Design of a Prototype Differential Die‐Away Instrument Proposed for Swedish Spent Nuclear Fuel Characterization
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Martinik, Tomas (ORCID:0000000165255328)↗
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Laser ablation spectrometry for studies of uranium plasmas, reactor monitoring, and spent fuel safety
Nuclear security is one of the defining challenges of our time. Nuclear threats range from deliberate dispersal of radioactive material to contaminate the vital infrastructure to diversion and smuggling of special nuclear material for clandestine nuclear programs and nuclear terrorism, respectively. There is an associated need to develop and sustain the nuclear forensics capabilities, which requires the understanding of complex processes that occur in plasmas of nuclear materials. The area of nuclear safety has seen a resurgence of public interest, and there is a concomitant need to safely store used nuclear fuel and detect structural material failure in nuclear power systems, especially in innovative reactor designs envisioned for future adoption. Laser-produced plasmas are complicated extreme environments that can generate intense and rich, highly specific signatures of nuclear and radiological materials, which can then be explored in a wide range of applications. They include interdiction and rapid detection of nuclear materials, including their isotopic composition, detection over long distances, laboratory simulation of weapons effects, monitoring the condition of structural materials in dry cask storage containers, and novel instrumentation for nuclear power systems. We present a compilation of recent representative examples of the application of laser spectroscopy, and laser-induced breakdown spectroscopy in particular, to nuclear safety and security problems. A case is made that spectroscopic techniques based on laser-produced plasmas offer complementary, and sometimes unique, capabilities that motivate the continued exploration of their efficient production and understanding of the signatures they produce
(020.3260) Isotope shifts, (140.3440) Laser-induce↗
MELCOR DEMONSTRATION ANALYSIS OF ACCIDENT SCENARIOS AT A SPENT NUCLEAR REPROCESSING PLANT.
Abstract not provided.
Wagner, Kenneth Charles↗
Quantify the Initial Enrichment, Burnup and Cooling Time of Individual Spent Fuel Assemblies for the NGSI Spent Fuel Nondestructive Assay Project [Slides]
Abstract not provided.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Predicted Liquid Atomization from a Spent Nuclear Fuel Reprocessing Pressurization Event.
Abstract not provided.
Brown, Alexander↗
Recycling of Dilute Deacetylation Black Liquor to Enable Efficient Recovery and Reuse of Spent Chemicals and Biomass Pretreatment Waste
Previous publications showed deacetylation/dilute alkaline pretreatment followed by mechanical refining (DMR) is an effective process for overcoming biomass recalcitrance. To successfully commercialize alkaline-based pretreatments, we pioneered a new process modification and strategy to recycle and reuse the weak black liquor (WBL) to reduce water, chemical, and energy usage while increasing the solids, sodium, acetate, sugars, and lignin contents in the weak black liquor. Our results suggest that the accumulation of both chemicals and organic matter did not reduce acetyl and lignin removal in alkaline pretreatment, resulting in comparable sugar yields in enzymatic hydrolysis. Sodium and potassium were found to be the two most important inorganic substances in the recycled weak black liquor. Moreover, the accumulated sodium and phenolic compounds did not inhibit the downstream fermentation and ethanol fermentation processes. Finally, techno-economic analysis (TEA) showed the recycling of weak black liquor could decrease the minimum ethanol selling price (MESP) by approximately 5 to 15 cents per gallon of ethanol when compared to a conventional non-recycling process. solids, sodium, acetate, sugars, and lignin contents in the weak black liquor. Our results suggest that the accumulation of both chemicals and organic matter did not reduce acetyl and lignin removal in alkaline pretreatment, resulting in comparable sugar yields in enzymatic hydrolysis. Sodium and potassium were found to be the two most important inorganic substances in the recycled weak black liquor. Moreover, the accumulated sodium and phenolic compounds did not inhibit the downstream fermentation and ethanol fermentation processes. Finally, techno-economic analysis (TEA) showed the recycling of weak black liquor could decrease the minimum ethanol selling price (MESP) by approximately 5 to 15 cents per gallon of ethanol when compared to a conventional non-recycling process.
09 BIOMASS FUELS↗
Gamma Scanning of 25 PWR Spent Fuel Rods in the High Burnup Spent Fuel Data Project
Explore the source record for details and available documents.
Montgomery, Rose↗
Research and Experiment on the Key Technologies of the Spent Fuel Storage in HTR-PM - 18486
High Temperature Reactor demonstration project of China (HTR-PM) is one of the major national science and technology projects, some key technologies for the design and operation of the demonstration project are formed to research according to the main technology. The spent fuel storage system is one of the main systems of the HTR-PM. In order to solve the technical problems related to the operation of spent fuel, several experiments have been conducted, including: spent fuel loading, getting back, spent fuel canister hoisting, storage well cover hoisting, storage canister drop and so on. In the study process, the test facility has been set up to carry on the operation and the experiment verification to ensure the reliability and safety of the spent fuel storage system. Spent fuel storage system undertakes spent fuel receiving, spent fuel canister transfer, storage, lifting and other functions in HTR-PM, the project research need to solve several major technical problems related to the spent fuel operation, which mainly include: Develop the transportable spent fuel canister for HTR-PM, the canister is corrosion resistance, impact resistance, and is in subcritical condition when loaded with fuel. Develop the spent fuel canister hoister, which could hoist the spent fuel canister safely and reliably. Develop the spent fuel storage well cover hoister, which could hoist the well cover safely and reliably. Study the small probability of the spent fuel canister in extreme condition of canister hoist falling accidents, to ensure the radiation protect safety and residual heat removal safety of the canister in case of the canister drop when lifting. Develop the spent fuel charging device and return device, which could load the spent fuel elements into the spent fuel canister, and could take out the fuel elements from the canister. The key technology and equipment of the spent fuel storage system in HTR-PM would be solved through the research of the project. Project research is very important for the implementation and safe operation of the HTR-PM project. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Analysis of Transportation Options for Commercial Spent Fuel in the U.S.
The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and high and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and high and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and highand associated transportation of spent nuclear fuel (SNF) and high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF)...
Kalinina, Elena↗
The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident
The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. Especially after the Fukushima NPP accident occurred, the importance of NPP safety analysis has been raised and there is more concern for the safety of NPPs in Taiwan. Because the earthquake and tsunami occurred, the cooling system of spent fuel pool failed and the safety issue of spent fuel pool generated in Japan's Fukushima NPP. Chinshan NPP is the first NPP in Taiwan which is BWR/4 plant and OLTP (Original Licensed Thermal Power) for each unit is 1775 MWt. Chinshan NPP finished SPU (stretch power uprate) and the operating power is 103.66% of the OLTP, which is 1840 MWt now. After Fukushima NPP accident occurred, in order to concern the safety of Chinshan NPP spent fuel pool, we performed the safety analysis of spent fuel pool by using MELCOR/SNAP which also assumed the cooling system of spent fuel pool failed. The geometry of Chinshan NPP spent fuel pool is 12.17 m x 7.87 m x 11.61 m and the initial condition is 60 deg. (water temperature) / 1.013 x 10{sup 5} Pa. And, the total power of the fuels is roughly 8.9 MWt initially. MELCOR is developed by Sandia National Laboratories. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor NPPs, including containment response and leakage of activity to the environment. A broad spectrum of severe accident phenomena in both BWRs and PWRs is treated in MELCOR in a unified framework. These include thermalhydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. According to Dr. Carbajo's paper, MELCOR was used to perform the study of spent fuel pool of Fukushima Daiichi Unit 4. Niina E. Koenoenen used MELCOR to establish the spent fuel pool model of a Nordic BWR. The loss-of-pool-cooling accidents was simulated and analyzed by this model. The above studies indicate that MELCOR is capable of handling the simulation of the spent fuel pool. An increasing number of researchers are using MELCOR code to analyze test facilities and nuclear power plants. Jun Wang et al. used MELCOR to perform core degradation simulation and MELCOR results were compared with CORA experimental data. Through this work, they reviewed the performance of MELCOR COR package in detail. Tae Woon Kim et al. developed MELCOR model of APR1400 (an advanced pressurized water reactor). They used this model to perform the LBLOCA transient simulation and sensitivity study. Longze Li et al. established the MELCOR model of CPR1000 (a Chinese pressurized reactor 1000-MW power plant). The station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV) transient was simulated and analyzed by this model. SNAP is a graphic user interface program that processes the inputs and outputs of MELCOR. In addition, there is an animation function in SNAP which can present the animation of analysis results. There are two main steps in this research. The first step was the establishment of Chinshan NPP spent fuel pool MELCOR/SNAP model. And the transient analysis of Chinshan NPP spent fuel pool MELCOR/SNAP model under the cooling system failure condition was performed. The next step was the comparison among the MELCOR, TRACE and CFD data. TRACE and CFD data were from INER reports. In addition, the animation model of Chinshan NPP spent fuel pool was presented by using the animation function of SNAP with MELCOR/SNAP analysis results. This study has developed MELCOR/SNAP model of Chinshan NPP spent fuel pool successfully. By using the above model, the safety analysis of the spent fuel pool was performed under the cooling system of spent fuel pool failed condition. The analysis results of MELCOR, TRACE and CFD were similar in this case. It indicated that there was a respectable accuracy in MELCOR/SNAP model. The analysis results depicted that the uncovered of the fuels occurred at 2.6 day and the metal-water reaction of fuels occurred roughly at 3.7 day after the cooling system failed. The above results indicated that the failure of cladding occurred after 3.7 day. This study's results can help to evaluate the safety issue of Chinshan NPP spent fuel pool. In addition, the cooling system and some mitigation measures of spent fuel pool is not simulated in the above model. However, we will do it in the future. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
US Spent (Used) Fuel Status, Management and Likely Directions- 12522
As of 2010, the US has accumulated 65,200 MTU (42,300 MTU of PWR's; 23,000 MTU of BWR's) of spent (irradiated or used) fuel from 104 operating commercial nuclear power plants situated at 65 sites in 31 States and from previously shutdown commercial nuclear power plants. Further, the Department of Energy (DOE) has responsibility for an additional 2458 MTU of DOE-owned defense and non defense spent fuel from naval nuclear power reactors, various non-commercial test reactors and reactor demonstrations. The US has no centralized large spent fuel storage facility for either commercial spent fuel or DOE-owned spent fuel. The 65,200 MTU of US spent fuel is being safely stored by US utilities at numerous reactor sites in (wet) pools or (dry) metal or concrete casks. As of November 2010, the US had 63 'independent spent fuel storage installations' (or ISFSI's) licensed by the US Nuclear Regulatory Commission located at 57 sites in 33 states. Over 1400 casks loaded with spent fuel for dry storage are at these licensed ISFSI's; 47 sites are located at commercial reactor sites and 10 are located 'away' from a reactor (AFR's) site. DOE's small fraction of a 2458 MTU spent fuel inventory, which is not commercial spent fuel, is with the exception of 2 MTU, being stored at 4 sites in 4 States. The decades old US policy of a 'once through' fuel cycle with no recycle of spent fuel was set into a state of 'mass confusion or disruption' when the new US President Obama's administration started in early 2010 stopping the only US geologic disposal repository at the Yucca Mountain site in the State of Nevada from being developed and licensed. The practical result is that US nuclear power plant operators will have to continue to be responsible for managing and storing their own spent fuel for an indefinite period of time at many different sites in order to continue to generate electricity because there is no current US government plan, schedule or policy for taking possession of accumulated spent fuel from the utilities. There are technical solutions for continuing the safe storage of spent fuel for 100 years or more and these solutions will be implemented by the US utilities that need to keep their nuclear power plants operating while the unknown political events are played out to establish future US policy decisions that can remain in place long enough regarding accumulated spent fuel inventories to implement any new US spent fuel centralized storage or disposition policy by the US government. (author)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Identify and Assess Technical Challenges in Safeguards Measurements of Spent Advanced Reactor Fuels
Advanced reactor (AR) designs use various nuclear fuel types that can be significantly different than conventional light-water reactor (LWR) fuels, including differences in sizes, compositions, and chemical forms (e.g., oxide, carbide, metal). Nearly all the proposed AR fuels use high-assay low-enriched uranium (HALEU), which will have higher enrichments (5–20 wt% 235 U) than LWR fuels (currently limited to <5 wt% 235 U). In advance of the wide use of these new fuel types around the world, international safeguards organizations such as the International Atomic Energy Agency (IAEA]) are working with some of the AR vendors to formulate safeguards approaches for these AR fuel cycles. As part of the overall safeguards approach, it is important to identify the potential technical challenges in performing safeguards verification measurements of these AR fuels (both fresh and spent fuels) in advance of the widespread adoption of these new fuel types, because new safeguards technologies can take several years to develop, test, and approve for use. This report documents work performed in fiscal year 2024 based on modeling and simulation to assess the performance of the existing safeguards measurement technologies for irradiated or spent AR fuel elements or items. This work is a continuation of the work performed in fiscal year 2023 that focused on fresh AR fuels. Spent AR fuels have a distinct difference from their LWR counterparts: unlike the spent LWR fuels typically stored in a water-filled pool, some spent AR fuels—such as tristructural-isotropic (TRISO)-based fuels—will most likely be stored in air-filled hot cells. Because most safeguards measurements on spent fuel performed to date have been conducted under water, the air-filled hot cell environment could present unique challenges to safeguards measurements. Fork detector (FDET) and Cerenkov viewing device (CVD) systems have been the two primary instruments used by the IAEA for several decades to measure spent LWR fuel assemblies stored in pools for safeguards verification purposes. Because the lower refractive index of air causes Cerenkov light to be of lower intensity in air than in water, existing CVDs are likely unable to perform safeguards verification measurements for spent fuel stored in an air-filled hot cell, as is the case for the TRISO-based spent fuel elements (e.g., pebbles, graphite fuel blocks). Unlike FDET measurements, CVD measurements do not require fuel be moved, so they are a simpler and faster to take than FDET measurements. The inability to perform CVD measurements on the TRISO-based AR fuel types presents a major technical challenge in the effort to use existing technology to perform safeguards measurements on spent AR fuels. This study was mainly conducted through the modeling and simulation of an FDET or an FDET-like system on five spent AR fuel types, including one metallic fuel type and four TRISO-based fuel types in both pebble and graphite block forms in their respective storage configurations and environments. Because the various AR fuel types have significantly different dimensions, FDET systems must be adapted to accommodate them. Partial defect tests were also simulated in this study to assess the FDET’s ability to detect potential fuel diversions. The FDET measures the fuel’s total passive neutron and gamma emissions. The simulated FDET results from spent AR fuel items are compared against results from a typical spent pressurized water reactor (PWR) assembly. High-purity germanium (HPGe) gamma detector measurements were also simulated for the spent AR fuel types and the PWR assembly because the signature photopeaks have been used in LWR safeguards verifications, although HPGe is usually not used to detect diversions because of the fuel’s self-attenuation effects on those photopeaks. The results indicate that these detectors have significant challenges in performing safeguards measurements of the spent AR fuel items, including incompatibilities between AR fuel items and existing FDETs, lower neutron count rates, lower sensitivities to fuel diversions in certain AR fuel items, and significantly higher interference from a neighboring fuel item when the measurement is performed in air. These results suggest that an alternative technology or significant and timely technology development is needed to perform adequate safeguards measurements of some of these AR fuel items.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation
Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Verification and Analysis of Spent Fuel Safety Information Database in Korea - 18252
The Public Engagement Commission on SNF management (PECOS) published the recommendation report on the spent fuel management of Korea in June, 2015. Based on this report, the national basic policy for high level radioactive waste management of Korea was established and promulgated, taking into consideration the national/international trends on policy and technology development. According to the national plan, the siting for deep geological disposal facility of high level radioactive waste will be proceeded by 2028. Once the site is selected at somewhere, centralized spent fuel storage facility will be constructed by 2035 at the same site while research program is ongoing to determine the compatibility of the site for disposal facility. According to the enforcement decree of the nuclear safety act, the definition of 'interim storage' is that a safe storage of spent fuel which is received from NPP operator for a specified period, until it is processed or permanently disposed of. It means that spent fuels should be taken over with integrated information of spent fuel from NPP operator to the KORAD (Korea Radioactive waste agency). The aim of this study is to confirm and analyze present and future spent fuel data in order to provide the basic information to the safety research on the transport and storage of spent fuel in Korea. AMORES (Automatic Multi-batch ORIGEN Runner for Evaluation of Spent fuel) was used for the aim of this study. All 16,160 bundles of PWR spent fuel database as of the end of 2015 and the expected spent fuel which will be generated from 2016 to 2053 are evaluated and verified based on national basic plan. Through scenario analysis, it was estimated that approximately 15,000 MTU of spent fuel will be transported to intermediate storage facilities by 2052. Therefore, 44%, 6%, and 33% of spent fuel will be left in Kori, Hanbit and Hanul sites, respectively. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Development of the Licensing Procedure and Regulatory Framework for the Spent Fuel Storage facility in Korea - 16423
The Public Engagement Commission on SNF management (PECOS) published the recommendation report on the spent fuel management of Korea in June, 2015. Based on this report, the future national policy for spent fuel management of Korea has been established, taking into consideration the national/international trends on policy and technology development. Regardless of the recommendation report of PECOS, Spent fuel storage casks are necessary in Korea. Therefore, various types of spent fuel storage casks are now being developed by Korea vendors. For example, Concrete storage casks and dual purpose metal casks are now under development by the KORAD (Korea Radioactive waste agency) for future use in the interim storage facility. Unfortunately, there is no independent licensing procedure for spent fuel storage casks in Nuclear Safety Act (NSA) because is the casks are considered an integral part of the interim storage facility safety equipment. Furthermore, the definition of radioactive waste in the NSA was changed recently. Now, spent fuel is separated from the definition of 'radioactive waste'. Spent fuel can only become radioactive waste by the decision through the meeting of the relative ministers of Korea government. For that reason, the licensing procedure for interim spent fuel storage facilities was separated from the current 'disposal facility, etc.' in accordance with the article 63 of NSA. The aim of this study is to develop the licensing procedure for the storage cask, and also to develop a revised draft of interim storage facility procedure within the NSA. Once this independent licensing procedure on the spent fuel storage cask is introduced to the NSA, it is expected that developers or operators can develop and commercialize the storage cask in Korea. In this study, an independent licensing procedure for the spent fuel storage facility was developed including the storage cask licensing procedure for the NSA. The procedure was composed of the permit for construction and operation of interim storage facility, criteria for permit, inspection criteria, periodic safety review, and decommissioning criteria, etc. The detailed contents of these requirements were also developed to introduce the regulatory framework for the interim storage facility. Particularly, periodic safety review and decommissioning criteria were developed in accordance with the recommendation of IRRS (International Regulatory Review Service) of IAEA. The developed licensing procedure for spent fuel storage facility will be introduced in NSA in response to the revised definition of NSA and the recommendation report of PECOS. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
A spent coffee grounds based biorefinery for the production of biofuels, biopolymers, antioxidants and biocomposites
Highlights: • Spent coffee grounds is used as a resource for the preparation of biofuels and value added products. • In-situ transesterification of spent coffee grounds show high biodiesel yield. • Biopolymer, biosorbent, antioxidant, and biocomposite are obtained from spent coffee grounds. • A biorefinery based on spent coffee grounds will positively impact the biobased economy. • Circular economy concept can be applied to spent coffee grounds valorisation. - Abstract: Spent coffee grounds are composed of lipid, carbohydrates, carbonaceous, and nitrogen containing compounds among others. Using n-hexane and n-hexane/isopropanol mixture highest oil yield was achived during soxhlet extraction of oil from spent coffee grounds. Alternatively, supercritical carbon dioxide can be employed as a green solvent for the extraction of oil. Using advanced chemical and biotechnological methods, spent coffee grounds are converted to various biofuels such as, biodiesel, renewable diesel, bioethanol, bioethers, bio-oil, biochar, and biogas. The in-situ transesterification of spent coffee grounds was carried out in a large scale (4 kg), which led to 80–83% biodiesel yield. In addition, a large number of value added and diversified products viz. polyhydroxyalkanoates, biosorbent, activated carbon, polyol, polyurethane foam, carotenoid, phenolic antioxidants, and green composite are obtained from spent coffee grounds. The principles of circular economy are applied to develop a sustanaible biorefinery based on valorisation of spent coffee grounds.
09 BIOMASS FUELS↗
Development of the computer code for multi-calculations of source term of spent fuel - 15351
Many PWR spent fuels have been transferred between the neighboring nuclear power plants at Kori, Hanbit, Hanwul site, and from nuclear power plant to the KAERI(Korea Atomic Energy Research Institute) for the research purpose in Korea. In case of the PHWR type spent fuels, these have been transferred from the spent fuel pool to an on-site dry storage facility. Total radioactivity of spent fuels has to be calculated before transport in advance by the regulation requirement. In case of PWR, KN-12 transport cask can transfer 12 spent fuel assemblies at one time and in case of PHWR, Hi-STAR 63 transport cask can transfer 120 spent fuel assemblies at one time. Therefore, total radioactivity of all transported spent fuel assemblies that are transported generally are calculated by using the ORIGEN-ARP to meet this regulatory requirement. Basically, ORIGEN-ARP calculates the one fuel assembly at one time. SCALE code including this code provides the batch mode to perform the multi-calculations of the source term. In this study, in order to use the batch mode of ORIGEN-ARP, ARP-auto program was developed to generate input files and batch file of ORIGEN-ARP automatically and to make the summary file extracting the result from output files of ORIGEN-ARP. This code enhances the efficiency at the field. The working time to complete the calculation of source terms of 120 spent fuel assemblies take only 20 minutes. In order to confirm the credibility of this code, total radioactivity of 12 spent fuel assemblies of 17x17 Westinghouse type were calculated and compared to the previous result. The initial enrichments of these spent fuels are 3.46 ∼ 3.50 w/o and burnups are 29 ∼ 43 GWD/MTU. As the results, the relative errors are about 0.9 ∼ 3.0%. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Development of Polymer Waste Form for Spent Ion-Exchange Resins
A large amount of radioactive waste streams has been treated by ion-exchange resins at most of nuclear facilities for many years. These activities leave radionuclides as well as chemical contaminants on the spent resins, and the spent resins are considered as low- and intermediate-level radioactive waste. Spent ion-exchange resins require special immobilization treatment to meet the waste acceptance criteria for disposal. Immobilization of spent ion-exchange resins can be processed by mixing or incorporating them with suitable solidifying materials, such as cement, bitumen, organic polymers, and other materials or a combination of them. The selection of immobilization options for spent ion-exchange resins must consider their physical, chemical, and radiological characteristics. In this study, direct immobilization method was used to solidify the spent ion-exchange resins using organic polymer for improving the waste content. The objective of this study is to test the immobilization of spent ion-exchange resins using polymer waste form and enhance the solidified waste content. The specific objective of this study was to develop polymer waste form to solidify spent ion-exchange resins. Leaching (ANSI/ANS 16.1) and compressive strength tests were performed to meet the acceptance criteria of waste form for disposal. Ion-exchange resin has been solidified physically and chemically stable in the polymer waste form. The leaching test (ANSI/ANS16.1) for two polymer waste form specimens were conducted. The leachability indexes of sample A and B showed to meet the acceptance criteria for disposal. The compressive strength tested showed higher values in polymer waste form prepared with ion exchange resin compared to CoCl{sub 2} polymer form (sample B). Although spent resin was successively solidified in polymer matrix, the future experiment will develop a polymer waste form as a pre-treatment of spent resin for size reduction using a ball milling (500 RPM, 280 min)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗