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DOE OSTI2023

SAS4A/SASSYS-1 Version 5.7 and 5.6 Release Notes

SAS4A/SASSYS-1 (SAS) is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. In this report, the major developments that were implemented in Versions 5.7 and 5.6 are summarized and an overview of all changes made between versions is presented. SAS Version 5.7 and 5.6 introduced a number of modeling improvements that allow users to capture important phenomena, such as pump heating and metallic fuel performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

SAS4A/SASSYS-1 Verification Testing for Sodium Fast Reactor Applications (Acceptance Testing Report)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Status of SAS4A/SASSYS-1 Software Development and Application (FY2024)

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over fifty years. It has been identified as a critical element of safety analysis capabilities for the U.S. Department of Energy and is utilized within industry to perform the transient safety analyses required to support the licensing of Liquid Metal-cooled Fast Reactors (LMFRs). This report summarizes the code development and update activities carried out during FY2024. In FY2024, programmatic activities focused on key improvements to software useability, such as enhanced user interfaces for reactivity feedback modeling, improvements in stability/useability of the Code Manual, and improvements to the acceptance testing infrastructure, including automation of acceptance testing and generation of the Acceptance Testing Report. To support end user applications, an open training was held, a semi-public forum was maintained, and a practical benchmarking and validation matrix was developed which allowed limitations of existing testing capabilities to be assessed. The existing fuel models were also enhanced with improved modeling capabilities and testing for the oxide and annular fuel models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Status of Multiple Channel Fuel Performance Capabilities Within the SAS4A/SASSYS-1 Safety Analysis Software

SAS4A/SASSYS-1 (SAS) is a fast-running simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. It is a critical element of safety analysis capabilities for the U.S. Department of Energy and is utilized within industry to perform the transient safety analyses required to support the licensing of Liquid Metal-cooled Fast Reactors (LMFRs). Although SAS is exceptionally fast for most transient scenarios, fuel performance calculations, along with the associated pre-transient characterization of the fuel pin, may be required for transient scenarios where fuel pin failure is hypothesized. Both the pre-transient characterization and the transient fuel performance calculation are necessary to properly quantify margins to potential fuel failure and assess the time spent potentially exceeding such margins during events. While safety analysis calculations with fuel performance models provide a more detailed characterization of the reactor during a transient, the pre-transient characterization can be time-consuming and computationally expensive. Often, large numbers of fuel pins have been exposed to similar pre-transient irradiation conditions. Similarly, the same pre-transient fuel characterization may be applicable to numerous transient conditions. This provides an opportunity to optimize the SAS computational framework such that pre-transient fuel characterization can be shared across multiple channels (fuel pins) and across multiple simulations, thus dramatically reducing overall computational costs. This report summarizes progress toward enhancing the SAS computational framework to support shared, multiple channel fuel performance characterizations intended to significantly reduce computational costs. Preliminary testing has shown that the computational time saved by using the pre-transient sharing capability is approximately equal to the time it takes to perform the pre-transient characterization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

SAS4A/SASSYS-1 Verification Testing for Sodium Fast Reactor Application: Acceptance Testing Report

AS4A/SASSYS-1 (SAS) is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over sixty years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. To support the dedication effort, this report has been generated to provide a detailed description of the available verification testing. The verification testing presented in this report captures functionality testing, focusing mainly on the testing of specific functions and algorithms for accuracy and precision of output, and interface testing, focusing mainly on the testing of critical input parameters and their valid ranges. Although SAS was developed to support the analysis of any liquid-metal-cooled nuclear reactor, the testing described in this document primarily focuses on the verification of SAS capabilities as they relate to a generic pool-type Sodium Fast Reactor (SFR).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Instrumentation Plan for the OSU High Temperature Test Facility, Revision 4

This instrumentation plan describes the instrumentation requirements and design for the Oregon State University (OSU) High Temperature Test Facility (HTTF). This report outlines HTTF instrumentation design specifications, data acquisition and control system (DACS) design specifications as well as a brief description of the instrumentation to be used in the HTTF. The HTTF is an integral test facility, which has been designed to model the behavior of interest for a Very High Temperature Gas Reactor (VHTR) during the Depressurized Conduction Cooldown (DCC) event. It also has the potential to conduct limited explorations into the progression of the Pressurized Conduction Cooldown (PCC) event and phenomena during normal operations. The facility is scaled 1:4 by height and 1:64 by volume. The maximum core power for the facility is approximately 2.2MW. The core region is of a modular design to allow for both the prismatic block core and the pebble bed core designs to be tested. However, only the prismatic block design is discussed in this report. A correctly scaled integral facility modeling the VHTR that can be used to examine the thermal hydraulic system behavior of the design basis and beyond design basis transients is of significant value. The experimental results obtained in such a program could be used to validate analytical tools and methods being proposed for use with the VHTR. Therefore, it is of importance that the HTTF be instrumented in such a way that the required data is collected to meet its overall goals of phenomena examination and code validation. In addition to the instrumentation used to collect data to achieve the overall goals of the HTTF test program, there is instrumentation in the HTTF, which is primarily used to ensure the safe operation of the test facility. This type of instrumentation shall be termed “Control Instrumentation” and is discussed in this report. This report discusses the DACS, which is a personal computer controlled system that is used to control both the collection and storage of the test matrix and control system data as well as electrical and mechanical operation of the test facility.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

SAS4A/SASSYS-1 Modeling Improvements for the Transition to Natural Circulation

SAS4A/SASSYS-1 (SAS) is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over fifty years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. In recent years, SAS has undergone a number of improvements to enable improved safety analyses that meet end users’ modernized needs while complying with the current regulatory environment. Improvements made in versions 5.6 and 5.7 released within the last year include the development of anisotropic Reynolds number dependent loss coefficients throughout the core and heat transport systems, the ability to distinguish the transition friction factor from the fully developed laminar and turbulent friction factors, and timedependent direct coolant and wall heating for pipe-like elements in the heat transport systems. While it was possible to capture loss coefficients, friction factors, and heat transfer from an element to a heat sink within SAS in previous versions of the code, users were required to end the simulation and restart it to adjust the input to account for any significant changes to the values during the transient. With these improvements, users can better capture flow reversal, pump heating, and the transition from forced to natural circulation without being limited to constant orifice coefficients, constant heat sinks, or the need to restart the simulation and modify input. In order to demonstrate the application of these improvements, a loss of flow transient is simulated for the Advanced Burner Test Reactor (ABTR).

SAS4A/SASSYS-1↗

DOE OSTI2025

Mechanical Property Assessment of Unirradiated Cladding After Exposure to Time at Temperature

Light water reactors in the United States are operated within inherent safety limits designed to account for anticipated operational occurrences (AOOs). Unlike more severe reactor transients, these events permit the potential reuse of reactor cladding, highlighting the nuanced role of thresholds in influencing reactor efficiency compared to design-basis or beyond-design-basis transient conditions. The threshold designated for AOO peak cladding and fuel temperatures are currently grounded around thermal hydraulic limits such as surpassing the critical heat flux (CHF), rather than in fuel-cladding material performance. This CHF criterion therefore represents a conservative approach to AOO transients during the reactor lifetime. Extending this threshold to higher temperatures than CHF can result in additional operation efficiency gains. Here, a material performance-based approach is assessed to better understand the physical material limits of short time scale exposures to elevated temperatures. This approach, termed time at temperature (t@T), references fuel cladding microstructural and mechanical response to rapid temperature transients in absence of radiation damage. The results of this work are expected to help identify margin for the United States reactor fleet, which could be utilized to increase operational efficiencies through decreased reactor outages.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Chemical Looping Combustion Sensitivity Analyses: CLOU Concepts

The objective of the National Energy Technology Laboratory (NETL) effort described in this report is to generate a reference-CLOU power plant performance and cost screening assessment and associated sensitivity analyses. To do this, a review of the CLOU literature was first performed to establish a complete design basis for a reference-CLOU power plant. While complete design-basis information is not available for any CLOU oxygen carrier, a representative design basis was compiled from a variety of literature sources. The design basis was then applied to generate reference-CLOU power plant material and energy balances, the unique CLOU process equipment was approximately sized and specified, equipment cost estimates were developed, and total power plant performance, capital investment, and cost of electricity (COE) estimates were compiled. Major performance and cost sensitivity parameters were identified, and performance and cost sensitivity analyses were performed. The results are these analyses relate not only the performance and cost potential of CLOU power generation, but also provide guidance to CLOU oxygen carrier and process researchers and developers.

20 FOSSIL-FUELED POWER PLANTS↗

DOE OSTI2024

Contribution of IAEA coordinated research projects to light water reactors advanced technology fuel testing and simulation

The Fukushima-Daiichi accident in 2011 highlighted the pressing need for enhanced design and safety analysis of nuclear fuels, especially under accident conditions in nuclear power plants (NPPs). To address this need, the international nuclear fuel community has developed several concepts of Accident Tolerant and Advanced Technology Fuels (ATFs) for light water reactors. The International Atomic Energy Agency (IAEA), at the request of its Member States, has initiated a series of three Coordinated Research Projects (CRPs) to aid in the development and testing of these ATFs. Further, these projects, known as FUMAC, ACTOF, and ATF-TS, focus on the experimentation and simulation of ATFs under various accident conditions, including design basis accidents and design extension conditions. This paper provides a comprehensive overview of the main activities, as well as the achieved or anticipated results of these three CRPs. The insights and advancements gained from these projects will enable IAEA Member States to further refine nuclear fission technology, thereby contributing to efforts aimed at mitigating climate change.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Transient Testing of Nuclear Fuels and Materials in United States

The US Department of Energy (DOE) has been engaged in an effort to develop and qualify next generation LWR fuel with enhanced performance and safety and reduced waste generation since 2010. This program, which has emphasized collaboration between the DOE, U.S. national laboratories and nuclear industry, was refocused from enhanced performance to enhanced accident tolerance following the events at Fukushima in 2011. Accident tolerant fuels have been specifically described as fuels that, in comparison with standard UO2-Zircaloy, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events. The program maintains an ambitious goal to insert a lead test assembly (LTA) of the new design into a commercial power reactor by 2022 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

FY2017 Updates to the SAS4A/SASSYS-1 Safety Analysis Code

The SAS4A/SASSYS-1 safety analysis software is used to perform deterministic analysis of anticipated events as well as design-basis and beyond-design-basis accidents for advanced fast reactors. It plays a central role in the analysis of U.S. DOE conceptual designs, proposed test and demonstration reactors, and in domestic and international collaborations. This report summarizes the code development activities that have taken place during FY2017. Extensions to the void and cladding reactivity feedback models have been implemented, and Control System capabilities have been improved through a new virtual data acquisition system for plant state variables and an additional Block Signal for a variable lag compensator to represent reactivity feedback for novel shutdown devices. Current code development and maintenance needs are also summarized in three key areas: software quality assurance, modeling improvements, and maintenance of related tools. With ongoing support, SAS4A/SASSYS-1 can continue to fulfill its growing role in fast reactor safety analysis and help solidify DOE’s leadership role in fast reactor safety both domestically and in international collaborations.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2022

Multi-Physics Coupled Seismic Safety Analysis of Molten-Salt Reactors

Pool-type Molten Salt Reactors (MSRs) are a promising advanced nuclear reactor concept relying on passive physical behavior to provide increased safety characteristics. The high heat capacity, high boiling point, unpressurized liquid salt contains the nuclear fuel and as it passes through the vessel achieves criticality and produces nuclear power. The salt then carries away the heat produced along with the delayed neutron precursors towards heat exchangers and primary pumps before entering the core. This forms a highly coupled multiphysics problem between neutronics and fluid flow, which makes modeling these reactors challenging. The safety/licensing case heavily relies on modeling and simulation to prove their safety, and significant effort has been expanded by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to develop the appropriate simulation tools. Their safety/licensing will require a thorough analysis of their behavior during earthquake-based transients. Earthquakes are a transient condition commonly experienced in many regions, and are safely survived by dozens of reactors every year. Nonetheless, significant conservatisms were introduced with limited predictive modeling and simulation available. The first-of-a-kind capability developed by this research will allow for engineers to analyze ranges of design-basis and beyond-design basis accident scenarios to assess the integrity of the reactor vessel and surrounding system, possibly allowing for limiting the costly conservatisms in the design. This projects uses MASTODON, Griffin, and Pronghorn in the NEAMS ecosystem to model the coupled multiphysics problem posed by an earthquake in a molten salt reactor. It marches through a number of number of coupling schemes, from uncoupled simulations to tightly two-way coupled simulations. The appropriate level of coupling for these simulations will then be determined based on the accuracy in the quantities of interest (QoI) and computational effort involved in each scheme. The QoIs chosen for this paper include the mass flow rate across the heat exchanger (for Pronghorn simulations), power output of the reactor (for Griffin simulations), and the maximum Von Mises stress (for MASTODON simulations). This summary paper reports on the current progress of this project, with 2D tightly coupled multiphysics results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2011

Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

AREVA Team Enhanced Accident Tolerant Fuels Concepts (Phase 1A Final Report)

In response to the Department of Energy (DOE) funded initiative to develop and deploy lead fuel assemblies (LFAs) of Enhanced Accident Tolerant Fuel (EATF) into a US reactor within 10 years, AREVA put together a team to develop promising technologies for improved fuel performance during off normal operations. This team consisted of the University of Florida (UF) and the University of Wisconsin (UW), Savannah River National Laboratory (SRNL), Duke Energy and Tennessee Valley Authority (TVA). This team brought broad experience and expertise to bear on EATF development. AREVA has been designing; manufacturing and testing nuclear fuel for over 50 years and is one of the 3 large international companies supplying fuel to the nuclear industry. The university and National Laboratory team members brought expertise in nuclear fuel concepts and materials development. Duke and TVA brought practical utility operating experience. This report documents the results from the initial “discovery phase” where the team explored options for EATF concepts that provide enhanced accident tolerance for both Design Basis (DB) and Beyond Design Basis Events (BDB). The main driver for the concepts under development were that they could be implemented in a 10 year time frame and be economically viable and acceptable to the nuclear fuel marketplace. The economics of fuel design make this DOE funded project very important to the nuclear industry. Even incremental changes to an existing fuel design can cost in the range of $100M to implement through to LFAs. If this money is invested evenly over 10 years then it can take the fuel vendor several decades after the start of the project to recover their initial investment and reach a breakeven point on the initial investment. Step or radical changes to a fuel assembly design can cost upwards of $500M and will take even longer for the fuel vendor to recover their investment. With the projected lifetimes of the current generation of nuclear power plants large scale investment by the fuel vendors is difficult to justify. Specific EATF enhancements considered by the AREVA team were; Improved performance in DB and BDB conditions; Reduced release to the environment in a catastrophic accident; Improved performance during normal operating conditions; Improved performance if US reactors start to load follow; Equal or improved economics of the fuel; and Improvements to the fuel behavior to support future transportation and storage of the used nuclear fuel (UNF). In pursuit of the above enhancements, EATF technology concepts that our team considered were; Additives to the fuel pellets which included; Chromia doping to increase fission gas retention. Chromia doping has the potential to improve load following characteristics, improve performance of the fuel pellet during clad failure, and potentially lock up cesium into the fuel matrix; Silicon Carbide (SiC) Fibers to improve thermal heat transfer in normal operating conditions which also improves margin in accident conditions and the potential to lock up iodine into the fuel matrix; Nano-diamond particles to enhance thermal conductivity; Coatings on the fuel cladding; and Nine coatings on the existing Zircaloy cladding to increase coping time and reduce clad oxidation and hydrogen generation during accident conditions, as well as reduce hydrogen pickup and mitigate hydride reorientation in the cladding. To facilitate the development process AREVA adopted a formal “Gate Review Process” (GR) that was used to review results and focus resources onto promising technologies to reduce costs and identify the technologies that would potentially be carried forward to LFAs within a 10 year period. During the initial discovery phase of the project AREVA took the decision to be relatively hands off and allow our university and National Laboratory partners to be free thinking and consider options that would not be constrained by preconceived ideas from the fuel vendor. To counter this and to keep the partners focused, the GR process was utilized. During this GR process each of the team members presented their findings to a board made up of technical experts from utilities, fuel manufacturing experts, fuel technical experts, and fuel research and development (R&D) experts. During the initial 2 years of the project there were several major accomplishments. These accomplishments, along with the implications for successfully implementing EATF, are; The experimental spark plasma sintering process (SPS) process was successfully used to produce fuel pellets containing either 10% SiC whiskers or nano-diamond particles. The ability to use this process enables the thermal margin enhancements of the fuel additives to be realized. Without the SPS process, the conventional process cannot support adding pellet additives in the required quantities; Coatings of Ti2AlC were successfully applied to Zircaloy-4 cladding. Testing of Ti2AlC coatings at Loss of Cooling Accident (LOCA) conditions showed reduced cladding oxidation compared to present un-coated Zircaloy-4 cladding. This achievement allows the presently used cladding system to be retained so that the 10 year schedule can be met. Having to implement a new cladding material will extend the development schedule beyond 10 years; Several documents were produced to support future development, testing, and licensing of EATF, including a design requirements traceability matrix, a draft business plan, a draft test plan, a draft regulatory plan, and the acceptance criteria for lead fuel assembly insertion into a commercial reactor. This preparatory work lays the foundation for ensuring the future development plans address all the areas required to test, license, and manufacture the new EATF; and In addition, the high velocity oxy-fuel and electrophoretic deposition (EPD) coating application processes were dropped from further consideration due to their inability to meet manufacturing criteria. This allows the resources to be focused on the most promising EATF concepts identified. Future development opportunities that were identified during this work include; The use of SiC or diamond requires that a new pellet production technique (Spark Plasma Sintering), be developed. This entails investment in developing, proving and implementing a new commercial pellet production process. Development of the process to apply thinner coatings is required; Coatings cannot be too “thick” or they will displace a significant volume of water in the core resulting in reduced thermal hydraulic characteristics; Application of the coating at high temperature can affect the Zircaloy substrate. This will require the development and implementation of a new cladding coating manufacturing process; and Replace the Cold Spray (CS) cladding coating application with the Physical Vapor Deposition (PVD) process to eliminate duplication of work and provide greater control over coating thicknesses. This can result in a reduction in the final cycle economic penalty of coatings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Overview of the U.S. DOE Accident Tolerant Fuel Development Program

The United States Fuel Cycle Research and Development Advanced Fuels Campaign has been given the responsibility to conduct research and development on enhanced accident tolerant fuels with the goal of performing a lead test assembly or lead test rod irradiation in a commercial reactor by 2022. The Advanced Fuels Campaign has defined fuels with enhanced accident tolerance as those that, in comparison with the standard UO2-Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations and operational transients, as well as design-basis and beyond design-basis events. This paper provides an overview of the FCRD Accident Tolerant Fuel program. The ATF attributes will be presented and discussed. Attributes identified as potentially important to enhance accident tolerance include reduced hydrogen generation (resulting from cladding oxidation), enhanced fission product retention under severe accident conditions, reduced cladding reaction with high-temperature steam, and improved fuel-cladding interaction for enhanced performance under extreme conditions. To demonstrate the enhanced accident tolerance of candidate fuel designs, metrics must be developed and evaluated using a combination of design features for a given LWR design, potential improvements to that design, and the design of an advanced fuel/cladding system. The aforementioned attributes provide qualitative guidance for parameters that will be considered for fuels with enhanced accident tolerance. It may be unnecessary to improve in all attributes and it is likely that some attributes or combination of attributes provide meaningful gains in accident tolerance, while others may provide only marginal benefits. Thus, an initial step in program implementation will be the development of quantitative metrics. A companion paper in these proceedings provides an update on the status of establishing these quantitative metrics for accident tolerant LWR fuel.1 The United States FCRD Advanced Fuels Campaign has embarked on an aggressive schedule for development of enhanced accident tolerant LWR fuels. The goal of developing such a fuel system that can be deployed in the U.S. LWR fleet in the next 10 to 20 years supports the sustainability of clean nuclear power generation in the United States.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Development of alternate methods of determining integrated SMR source terms

The Nuclear Energy Institute (NEI) Small Modular Reactor (SMR) Licensing Task Force (TF) has been evaluating licensing issues unique and important to iPWRs, ranking these issues, and developing NEI position papers for submittal to the U.S. Nuclear Regulatory Commission (NRC) during the past three years. Papers have been developed and submitted to the NRC in a range of areas including: Price-Anderson Act, NRC annual fees, security, modularity, and staffing. In December, 2012, NEI completed a draft position paper on SMR source terms and participated in an NRC public meeting presenting a summary of this paper, which was subsequently submitted to the NRC. One important conclusion of the source term paper was the evaluation and selection of high importance areas where additional research would have a significant impact on source terms. The highest ranked research area was iPWR containment aerosol natural deposition. The NRC accepts the use of existing aerosol deposition correlations in Regulatory Guide 1.183, but these were developed for large light water reactor (LWR) containments. Application of these correlations to an iPWR design has resulted in greater than a ten-fold reduction of containment airborne aerosol inventory as compared to large LWRs. Development and experimental justification of containment aerosol natural deposition correlations specifically for the unique iPWR containments is expected to result in a large reduction of design basis and beyond-design-basis accident source terms with concomitantly smaller dose to workers and the public. Therefore, NRC acceptance of iPWR containment aerosol natural deposition correlations will directly support the industry’s goal of reducing the Emergency Planning Zone (EPZ) for SMRs. Based on the results in this work, it is clear that thermophoresis is relatively unimportant for iPWRs. Gravitational settling is well understood, and may be the dominant process for a dry environment. Diffusiophoresis and enhanced settling by particle growth are the dominant processes for determining DFs for expected conditions in an iPWR containment. These processes are dependent on the areato-volume (A/V) ratio, which should benefit iPWR designs because these reactors have higher A/Vs compared to existing LWRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Evolution of Microstructure and Surface Characteristics of FeCrAl alloys when Subjected to Flow Boiling Testing

FeCrAl alloys are candidate materials for manufacturing accident-tolerant fuel (ATF) cladding intended for light water reactors to increase fuel reliability and safety during design-basis and beyond-design-basis accident scenarios. The evolution of the materials' surface characteristics, microstructure, and mechanical properties when exposed to the Critical Heat Flux (CHF) in flow boiling testing is crucial for safety analysis while providing insights into their thermal-hydraulic performance in nuclear reactors. After CHF, the surface chemistry of two FeCrAl alloys, APMT and C26M, was studied to understand their evolution at the early stage of high-temperature excursions in short time periods. The results indicated a thin layer composed of oxides and hydroxides of Al, Cr, and Fe with varying proportions at different depths in the layer, as indicated by X-ray photoelectron spectroscopy (XPS) and depth profiling. The cross-sections prepared by focused ion beam (FIB) revealed the growth of an oxide layer, in the range of 90-180 nm thick, on the alloys' surfaces. The evolution of the materials' surface chemistry also led to a noticeable post CHF excursion increase in their wettability, with a slight increase in roughness. Further, the investigation of the materials' mechanical properties indicated a modest increase in hardness by 10-15% as well as an increase in their yield strength, as evidenced by the microindentation and ring compression tests conducted before and after CHF testing. Scanning electron Microscopy (SEM) and X-ray diffraction (XRD) were used to investigate microstructural features of the materials and their changes after CHF treatment.

36 MATERIALS SCIENCE↗