In addition, NRG operated the Low Flux Reactor (LFR), a small 30 kWmax Argonaut type reactor which was used for research and education purposes. Over the past decade several projects on waste management and decommissioning were carried out requiring various approaches, including the radiochemical characterisation of the different emerging waste streams as required by the Dutch regulator. This paper focuses on the practices and technical aspects of the destructive and non-destructive characterization of various types of low- and intermediate-level waste arising from a number of nuclear activities performed at the premises of NRG, i.e. the Dutch legacy waste management program, the treatment of ion exchange resins, and the decommission of the Low-Flux Reactor. (authors)

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DOE OSTI2017

Conceptual Design for a Short-Lived Low- and Intermediate-Level Waste Repository at the Al Tuwaitha Nuclear Center, Baghdad, Iraq - 17189

The European Commission under their Instrument for Nuclear Safety Cooperation (INSC) has initiated a project supporting Iraq's efforts for the safe disposal of short lived low- and intermediate-level radioactive waste (SL-LILW) arising from the decommissioning of Iraqi nuclear installations. The INSC project includes both the development of an initial conceptual level basic design and preliminary safety assessment (PSAR) report and the subsequent detailed engineering design and supporting intermediate safety assessment report (ISAR). The project is being executed by a Consortium led by NUKEM Technologies Engineering Services GmbH and joined by DBE Technology GmbH of Germany; JAVYS a.s. of Slovakia and MCM Environmental Services Ltd of the United Kingdom. The current paper focuses on the development of the conceptual level basic design. In previous decades Iraq had a significant nuclear program which encompassed ten nuclear sites spread out across the country. The largest site was located at the Al-Tuwaitha site near Baghdad where 18 nuclear facilities were located. The facilities suffered significant damage from bombing and subsequent looting during the first Gulf War in 1991. Currently, the Ministry of Science and Technology (MoST) is the radioactive waste management operator in Iraq and also the license holder for the ten Iraqi nuclear sites, all of them heavily damaged. Because of the ongoing security concerns in Iraq MoST has selected the Al Tuwaitha Nuclear Center as the preferred and best suited location for development of an SL-LILW repository. As an initial step in development of the conceptual design various repository design options for the site were examined and assessed. The assessment considered similar facilities that have been implemented internationally. At the conclusion of the assessment and in conjunction with the Iraqi Beneficiaries from MoST, a near-surface vault design was selected for the facility. The design of the facility relies on experience gained from existing low-level waste repositories at Centre de l'Aube in France and the El Cabril solid radioactive waste disposal facility in Spain as well as on more recent experience gained in the design of the Bulgaria National Disposal Facility (NDF) under construction near the Kozloduy Nuclear Power Plant. Similar to these reference designs, the concept for the Al Tuwaitha Disposal Facility (ATDF) will rely on a series of reinforced concrete vaults into which containerised waste will be placed. Each vault will consist of two chambers designed specifically to hold the waste. Central drainage points in each waste chamber will lead to an infiltration water collection system designed to detect any failures in the waste containment. Each disposal vault will be sealed after filling by installation of a reinforced concrete closure slab. Once all of the waste has been disposed of the site will be decommissioned and a multilayer protective cover will be installed over the sealed disposal vaults. The multilayer cover will act to further isolate the waste from the environment by limiting the access of water to the waste. Due in part to the presence of shallow groundwater at the site and to avoid potential damage to the facility by flooding from the nearby Tigris River the concept for the Al Tuwaitha Disposal Facility requires construction of the disposal vaults on a raised platform several meters above ground level. Construction of this platform will be integrated into the conceptual safety design and materials will be selected which further enhance the isolation function of the disposal system by hindering or delaying the release of radionuclides to the environment. The infiltration water collection system, which will be constructed in a specifically designed accessible gallery embedded and fully integrated into the platform, will provide access to inspection points located beneath each disposal chamber. The ATDF when completed will be a further demonstration of the adaptability of the disposal vault concept for use in isolating SL-LILW from the environment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Radiological Verification for the Remediation of LLRW at the Port Granby Project - 19210

The Port Granby Project is part of the Port Hope Area Initiative (PHAI) that Canadian Nuclear Laboratory (CNL) is using to manage long-term historic low level radioactive waste (LLRW) resulting from 60 years of uranium and radium processing operations in the urban setting of Port Hope, Ontario. Port Granby remediation scope includes constructing a new, secure long-term waste management facility (LTWMF) and supporting infrastructure and excavating and relocating LLRW and marginally contaminated soil from the Port Granby Waste Management Facility (PGWMF) which accepted waste until the 1980's to the newly constructed LTWMF. The PGWMF contains an estimated 450,000 cubic meters of LLRW deposited in unmarked and unlined trenches throughout the southern portion of the site that borders Lake Ontario. Project requirements include excavating contaminated material, verifying that the trench bottoms are below action levels using Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) guidance, and final restoration. This paper describes challenges of meeting remediation verification requirements using traditional field and laboratory confirmation methods in areas of deep trench areas located in rugged terrain with ponded water and mud and presents a process that has expedited the traditional MARSSIM method for verification with greater productivity, and with enhanced safety and reliability as compared to the use of traditional methods alone. This enhanced process has been successfully employed for in-situ verification so that the excavated area can be backfilled and returned to a vegetative state. This enhanced process has been examined by federal regulatory staff through the course of their onsite audits and outcomes from these examinations have all been positive. The method involves a visual inspection and traditional walkover of the area to identify any areas with radiation measurements above the region's background followed by a full area survey performed using Wood's proprietary Orion ScanPlotSM system. The ScanPlotSM data is evaluated, mapped and overlaid on top of an aerial image of the surveyed area allowing remaining radiological material at surface or near surface to be quickly and easily identified for removal. Field staff then collect soil samples from evenly distributed points within the surveyed area and field screen the samples using an X-ray fluorescence (XRF) analysis for the three primary contaminates of potential concern (CoPCs) (arsenic, uranium and lead) that are indicative of the historical contamination. Samples passing XRF field screening are packaged and sent to a local laboratory for a full suite of analyses and checked against the CoPC list to assess whether the soil matches the profile for nearby native soils and passes the remediation goals. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

A Framework Approach to Evaluate Cross-Cultural Adaptation of Public Engagement Strategies for Radioactive Waste Management - 13430

The complex interplay of politics, economics and culture undermines attempts to define universal best practices for public engagement in the management of nuclear materials. In the international context, communicators must rely on careful adaptation and creative execution to make standard communication techniques succeed in their local communities. Nuclear professionals need an approach to assess and adapt culturally specific public engagement strategies to meet the demands of their particular political, economic and social structures. Using participant interviews and public sources, the Potomac Communications Group reviewed country-specific examples of nuclear-related communication efforts to provide insight into a proposed approach. The review considered a spectrum of cultural dimensions related to diversity, authority, conformity, proximity and time. Comparisons help to identify cross-cultural influences of various public engagement tactics and to inform a framework for communicators. While not prescriptive in its application, the framework offers a way for communicators to assess the salience of outreach tactics in specific situations. The approach can guide communicators to evaluate and tailor engagement strategies to achieve localized public outreach goals. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

GLASSES CONTAINING IRON (II III) OXIDES FOR IMMOBILIZATION OF RADIOACTIVE TECHNETIUM

Technetium-99 (Tc-99) has posed serious environmental threats as US Department of Energy's high-level waste. This work reports the vitrification of Re, as surrogate for Tc-99, by iron-borosilicate and iron-phosphate glasses, respectively. Iron-phosphate glasses can dissolve Re as high as {approx} 1.2 wt. %, which can become candidate waste forms for Tc-99 disposal, while borosilicate glasses can retain less than 0.1 wt. % of Re due to high melting temperature and long melting duration. Vitrification of Re as Tc-99's mimic was investigated using iron-borosilicate and iron-phosphate glasses. The retention of Re in borosilicate glasses was less than 0.1 wt. % and more than 99 wt. % of Re were volatilized due to high melting temperature and long melting duration. Because the retention of Re in iron-phosphate glasses is as high as 1.2 wt. % and the volatilization is reduced down to {approx}50 wt. %, iron-phosphate glasses can be one of the glass waste form candidates for Tc (or Re) disposal. The investigations of chemical durability and leaching test of iron-phosphate glasses containing Re are now underway to test the performance of the waste form.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Improvement of Discharge Regulation System for Liquid and Gaseous Radioactive Materials in Korea - 18050

In this study, we analyzed radioactive waste management practices and discharge regulations for liquid and gaseous radioactive materials generated from the nuclear power plants. Liquid and gaseous radioactive materials are discharged into the environment in a controlled and monitored state after proper treatment. In order to prevent radiological hazards due to the release of radioactive materials, the standards for effluent control limits and off-site dose constraints at the boundary of exclusion area shall be met. In December 2015, the Nuclear Safety Act was partially amended to add the 'Discharge Plan for Liquid and Gaseous Radioactive Materials' to the application form for the power reactor operating licenses. The Discharge Plan shall include the total amount of liquid and gaseous radioactive materials per site, period (annual, quarterly, etc.), and radionuclide group. Therefore, nuclear power plant operators are required to set in advance the total amount of liquid and gaseous radioactive materials for each radionuclide group per site, and keep them below the control limits during the operation of the facilities. Accordingly, it is expected that public trust regarding regulations on effluent discharge as well as effluent control itself can be further improved. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Waste Estimation from a Wide-Area Radiological Incident: The Impact of Geography and Urban Footprint - 16256

In the planning and preparedness stages for a response to a radiological incident, it is important to include waste management considerations when developing the decontamination/demolition/cleanup approach because waste management can be a driver for time and cost for return to normalcy. Waste management is also inextricably linked to the geography and urban footprint of the impacted area. The U.S. Environmental Protection Agency's (EPA's) Waste Estimation Support Tool (WEST) is a novel application based on the Federal Emergency Management Agency (FEMA) Hazus-MH software. WEST enables users to estimate the characteristics, amount, and residual radioactivity of waste generated from remediation and cleanup activities after a radiological incident, including incidents caused by radiological dispersal devices and improvised nuclear devices, as well as nuclear power plant accidents. These waste estimates are generated as a function of user-defined decontamination strategies and are specific to a geographic location. This paper will describe the recently released version 3 update to WEST, and how differences in geography and urban footprints can impact the waste that is generated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Area V: Waste Management and Subsurface Science (FY23 Q2)

The objective of Topic Area V is to evaluate the feasibility and safety of an intermediate-depth borehole for disposal of IAEC nuclear waste. The IAEC nuclear waste is currently stored in above ground and/or in near-surface facilities that provide short-term radiological worker safety but do not provide long-term, full security and safeguards protection for the nuclear material. Disposal of the IAEC nuclear waste several hundred meters below the surface in a sealed borehole would provide long-term security and safeguards due to its depth underground, as well as long-term radiological safety, due to its isolation in a stable geological formation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Dounreay (UK) Shaft and Silo Decommissioning - Remote Handling in an Extreme and Uncertain Environment - 18420

DSRL is a wholly-owned subsidiary of the Cavendish Dounreay Partnership Ltd, a consortium of Cavendish Nuclear, CH2M and AECOM. It is funded by the Nuclear Decommissioning Authority (NDA) to deliver the site closure programme. Today, Dounreay is a site of construction, demolition and waste management, all of it designed to return the site to an interim end state. Dounreay is recognised as the most complex nuclear site closure project in Europe. The experimental nature of many of the sites redundant facilities means the clean-up and demolition requires innovation as well as great care. It is driving innovation in the approach to remote handling and robotics in the highly complex, hazardous and uncertain environment, all within tight financial constraints by UK Government. Veolia Nuclear Solutions (VNS) is instrumental in delivery of the overall solutions for the shaft and silo decommissioning project working with other members of industry and directly for DSRL to deliver the overall concept design for waste treatment as well as specialised equipment to recover highly hazardous and uncertain materials in this unique project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

Compliance Monitoring Implementation Plan for 40 CFR §191.14(b), Assurance Requirement (Rev. 6)

This Compliance Monitoring Implementation Plan outlines monitoring activities conducted by the U.S. Department of Energy (DOE) at the Waste Isolation Pilot Plant (WIPP) to demonstrate compliance with the U.S. Environmental Protection Agency (EPA) disposal regulations at Title 40 Code of Federal Regulations (CFR) Part 191, "Environmental Protection Standards for the Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes," Subparts B and C; and the EPA criteria for certifying compliance at 40 CFR Part 194, "Criteria for the Certification and Recertification of the Waste Isolation Pilot Plant's Compliance with the Disposal Regulations," Certification Decision, Final Rule. This plan does not address | monitoring activities intended to demonstrate compliance with 40 CFR Part 191 Subpart A. WIPP is a mined repository designed for the permanent disposal of defense-related transuranic (TRU)waste located in the Chihuahuan Desert, 26 miles east of Carlsbad, New Mexico. The suitability of the WIPP site for TRU waste disposal is supported by more than three decades of environmental studies. Monitoring the WIPP facility is one of the DOE's top priorities. Monitoring activities are implemented in compliance with various federal and state of New Mexico regulatory and operational safety requirements. These activities are conducted to ensure environmental protection, public and worker health and safety, and proper characterization of the disposal system. Monitoring activities will continue at WIPP through the operational period and until well after closure of the facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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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