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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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DOE OSTI2013

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The hydrogen behavior analysis in the water-ingress accident of the 250 MW pebble-bed modular high temperature gas-cooled reactor - 18639

The demonstration power plant of Chinese 250 MW high temperature gas-cooled reactor pebbled-bed module (HTR-PM) which is being under construction in the Shidao Bay of Shandong province has made breakthrough in all of the critical technologies, and is willing to operate in 2017. HTR-PM does not have the risk of large amount of hydrogen generation resulted from the zirconium water reaction due to its special design of graphite core structure and helium coolant of the primary circuit, however, considering that the HTR-PM has a small containment and does not design the deliberate ignitions which is the main hydrogen management measure in light water reactors, it is still necessary to study the hydrogen behavior in the water-ingress accident in which hydrogen will probably be generated. In this paper, the hydrogen generation amount and its distribution in the HTR-PM containment are analyzed with the TINTE code and GASFLOW code respectively in the water ingress accident caused by the steam generator heating tubes rupture. The calculation results indicate that there are total about 16.6 kg and 21 kg hydrogen produced due to graphite corrosion in the typical design basis water ingress accident and beyond design basis accident. The containment sub-pressure ventilation system makes sure the hydrogen concentration in the containment is less than 0.2% during the accident. There is no hydrogen explosion risk in the HTR-PM. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

SANDD: A directional antineutrino detector with segmented 6 Li-doped pulse-shape-sensitive plastic scintillator

We present a characterization of a small (9-liter) and mobile 0.1% 6 Li-doped pulse-shape-sensitive plastic scintillator antineutrino detector called SANDD (Segmented AntiNeutrino Directional Detector), constructed for the purpose of near-field reactor monitoring with sensitivity to antineutrino direction. SANDD comprises three different types of module. A detailed Monte Carlo simulation code was developed to match and validate the performance of each of the three modules. The combined model was then used to produce a prediction of the performance of the entire detector. Analysis cuts were established to isolate antineutrino inverse beta decay events while rejecting large fraction of backgrounds. The neutron and positron detection efficiencies are estimated to be 34.8% and 80.2%, respectively, while the coincidence detection efficiency is estimated to be 71.7%, resulting in inverse beta decay detection efficiency of 20.0% ± 0.2%(stat.) ± 2.1%(syst.). Finally, the predicted directional sensitivity of SANDD produces an uncertainty of 20° in the azimuthal direction per 100 detected antineutrino events.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2019

Thermal Energy Storage Selection for Near Term Nuclear Integration

Thermal energy storage (TES) integration with existing and future light water reactors (LWRs) has the capability to increase the deployment options and uses for the large amounts of heat produced in nuclear reactor cores. The present method of exclusively producing electricity has become less economical with increases in renewable energy penetration leading to times of low or negative electricity prices. Some response to this new grid reality has been to begin load following nuclear reactors, an unideal choice due to the relatively small modular cost of producing nuclear energy compared to the revenue that could be received by using all of the heat produced. To support research in this area, Idaho National Laboratory (INL) has joined with Utah Associated Municipal Power Systems (UAMPS) on the Carbon Free Power Project (CFPP) to contract the Joint Use Modular Plant (JUMP). CFPP is planned to be a 12-module NuScale operating site built on INL property, with the first module dedicated to JUMP for research, development, and demonstration purposes. The JUMP module coming online around the beginning of fiscal year 2027 in the current project timeline. One primary goal of JUMP is to demonstrate TES integrated with an operating LWR in a real-world environment. TES integration will allow for the demonstration of nuclear power’s capability to participate in grid load following without reactor cycling, process heat integration, and electrical ancillary market opportunities. The goal of this research is to support both SMR deployment opportunities as well as support LWR retrofit capabilities. In a discussion limited to load following or energy arbitration, “charging” and “discharging” can be simply defined. TES charging refers to time when energy is removed from the nuclear reactor power conversion cycle and directed to be stored in the TES storage medium resulting in less available energy available for grid output. Discharging refers to the process of releasing energy stored in the TES medium and having an increase in grid. This paper quantifies energy storage technologies for use with nuclear reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

United States Nuclear Manufacturing Infrastructure Assessment

This report provides a review and assessment of the existing U.S. infrastructure capabilities to manufacture on a commercial scale, the major components and unique materials required by the various U.S. designs for advanced nuclear reactors and small modular reactors (SMRs). The approach was to compare the needs of SMRs and advanced reactors against the capability of U.S. manufacturers to ascertain gaps that could inhibit U.S. industrial support for deployment of SMRs or advanced reactors. This report focused on the reactor plant designs being developed by NuScale Power (the NuScale Power Module), TerraPower (both the Traveling Wave Reactor and the Molten Chloride Fast Reactor to a limited extent), and X-energy (the Xe-100). To gather the information, MPR visited each of these companies and selected candidate suppliers to tour their facilities and conduct in-person interviews. Information was also obtained by telephone interviews, internet research, and literature review of technical reports. The overarching conclusion from this study is that the U.S. has existing capability or can readily develop capability to support and sustain most, but not all, aspects of commercial deployment of advanced reactors and SMRs. The greatest concerns for capability gaps in the U.S. infrastructure for supply of components for SMRs and advanced reactors are associated with reactor pressure vessels, steam generators, and fuel fabrication. Foreign vendors can supply reactor pressure vessels and steam generators. Issues with fuel fabrication are associated with uranium enrichment and development of fabrication methods for the innovative fuel concepts planned for advanced reactors. The technical capabilities of U.S. manufacturers appear satisfactory with the possible exceptions noted above. However, the capacity of the existing U.S. manufacturing infrastructure may be insufficient for a high rate of construction. Manufacturers contacted as part of this study indicated a willingness to expand capacity if sufficient orders are in place. The review of U.S. infrastructure included a detailed review of capabilities in many different processes including forging, tube drawing, machining, fabrication, and testing. In addition, MPR researched the advanced manufacturing processes for additive manufacturing (i.e., 3D printing) and powder metallurgy - hot isostatic pressing. Reactor plant designers and candidate suppliers are planning to rely on traditional processes for the initial plants. If advanced manufacturing processes are developed and accepted by the nuclear industry, these processes may be used in the future. In the immediate term, ongoing research (much of which is funded by DOE) is necessary to develop and refine advanced manufacturing methods. This report includes several recommendations for DOE that will advance deployment of SMRs and advanced reactors from a manufacturing perspective and other areas identified during the course of this project.

36 MATERIALS SCIENCE↗
DOE OSTI2021

NRC Multiphysics Analysis Capability Deployment (FY2021 - Part 2)

This report details the progress and activities of Idaho National Laboratory (INL) on the Nuclear Regulatory Commission (NRC) project “Development and Modeling Support for Advanced Non-Light Water Reactors.” Task 4b was completed for this report. INL developed a sample problem showing how to use Serpent 2 to calculate macroscopic cross sections for use in Griffin for a typical sodium fast reactor Unprotected Loss of Flow (ULOF) transient. The complicating factor is the relatively large axial gradient of the coolant (and hence fuel) temperature and its evolution during the transient. A 3D Griffin model of the Advanced Burner Test Reactor (ABTR) is coupled to the System Analysis Module (SAM) to perform the ULOF transient. The model includes various explicit feedback mechanisms for fast reactors, including Doppler, radial expansion from the displacement of the support plate, and axial expansion from the displacement of the fuel pins. The results for the ULOF transient are consistent with published values. We observed small effects in the transient results that arise from the superhomogenization equivalence correction of uniform and nonuniform temperature data sets, but they diminish as more dominant reactivity mechanisms are added to the model. Potential improvements to the neutronics model include: adding the differential control rod and sodium feedback mechanisms, implementing cross section re-homogenization based on the partial volume of the various materials, and enhancing the kinetics parameters. Furthermore, the fluids model only includes four subchannels to represent the entire active core due to performance limitations in SAM when attempting to model one subchannel per assembly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Coupling of Thermal Energy Storage with Small Modular Reactors

Thermal energy storage has been proposed as a load management strategy for Small Modular Reactors (SMRs) deployed on constrained grids and/or in hybrid energy systems (1). Hybrid energy systems can include process steam applications, and the presence of intermittent energy sources such as wind and solar. Under these conditions, the reactor can be subjected to significant time varying electric loads. One strategy for accommodating these loads is to operate the reactor in a load follow mode where reactor power is modulated to match the electric demand. In addition to lost energy potential, load follow operation can result in additional stresses on the fuel and other mechanical components. A more attractive approach is to operate the reactor at or near steady state and bypass excess steam to a thermal energy storage system. The thermal energy can then be recovered later, either for electric generation during periods of peak electric demand, or for uses in process steam applications. In previous work, the design of a sensible heat Thermal Energy Storage (TES) system was proposed for SMR applications (2). In this paper we examine coupling of the TES system to a SMR, typical of Integral Pressurized Water Reactor (IPWR) designs currently under development. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Thermal Hydraulic Modeling of an Advanced Nuclear Reactor using open-source MOOSE tools

The aim of this study is to develop model of a small High Temperature Gas cooled Reactor (HTGR) including the balance of plant. This microreactor produces electricity using the thermal power of the nuclear reaction. This work utilizes the MOOSE Multiphysics simulation tools, which are mainly developed at Idaho National Laboratory (INL, Idaho, United States of America). Its thermal-hydraulics and Heat conduction Modules are used to study the fluids behavior in the primary loop and power conversion system and their interactions with the heating structures. More specifically, one verifies that the temperatures, pressures and mass flow rates of the fluids in both loops are consistent and that, at the same time, all the power transfers occur as expected. Moreover, the various mechanical components characteristics (turbine, compressor, pump) are adapted to the reactor operating conditions. This first analysis is conducted using simplified and one-dimensional model for the core. Ultimately, the obtained results are used to build a higher fidelity core model. This step aims to verify that the initial simplified simulation is consistent with the three-dimensional core modeling. In addition, the calculated material temperatures are checked. This study provides a fairly complete model of the thermal hydraulic phenomena for a particular design of High Temperature Gas cooled Reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Once-Through Steam Generator Model Analysis Using Python and Advanced Optimization Tools (Summer Internship Report)

This study focuses on the parametric analysis of design parameters for a once-through steam generator (OTSG) model, using python and advanced optimization tools to facilitate applications such as the flowing autoclave steam generator (FASG) test cases. Building on previous research involving another OTSG with a different design, this project aims to enhance our understanding of how steam generators (SGs) behave and how their outputs are influenced by changes in design. The reason for this design change is to allow for more precise modeling and optimization of SG performance, to provide a comparative analysis between the two designs, and to set up the model for integration with the FASG test case. The OTSG python-model is a mathematical representation (including fluid flow and heat transfer equations/models/correlations) of a steam-generating unit in a pressurized water reactor-type small modular reactor system. Design studies involve changing the model’s input design parameters to observe the resulting effects on the output of the system. By using advanced optimization tools, such as the Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory, detailed design parametric studies and model optimization were performed. Six input parameters—pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid), respectively, of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (i.e., ±1%, ±5% and ±10% relative changes) for 600 samples. The analysis provides valuable insights into SG optimization and can be used for sensor placement optimization to effectively monitor and obtain experimental data in other tests.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2019

Atomic layer etching of SiO 2 with Ar and CHF 3 plasmas: A self-limiting process for aspect ratio independent etching

With ever increasing demands on device patterning to achieve smaller critical dimensions, the need for precise, controllable atomic layer etching (ALE) is steadily increasing. In this work, a cyclical fluorocarbon/argon plasma is successfully used for patterning silicon oxide by ALE in a conventional inductively coupled plasma tool. The impact of plasma parameters and substrate electrode temperature on the etch performance is established. We achieve the self-limiting behavior of the etch process by modulating the substrate temperature. We find that at an electrode temperature of -10°C, etching stops after complete removal of the modified surface layer as the residual fluorine from the reactor chamber is minimized. Finally, we demonstrate the ability to achieve independent etching, which establishes the potential of the developed cyclic ALE process for small scale device patterning.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Fabrication and Testing of Complex Silicon Carbide Composite Structures

The ability to fabricate larger complex silicon carbide composites is an enabling technology for advanced reactors. Processing techniques to make complex SiC/SiC structures must be versatile enough to facilitate fabrication of both very large components (>1 m), as well as very small feature sizes (<1 mm), minimize the amount of machining required to complete the final part, and must also provide composite material with similar performance to bulk properties previously established for planar SiC/SiC composite material. This paper presents results on fabrication of complex silicon carbide structures and targeted testing aimed specifically at understanding how complex features impact composite performance. Silicon carbide (SiC) and silicon carbide matrix, silicon carbide fiber-reinforced composites (SiC/SiC) are a strong candidate for cladding and core structure used in advanced reactor concepts such as General Atomics' (GA) Energy Multiplier Module (EM{sup 2}) as well as accident tolerant fuel (ATF) in LWRs. SiC retains its strength up to very high temperatures, is stable under irradiation, and shows favorable neutronics and oxidation resistance compared to alloy alternatives. However, unlike metals, SiC cannot be easily machined, formed or welded to make the more complex core components needed to fully realize the benefits of advanced high temperature reactors. TEP processing provides an appealing route to achieving SiC/SiC composite structures with enhanced feature complexity. This is a critical development in the maturation of SiC/SiC for in-core structural materials in advanced reactors. This processing route provides an avenue to make large, dense, complex structures. Experiments have shown retention of strength and pseudo-ductility with hole spacing as small as 4 mm with potential for still smaller feature spacing. The results indicate that feature spacing is not limited by processing effects, but instead by the fiber architecture. The lower limit of feature spacing will likely have implications on potential core loading limits in advanced nuclear reactors utilizing SiC/SiC. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2023

Unique safety features and licensing requirements of the NuScale small modular reactor

Small modular reactors (SMR) offer a novel approach to the construction and operation of nuclear power plants. The NuScale VOYGR™ plant uses a simplified SMR design that is based on proven light-water reactor technology with substantial improvements in nuclear safety. It consists of a 250 MWt reactor core housed with other primary system components in an integral reactor pressure vessel surrounded by a steel containment vessel, all of which is immersed in a large pool of water that also serves as the ultimate heat sink. At the core of the NuScale safety case are three primary safety systems: the decay heat removal system, the emergency core cooling system, and the containment. The ability of the NuScale Power Module (NPM) passive safety systems to remove core decay heat for an unlimited duration is demonstrated through analysis of a beyond-design-basis extended loss of AC power with no replenishment of water to the ultimate heat sink or operator actions. The NuScale methodology to evaluate an indefinite loss of AC power uses the proprietary NRELAP5 systems analysis computer code. Analysis results show that the reactor coolant system liquid level above the core is maintained and that containment pressure remains below the vessel design pressure. Once full passive air cooling is established, containment pressure and temperature will decrease over time with decreasing core decay heat. NuScale received standard design approval in September 2020 and design certification in January 2023 for its 50 MWe NPM configured as a 12 module plant. NuScale is currently seeking standard design approval to increase its core power to 250 MWt, nominally 77 MWe per module, in a 6-module plant (VOYGR™-6) configuration. The high-level safety of NuScale’s SMR technology is foundational to a new standard of nuclear power plant resilience.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Modular Design of Processing and Storage Facilities for Small Volumes of Low and Intermediate Level Radioactive Waste including Disused Sealed Sources - 12372

There are a number of IAEA Member States generating relatively small quantities of radioactive waste and/or disused sealed sources in application of nuclear techniques in medicine, industry and research and in nuclear research centres having small research reactors. At present many of these Member States do not have facilities for processing and storing their radioactive wastes; notably in those countries with small quantities of generated radioactive wastes. In other Member States the existing waste processing and storage facilities (WPSF) are in need of varying degrees of upgrading in order to address new waste streams, incorporate new waste processing technologies, or expand interim storage capacities. The IAEA has developed a modular design approach for a WPSF that is based on a variety of modules for different waste stream treatment and conditioning processes. The modular WPSF design is elaborated in a substantial Design Engineering Package that will be published by IAEA as a technical report. The Design Engineering Package enables users to select the optimum waste processing and storage modules to meet their needs, and to specify the requirements for procurement of individual modules and their integration into a waste processing and storage facility. The Design Engineering Package is planned for publication by the IAEA in 2012 and is presented as: - A Design Engineering Package Summary document. - A supporting CD that contains: - Process module general specifications. - Process module interface specifications. - Design Engineering Package for process modules. - Sample technical specifications for design and construction of modular processing facility. - Design Engineering Package for storage modules. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Spatial homogenization of thermal feedback regions in Monte Carlo reactor calculations

An integrated thermal-hydraulic feedback module has previously been developed for the Monte Carlo transport solver, MC21. The module incorporates a flexible input format that allows the user to describe heat transfer and coolant flow paths within the geometric model at any level of spatial detail desired. The effect that the varying levels of spatial homogenization of thermal regions has on the accuracy of the Monte Carlo simulations is examined in this study. Six thermal feedback mappings are constructed from the same geometric model of the Calvert Cliffs core. The spatial homogenization of the thermal regions is varied, giving each scheme a different level of detail, and the adequacy of the spatial homogenization is determined based on the eigenvalue produced by each Monte Carlo calculation. The purpose of these numerical experiments is to determine the level of detail necessarily to accurately capture the thermal feedback effect on reactivity. Several different core models are considered: axial-flow only, axial and lateral flow, asymmetry due to control rod insertion, and fuel heating (temperature -dependent cross sections). The thermal results generated by the MC21 thermal feedback module are consistent with expectations. Based upon the numerical experiments conducted it is concluded that the amount of spatial detail necessary to accurately capture the feedback effect on reactivity is relatively small. Homogenization at the assembly level for the Calvert Cliffs PWR model results in a similar power defect to that calculated with individual pin-cells modeled as explicit thermal regions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Highly Available Nuclear Power for Mission-Critical Applications

Some energy consumers require power on an anytime, all-year-round basis with a high level of certainty, including defense installations, isolated communities, and some industrial processes. For these customers, interruptions in electricity or heat can mean substantial financial loss or even loss of life. In the absence of grid-scale energy storage, a high level of power availability can be accomplished only through the robustness and redundancy of power generators. The NuScale small modular reactor design is well suited to provide highly available power because of several features related to both the nuclear steam supply system and the overall plant design. In analogy to Redundant Array of Independent Disks (RAID) systems used to provide highly reliable data storage, a NuScale plant can assure sustained power generation by virtue of its Redundant Array of Integral Reactors (RAIR).This paper describes the NuScale RAIR plant features and summarizes the results of a rigorous analysis of RAIR availability as a function of power or, conversely, the RAIR plant output power as a function of power availability. The analysis utilized the Matrix Laboratory code (MATLAB) and included probability distributions for the frequency and duration of module outages due to planned and unplanned events. Further, the study evaluated the impact of implementing turbine bypass rather than module shutdown and using one or more modules to supply house loads in the case of loss of off-site power (LOOP). Availability results are presented for a 12-module RAIR plant with and without turbine bypass enabled during a LOOP and for different possible connections to the off-site power distribution grid and dedicated service loads. Results indicate that a very high level of availability can be achieved at relatively high power output levels, regardless of turbine bypass and dedicated load connection, compared to the operating fleet.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Report on the Initial Fundamental Sodium-CO 2 Interaction Experiment

The U.S. Department of Energy (U.S. DOE) Advanced Reactor Concepts (ARC) Program is developing the Sodium-Cooled Fast Reactor (SFR) as a reactor that could dramatically improve performance in sustainability, safety, economics, security, and proliferation resistance. The near-term Program goal is to reduce technical barriers to improve the prospects of advanced technology reactor systems. In the long-term, integrated advanced reactor system designs will be developed that implement the chosen missions and complete the testing and demonstration necessary for licensing. The supercritical carbon dioxide (S-CO 2 ) Brayton cycle, coupled with an SFR, has been identified as a new and innovative energy conversion technology that could contribute to meeting these objectives especially in terms of improving economics. One appealing feature of this energy conversion system is the smaller footprint that the hardware requires relative to the traditional superheated steam cycle, which is in part due to the use of Printed Circuit Heat Exchanger TM (PCHE TM ) compact diffusion-bonded heat exchangers as the heat source heat exchanger (sodium-to-CO 2 ) as well as the recuperator and cooler modules. Although PCHEs have a high degree of structural integrity, the potential for leaks to develop between the sodium and CO 2 coolant channels in the secondary heat exchanger cannot be ruled out, leading to discharge of high pressure CO2 into the secondary coolant circuit. Due to the robustness of the PCHE design, catastrophic failure leading to CO2 jet blowdown into the intermediate sodium loop is not deemed likely. Rather, small cracks (or micro-leaks) may develop in which CO 2 will bleed into the secondary system at a relatively low rate and chemically react with the sodium. The 2012 Sodium Fast Reactor Research Plan produced by Sandia National Laboratory (Denman et al., 2012) identified the interaction between sodium and CO 2 as being among the highest priority areas of research for SFR accident initiators. In fact, the report rated both the experimental database and modeling capabilities of Na-CO 2 interactions at the lowest current state of knowledge compared with the other listed high-priority accident initiator items. In recognition of the anticipated failure mode for a PCHE sodium-to-CO 2 heat exchanger, an Argonne experiment program, called SNAKE, was initiated in FY 2010 to investigate the reaction characteristics between sodium and CO 2 under micro-leak conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Ultra-High-Performance Concrete And Advanced Manufacturing Methods For Modular Construction

Small modular reactors (SMR) allow for less onsite construction, increase nuclear material security, and provide a flexible and cost-effective energy alternative. SMR can be factory-built as modular components, and shipped to desired locations for fast assembly. This project successfully developed a new class of ultra-high performance concrete (UHPC), which features a compressive strength greater than 22 ksi (150 MPa) without special treatment and self-consolidating characteristics desired for SMR modular construction. With an ultra-high strength and dense microstructure, it will facilitate rapid construction of steel plate-concrete (SC) beams and walls with thinner and lighter modules, and can withstand harsh environments and mechanical loads anticipated during the service life of nuclear power plants. In addition, the self-consolidating characteristics are crucial for the fast construction and assembly of SC modules with reduced labor costs and improved quality. Following the UHPC material development, the capacity of producing self-consolidating UHPC in mass quantities was investigated and compared to accepted self-consolidating concrete standards. With slightly adjusted mixing procedure using large-scale gravity-based mixers (compared with small-scale force-based mixer), the self-consolidating UHPC has been successfully processed at six cubic yards; the product met both minimum compressive strength requirements and self-consolidating concrete standards. Steel plate-UHPC beams (15 ft. long, 12 in. wide and 16 in. deep) and wall panels (40 in. X 40 in. X 3 in.) were then constructed using the self-consolidating UHPC without any external vibration. Quality control guidelines for producing UHPC in large scale were developed. When the concrete is replaced by UHPC in a steel plate concrete (SC) beam, it is critical to evaluate its structural behavior with both flexure and shear-governed failure modes. In recent years, SC has been widely used for buildings and nuclear containment structures to resist lateral forces induced by severe earthquakes and heavy winds. SC modules have good potential for SMR because of their cost-effectiveness and reduced construction time. However, the minimum shear reinforcement (i.e. cross tie) ratio needs to be determined for the steel plate-UHPC (S-UHPC) beams to exhibit a ductile failure mode. In this project, S-UHPC beams were designed and constructed. The beams were tested to evaluate structural capacity and identify the minimum cross ties ratios. In addition, as the bond between UHPC and steel plate is essential for ensuring structural integrity under shear and flexure, it was measured and examined in this project through digital image correlation system and smart piezoelectric aggregate sensors. Large-scale testing and finite element simulation were also performed on S-UHPC wall panels. New bond slip-based constitutive models of steel plate were developed for S-UHPC, which were used in finite element analysis program to predict S-UHPC behavior under shear. The results were well validated through experimental data. The long-term durability of UHPC were established in this project. UHPC specimens were tested under free shrinkage, restrained shrinkage, elevated temperature, water permeation, chloride diffusion, corrosion, and alkali silica reaction. UHPC has demonstrated significantly improved durability compared with control concrete specimens. This research led to a new generation of steel plate-UHPC modules for SMR that can provide large benefits to the electric power industry. Taking advantage of the high strength and durability of UHPC, their modularity and ease of assembly can address the high cost barriers of typical nuclear power plants.

42 ENGINEERING↗