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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.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Expanded Validation of Uranium Systems with the KENO Monte Carlo Codes and SCALE 6.2.4 [Slides]

There is likely a trend in bias for uranium-fueled systems as a function of spectrum, but probably not as a function of enrichment. Some enrichment trends appear, but are potentially the result of outlier experimental results; further investigation is needed where possible. There are good examples in this data of markedly different biases from different systems. Future work includes review and vetting for inclusion in the VALID library and analysis with TSUNAMI.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Multi-group Examination of Nickel-Reflected HEU System [Slides]

Researchers noted an unusually large bias for 8-in. nickel reflected HEU sphere in HMF-003 between the 252- group library and the CE library in SCALE 6.2.4. The bias was investigated by reviewing reactions that $k_{eff}$ is sensitive to using TSUNAMI and collecting reaction rate data using tools within SCALE. For this system, bias is primarily due to the elastic scattering in nickel. Researchers compared new multigroup structures in SCALE 6.3. For this system, reduction in CE-to-MG bias seen in the 302-group and further improved in the 1597-group structure. When researchers compared ENDF/B-VIII.0 in SCALE 6.3, library showed improved results with new nickel evaluation. However, the MG-to-CE bias can still be large as large biases can occur in any system. This example highlights the importance of validating results with measured systems.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Planetary Defense: Hazards Modeling [Slides]

Hydrocodes were quickly adopted for planetary impact simulations and are now being combined with other codes (e.g. tsunami, atmospheric transport, general circulation, etc.) for hazard assessment and quantification.

58 GEOSCIENCES↗
DOE OSTI2024

RACEE: Port Lions Community Energy Efficiency Project

The City of Port Lions is one of several villages located on Kodiak Island in the Gulf of Alaska. Founded in 1964, Port Lions was built through the joint efforts of the Bureau of Indian Affairs, Lions International, the Public Health Service, and other organizations for the displaced residents of Afognak which was destroyed by a tsunami caused by the Good Friday Earthquake.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2025

Design of Critical Experiments Involving Graphite elements at the IPEN/MB-01 Plate-Type Core

A series of critical experiments were designed involving the use of characterized graphite elements in the new plate-type core of the IPEN/MB-01 reactor. Eleven configurations with different graphite quantities are considered, each producing distinct effects on neutron fluxes and offering opportunities to analyze varying sensitivities to k eff in the system. The analysis shows that the experiments can be performed with acceptably low k eff uncertainties considering an adequate characterization of the graphite elements to be used. The sensitivity results obtained using SCALE TSUNAMI calculations show that the experiments have a notable k eff sensitivity coefficient to graphite. In the configurations tested, the k eff sensitivity to graphite TSL is low. This experiment will be useful for criticality safety validation of applications using uranium fuel enriched to around 19.75 wt % 235 U, light water and graphite. It will also be useful to gain more insight into graphite material properties and their effect on k eff as a result of using the well-characterized graphite elements. The execution of the experiments is planned for the summer of 2025. The 11 designed configurations serve as a basis for the final design, and fewer configurations will be executed. Once executed, these critical experiments will be evaluated and submitted for publication in the International Criticality Safety Benchmark Evaluation Project Handbook, supporting the DOE/NRC Collaboration for Criticality Safety Support for Commercial-Scale HALEU for Fuel Cycles and Transportation project goal.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Microreactor Assembly Transportation Cask Model Description for Criticality Safety Validation Basis Assessment

Criticality safety analyses are completed on a transportation cask used for microreactor assembly shipment to provide an example of model and analysis to industry for reproducing this type of study on their microreactor fuel shipment. The fuel assembly considered is based on a gas-cooled microreactor (GC-MR), which utilizes HALEU fuel in the form of TRISO particles and utilizes various design options considered in industry designs. Various versions of this GC-MR assembly were studied, with and without YH2 moderator, providing similar conclusions. The shipment cask design is revised based on an existing design ES-3100, developed by Y-12 for the transport of highly enriched uranium (HEU), but is enlarged to hold the GC-MR fuel assembly. Criticality safety analysis for the cask/GC-MR fuel assembly package was performed using the CSAS6 sequence of SCALE6.3.2, utilizing the ENDF/B-VII.1 based continuous energy neutron library, and the analysis strictly follows the guideline from NRC reference reports. Different scenarios, e.g. normal operation, undamaged cask with water flooded, damaged cask with optimal water moderation, have been analyzed and it could be concluded the package would always have a large margin of subcriticality even packed in an infinite array. Sensitivity and similarity analyses are also performed using the TSUNAMI sequence of SCALE6.3.2, and the similarity analysis uses all the experiments from the ICSBEP Handbook with Intermediate and Mixed Enriched Uranium (IEU) and Low Enriched Uranium (LEU) systems together with additional ones that are sponsored by the DNCSH program. These similarity analyses indicate that dry cases have no similar benchmark experiments (ck values greater than 0.8), which may become problematic if more assemblies are shipped together (or a fully loaded core is shipped) and margin to criticality is reduced. However, the damaged cask models with flooded assemblies exhibited similarities to many experiments with ck values greater than 0.8.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Microreactor Core Transportation Cask Model Description for Criticality Safety Validation Basis Assessment (Rev. 1)

Criticality safety analyses are completed on transportation casks used for microreactor whole core shipment to provide examples of models and analyses to industry, regulators, and nuclear community at large to be used in verification and validation analyses of similar applications. The microreactors considered are based on a Gas-Cooled Microreactor (GCMR) and a Heat-Pipe Microreactor (HPMR), both utilize HALEU fuel in the form of TRISO particles and various other design options considered in industry microreactor designs. Variant design options of GCMR and HPMR were also investigated to provide a wider application range for each technology. Criticality safety analyses for the GCMR and HPMR packages were performed using the CSAS6 sequence of SCALE 6.3.2 with the ENDF/B-VII.1-based continuous energy neutron libraries. Different scenarios were investigated, including normal operation and water flooded conditions to represent nominal and hypothetical accident scenarios. Sensitivity and similarity analyses are also performed using the TSUNAMI sequence of SCALE 6.3.2, and the similarity analysis uses all the experiments from the ICSBEP Handbook with Intermediate and mixed Enriched Uranium (IEU) and Low Enriched Uranium (LEU) systems, together with additional experiments that are sponsored by the DNCSH program. Many experiments were found marginally similar to the GCMR and HPMR models, with similarity index (ck) values greater than 0.8 but less than 0.9. Among all the experiments analyzed, no case has a ck value greater than 0.9, indicating that additional critical experiments might be needed to further validate the criticality safety models for microreactors transport packages.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

U.S. Efforts in Support of Examinations at Fukushima Daiichi - November 2025 Meeting Notes

Fifteen years have passed since the March 11, 2011, the Great East Japan Earthquake (GEJE) triggered a tsunami that inundated the Fukushima Daiichi Nuclear Power Station (Daiichi or 1F) and led to severe accidents in multiple units at the plant. During these fifteen years since these events, Daiichi forensics examinations and evaluations have revealed many insights for enhancing the safety of nuclear power plants worldwide. However, uncertainties remain regarding the events after the GEJE. Additional Daiichi forensics information is required to inform future Decontamination and Decommissioning (D&D) activi- ties conducted by the plant owner, Tokyo Electric Power Company Holdings, Incorporated (TEPCO Hold- ings). This forensics information also has important implications for the safety and operation of U.S. commercial nuclear power plants. A collaborative U.S. and Japanese effort, which includes approximately 50 experts in reactor safety and plant operations, was initiated in 2014 by the Department of Energy Office of Nuclear Energy (U.S. DOE-NE) to identify Daiichi examination needs and evaluate recent Daiichi examination data to address these needs. This collaborative effort, which is beneficial to the U.S. and Japan, continues as Japan moves forward with D&D activities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Makah Tribe Strategic Energy Plan

The U.S. Department of Energy’s (DOE) Energy Transitions Initiative Partnership Project (ETIPP) connects remote and island communities, regional partners, and the DOE national laboratories to support communities as they seek to build resilience in their energy systems. The Makah Tribe faces several energy challenges, including frequent power outages and the potential for an extended outage due to an earthquake or tsunami. The Tribe joined ETIPP in 2022 to address those challenges, seeking to build energy resilience and sovereignty in the community. The Makah Tribe, Spark Northwest, the Pacific Northwest National Laboratory (PNNL), and the National Renewable Energy Laboratory (NREL) collaborated to develop a strategic energy plan as part of the second cohort of ETIPP communities. The long-term energy vision of the Makah Tribe includes increasing energy efficiency in the community, improving energy management capacity, and developing the renewable energy generation and storage sufficient to independently power the Reservation for one year. Additionally, the ETIPP team worked with Makah leadership, staff, and community members to identify a set of community priorities, values, and goals to guide energy development as the Tribe takes the incremental steps toward their vision for energy sovereignty. Those energy values include ecosystem-based management, energy sovereignty and project ownership, workforce development and capacity, economic opportunity, community wellbeing and priorities, and emergency disaster resilience. To understand what would be needed for a year for energy independence, the PNNL team conducted an assessment to determine the current energy usage of the Tribe and also modeled several scenarios for future energy use. Using the energy usage values, the team estimated how two types of renewable energy technology, specifically locally deployed solar and small-scale wind, could contribute towards the energy independence goal. The energy baseline and resource assessment produced the following key findings:

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2026

Microreactor Core Transportation Cask Model Description for Criticality Safety Validation Basis Assessment (Rev. 2)

Criticality safety analyses are completed on transportation casks used for microreactor whole core shipment to provide examples of models and analyses to industry, regulators, and nuclear community at large to be used in verification and validation analyses of similar applications. The microreactors considered are based on a Gas-Cooled Microreactor (GCMR) and a Heat-Pipe Microreactor (HPMR), both utilize HALEU fuel in the form of TRISO particles and various other design options considered in industry microreactor designs. Variant design options of GCMR and HPMR were also investigated to provide a wider application range for each technology. Criticality safety analyses for the GCMR and HPMR packages were performed using the CSAS6 sequence of SCALE 6.3.2 with the ENDF/B-VII.1-based continuous energy neutron libraries. Different scenarios were investigated, including normal operation and water flooded conditions to represent nominal and hypothetical accident scenarios. Sensitivity and similarity analyses are also performed using the TSUNAMI sequence of SCALE 6.3.2, and the similarity analysis uses all the experiments from the ICSBEP Handbook with High Enriched Uranium (HEU), Intermediate and mixed Enriched Uranium (IEU) and Low Enriched Uranium (LEU) systems. Many ICSBEP experiments were found marginally similar, with similarity index (ck) values greater than 0.8, to the GCMR and HPMR cask models, especially in flooded conditions that may be more constraining due to reduced reactivity margins. Some experiments using TRISO fuel particles and graphite moderator, including the recent THETA and Deimos experiments from LANL, together with some IRPhEP experiments, were also used for the similarity analyses. These TRISO-fueled and graphite moderated experiments behaved similarly to the GCMR models, with maximum ck value greater than 0.9, but they are less similar to the HPMR models. This observation indicates that additional critical experiments might be needed to further validate the criticality safety models, especially for heat-pipe based microreactor transport packages and for non-flooded transportation configurations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Microreactor Assembly Transportation Cask Model Description for Criticality Safety Validation Basis Assessment (Rev. 3)

Criticality safety analyses are completed on a transportation cask used for microreactor assembly shipment to provide an example of model and analysis to industry for reproducing this type of study on their microreactor fuel shipment. The fuel assembly considered is based on a gas-cooled microreactor (GC-MR), which utilizes HALEU fuel in the form of TRISO particles and utilizes various design options considered in industry designs. Various versions of this GC-MR assembly were studied, with and without YH 2 moderator, providing similar conclusions. The shipment cask design is revised based on an existing ES-3100 design, developed by Y-12 for the transport of highly enriched uranium (HEU), but is enlarged to hold the GC-MR fuel assembly. Criticality safety analysis for the cask/GC-MR fuel assembly package was performed using the CSAS6 sequence of SCALE6.3.2, utilizing the ENDF/B-VII.1 based continuous energy neutron library, and the analysis strictly follows the guideline from NRC reference reports. Different scenarios, e.g. normal operation, undamaged cask with water flooded, damaged cask with optimal water moderation, have been analyzed and it could be concluded the package would always have a large margin of subcriticality even packed in an infinite array. Sensitivity and similarity analyses are also performed using the TSUNAMI sequence of SCALE6.3.2, and the similarity analysis uses all the experiments from the ICSBEP Handbook with Highly Enriched Uranium (HEU), Intermediate and Mixed Enriched Uranium (IEU) and Low Enriched Uranium (LEU) systems, together with additional ones that are sponsored by the DNCSH program, and selected IRPhEP experiments using TRISO fuel and graphite moderator. These similarity analyses indicate that the dry nominal design has no similar benchmark experiments (ck values greater than 0.8), which may become problematic if more assemblies are shipped together (or a fully loaded core is shipped) and margin to criticality is reduced. However, the cask models with flooded assemblies exhibited similarities to many experiments with c k values greater than 0.8.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Demonstration of Optimal Benchmark Selection Website and Validation of the q c Coverage Metric Using HEU-SOL-THERM-013-003 Experiment

In the work documented in this interim report, the experiment selection toolkit web site was demonstrated and q C coverage metric methodology was validated for IEU-MET-FAST-002-001, MIX-COMP-THERM 004-004, and HEU-SOL-THERM-013-003 experiments. 𝑞 𝐶 is an information-theoretic measure based on mutual information that quantifies the ability of candidate benchmark experiments to reduce the bias and uncertainty of a target criticality safety application. The metric and an accompanying open-source Python toolkit with a web-based interface were tested against a benchmark set of 425 experiments drawn from the International Criticality Safety Benchmark Evaluation Project Handbook. The interface is hosted at https://edim.covdef.com. It accepts sensitivity data files produced by the TSUNAMI-IP module of the SCALE code system and supports both (i) deterministic analysis using the ENDF/B-VII.0 covariance library and (ii) stochastic analysis based on user-supplied keff samples. Demonstrations on representative applications across a range of material composition, spectrum, and form show that q C -guided benchmark selection achieves greater uncertainty reduction with fewer experiments and yields more stable posterior bias and uncertainty estimates than traditional similarity coefficient ( c k )–based selection, while also capturing valuable low-ck experiments that one-to-one metrics overlook.

Abdel-khalik, Hany S. [Indiana Univ.-Purdue Univ. ↗
DOE OSTI2022

Block-structured, equal-workload, multi-grid-nesting interface for the Boussinesq wave model FUNWAVE-TVD (Total Variation Diminishing)

Abstract. We describe the development of a block-structured, equal-CPU-load (central processing unit), multi-grid-nesting interface for the Boussinesq wave model FUNWAVE-TVD (Fully Nonlinear Boussinesq Wave Model with Total Variation Diminishing Solver). The new model framework does not interfere with the core solver, and thus the core program, FUNWAVE-TVD, is still a standalone model used for a single grid. The nesting interface manages the time sequencing and two-way nesting processes between the parent grid and child grid with grid refinement in a hierarchical manner. Workload balance in the MPI-based (message passing interface) parallelization is handled by an equal-load scheme. A strategy of shared array allocation is applied for data management that allows for a large number of nested grids without creating additional memory allocations. Four model tests are conducted to verify the nesting algorithm with assessments of model accuracy and the robustness in the application in modeling transoceanic tsunamis and coastal effects.

Choi, Young-Kwang↗
DOE OSTI2023

Observations of extreme wave runup events on the US Pacific Northwest coast

Extreme, tsunami-like wave runup events in the absence of earthquakes or landslides have been attributed to trapped waves over shallow bathymetry, long waves created by atmospheric disturbances, and long waves generated by abrupt breaking. These runup events are associated with inland excursions of hundreds of meters and periods of minutes. While the theory of radiation stress implies that nearshore energy transfer from the carrier waves to the infragravity waves can also lead to very large runup, there have not been observations of runup events induced by this process with magnitudes and periods comparable to the other three mechanisms. This work presents observations of several runup events in the US Pacific Northwest that are comparable to extreme runup events related to the other three mechanisms. It also discusses possible generation mechanisms and shows that energy transfer from carrier waves to bound infragravity waves is a plausible generation mechanism. In addition, a method to predict and forecast extreme runup events with similar characteristics is presented.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Panel Session 4 and 15: Japan Fukushima Daiichi D and D Update and Technological Challenges at Japan Fukushima Daiichi D and D and Update on Nuclear Overview and Development in Japan - Nuclear Fuel Cycle

Sessions 4 and 15 represent a two-part panel series that discusses progress and challenges associated with cleanup at Fukushima. Severe limitations on availability of original panelists from Japan due to strict restrictions put in place to alleviate the spread of coronavirus necessitated changes to both panels. The result was a significantly modified panel for Session 04 (shown above) and the elimination of all panelists for Session 15. The 4 and 15 Panel Sessions provide an overview of activities related to both the progress and challenges of cleanup and decommissioning of the Fukushima Daiichi Nuclear Power Station (NPS) in Japan. Five panelists discussed perspectives of the cleanup following a Tokyo Electric Power Company (TEPCO) video showing the progress on site since the devastating Great East Earthquake and tsunami that caused the explosions at three of the six reactors on the site. Three of the five panelists discussed on-going work being performed for the effort, while the other two provided expert perspectives of on strategic efforts at the site. The panel was attended by over 80 technologists and policy makers spanning the globe and was opened by Dr. Monica Regalbuto of Idaho National Laboratory and a short video provided by TEPCO. The video described changes at the site that spanned the cleanup efforts from stabilizing water intrusion into the contaminated reactor buildings to construction of new administrative facilities. The video explained the processes underway to retrieve spent fuel rods and challenges in retrieval of the compromised fuel debris. The video highlighted working condition improvements that included establishment of rest housing and a small convenience store on the site, and the rollback of protective equipment around the site due to decreases contamination. Panelists with presentations: Revision of 'the Mid-and-Long-Term Road-map towards the Decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station' (Paul Dickman); Sharing UK experience at Fukushima Daiichi (Adrian Simper); SRNL Japan (Andrew Fellinger); ABLE's Initiative to Dismantle the Exhaust Stack (Daniel Walter); JAEA R and D in Fukushima (Tokio Fukahori); TEPCO - Overview and Update of the Fukushima Decommissioning Process (Monica Regalbuto); Toshiba's Involvement in the Decommissioning of the Fukushima Daiichi Nuclear Power Plant (Yasuhiro Yuguchi); Remote Dismantling of the Exhaust Stack At Fukushima Dai-ichi NPS (Takashi Okutsu)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

The Intern Professional Development Center: A Model for Improving Student Engagement Within Industry - 20006

Many innovations and technical improvements have been made over the last half-century regarding our understanding of nuclear materials, with some of the brightest minds in the world collectively working towards a cleaner, brighter, healthier future. However, there is a demographical oddity with the enormous technical workforce that helped shape this new world: they are aging, collectively. Known as the 'silver tsunami,' mass retirements are in the near future for many technical industries, something that the industrial community is not entirely prepared to accommodate. There is an urgent need for a tech-savvy generation to replace the baby boomers, and there must be an uninterrupted hand-off from one generation to the next of institutional knowledge and best practices. Further, it is not enough to simply replace every retiring engineer with a younger one in a 'cold hand-off' - this information transfer takes time. Many educational communities, such as those in the Tri-Cities, Washington, are rising to the challenge, with STEM-focused programs, introducing students to technical concepts and potential career opportunities at an early age. However, while individually, academic institutions search for internships and industry has open positions for interested students, no single clearinghouse exists in the Mid-Columbia region to match talent with opportunities, regardless of the sending or receiving organizations. To fill this void, AYB Drafting (AYB) has established a career development and personal growth center, the Intern Professional Development Center (IPDC). The IPDC, located in Richland, Washington, addresses the problem of what has been referred to as the 'Graying of America,' as it bridges the gap between education and career for the up and coming workforce, creating valuable efficiencies for the young workforce and the industry alike. The operating principles and details of the initial IPDC launch are discussed in the following text, as well as the apparent success and overwhelming positive feedback received in just the first year of operation. (authors)

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2020

Investing in STEM Benefits Community, Company, and Industry - 20243

Like many organizations and groups, the Department of Energy Environmental Management program and the nuclear industry are facing an aging employee population and record numbers of employee retirements. This 'silver tsunami' of retirees means the industry as a whole is recruiting graduates at a fast pace. So, how do we ensure the nuclear industry has a robust pipeline of graduates with the necessary background? Fluor Idaho's investment in science, technology, engineering, and mathematics (STEM) education programs helps promote the next generation of technical professionals, benefiting both the company and the industry. Through community outreach initiatives and student mentoring programs, Fluor Idaho promotes kids' interest in STEM. The programs are designed to spark STEM interest in grade school and then foster and grow that interest through high school and higher education. After igniting student excitement about STEM topics in school, Fluor Idaho focuses on providing work-specific opportunities including mentoring and intern programs. These programs give students a better look at real-world job opportunities while still stimulating their interest in STEM. These programs ultimately provide career opportunities for the students and a pipeline of workers. Fluor Idaho's strategy for its summer interns has been to place students in a wide variety of projects - from the commissioning of a first-of-a-kind nuclear processing facility, to operating nuclear waste management facilities, to developing computer programs and applications - and give them responsibilities that make them feel like valued contributors. This approach has created a positive and rewarding summer-intern program experience for the students. Fluor Idaho's investment in its future today is reducing the knowledge gap - or cliff - that many contractors in the nuclear industry are facing. It's a worthwhile investment that is readily visible within its many cleanup projects. The next generation of technical professionals is making an impact now - benefiting both the company and the nuclear industry. This paper will review Fluor Idaho's programs, including STEM community outreach and education and mentoring programs. It will also present statistics about the reach of those programs and their successes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

DEM and MELT In-Can Vitrification Process for Fukushima Daiichi Water Treatment Secondary Waste - 20034

Fukushima Daiichi Nuclear Power Station was severely damaged by a tsunami induced by the earthquake on March 11, 2011. A large amount of water treatment secondary waste has been generated from the purification work of contaminated water that consists of the water flowing into the reactor building. The Mid and Long-Term Road map specifies that solid waste at Fukushima Daiichi Nuclear Power Station should be stored safely and reasonably according to its characteristics. In order to meet this goal, a versatile treatment solution is needed, and the In-Can vitrification seems to be a suitable solution. The DEM and MELT In-Can vitrification process is being developed and qualified in France to treat waste arising from decommissioning and dismantling operations. This qualification is carried out through the consortium of the DEM and MELT project coordinated by CEA which gathers Orano, ECM technologies and Andra. The DEM and MELT In-Can vitrification process is a robust, simple and versatile in situ vitrification process. It is designed for high or intermediate level waste and is compact enough to be implemented in an existing facility or close to the waste to be treated. It is developed to treat liquid and solid waste, to produce a small amount of secondary waste and to minimize investment and operating costs. The DEM and MELT In-Can vitrification process is also developed with a modular design in order to be adapted to nuclear operators needs and requirements. The In-Can melting process has already been developed and fully qualified by the CEA for Alpha liquid waste and can be rated TRL 7 for this scope. The In-Can vitrification technology is considered as a qualified solution that could be applied for the treatment of waste resulting from the D and D work in Fukushima Daiichi Nuclear Power Station. The feasibility evaluation of the DEM and MELT In-Can vitrification process for Fukushima Daiichi water treatment secondary waste is a project led by Orano, CEA and ANADEC. This R and D project is performed by the funds from the Ministry of Economy, Trade and Industry as The Subsidy Program 'Project of Decommissioning and Contaminated Water Management'. This paper presents the first results of this feasibility evaluation, from the scenarios and glass formulation definition to the industrial implementation feasibility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗