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NASA NTRS1976

Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems

A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. Neutronic feasibility of dual mode operation and smaller reactor sizes than those previously investigated are shown to be possible. A heat transfer analysis of one such reactor shows that the dual-mode concept is applicable when power generation mode thermal power levels are within the same order of magnitude as direct thrust mode thermal power levels.

Chow, S.↗

DOE OSTI2020

Razorback – A reactor transient analysis code for large rapid reactivity additions in a natural circulation research reactor

Research reactors play an important role in higher education, scientific research, and medical radioisotope production around the world. It is thus important to ensure the safety of facility workers and the public. This work presents a new reactor transient analysis code referred to as Razorback. The code has been developed for the evaluation of large rapid reactivity addition in research reactors. Its initial focus is the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. Results have been validated against ACRR pulse operations. Razorback models the reactor kinetics, fuel element heat transfer, fuel element thermal expansion, and natural circulation coolant channel thermal-hydraulic response. Simulation results for ACRR pulse operations are shown to agree very well with operational data obtained from the ACRR. Finally, Razorback is expected to be a valuable tool for ACRR pulse performance prediction and ACRR reactor safety analyses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS2005

Development of a gravity-independent wastewater bioprocessor for advanced life support in space

Operation of aerobic biological reactors in space is controlled by a number of challenging constraints, mainly stemming from mass transfer limitations and phase separation. Immobilized-cell packed-bed bioreactors, specially designed to function in the absence of gravity, offer a viable solution for the treatment of gray water generated in space stations and spacecrafts. A novel gravity-independent wastewater biological processor, capable of carbon oxidation and nitrification of high-strength aqueous waste streams, is presented. The system, consisting of a fully saturated pressurized packed bed and a membrane oxygenation module attached to an external recirculation loop, operated continuously for over one year. The system attained high carbon oxidation efficiencies often exceeding 90% and ammonia oxidation reaching approximately 60%. The oxygen supply module relies on hydrophobic, nonporous, oxygen selective membranes, in a shell and tube configuration, for transferring oxygen to the packed bed, while keeping the gaseous and liquid phases separated. This reactor configuration and operating mode render the system gravity-independent and suitable for space applications.

NASA Discipline Life Support Systems↗

NASA NTRS2005

A molten Salt Am242M Production Reactor for Space Applications

The use of Am242m holds great promise for increasing the efficiency nuclear thermal rocket engines. Because Am242m has the highest fission cross section of any known isotope (1000's of barns), its extremely high reactivity may be used to directly heat a propellant gas with fission fragments. Since this isotope does not occur naturally, it must be bred in special production reactors designed for that purpose. The primary advantage to using molten salt reactors for breeding Am242m is that the reactors can be reprocessed continually yielding a constant rate of production of the isotope. Once built and initially fueled, the reactor will continually breed the additional fuel it needs to remain critical. The only feedstock required is a salt of U238. No enriched fuel is required during normal operation and all fissile material, except the Am242m, is maintained in a closed loop. For a reactor operating at 200 MW several kilograms of Am242m may be bred each year.

Emrich, William↗

NASA NTRS2023

Development of an Efficient Alternative to Recovery O2 From Metabolic CO2 Via Electrolysis Operated at Ambient Temperature and Driven By A Highly Selective Catalysis

The current State of Art (SOA) on oxygen recovery onboard the Environmental Control and Life Support System (ECLSS) at the International Space Station (ISS) is complex, heavy, and power consuming system that recovers approximately 50% of the oxygen (O 2 ) from metabolic carbon dioxide (CO 2 ). For future long duration beyond low earth orbit missions (LEO), O 2 recovery systems will need to be highly reliable, efficient, and recover a minimum of 75% O 2 from metabolic CO 2 . An alternative technology development effort currently underway at NASA Marshall Space Flight Center (MSFC) has the potential to significantly increase O 2 recovery currently limited to 50% (Sabatier) and reduce the complexity of ECLSS O 2 recovery. MSFC and University of Texas in Arlington (UTA) have jointly designed and fabricated a microfluid electrochemical reactor (MFECR) that operates at ambient conditions and utilizes a proprietary catalysis highly selective on reducing CO 2 to ethylene (C 2 H 4 ) at the cathode while O 2 is generated at the anode. The MFECR would replace three pieces of hardware for future ECLSS architectures: the current CO 2 Reduction Assembly (CRA) (Sabatier reactor), the Plasma Pyrolysis Assembly (PPA), and the Oxygen Generation Assembly (OGA). It is designed to interface directly with the CO 2 Removal Assembly (CDRA) and the Water Processing Assembly (WPA) to supply CO 2 reactant and water replenish respectively. This is expected to substantially improve sustainability of the ISS ECLSS and reduce requirement on power and weight. Here, we discuss the current development and evaluation efforts on different alternatives on not only the configuration and setup of the MFECR at an Engineering Design Unit (EDU) scale but also the selection of component materials.

Jesus A Dominguez↗

DOE OSTI2016

Coupled Calculations in Helical Steam Generator: Validation on Legacy Data

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Demonstration of Fully Coupled Calculations in Helical Steam Generator: Toward Predictions of Fluid Elastic Instability

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past seven years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In the last two years a significant effort was devoted to 1) develop and validate a fully coupled capability in SHARP for Fluid-Structure calculations (FSI) and 2) validate the use of SHARP for one-way coupled calculations in helical steam generators on legacy data available as well as novel experiments. In this report we discuss an extension of previous one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. It was demonstrated that in the buffeting regime one-way coupling was judged sufficient because the pressure loads do not cause substantial displacements. However, higher speeds led to a reduction in accuracy and the need for a fully coupled capability for velocity or approaching the critical velocity and fluid elastic instability. In this report we demonstrate such fully coupled capability on the Argonne test case discussed in previous reports.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Coupled Calculations of the NuScale FIV Tests – Pre-test Simulations and Sensitivity Analysis

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict FIV based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the conceptual design stage. Over the past several years, the Reactor Product Line has developed a FIV capability in SHARP. SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. In previous reports we have discussed the implementation and validation of an FIV capability for helical steam generators and fuel assemblies. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo for the NuScale FIV tests. In the turbulent fretting regime one-way coupling is judged to be sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators per the low flows expected in integral PWRs. The simulations precede the tests, as experimental data was not available at the delivery of this report. Several assumptions had to be made on the support-tube interaction and multiple tests and sensitivity analysis have been performed. Additional work remains to be done and it is discussed in the future work section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Modeling the Molten Salt Reactor Experiment with the RELAP5-3D code

The Molten Salt Reactor Experiment (MSRE) is a liquid-fueled, graphite-moderated, thermal reactor that was operated at Oak Ridge National Laboratory (ORNL) between 1965 and 1969 at a power of ∼8 MWt (thermal). The fuel (as enriched UF{sub 4}) was dissolved in the coolant, a fluoride liquid salt. In the MSRE, the fuel and the coolant were together in the same fluid-the molten salt. The objective of this effort is to modernize and recapture dynamic modeling capability for molten salt reactors. As part of this effort, the MSRE has been modeled with the code RELAP5-3D. This paper presents calculated results and compares them to available data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

One-Way Coupled Flow-Induced Vibration Analysis for a Twisted-Tube Heat Exchanger

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Assessing the interfacial corrosion mechanism of Inconel 617 in chloride molten salt corrosion using multi-modal advanced characterization techniques

The United States Department of Energy (DOE) has committed to expanding the domestic clean energy portfolio in response to the rising challenges of energy security in the wake of climate change. Accordingly, the construction of a series of Generation IV reactor technologies are being demonstrated, including sodium-cooled, small modular, and molten chloride fast reactors (MCFRs). To date, there are no fully qualified structural materials for constructing MCFRs. A number of commercial structural alloys have been considered for the construction of MCFRs, including alloys from the Inconel and Hastelloy series. Informed qualification of structural materials for the construction of MCFRs in the future can only be ensured by expanding the current fundamental knowledgebase of information pertaining to material performance under environmental stressors relevant to operation of the reactor, including corrosion susceptibility. The purpose of this investigation is to illustrate how a correlative multi-modal electron microscopy characterization approach, including the novel application of focused-ion beam 3D reconstruction capabilities, can elucidate the corrosion mechanism of a candidate structural material Inconel 617 for MCFR in NaCl-MgCl 2 eutectic salt at 700°C for 1,000 h. Evidence of intergranular corrosion, Ni and Fe dealloying, and Cr-O enrichment along the grain boundary, which most likely corresponds to Cr 2 O 3 , is a phenomenon that has been documented in other Ni-based superalloys exposed to chloride molten salt systems. Additional corrosion products, including the formation of insoluble MgAl 2 O 4 , within the porous network produced by the salt attack is a novel observation. In addition, Mo 3 Si 5 and τ 2 precipitates are detected in the alloy bulk and are dissolved by the salt. Furthermore, the lack of detection of design γ' precipitates in Inconel 617 after 1,000 h could indicate that the molten salt corrosion mechanism has indirectly induced a phase transformation of Al 2 TiNi (τ 2 ) and Ni 3 (Al,Ti) (γ’) phase. This investigation provides a comprehensive understanding of molten salt corrosion mechanisms in a complex material system such as a commercial structural alloy for applications in MCFRs.

36 MATERIALS SCIENCE↗

NASA NTRS1980

Factors affecting cleanup of exhaust gases from a pressurized, fluidized-bed coal combustor

The cleanup of effluent gases from the fluidized-bed combustion of coal is examined. Testing conditions include the type and feed rate of the coal and the sulfur sorbent, the coal-sorbent ratio, the coal-combustion air ratio, the depth of the reactor fluidizing bed, and the technique used to physically remove fly ash from the reactor effluent gases. Tests reveal that the particulate loading matter in the effluent gases is a function not only of the reactor-bed surface gas velocity, but also of the type of coal being burnt and the time the bed is operating. At least 95 percent of the fly ash particules in the effluent gas are removed by using a gas-solids separator under controlled operating conditions. Gaseous pollutants in the effluent (nitrogen and sulfur oxides) are held within the proposed Federal limits by controlling the reactor operating conditions and the type and quantity of sorbent material.

Rollbuhler, R. J.↗

DOE OSTI2017

Surrogate Models for TREAT Transient Calculations

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is currently preparing for resumption of operation after more than 20 years in a long-term standby state. In parallel with preparation for operations and a new transient testing program, the DOE-NE Advanced Modeling and Simulation program (NEAMS) has been supporting the development of three-dimensional transient multi-physics modeling capabilities within the reactor physics application MAMMOTH [1]. Recently, the MAMMOTH reactor physics team has begun development of uncertainty and sensitivity analysis for the TREAT core. This research will be used to support the analysis of reactor operation and experiment planning. Traditional approaches, i.e., Monte Carlo simulations, for such tasks require a large number of model evaluations under both steady state and transient conditions. However, exhaustive sets of sampling runs are simply not practical here since evaluations of each model can take a substantial amount of time. A solution to this problem is to build a surrogate model employing a smaller number of model evaluations. The surrogate model is designed to provide an accurate and fast approximation to the detailed TREAT model. In this manner, the surrogate model can be used in place of the TREAT transient model for uncertainty and sensitivity analysis at a fraction of the computational cost. The high dimensional model representation (HDMR) [2] method is employed here for approximating the responses of interest in terms of functions of lower dimensions. This method has been implemented in RAVEN [3] and all TREAT transient calculations are performed using the multiphysics code MAMMOTH. In addition, a code interface has also been developed in order to couple RAVEN with MAMMOTH. The surrogate model is then tested and compared for the MAMMOTH model of the TREAT startup testing program Transient Test #15 core [4], and its accuracy is measured by its ability to estimate the mean, coefficient of variance, skewness, excess kurtosis and probability density function of the outputs of Transient Test #15 model, as well as root mean square error metric.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

The role of biochar in the degradation of sugars during fast pyrolysis of biomass

With appropriate pretreatment, sugars can be a major product from fast pyrolysis of lignocellulosic biomass. Analytical pyrolysis of pure cellulose can produce up to 60 wt% yield of levoglucosan although yields are significantly lower in continuous pyrolysis at larger scales. Secondary reactions of vaporized levoglucosan are thought to be responsible for this loss of sugar yield, suggesting changes in the design and operation of pyrolysis reactors to minimize these reactions. Micropyrolysis experiments were performed to better understand the mechanism of sugar degradation in the presence of biochar. A 57% loss in levoglucosan yield was observed for cellulose overlain with untreated biochar powder compared to the pure cellulose control sample. The addition of biochar derived from pyrolysis of untreated corn stover to a fluidized bed pyrolyzer reduced sugar yields from cellulose from 61.3 wt% to 21.3 wt% and 41.5–11.6 wt% for conventional and autothermal operation, respectively. The significant drop in sugar yield due to biochar interaction inspired change in feeder configuration for the fluidized bed pyrolyzer to reduce vapor-char interactions. Biomass feeding was changed from in-bed to above-bed injection, which allowed significant devolatilization to occur above the layer of biochar that exists at the surface of the bed. Finally, by reducing secondary reactions, bio-oil and sugar yields increased by 7.9% and 14%, respectively, for autothermal pyrolysis.

09 BIOMASS FUELS↗

DOE OSTI2017

Serrated Yielding in Alloy 709 at 500 deg. C

Gen-IV nuclear power reactors are being designed to improve the efficiency, safety, reliability, and lifetime in comparison with the current fleet of nuclear power reactors. Hence, candidate structural materials for Gen-IV reactors need to have superior mechanical properties to withstand harsh reactor conditions including higher operating temperature, higher radiation dose, and extremely corrosive environment. Sodium Fast Reactor (SFR) is one of the Gen-IV reactor concepts and it has a high technology readiness level for deployment. It uses liquid sodium as coolant and fast neutron spectrum to induce fission with core outlet temperature of no higher than 550 deg. C. In general, austenitic stainless steels have an excellent corrosion resistance and good creep properties due to high chromium and nickel contents. Alloy 709 is an austenitic stainless steel with base Fe-20Cr-25Ni stabilized with nitrogen and strengthened with niobium. Its excellent mechanical properties relative to similar steels (304 and 316) including high strength at elevated temperature, corrosion resistance, thermal stability, sodium compatibility and creep properties make it an excellent candidate material for the nth-of-a-kind SFR structural applications. Understanding the plastic deformation behavior at service temperature for Alloy 709 is essential to predict the performance of the material under the operating conditions. In this study, the effects of the temperature and strain-rate on the uniaxial quasi-static tensile plastic deformation of Alloy 709 are discussed with special emphasis on dynamic strain aging and work hardening behavior.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Finite element analysis of compliant cladding and base metal systems

This report describes a roadmap towards a design method for high temperature nuclear cladded components that does not require long term testing of clad materials. Molten Salt Reactors (MSRs) will operate using corrosive fuel or coolant salts. The existing Class A metallic materials qualified for ASME Boiler and Pressure Vessel Code Section III, Division 5 design are not optimal for corrosion resistance when exposed to MSR salts. Corrosion could then severely limit the service life of MSRs. One option to alleviate this could be to Code qualify new corrosion resistant materials. However, this is a lengthy and expensive process and MSR developers have a near term need for a design solution that enables reactor designs with long service lives. An alternative would be to use cladded components – overlay some existing Class A material with a thin layer of some corrosion resistant material. This report describes the outlines of a design strategy that would enable this approach. The strategy relies on approximate design analysis methods for certain types of clad/base systems that avoids requiring long-term mechanical properties for the clad materials. This would allow the use of cladded components in the near term, without requiring long testing programs. However, it will limit the types of clad materials that can be used. This report describes the design framework, demonstrates the effectiveness of the approximate analysis methods, and confirms the approach using full scale finite element analysis of prototypical cladded high temperature components. It also describes a roadmap towards a complete design method and surveys available clad materials and bonding methods for suitably, considering only mechanical and not environmental effects.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

FFTF Acceptance and Startup Testing for GAIN

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following the demise of the breeder reactor program in the United States, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. FFTF operations ceased in April 1992 after a determination by DOE that no combination of proposed missions was financially feasible over a ten-year period. The reactor is currently deactivated and in a long-term surveillance and maintenance (S&M) mode. This report provides information on the extensive and rigorous process that was used to conduct turnover from construction followed by acceptance and startup testing of the FFTF. This paper is in support of the Gateway for Accelerated Innovation in Nuclear (GAIN), which provides the nuclear energy community with access to the technical, regulatory, and financial support necessary to move new or advanced nuclear reactor designs toward commercialization while ensuring the continued safe, reliable, and economic operation of the existing nuclear fleet. The information obtained from the design, startup, and operation of the FFTF provides valuable insight for follow-on reactor projects, such as the Versatile Test Reactor (VTR), in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, reactor startup and operations, and reactor deactivation and decommissioning (D&D). The focus of this report is on the process used to startup the FFTF and to ensure that operations could be conducted efficiently and safely. A reference section is provided of documents detailing the successful turnover and testing process implemented for startup of the reactor and its supporting systems. The documents listed can be retrieved upon request and are believed useful for future reactor startup endeavors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

A Novel Method for Controlling Crud Deposition in Nuclear Reactors Using Optimization Algorithms and Deep Neural Network Based Surrogate Models

This work presents the use of a high-fidelity neural network surrogate model within a Modular Optimization Framework for treatment of crud deposition as a constraint within light-water reactor core loading pattern optimization. The neural network was utilized for the treatment of crud constraints within the context of an advanced genetic algorithm applied to the core design problem. This proof-of-concept study shows that loading pattern optimization aided by a neural network surrogate model can optimize the manner in which crud distributes within a nuclear reactor without impacting operational parameters such as enrichment or cycle length. Several analysis methods were investigated. Analysis found that the surrogate model and genetic algorithm successfully minimized the deviation from a uniform crud distribution against a population of solutions from a reference optimization in which the crud distribution was not optimized. Strong evidence is presented that shows boron deposition in crud can be optimized through the loading pattern. This proof-of-concept study shows that the methods employed provide a powerful tool for mitigating the effects of crud deposition in nuclear reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗