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DOE OSTI2018

RISK-INFORMED EXTERNAL HAZARDS ANALYSIS BY EEVE TOOLKIT

Introducing risk-informed safety analyses for external events can bring sensible benefits to the nuclear power industry easing the regulatory burden introduced in the aftermath of the Fukushima accident. In the framework of the US-DOE Light-Water Reactor Sustainability program, Risk-Informed Margin Characterization project (LWRS-RISMC), research activities were conducted at INL for developing and applying a toolkit for performing risk-informed external hazards analysis. The “EEVE” (External EVEnts) toolkit, a chain of INL and state-of-the-art codes, was set-up and applied for simulating earthquake-induced transients with internal flooding in a PWR. This was done using methodology set-up for evaluating new data and determining critical areas that would benefit most from advanced analysis methods. The methodology is agnostic and it can be applied to other external hazards than just seismic and flooding (e.g., high winds, intense precipitation, etc.). The first step was the deterministic calculations for evaluating the propagation of a set of earthquakes and for determining the effects on a fire suppression system in an auxiliary building. These deterministic calculations were performed using the LS-DYNA and the OPENSEE structural mechanics codes. Probability failures and event trees of a generic 3-loops Westinghouse PWR were derived from a generic SAPHIRE model. Seismically induced Loss of Offsite Power, Station Blackout, and fire suppression piping failure events were considered. SAPHIRE code allowed a pre-screening of all possible branches of the event tree and a selection of those with the largest Core Damage Frequency increase for further analysis using dynamic PRA methods. INL's EMRALD code performed dynamic PRA calculations of the identified sequences, integrating risk analysis with on-line deterministic safety analysis results (seismically-induced internal flooding calculations by NEUTRINO code and system analysis by RELAP5-3D code). RELAP5-3D Best-Estimate calculations coupled with RAVEN statistical code were used for deriving limit surfaces as function of relevant transient parameters. This data was used by EMRALD for determining the final core status (core damage/core safe). Finally, the coupling of seismic, flooding, system thermal-hydraulic, uncertainty quantification and probabilistic risk analysis for an advanced risk-informed external hazards analysis was demonstrated. Results showed the sensible decrease in the level of conservatism achieved using the developed risk-informed methodology, which allowed also improving the evaluation of the failure modes and failure probabilities of different components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

NEUP Final Report: Multilayer Composite Fuel Cladding and Core Internals for LWR Performance Enhancement and Severe Accident Tolerance

Accident tolerance of LWR fuels and structures is of paramount importance, as highlighted by the accident at the Fukushima Daiichi nuclear power station. The ability of fuel cladding and core internals to resist runaway oxidation during a beyond design basis accident (BDBA), as well as to minimize corrosion during steady state operation and design basis accidents (DBAs), determines its degree of accident tolerance. A gap was identified in the accident tolerant fuel (ATF) concepts under consideration, as most consisted either of coatings which may not hold their integrity under prolonged operation, or different materials altogether which incur significant neutronic penalties. A compromise was therefore sought – by minimizing the amount of coating material while ensuring a strong microstructural bond, it was thought that a Zircaloy-steel layered composite would achieve the best of both approaches.The goal of this NEUP project was to develop a multi-metallic layered composite (MMLC) tailored to addressing accident tolerance of LWR fuel cladding and core internal structures using an innovative fabrication technology. The MMLC developed in this program is expected to enhance the accident tolerance of LWRs, thereby reducing the cost by recovering lost operating margins and/or increasing operating windows of peak cladding temperature, peak linear power, reducing steady-state corrosion, and enhancing severe accident tolerance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Methods and Usability Enhancements in Shift for Non-LWR Applications

Several development and analysis tasks were undertaken in FY21 under the Nuclear Energy Advanced Modeling and Simulation program to enhance modeling of non light–water reactors (LWRs) with Shift. Specifically, these efforts targeted enhancements for tristructural isotropic (TRISO) fuel modeling. A new Shift user interface was developed that allows for much better usability and ease of modeling for non-LWR problems and TRISO fuel. Performance studies were conducted using an HTR-10 fuel pebble model by comparing different geometry packages in Shift, KENO-VI, and Serpent. These studies showed that the new geometry package in Shift performs well compared to Serpent for TRISO fuel modeling with consistent tracking options between both packages. The studies also identified the most critical areas of improvement for more efficiently performing Monte Carlo transport on TRISO fuel models with Shift. Tally calculations in Shift were optimized for non-LWR cross section generation and depletion calculations, and areas for further optimization and accuracy improvements were identified. Finally, initial collaboration efforts were formed between Idaho National Laboratory, Argonne National Laboratory, and the Nuclear Regulatory Commission to use Shift for Comprehensive Reactor Analysis Bundle support.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

White Paper: Research & Development for the Time at Temperature Approach

Recent advancements in nuclear power research are greatly improving reactor safety and performance through the development of Accident Tolerant Fuel (ATF) and Low-Enriched Uranium Plus (LEU+). These innovations can address Departure from Nucleate Boiling (DNB) margins, which are vital for reactor safety. DNB happens when the coolant switches to film boiling, significantly decreasing heat transfer and posing a risk of fuel cladding failure. The U.S. Nuclear Regulatory Commission (NRC) employs conservative DNB criteria, which can potentially restrict the operational flexibility and efficiency of reactors. The Time at Temperature (TaT) approach could provide a more detailed and adaptable operational guideline by establishing acceptable time-temperature limits, accounting for the duration a material can withstand elevated temperatures without losing its integrity. This method allows reactors to operate more efficiently and safely, offering additional operational margins, faster power adjustments, and improved fuel cycle economics. TaT criteria allow for higher power levels and more flexible responses to operational transients, particularly applicable for anticipated operational occurrences (AOOs) that result in short durations of post-DNB conditions. It enhances plant operational flexibility, allows faster startup times, and enables quicker power level adjustments, optimizing fuel loading patterns and improving fuel cycle economics. Implementing TaT limits reduces core design constraints, lowers fuel usage, and reduces costs, essential for the long-term sustainability of Light Water Reactors (LWRs). TaT maximizes the use of advanced fuel technologies like ATF and LEU+, further enhancing their economic and environmental benefits. To apply the TaT approach in existing LWRs, collaborative research activities among various DOE-sponsored programs are essential. These efforts should incorporate fuel experiments, physics-based high-fidelity modeling, ML-based surrogate modeling, and optimization techniques. This whitepaper proposes four research and development areas: 1) Investigation of the feasibility of new operations of LWR with updated safety limits; 2) Assessment of reactor operation limits through uncertainty reduction; 3) Evaluation of power uprate in virtual environment; and 4) Lattice and reactor core design for power uprate. Each area includes why this research is in need and a suggested scope of work. These comprehensive research areas ensure practical and beneficial advancements for existing reactors, translating innovations in nuclear fuel and cladding technology into improved reactor performance and safety.

42 - ENGINEERING↗

DOE OSTI2022

Improvement of two-phase closure models in CTF using Bayesian inference

Under the Consortium for Advanced Simulation of Light Water Reactors (CASL) program, extensive capabilities have been developed in CTF to analyze light-water reactors (LWRs) for normal operating conditions, departure from nucleate boiling (DNB), and system transients. However, further improvements are required in the modeling and simulation of boiling water reactors (BWRs), which is a focus of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. In this work, CTF validation results were used to optimize selected modeling coefficients by calibrating to experimental data using a Bayesian inference approach. Here, calibration studies were conducted to improve (vapor) void fraction prediction without worsening the two-phase pressure drop prediction, as well as to improve the two-phase pressure drop prediction. Calibration was performed for interfacial drag and wall shear models. Surrogates were developed to alleviate the computational expense required for sampling the parameter space using Markov chain Monte Carlo (MCMC). An assessment performed with calibrated models demonstrated an improvement of CTF in its prediction of key parameters such as void fraction and two-phase pressure drop.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Comparison of experimental and simulated critical heat flux tests with various cladding alloys: Sensitivity of iron-chromium-aluminum (FeCrAl) to heat transfer coefficients and material properties

In this paper we analyze the differences between previous transient critical heat flux (CHF) experiments using iron-chromium-aluminum (FeCrAl), Inconel 600, and stainless steel 316 (SS316) alloy test sections and best-estimate modeling results from widely-used nuclear engineering systems and subchannel analysis tools. FeCrAl is an Accident Tolerant Fuel (ATF) candidate cladding material. The thermal hydraulic performance and safety characteristics of FeCrAl are being evaluated to determine viability as a cladding material in Light Water Reactors (LWRs). In this study, the results of the CHF experiments conducted at atmospheric pressure and fixed inlet coolant temperature and mass flux are compared to models built in the fifth version of the Reactor Excursion and Leak Analysis Program (RELAP5-3D) and CTF, the modernized version of COBRA-TF developed by the Consortium for Advanced Simulation of LWRs (CASL). Results from RELAP5-3D and CTF showed differences from the experiments and from each other in predicting CHF. In the Inconel 600 case, both computational tools overpredicted CHF, which led to an underprediction in the tube outer surface temperature. In the SS316 and FeCrAl cases, CHF was underpredicted by the codes, leading to an overprediction of the tube outer surface temperature. To understand the discrepancies in CHF and post-CHF predictions, studies were performed using RELAP5-3D and RAVEN to determine the sensitivity of CHF and peak test section temperature, an analog to peak cladding temperature (PCT), to heat transfer coefficients, a CHF multiplier, and uncertainties in the thermal conductivity and volumetric heat capacity. We found that CHF depends most strongly on the CHF multiplier and thermophysical properties. A combination of these factors that produced the best match to the experiment based on CHF, PCT, and the total energy deposited into the tube was determined. The best match parameters were able to provide best-estimate predictions of the CHF and integral heat flux, but were still conservative when predicting the PCT. The best match set of parameters developed in this paper are intended only as a demonstration of an approach that could be applied in the future with a larger set of experiments to produce more accurate models of CHF and post-CHF behavior.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2013

Nuclear Energy Return on Energy Investment

This report provides a methodology and requisite data to assess the potential Energy Return on (Energy) Investment (EROI) for nuclear fuel cycle alternatives, and applies that methodology to an example “once-through” fuel cycle using low enrichment uranium (LEU) in conventional light water cooled reactors (LWRs). Within the DOE Office of Nuclear Energy Science and Technology - Fuel Cycle Technologies program, the Fuel Cycle Options Campaign “performs integrating analyses of nuclear energy and fuel cycle systems to inform fuel cycle R&D, programmatic decisions, strategy formulation, and policy development”. Campaign objectives include development of relevant fuel cycle metrics, and development of tools and associated data for analysis of fuel cycle systems. This study represents an extension of a prior evaluation of EROI as a metric for fuel cycle facilities, processes and technologies. That prior study [Simon 2011] addressed the energy return on the addition of fuel recycle to an existing nuclear energy system. Limited to just the addition of fuel recycle, that study did not include all the energy investments required to create, operate and decommission the underlying nuclear fuel cycle, such as uranium mining, fuel fabrication, reactor construction, and used fuel disposition. This extension of the prior work adds these remaining pieces of the fuel cycle to provide a basic evaluation framework and initial data to enable evaluation of EROI for nuclear energy in general. It is intended as a basis for evaluation of alternative fuel cycle options in the future with the addition of pertinent details for other fuel cycle scenarios. In the prior work, a spreadsheet tool was constructed for the specific purpose of this energy input/output analysis of nuclear fuel cycle facilities. This tool allows the user to enter the parameters of the nuclear fuel cycle and assumptions about energy use in reprocessing. In the current analysis, energy consumption for an entire nuclear energy enterprise is considered, including facility construction, materials, facility operation, and decommissioning. The analysis tool has been extended to include the entire fuel cycle for a representative scenario of a once-through LWR nuclear energy system. Energy content data have been developed for an initial representation. In many cases where there is a valid range of plausible energy content, the range is discussed and a representative value selected for demonstration of the evaluation methodology. In this analysis, energy produced is the energy output of power reactors within the fuel cycle. The spreadsheet calculates both Primary EROI and Final EROI for the fuel cycle entered. The EROI is the ratio of the output energy divided by the consumed energy. The intent for this study was to develop the methodology and analysis tool for a complete fuel cycle. It is not the intent of the study to fully explore the wide range of potential energy intensities, or to reconcile the disparate values found in the literature. Representative numbers were used in this demonstration. The methodology and tool do provide a framework for future exploration of the key energy intensity values and for conducting sensitivity studies on specific values, either to assess improved understanding of the values, or to explore the potential for alternative technologies to impact EROI. The Primary and Final Energy EROI values calculated for the representative scenario were ~ 52 and 24, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

CFD-grade Experiments Addressing Thermal Fatigue in Nuclear Reactor Branch Lines

Thermal fatigue has been identified as one of the main degradation mechanisms contributing to Loss of Coolant Accident frequencies in Light Water Reactors (LWRs). Thermal fatigue in stagnant branch lines departing from hot or cold legs of LWRs has become of high interest to the nuclear industry, as the current management program has had mixed effectiveness in preventing loss of power incidents. In fact, at least 8 such events involving branch lines have been registered at US plants over the past 2 years. Thermal cycling in a branch line occurs at the fluctuating boundary created by the penetration of hot fluid from the main flow into the cold stagnant fluid of the branch line. Penetration occurs as the fast moving flow of the main line crosses the opening of a branch line, causing cavity flow. Turbulent cavity flow dissipates to a disturbed flow further into the branch line and then to a laminar spiral flow. At the extent of the spiral flow, hot and cold fluids stratify to create the fluctuating boundary of interest. Qualitative experimental techniques, such as visualization of tracer particles, have been employed in past research efforts to capture the large scale evolution from cavity flow to the formation of the fluctuating stratified layer. The insight gained has been crucial in understanding the fundamentals of thermal cycling and stratification in branch lines; however, this qualitative information is insufficient for improving the current methods of predicting thermal fatigue in branch lines. High resolution quantitative measurements of the flow, capturing the small scale motions that affect the thermal cycling at the pipe wall, are needed to shed more light on the phenomenon and improve the industry fatigue management program, especially in terms of reliability. In the work presented here, time-resolved particle image velocimetry (PIV) is employed to investigate the flow in a stagnant branch line. Both time-averaged and time-dependent velocity fields of interest were captured, providing the groundwork for the experimental campaign that will be translated into model enhancement. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Systematic Technology Evaluation Program for SiC-based BWR Channel Box

Silicon carbide (SiC) –based accident-tolerant fuel (ATF) cladding of light water reactors (LWRs) has become a topic of great interest in nuclear material development since the Fukushima Dai-ichi Nuclear Power Plant accident. Extensive development of SiC cladding has identified benefits of the concept, including its inherent accident resistance and neutronic performance. Advancements achieved during the development of SiC-based ATF cladding have made it attractive for use in accident-tolerant cores of LWRs. In this document, we consider the use of SiC-based materials to replace the zirconium alloy used for the channel boxes of boiling water reactors. Both accident resistance and prevention of the bowing observed in current grades of channel boxes are needs driving the exploration of a new channel box system. To provide a pathway for channel box development, this document lays out (1) a current technology overview, (2) the technology gap, and (3) a systematic technology evaluation plan. The objectives of the technology evaluation plan are to address key feasibility issues revealed by the knowledge gap analysis and to contribute to advancing the technology readiness level. Further development will be required in the technology implementation phase. The technology evaluations include understanding the behavior of channel boxes under both normal and off-normal conditions, identifying potential mitigation strategies for the issues if necessary, and improving channel box performance based on the knowledge obtained. The structure of the technology evaluation plan and each task are shown in Figure S.1. The knowledge gap analysis identified key research areas, including irradiation effects, hydrothermal corrosion, accident behavior, and design and manufacture. These are common issues between SiC-based channel boxes and SiC-based fuel cladding, although technical evaluations specific for the channel box components are required. This document also describes details of the technical feasibility evaluation methods and plans.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Preliminary Process and Instrumentation Design of Advanced Reactor Integration with Refineries and Hydrogen Production Facilities

With the ongoing push to decarbonize energy use and especially greenhouse gas emissions across all sectors, there are incentives to investigate how nuclear reactors may be used to generate clean energy and be used in various energy economies beyond just the electrical grid. Two initial integrations, high temperature steam electrolysis (HTSE) and oil refineries, are investigated in this first DOE Integrated Energy Systems (IES) program detailed industrial integration design report. Increasingly detailed reports are anticipated both for the industries discussed in this report and for additional industries in the future of the program. This report is a robust starting point showing how integration thermodynamic analysis establishes the requirements on the reactor and methods by which those requirements can be evaluated for specific reactor designs. Two Advanced Reactor Demonstration Program awardees are selected as representative designs for their respective technologies: NuScale for light-water reactors (LWRs) and X-Energy for high temperature gas reactors (HTGRs). Other reactor technologies or specific reactor configurations would require specific analysis similar to what is done in this report, thus this report can be a reference point by which to extend this work to other nuclear plant designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Nuclear Fuel Cycle and Supply Chain (NFCSC) Technical Monthly September FY-24

AFC hosted the Light Water Reactor (LWR) Fuels Research Workshop and Electric Power Research Institute (EPRI) Collaborative Research on Advanced Fuel Technologies (CRAFT) Meeting September 9-13, 2024. The Department of Energy's (DOE) Advanced Fuels Campaign (AFC) Accident Tolerant Fuel (ATF) program is pivotal in advancing clean energy through collaborative innovation in fuel technology for light water reactors (LWRs). For over a decade, AFC has engaged a wide array of stakeholders—researchers, industry participants, and regulatory bodies—both within the U.S. and internationally to develop ATF technology. This initiative aims to enhance the safety and economic performance of both Pressurized Water Reactors (PWRs) and LWRs. To foster community relationships and disseminate research, AFC initiated annual LWR Fuel workshops in 2024, supplementing ongoing industry-led CRAFT workshops.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Fully Ceramic Microencapsulated Fuel Development for LWR Applications

The concept, fabrication, and key feasibility issues of a new fuel form based on the microencapsulated (TRISO-type) fuel which has been specifically engineered for LWR application and compacted within a SiC matrix will be presented. This fuel, the so-called fully ceramic microencapsulated fuel is currently undergoing development as an accident tolerant fuel for potential UO2 replacement in commercial LWRs. While the ability of this fuel to facilitate normal LWR cycle performance is an ongoing effort within the program, this will not be a focus of this paper. Rather, key feasibility and performance aspects of the fuel will be presented including the ability to fabricate a LWR-specific TRISO, the need for and route to a high thermal conductivity and fully dense matrix that contains neutron poisons, and the performance of that matrix under irradiation and the interaction of the fuel with commercial zircaloy clad.

Snead, Lance Lewis↗

DOE OSTI2014

Analysis of Phase Transformation Studies in Solute Addition Alloys

Components serving in a nuclear reactor plant must withstand a very harsh environment including extended time at elevated temperatures, neutron irradiation, stress, and/or corrosive media. The many modes of degradation are complex and very depending on material and service locations. However, understanding and managing materials degradation is paramount for the continued safe and reliable operation of nuclear power plants. For the reactor core and primary systems, several key areas have been identified. Thermomechanical considerations such as aging and fatigue must be examined. Irradiation-induced processes must also be considered for higher fluences under the life extension consideration of the current light water reactors (LWRs), which particularly includes the influence of radiation-induced segregation (RIS), swelling, and precipitation on component hardening and embrittlement. Corrosion takes many forms within the reactor core, although irradiation-assisted stress corrosion cracking (IASCC) and primary water stress corrosion cracking (PWSCC) are of high interest in extended life scenarios. Research in these areas can build upon other ongoing programs in the LWR industry as well as other reactor materials programs (such as fusion and fast reactors) to help resolve these issues for extended LWR life. Type 304 and 316 stainless steels are classic structural materials used in LWRs. A set of well designed type 304 and 316 variants by changing the amount of a single solute element were irradiated using the BOR-60 fast breeder reactor at ~320°C in Russia. This set of alloy samples has been used in this task to characterize detailed microstructural evolution and measure magnetic phase, swelling and hardness. This report summarizes the detailed microstructural results, characterized using conventional and scanning transmission electron microscopy (CTEM/STEM) and energy dispersive X-ray spectroscopy (EDS), and the newly developed model of strength factor for estimating radiation hardening. The results provide invaluable insights into understanding degradation mechanisms of key components and assists in the prediction of the life of similar materials and the base knowledge required to develop advanced radiation resistant alloys. The effect of solute elements on microstructural evolution under neutron irradiation is explored. The approximate correlation between microstructure changes and radiation hardening is established. The major results and implications are discussed below.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2015

PROGRESS IN THE INVESTIGATION OF NUCLEAR POWER PLANT CABLE AGING

The thousands of miles of polymer-insulated cables used for power, control and communications in nuclear power plants were originally qualified for forty years of plant operation. Continued use of cable in existing plants for periods much beyond their original qualified life will require a detailed understanding of cable degradation under plant operating conditions to assess and mitigate risk of continued use. The effect and relative importance of individual and combined environmental stresses including radiation, heat and moisture on reliable cable performance is essential to inform condition monitoring, lifetime prediction and required cable replacement planning. We present an overview and update of the nuclear cable aging program at the Pacific Northwest National Laboratory and the Oak Ridge National Laboratory supported by the U.S. Department of Energy, Office of Nuclear Energy, Light Water Reactor Sustainability program.

cable, insulation, polymer, EPR, XLPE, aging, degr↗

DOE OSTI2013

Irradiation of SiC Clad Fuel Rods in the HFIR

During 2009 and- 2010, new test capability for the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was developed that allows testing of advanced nuclear fuels and cladding under prototypic light-water-reactor (LWR) operating conditions (i.e., cladding and fuel temperatures, fuel average linear heat generation rates, and cladding fluence). For the initial experiments for this test program, ORNL teamed with commercial fuel/cladding vendors who have developed an advanced composite-wound SiC cladding material for possible use in LWRs. The first experiment, containing SiC-clad UN fuel, was inserted in HFIR in June 2010, and the second experiment, containing SiC-clad UO2 fuel, was inserted in October 2010. Two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in November 2011 at an estimated fuel burnup of approximately 10 GWd/MTHM; and two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in February 2013 at an estimated fuel burnup of approximately 20 GWd/MTHM. These capsules are currently awaiting PIE. This paper will describe the experiment, as-run operating conditions for these capsules, and current PIE plans and status.

Irradiation SiC fuel rods HFIR↗

DOE OSTI2013

Utilization of Used Nuclear Fuel in a Potential Future US Fuel Cycle Scenario - 13499

To date, the US reactor fleet has generated approximately 68,000 MTHM of used nuclear fuel (UNF) and even with no new nuclear build in the US, this stockpile will continue to grow at approximately 2,000 MTHM per year for several more decades. In the absence of reprocessing and recycle, this UNF is a liability and needs to be dealt with accordingly. However, with the development of future fuel cycle and reactor technologies in the decades ahead, there is potential for UNF to be used effectively and efficiently within a future US nuclear reactor fleet. Based on the detailed expected operating lifetimes, the future UNF discharges from the existing reactor fleet have been calculated on a yearly basis. Assuming a given electricity demand growth in the US and a corresponding growth demand for nuclear energy via new nuclear build, the future discharges of UNF have also been calculated on a yearly basis. Using realistic assumptions about reprocessing technologies and timescales and which future fuels are likely to be reprocessed, the amount of plutonium that could be separated and stored for future reactor technologies has been determined. With fast reactors (FRs) unlikely to be commercially available until 2050, any new nuclear build prior to then is assumed to be a light water reactor (LWR). If the decision is made for the US to proceed with reprocessing by 2030, the analysis shows that the UNF from future fuels discharged from 2025 onwards from the new and existing fleet of LWRs is sufficient to fuel a realistic future demand from FRs. The UNF arising from the existing LWR fleet prior to 2025 can be disposed of directly with no adverse effect on the potential to deploy a FR fleet from 2050 onwards. Furthermore, only a proportion of the UNF is required to be reprocessed from the existing fleet after 2025. All of the analyses and conclusions are based on realistic deployment timescales for reprocessing and reactor deployment. The impact of the delay in recycling the UNF from the FRs due to time in the core, cooling time, reprocessing, and re-fabrication time is built into the analysis, along with impacts in delays and other key assumptions and sensitivities have been investigated. The results of this assessment highlight how the UNF from future reactors (LWRs and FRs) and the resulting fissile materials (U and Pu) from reprocessing can be effectively utilized, and show that the timings of future nuclear programs are key considerations (both for reactors and fuel cycle facilities). The analysis also highlights how the timings are relevant to managing the UNF and how such an analysis can therefore assist in informing the potential future R and D strategy and needs of the US fuel cycle programs and reactor technology. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Status of the Norwegian thorium light water reactor (LWR) fuel development and irradiation test program

Thorium based fuels offer several benefits compared to uranium based fuels and should thus be an attractive alternative to conventional fuel types. In order for thorium based fuel to be licensed for use in current LWRs, material properties must be well known for fresh as well as irradiated fuel, and accurate prediction of fuel behavior must be possible to make for both normal operation and transient scenarios. Important parameters are known for fresh material but the behaviour of the fuel under irradiation is unknown particularly for low Th content. The irradiation campaign aims to widen the experience base to irradiated (Th,Pu)O{sub 2} fuel and (Th,U)O{sub 2} with low Th content and to confirm existing data for fresh fuel. The assumptions with respect to improved in-core fuel performance are confirmed by our preliminary irradiation test results, and our fuel manufacture trials so far indicate that both (Th,U)O{sub 2} and (Th,Pu)O{sub 2} fuels can be fabricated with existing technologies, which are possible to upscale to commercial volumes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗