Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Small Modular”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Simulation of the NuScale SMR and Investigation of the Effect of Load-Following on Component Lifetimes

The NuScale SMR has been modelled using the Virtual Environment for Reactor Applications (VERA) multiphysics environment and the results compared with the publicly reported data in the Design Certification Application (DCA). The results show an excellent agreement for the compared axial and radial power distributions, temperature coefficients of reactivity, boron and control rod worths, and fast neutron flux. This NuScale model is then used to investigate the effect of different operational modes on reactor components to determine how flexible load-following operation may affect control rod and reactor pressure vessel (RPV) lifetimes. The control rod degradation is confirmed to primarily affect the silver-indium-cadmium (AIC) rod tip. The degradation rate is observed to follow a non-linear function of core power level where the increase in degradation decreases with insertion depth. For the variation in core power levels expected with current load-following schemes, the total control rod degradation is found to be mild, at 5-10% of usable life per cycle for a reactor operating at <80% power. Nonetheless, this enables load following strategies to be confirmed and/or modified to ensure that control rods do not need to be replaced during the 60+ year life of the reactor. The RPV degradation was found to be almost directly proportional to the core power level and was not overly sensitive to flux shape perturbations. Future work is planned using these damage functions to optimize operation over multiple NuScale SMR units and develop strategies for prognostics and health management.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Fuel Behavior Implications of Reactor Design Choices in Pressurized Water SMRs

Small pressurized water reactors (PWRs) can feature boron free operation, natural circulation mode, reduced height assemblies and/or long refueling cycles. This paper attempts to explore core design optimization for each of these design evolutions. In consequence, five core design layouts are developed incorporating boron free operation with continuous control rods insertion, natural circulation with low burnup/low power density design, natural circulation with high burnup/low power density design, forced circulation with standard core power density design, and forced circulation with high power density design. These cores’ performance is compared to a standard 4-loop PWR. The design process aims to improve the fuel cycle cost under safety constraints through core design optimization using CASMO4E/SIMULATE3 reactor physics codes and FRAPCON4.1 fuel performance assessment tool. Core modeling assumes standard 17x17 PWR fuel assemblies loaded with low enriched uranium (LEU) up to 5wt% or LEU+ (i.e., below 10wt% enrichment) pellets with gadolinium oxide (Gd2O3) as the burnable poison. Satisfactory core and fuel performances are obtained for all the designed cores under steady state and considered overpower transients. For low power density operation, long cycle lengths are achieved reaching a 2.5- and a 5-year cycles and peak rod-average burnup is pushed to 83 MWd/kgU. Other cycle lengths are maintained at 18 months. Boron free operation exhibits the ability to achieve longer cycle lengths at the cost of higher peaking factors leading to high local power and fuel temperatures which prevents sizable power uprates and is deemed uneconomical. Fuel assembly height reduction allows coolant velocity retrofit which enables higher core power density without violating structural integrity of the fuel assembly. As a result, a core power density of 123 kW/l is reached where total cladding hoop strain becomes the limiting parameter.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Multiphysics analysis system for heat pipe cooled micro-reactors employing PRAGMA-OpenFOAM-ANLHTP

A multiphysics analysis system for neutronics/thermo-mechanical/heat-pipe analysis of heat pipe cooled micro-reactors was developed using the PRAGMA code as the neutronics engine. PRAGMA, which used to be a GPU-based continuous-energy MC code for power reactor applications, now has an extended geometry package to handle geometries with unstructured meshes generated by the ANSYS Design-Modeler and Meshing. The NVIDIA ray tracing engine OptiX was exploited for efficient neutron transport on unstructured geometry. On the multiphysics side, the open-source CFD tool OpenFOAM and one-dimensional heat pipe analysis code ANLHTP were adopted. The manager-worker system based on the MPI dynamic process management (DPM) model enables efficient coupling of codes employing different parallelization schemes. With all the features, the multiphysics analysis of the one-sixth symmetrical MegaPower 2D core was performed and it demonstrated the benefit of the tight integration of the three-way coupling system and one-to-one geometry coupling strategy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Error analysis of a hybrid control drum worth model

This paper presents a perturbation-based model for control drum worth prediction which employs both physics-based and statistics-based components. Control drums, or control shims, are cylindrical in shape and span the axial length of the core. A portion of the cylinder is coated in neutron absorbing material and the drum can rotate to introduce the absorbing material to the body of the core to reduce reactivity. The model can be expensive to create due to the requirement for full-core Monte Carlo eigenvalue calculations. Therefore, it is important to analyze how the errors in Monte Carlo calculated k{sub eff} used for model training affect model performance. It was found that the error in predicted criticalities could average to 70 pcm in the most complex form of the model and 215 pcm in the simplest form of the model. Furthermore, it was found that the Monte Carlo uncertainty in quantities calculated with Serpent used to train the models had minimal impact on the error observed from the model. Lastly, one of the forms of the hybrid model could be trained in considerably less computational time if the Monte Carlo calculations were run to higher uncertainty in k{sub eff} with a small penalty to model performance.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Analysis of near-field and far-field aerosol dispersion for microreactors

The current paper presents a simulation-based analysis of aerosol dispersion in the near-field and far-field of a generic, conceptual microreactor operating at pressures close to the ambient pressure; therefore, in the event of an accident that causes radionuclide leakage from the microreactor containment, the radionuclide particles are less likely to travel too far from the reactor, as opposed to conventional reactors. Accordingly, the presented work provides estimates of average and 95-percentile values of the relative effluent concentration. A parametric study is then performed to narrow down the parameters which affect the aerosol dispersion characteristics most significantly. Simulations were performed in the computer code ARCON96, and the parameters found to affect aerosol dispersion characteristics are the atmospheric stability class, and the distance between the release point and the receptor. It is recommended that the computer code RADTRAD be used to calculate actual dosage over distance, using the outputs from ARCON96 as inputs, along with reactor-specific core term inventories. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

An economics-by-design approach to a radiant integrated thermophotovoltaic microreactor system

This paper presents an application of the economics-by-design approach to the Radiant Integrated Thermo-photovoltaic Microreactor System (RITMS). The RITMS design is unique in that it directly couples a critical fission reactor with thermo-photovoltaic (TPV) panels for high efficiency energy conversion. This significant shift from electric conversion using traditional dynamic heat pumps leads to a simpler and more reliable system without turbomachinery and high pressurization. In working towards wrapping up the early design work, the economics-by-design approach, which centers economic competitiveness as the optimization parameter, is well suited in making a final determination on viability. This paper describes the computational sequence that was developed to couple the radiative and conductive heat transfer and feed operational performance to cost estimation. The framework was applied to maximize the power of the system, while minimizing the fuel enrichment. This method is applied to a reference RITMS design as part of a parametric sensitivity study, which revealed that the system is under moderated and that single unit plants could produce power as low as 300 dollar/MWh. This price point supports the notion that early in the RITMS design implementation, adoption into niche markets as a first-of-a-kind technology is possible. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Critical configuration and physics measurements for graphite reflected assemblies of U(93.15)O 2 fuel rods (1.506-CM pitch)

A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods.

1.506-cm pitch UO2 Assembly↗
DOE OSTI2012

Critical Configuration for Beryllium Reflected Assemblies of U(93.15)O 2 Fuel Rods (1.506-CM Pitch and 7-Tube Clusters)

A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Critical Configuration and Physics Measurements for Assemblies of U(93.15)O 2 Fuel Rods (1.506-cm Pitch)

A series of critical experiments were completed from 1962–1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967.a The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO 2 fuel pellets, surrounded by a graphite reflector. Measurements were performed to determine critical reflector arrangements, relative fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector” (see Reference 1). The experiment studied in this evaluation was the second of the series and had the fuel rods in a 1.506-cm-triangular pitch. One critical configuration was found (see Reference 3). Once the critical configuration had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U,bc and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configuration are described in Sections 1.3, 1.4, and 1.7, respectively.

1.506-cm pitch↗
DOE OSTI2013

Critical Configuration and Physics Measurements for Beryllium Reflected Assemblies of U(93.15)O 2 Fuel Rods (1.506-cm Pitch)

Cadmium ratios were measured with enriched uranium metal foils at various locations in the assembly with the fuel tube at the 1.506-cm spacing. They are described in the following subsections. The experiment configuration was the same as the first critical configuration described in HEU-COMP-FAST-004 (Case 1). The experimenter placed 0.75-cm-diameter × 0.010-cm-thick 93.15%-235U-enriched uranium metal foils with and without 0.051-cm-thick cadmium covers at various locations in the core and top reflector. One part of the cadmium cover was cupshape and contained the uranium foil. The other part was a lid that fit over the exposed side of the foil when it was in the cup shaped section of the cover. As can be seen in the logbook, two runs were required to obtain all the measurements necessary for the cadmium ratio. The bare foil measurements within the top reflector were run first as part of the axial foil activation measurements. The results of this run are used for both the axial activation results and the cadmium ratios. Cadmium covered foils were then placed at the same location through the top reflector in a different run. Three pairs of bare and cadmium covered foils were also placed through the core tank. One pair was placed at the axial center of a fuel tube 11.35 cm from the center of the core. Two pairs of foils were placed on top of fuel tubes 3.02 and 12.06 cm from the center of the core. The activation of the uranium metal foils was measured after removal from the assembly using two lead shielded NaI scintillation detectors as follows. The NaI scintillators were carefully matched and had detection efficiencies for counting delayed-fission-product gamma rays with energies above 250 KeV within 5%. In all foil activation measurements, one foil at a specific location was used as a normalizing foil to remove the effects of the decay of fission products during the counting measurements with the NaI detectors. The normalization foil was placed on one NaI scintillator and the other foil on the other NaI detector and the activities measured simultaneously. The activation of a particular foil was compared to that of the normalization foil by dividing the count rate for each foil by that of the normalization foil. To correct for the differing efficiencies of the two NaI detectors, the normalization foil was counted in Detector 1 simultaneously with the foil at position x in Detector 2, and then the normalization foil was counted simultaneously in Detector 2 with the foil from position x in Counter 1. The activity of the foil from position x was divided by the activity of the normalization foil counted simultaneously. This resulted in obtaining two values of the ratio that were then averaged. This procedure essentially removed the effect of the differing efficiencies of the two NaI detectors. Differing efficiencies of 10% resulted in errors in the ratios measured to less than 1%. The background counting rates obatined with the foils used for the measurements on the NaI detectors before their irradiation measurement were subtracted from all count rates. The results of the cadmium ratio measurements are given in Table 1.3-1 and Figure 1.3-1. “No correction has been made for self shielding in the foils” (Reference 3).

1.506-cm pitch↗
DOE OSTI2014

Critical Configuration and Physics Measurements for Beryllium Reflected Assemblies of U(93.15)O 2 Fuel Rods (1.506-cm Pitch and 7-Tube Clusters)

Cadmium ratios were measured with enriched uranium metal foils at various locations in the assembly with the fuel tube at the 1.506-cm spacing. They are described in the following subsections. The experiment configuration was the same as the first critical configuration described in HEU-COMP-FAST-004 (Case 1). The experimenter placed 0.75-cm-diameter × 0.010-cm-thick 93.15%-235U-enriched uranium metal foils with and without 0.051-cm-thick cadmium covers at various locations in the core and top reflector. One part of the cadmium cover was cupshape and contained the uranium foil. The other part was a lid that fit over the exposed side of the foil when it was in the cup shaped section of the cover. As can be seen in the logbook, two runs were required to obtain all the measurements necessary for the cadmium ratio. The bare foil measurements within the top reflector were run first as part of the axial foil activation measurements. The results of this run are used for both the axial activation results and the cadmium ratios. Cadmium covered foils were then placed at the same location through the top reflector in a different run. Three pairs of bare and cadmium covered foils were also placed through the core tank. One pair was placed at the axial center of a fuel tube 11.35 cm from the center of the core. Two pairs of foils were placed on top of fuel tubes 3.02 and 12.06 cm from the center of the core. The activation of the uranium metal foils was measured after removal from the assembly using two lead shielded NaI scintillation detectors as follows. The NaI scintillators were carefully matched and had detection efficiencies for counting delayed-fission-product gamma rays with energies above 250 KeV within 5%. In all foil activation measurements, one foil at a specific location was used as a normalizing foil to remove the effects of the decay of fission products during the counting measurements with the NaI detectors. The normalization foil was placed on one NaI scintillator and the other foil on the other NaI detector and the activities measured simultaneously. The activation of a particular foil was compared to that of the normalization foil by dividing the count rate for each foil by that of the normalization foil. To correct for the differing efficiencies of the two NaI detectors, the normalization foil was counted in Detector 1 simultaneously with the foil at position x in Detector 2, and then the normalization foil was counted simultaneously in Detector 2 with the foil from position x in Counter 1. The activity of the foil from position x was divided by the activity of the normalization foil counted simultaneously. This resulted in obtaining two values of the ratio that were then averaged. This procedure essentially removed the effect of the differing efficiencies of the two NaI detectors. Differing efficiencies of 10% resulted in errors in the ratios measured to less than 1%. The background counting rates obatined with the foils used for the measurements on the NaI detectors before their irradiation measurement were subtracted from all count rates. The results of the cadmium ratio measurements are given in Table 1.3-1 and Figure 1.3-1. “No correction has been made for self shielding in the foils” (Reference 3).

reactivity worth measurements↗
DOE OSTI2014

An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

09 BIOMASS FUELS↗
DOE OSTI2014

Local-Level Prognostics Health Management Systems Framework for Passive AdvSMR Components. Interim Report

This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical AdvSMR passive components (to establish condition indices for monitoring) with model-based prognostics methods. The focus of the PHM methodology and algorithm development in this study is at the localized scale. Multiple localized measurements of material condition (using advanced nondestructive measurement methods), along with available measurements of the stressor environment, enhance the performance of localized diagnostics and prognostics of passive AdvSMR components and systems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

U.S. Forward Operating Base Applications of Nuclear Power

This paper provides a high level overview of current nuclear power technology and the potential use of nuclear power at military bases. The size, power ranges, and applicability of nuclear power units for military base power are reviewed. Previous and current reactor projects are described to further define the potential for nuclear power for military power.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Prototypic Enhanced Risk Monitor Framework and Evaluation - Advanced Reactor Technology Milestone: M3AT-15PN2301054

This research report summaries the development and evaluation of a prototypic enhanced risk monitor (ERM) methodology (framework) that includes alternative risk metrics and uncertainty analysis. This updated ERM methodology accounts for uncertainty in the equipment condition assessment (ECA), the prognostic result, and the probabilistic risk assessment (PRA) model. It is anticipated that the ability to characterize uncertainty in the estimated risk and update the risk estimates in real time based on equipment condition assessment (ECA) will provide a mechanism for optimizing plant performance while staying within specified safety margins. These results (based on impacting active component O&M using real-time equipment condition information) are a step towards ERMs that, if integrated with AR supervisory plant control systems, can help control O&M costs and improve affordability of advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave similarly and retain the same ex-core features and characteristics. While INL has developed Designs A and B, LANL is evaluating a Hot Isostatic Pressed reactor configuration to overcome the drawbacks of the monolithic core design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO 2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO 2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave similarly and retain the same ex-core features and characteristics. While INL has developed Designs A and B, LANL is evaluating a Hot Isostatic Pressed reactor configuration to overcome the drawbacks of the monolithic core design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗