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

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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Methods and apparatuses for operating nuclear reactors and for determining power levels in the nuclear reactors

An apparatus may include two or more electrical conductors, one or more signal devices, and an analyzer. The one or more signal devices may be adapted to apply one or more signals to the two or more electrical conductors and receive one or more signals from the two or more electrical conductors. The analyzer may be adapted to determine power level in the nuclear reactor using at least one of the applied signals and at least one of the received signals. A method of determining power level in a nuclear reactor may include: measuring impedance values of two or more electrical conductors disposed in the nuclear reactor and using the measured impedance values to determine the power level. A method of operating a nuclear reactor may include: measuring the impedance values and using the measured impedance values to determine parameters of the nuclear reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Nuclear Forensics for Irradiated Materials using Kernel Regression Models

The ultimate purpose of nuclear forensics is to locate the origin of nuclear and other radioactive materials out of regulatory control used in nuclear security events such as classical forensics. For this purpose, some specific characteristics of interdicted materials (referred to as 'signatures') should be defined to compare the known sample database. In some cases, involved with irradiated material, the signatures could be specific amount of fission product inventory depending on reactor operational histories such as reactor type, initial enrichment, burn up rate of nuclear fuel and cooling period after irradiation in the reactor. In order to make inference model for irradiated material for nuclear forensics, we suggested 3 ways as follows; (i) analyzing ORIGEN produced nuclear substances for the fission product inventory, (ii) classifying the range of reactor type, (iii) predicting the operational history from the fission product inventory. In terms of nuclear substance analysis, the construction of spent fuel database has been internationally carried out since ITWG (Nuclear Smuggling International Technical Working Group) was established in 1995. However, enough sample data cannot be obtained in Korea to determine reactor type and operational histories because analyzing nuclear substances have been limited due to the international regulation. Thus, this study does not mention how to obtain mass or ratios of nuclides for the spent fuel. We just have generated simulation sample data as a fission product inventory using ORIGEN to construct spent fuel database. When unidentified radioactive materials are discovered, it is assumed that mass or ratios of the nuclides are analyzed by certain methods which are available in Korea as well as other countries. Then, a reactor type is determined, and finally, regression model can locate the origin or source of the radioactive materials by inferring operational histories using the fission product inventory. There are some studies on the classification of reactor type using statistical methods. As a part of predicting the operational history, this paper focuses on the development of spent fuel inference algorithm for nuclear forensics to estimate reactor operational histories that reflect characteristics of a reactor from the fission product inventory. The spent fuel inference algorithm for nuclear forensics has been developed on the basis of Kernel Regression (KR) which can conditionally generate the estimated output data by giving weight depending on geometric distance between sample data and input data. Also the KR which comprises Auto-Associative KR and Inferential KR have some benefits that can estimate only using data without assumption for specific physical model and re-learning process. Auto-associative KR (AAKR) was used to sort out outliers of input data and Inferential KR (IKR) was carried out to predict the operational histories for spent fuel. All of regression algorithms were performed in R programming tool. From the validation of both AAKR and IKR, it is expected that kernel regression has enough accuracy to be applied to nuclear forensics for spent fuel. Once the nuclear substance is measured to fill the fission product inventory with a certain reactor type, this kernel regression algorithm would predict the operational histories of spent fuel with high accuracy. Finally, this information is expected to be useful to find the origin of irradiated materials. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Variable Nuclear Electricity with Base-Load Reactor, Closed Brayton Power Cycle and Firebrick Resistance-Heated Energy Storage (Fires)

Power cycle choices are determined by markets and available technologies. Two changes are occurring in the electricity sector. First, the world is beginning to move away from fossil fuels because of concerns about climate change. What replaces fossil fuels for variable electricity production? Second, there have been extraordinary technological advances in gas turbine technology. The changes in market requirements and technology suggest the need to reexamine nuclear power cycles. One option is to couple high-temperature base-load nuclear reactors to closed Brayton cycles (helium, nitrogen, etc.) with Firebrick Resistance-Heated Energy Storage (FIRES). The reactor operates at base-load but not the power cycle. When electricity prices are low, firebrick is electrically heated to high temperatures using base-load electricity and/or electricity from the grid. The gas turbine can use nuclear heat for base-load operation and produce added peak electricity by sending the compressed heated gas through hot firebrick after nuclear heating to further raise power cycle gas temperatures and power output from the turbine-creating a nuclear plant with a thermodynamic topping (NUTOP) cycle. If such cycles can be developed, they would increase plant revenue and enable a low-carbon electricity grid. Such cycles can be coupled to Fluoride-salt-cooled High-temperature Reactors (FHRs), molten salt reactors (MSRs), high-temperature gas-cooled reactors (HTGRs), sodium-cooled fast reactors (SFRs) and lead-cooled fast reactors (LFRs). We discuss (1) market changes, (2) previous work on Nuclear Air-Brayton Combined cycles (NACC) with base-load reactors and variable power to the electricity grid and (3) options for closed Brayton cycles with variable power while the reactor operates at base load. Market (customer) requirements ultimately drive selection of power cycles. A low-carbon electricity grid requires dispatchable nuclear electricity to replace the traditional role of fossil fuels in the electricity grid. While reactors can be operated at part load, this is expensive for a capital-intensive low-operating-cost technology. Brayton power cycles with FIRES have the potential capability to provide that variable electricity while the reactor operates at base load to (1) improve economics and (2) enable a zero-carbon electrical grid. Open Brayton cycles have been investigated. Work has now started on developing closed Brayton cycles with FIRES. Only limited work has been done to date has been done for closed cycles with FIRES. (authors)

42 ENGINEERING↗
DOE OSTI2018

Unit mechanisms of fission gas release: Current understanding and future needs

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel properties and, once the gas is released into the gap between the fuel and cladding, lowering gap thermal conductivity and increasing gap pressure. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are being applied to provide unprecedented understanding of the unit mechanisms that define the fission product behavior. In this article, existing research on the basic mechanisms behind the various stages of fission gas release during normal reactor operation are summarized and critical areas where experimental and simulation work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior during reactor operation and to design fuels that have improved fission product retention. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.

computer simulation↗
DOE OSTI2022

Development of Surveillance Test Articles for Materials Degradation Management in MSR Environment

Information on materials degradations during advanced reactor operations is limited Effects of materials degradations during reactor operations are synergistic, involving: Irradiation, corrosion, elevated temperature exposure and stress (creep-fatigue loading) Establishment of surrogate materials surveillance program for the management of materials degradations would be an important pathway in support of the timely licensing of advanced reactors

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluation of Changes to the MTR Fuel Dissolution Flowsheet to Increase the Number of L-Bundles Charged to a Dissolver

Material Test Reactor (MTR) fuels are dissolved in the H-Canyon facility using a Hg-catalyzed, HNO₃ dissolution flowsheet. The flowsheet is based on experimental work in which laboratory-scale experiments were used to provide bounding values for both offgas and H₂ generation rates. The generation of H₂ gas, primarily from the dissolution of Al, is controlled by limiting the surface area exposed to the solution. The number of L-Bundles charged and their submergence level defines the exposed surface area which is directly proportional to the H₂ generation rate. For the MTR fuel dissolution flowsheet, the maximum number of L-Bundles which can be charged to a dissolver was specified as a function of the Al concentration in the solution and the submergence level of the bundle. The MTR fuels are currently dissolved in the 6.1D dissolver. To reach the desired terminal Al concentration, a total of 12 L-Bundles of fuel are dissolved per batch. Depending on the submergence level being used in the dissolver, the 12 bundles are dissolved in three charges of four bundles per charge or in a sequence of three, four, and five bundles. However, if six L-Bundles could be dissolved in the initial charge, a dissolver batch could be completed in two dissolutions instead of three. To assess the feasibility of dissolving six bundles in the first charge, the Savannah River National Laboratory evaluated changes to the MTR flowsheet which increase the number of bundles which can be charged to a dissolver. Since the development of the existing MTR fuel dissolution flowsheet, improved offgas measurement methods have been developed to more accurately calculate the offgas composition and generation rate during laboratory-scale dissolutions. In the first part of this study, improvements in the calibration model for the Raman spectrometer (used for offgas characterization) and a more effective data smoothing routine were applied to existing data. The uncertainty in the offgas composition data used to define the MTR fuel dissolution flowsheet were subsequently re-evaluated which resulted in an approximate 50% reduction in the uncertainty of the H₂ gas concentration measurements. The reduced uncertainty was used in new calculations to define the maximum number of L-Bundles which can be charged to either the 6.1D or 6.4D dissolver as a function of the Al concentration. With the reduced uncertainty in the H₂ concentration measurements, a flowsheet was defined which supported initially charging 6 bundles to a dissolver. This flowsheet is based on the use of 0.002 M Hg, an L-Bundle immersion depth of 54 in., an air sparge/purge flowrate of 40 scfm, an iodine reactor operating temperature of 200 °C, and not exceeding 60% of the H₂ LFL in the dissolver offgas. In the second phase of this study, calculations were performed to assess the impact of changing various operating parameters to increase the number of L-Bundles charged to an H-Canyon dissolver. Applying the improved offgas measurement techniques including the reduced uncertainty in the H₂ concentration measurements to the existing data, the number of bundles of MTR fuel which can be charged to either the 6.1D or 6.4D dissolver were calculated based on varying the bundle immersion depth (44-60 in.), air sparge/purge flowrate (40-60 scfm), iodine reactor operating temperature (150-200 °C), and the percentage of the H₂ LFL which must not be exceeded (60-76%). Qualitatively, the percentage of the LFL had the largest influence on the maximum number of bundles which can be charged to a dissolver followed by the air sparge/purge flowrate, L-Bundle immersion depth, and the iodine reactor operating temperature. During the evaluation of each variable, the other variables were held at the baseline flowsheet conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Whole core analysis of the single-fluid double-zone thorium molten salt reactor (SD-TMSR)

The SD-TMSR (2,250 MW th ) is a Single-fluid Double-zone Thorium-based Molten Salt Reactor. The active core of the SD-TMSR is divided into the inner zone (486 fuel tubes) and the outer zone (522 fuel tubes) to improve the Th-U3 breeding performance. This work adopted the SERPENT-2 Monte Carlo code to analyze the whole core model of the SD-TMSR. Built-in SERPENT-2 capabilities simulated online reprocessing and refueling and calculated the multiplication factor and Breeding Ratio (BR). We found that the molten salt Temperature Coefficient of Reactivity ( α T ) was negative for initial and equilibrium states. This study investigated the variation of the multiplication factor, BR, and build-up of important nuclides in the core as a function of burnup. Under online reprocessing and refueling, we studied the variation of the reactivity during 60 years of reactor operation. Additionally, the neutron spectrum shift during the reactor operation was calculated. Finally, these simulations determined the appropriate 232 Th and 233 U feed rates for maintaining criticality and enabled analysis of the overall SD-TMSR fuel cycle performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

A model to assess Zircaloy’s mechanical property changes following a transient beyond critical heat flux

Maintaining the integrity of nuclear fuel rods is essential for ensuring public health and safety in nuclear power generation. During reactor operation, this integrity is confirmed by demonstrating compliance with established regulatory acceptance criteria. For moderate-frequency events, such as limiting transients and anticipated operational occurrences (AOOs), the current fuel integrity criterion is based on preventing boiling transition. This criterion assumes that prevention of boiling transition will prevent excessive cladding heating and, thus, fuel failure during normal operations. While conservative, this approach places significant constraints on core design, fuel cycle economics, and a plant’s ability to perform major power uprates, leading to suboptimal fuel utilization and inefficient carbon-free energy production. A more efficient approach could be achieved by revising the failure criterion to a material-specific limit rather than strictly preventing the boiling transition, since boiling transition per se is not a cause of fuel cladding failure. Here, as a result, a new licensing framework based on material properties, termed time-at-temperature (t@T), is needed. This approach would allow for brief periods of post–critical heat flux operation during an AOO without compromising safety. Implementing the t@T licensing strategy requires a robust technical foundation in material properties, which must be established through comprehensive data collection on both unirradiated and irradiated fuel and cladding materials. This foundation would enable the development of a safety basis that ensures safe operation while providing greater flexibility and efficiency for reactor operation. This paper documents a thorough review of the available data to establish a baseline knowledge that can inform the development of cladding mechanical models, as well as identify experimental data gaps that need to be addressed in future research. Machine learning and data informatics were utilized to extract the importance of parameters on the t@T parameter. Industry tools were used to perform baseline analyses to define the relevant transient conditions for data analysis. The subsequent review successfully identified applicable experimental data, as well as sufficient data to evaluate changes in cladding mechanical properties following an AOO transient. Rather than developing new models, this work coupled existing irradiation annealing and recrystallization models to calculate changes in hardness, yield stress, and ultimate tensile stress following an AOO event. The findings from this review were summarized to highlight the experimental data needs required to fill remaining gaps and support the development of future t@T licensing methodologies.

Cladding performance↗
DOE OSTI2025

Autonomous Operations for Advanced Reactors Utilizing Supervisory Control

Automation is a critical tenet of reactor plant operations as reliance on nuclear energy increases. Nuclear power plants require a large workforce which does not scale with output; that is, the cost per megawatt increases as reactor output becomes smaller. The economic viability of advanced reactors, particularly small modular reactors (SMRs) and microreactors, requires a significantly reduced onsite workforce. The logical solution is establishing a systematic process of elimination of reliance on human operators, and to the extent possible, replacing these actions with automated functions. In this paper, we propose a method for such transformation to establish a robust technical basis to enable transition to autonomy. Our method is based on finite state automata (FSA)—also known as finite state machines (FSMs). Relying on this method allows us to exploit the rich set of mathematical proofs available in the field of regular languages. FSA are one of the mathematical tools to model discrete event systems (DES). These properties are applied to produce an automated startup controller for the Massachusetts Institute of Technology Research Reactor (MITR). The startup procedure is captured in terms of discrete changes from one state to another while an independent supervisory control system directs the sequence of states and alerts a human in the event of an abnormal operation. First, the design and behavior of the MITR rod control system were modeled in Simulink. Then, the startup procedure was applied to the rod control system and the DES performed a startup by procedurally withdrawing rods to the subcritical position. The simulation also stops rod motion in response to an uncontrollable event and restarts rod motion once the event has been cleared.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2020

Use of Multi-Attribute Utility Analyses to Assess Transportation Routes from ISFSIs - 20535

With two consolidated interim storage facilities for used nuclear fuel (UNF) having applied for storage licenses, there is the potential need to transport UNF in the U.S. from shutdown reactor sites to complete the site cleanup and from operating reactor sites who may be running out of storage options (e.g., spent fuel pools are nearly full and dry storage facilities are reaching licensed capacity). At some of the shutdown reactor sites, only the independent spent fuel storage installation (ISFSI) exists and often there remains a degraded transportation infrastructure for performing shipments from these sites. At other sites (e.g., operating reactor plants), the existing transportation infrastructure on the site is unchanged and sufficient for moving UNF, but the environment around the route may have changed (e.g., what was once a rural setting has transformed into a more urban setting) and/or local stakeholders want their input taken into consideration in UNF transportation activities. So, although preliminary evaluations have been performed to identify viable modes of transport from these sites using rail, barge, and/or heavy haul truck, if multiple options exist for moving the UNF from the site (e.g., multiple truck routes or a truck and a barge route both exist) then an evaluation that takes input from multiple stakeholders should be considered to identified the preference (e.g., ranking) of routes from a site. Orano has supported the development of several studies examining the viable routes for removing UNF from some of these sites and has utilized a multi-attribute utility analysis (MUA) to establish a ranking of routes and associated modes to move the UNF from the ISFSIs to a Class I rail carrier. The ultimate result from the MUA is a list(s) of the most to the least favored/preferred routes from the site. This paper provides an overview of the MUA methodology and provides an example of its application to a site with a shutdown reactor. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Brittle nature and the related effects of zirconium hydrides in Zircaloy-4

The use of zirconium alloys such as Zircaloy-4 in nuclear reactors is limited by brittle hydride formation, for example in harsh reactor operating conditions such as those in the pressurized water reactors. Even though much research has studied this phenomenon, a more systematic, in-depth study of the effects of hydrides on the properties of Zircaloy-4, using various analytical techniques, can add more details to the understanding of these effects. In the current study, hydrogen charging was performed on Zircaloy-4 at different levels compatible with commercial reactor operating conditions and conditions in accident scenarios. Effects of hydriding on the Zircaloy-4 (α-(hcp)-Zr) tensile properties, together with an in-depth study of any relationship between tensile and the microstructural properties of the material, are discussed. While a moderate pinning effect on the equiaxed α-Zr grain size was observed, no significant change in bulk texture of the Zry-4 was revealed as the hydrogen level increased up to nominally 1230 wppmH. At a nominal value of 490 wppmH, Zircaloy-4 fractured at a low strain value just above the yielding point, and brittle tensile fracture was observed in a sample H-charged to nominally 1230 wppmH. The low ductility and brittle fracture of the samples were attributed to the presence of brittle inter-grain boundary hydrides. Finally, it was observed that brittle fracturing was also supported by intra-grain boundary hydrides that were affected by the tensile deformation of the sample.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Interface Tracking Investigation of Geometric Effects on the Bubbly Flow in PWR Subchannels

We report that absorbing heat from the fuel rod surface, water as coolant can undergo subcooled boiling within a pressurized water reactor (PWR) fuel rod bundle. Because of the buoyancy effect, the vapor bubbles generated will then rise along and interact with the subchannel geometries. Reliable prediction of bubble behavior is of immense importance to ensure safe and stable reactor operation. However, given a complex engineering system like a nuclear reactor, it is very challenging (if not impossible) to conduct high-resolution measurements to study bubbly flows under reactor operation conditions. The lack of a fundamental two-phase-flow database is hindering the development of accurate two-phase-flow models required in more advanced reactor designs. In response to this challenge, first-principles–based numerical simulations are emerging as an attractive alternative to produce a complementary data source along with experiments. Leveraged by the unprecedented computing power offered by state-of-the-art supercomputers, direct numerical simulation (DNS), coupled with interface tracking methods, is becoming a practical tool to investigate some of the most challenging engineering flow problems. In the presented research, turbulent bubbly flow is simulated via DNS in single PWR subchannel geometries with auxiliary structures (e.g., supporting spacer grid and mixing vanes). The geometric effects these structures exert on the bubbly flow are studied with both a conventional time-averaging approach and a novel dynamic bubble tracking method. Lastly, the new insights obtained will help inform better two-phase models that can contribute to safer and more efficient nuclear reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Multiscale Investigation of Thermal Transport in UO{sub 2} Fuel using Modeling and Simulation

The thermal conductivity of the UO{sub 2} fuel in light water reactors (LWRs) defines how efficiently heat is transported from the fuel to the coolant, and thus directly influences the reactor efficiency. Unfortunately, the same fission events that generate heat also cause radiation damage and fission product formation that decreases the thermal conductivity. This reduction in thermal conductivity reduces the safety of the reactor in addition to its efficiency, because excess heat generated during an accident cannot be transported from the fuel, resulting in high temperatures that can result in melting. Because thermal conductivity plays such an important role in reactor efficiency and safety, many experiments have been conducted to measure the thermal conductivity of UO{sub 2} and how it changes during reactor operation. Experiments on unirradiated fuel have resulted in a large set of data indicating how the bulk thermal conductivity changes as a function of temperature, which was used to develop a semiempirical model. Additional experiments have been conducted to determine how the thermal conductivity of UO{sub 2} changes as a function of burnup, leading to burnup dependent models fitted to the data. While these empirical models perform well within the conditions at which the data was collected, they cannot be extrapolated outside of those conditions. In addition, while burnup is a measure of the amount of fissioning that has occurred, it cannot account for the impact of the operating conditions on the thermal conductivity. While experimental data provides an empirical view on the value of the thermal conductivity in the fuel during reactor operation, a more mechanistic understanding of how temperature and burnup impact the transport of heat through the fuel would be extremely valuable. Such an understanding would allow for a more accurate prediction of fuel performance, but it could also be used to direct modifications to the fuel that would raise the thermal conductivity or reduce the impact of burnup. In this work we summarize recent efforts to use modeling and simulation at the atomistic and mesoscales to obtain a mechanistic understanding of thermal transport in UO{sub 2}. The thermal conductivity of LWR fuel is one of the most critical material parameters impacting the efficiency and safety of the reactor. Thus, a significant effort is being made to develop a mechanistic understanding of the impact of temperature and burnup on it. In this work, we have provided a short summary of modeling and simulation efforts being carried out at the atomistic and mesoscale to understand this critical property. These research efforts are providing a fundamental understanding of the thermal transport in unirradiated UO{sub 2}, as well as the impact of radiation damage and fission products on thermal transport. These efforts have been focused on UO{sub 2}, but they also are laying the framework for conducted similar simulations to investigate new accident tolerant fuel concepts. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Automatic Modeling of VVER-1000 Power Maneuvers

A special version of the IR code is described. The purpose of this code is to generate recommendations for the reactor operator when operating in a maneuvering regime, automate boron regulation, and study the maneuvering characteristics of VVER-1000 fuel loads. The results of modeling of operation in a daily load schedule are presented for the example of a stationary run of VVER-TOI. The impact of variation of the parameters of power maneuvers (run, operating time a reduced power, loading rate) on the boron regulation parameters is analyzed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

An Autonomous Critical Data Extrapolator for the AGN-201m

Nuclear nonproliferation serves as a key goal, being undertaken by the International Atomic Energy Agency (IAEA). To recognize proliferation there are two pathways that states, who intend to use nuclear material for malicious purposes can take, diversion can misuse. Diversion is when fissile nuclear material is declared to the IAEA for non-weapon purposes, but then covertly removed. If the source of nuclear material, that is not declared and not fissionable, is placed inside the reactor core to create fissile material used to create weapons then the state is using the second pathway of proliferation, misuse. With the emerging development in areas of simulation and machine learning the creation of virtual models of reactor systems, digital twins, serve as a potential method to identify proliferation through detecting anomalous behavior in the reactor. A digital twin for a physical nuclear reactor has never been developed, as digital twins serve as an emerging technology. To investigate the process for development and use of a digital twin for a nuclear reactor Idaho State University’s AGN-201m serves as the nuclear reactor used for development of this digital twin. A data acquisition system has been installed to the reactor system allowing for the transfer of collected data from a reactor operation to Idaho National Laboratory’s Deeplynx data warehouse. When utilizing data to train reactor physics and machine learning models, a significant challenge encountered is the initial state of the data. Nuclear proliferation will have the capacity to be detected when the reactor immediately starts up, nor will it occur after the reactor shuts down. Generally, it will be detected when the reactor is operating at some desired power over a sufficient period for that specific reactor design. For the AGN-201m this will be when the reactor is critical (generally 1 mW or above) for a timespan that is within or less than the range of a regular business day. Datasets sent to Deeplynx have had to be manually cut to when the reactor is critical based on plots of power levels. This method is inefficient and laborious, especially when using multiple datasets at once to train a model. To provide a more streamlined approach an automated critical data extrapolator is developed, with capabilities of recognizing when the reactor operation first reaches criticality, and when the reactor undergoes a SCRAM and is shutdown.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Development of Critical Experiments to Benchmark Moderator Temperature Reactivity Worth

The simulation of nuclear reactors requires being able to accurately model several different interrelated physical processes including neutronics, heat production, fluid/thermal physics, structural mechanics, fuel behavior, chemistry, and balance-of-plant. In order to design and use new types of nuclear reactors, modern codes have been developed which couple together the behavior of these different physical models to accurately consider the feedback effects between different physical effects. The SHARP toolset, part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) project, is an example of an effort to combine models of neutronics, thermal-hydraulics, and structural mechanics in a coupled simulation to model any type of nuclear reactor as accurately as possible. The validation of multi-physics coupling techniques requires new experimental benchmark experiments specifically designed for that purpose. As a preliminary step towards design of these experiments, a series of measurements have been performed to demonstrate the temperature-dependent neutronic behavior of a well-characterized reactor assembly for comparison with the available codes. The facility adopted for these experiments is Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI). The RCF is unique in several aspects and provides several advantages to this type of experiment. Built in 1954 by the American Locomotive Company (ALCO), the RCF has been used to design small compact reactors. RPI acquired the facility in 1963, and the highly-enriched uranium fuel was later swapped for low-enriched fuel pins that were excess from the SPERT facility experiments. Nowadays, The RCF is regarded as a teaching and research facility, permitting the training of licensed Senior Reactor Operators, and offering a critical experiments laboratory course to students. The reactor core standard configuration consists of 332 or 333 UO{sub 2} fuel pins, containing a total of 35.2 grams of U- 235 each (4.81 % enriched). The pins are arranged in a regular lattice with a pitch of 1.6256 cm. The number and configuration of these pins are flexible, and the reactor's overall thermal power is limited to no more than 15 W. As the low thermal output of the reactor does not induce any measurable temperature changes, the effect of temperature in the reactor is observed by applying two 18- kW electric heaters to the water moderator, with mechanical agitators to maintain uniformity in moderator temperature. Before beginning experiments, each fuel pin inside the core was surveyed for mechanical defects (bending or warping) that may contribute to uncertainty in core behavior. Every fuel pin has been individually removed from the core and checked for any irregularities on its surface (caused by shocks or usage). About 12 % of the 333 pins were found to have a small degree of warping and were replaced with excess pins available in the facility. In doing this, we made sure all the pins are identical, thereby increasing the accuracy of future models. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Post-quench ductility of zirconium alloy cladding materials - 2016-0088

The ductility of zirconium-based alloys after steam oxidation at high temperatures, followed by quenching in water, is of significant interest for the nuclear industry. High-temperature steam oxidation and quenching of nuclear fuel cladding occurs during postulated loss-of-coolant accidents. The industry is currently establishing in-reactor operation limits for peak cladding temperature and time at temperature based on the measured ductile-to-brittle transition of different cladding materials, both in the as-fabricated condition and using hydrogen charging as a surrogate for in-reactor operation. Aiming to study possible differences in post-quench ductility caused by differences in chemical composition, extensive testing on Zircaloy-4, ZIRLO{sup R}, and Optimized ZIRLO{sup TM} cladding materials as a function of hydrogen content, extent of oxidation, and peak cladding temperature was performed. The results showed similar ductile-to-brittle transition for the alloys studied despite differences in chemical composition, with a decrease in the extent of oxidation to reach the ductile to brittle transition with increasing hydrogen content. Microstructural characterization was performed to investigate the evolution of the oxygen-enriched layer that forms underneath the oxide, both during the time at high temperatures and during cooling. Significantly different morphologies of the oxygen-enriched layer were observed with different cooling rates, particularly for alloys containing niobium. Possible mechanisms for the development of the oxygen-enriched layer are discussed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Post quench ductility of Zirconium alloy cladding materials - 6.2 STP-2016-0088

The ductility of zirconium-based alloys after steam oxidation at high temperature, followed by quenching in water, is of significant interest for the nuclear industry. High temperature steam oxidation and quenching of nuclear fuel cladding occurs during postulated loss-of-coolant accidents (LOCA). The industry is currently establishing in-reactor operation limits for peak cladding temperature and time at temperature based on the measured ductile-to-brittle transition of different cladding materials, both in the as-fabricated condition and using hydrogen charging as a surrogate for in-reactor operation. Aiming to study possible differences in post-quench ductility (PQD) caused by differences in chemical composition, extensive testing has been performed on Zircaloy-4, ZIRLO{sup R} and Optimized ZIRLO{sup TM} cladding materials as a function of hydrogen content, extent of oxidation and peak cladding temperature. The results show similar ductile-to-brittle transition for the alloys studied, in spite of the differences in chemical composition, with a decrease in the extent of oxidation to reach the ductile-to-brittle transition with increasing hydrogen content. Microstructural characterization was performed to investigate the evolution of the oxygen enriched layer that appears underneath the oxide, both during the time at high temperature and during cooling. Significantly different morphologies of the oxygen enriched layer are observed with different cooling rates, particularly for alloys containing niobium. Possible mechanisms for the development of the oxygen enriched layer are discussed (authors)

36 MATERIALS SCIENCE↗