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DOE OSTI2020

Microreactor Agile Nonnuclear Experimental Testbed Test Plan

Microreactors are an attractive technology option for kick-starting nuclear innovation if they can be operated at high temperature, yielding high power conversion thermal efficiencies comparable or better than in commercial light water reactors. Microreactors are currently the smallest variation of Small Modular Reactors (SMRs). SMRs are “newer generation reactors designed to generate electric power up to 300 MWe and whose components and systems can be shop-fabricated and then transported as modules to the sites for installation as demand arises.” (IAEA, 2016). Vendors are developing microreactor designs to provide an affordable, potentially mobile source of electricity - see Fig. 1 for an example of a microreactor on a semi-truck. Various microreactor designs are possible including heat pipe- and gascooled options, which are the focus of the nonnuclear testing described in this document. In heat pipe microreactors, high-temperature heat pipes using liquid sodium or potassium working fluid transport fission heat from the core to a heat removal section which in turn transfers heat to the power conversion system working fluid. In a gas-cooled design, He or other gas will flow through a solid monolith of material and transfer heat as the temperature of the gas increases through a heat exchanger to a power conversion unit. The logistics of all these processes will be examined and tested through a series of articles at the nonnuclear test bed at Idaho National Laboratory (INL), the Microreactor Agile Nonnuclear Experiment Testbed (MAGNET) facility. Microreactors designed to produce power of 0.1-20 MWt offer the potential for more affordable nuclear energy for a range of applications. In a heat pipe microreactor, heat pipes, fuel rods, and/or moderator are intermixed in the reactor core assembly. Heat pipes extend from the core region into the heat removal section where the power conversion unit working fluid flows through holes or channels, transferring heat from the heat pipes to the working fluid. In a gas-cooled microreactor, gas flows through the solid monolith region and up into the heat exchanger region, transferring heat to the working fluid. For initial testing, the heat removal working fluid can be a low pressure gas for testing that addresses thermal stresses. In the final application, heat addition to the power conversion working fluid typically occurs at high pressure, supporting operation of an air-Brayton, supercritical CO 2 (SCO2), or He-recuperated Brayton cycle. Various stages of the steps above will be demonstrated through the tests described in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Critical response to M. Worrall et al. Published in Annals of Nuclear Energy 207 (2024) 110731

M. Worrall et al. recently published a manuscript titled “Fast neutron irradiation capability in existing thermal test reactors” (Worrall, 2024) that summarizes an irradiation vehicle design that would boost the fast neutron flux in the Advanced Test Reactor (ATR) for testing of nonfuel materials in a neutron flux energy spectrum that is more representative of fast reactors. Here, the authors compare their design with a separate vehicle design that they conceived of that would be implemented within the High Flux Isotope Reactor (HFIR). They analyzed both designs and drew conclusions on the most realistic near-term options for shielded nonfuel material irradiations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Upgrading Limiting Peak-Power Analysis Techniques with Modern Validation and Uncertainty Quantification for the Advanced Test Reactor

Here, this work demonstrates the acceptability of the 2D deterministic transport code, HELIOS, to replace the legacy diffusion code, PDQ, for computing the peak-power performance parameters of the Advanced Test Reactor (ATR). The 95% Confidence Rule, commonly used in the commercial reactor sector, is explored to develop the so-called “reliability factors” which provide statistical confidence that the peak-power limits within the hottest location along a fuel plate, referred to as the hot-stripe, will not be exceeded. Additionally, an alternative “legacy” methodology was explored that attempts to mimic the exact PDQ analysis process used for defining the peak-power limits. The legacy methodology, involves interpolating power between regions at azimuthal boundaries subtending the regions of interest. In order to apply the 95% Confidence Rule, a statistically significant calculation-to-measurement bias must first be established. Unlike the commercial world where thousands of power observations can be collected every cycle using on-line flux mapping instrumentation, the ATR power distribution must be measured during “depressurized” zero-power measurements using fission wires in polyethylene wands. In 2012, fission wire activation data was collected during a flux run in the Advanced Test Reactor – Critical facility. Also to improve statistical validity, archival data from ATR zero power flux runs from 1977, 1986, and 1994 were digitized from scanned reports and used to create new benchmark models. Borrowing from least-squares adjustment methods commonly used for neutron activation spectroscopy, adjusted fission wire powers were calculated for all four datasets. The mean and standard deviation of the bias between a priori calculated and adjusted wire-powers was then taken as the bias and uncertainty used in the 95% Confidence Rule.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Evaluation of Reduced Order Model for HT-9 Creep and Modifications to Current HT-9 Creep Model in BISON

In order to leverage existing reduced order models (ROM) for the modeling of LANL-developed HT-9 cladding constitutive behavior, mechanistic-based models have been implemented into BISON. It is posited that the increased fidelity of the ROM will show improvement over engineering-scale models in terms of comparison against experimental measurements. INL will assist LANL in implementing an HT9 ROM in BISON. Once the HT-9 mechanistic constitutive model can be leveraged in nuclear performance simulations, the numerical results of fast reactor models will be analyzed and compared against the Fuels Irradiation & Physics Database (FIPD) and separate effects tests to attempt a validation process. INL and LANL will work together to perform any necessary improvement that is identified during the implementation and use processes. Successful completion of the milestone will enable the general use of validated mechanistic ROMs for HT-9 cladding in BISON. This report provides a brief introduction to the Los Alamos Reduced Order Model Applied to Nonlinear Constitutive Equations (LAROMANCE) code, mechanical testing used to illustrate its creep and plastic deformation predictions for HT-9 cladding, and a comparison with existing models implemented into BISON. Current models have been modified to more appropriately account for primary thermal creep and results for corresponding testing are included in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Preconceptual Design of Irradiated Fuel Salt Management System

The National Reactor Innovation Center was established by the U.S. Department of Energy to accelerate the demonstration and deployment of advanced nuclear reactors. To meet this mission, the National Reactor Innovation Center has developed the Laboratory for Operations and Testing in the U.S. (LOTUS) to support the first fast-spectrum molten salt reactor demonstration. The goal of this research is to understand the system requirements that may be applied to a molten salt reactor experiment in the LOTUS test bed and to perform systems analyses and preconceptual development of a storage container for the management of the irradiated molten salt fuel. This study proposes a set of irradiated fuel salt management processes for the molten salt reactor, from defueling to storage or disposal. For that, the historic Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has served as a crucial source of data for a mock application. This paper proposes a potential geometry for storing irradiated fuel salt based on MSRE data and the Monte Carlo N-Particle radiation transport code. The geometry was analyzed for criticality safety under abnormal situations, such as water ingress and interaction between containers. Based on the determined configuration of the storage system, the decay heat produced from the irradiated fuel salts was calculated to provide a reasonable time scale of the dormancy period. Using the same geometry assumed above, the neutron and gamma dose rates from the irradiated fuel salt at the end of the dormancy period were estimated, which will be useful for understanding the handling and radiation protection requirements for handling the irradiated fuel salt. Lastly, a reactor in the LOTUS test bed could use chloride-based fuel as opposed to the fluoride-based fuel used in MSRE. Thus, a literature study provided data to understand the similarities and differences between the characteristics of fluoride- and chloride-based salts in terms of salt chemistry, material characteristics, postirradiation behavior, neutronics, thermal hydraulics, and heat removal. This study, together with the proposed irradiated fuel salt management processes, preliminary criticality calculations, decay heat, and dose rate estimates, provides a strong basis for future investigation into the management of irradiated fuel salts generated during molten salt reactor technology demonstrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

An Evaluation and Qualification of U.S.-Based Research Reactors for Irradiation Capabilities Supporting Advanced Nuclear Systems

Irradiation experiments are a prerequisite for evaluating nuclear reactor system designs, analyzing the performance of these systems, and obtaining licenses. Likewise, irradiation facilities are necessary for producing the radioisotopes used in industrial and medical applications. Recent developments in modeling and simulation capabilities and advancements in computational resources have further enabled the design of irradiation experiments for evaluating radiation-induced phenomena and determining nuclear fuel, material, and system design and safety criteria pertaining to both normal and accident scenarios. These computational tools and models require comprehensive experimental datasets acquired under prototypic radiation conditions—for exploring material and system performance under the uniquely harsh environments found in nuclear reactors—to enable verification and validation for qualification and licensing purposes. However, qualification of irradiation experimental facilities, primarily research and test reactors (RTRs), necessitates that their performance be evaluated based on the irradiation environment (e.g. flux, power, testing capabilities) using an appropriate scoring matrix. Although many university campus RTRs are available for research and development (R&D) activities and initiatives, this study focuses on evaluating and qualifying the irradiation facilities (mostly RTRs) within the United States that are suitable for advanced nuclear fuel, material, and system irradiation experiments aimed at establishing operational-performance limits and informing component and fuel designs so as to improve operational efficiencies and mitigate proliferation vulnerabilities, as well as for radioisotope production aimed at multipurpose applications. As a result, the findings of the present study support the acceleration of nuclear fuel and material qualifications, thus hastening new and advanced nuclear energy system demonstrations and radioisotope production efforts by using extended R&D.

irradiation experiment↗

DOE OSTI2025

Advanced Test Reactor Neutron Dosimetry Report - 173B (Rev. 1)

Results of neutron dosimetry measurements in ATR for cycle 173B-1&2 are presented. Cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

173B↗

DOE OSTI2021

Infrastructure and Testing Needs for Molten Salt Reactor Safeguards

Molten salt reactors (MSRs) are an attractive advanced reactor technology because of their increased fuel utilization and efficiency, the ease of fueling, and the potential to use alternative fuel cycles, including thorium and breeder/burner scenarios. Since an MSR has never been domestically operated for commercial electricity production, minimal infrastructure exists to support performance and materials testing, reactor licensing (including research, test, and operating reactors), and technology development in general. These factors could present challenges for the commercialization of the proposed design technology. Several research and development efforts and investments have begun to identify and supplement the infrastructure to accommodate the molten salt concept, including the United States Department of Energy Office of Nuclear Energy Molten Salt Reactor Program. This report reviews safeguards technology and domestic infrastructure to support the evaluation of nuclear safeguards (primarily material control and accounting) measures and approaches for salt-fueled MSRs. Recommendations for potential priority infrastructure investments, including safeguards measurement test beds and future research, are presented.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Critical heat flux impact due to cladding surface condition

This paper demonstrates how the cladding surface condition can impact fuel rod heat transfer and critical heat flux (CHF) performance at Pressurized Water Reactor (PWR) type conditions. The cladding surface that is investigated here is a surface with a crud deposit similar to ones observed in a PWR reactor. Tests are performed in the Westinghouse WALT (Westinghouse Advanced Loop Tester) loop test facility where a single ZIRLO clad electrically heated rod with approximately one-foot length is tested at PWR thermal hydraulic conditions and chemistry. Tests are performed for a normal clean rod surface, an oxide surface and with CRUD deposits to investigate the impact on fuel rod heat transfer and CHF performance. The CRUD deposit is characterized for thickness, chimney size and porosity. Test results are presented for the different cladding surface conditions along with comparisons to evaluations performed in the literature for different surfaces at pool boiling conditions. The results indicate that the WALT loop test facility can produce CHF test results which are repeatable for the same heater rod and for different rods with the same surface condition. CHF test results for a ZIRLO cladding surface with a CRUD deposit near 20 microns typical of PWR deposits show similar CHF performance to clean rod surfaces. A noticeable increase in CHF performance was observed for a ZIRLO cladding surface with a thermal oxide layer based on one CHF measurement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

MRS: Surveillance Test Article Development

The presentation covers work performed in FY23 on surveillance test article development. Surveillance test articles have been used of integrity assessment in light water reactor plants, this work seeks to develop a specimen that is passively loaded and could be used in advanced nuclear reactors for structural material health monitoring.

36 MATERIALS SCIENCE↗

DOE OSTI2017

Availability of Neutronics Benchmarks in the ICSBEP and IRPhEP Handbooks for Computational Tools Testing

In the past several decades, numerous experiments have been performed worldwide to support reactor operations, measurements, design, and nuclear safety. Those experiments represent an extensive international investment in infrastructure, expertise, and cost, representing significantly valuable resources of data supporting past, current, and future research activities. Those valuable assets represent the basis for recording, development, and validation of our nuclear methods and integral nuclear data [1]. The loss of these experimental data, which has occurred all too much in the recent years, is tragic. The high cost to repeat many of these measurements can be prohibitive, if not impossible, to surmount. Two international projects were developed, and are under the direction of the Organisation for Co-operation and Development Nuclear Energy Agency (OECD NEA) to address the challenges of not just data preservation, but evaluation of the data to determine its merit for modern and future use. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was established to identify and verify comprehensive critical benchmark data sets; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data [2]. Similarly, the International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications [3]. Annually, contributors from around the world continue to collaborate in the evaluation and review of select benchmark experiments for preservation and dissemination. The extensively peer-reviewed integral benchmark data can then be utilized to support nuclear design and safety analysts to validate the analytical tools, methods, and data needed for next-generation reactor design, safety analysis requirements, and all other front- and back-end activities contributing to the overall nuclear fuel cycle where quality neutronics calculations are paramount.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS2004

Deflection Measurements of a Thermally Simulated Nuclear Core Using a High-Resolution CCD-Camera

Space fission systems under consideration for near-term missions all use compact. fast-spectrum reactor cores. Reactor dimensional change with increasing temperature, which affects neutron leakage. is the dominant source of reactivity feedback in these systems. Accurately measuring core dimensional changes during realistic non-nuclear testing is therefore necessary in predicting the system nuclear equivalent behavior. This paper discusses one key technique being evaluated for measuring such changes. The proposed technique is to use a Charged Couple Device (CCD) sensor to obtain deformation readings of electrically heated prototypic reactor core geometry. This paper introduces a technique by which a single high spatial resolution CCD camera is used to measure core deformation in Real-Time (RT). Initial system checkout results are presented along with a discussion on how additional cameras could be used to achieve a three- dimensional deformation profile of the core during test.

Stanojev, B. J.↗

DOE OSTI2017

Upcoming Irradiation of Ultrasonic and Fiber Optic Sensors in the MITR

Fiber optic and ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. These technologies, which have performed successfully in out-of-pile tests, have not been well qualified in a test reactor environment. The uncertainty of successful operation of instruments at this level of development has typically been a significant barrier to in-core deployment. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of these technologies to provide enhanced sensors for irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. The National Scientific User Facility (NSUF) is sponsoring a project to evaluate the performance of promising temperature sensors based on ultrasound and on fiber optics during an upcoming irradiation test in the Massachusetts Institute of Technology Research Reactor (MITR). Several types of thin-wire metallic waveguide ultrasonic thermometers will be tested along with fiber optic temperature sensors using regenerated Fiber-Bragg Grating technology. Additionally, samples of several new optical materials which are potentially much more radiation tolerant than commonly used materials, such as silica, will be included and analyzed post-irradiation. The irradiation is a collaborative effort between researchers from Idaho National Laboratory, the French Alternative Energies and Atomic Energy Commission, the University of Pittsburgh, the Massachusetts Institute of Technology, and AFO Research, Inc.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

NASA NTRSDate not supplied

An Overview of Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Capabilities, Upgrades, and Overlaps in Advanced Material Testing Areas

The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) was designed to perform non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and materials. NTREES can simulate the thermal hydraulic environment within the reactor of an NTR and has been leveraged to provide data on the thermochemical, thermomechanical and thermohydraulic performance of components and materials. Active upgrades to the NTREES facility include installing a DC power supply for DC powered joule heating, a supply gas chiller for a cryogenically cool gas supply, gas pre-heaters with sample holders for testing samples too small to be inductively heated by bathing them in hot gas, and a high temperature digital image correlation (DIC) capability for measuring high temperature strains real time. In addition to NTR testing, given the commonality of high enthalpy flow environments between the reactor of an NTP engine and the aerothermal heating of high-speed flight, the unique capabilities of NTREES to provide high temperature gas flows at high pressure without combustion or electrode spallation byproducts may lend itself to additional usage of the facility beyond the original intent.

Michael P Schoenfeld↗

DOE OSTI2026

Advanced Test Reactor Neutron Dosimetry Report - Cycle 175B-1

Results of neutron dosimetry measurements in ATR for cycle 175B-1 are presented. A single set of cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

175B↗

DOE OSTI2019

Considerations for Canister Design and Drop Testing

This paper discusses some considerations for the proposed design and analysis of canisters used to ship reactor debris from the Fukushima Daiichi nuclear plant. Energy-limited events, such as cask drops and other dynamic impacts, are evaluated much differently than other structures using static analysis. Static analysis of structures makes sense for engineers because static analysis is intuitive and material typically behaves in a linear elastic manner. Dynamic analyses are typically not intuitive because of non-linear material behavior. This is why canister designs need to be evaluated by dynamic finite element analysis and/or laboratory drop tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

SAS4A/SASSYS-1 Lead Verification Test Suite

SAS4A/SASSYS-1 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 forty-five years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. Although SAS4A/SASSYS-1 was developed to support the analysis of any liquid-metal-cooled nuclear reactor, it has primarily been utilized to design and analyze Sodium Fast Reactors (SFRs). As a result, most of the verification basis for SAS4A/SASSYS-1 utilizes sodium as a coolant and geometry descriptions that are typical of a SFR facility. In this work, new verification problems are defined to extend the applicability of the SAS4A/SASSYS-1 verification basis into Lead Fast Reactor (LFR) design space. A review of the existing SAS4A/SASSYS-1 SFR verification test suite is performed to determine the number of test cases that are SFR specific. SFR specific test cases are replicated using lead as a coolant and system layouts that are representative of LFRs. This report provides definitions and reference solutions for seven new test cases that scope lead-specific features and capabilities of the software. For a verification test to be considered acceptable, comparisons of analytical solutions to SAS4A/SASSYS-1 predictions must produce negligible or justifiable errors. All LFR specific test cases are found to be acceptable.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Safeguards by Design Projects (Final Report-FY-22)

This University Engagement project challenged engineering students at universities, that do not have Bachelor degree programs in nuclear engineering but do have research reactors and some nuclear engineering coursework, to incorporate Safeguards by Design concepts into their Senior Capstone Design Project. This University Engagement project was part of the U. S. Department of Energy’s (DOE) National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation, Office of International Nuclear Safeguards, Next Generation Safeguards Initiative (NGSI), Human Capital Development (HCD): University Engagement Program. This program exposed university students with Mechanical Engineering majors and Nuclear Engineering minors to the concepts of international nuclear safeguards. In FY22, three teams at the University of Rhode Island and two teams at the University of Texas - Austin participated in researching, designing, building, and testing projects to support international nuclear safeguards measurements or verification. The projects involved engaging in activities at the university’s research reactors. All the projects engaged students with prototyping a design and/or tool for application at the Universities’ reactor. At the end of the course, most of the students expressed the experience was positive and they learned more about international nuclear safeguards and applying requirements than they had previously encountered. This school year the projects were further complicated by the COVID-19 pandemic. Both universities had limited in classes on campus, still relying on Zoom classes, and limited direct student/professor interactions. Furthermore, Los Alamos National Laboratory (LANL) greatly restricted travel, therefore making it impossible to visit the students at the end of the semester for the review of their design projects. The final design and review meeting for the projects happened via meetings over the internet. Additionally, the teams did build and test some prototypes but could only do so in a limited capacity.

42 ENGINEERING↗