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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

A hybrid surrogate modeling framework for the Digital Twin of a Fluoride-salt-cooled High-temperature Reactor (FHR)

While nuclear energy is a non-greenhouse-gas emitting energy source, expensive operational costs due to the high-level of safety requirements decreases their competitiveness in the sustainable energy market. Advanced reactor concepts paired with Digital Twins aim to increase the commercialization gains of nuclear energy by reducing operational costs, increasing reactor reliability and enhancing power generation. To support Digital Twin tasks such as real-time autonomous control, proactive maintenance monitoring or optimizing power demand operations, a fast and accurate virtual representation of the Nuclear Power Plant (NPP) is required. The computational cost of high-fidelity, physics-based models are unsuitable for real-time analysis or scalability. Here, in this work, a hybrid surrogate modeling framework is developed fora Fluoride-salt-cooled High-temperature Reactor (FHR) that leverages physics-inspired models for key reactor components and uses data-driven methods for rapid system state space prediction. The Xenon reactivity feedback model is integrated to inform the surrogate model about the reactor core and the homologous pump theory model is the basis for representing pump degradation. Using a detailed, two dimensional thermal hydraulics model to generate data on the FHR, we train a network of Vectorized Autoregressive Moving-Average with eXogenous input (VARMAX) models to predict the remaining state values. The result is a surrogate model that provides a detailed reactor state representation of 41 system states and a pump degradation analysis. The framework is applied to Load Follows profiles, yielding high accuracy and a speedup that is more than 4000x faster compared to the higher- fidelity thermal hydraulics model, enabling real-time operational intelligence and applications in long horizon predictions. While the surrogate model framework is demonstrated for the particular case of FHR, the hybrid physical/data-driven modeling approach including the network of surrogates and the underlying modularity has the potential to be applied to other physical asset systems.

Digital Twins↗
DOE OSTI2025

Decay heat analysis for advanced reactor spent fuel transportation and storage applications

Accurate characterization of nuclide inventories and decay heat in spent nuclear fuel is critical for ensuring its safe handling, storage, transportation, and disposal. Although extensive research has been conducted on light-water reactor fuel, advanced reactors present unique challenges due to their diverse core configurations, fuel characteristics, neutron energy spectra, and burnup levels. Building upon previous efforts that developed representative reactor core models for various advanced reactor types and fuels, this study evaluates reactor-specific decay heat characteristics. The results highlight significant variations across advanced reactor types as well as across reactor designs within the same reactor type, and they provide comparison to typical commercial light-water reactor fuel. For example, thermal-spectrum reactor fuels were observed to have an approximately 100-fold decrease in decay heat over the first decade of cooling, whereas the reduction was 10-fold for fast-spectrum reactor fuels. Mass-specific decay heat at discharge can differ by three orders of magnitude among the fast and thermal reactor systems considered. Overall, for the analyzed advanced reactor fuel, fewer than 17 nuclides account for over 99% of total decay heat at 0.5 years, and that number drops to fewer than 7 nuclides at 100 years of cooling. By quantifying reactor-specific decay heat trends and nuclide contributions, this work provides a technical basis to support the development of spent fuel management strategies for advanced reactor fuels as well as safety evaluations for storage, transportation, and long-term waste disposal.

Advanced reactors↗
DOE OSTI2024

Comparative Analysis of Heat Exchangers for High Temperature High Pressure Helium Component Testing

Helium gas loops have been designed and built to gain a better understanding of the gas thermohydraulic phenomena that take place in a helium system. Some of these loops are used for validation and testing of components for high temperature gas-cooled reactors (HTGRs). However, most of them operate at lower pressure and temperature than an HTGR. While these loops can provide valuable information about gas-cooled reactor components, the operating envelope of the experiment is constrained by the maximum operating conditions of the helium loop. In response to the lack of an experimental facility that can provide the infrastructure needed to validate and test components at nominal pressures and temperatures of HTGRS, the HElium Component Testing Out-of-pile Research (HECTOR) facility was designed at Idaho National Laboratory with the assistance of University of Idaho and Walsh Engineering. With the capability to test at temperatures up to 800°C and pressures of 8MPa, HECTOR serves as a critical tool for the advancement of HTGR technology. The facility's primary role is to provide a controlled, high-fidelity environment for the assessment of component resilience and efficiency under nominal HTGR conditions. In the quest to enhance the efficiency and performance of HECTOR, a comparative analysis of three distinct types of heat exchangers—shell and tube, offset strip fin, and printed-circuit—was conducted, focusing primarily on two critical metrics: the required surface area and pressure drop characteristics. The shell and tube heat exchanger, renowned for its robust design and widespread industrial application, was evaluated against the offset strip fin and the cuttingedge printed-circuit heat exchangers, both of which are lauded for their compactness and thermal effectiveness. This comparative study aims to provide detailed insights into the thermal management capabilities of each heat exchanger type under the conditions inherent to HECTOR, thereby facilitating an informed selection for systems demanding high operational integrity and efficiency.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Bi‐Continuous W‐Rich Refractory High Entropy Alloy‐Cu Composite: Toward Material Innovation of Nuclear Reactor Coolant System

Abstract Refractory high‐entropy alloys (RHEAs) are considered promising candidate materials for next‐generation nuclear reactors due to their superior mechanical strength, irradiation resistance, and thermal stability at high temperatures. However, the significant positive heat of mixing between refractory alloying elements and Cu, commonly used in cooling systems, poses challenges in forming composite structures. This study addresses the issue using a liquid metal dealloying (LMD) process. A precursor alloy (WTaVTi) with a directional dendrite‐interdendrite structure is fabricated and reacted with molten Cu at 1200 °C for 96 h. This approach produced a RHEA‐Cu composite with a stable interface between RHEA (W 31.5 Ta 30.9 V 21.4 Ti 14.3 ) and Cu, featuring a spontaneously formed W‐rich interlayer that enhances interfacial bonding. The composite showed excellent irradiation resistance, with 30% less swelling under α‐ion irradiation than pure W. It also exhibited low thermal conductivity at room temperature, but reached ≈120 W m −1 ·K −1 at ≈650 °C, surpassing pure W. This temperature‐dependent rise in κ, with a positive gradient of +0.075 W m −1 ·K − 2 , is attributed to decreasing diffuse mismatch at elevated temperatures. The large‐scale reaction and stable microstructure achieved through LMD process highlight its industrial potential. This work offers a strategy for developing high‐performance materials by combining RHEA's radiation resistance with Cu's thermal conductivity for extreme environments.

Chemistry↗
DOE OSTI2023

Tensile behavior of diffusion bonded AA6061 - AA6061 with variation in cooling method

Hot isostatically pressed AA6061 cladding is an important structural component of the high performance, Zr-laminated U-10Mo monolithic fuel system for the application in research and test reactors. In this study, the mechanical behavior of two diffusion bonded aluminum alloy, AA6061, was examined using tensile testing. Solid-to-solid diffusion bonding between two pieces of AA6061 was performed by isothermal annealing at 560 °C for 1.5 h, and diffusion couples were subsequently cooled via three different cooling methods: furnace cooling, air cooling, and water quenching. Dog-bone shaped tensile specimens, with 10 mm in gauge length (with diffusion bonded interface in the middle), and 1.5 × 1.5 mm 2 gauge cross-sections, were fabricated from the diffusion bonded AA6061 by electro-discharge machining. Yield strength (% EL at failure) of furnace cooled, air cooled and water quenched tensile specimens determined was 82 ~ 89 MPa (10 ~ 30%), 112 ~ 116 MPa (10 ~ 14%), and 149 ~ 164 MPa (10 ~ 17%), respectively. This variation in mechanical behavior was examined with cooling-rate dependent, concentrated precipitation of Mg 2 Si at the diffusion bonded interface, with due respect for mechanical properties of the AA6061 alloy that inherently vary as a function of cooling rate from 560 °C. Finite element analysis using ABAQUS was employed to augment experimental findings with the appropriate microstructural constituents and alloy properties. Results suggest that the strength is dominated by matrix/bulk properties of AA6061, while ductility is strongly influenced by the cooling method dependent presence of Mg 2 Si precipitates at the interface.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Comparison of Irradiated TRISO Fuel Radioactivity from Multiple Advanced Reactor Designs

The SCALE code system was used to model, deplete, and compare several different TRISO-fueled reactor designs: a helium-cooled prismatic reactor, a helium-cooled pebble-bed reactor (PBR), and a Fluoride-Lithium-Beryllium (FLIBE) molten-salt-cooled PBR. The purpose of this comparison was to understand how differences in the reactor designs affect the radioactivity of the fuel after discharge and whether those differences are significant. First, the various reactor designs were build and depleted in the TRITON module for each design and fuel enrichment. Then, the TRITON outputs were used to create burn-up dependent reactor libraries. These libraries were then used by ORIGEN to determine the activities of discharged fuel for each reactor, which were compared to generic Westinghouse 17x17 fuel. Overall, the results showed that short term decays are dominated by reactors with higher operating powers, and the reactor type, initial fuel enrichment, and maximum burn-up are of only secondary importance. Although this analysis only focuses on activities in becquerels, these dependencies are consistent with the expected behavior of decay heat. However, analysis of long- term time periods post-irradiation shows that the reactor type and maximum burn-up have strong impacts on the activities; initial fuel enrichment has a secondary impact while operating power is inconsequential. These results would be useful for analyses, such as dose assessment and modeling in post- release scenarios; normal fuel handling operations; and spent fuel transport, storage and disposal. Of particular interest, the results in this report show that analyses that focus on spent nuclear fuel of advanced reactors need to consider each parameter carefully. Unsurprisingly, if the correct operating power is not used in short term analyses, the results will not be correct. Perhaps unexpectedly, however, if the correct reactor type is not used, then the long term results will also be incorrect, especially for areas such as permanent disposal. Even though this report focuses on the total activity of nuclear fuel, it provides initial results on the effects of various input parameters and also provides a framework to extend the work into other analyses of spent fuel from advanced reactors, especially those employing TRISO fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Multiphysics Analyses of the Protected and Unprotected Loss of Forced Cooling Accidents in the HTR-PM

Here, we present the multiphysics simulation results for the protected pressurized and depressurized loss of forced cooling (PLOFC and DLOFC) events in the High-Temperature gas-cooled Reactor--Pebble-bed Module (HTR-PM) equilibrium core using the Griffin-Pronghorn coupled code system. Additionally, this paper discusses the strategy for estimating the spontaneous fission neutron source needed for unprotected events and re-criticality calculations. The solutions of the protected PLOFC and DLOFC events were verified against similar solutions obtained for temperature evolutions from the open literature. Both the average and maximum pebble surface temperatures behaved as expected during the DLOFC and remained below 1800 K. The PLOFC results are highly dependent on the ability to resolve the natural circulation in the core, which is impacted by the mesh resolution in Pronghorn. Furthermore, we present the results of the unprotected DLOFC transient to predict the timing of the re-criticality event, which occurred 47 hours after the onset of the transient, and the new steady-state power of 1.3 MW.

42 - ENGINEERING↗
DOE OSTI2025

Solid structure of Li 2 BeF 4 (FLiBe) from room temperature to melting studied by neutron and X-ray diffraction

Molten fluoride salts such as Li 2 BeF 4 (FLiBe) are used in molten salt reactors, fluoride-salt-cooled high-temperature reactors and fusion reactors as a fuel solvent, coolant and/or tritium breeding medium. In engineered systems that use molten salt, solid-state material will be present during melting and freezing scenarios, and therefore the temperature-dependent properties of the solid and solid/liquid phase transition merit investigation. To observe the behavior of the solid state of Li 2 BeF 4 from room temperature to melting, this work used neutron and X-ray diffraction to measure the changes in the lattice parameters and volume of the crystalline unit cell and compared the results with prior low-temperature data for solid Li 2 BeF 4 . From neutron diffraction data it is also possible to identify anisotropy: centimetre-scaled crystals align preferentially with the a axes parallel to the direction of freezing front propagation, and the c axes expand 54% more than the a axes. This work provides the lattice constants as a function of temperature, quantifies the thermal expansion, and determines the equation describing the change in density for solid Li 2 BeF 4 from room temperature to 459°C to be ρ solid (kg m −3 ) = 2182 (3) − 0.115 (2) T (°C) and the volume expansion upon melting to be less than 5%. This density changes depending on molecular weight and enrichment.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Improvements of SAM Heat Transfer Models for Molten Salt-Cooled Pebble Bed Reactors

To support the development and utilization of the SAM code for molten salt-cooled pebble bed reactor safety analyses and licensing, selected improvements were made to the one-dimensional system-level heat transfer modeling in SAM with the goal of better capturing heat transfer phenomena specific to these designs. In consideration of molten salt as a coolant, new models were introduced to capture the enhanced radiative heat transfer that must be considered at the higher working temperatures. Radiative heat transfer is handled not only between two structural surfaces, but also in combination between the surfaces and the molten salt fluid, which has a higher opacity and absorption of radiative heat as compared to water. Several examples of analytical verification of the new radiative heat transfer methods are included in this report. Additionally, convective heat transfer correlations were selected from literature and included in SAM to account for the enhanced heat transfer expected in pebble bed geometry flows. These correlations are verified and demonstrated in the present work with a reference FHR example.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Degradation and Failure Phenomena of Advanced Reactor Fuel Concepts: Sodium-Cooled Fast Reactor Metallic Fuel

The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of new fuel types in advanced nuclear power reactors that would be designed and built in the United States. Pacific Northwest National Laboratory (PNNL) is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs that would be deployed in these reactors. This report focuses specifically on the metallic fuel that is being considered for sodium-cooled fast reactors and specifically on mechanisms that would cause damage or failure to the fuel under reactor operating conditions and design basis accident conditions. There is historic experience with both metallic and oxide fuels in experimental sodium-cooled fast reactors, but this report will focus solely on metallic fuel. Currently two U.S.-based reactor designers are engaged with the NRC in the application or pre-application review and considering a sodium-cooled fast reactor. TerraPower is engaged with NRC in pre-application review of its Natrium reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel clad in HT9 stainless steel. ARC Clean Technology is engaged with NRC in pre-application review of its ARC-100 reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel with steel cladding. This report will focus on this fuel system specifically, with broader information given regarding other metallic fuel systems with other stainless steel alloy claddings. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of metallic fuel concepts for sodium-cooled fast reactors, including fuel performance characteristics that may not be addressed within existing regulatory documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Thermal Gradient Effects on Redox Evolution and Volatility-Driven Fractionation in Ternary U/Ce/Cs Condensates

Understanding how thermal history influences redox evolution and chemical fractionation is essential for characterizing high-temperature condensation in complex materials, including nuclear debris. Here, we tested the hypothesis that distinct thermal regimes in a plasma flow reactor influence redox pathways and elemental partitioning in ternary U/Ce/Cs systems. A configurable plasma flow reactor was modified with an external tube furnace to impose two distinct thermal gradients: continuous ambient cooling and a furnace-assisted thermal hold-up near 1400 K followed by rapid cooling. Transmission electron microscopy characterized phase identity, morphology, and nanoscale element distributions, while inductively coupled plasma-mass spectrometry quantified bulk elemental ratios. Across both thermal regimes, uranium and cerium condensed as UO 2 and CeO 2 as dominant refractory oxide products. Uranium partially oxidized to α-UO 3 during extended ambient cooling, while furnace-assisted hold-up preserved UO 2 and produced partial reduction of cerium to Ce 2 O 3 . Cesium remained volatile upstream and condensed later in the reactor, forming Cs 2 O and Cs-uranate phases with the highest incorporation after thermal hold-up. Bulk ICP-MS measurements supported these observations. U/Ce ratios remained comparatively stable and Cs displayed delayed and apparent transient enrichment that matched the nanoscale measurements. This integrated approach provides a quantitative method for linking thermal gradients to redox evolution and volatility-driven fractionation. These results show how the plasma flow reactor can identify where equilibrium descriptions remain adequate and where kinetic effects from residence time and temperature history must be considered when interpreting condensation behavior in multicomponent systems.

and nuclear chemistry↗
DOE OSTI1994

Nuclear Safety [Vol. 35, No. 2, July-December 1994]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 179 Consideration of Postaccident Consequences in the Determination of Safety Objectives for Future Nuclear Power Plants in France, D. Queniart, A. Sugier, and J. Lochard; ACCIDENT ANALYSIS: 187 Nuclear Safety Research: The Phebus FP Severe Accident Experimental Program, P. von der Hardt, A. V. Jones, C. Lecomte, and A. Tattegrain; 205 Containment Performance Analysis of the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory, S. H. Kim, R. P. Taleyarkhan, and V. Georgevich; 213 Assessment of Fission Product Deposits in the Reactor Coolant System: The DEVAP Program, G. Le Marois and M. Megnin; 222 Erratum to “A Review of the Available Information on the Triggering Stage of a Steam Explosion," Vol. 35, No. 1, CONTROL AND INSTRUMENTATION: 223 Effects of Normal Aging on Calibration and Response Time of Nuclear Plant Resistance Temperature Detectors and Pressure Sensors, H. M. Hashemian; DESIGN FEATURES: 235 Defense in Depth Against the Hydrogen Risk—A European Research Program, F. Fineschi; ENVIRONMENTAL EFFECTS: 246 Technical Note: A Preliminary Analysis of the Risks to Hong Kong Resulting from Potential Accidents of Daya Bay Nuclear Power Plant, Z. Shi and X. Wei; OPERATING EXPERIENCES: 253 Reactor Shutdown Experience, Compiled by J. W. Cletcher; SPECIAL SECTION ON TMI-2 VESSEL INVESTIGATION PROJECT: 256 Three Mile Island—New Findings 15 Years After the Accident A. M. Rubin and E. Beckjord; 269 Relocation of Molten Material to the TMI-2 Lower Head, J. R. Wolf, D. W. Akers, and L. A. Neimark; 280 Insight Into the TMI-2 Core Material Relocation Through Examination of Instrument Tube Nozzles, L. A. Neimark; 288 Physical and Radiochemical Examinations of Debris from the TMI-2 Lower Head, D. W. Akers and B. K. Schuetz; 301 Results of Metallographic Examinations and Mechanical Tests of Pressure Vessel Samples from the TMI-2 Lower Head, D. R. Diercks and G. E. Korth; 313 Margin-to-Failure Calculations for the TMI-2 Vessel, J. Rempe, L. Stickler, S. Chavez, G. Thinnes, R. Witt, and M. Corradini; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 328 1993 Accident Sequence Precursor (ASP) Program Results, L. N. Vanden Heuvel, J. W. Cletcher, D. A. Copinger, J. W. Minarick, B. W. Dolan, and P. D. O’Reilley; RECENT DEVELOPMENTS: 339 Reports, Standards, and Safety Guides, D. S. Queener; 345 Proposed Rule Changes as of June 30, 1994; ANNOUNCEMENTS: 234 Thirty-First Annual Meeting of the National Council on Radiation Protection and Measurements; 356 1995 International Incineration Conference; 358 Fifth International Controls and Instrumentation Conference; 359 ANS International Topical Meeting on Safety of Operating Reactors; 359 Fifth International Conference on Nuclear Criticality Safety; 360 International Conference on Probabilistic Safety Assessment Methodology and Applications; 351 The Authors; 357 Reviewers of Nuclear Safety, Vol. 35.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Molten Salt Reactor Signatures and Modeling Study

Molten salt reactor (MSR) technologies, either liquid fueled and cooled or only liquid cooled, pose specific, unique challenges for safeguards of the special nuclear material during the operation, fueling, and maintenance of the reactor. MSRs are one type of Generation IV technologies being invested in and considered for U.S. domestic fabrication primarily for electricity and process heat production. These designs have generated growing commercial interest for several reasons, including high (≈40%) thermal efficiency, ease of fueling, improved use of uranium fuel, potential utilization of thorium fuel, and proposed inherent safety features. U.S. companies have several planned designs that differ in fuel, cooling, and neutron energy spectrum. Driven by commercial interest and the intent of licensing MSRs, the U.S. Nuclear Regulatory Commission (NRC) has developed a vision and strategy to accommodate non-light water reactors (LWR), which include MSRs (U.S. Nuclear Regulatory Commission 2019). In addition, the Department of Energy (DOE) through the Office of Nuclear Energy (NE) Office of Advanced Reactor Technologies (ART), …sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. Reactor types considering the use of salts, liquid metals, or gases for coolant fall under both ARC and aSMR. Therefore, Research Design & Development is being pursued by DOE-NE through national laboratories, universities, and international and industrial collaborations. Additionally, the U.S. is a member of the Gen IV International Forum (GIF). The GIF is a cooperative, multinational organization to guide and carry out research and development needed for the GEN IV reactor systems (Forum 2018). GIF evaluated numerous reactor concepts and down-selected to the six most feasible advanced reactor technologies: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), MSR, supercritical watercooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). In support of the growing interest domestically and internationally, the Materials Protection, Accounting, and Control Technologies (MPACT) campaign, under the DOE-NE Fuel Cycle Technologies (FCT) program, engages in R&D activities by developing advanced instrumentation and analysis for safeguards and security of modern, advanced nuclear fuel cycle (non-LWR) facilities. Because of the historic experience in the operation of the Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) (Robertson, MSRE Design and Operations Report Part I 1965), Oak Ridge National Laboratory (ORNL) is heavily engaged in the various R&D activities through the DOE complex related to MSRs including national technical leadership of the DOE-NE MSR campaign. This report discusses and presents the outcomes of the FY19 MPACT MSR Safeguards task. The challenges presented by MSRs for nuclear material accountancy and control (NMAC) and associated safeguards will be investigated. The objective of this research is to explore and compile the safeguards requirements and identify measurement signatures through an initial high-level MSR design and develop complementary advanced simulation and modeling capabilities. A high-level ORNL-developed MSR design called the Molten Salt Demonstration Reactor (MSDR) (Bettis, Alexander and Watts 1972) was used as the target reactor design for this research. The MSDR model incorporates technology from the MSRE and the Molten Salt Breeder Reactor (Robertson, Conceptual Design of a Single-Fluid Molten-Salt Breeder Reactor 1971). But the MSDR is a 750 MWth graphite moderated liquid fueled (low-enriched uranium) MSR compared to the MSRE’s of 7.5 MWth. The focus of this report is to discuss the evaluation of novel signatures, correlations, and indicators to understand the applicability of current safeguards instrumentation to MSRs using the modeling results from the MSDR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Safety-Related Instrumentation and Control Pilot Upgrade: Conceptual - Detailed Design Phase Report and Lessons Learned

Historically, regulatory barriers have precluded nuclear plant safety-related (SR) Instrumentation and Control (I&C) system modernizations. These barriers have been largely addressed by industry and United States Nuclear Regulatory Commission (NRC) efforts, including development of the License Amendment Request (LAR) Alternate Review (AR) process for digital I&C upgrades. The U.S. Department of Energy (DOE) and the Idaho National Laboratory (INL) are supporting a SR I&C pilot upgrade at the Limerick Generating Station (LGS) to demonstrate this process and capturing associated lessons learned for the benefit of industry. In 2021, a paper summarizing lessons learned as captured in INL/EXT-20-59809, Safety Related Instrumentation & Control Pilot Upgrade Initial Scoping Phase Implementation Report and Lessons Learned was presented at the 12th NPIC & HMIT conference. It is recommended that readers of this paper review reference 1 or the associated Nuclear Journal paper for context. This paper summarizes activities performed and lessons learned for the conceptual design phase and portions of the detailed design phase as captured in INL/RPT-23-XXXX, “Safety-Related Instrumentation and Control Pilot Upgrade: Conceptual - Detailed Design Phase Report and Lessons Learned.” The portions of the effort covered by reference 2 include conceptual design phase activities from December 2020, through completion in November 2021, and detailed design activities through September 2022. These efforts and associated detailed documentation in reference 3 are intended to support similar industry upgrades as a foundation stone for a Digital Transformation that will improve plant safety, reliability, and operational performance while lowering plant total cost of ownership.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Insights into Methodologies and Stochastic Optimization of Thermal Energy Storage-Coupled Advanced Reactor Systems: A Comparison of Methods for Accessing Long-Term Sub-System Sizing Adequacy

This paper investigates the potential of coupling Thermal Energy Storage (TES) with Advanced Reactors (ARs) to address uncertainties posed by climate change in deep decarbonized power systems. The TES Use-case Team at Idaho National Laboratory (INL) has examined the potential of storing thermal energy from ARs during low demand periods and optimizing discharge during peak-priced hours, in both steady-state and transient conditions. Building on this groundwork, this study bridges the gaps in optimal sizing of the sub-system of TES-coupled AR systems using Risk Analysis Virtual Environment (RAVEN) and Holistic Energy Resource Optimization Network (HERON), INL?s framework for grid optimization. By applying this framework, we present statistically-robust optimal charge, discharge including balance of plant (BOP), and storage sizing for the High-Temperature Gas-Cooled Reactor (HTGR) with 203 MWth output. To this end, we generated synthetic price samples for 30 years using 2018 ? 2021 real-time market data from ERCOT, PJM and MISO. Our results reveals that the TES-coupled HTGR system is highly effective in maximizing revenue from electricity sales. We observed a substantial increase of 40 % in ERCOT and a noteworthy 15 % increase in PJM and MISO when compared to the conventional BOP without TES. This improvement is achieved through regionally-tailored sub-system sizing, which ranges from 398 to 416 MWth for the discharge system and 610 to 1029 MWth for the TES. We find that the average electricity price directly impacts the overall economics, while price volatility influences storage size. Additional sensitivity analyses were performed to access the impact of key assumptions on system economics and sizing, differentiating the optimization window (i.e., 24 ? 219 hours of chronological observations) and by imposing storage continuity condition in tracking TES cycles. We observed that at the 120-hour of the optimization window, a reasonable balance between computation time and accuracy was achieved. Our analysis also highlights the significance of conducting multi-day cycle analysis (> 120-hour) for TES to capture interaction between electricity prices and storage dynamics, providing a comprehensive understanding of TES behavior that AR developers should integrate into their plant designs.

25 ENERGY STORAGE↗
DOE OSTI2022

SAS4A/SASSYS-1 Validation with EBR-II Tests Performed During the SHRT Testing Program

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates. This report documents the results of modeling and simulation of tests from the Shutdown Heat Removal Test program to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. SHRT-17 was a protected loss of flow test where a loss of electrical power to all sodium coolant pumps was simulated to demonstrate the effectiveness of natural circulation cooling characteristics. SHRT-45R was an unprotected loss of flow test where the control rod scram function of the plant protection system was disabled to demonstrate the effectiveness of EBR-II’s passive reactivity feedbacks. BOP-301 and BOP-302R were unprotected loss of heat sink tests that demonstrated how reactivity feedback effects driven by changes to the core inlet temperature can shut down the fission process. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project for benchmark analysis of the EBR-II Shutdown Heat Removal Tests. The SAS4A/SASSYS-1 models originally developed for the CRP have been modernized to utilize new code features and modeling practices. Differences between the predicted and measured test data were explored and the causes of these differences were identified as being due to modeling approximations, uncertainty in transient component behavior, and instrumentation uncertainties. Overall, it was concluded that for SHRT-17, SHRT-45R, BOP-301, and BOP-302R there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Advances in Integral and Separate Effects Experiments for Water Cooled Small Modular Reactors

This paper focuses on the progress being made in water-cooled small modular reactor (SMR) advanced integral effects and separate effects experiments for reactor licensing. SMRs, considered modular in design, are mostly factory-built and then shipped to the reactor site. Of the several types of SMR designs available, water-cooled SMRs are likely to receive regulatory approval faster than others, as most of the technologies involved (e.g., the fuel and coolant technologies) are matured. However, the unique safety systems of SMRs, which depend on specific SMR design features, require integral and separate effects experiments to achieve reactor licensing. Thus, of the many SMR designs being proposed, only a few have successfully undergone licensing and reached the final development and demonstration stage. Many nuclear vendors and newcomer companies are investing millions of dollars to develop integral and separate effects testing facilities for preparing final safety analysis reports to include in licensing applications. Development and analysis of these experimental facilities is costly and takes about four to five years. The unique challenges involved can be reduced when stakeholders synergistically apply lessons learned, knowing the critical role played by advancements in experiments that support the licensing of SMR safety systems. Identification of knowledge/research gaps with the phenomena identification and ranking table (PIRT) and designing experimental facilities focusing on the phenomena of interest (POI) and figures of merit (FOMs) are pivotal to select the critical path to successful design demonstration and licensing application.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI1993

Nuclear Safety [Vol. 34, No. 1, January-March 1993]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 An Integrated Regional Approach To Risk Management of Industrial Systems, S. Chakraborty and R. Stratton; 9 Annual Technical Meeting of the NRC Cooperative Severe Accident Research Program, E. G. Silver; ACCIDENT ANALYSIS: 13 Deterministic Severe Accident Criteria as Severe Accident Design Criteria and Policy for the New Production Reactor-Heavy Water Reactor, P. T. Rhoads; 20 Proposed Deterministic Severe Accident Criteria for the Heavy Water Reactor-New Production Reactor Containment, K. D. Bergeron, S. E. Slezak, and C. E. Leach; 33 Quasi-Static Core Liquid Level Depression and Long-Term Core Uncovery During a PWR LOCA, Y. Kukita, R. R. Schultz, H. Nakamura, and J. Katayama; 49 LIRA: An Advanced Containment System to Minimize the Accidental Radioactivity Releases, A. Turricchia; 63 Errata to “A Review of Hydrogen Production During Melt/Water Interaction in LWRs,” Vol. 33, No. 4; CONTROL AND INSTRUMENTATION: 64 The Nuplex 80+™ Advanced Control Complex from ABB Combustion Engineering, F. Ridolfo, D. Harmon, and K. Scarola; DESIGN FEATURES: 76 A Scheme for Passive Isolation of the Containment of a Reactor, A. K. Ghosh, V. V. Raj, and K. Kakodkar; 84 Utility Requirements for Safety in the Passive Advanced Light-Water Reactor, T. U. Marston, W. H. Layman, and G. Bockhold, Jr.; WASTE AND SPENT FUEL MANAGEMENT: 91 Activities Related to Waste and Spent Fuel Management, Compiled by M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 103 Testing Deficiencies in Auxiliary Systems Feedwater Systems, J. D. Kueck; 110 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 113 Selected Safety-Related Events, Compiled by G. A. Murphy; 115 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; RECENT DEVELOPMENTS: 133 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 145 Reports, Standards, and Safety Guides, D. S. Queener; 150 Proposed Rule Changes as of Sept. 30, 1992; ANNOUNCEMENTS: 75 Harvard Short Course on Risk: Science, Assessment, and Management; 102 International Conference on Reactor Physics and Reactor Computations; 102 The Fourth International Conference on Nuclear Waste Reprocessing and Waste Management [RECOD ’94]; 132 International Symposium on the Radiological Exposure of the Population of the European Community from Radioactivity in the Mediterranean Sea [MARINA-MED]; 162 International High-Level Radioactive Waste Conference ‘Technology Integration—Above and Below”; 154 The Authors; 158 Indexes to Nuclear Safety, Volume 33.

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