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At least 1,243 records · Page 69
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The UFDC Storage and Transportation staffs are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While low burnup fuel [that characterized as having a burnup of less than 45 gigawatt days per metric tonne uranium (GWD/MTU)] has been stored for nearly three decades, the storage of high burnup used fuels is more recent. The DOE has funded a demonstration project to confirm the behavior of used high burnup fuel under prototypic conditions. The Electric Power Research Institute (EPRI) is leading a project team to develop and implement the Test Plan to collect this data from a UNF dry storage system containing high burnup fuel. The Draft Test Plan for the demonstration outlines the data to be collected; the high burnup fuel to be included; the technical data gaps the data will address; and the storage system design, procedures, and licensing necessary to implement the Test Plan. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must closely mimic real conditions high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage. To document the initial condition of the used fuel prior to emplacement in a storage system, “sister ” fuel rods will be harvested and sent to a national laboratory for characterization and archival purposes. This report supports the demonstration by describing how sister rods will be shipped and received at a national laboratory, and recommending basic nondestructive and destructive analyses to assure the fuel rods are adequately characterized for UFDC work. For this report, a hub-and-spoke model is proposed, with one location serving as the hub for fuel rod receipt and characterization. In this model, fuel and/or clad would be sent to other locations when capabilities at the hub were inadequate or nonexistent. This model has been proposed to reduce DOE-NE’s obligation for waste cleanup and decontamination of equipment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Monte Carlo fuel burnup analyses of the Accelerator-Driven Subcritical systems

Here, we discuss the Monte Carlo fuel burnup analysis performed for the Argonne National Laboratory (ANL) Accelerator-Driven Subcritical (ADS) conceptual design. The ANL ADS conceptual design was determined to dispose of the 131 tons of Minor Actinides (MA) fuel of the current United States Spent Nuclear Fuel (SNF) inventory. It uses a neutron source from 1 GeV 25MW proton beam with the liquid lead target to drive the subcritical fission assembly. It utilizes the liquid metal slurry fuel concept to load the MAs fuel. The subcritical fission assembly has a neutron multiplication factor of 0.98. The total fission power generated in the system is about 2.67 GW. The system is designed to operate for 35 full power years. Monte Carlo fuel burnup analyses were performed to calculate the amount of MAs consumed by the ANL ADS conceptual design. The MCNP6 computer program was used to simulate the high energy physics of the protons striking the liquid lead target for generating the neutron source. The MCB5 and SERPENT Monte Carlo burnup computer programs were used to simulate the fuel burnup during the ADS operation. In the ANL ADS conceptual configuration, the reactivity loss during the fuel burnup is compensated by adding fresh fuel into the subcritical assembly. In this paper, both batch and continuous fueling modes are examined. The fresh fuel particles are loaded into the slurry fuel carrier at the beginning of each burnup step of the batch-fueling mode and loaded continuously into the slurry fuel carrier during the operation in the continuous fueling mode. Monte Carlo fuel burnup analyses show that one ANL ADS conceptual design can consume about 25.7 tons of MAs fuel during the 35 full power years with the continuous fueling mode. About five units is required to dispose of the 131 tons of MAs

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Development of Metallic Fuels for Actinide Transmutation

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10 and 60%. In general, the performance of all of these substantially disparate metallic fuel alloys has been observed to be excellent, and their irradiation behaviors are generally consistent with historic norms for metallic fuels without minor actinide additions and having lower Pu or Zr contents. Future work is being undertaken with a view toward increasing the burnup potential of metallic fuels even more. Design innovations under investigation include: 1) lowering the fuel smear density in order to accommodate more swelling, 2) annular fuel geometry to eliminate the need for a sodium bond, 3) minor alloy additions to stabilize lanthanide fission products inside the fuel and prevent their transport to the cladding where they can participate in fuel-cladding chemical interaction (FCCI), and 4) coatings/liners on the cladding inner surface to mitigate FCCI and enable higher temperature operation. This paper will present the current state of development of metallic fuels for actinide transmutation in the US. Highlights will include recent results from metallic fuel casting experiments, experiments to identify alloy additions to immobilize lanthanide fission products, and postirradiation examinations of annular metallic fuels at low burnup.

actinide transmutation↗
DOE OSTI2017

The characterization of microstructure and chemistry of transmutation fuels

The goal of the fast reactor development programs is fabrication and certification of a nuclear fuel system that performs all functions of a conventional fuel while destroying recycled actinides. Metal fuels, such as U-Pu-Zr fuels, are considered to be candidates for this application. The physical and mechanical properties, as well as behavior laws and correlations of metal alloy fuel, all of which define the behavior and performance of metal alloy fuels has been investigated in detail. The results indicated that the prospect for fueling a future fast spectrum reactor using metal fuels is excellent. There are several challenges associated with the fuel implementation in reactor environment. One of the limiting factors in lifetime and safety of thermal and fast reactor systems is fuel swelling and resulting fission product transport to cladding. Understanding diffusion kinetics and phases formed between metal fuel and cladding upon exposure to high temperatures typical in reactor environment is critical for ensuring integrity, safety, and performance of the material, advancement of this type of fuel in the future, and securing long-term success of the nuclear fleet. The first step to understanding fuel-cladding interaction is investigation of an as-cast fuel microstructure. Despite the importance of these fuels, detailed description of microstructures in these materials is extremely limited. The majority of published results are based on characterization of as-cast microstructures of these fuels in scanning electron microscopes (SEM). In this work, we present the results from microstructural characterization of metallic Pu-based fuels, which is essential in the validation of current and future atomistic and meso-scale modeling efforts. A firm understanding of microstructural evolution in reactor materials is imperative and will lead to deeper understanding of in-reactor degradation behavior, creating a new generation of radiation-resistant materials and fuels with the potential for significant increases in burn-up.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Apparatus and method for mixing fuel in a gas turbine nozzle

A nozzle includes a fuel plenum and an air plenum downstream of the fuel plenum. A primary fuel channel includes an inlet in fluid communication with the fuel plenum and a primary air port in fluid communication with the air plenum. Secondary fuel channels radially outward of the primary fuel channel include a secondary fuel port in fluid communication with the fuel plenum. A shroud circumferentially surrounds the secondary fuel channels. A method for mixing fuel and air in a nozzle prior to combustion includes flowing fuel to a fuel plenum and flowing air to an air plenum downstream of the fuel plenum. The method further includes injecting fuel from the fuel plenum through a primary fuel passage, injecting fuel from the fuel plenum through secondary fuel passages, and injecting air from the air plenum through the primary fuel passage.

42 ENGINEERING↗
DOE OSTI2014

Subtask 3.11 - Production of CBTL-Based Jet Fuels from Biomass-Based Feedstocks and Montana Coal

The Energy & Environmental Research Center (EERC), in partnership with the U.S. Department of Energy (DOE) and Accelergy Corporation, an advanced fuels developer with technologies exclusively licensed from Exxon Mobil, undertook Subtask 3.11 to use a recently installed bench-scale direct coal liquefaction (DCL) system capable of converting 45 pounds/hour of pulverized, dried coal to a liquid suitable for upgrading to fuels and/or chemicals. The process involves liquefaction of Rosebud mine coal (Montana coal) coupled with an upgrading scheme to produce a naphthenic fuel. The upgrading comprises catalytic hydrotreating and saturation to produce naphthenic fuel. A synthetic jet fuel was prepared by blending equal volumes of naphthenic fuel with similar aliphatic fuel derived from biomass and 11 volume % of aromatic hydrocarbons. The synthetic fuel was tested using standard ASTM International techniques to determine compliance with JP-8 fuel. The composite fuel thus produced not only meets but exceeds the military aviation fuel-screening criteria. A 500-milliliter synthetic jet fuel sample which met internal screening criteria was submitted to the Air Force Research Laboratory (AFRL) at Wright–Patterson Air Force Base, Dayton, Ohio, for evaluation. The sample was confirmed by AFRL to be in compliance with U.S. Air Force-prescribed alternative aviation fuel initial screening criteria. The results show that this fuel meets or exceeds the key specification parameters for JP-8, a petroleum-based jet fuel widely used by the U.S. military. JP-8 specifications include parameters such as freeze point, density, flash point, and others; all of which were met by the EERC fuel sample. The fuel also exceeds the thermal stability specification of JP-8 fuel as determined by the quartz crystalline microbalance (QCM) test also performed at an independent laboratory as well as AFRL. This means that the EERC fuel looks and acts identically to petroleum-derived jet fuel and can be used interchangeably without any special requirements and thus provides a pathway to energy security to the U.S. military and the entire nation. This subtask was funded through the EERC–DOE Joint Program on Research and Development for Fossil Energy-Related Resources Cooperative Agreement No. DE-FC26- 08NT43291. Nonfederal funding was provided by Accelergy Corporation.

Sharma, Ramesh↗
DOE OSTI2023

A Plan to Qualify New Fuel for the High Flux Isotope Reactor for Material Minimization

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is one of five high power research reactors the Office of Material Management and Minimization (M3) Program, Office of Conversion is working to convert from using highly enriched uranium (HEU) fuel to using low-enriched uranium (LEU) fuel. This effort stems from the primary objective within the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) to achieve permanent threat reduction by minimizing, and when possible, eliminating weapon-usable nuclear material around the world. Under M3’s Office of Conversion, the U.S. High Performance Research Reactor (USHPRR) Project is pursuing fuel qualification and licensing of the high-performance reactors to operate with LEU fuels. All high-performance reactors except HFIR will be converted to LEU monolithic uranium-molybdenum alloy fuel. HFIR will be evaluated for conversion to LEU using a uranium silicide fuel, namely, U3Si2-Al dispersion fuel.The mission of the USHPRR Project is to develop the technology needed to reduce, and eventually eliminate, worldwide use of HEU in civilian applications. The goal is to develop the technical means needed to use low enriched uranium (LEU) instead of HEU fuel in research and test reactors without significant penalties in performance, economics, or safety of the reactors. The USHPRR Project has four major elements, called Pillars: Fuel Qualification (FQ) managed at Idaho National Laboratory (INL), Fuel Fabrication (FF) managed at Pacific Northwest National Laboratory (PNNL), Reactor Conversion (RC) managed at Argonne National Laboratory (Argonne), and Cross-Cutting (CC) managed at Savannah River National Laboratory (SRNL). FQ is responsible for the qualification of the fuel type. RC is responsible for supporting reactor conversion analysis and overseeing licensing submittals leading to conversions of domestic reactors to LEU fuel. For the FQ effort, FQ (INL) worked in collaboration with RC (Argonne) and ORNL to develop the plan for the uranium silicide fuel qualification for HFIR.The resulting HFIR Fuel Qualification Plan provides the general approach for the USHPRR team to move the selected uranium silicide fuel design for HFIR conversion through qualification. Authorization and use in HFIR will be approved through the DOE’s Office of Science. Uranium silicide fuel was previously qualified in NUREG-1313 at an approximate maximum heat flux of 1.4 MW/m2 and a maximum fuel section temperature of about 130°C. In addition to the different regulator process utilized by DOE, these upper limits will be exceeded in HFIR; therefore, further testing will be necessary to ensure the fuel can meet HFIR qualification requirements. The HFIR fuel loading may exceed 4.8 gU/cm3 which was determined in the NUREG-1313 safety evaluation to be acceptable for use in non-power NRC-regulated reactors provided there exist no other safety considerations. In addition, the uranium silicide fuel will need to be qualified in a HFIR-specific design. This plan includes the currently available information from the USHPRR Project Functions and Requirements document and expands these requirements to ensure that planned tests have traceable results providing evidence that the requirements have been met. Data collection methods are discussed as well as the process to show that the requirements have been met. This document is designed to provide a pathway for researchers to obtain data necessary and at the appropriate quality level for HFIR fuel qualification.

Shokes, Tamara↗
DOE OSTI2017

Statistical Analysis in Support of Control Fuel Temperature for the Advanced Gas Reactor Experiments

There are four completed tests among a series of Advanced Gas Reactor (AGR) irradiation experiments that are being conducted within the Advanced Reactor Technologies (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaigns are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and hypothetical accident conditions, and support development and validation of fuel performance and fission product transport models and codes.1 Since this tristructural isotropic fuel is qualified to be used in high temperature gas-cooled reactors (HTGRs), it is essential that the testing conditions of fuel irradiations are designed to mimic those in the conceptual gas reactors. As a result, temperatures of the AGR test fuel have to be maintained within a specified range that is typical of the HTGRs. In each AGR capsule, the test fuel temperature is controlled independently using a variable fraction neon and helium gas mixture that fills the gap volume for each capsule. The temperature as measured by a thermocouple (TC) terminated in the graphite holder (which houses fuel compacts) is used as a feedback for the gas flow controllers. Direct temperature measurement for fuel compacts is not possible because contact between a TC head and fuel particle could lead to undesirable failures. Therefore, an ABAQUS based finite element thermal model was created to predict the daily average fuel temperatures during the entire irradiation period.2,3 It is well-known that extended exposure to neutron irradiation and high temperatures leads to changes in geometries of fuel compacts and graphite holder, which result in changes of the gas gap leading to a change in the relationship between TC and fuel temperatures. In addition, TCs are also expected to drift and sometimes fail under irradiation conditions. These facts make maintaining test fuel temperature within a specified range challenging. This paper describes how the regression analysis methods are used in conjunction of calculated fuel temperatures from early irradiation periods (when irradiation effect is relatively small) to effectively control fuel temperature for the remaining irradiation, when irradiation effect became profound. This is particularly important in the event of all TCs in a capsule fail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Certification of alternative aviation fuels and blend components

Aviation turbine engine fuel specifications are governed by ASTM International, formerly known as the American Society for Testing and Materials (ASTM) International, and the British Ministry of Defence (MOD). ASTM D1655 Standard Specification for Aviation Turbine Fuels and MOD Defence Standard 91-91 are the guiding specifications for this fuel throughout most of the world. Both of these documents rely heavily on the vast amount of experience in production and use of turbine engine fuels from conventional sources, such as crude oil, natural gas condensates, heavy oil, shale oil, and oil sands. Turbine engine fuel derived from these resources and meeting the above specifications has properties that are generally considered acceptable for fuels to be used in turbine engines. Alternative and synthetic fuel components are approved for use to blend with conventional turbine engine fuels after considerable testing. ASTM has established a specification for fuels containing synthesized hydrocarbons under D7566, and the MOD has included additional requirements for fuels containing synthetic components under Annex D of DS91-91. New turbine engine fuel additives and blend components need to be evaluated using ASTM D4054, Standard Practice for Qualification and Approval of New Aviation Turbine Fuels and Fuel Additives. This paper discusses these specifications and testing requirements in light of recent literature claiming that some biomass-derived blend components, which have been used to blend in conventional aviation fuel, meet the requirements for aviation turbine fuels as specified by ASTM and the MOD. The 'Table 1' requirements listed in both D1655 and DS91-91 are predicated on the assumption that the feedstocks used to make fuels meeting these requirements are from approved sources. Recent papers have implied that commercial jet fuel can be blended with renewable components that are not hydrocarbons (such as fatty acid methyl esters). These are not allowed blend components for turbine engine fuels as discussed in this paper.

09 BIOMASS FUELS↗
DOE OSTI2016

Effects of Coolant Void Condition on Melt Fragmentation using Simulants in Initial Phase of Metal-fueled SFR Severe Accident

Metal fuel has inherent safety characteristics, especially in severe accident conditions. However, the metal-fueled SFR severe accident scenarios need to be categorized since overall severe accident mitigation strategies have not been prepared. In addition, these mitigation strategies should be classified in specific phases. In an initial phase of metal-fueled SFR severe accident scenarios, fuel melting, and cladding failures would occur. After the occurrence of the cladding breach, the molten metal fuel would be ejected into a coolant channel due to fission gas in fuel cavity or rapid evaporation of sodium bond. This is called Ex-pin phenomena. There are few studies considering the Ex-pin phenomena of metal fuel. Related representative studies are transient reactor test facility (TREAT) M series tests. The TREAT M series tests provided fundamental metal fuel performance in transient condition. They informed partial experimental data such as the amount of melt ejected into coolant channel after the cladding failure. However, these experimental data did not examine the melt behavior during the Ex-pin phenomena. In metal-fueled SFR, it is a known fact that the negative reactivity feedback occurs in the initial phase of severe accident. In an unprotected loss-of-flow (ULOF) accident, the void-reactivity feedback may be positive depending on where in the reactor core the sodium voiding occurs. However, molten metal fuel would be better dispersed upward compared to those of an oxide fuel and significant negative reactivity feedback could be induced through such behaviors. Thus, a key safety strategy to mitigate the accident progression at the initial phase of severe accident is to discharge melt upwards out of the core. In this paper, the upward discharge of melt is defined as melt levitation. Such melt levitation could ensure early termination of severe accident. There could be different melting behaviors inside the pin depending on reactor types or the type of fuel. In case of pressurized water reactor (PWR) oxide fuel-water system, the rapid power excursion would initiate melting of fuel at the surface of the pellets. In this paper, the ULOF accident for the metal-fueled SFR is considered as an accident scenario. To induce the melt levitation, the driving force should be clarified. Heo et al. observed a built-up vapor pressure was suggested as one of the driving forces for the upward dispersion of the molten materials. The following factors also have the potential to induce the melt levitation; the pressure induced by fission gas in the cavity or rapid boiling of sodium bond, natural circulation flow of the sodium, the sodium vapor flow generated by the difference between inlet and outlet plenum in the reactor core, the built-up vapor pressure due to the fuel-coolant interaction etc. These factors are balanced each other. They should be examined through the Ex-pin experiments. Before conducting a parametric study for each factor, the general behavior of the molten metal fuel should be investigated in the coolant channel. Since rapid sodium coolant boiling could occur in the initial phase of severe accident scenarios, the molten fuel would be ejected and fragmented in voided channel. Unlike previous studies concerning an un-voided channel, differences in melt behavior in the voided channel are expected. In present study, molten material behavior was evaluated to reveal their characteristics in voided channel. The Ex-pin experiments were performed in both un-voided and voided channel, thus such behavior differences were explicitly observed. In the lab-scale experiments, the un-voided and voided channel were reflected by coolant-filled and air-filled channel, respectively. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Evaluation of PBR Spent Fuel Criticality and Dose Rate Compliance for Storage and Transportation

Spent tri-structural isotropic (TRISO)–based fuels have a strong track record in storage and transportation without documented incidents. This work seeks to reduce uncertainty to aid in more informed spent fuel management of TRISO-based fuels by modeling both fresh and spent pebble bed reactor (PBR) fuel and comparing the results to the regulatory standards from 10 CFR 71. SCALE was used for all modeling due to it having fast and accurate methods for handling PBR fuel modeling, as well as having an efficient method for shielding calculations in monaco with automated variance reduction using importance calculations (MAVRIC), which utilizes the consistent adjoint-driven importance sampling (CADIS) and the forward-weighted consistent adjoint-driven importance sampling (FW-CADIS) methods. KENO-VI was used for all criticality calculations, TSUNAMI was used for uncertainty quantification on k-effective, TRITON and the Oak Ridge isotope generation code (ORIGEN) were both used for depletion of the fuel, and MAVRIC was used for shielding calculations. For criticality assessments, this study focused on the requirement that the value of the neutron multiplication factor, k-effective (k-eff), would not exceed a peak value of 0.95, including uncertainty, with 95% confidence. Criticality was initially examined by modeling fresh fuel from three different designs—HTR-10 fuel, PBMR-400 fuel, and demonstration fuel representative of a TRISO-fueled modern high-temperature gas reactor (HTGR) design, henceforth referred to as Demo HTGR—and placing them into various sized containers with conditions described in 10 CFR 71 to quantify the peak k-eff state. When the peak value of 0.95 k-eff was exceeded, mitigation methods were examined in those scenarios. Burnup credit, pebble displacement in areas of strong neutron multiplication, and random pebble replacement using pebbles of various compositions and replacement fractions were examined. In summary, the criticality of PBR fuels can be well accounted for by restricting container size, taking credit for burnup, or by displacing/replacing pebbles. Uncertainty of the k-eff due to nuclear data uncertainties was recorded at ~0.6644%Δk/k, or roughly 664% mil (pcm). The nuclear data–induced uncertainty was relatively small and should not require significant modification in the design to be accounted for. Revisions to the evaluated nuclear data file values have been shown to have a larger impact than nuclear data–induced uncertainty. For dose rate aspects, U.S. Nuclear Regulatory Commission regulations require a maximum dose rate of 10 millirem per hour (mrem/h) at 2 meters. In examining the dose rate behavior of spent PBR fuel, the representative Demo HTGR fuel was modeled exclusively due to it possessing the highest target burnup of the examined fuels. Equilibrium cycle modeling methods were used to produce a higher-fidelity discharge isotopic composition than simple assumptions, such as reflected pebbles. The discharge composition was used as a source term in the fixed-source transport shielding calculations, and dose rates were calculated at 2 m for the shortest possible cooling time. The low concentration of fuel material led to dose rates that were in line with regulatory limits, despite the high burnup when compared to traditional light water reactor fuels. In conclusion, the methods employed in this study would require more work to further verify and validate and are limited to the criticality and dose rate analyses performed.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Overview of the SAS4A Metallic Fuel Models and Extended Analysis of a Postulated Severe Accident in the Prototype Gen-IV Sodium-Cooled Fast Reactor

The SAS4A safety analysis code, originally developed for the analysis of postulated severe accidents in oxide fuel sodium-cooled fast reactors (SFRs), has been significantly extended to allow the mechanistic analysis of severe accidents in metallic fuel SFRs. The SAS4A metallic fuel models simulate the metallic fuel thermomechanical and chemical behavior and track the evolution and relocation of multiple fuel and cladding components during the pretransient irradiation and during the postulated accident, allowing an accurate description of the changes in the local fuel composition. The local fuel composition determines the fuel thermophysical properties, such as freezing and melting temperatures, which in turn affect the fuel relocation behavior and ultimately the core reactivity and power history during the postulated accidents. Models describing the fuel-cladding interaction and eutectic formation, the effects of the in-pin sodium on the in-pin fuel relocation, and the postfailure reentry of the molten fuel and fission gas from the pin plenum have also been added. This paper provides an overview of the SAS4A key metallic fuel models emphasizing the postfailure metallic fuel relocation models included in the LEVITATE-M module of SAS4A. The capabilities of the SAS4A metallic fuel models are illustrated through an extended SAS4A analysis of a postulated unprotected loss-of-flow and transient-overpower accident in the metallic fuel prototype Gen-IV sodium fast reactor. The results show that the maximum relative power reached during the postulated accident is 1.19 P 0 . The favorable characteristics of the metallic fuel cause a significant decrease in net reactivity and relative power due to prefailure in-pin fuel relocation. Furthermore, negative net reactivity values persist after cladding failure, and the postfailure fuel relocation events occur at low and decreasing power levels.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Strategy for Used Fuel Acquisition

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The Storage and Transportation staffs within the UFDC are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While both wet and dry storage have been shown to be safe options for storing UNF, the focus of the program is on dry storage at reactor or centralized locations. Because limited information is available on the properties of high burnup fuel (exceeding 45 gigawatt-days per metric tonne of uranium [GWd/MTU]), and because much of the fuel currently discharged from today’s reactors exceeds this burnup threshold, a particular emphasis of this program is on high burnup fuels. Since high burnup used fuels have only been loaded into dry storage systems in the past decade or so, these materials are available to the UFDC for testing in only very limited quantities. Much of what is available has come via NRC testing programs. Some of these fuels may have achieved "high burnup," but that does not mean they were designed for high burnup use (e.g. lower enrichments, smaller plenum spaces, extra reactor cycles). The handling and transfer of these materials from utility to laboratory has not always been prototypical of how used nuclear fuel is prepared for dry storage; these fuels are not subjected to the same vacuum drying conditions that can lead to changes in hydride morphology that will affect the mechanical properties of the fuel. It is recognized that sources of used high burnup fuel that can be handled in a manner consistent with how fuel is readied for dry storage is essential to the mission of the UFDC. This report documents what types of fuel are of interest to the campaign, and how those fuels could be acquired and shipped to the Idaho National Laboratory (INL) for incorporation into the campaign R&D mission. It also identifies any gaps in INL capabilities that might preclude working with one fuel type or another.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Feasibility study for detection of fuel assemblies state inside sealed dry storage casks using external gamma flux measurements - 25445

The license for the storage of spent nuclear fuel assemblies in dry casks is limited to 40 years. In Germany, a site for a final repository of spent fuel is not yet available. Therefore, this license has to be renewed for prolonged storage period. Currently, there is no experience with dry cask storage worldwide that goes beyond a period of 40 years. Beside regulatory and security aspects, there are questions concerning the long-term integrity of the spent fuel assemblies as this is of relevance for final transportation and reloading to final waste repository casks. Once the cask is filled and sealed, the knowledge about the state of the fuel assemblies is limited. This study investigates the feasibility of using external gamma flux measurements to detect the state of the spent fuel assemblies inside the cask. Monte-Carlo simulations were performed to evaluate the gamma flux distribution outside the sidewall of a cask with all intact fuel assemblies and a cask with one damaged fuel assembly, mimicking a fuel assembly with expanded rods and fuel relocation. The evaluation was performed for different location of the damaged fuel assembly and different storage times. The results of the investigation showed that the intensity and the shape of the gamma flux outside the sidewall of the cask are sufficient to identify the damaged fuel assembly, in case it is located near the inner wall of the cask. The detection of the inner damaged fuel assemblies, however, is less feasible. Due to the self-shielding of the fuel assemblies, the gamma photons which come from the inner fuel assemblies stopped within the fuel basket and assemblies materials. The gamma photons which do survive the fuel assemblies self-shielding reach the inner cask wall with not enough energy to escape the shielding of the thick cast iron wall. In general, it can be concluded that the more the damaged fuel assembly is close to the cask inner walls and that the damage is in a large scale (i.e., significant fuel relocation), the more it can be detected by external gamma flux measurements. (authors)

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

Development and demonstration of a methodology to evaluate high burnup fuel susceptibility to pulverization under a loss of coolant transient

For economic reasons, the US nuclear industry is renewing efforts to build a technical basis to extend rod average burnup limits above the current regulatory burnup limit of 62 GWd/MTU. The primary driver is to increase pressurized water reactor cycle lengths to 24 months, reducing the number of fresh fuel assemblies and core design constraints, thereby making core energy utilization more efficient. However, fuel pellet fragmentation and pulverization, termed high burnup fuel fragmentation (HBFF), has been observed in the high burnup (>90 GWd/MTU) Halden loss-of-coolant-accident (LOCA) integral test series. The issue gained attention when fuel fragmentation and pulverization were also observed closer to the current US regulatory limit during the US Nuclear Regulatory Commission (NRC) sponsored out-of-core integral test at Studsvik Nuclear in early 2011. This led to NRC concerns with potential changes to fuel and core designs relative to fuel pellet pulverization. In a letter to the NRC Commissioners, the staff specifically identified a need to “…define the boundary of safe operation for key fuel design and operating parameters,” stating that “the staff is challenged to evaluate the acceptability of future fuel design advancements and fuel utilization changes.” As such, it can be concluded that HBFF and potential dispersal into the reactor coolant system introduces additional complications in light-water reactor (LWR) fuel safety evaluations. However, it is not clear how much fuel will be susceptible to HBFF; nor has there been a methodology developed to evaluate fuel susceptibility to HBFF. To that end, this paper proposes an analysis methodology to assess fuel susceptibility to HBFF during LOCA scenarios. The work presented here uses the BISON fuel performance code to evaluate a representative pressurized water reactor fuel rod exposed to a rod average burnup of 75 GWd/MTU. Sensitivity studies investigated the impact of the peak cladding temperature, transient fission gas released, and pre-transient fission gas release on cladding ballooning and burst timing. Subsequently, a methodology to assess fuel susceptibility to HBFF will be developed based on experimental data published in the open literature. The methodology will then be demonstrated by calculating the mass of fuel susceptibility to HBFF. The BISON results conclude that increasing peak cladding temperature drastically decreased time to failure, and decreased balloon size both of which have been confirmed experimentally. Additionally, the effect of pre-transient and transient fission gas release affected cladding balloon size and burst timing. Finally, fuel susceptibility to HBFF significantly decreased as a function of peak cladding temperature.

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DOE OSTI2019

Thermal Transport and Fracture Behavior of Sintered Fuel Pellets: Experimental Validation of NEAMS Tool MARMOT

The project targets at experimental validation of NEAMS tool MARMOT for predicting thermal transport and fuel fracture. The key components of the project include: (1) sintering UO 2 samples with well-controlled microstructures (grain size, pore, fission bubble, porosity and pore distribution either in intra- or inter-granular pores); (2) thermal conductivity measurements of these wellcharacterized polycrystalline samples; (3) indentation testing of sintered UO 2 with various grain size, porosity and stoichiometry to obtain fracture and crack propagation mechanisms (inter or intragranular fracture); and (4) validation and uncertainty quantification of the MARMOT thermal transport and fracture models. The high quality experimental data will be used to validate the predictions of MARMOT for thermal transport and fracture, using average data and local microstructure information. In this work, UO 2 fuels with controlled microstructure (grain size, pore size/structure and fuel chemistry) were successfully prepared by high energy ball milling of the hyper-stoichiometric UO 2+x powders purchased commercially, and consolidated by spark plasma sintering. Dense nano-sized UO 2 fuel pellets with controlled stoichiometry were sintered by SPS. The effects of different SPS sintering parameters on the pore size and distribution were investigated, and the correlation among the SPS parameter – microstructure control – stoichiometry/defect structure was established. To further control the pore structure and fuel chemistry, systematic thermal annealing of the sintered fuels with the controlled pore structure was performed in a reduced. The controlled microstructure and fuel chemistry enable the investigation of the separate effects of pore structure and fuel chemistry on thermal transport behavior. The mechanical properties of the sintered UO 2 fuels were tested by nanoindentation and microindentation testing at different temperatures. The fracture behavior of the sintered fuels as functions of grain structure and temperature was investigated. Nanocrystalline UO 2 display higher hardness than microcrystalline counterpart, consistent with the Hall-Petch strengthening mechanism. Greater Young’s modulus and fracture toughness are also identified for the nanocrystalline UO 2 , and hardness and Young’s modulus decrease with temperature, suggesting better ductility of oxide fuels at greater temperature at small length scale. Hyper-stoichiometric UO 2 specimen displays higher hardness and fracture toughness than stoichiometric UO 2 , due to the impediment of the crack propagation by the oxygen interstitial atoms. These results are useful in understanding the mechanical properties of the high burn-up structure (HBS) formed in nuclear fuels during reactor operation, and also provide critical experimental data as the input for the development and validation of the MARMOT fracture model of nuclear fuels. The thermal properties of the sintered UO 2 fuel pellets with controlled grain structure, fuel chemistry and pore structures are measured by laser flash and the impact of the microstructure characteristics on the thermal transport properties is elucidated. Through a close collaboration with the MARMOT team at Idaho National Laboratory, the thermal properties of the sintered oxide fuel pellets with controlled microstructures are measured experimentally, and modeled using the developed MARMOT thermal transport models for a joint effort of demonstrating high spatial resolution measurements at the microstructural level on as-fabricated nuclear fuels and materials and obtain critical experimental data for model validation and properties predictions. In this work, we also conducted sensitivity analysis and uncertainty quantification (UQ) on a mesoscale simulation that has been applied to understand the effective thermal conductivity of UO 2 reactor fuel using the MARMOT tool. We also conduced UQ on mechanistic macroscale models that have been developed to predict the effective thermal conductivity. The most sensitive parameter is the thermal conductivity of bulk UO 2 for the thermal conductivity models. In the future, the predicted simulation distributions need to be compared to experimental data for validation of the models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Post-Irradiation Examination for Conventional and Advanced Nuclear Fuel

During irradiation in a nuclear reactor, the fuel pellet experiences non-uniform fission rate and strong thermomechanical loads, which result in alterations and restructuring occurring non-homogeneously in the fuel volume. The irradiated fuel is characterized by a highly heterogeneous structure, with a correspondingly heterogeneous distribution of phases and properties. It is a big technical challenge for a hot cell facility to be able to cover effectively a broad variety of fuel types and compounds, characterized by different compositions, physico-chemical properties, geometries and configurations. The first challenge is to provide testing environments and procedures which do not alter or degrade the original irradiated fuel properties and are technically suitable for hot cell or shielded facility operation. For instance, non-oxide fuel compounds require protective atmosphere to avoid sample oxidation during storage, preparation and testing. Moreover, suitable adaptations of tools and measuring devices must be available to characterize fuel configurations which differ from the conventional cylindrical pellet-cladding geometries (e.g. spherical fuel elements for pebble bed HTR and TRISO coated fuel particles). Finally, adequate scientific methods of investigation must be available to measure properties and study the behaviour of highly heterogeneous irradiated fuel. In addition to basic techniques for non-destructive and destructive examination of nuclear fuel systems (e.g. covering fuel and cladding or fuel and coatings, or other configurations), 'in-depth' methods are applied to determine physical, thermomechanical and micro-analytical properties of specific phases/features present in irradiated fuel. This paper presents relevant trends and examples from fuel safety studies performed at JRC-Karlsruhe and in other laboratories which highlight the 'added value' obtainable by applying combined techniques, high resolution tools or 3D methods of analysis for the experimental 'multiscale' characterization of irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Transient performance of accident tolerant fuel in design basis accidents

A set of transient analysis has been performed to assess the performance of candidate accident tolerant fuel (ATF) against reference UO{sub 2} fuel with Zircaloy cladding under design basis accident (DBA) conditions. This study examined two ATF options, U{sub 3}Si{sub 2} fuel with SiC cladding and UO{sub 2} fuel with APMT cladding. The accident analyses were conducted using the system code TRACE for a three-loop PWR plant. PARCS standalone full-core calculations were done for each fuel/cladding combination to provide inputs to the TRACE point-kinetics model. The design basis accidents analyzed were a reactivity insertion accident (RIA) and a large break loss-of-coolant accident (LBLOCA). The RIA was simulated by the insertion of a reactivity ramp of $1 (one dollar) in 0.1 s at hot full power (HFP) condition. The LBLOCA was a simulation of a double-ended pipe break in one of the cold legs of the primary loops. Differences in thermal responses of the fuel and the cladding are attributed to their thermal properties. The U{sub 3}Si{sub 2} fuel with its lower volumetric heat capacity and higher thermal conductivity (i.e. higher thermal diffusivity) generally results in a lower fuel temperature than the UO{sub 2} fuel and a correspondingly faster thermal response. Additional calculations for the U{sub 3}Si{sub 2} fuel shows that in a RIA the cladding material, SiC versus Zr, has little impact on the maximum fuel centerline temperature and the PCT (peak cladding temperature). An increase in the hot rod power factor is shown to raise the centerline fuel temperature in a RIA but has little impact on the PCT. The analysis results show that for a relatively fast decay heat driven transient, such as the LBLOCA, the thermal-hydraulics and fuel/cladding thermal properties dictate the overall response more so than the reactivity feedback. The results show that the PCTs in a LBLOCA are similar for all fuel/cladding combinations considered in this study. The LBLOCA analysis also notes the impact of the asymmetry in the system response. The fuel rods in the sector of the core connected to the hot leg of the broken loop (Sector 1 in the PWR model) is shown to have a much lower PCT than fuel rods in the other five sectors of the core. Results of this study indicate that for the two DBAs considered (RIA and LBLOCA) the transient behavior and the resulting peak cladding temperature of the two ATF options (U{sub 3}Si{sub 2} fuel with SiC cladding and UO{sub 2} fuel with APMT cladding) are similar to the standard light-water reactor fuel (UO{sub 2} fuel with Zircaloy cladding). (authors)

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