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At least 1,117 records · Page 62
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Materials Irradiation Facility (MIF) at the High Flux Isotope Reactor (HFIR)

The Material Irradiation Facility (MIF) at the Oak Ridge National Laboratory (ORNIL) High Flux Isotope Reactor (HIFR) controls and monitors instrumented irradiation experiments for materials and fuel research. In 2013 ORNIL founded a Laboratory Directed Research and Development project to reinvigorate the use of instrumented experiments by redesigning MIF to enhance its capabilities. MIF improvements include an updated and more sophisticated experiment monitoring system, a modular design allowing easier addition of simultaneous experiments, and features that allow rapid incorporation of additional instrumentation Interest in materials and fuel irradiation has increased significantly over the last decade. Much of this interest is related to investigating the performance of advanced materials for fusion environments, designing and developing improved fuel and cladding materials, and examining material characteristics and behavior under irradiation conditions. At ORNL, the irradiation engineering team has more than 60 years of experience designing and fabricating various types of irradiation experiments. Rabbit capsule experiments are small scale and relatively inexpensive. These capsules are typically designed to be irradiated in the flux-trap region of the core. Full-length target capsules are larger scale and could be either instrumented or non-instrumented. These more complicated capsules can be irradiated in any of the experiment facilities at HFIR. Instrumented experiment irradiation provides significant advantages for researchers. The unique advantage is that the researcher can monitor and acquire in situ data and change capsule operating conditions in real time during the irradiation. This capability enables tight control of desired experimental conditions so that they match the design specifications throughout the irradiation. Also, the experiment conditions (e.g., temperature and pressure) can be altered during the test, enabling significant flexibility and the ability to collect considerably more data from a single experimental capsule. This capability allows measurement of a variety of material properties (conductivity, fission product composition, etc.) that otherwise would only be available upon post-irradiation examination, where non-equilibrium conditions brought on by irradiation do not exist. The data generated by these experiments contribute significantly to the understanding of material behavior under irradiation. MIF is an experimental platform that can host fully instrumented and monitored experiments in a high neutron flux environment. Its primary functions are to provide clean inert gas (helium, neon, argon) to experiments at controlled pressures and flows, monitor irradiation conditions (typically temperatures and pressures) during irradiation, and monitor effluent gases from experiments. This facility allows monitoring of chemical processes during a fueled irradiation, where this becomes important for detecting fuel failures. This paper reports the efforts towards reestablishing the value of HFIR for instrumented irradiation experiments and the improvements to MIF that ultimately reduce the costs of instrumented irradiation. This project advances ORNL materials capabilities to new levels by providing a state-of-the art, robust and flexible facility that can be expanded as customer requirements change, while supporting the goals of the fusion, advanced fuels, and structural materials programs. (authors)

42 ENGINEERING↗
DOE OSTI2016

Overview of the US Nuclear Science User Facilities

Specialized and often unique and expensive instrumentation, adapted for use on radioactive materials, is essential to conduct research in the area of nuclear energy. Much of this equipment is located in facilities dedicated for such work that have been built and maintained over many decades. The Nuclear Science User Facilities (NSUF) was established in 2007 under the US Department of Energy Office of Nuclear Energy (DOE-NE) in order to bring these capabilities together with those innovative ideas from researchers at universities, national laboratories, small business, and industry that might not have regular access to reactors, hot cells, transportation casks, radiologically qualified instrumentation, and associated expertise. Early descriptions of the originally named Advanced Test Reactor National Scientific User Facility have been given. Through a competitively reviewed proposal process, the NSUF offers access to these capabilities and expertise at no cost to the user to promote research that will address the mission of DOE-NE. As a national organization comprised of a number of partner facilities across the US and located at National Laboratory, university, and industry sites, the NSUF supports materials science studies directed at irradiation effects in nuclear fuels and materials as well as in-pile and post irradiation examination (PIE) instrumentation development. The NSUF has created a Nuclear Fuels and Materials Library (NFML) of previously irradiated materials that users may access for their studies. Expanding the library for future applications of advanced technologies and novel concepts is an important objective of the NSUF. The NSUF has established and maintains the Nuclear Energy Infrastructure Database (NEID) as a tool for both DOE-NE and the nuclear energy research community. The extensive catalogue of capabilities is intended to aid DOE-NE, NSUF, and researchers more effectively and efficiently utilize the available capabilities. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2014

Radiation Protection for the First Korea Heavy-ion Medical Accelerator (KHIMA) Facility - Paper 7

The number of particle therapy facilities is increasing worldwide due to the encouraging results obtained at research facilities such as the Lawrence Berkeley National Laboratory (USA), the Helmholtz Center for Heavy Ion Research GSI (Germany), and the Heavy Ion Medical Accelerator (HIMAC) in Chiba (Japan). The first hospital-based heavy-ion therapy facility is under construction in Busan, Korea, at the Korea Institute of Radiological and Medical Sciences [KIRAMS]. The project was originally designed to develop a compact superconducting cyclotron for hadron therapy able to deliver forefront beams of heavier ions, essentially carbon ions of energy, 430 MeV/u. Radiation protection issues play an important role in every stage of the project such as civil construction planning, control system design, and special technological plants. Monte Carlo transport codes are the best tools to solve these issues and the FLUKA2011 and MCNPX2.7.0 codes are used in this work. Beam loss distributions of cyclotron installations in combination with passive or active beam shaping methods have a particularly high implication on the protection layout of the facility. Methods for shielding and activation calculations are described with an emphasis on carbon ion beams. All the tools and databases require verification. Benchmarking is an instrumental process for verification. An overview of benchmarking efforts will also be given and the need for further benchmarks is discussed. (authors)

43 PARTICLE ACCELERATORS↗
DOE OSTI2014

Shielding design for a chip irradiation test facility at SNS - Paper 119

As digital and analog circuits are more and more sophisticatedly integrated, they become increasingly susceptible to single event upsets (SEU), a cause of electronic equipment malfunction induced by cosmic rays with particular hazard potential for aviation equipment. Electronic chips sensitive to SEU have to be tested in a simulated comic ray environment to ensure their reliability. It is found that the spallation induced fast neutron spectrum is very close to the neutron spectrum of the cosmic ray. Hence, a conceptual design study of a chip irradiation test facility is funded by FAA to be carried out at SNS, the world's most powerful accelerator driven facility for neutron scattering studies on materials. Compared to existing/being construct chip irradiation test facilities at LANL and ISIS, SNS can provide wider neutron energy range and much higher fluxes. This study focuses on the aspect of shielding design for such a chip irradiation facility. A number of medium-weighted and heavy materials were examined for efficiency in reducing radiation dose rate. In general, the smaller the target monolith is while satisfying the personnel-protection-driven radiation dose rate limit at its outside, the closer to the target chips could be placed and thus less beam power it demands to satisfy a given irradiation flux requirement. Before proceed to the shielding design, the neutron source coming off the target has to be determined for selecting a best spot for chip irradiation test, in which the incoming fast neutron spectrum approximate most to the atmospheric radiation condition. It also provides a source term for the following shielding design of the irradiation enclosure. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2015

'Criticality events' database for fuel cycle facilities in support of the IRSN nuclear criticality safety guide - 14384

IRSN has developed a database gathering the events which have affected the safety of French fuel cycle facilities (from conversion to waste storage, laboratories and experimental facilities, facilities under dismantling and decommissioning). Among all the recorded events, it is possible to extract in particular those impacting the criticality safety. In order to extend the analysis of these events, the IRSN criticality safety department has developed a specific database dedicated to criticality risks. The 'criticality events' database is continuously updated on the basis of the declared events and their analysis by engineers from the criticality safety department. At this time, about 380 events (on 540 identified since 1980) are described in details in this database. Thus, the number of events in facilities can be between 1 and 75. It is possible to establish a link, as a feedback, between the 'criticality events' database and the sequences of events identified in the IRSN nuclear criticality safety guide. Indeed, the purpose of this guide is to be a support for criticality safety analysis or assessment. The guide is in the form of diagrams and gives advice to perform an assessment of criticality risks for each criticality control mode and for the reference fissile medium. Thus, the guide presents the parameters to be considered in the analysis and the typical sequences (written in the form of questions) related to the most current failures to be investigated. With the objective of being able to illustrate the nuclear criticality safety guide, it was decided to select the events of the database which are representative of the sequences. Thus, 70 sequences have been illustrated up to now by one or several events analyzed in the 'criticality events' database. Furthermore, this action allows 'extending' the nuclear criticality safety guide with sequences which had not yet been identified (when an event cannot illustrate a failure listed in the diagrams). This paper will describe the approach used to build the 'criticality events' database, the criteria used to classify events and the link with the IRSN nuclear criticality safety guide. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Recycling facility periodic safety review: Application of a methodology used for the nuclear criticality safety review of geometrically safe or favorable equipment - 14610

In France, for all fuel cycle facilities, the reference safety frame is continuously adapted to take into account technical developments such as changes in process and/or maintenance operations. Every ten years according to French law, the facility operator is required to launch a process of Periodic Safety Review (PSR), whose conclusions are sent to the Nuclear Safety Authority (NSA). This PSR is split in two parts: first, a review of the facility to its safety documentation and authorization, second, a safety reassessment. Actually, this PSR takes account for topics (actual aging for instance) which require several years of feedback from experience to be integrated in safety evaluations, or which cannot be done on a daily basis (documentation reorganization). It also incorporates the new regulatory safety requirements. A specific methodology was developed for the first PSR of UP3-A since the publication of the law. UP3-A is a used fuel reprocessing facility of La Hague plant. This methodology, adapted for large plants with numerous process and fissile media, is based on a two-step approach intending to focus the analysis on the real hard spots. First, 'sensitive' systems are identified. Second, a specific review is performed on these systems. After a reminder of the methodology, this paper presents the results of its application on the UP3-A scope. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

What is involved in radiation protection at accelerator facilities?

Radiation protection for accelerator facilities involves a range of actions to operate the facility within safe boundaries (often called an accelerator safety envelope) and it starts from the very early stage of facility development. Each facility will establish fundamental safety principles, requirements, and measures to control the radiation exposure to people and the release of radioactive material to the environment.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Site Integration and Regulatory Considerations for a Nuclear Power Plant Colocated with Industrial Facilities: Colocation Studies for a Petroleum Refinery, Methanol Plant, and Wood Pulp Plant

This research explores the colocation of nuclear power plants (NPPs) with industrial applications. Three existing industrial sites were considered to demonstrate the siting process and illuminate technological gaps for future work. The three applications demonstrated for colocation here are a petroleum refinery, a methanol production plant, and a pulp and paper plant. This study uses a modified version of the EPRI siting criteria to explore the geological and demographic characteristics of the location of the current industrial site, as well as exploring external hazards from the industrial plant and its surrounding land use. Data was collected from public databases to estimate site characteristics. We then discuss how the site characteristics may impact the ability to colocate an NPP with an industrial application. The application site and 5 additional sites were explored for each application to give a general indication of the siting implications for an NPP in each area. The hazards for each industrial application was also explored to determine how colocation may impact reactor safety. The following gaps have been identified and should be explored in future research on colocation of NPPs with petroleum refineries, methanol plants, and pulp and paper plants: - There is a variety of industrial use, hazards, and pipelines in the surrounding area. A more thorough review of these hazards should be considered for colocation. - In general, the whole region around some applications seems to have softer soil, with implications for large site preparation costs. Further site investigations should prioritize looking into the geotechnical conditions. - Applications along coastlines are susceptible to flooding and hurricanes. The benefits of colocation should be weighed against the potential design implications. - The benefits of natural gas pipeline infrastructure in place should be explored further. If heat supply from the NPP is not required or not feasible due to the distance between the NPP and the application, there may be an opportunity to supply hydrogen to the plant through an existing pipeline. - Because there are several collocated industrial plants in the regions for the refinery and methanol plant, the benefits of sharing resources from the NPP should be explored further. This may open up additional sites for colocation. The following knowledge gaps were identified for the colocation of NPPs with these three industries, and industrial applications in general. These gaps are: - While the STAND tool contains many important characteristics for the reactor siting process, it is not calibrated for the colocation of NPPs with industrial facilities. - There are aspects of both the NPP and industrial application that need to be quantified for a siting analysis. Particularly, we need to understand the water intake requirements for NPPs and each application. - Further work may focus on adapting the STAND site comparison methodology to comparison of sites for co-location. This will involve using the data documented in this report as a starting point and performing a comprehensive and quantitative comparison. - Without spending significant resources, it would be impossible to gather data for each site to evaluate all aspects of siting. One approach to finding data and understanding its implications to siting is looking at FSARs for existing plants. For example, most sites considered in this study have small Vs30 values, indicating soft soil. However, there are NPPs located in the vicinity of most of the sites (e.g., Waterford Steam Electric Station near New Orleans) and reviewing available site characteristics and geotechnical data for these NPPs, might provide further information for siting. - The siting analysis in this study indicates that colocation of the NPP with the industrial site could be difficult based on external hazards, cooling requirements, weather, or population. We need to determine the impact of distance between the two facilities on cost and quality of energy transport. - This study did not touch on socioeconomic impacts for NPP colocation with industrial facilities. The input-output analysis methodology could be applied to the communities referenced in this study to determine the socioeconomic impact of these projects. - Similarly, the impacts of colocation on emergency planning was not explored in this study. The impacts on emergency planning infrastructure are somewhat related to the socioeconomic impacts, and could be explored using a similar methodology. - This study also did not address physical and cybersecurity, which will be important aspects of co-location [ref] . Cybersecurity will be important, regardless of the distance, but physical security will be important if the facilities are located very closely. Physical security might also be important for the steam lines between the plants, unless they are determined to be non-safety significant. - In many site l

08 HYDROGEN↗
DOE OSTI2024

Criticality Safety Support for Deactivated Facilities at Savannah River Site

The Savannah River Site (SRS) has a number of nuclear facilities. Some date back to the mid-1950s while others are actively being built today. As facilities have aged and missions changed, a number of the older nuclear facilities have gone through various stages of deactivation, decontamination, decommissioning, and dismantlement (D4). Due to the nature of the site’s historic missions a number of these facilities processed and utilized significant quantities of fissionable materials. D4 activities have now spanned multiple decades and as such have spanned multiple regulatory and industry standards have variety of documentation levels associated with them.

Stover, Tracy E.↗
DOE OSTI2015

Technical Approach for Determining Key Parameters Needed for Modeling the Performance of Cast Stone for the Integrated Disposal Facility Performance Assessment

Washington River Protection Solutions (WRPS) and its contractors at Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) are conducting a development program to develop / refine the cementitious waste form for the wastes treated at the ETF and to provide the data needed to support the IDF PA. This technical approach document is intended to provide guidance to the cementitious waste form development program with respect to the waste form characterization and testing information needed to support the IDF PA. At the time of the preparation of this technical approach document, the IDF PA effort is just getting started and the approach to analyze the performance of the cementitious waste form has not been determined. Therefore, this document looks at a number of different approaches for evaluating the waste form performance and describes the testing needed to provide data for each approach. Though the approach addresses a cementitious secondary aqueous waste form, it is applicable to other waste forms such as Cast Stone for supplemental immobilization of Hanford LAW. The performance of Cast Stone as a physical and chemical barrier to the release of contaminants of concern (COCs) from solidification of Hanford liquid low activity waste (LAW) and secondary wastes processed through the Effluent Treatment Facility (ETF) is of critical importance to the Hanford Integrated Disposal Facility (IDF) total system performance assessment (TSPA). The effectiveness of cementitious waste forms as a barrier to COC release is expected to evolve with time. PA modeling must therefore anticipate and address processes, properties, and conditions that alter the physical and chemical controls on COC transport in the cementitious waste forms over time. Most organizations responsible for disposal facility operation and their regulators support an iterative hierarchical safety/performance assessment approach with a general philosophy that modeling provides the critical link between the short-term understanding from laboratory and field tests, and the prediction of repository performance over repository time frames and scales. One common recommendation is that experiments be designed to permit the appropriate scaling in the models. There is a large contrast in the physical and chemical properties between the Cast Stone waste package and the IDF backfill and surrounding sediments. Cast Stone exhibits low permeability, high tortuosity, low carbonate, high pH, and low Eh whereas the backfill and native sediments have high permeability, low tortuosity, high carbonate, circumneutral pH, and high Eh. These contrasts have important implications for flow, transport, and reactions across the Cast Stone – backfill interface. Over time with transport across the interface and subsequent reactions, the sharp geochemical contrast will blur and there will be a range of spatially-distributed conditions. In general, COC mobility and transport will be sensitive to these geochemical variations, which also include physical changes in porosity and permeability from mineral reactions. Therefore, PA modeling must address processes, properties, and conditions that alter the physical and chemical controls on COC transport in the cementitious waste forms over time. Section 2 of this document reviews past Hanford PAs and SRS Saltstone PAs, which to date have mostly relied on the lumped parameter COC release conceptual models for TSPA predictions, and provides some details on the chosen values for the lumped parameters. Section 3 provides more details on the hierarchical modeling strategy and processes and mechanisms that control COC release. Section 4 summarizes and lists the key parameters for which numerical values are needed to perform PAs. Section 5 provides brief summaries of the methods used to measure the needed parameters and references to get more details.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Final Design Report for the RH LLW Disposal Facility (RDF) Project

The RH LLW Disposal Facility (RDF) Project was designed by AREVA Federal Services (AFS) and the design process was managed by Battelle Energy Alliance (BEA) for the Department of Energy (DOE). The final design report for the RH LLW Disposal Facility Project is a compilation of the documents and deliverables included in the facility final design.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Assessment of Space Nuclear Thermal Propulsion Facility and Capability Needs

The development of a Nuclear Thermal Propulsion (NTP) system rests heavily upon being able to fabricate and demonstrate the performance of a high temperature nuclear fuel as well as demonstrating an integrated system prior to launch. A number of studies have been performed in the past which identified the facilities needed and the capabilities available to meet the needs and requirements identified at that time. Since that time, many facilities and capabilities within the Department of Energy have been removed or decommissioned. This paper provides a brief overview of the anticipated facility needs and identifies some promising concepts to be considered which could support the development of a nuclear thermal propulsion system. Detailed trade studies will need to be performed to support the decision making process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Compressed Natural Gas Vehicle Maintenance Facility Modification Handbook

To ensure the safety of personnel and facilities, vehicle maintenance facilities are required by law and by guidelines of the National Fire Protection Association (NFPA) and the International Fire Code (IFC) to exhibit certain design features. They are also required to be fitted with certain fire protection equipment and devices because of the potential for fire or explosion in the event of fuel leakage or spills. All fuels have an explosion or fire potential if specific conditions are present. This handbook covers the primary elements that must be considered when developing a CNG vehicle maintenance facility design that will protect against the ignition of natural gas releases. It also discusses specific protocols and training needed to ensure safety.

cng, maintenance facility, compressed natural gas,↗
DOE OSTI2017

Code eOMEGA - Decommissioning Costing of Nuclear Facilities Based on the ISDC - 17230

The paper is focused on the presentation of the development process, calculation procedure and main features of the eOMEGA code used for decommissioning costing of nuclear facilities. The methodology for decommissioning cost calculation implemented to the eOMEGA code is fully in line with the international recommendations, best practices and is completely based on the ISDC - The International Structure for Decommissioning Costing (ISDC) of Nuclear Installation. ISDC represents the standard and common platform for presenting the results of the cost estimates for decommissioning projects related to any type of nuclear facility. The main goal of the eOMEGA costing code is to provide the ISDC based flexible, open, modern and user-friendly tool for transparent and reliable decommissioning costing for any type of nuclear facility with any systems, structures and radiological situation at the end of the operational period. Moreover, the eOMEGA decommissioning costing code is intended to be one of the modules of the universal eOMEGA platform which should cover also the other activities within the back-end of nuclear power engineering e.g. decommissioning funding, back end of nuclear fuel cycle, site remediation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Controversial Projects and Potential Implications for the Consent-Based Siting of Nuclear Waste Management Facilities - 17345

The U.S. Department of Energy's (DOE) goal is to develop solutions for the long-term, sustainable management of the nation's spent nuclear fuel and high-level radioactive waste. The DOE is planning for an integrated waste management system to transport, store, and dispose of spent nuclear fuel and high-level radioactive waste from commercial electricity generation, as well national defense activities. To achieve this goal, the DOE is developing a process to site facilities collaboratively with the public, communities, stakeholders, and governments at the state, tribal, and local levels. The DOE is seeking the help of all Americans in developing a consent-based approach to siting that is fair and reflective of public input. To inform DOE's efforts to develop a consent-based process for siting facilities an effort was undertaken to review other contentious projects including contaminated site remediation and facilities and infrastructure deployment to understand the role of different parties involved and of public involvement for both successful and unsuccessful projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Considerations for recovery of large areas of a nuclear facility for re-use - 15147

This paper highlights a number of factors an organization might consider for recovery of a facility, using a successful example from the UK. In 2013 a power failure led to the shutdown of the ventilation system that caused loose contamination to be deposited over large sections of man-accessible areas of plant after the evacuation of personnel. The successful initiation of emergency Command and Control arrangements during the Response phase, saw several key responders brought together to offer technical direction across many disciplines. The actions of the Command and Control unit prevented any further spread contamination to other areas of the facility and any external release. This enabled the return of the plant to an operational status which forms an important mission enabler for the site hazard reduction programme. Intervention at an early stage during the Response phase of the incident saw the cessation of traditional dilute chemical wet decontamination approach that would otherwise leave a legacy of slow leaching contamination; to that of containment and ultimate removal of loose contamination using strippable coatings. Doing so established a more rapid and reliably 'clean' access or pathway into the plant to establish the full extent of the situation. The scale and scope of the decontamination requirement was steadily revealed through a number of controlled re-entries to the plant. A number of other techniques were considered and employed to address the different areas making use of available plant resources to deliver the successful 'self-help' recovery of the plant using multi-disciplinary teams. As Response switched to Recovery, it was important to set out and agree which strategic option should be adopted to return the facility to normal operation. Equally, it was necessary to verify which equipment had been affected by the power outage and repair as required. Furthermore, equipment with designated safety functions had to be maintained alongside the Recovery activities. A careful balance of priorities demanded an unparalleled depth of plant knowledge to address this, whilst still keeping site and external stakeholders informed. The engagement by the plant workforce to take on the decontamination challenge to enable recovery of their plant supported by other teams formed a critical step in the speed of recovery. (authors)

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

Best Practice on Facility Characterisation from a Material and Waste End- State Perspective - 16041

Radiological characterisation plays a significant role in the process of decommissioning of shut-down nuclear facilities in order to ensure the protection of the environment and radiation safety. At all stages of a decommissioning programme or project, adequate radiological characterisation is of crucial importance, not least from a material and waste perspective. The radiological characterisation is a key element for planning, controlling and optimising decommissioning and dismantling activities. Experience has shown that data and information from the operation of a facility can - supplemented by recently collected and analysed data and information - be of crucial importance for decisions on waste management and for characterisation of radioactive waste. Once the dismantling has been done, some information may be hard, costly or even impossible to obtain later in the waste management process. This was the reason why the Working Party on Decommissioning and Dismantling (WPDD) of the OECD Nuclear Energy Agency (NEA) decided in late 2013 to extend the mandate of the Task Group on Radiological Characterisation and Decommissioning (TGRCD) for a second phase focusing on nuclear facility characterisation from a waste and material end-state perspective whereas the first phase focused on overall strategies of radiological characterisation. This paper gives an overview of the activities and findings within both phases up to now. (authors)

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

Thermo-Hydro-Geochemical Evolution Influences on the Concrete Cask in Long-Term Storage Facility of Spent Fuel - 16171

Due to humid environment and weather condition in Taiwan's island, as well as the acid rain environment resulting from sulfur dioxide emissions, the concrete cask might be damaged or degraded under the effect of humid, salt, acid rain, and sulfate attack on cementitious materials. These degradation factors could influence the safety design of the long-term spent fuel storage facility. Due to the long-term decay heat of spent fuel and severe environments to be encountered, the thermo-hydro-geochemical evolution could play an important role in the safety assessment of concrete-cask storage facility. The effects of humid, salt, acid rain, and sulfate attack on the degradation of cementitious materials of concrete-cask facility in thermo-hydro-geochemical evolution under rain water quality conditions have been investigated. The results show that the main processes responsible for concrete degradation are species induced from hydrogen ion, sulfate and chloride. For most storage and disposal casks and conditions, the American Concrete Institute nuclear safety code requirements (ACI-349 Appendix A) specify a bulk-concrete shield temperature limit of 150 deg. F (65.5 deg. C). However, localized temperatures at the inner walls of the shielding casks could reach 200-300 deg. F (93-149 deg. C). Simulation of geochemical degradation reactions for 100 years of cask storage under two temperature scenarios was performed. The results indicate that the hydrogen ion, sulfate anion, and chloride do not result in the dissolution of portlandite, or formation of ettringite and Friedel's salt. Therefore, these results indicate the concrete cask may remain in a durable condition. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗