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DOE OSTI2015

Applying fast calorimetry analysis on spent fuel calorimeter - 15620

Swedish Nuclear Fuel and Waste Management Company (SKB) is responsible for the back-end of the Swedish nuclear industry. This includes transport of spent fuel by boat from the power plants to the Central Interim Storage Facility for Spent Fuel (Clab), and in the future, encapsulation of the spent fuel and further transportation to a geological repository. In order to ensure safety for the planned geological repository, the decay heat has to be well known. For this purpose the SKB has performed calorimetric measurements on over 200 pressurized water reactor (PWR) and boiling water reactor (BWR) assemblies since 2003. The results from earlier analyses have been used to improve computer codes predicting heat as well as to acquire deeper knowledge about the fuel inventory. The work complements research being done as part of the Next Generation of Safeguards Initiative - Spent Fuel project, whose purpose is to develop and test nondestructive assay (NDA) technologies to improve safeguards measurements of spent nuclear fuel. The project is working to accomplish multiple goals, one of which is to determine the heat content of spent fuel assemblies. To shorten the measurement time, the temperature increase method was used instead of the equilibrium temperature method. In the temperature increase method, the temperature inside of the calorimeter is lowered below the surrounding water temperature before the measurement, and the temperature is allowed to increase due to the higher surrounding water temperature during the measurement. The rate of the temperature increase when the temperature is the same as the surrounding temperature is then compared to a calibration curve where an electric heater has been used in the same way. The price for the higher throughput is less accuracy and dependency of fuel geometry since the heat flow will be different for different fuel types, especially between PWR and BWR assemblies. Recently at Los Alamos National Laboratory, more sophisticated analysis methods have been considered for the use of calorimetry for treaty verification. These methods aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy and eliminate the dependency on fuel geometry. Even though the techniques show promise for future calorimetry of spent fuel, certain design features of the current system have been identified to interfere with the accuracy of equilibrium measurements in general such as changing pool temperatures and heat leakage during calibration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Mass Inventory Balance and Fuel Breeding Analysis on Transuranic Recycling Option of FBR

Maintaining sustainability of nuclear fuel can be maintained by utilizing new fresh fuel resources, utilization of spent fuel resource, and obtaining fuel breeding resource in the reactors. Utilization of used fuel from recycling resource and fuel breeding resource becomes the key issue which affects to the reprocessing and refueling facilities. Some minor actinides (MA) have been developed to be recycled as well as some different compositions of plutonium for some purposes such as converted fuel for extending longer reactor operation, burning and transmutation of MA, fuel breeding of FBR as well as for proliferation resistance purposes. This evaluation intends to investigate the mass inventory balance focused on plutonium (Pu) and minor actinide (MA) inventory and reactor behavior in relation to reactor criticality, excess reactivity and fuel breeding for several fuel types. Time dependence as well as irradiation process during reactor operation gives the significant impact for changing the behavior of fuel composition which affects to the behavior of reactor itself. Reactor operation time effect on this analysis has been analyzed from the beginning of cycle up to the equilibrium condition of each fuel type compositions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Steady-State Subchannel Thermalhydraulic Assessment of Thorium-Based Fuel Concepts for Use in Pressure Tube Heavy Water Reactors

Thorium is an abundant alternative fertile nuclear fuel that could potentially be implemented in pressure tube heavy water reactors (PT-HWR) in the near future. One of the key stages in developing the thorium-based fuels for PT-HWRs is the reactor core concept, which requires the collaboration of different disciplines, such as neutron physics, thermal-hydraulics, materials, chemistry, safety, and others. Currently there are numerous core concepts under investigation at the Canadian Nuclear Laboratories that involve several types of thorium-based fuels, and that could be implemented in a 700-MWe-class PT-HWR. Four fuel bundle concepts are considered, including i) the reference (or nominal) 37-element bundle; ii) a 37-element modified bundle, with the center element using a different fuel material; iii) a 35-element bundle design, and iv) an 18-element internally cooled annular fuel (ICAF-18) fuel bundle. This study presents preliminary results of the steady-state thermalhydraulic subchannel analyses of the hot channels of different PT-HWR core concepts using thorium-augmented or thorium-based fuels. A follow up study will be initiated to improve the assessment of the proposed cores and associated fuel bundle concepts for thorium-augmented and thorium-based fuels. The computational model will be revisited to include; - Expanding the analysis to cover different power profiles at different burnup stages of the bundle. - Expand the analysis to cover 28.5 kg/s. - Reduce the eccentricity of the 35-element fuel bundle. - Assess the effect of the phenomena not yet included in the subchannel model (such as the effect of cold wall, and gamma heat). - Expand the analysis to include the effect of spacers and junctions. - Compare bundle concepts using other performance indicators (such as burnup, breeding ratio), as optimizing the fuel bundle geometry is a multidisciplinary problem.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Correlation of rocket propulsion fuel properties with chemical composition using comprehensive two-dimensional gas chromatography with time-of-flight mass spectrometry followed by partial least squares regression analysis

There is an increased need to more fully assess and control the composition of kerosene based rocket propulsion fuels, namely RP-1 and RP-2. In particular, it is crucial to be able to make better quantitative connections between the following three attributes: (a) fuel performance, (b) fuel properties (flash point, density, kinematic viscosity, net heat of combustion, hydrogen content, etc) and (c) the chemical composition of a given fuel (i.e., specific chemical compounds and compound classes present as a result of feedstock blending and processing). Indeed, recent efforts in predicting fuel performance through modeling put greater emphasis on detailed and accurate fuel properties and fuel compositional information. In this regard, advanced distillation curve (ADC) metrology provides improved data relative to classical boiling point and volatility curve techniques. Using ADC metrology, data obtained from RP-1 and RP-2 fuels provides compositional variation information that is directly relevant to predictive modeling of fuel performance. Often, in such studies, one-dimensional gas chromatography (GC) combined with mass spectrometry (MS) is typically employed to provide chemical composition information. Building on approaches using GC-MS, but to glean substantially more chemical composition information from these complex fuels, we have recently studied the use of comprehensive two dimensional gas chromatography combined with time-of-flight mass spectrometry (GC × GC - TOFMS) to provide chemical composition data that is significantly richer than that provided by GC-MS methods. In this report, by applying multivariate data analysis techniques, referred to as chemometrics, we are able to readily model (correlate) the chemical compositional information from RP-1 and RP-2 fuels provided using GC × GC - TOFMS, to the fuel property information such as that provided by the ADC method and other specification properties. We anticipate that this new chemical analysis strategy will have broad implications for the development of high fidelity composition-property models, leading to an optimized approach to fuel formulation and specification for advanced engine cycles.

fuel↗

DOE OSTI2017

Comparison of the greenhouse gas and criteria pollutant emissions from the SF-BREEZE high-speed fuel-cell ferry with a diesel ferry

A theoretical comparison is made of the “well to waves” (WTW) greenhouse gas (GHG) and criteria pollutant emissions from the SF-BREEZE high-speed hydrogen PEM fuel cell ferry and the VALLEJO ferry powered by traditional diesel engine technology but constrained to Tier 4 emissions standards. The emissions were calculated for a common maritime mission, the current ferry route between Vallejo CA and San Francisco CA. Calculations are made of the energy required for the SF-BREEZE and VALLEJO to perform the mission route profile. The SF-BREEZE requires 10.1% more fuel energy than the VALLEJO, primarily due to the SF-BREEZE being heavier. Estimates are made for the SF-BREEZE GHG emissions associated with five LH2 fuel production pathways including renewable and non-renewable (fossil-fuel based) methods. Estimates are also made for GHG emissions associated with fossil-diesel production and delivery as well as those for biodiesel, which can be considered a renewable “drop-in” fuel replacement for conventional diesel fuel. We find that the GHG emissions for the SF-BREEZE using non-renewable LH 2 are significantly higher than for the Tier 4 diesel-fueled VALLEJO on a per passenger basis. However, using renewable LH 2 , the GHG emissions for the SF-BREEZE ferry are reduced 75.8% compared to the diesel-fueled VALLEJO operating at Tier 4 emissions standards. We also compare the criteria pollutant emissions (NO x , HC, PM 10 ) for the SF-BREEZE to that of the VALLEJO held to Tier 4 emissions standards fueled by diesel fuel or biodiesel. Hydrogen PEM fuel cell technology dramatically reduces NO x and HC emissions below the most advanced Tier 4 criteria pollutant emissions requirements regardless of whether the LH 2 is made by NG reforming or via water electrolysis using 70% renewable energy. Renewable LH 2 made with greater than 84% renewable process energy is needed to also drop the SF-BREEZE PM 10 emissions below that of Tier 4 for high-speed fuel cell ferry transportation. Overall, the results show that operating a hydrogen fuel cell ferry on nearly 100% renewable hydrogen provides the dramatic reduction in GHG and criteria pollutant emissions commensurate with the problems of global climate change and maritime air pollution worldwide.

33 ADVANCED PROPULSION SYSTEMS↗

DOE OSTI2020

SENSITIVITY ANALYSIS OF FUEL PHYSICAL PROPERTY EFFECTS ON SPARK IGNITION ENGINE PERFORMANCE

Alternative fuels are of interest to automakers and regulators due to their potential to reduce net greenhouse gas emissions from transportation sources. Alternative fuels also have fuel properties which may enable advanced combustion modes with higher engine thermal efficiencies. There has been previous work to identify the relationship between various fuel properties and engine performance, but most of this work has been experiments or simulations where the change in properties was obtained through changing the fuel composition, making isolating the effects of individual fuel properties difficult. In this study, numerical simulations have been used to investigate the effects of individual fuel physical properties such as viscosity or heat of vaporization (HoV) on engine performance. Simulations have been performed of two different engine platforms, the first an optical, single-cylinder research engine and the second a multi-cylinder production engine. Both engines are direct-injection spark-ignition engines with pent-roof heads and are designed for automotive applications. Each engine was run at a different operating condition, one stable and one knock-limited. Different base fuels provided a variety of simulated conditions. Up to six different fuel properties were varied as part of Global Sensitivity Analyses performed for each of the engines with multiple performance targets including thermal efficiency, combustion efficiency and combustion phasing. Results show trends that are largely consistent with previous experimental findings using multiple fuels. The engine thermal efficiency was primarily sensitive to the fuel's HoV, with other fuel physical properties having smaller effects. For optical engine results, the magnitude of the effect was greater in this study than expected based on previous experimental results were many fuel physical and chemical properties were varied simultaneously. However, for the multi-cylinder production engine, the relationship between thermal efficiency and HoV was slightly smaller.

engine simulation↗

DOE OSTI2016

THE APPLICATION OF MAMMOTH FOR A DETAILED TIGHTLY COUPLED FUEL PIN SIMULATION WITH A STATION BLACKOUT

Accurate calculation of desired quantities to predict fuel behavior requires the solution of interlinked equations representing different physics. Traditional fuels performance codes often rely on internal empirical models for the pin power density and a simplified boundary condition on the cladding edge. These simplifications are performed because of the difficulty of coupling applications or codes on differing domains and mapping the required data. To demonstrate an approach closer to first principles, the neutronics application Rattlesnake and the thermal hydraulics application RELAP-7 were coupled to the fuels performance application BISON under the master application MAMMOTH. A single fuel pin was modeled based on the dimensions of a Westinghouse 17x17 fuel rod. The simulation consisted of a depletion period of 1343 days, roughly equal to three full operating cycles, followed by a station blackout (SBO) event. The fuel rod was depleted for 1343 days for a near constant total power loading of 65.81 kW. After 1343 days the fission power was reduced to zero (simulating a reactor shut-down). Decay heat calculations provided the time-varying energy source after this time. For this problem, Rattlesnake, BISON, and RELAP-7 are coupled under MAMMOTH in a split operator approach. Each system solves its physics on a separate mesh and, for RELAP-7 and BISON, on only a subset of the full problem domain. Rattlesnake solves the neutronics over the whole domain that includes the fuel, cladding, gaps, water, and top and bottom rod holders. Here BISON is applied to the fuel and cladding with a 2D axi-symmetric domain, and RELAP-7 is applied to the flow of the circular outer water channel with a set of 1D flow equations. The mesh on the Rattlesnake side can either be 3D (for low order transport) or 2D (for diffusion). BISON has a matching ring structure mesh for the fuel so both the power density and local burn up are copied accurately from Rattlesnake. At each depletion time step, Rattlesnake calculates a power density, fission density rate, burn-up distribution and fast flux based on the current water density and fuel temperature. These are then mapped to the BISON mesh for a fuels performance solve. BISON calculates the fuel temperature and cladding surface temperature based upon the current power density and bulk fluid temperature. RELAP-7 then calculates the fluid temperature, water density fraction and water phase velocity based upon the cladding surface temperature. The fuel temperature and the fluid density are then passed back to Rattlesnake for another neutronics calculation. Six Picard or fixed-point style iterations are preformed in this manner to obtain consistent tightly coupled and stable results. For this paper a set of results from the detailed calculation are provided for both during depletion and the SBO event. We demonstrate that a detailed calculation closer to first principles can be done under MAMMOTH between different applications on differing domains.

36 MATERIALS SCIENCE↗

DOE OSTI2022

AN EXPERIMENTAL STUDY OF FUEL SELECTION FOR A GASOLINE MULTI-MODE

In this study of fuels of different reactivities and their performance in a low-temperature combustion (LTC) engine. The engine was a 2.2L CRDI engine code D4HB provided by industry partner Hyundai America Technical Center inc. (HATCI). The engine was instrumented with in-cylinder pressure sensors allowing for monitoring of the combustion process. First, the engine was operated with standard US pump grade gasoline, Research Octane number (RON) 91 E10, to find initial operation conditions as well as control points of stable operation for daily checks on system conditions. Tested were 8 different fuel blends, the four base blends were provided by Phillips 66 (P66) with a RON value ranging from 60 to 91, and three other variations of these fuels were formulated by splash blending alcohols, iso-butanol, and ethanol, with the RON60 fuel and the final blend, was US RON91 E10 pump gasoline. Using these 8 fuels, the low load performance was analyzed allowing a selection of one of the fuels to be made to proceed with experimentation. To study the low load performance of each fuel, the load was decreased until combustion became too unstable to continue. This was done at 1500 rpm with all of the fuels. At 1200 rpm, the same methodology was applied, though all of the fuels were not tested, only the fuels which showed promise at 1500 rpm. Meaning that the lowest targeted load was achieved without any aid to the combustion process other than advancing combustion, increasing intake air temperature (IAT) for example was not done to maintain consistency across all fuels. The same methodology was applied at 800 rpm based upon the 1200 rpm results to see the ability of the engine to idle with the given fuel. It was found that fuel with a RON of 80 showed good low load performance while allowing for high load performance to be maintained. A blend of 25% Iso-Butanol and 75% RON 60 equated to a RON of 80 and showed good low load performance and showed peak indicated thermal efficiency (ITE) numbers of 46% at 17 bar indicated mean effective pressure (IMEP) and 2200 rpm.

02 PETROLEUM↗

DOE OSTI2024

Concept of Operations for Advanced Reactor Spent Nuclear Fuel Management

This presentation presents a preliminary description of a concept of operations to incorporate advanced-reactor spent nuclear fuel (SNF) into an integrated waste-management system (IWMS). The evaluation includes SNF from four advanced-reactor concepts with the following fuel types: (1) small modular reactors using oxide fuels, (2) tristructural-isotropic (TRISO) fuels, (3) metallic fuels, and (4) fuel salts. To provide context for the proposed concept of operations for advanced reactors, a comparison is made with traditional light-water reactors (LWRs) to identify potential gaps in the IWMS. The technical differences between advanced reactors and LWRs are assessed to determine the feasibility of managing advanced-reactor waste streams using existing operations and technology. This presentation emphasizes fuel types from Advanced Reactor Demonstration Program reactors: Xe 100, which uses TRISO fuel and Natrium, with its metallic fuels while also analyzing management options for molten-salt reactors and advanced light-water reactors (ALWRs). Understanding the storage, transportation, and disposal requirements of SNF is dependent on both the quantity and characteristics of the SNF generated by nuclear reactors. This presentation provides a high level overview, comparing the anticipated concept of operations for different SNFs from advanced reactors. The IWMS includes at-reactor storage, transportation, potential off-site storage, potential treatment, and disposition. To assess the potential effect of advanced-reactor concept of operations on the IWMS, estimates were made for fuel characteristics that contribute to storage, transportation, disposal, and possible treatment of advanced-reactor SNF. These include canister heat load, dose rates, and criticality-control limits, which are important for determining the condition and configuration of the advanced reactor SNF. At-reactor storage of LWR SNF traditionally involves a spent-fuel pool (SFP) before transfer to an independent spent-fuel storage installation. However, some advanced-reactor concepts, particularly those using TRISO and salt fuels, do not anticipate the use of an SFP. This difference in at-reactor storage could impact the IWMS. Additionally, transportation of advanced-reactor SNF may include additional processes tied to potential off gassing, and transportation of microreactor SNF may occur within the reactor vessel. Some advanced-reactor SNF could also undergo treatment to meet requirements of an acceptable waste form for disposition, and the treatment location will be a major contributor to efficiently performing IWMS responsibilities. Moreover, the quantity of SNF generated is an important consideration for IWMS because it could affect the size of the transportation fleet and potential off-site storage requirements. Additionally, volume and heat load are the primary drivers for SNF disposition. This presentation compares potential packaging options for advanced reactor SNF with traditional LWR packaging to provide a high-level comparison for the needs of the IWMS for advanced-reactor SNF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Postirradiation examination of recently irradiated metallic fuel concepts

In fast neutron spectrum nuclear reactors, metallic uranium alloys have often been chosen because of their high uranium density, high thermal conductivity, and several reactor kinetic safety benefits. Metallic uranium is often alloyed to raise the melting temperature or to modify the microstructure which improves irradiation performance. Recent irradiations have explored additional alloys and geometric forms beyond what has historically been irradiated (U-10 weight % Zr, 75% smeared density, sodium bonded fuel). Postirradiation examination results and how these results can be incorporated into modelling and simulation will be presented. The evaluated fuels fall into four categories alternative alloying metals, additives, sodium bond removal, and alternative smear densities. For alternative alloys, Zr has been substituted for 10% Mo or 5% Mo, 4.3%Ti, 0.7%Zr to stabilize the cubic phase of U and prevent constituent redistribution seen in U-Zr alloys. Unfortunately these alloys have significant, undesirable interactions with cladding during irradiation. Additives, such as Pd, have been suggested to bind lanthanide fission products, which interact with cladding limiting fuel performance, into high temperature intermetallic compounds. Irradiation results indicate there may be targeted applications such as high burnup fuel or recycled fuel with lanthanide carry-over where additives are useful. Sodium bonding is considered undesirable at the back end of the fuel cycle where sodium bonded fuel must be treated to remove sodium prior to geological disposition. Helium bonded metallic fuel of several alloys has been irradiated to better understand the implications of removing sodium bonding. Current results show that with proper dimensional tolerances on the final fuel slug, helium bonding is feasible. Low smear density fuel is necessary to enable high (30 atom %) burnup. Both solid and annular low smear density fuels have also been examined, and their performance is also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Top Fuel 2016 Proceedings

Top Fuel, also called Light Water Reactor (LWR) Fuel Performance Meeting, is an annual topical meeting whose primary objective is to bring together leading specialists in LWR fuel from around the world to analyze advances in nuclear fuel technology and to use the findings of the latest cutting-edge research to help manufacture high performance nuclear fuels of today and tomorrow. The conference hosts 2 plenary sessions, 2 special sessions, and numerous technical sessions organized into 5 technical track ares: Track 1 - Fuel performance reliability, operations and maintenance experience; Track 2 - Advances in fuel technologies (e.g., enhanced accident tolerant fuel); Track 3 - Transient and off-normal fuel behavior; Track 4 - Used fuel storage, transportation, and reuse/recovery; and Track 5 - Fuel modeling and analysis. These proceedings gather 169 papers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Vacuum drying of spent fuel storage canisters/casks - 'just how much water could be left in there, anyway?' - 15233

The process of moving spent fuel assemblies from wet storage in the spent fuel pool to dry storage in an inert gas environment within a canister or cask typically involves two main stages. In the first stage, the bulk of the water is removed by a draining or siphoning operation, conducted with pressure inside the canister maintained at 1.3 to 2 atm. This process removes most of the water in the cask cavity, but residual water can be left behind, in three main regions. The largest region, typically, is the bottom of the cavity below the end of the siphon tube. A second general class of locations, also located at the bottom of the cask cavity, consists of the 'dashpots' at the ends of the various non-fuel tubular structures within the fuel assemblies. For PWR assemblies, these are the control rod guide sleeve tubes and instrumentation thimbles. For BWR assemblies, these are the water rods of various designs and control blade ends in some configurations. A third class of locations consists of porous material within the cavity that can absorb water, such as thin ceramic neutron poison plates, and ostensibly intact fuel rods that nevertheless might have pinholes or fine cracks in the cladding that could allow pool water to creep inside the fuel rod. To remove residual water from these locations, the second stage of drying operations utilizes vacuum drying, by reducing the pressure sufficiently to convert the water to vapor that can then be pumped out of the canister cavity. In this operation, the cask cavity pressure is slowly reduced from the nominal condition of (1.3-2.0 atm (1000-1520 torr), to a final target pressure below 3 torr (0.004 atm). This is typically accomplished in three to five stages, with prescribed 'hold' points defined in procedures or in the Technical Specifications, to preclude sudden pressure spikes due to flashing of water to steam, or development of large local temperature gradients in the fuel rods or other cask internal components. This paper discusses the various locations in the system where residual water may 'hang up' in the cask, and presents a discussion of the physical basis for the effectiveness of vacuum drying for removing it. Results and conclusions based on realistic qualitative and quantitative evaluations of the transient heat transfer behavior of the system are presented, based on detailed modeling of a range of spent fuel canister designs and decay heat loads, most notably the 24P Dry Shielded Canisters (DSCs) stored at the Rancho Seco Independent Spent Fuel Storage Installation (ISFSI), the NAC-MPC modules at Maine Yankee, the 24P DSCs stored at Calvert Cliffs, and the Holtec HI-STORM modules at Hope Creek. The evaluations presented here show that ice formation within a spent fuel canister is highly unlikely, due to the availability of a large thermal source term in the form of the decay heat load of the spent fuel. The formation of ice is an operational concern, since it mainly affects operation of equipment external to the canister (i.e., the hoses, valves, and fittings of the vacuum drying equipment.) Evaluations are presented for quantifying the amount of residual water that could remain in the cask cavity, even after successful drying operations, since vacuum drying is a limit process, and by definition, cannot remove all of the water from the cavity, no matter how long the operation goes on, even if all of it is in vapor form. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Total Integrated Dose to Spent Fuel Pool Level Instrument Probes and Refueling Floor Personnel Dose Rates Meeting NRC Order EA-12-051 for a Typical Boiling Water Reactor - Paper 22

In reviewing the events after the March 11, 2011 Fukushima accident, the Nuclear Regulatory Commission (NRC) issued Order EA-12-051, Reference 1, requiring new reliable spent fuel pool (SFP) instrumentation for pool level determination under beyond design basis event conditions. In 2013, URS prepared a calculation for a typical Boiling Water Reactor to determine the Total Integrated Dose (TID) potentially to be imparted to radiation sensitive underwater probes (one primary and one backup) planned for meeting this Order. Additionally, URS calculated the radiological dose rates to an individual standing on the refueling floor for various SFP fill levels. NEI 12-02, Reference 2, defines the levels at which SFP condition identification is necessary as the following: Level 1 - level that is adequate to support operation of the normal fuel pool cooling system. Level 1 represents the higher of the following two points: - The level at which reliable suction loss occurs due to uncovering of the coolant inlet pipe, weir or vacuum breaker (depending on the design), or - The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis. Level 2 - level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. Level 2 is based on either of the following: - 10 feet (+/- 1 foot) above the highest point of any fuel rack seated in the spent fuel pools, or - A designated level that provides adequate radiation shielding to maintain personnel radiological dose levels within acceptable limits while performing local operations in the vicinity of the pool. This level shall be based on either plant-specific or appropriate generic shielding calculations, considering the emergency conditions that may apply at the time and the scope of necessary local operations, including installation of portable SFP instrument channel components. Level 3 - level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. Level 3 corresponds nominally (i.e. +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. The URS calculation did not define these levels or determine instrumentation setpoints. However, it determined refueling floor dose rates that could be used as a tool to determine the appropriate water depths for these levels. In determining the TID for the SFP probes and the radiological dose rates to the individual, the following radioactive sources were considered: 1. Spent Nuclear Fuel (SNF) stored in the spent fuel storage racks, 2. Local Power Range Monitors (LPRMs) stored above the top of the racks on hangers on the SFP walls 3. Radioactive material laden filters stored on platforms on top of the racks, and 4. Control Rod Blades (CRBs) stored above the top of the racks on hangers on the SFP walls. The TID for the SFP probes includes both the dose from normal operation (pool full of water) for the remaining plant life (an assumed 40 years, conservatively including an allowance for lifetime extension SFP loading) and, consistent with the NEI 12-02 guidelines, Reference 2, the assumed drop in elevation of the pool water level to the top of the spent fuel racks for a 6 hour period. The acceptable TID for the radiation-sensitive probe as defined by the manufacturer was 2x10{sup 9} rad. The individual on the pool deck was assumed to be standing on the edge of the pool with an unobstructed view of its contents, as a worst case scenario. Dose rates to this individual determined from direct shine from the same 4 sources were determined for SFP water levels varying from 1 ft above the SNF racks to 21 ft above the racks in one foot increments. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor have been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235 U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Separate effects irradiation testing of miniature fuel specimens

We report that qualification of new nuclear fuels is necessary for their deployment and requires a thorough understanding of fuel behavior under irradiation. Traditionally, nuclear fuels have been qualified by performing exhaustive integral tests under a limited range of prototypic conditions designed for their specific reactor application. While some integral fuel testing is essential, basic data on behavior and property evolution under irradiation can be obtained from separate effects tests. These irradiations could offer reduced cost, reduced complexity, and in the case of accelerated testing, reduced time to achieve a given burnup. Furthermore, it may be desirable to design test irradiations capable of deconvoluting the myriad effects of burnup, temperature gradients, and other factors inherent to integral irradiation tests. Oak Ridge National Laboratory has developed an experimental capability to perform separate effects irradiation testing of miniature fuel specimens in the High Flux Isotope Reactor (HFIR): the “MiniFuel” irradiation vehicle. The small size (<4 mm 3 ) of the fuel specimens simplifies the design, analysis, and post-irradiation examination. By reducing the fuel mass, the total heat generated inside the experiment vehicle can be dominated by gamma heating in the structural materials instead of fission heating in the fuel. This essentially decouples the fuel temperature from the fission rate, allowing for highly accelerated testing (3X-18X the burnup rate of a typical light water reactor for 235 U enrichments varying from 0.22 wt% to 8 wt%) and an extremely flexible experiment design that can accommodate a wide range of fuel temperatures (~100 °C to >1200 °C), compositions, enrichments, and even geometries without requiring detailed analyses for each fuel variant. In conclusion, this paper summarizes the experiment design concept, evaluates potential applications for specific fuel forms, and briefly describes the first set of experiments on uranium nitride kernels that have been assembled and are currently being irradiated in the HFIR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Automatic information extraction from neutron radiography imaging to estimate axial fuel expansion in $\mathrm{EBR-II}$

Approximately 130,000 metal fuel pins were irradiated in the Experimental Breeder Reactor II (EBR-II) during its 30 years of operation to develop and characterize existing and prospective fuels. For many of the metal fuel irradiation experiments, neutron radiography imaging was performed to characterize fuel behavior, like fuel swelling. However, due to the lack of technology or resources, many of the images have not been processed or were processed manually through visual examination. This paper represents first-attempt to develop an image processing algorithm capable of automatically extracting information regarding the degree of fuel swelling from neutron radiography imaging. The algorithm was applied to 120 images of three different metallic fuel pin compositions—U-10Zr, U-8Pu-10Zr, and U-19Pu-10Zr. The algorithm performs operations of image intensity adjustment, image binarization, region finding, and labeling to extract information about fuel swelling. The average growth for U-10Zr was found to be 8.49% with 95% Confidence Interval (CI) [8.33 –8.66%], for U-8Pu-10Zr — 7.50% with 95% CI [7.23 – 7.78%], and for U-19Pu-10Zr — 3.15%, with 95% CI [2.40 – 3.91%]. The results obtained by applying this automatic image processing algorithm are consistent with previously reported studies of the same types of fuels. The automatic image processing algorithm will be expanded to include thousands of available neutron radiography images and different types of fuels to investigate empirical dependencies of fuel swelling, which can be subsequently applied to advanced fuel modeling. Results from this study can later be compared to BISON simulations to further benchmark modeling efforts and develop assessment cases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

BISON fuel performance modeling optimization for experiment X447 and X447A using axial swelling and cladding strain measurements

With the recent need to qualify new reactor designs such as the Versatile Test Reactor (VTR), fuel performance calculations need to be performed to determine safety criteria of the proposed designs. In order to validate the fuel performance results obtained by a fuel performance code, BISON, for new reactor designs, legacy fuel from EBR-II and FFTF MFF with Post -Irradiation Examination (PIE) data need to be used as validation cases to benchmark models. Here in this work, BISON has been paired with the Fuels Irradiation & Physics Database (FIPD) and IFR Materials Information System (IMIS) to supply PIE data for comparison with simulations of EBR-II experiments X447/X447A. X447/X447A were assessed by implementing models for Fuel Cladding Chemical Interaction (FCCI) within BISON and optimizing the friction coefficient between the fuel surface and the cladding, the anisotropic swelling factor, and the HT9 first thermal creep scalar (which scales the first term in the HT9 creep equation) to best match the PIE axial fuel swelling height and cladding profilometry for all pins in X447/X447A. The optimal values were found using a generic algorithm developed to select different values for the three parameters until end criteria was met and error couldn’t be reduced further. The BISON-simulated cladding profilometry was evaluated using Standard Error of the Estimate (SEE) to account for the profile shape of the cladding profilometry. Optimal values for the friction coefficient, anisotropic fuel swelling factor, and HT9 first thermal creep scalar were found to best fit the BISON simulation results to the PIE measurements found in IMIS and FIPD. Improvements to current models are suggested to account for the underprediction of fuel swelling at low burnups and the overprediction of fuel swelling at higher burnups observed for the axial fuel swelling height. Although two pins in EBR-II X447/X447A (DP70 and DP75) were known to fail due to FCCI, none of the pins simulated in BISON reached a cumulative damage fraction (CDF) above 0.008 with FCCI correlations coupled in the BISON simulations. The error estimate generated for all pins in X447/X447A using optimal values was 209 µm, which is deemed acceptable.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by conversion ratio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loading scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗