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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 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Decommissioning Characterisation Through Compressive Gamma-Ray Imaging - 19308

We present the imaging results from two novel single pixel gamma-ray imaging systems, developed at ANSTO, in order to support the characterisation phase of decommissioning activities. The ability to remotely locate, identify, and quantify gamma emitting radionuclides can provide crucial information for the development of an efficient decommissioning plan. The imaging systems have been designed around the theory of compressed sensing and make use of a single spectroscopic detector and a set of dual, rotating, nested tungsten masks. The imaging system designs provide a large field of view and cover an energy range of 40 keV to 1.5 MeV. The first system has been specifically developed to image the High Flux Australian Reactor (HIFAR) reactor vessel with a hemispherical 2π field of view and provided quantitative information on the location of the gamma-ray emitting radionuclides present, whilst operating in a ∼10 Sv/h dose rate environment. The second system is a portable, high sensitivity imager with a cylindrical 360 deg. x 90 deg. field of view used for localising and identifying low to medium dose rate sources in the remaining facility infrastructure. Gamma-ray images were then accurately overlaid on optical images to allow for easy visualisation of the location of gamma emitting radionuclides. The spectroscopic gamma-ray imaging of point sources has been demonstrated with only 1/10. of the measurements normally required by traditional aperture based imaging techniques. The compressive gamma-ray imaging technique provides a novel and highly configurable spectroscopic imaging capability for applications in the nuclear industry. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

To Decommission or Not To Decommission? That is the Question... Revisited

Major Highlights Comparing 1998 to 2014: Favoring Plant Shutdown: Electric utility deregulation has created greater competition; Markets have been disrupted by cheap natural gas; Despite high availability, cost of electric power from natural gas still below break-even cost of nuclear generation; Emerging ISO acceptance of fluctuating renewable energy on the grid. Favoring DECON: D and D technologies have matured; Reactor Vessel and Internals segmentation methods still being refined; Nuclear workforce is 16 years older and leaving the nuclear industry (Impact on D and D TBD); Avoiding risk of more restrictive regulation and higher LLW disposal costs. Favoring SAFSTOR: Full funding of NDT still major driver for deferring immediate dismantlement. SAFSTOR vs. near term D and D influenced by: Status of and confidence in decommissioning fund (NDT) and cost estimate; Local and state politics; Public pressure. Most early shutdown plants tend to have under-funded decommissioning funds and opt for SAFSTOR. SAFSTOR allows for fund growth through interest compounding, but must assume that cost increase is less than fund growth: May not be true across the board, Much uncertainty/risk in waste disposal costs, regulatory requirements, etc. when projected out several decades. Once initiated SAFSTOR decision is not easily changed to DECON: Plant infrastructure equipment and components not readily available, RCS chemical decontamination can be very challenging, Key/knowledgeable plant personnel are gone. Consideration for phased approach to DECON: Remove >95% of the radiological source term under today's known regulations and cost by full System Decontamination of the primary circuit systems, and removal of NSSS and decontamination of the Containment Building; Improves risk profile in those areas most vulnerable to excessive cost growth; Able to defer D and D of the remainder of the plant to allow for the NDT to grow with interest.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Combined Raman imaging and {sup 18}O tracer analysis for the study of high temperature oxidation of pre-oxidized Zircaloy-4 in spent fuel pool accident - 6.4 STP-2016-0037

Oxidation at high temperature of Zr-based cladding materials is expected to be the primary cause of the fuel assemblies' degradation in Spent Fuel Storage Pool loss of cooling accidents. Differently from Loss Of Coolant Accident (LOCA) in reactor vessel, presence of air in the atmosphere in case of a spent fuel pool accident is known to be an aggravating factor because of the 'catalytic' role of nitrogen on the oxidation. 'Low temperature' oxidation scales simulating corrosion oxide formed during normal operation have been observed to have a protective effect regarding subsequent high temperature oxidation, in steam as well as in air and air + steam atmospheres. However, in the presence of nitrogen, this protective effect is lost much earlier than in pure steam or pure oxygen. We present results on oxygen transport through such corrosion oxide layers studied using an {sup 18}O tracer technique. Experiments were performed in {sup 18}O{sub 2} alone as well as in mixed {sup 18}O{sub 2} + N{sub 2} atmospheres, at 850 deg. C, a temperature for which nitrogen is known to have a strong effect on the Zr-based alloys oxidation. Post-test investigations of the specimens were performed by means of micro-Raman profiling or imaging, giving simultaneous information on the {sup 18}O distribution and on structural factors such as composition and stresses in the scales. Raman imaging gave clear evidence for different characteristic distributions of {sup 18}O in the scales. Some of them have been correlated with the development of cracks and porosity in the oxide which allows oxygen to locally penetrate in the scales. Although nitrogen appears to have no or limited influence on the oxygen diffusion, it is observed to reach the metal/oxide interface much faster than oxygen. A mechanism explaining the role of nitrogen on the loss of the protective effect and the kinetic acceleration observed by TGA is proposed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Combined Raman imaging and {sup 18}O tracer analysis for the study of zircaloy-4 high-temperature oxidation in spent fuel pool accident - 2016-0037

Oxidation at high temperature (HT) of cladding materials is expected to be the primary cause of the fuel assemblies degradation in spent fuel storage pool loss of cooling accidents. Unlike a loss-of-coolant accident (LOCA) in a reactor vessel, the presence of air in the atmosphere is expected in a spent fuel pool accident and is known to be an aggravating factor because of the 'catalytic' role of nitrogen on the oxidation. In steam LOCA conditions, a low-temperature (LT) oxidation scale simulating the corrosion oxide formed during in-service use of the assemblies has been observed to have a protective effect regarding subsequent HT oxidation. In the presence of nitrogen, HT oxidation kinetic experiments show that the protective effect of an LT corrosion scale also exists but is lost much earlier than in pure steam. Oxygen transport through LT oxide layers has been studied using the {sup 18}O tracer technique. We performed experiments in {sup 18}O{sub 2} alone as well as in mixed {sup 18}O{sub 2}+N{sub 2} atmospheres, at 850 deg. C. Micro-Raman imaging, both at the specimen surfaces and on cross sections gave clear evidence for different characteristic distributions of {sup 18}O in the scales. Some of these distributions have been correlated with the presence of cracks and porosity in the LT oxide, which allows oxygen to locally penetrate in the scales. In cases in which no radial crack is present, apparent oxygen diffusion coefficients in the oxide were derived from fitting {sup 18}O diffusion profiles. Nitrogen appears to have no or only a limited influence on the oxygen diffusivity, but it is observed to reach the metal-oxide interface faster than oxygen, at which point it reacts with the metal and the sub-stoichiometric oxide to form zirconium nitride.The later conversion of the ZrN into ZrO{sub 2} might lead to the destabilization of the LT scale and to the loss of its protective effect. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2014

A Code of Radiation Damage by Gamma - Paper 62

In recent years, the study of radiation damage of materials has seen a thrust because of the expectation of the high rate of radiation damage in fusion reactors, accelerator driven sub critical systems and damage of a control system of space vehicles by the high energy neutrons and gamma showers. Embrittlement of reactor vessels, corrosion of steel and swelling of fuel rods are the well known empirical examples of severe radiation damage. It is also realized that damage by high energy gammas is in no way insignificant, and it is equivalent to the damage by the neutrons. Secondly, value addition of certain materials like enhancement of electrical resistance, mechanical strength, color enhancement of precious stones on irradiation by neutrons and gammas and polymerization in the presence of gamma radiation add to the necessity of a Monte Carlo (MC) code of radiation damage in a wide range of energy. The JAIPU code, for the study of radiation damage by neutrons up to 10 MeV energy, incorporates features like development of secondary atomic cascade and PKA energy corrected for the ionization lose and production of Frenkel pair. MC estimation of both the primary and secondary displacements has been compared with the Frenkel pairs calculated using the NRT model. In the present work, JAIPU code is extended for irradiation of materials by high energy gammas. Here cascade generated by the incident gamma produces electrons by way of photo-electric, Compton and pair production until it is absorbed. The recoiled Frenkel pairs produced in electron + atom collision have been considered for the purpose of atomic cascade. Standard cross section libraries for gamma and electron interaction are used while for atom + atom collision we have made use of the IOTA code. In addition, in the code, the lattice structure of the irradiated material is simulated to know the positions of the photon interactions with atoms and the energy of electron it adds an extra feature of the position and size of the voids and the clusters produced on irradiation. The code has the possibility of calculation of 'absorbed radiation dose' and 'KERMA' and the interatomic distances of the displaced atoms. It may be pointed out that the code uses only the MC simulation of the number of displacements unlike several other approaches where the NRT model with PKA energy is used for gross estimation of the number of the displaced atoms. Thus, the code keeps an account of the positions of displaced atoms for the in-depth modification of the irradiated material. 1 MeV gamma seems to damage Fe up to several millimeters of its depth and this is mostly by the recoiled Compton electrons than photo electrons. The code is in the developmental stage and it has the capability of applications in the developing field of Radiology besides the shielding and other studies of radiation damage. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Self-powered Wireless Dual-mode Langasite Sensor for Pressure/Temperature Monitoring of Nuclear Reactors

Safety issues in nuclear power plants are of extreme importance. Numerous sensors are used within nuclear power plants for sensing and monitoring of its critical operation parameters. As currently implemented, during massive losses of power across the plant in accidents, the backup batteries for the monitoring sensors will ultimately be drained and can cease to operate. The overall objective of this NEUP project is to develop a novel self-powered wireless hybrid sensor which can accurately monitor both pressure and temperature using a single device, even in the extreme harsh environments of severe nuclear accidents, without the requirement of external electricity. At Virginia Tech side, a heat pipe assisted thermoelectric generator (TEG) energy harvester and associated power management and data transmission circuits were developed. The energy harvester can produce electrical power continuously, even in loss of plant and external power situations, providing power for important data transmission for accident diagnosis. The self-powered sensing system could be integrated directly onto key nuclear components including pipes, pump housings, heat exchangers, reactor vessels, and shielding structures inside and outside of the reactor core. Two TEG based energy harvesters which can work in relatively low (50-250 ºC) and high temperature ranges (250-350 ºC) were established and tested. In the prototypes, heat pipes were introduced to reduce the heat resistances of the device at the hot and cold ends. The power output of the heat pipe assisted TEG module is 2.22 W, which is 6 times more than one using an aluminum rod. In-lab tests showed that these devices canmeet the requirement to power the sensors and circuit during normal and off-normal conditions.Amodel was then built to analyze the thermal network in the energy harvester, and was used to optimize its performance and reduce its geometry size. The design and testing of the vibrational energy harvester were completed. The harvester, has shown to effectively increase the vibrational potential of the main coolant pump, though it has been shown that thermal energy has a much higher potential for the powering of sensors. In this project, a self-start charger circuit for the energy harvester was designed to regulate and manage the power. An energy storage element was incorporated into the energy harvesting module for sensors and wireless transceivers in an off-normal situation. The signal conditioning and wireless communication electronic circuits were developed and tested in the lab. An important consideration for an energy harvester within a nuclear power plant is the level of radiation it will encounter during operation. The two main sources of radiation in a nuclear power plant are gamma and neutron radiation. Aradiation experiment was conducted in the Westinghouse Electric Company during which the prototypes were placed in a 106 rad radiation environment to test their ability to operate in the extreme environment. The results showed that this level of radiation had little effect on the performance of the TEGmaterial and heat pipe heat exchanging system. Its influence on the circuit was still under research, however the need for more development on radiation hardened electronics is needed. Pressure measurement based on piezoelectric crystal resonators has advantages over the commercially available methods like capacitive, electromagnetic, optical, potentiometric, and piezo resistive strain gauges. Since the output of piezoelectric based sensors are digital in format, it has a very high resolution, high accuracy, and long-term stability. Quartz is one of the commonly used piezoelectric material in digital pressure sensors design. This material is elastic in nature, providing high stability, high repeatability, high elasticity, repeatable mechanical behavior for many cycles of operation, and is free of hysteresis. The quartz resonator has a high Q, which means that very high vibration can be driven with the very low electric power. Langasite (LGS) is a new promising piezoelectric material that combines the advantageous properties of quartz with a better performance at high temperatures. LGS will not undergo any phase transitions up to its melting temperature of 1473 C, and hence, the application of Langasite in measuring pressure at high temperatures is promising. In this project, the design and fabrication of the piezo oscillator crystal for temperature and pressure measurement has been completed. The entire sensor has been fabricated and tested to see its sensitivity to temperature and pressure changes. The design and optimization of the radiation shielding systems that can effective protect the duel-mode Langasite sensor inside reactor core were demonstrated. To maximize the radiation shielding effectiveness, different types of materials are investigated based upon recent developments of radiation shielding systems including aluminum + Boron carbide and bismuth borosilicate glass B2O3 in addition to tungsten/B4C. New materials based upon metatheses is included as potential shielding to mitigate neutron fluencies. The radiation shielding effectiveness of selected composites are evaluated within a broader neutron spectrum from thermal, epithermal, fast (1 MeV) and 15 MeV neutrons. Performance analysis was also conducted to evaluate the effects of selected composite systems.Aconceptual design was also proposed for the packaging and integration of the radiation shielding system with the sensor, and its applicability within the reactor core was discussed. At end of the project, all components necessary for a complete sensing and monitoring system, including a langasite sensor, electrodes, the TEG harvester, an energy storage device, and power management and transmission electronics were integrated to demonstrate feasibility in powering the sensor. Finally, integration of each component into a full package, and a full demonstration were done.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Best Practices for Shielding Analyses of Activated Metals and Spent Resins from Reactor Operation

This report investigated best practices for performing shielding evaluations of Type B waste packages, as defined in 10 CFR Part 71 regulations on packaging and transportation of radioactive material, or packages for which the contents are not defined before loading and may include a broad range of nuclides, geometries, and non-fuel materials. The following non-fuel waste streams were analyzed: Activated metals from decommissioned commercial power reactors, including Type 304 stainless steel, reactor vessel steel, and Inconel, Control blades from boiling water reactors, Neutron-activated corrosion products on surfaces of activated metals, and Spent resins from power plant operations. Measured elemental compositions, including major constituents and impurities, for steel and Inconel samples from commercial power reactors were used in activation calculations to determine radionuclide inventories in activated metals. For a simplified cask model, 60 Co contribution to the total external package dose rate at 30 days after shutdown varied from approximately 60% to 95%, depending on the activated metal, initial cobalt impurity concentration in the metal, and the thickness of the overpack gamma shield. Its maximum contribution to the total external package dose rate of approximately 100% was reached within the time interval of 2 to 5 years after shutdown and was maintained for up to 45 to 60 years after shutdown, depending on material, initial cobalt impurity concentration, and shield thickness. Thereafter, the 60 Co contribution to external package dose rate decreased with increasing decay time. Cobalt-60 is primarily produced by neutron reactions with the cobalt impurity in steel and Inconel. Other important radionuclides in activated metals contributing to package external dose rate are radionuclides with relatively short decay times, including 51 Cr, 59 Fe, 58 Co, and 54 Mn. These radionuclides may be represented as an equivalent 60 Co activity/source because 60 Co gamma ray emissions are bounding in terms of source strength and energy to other important radionuclides identified in the analyzed activated metals. Approaches for modeling the neutron-activated corrosion products that may be attached to activated reactor components were analyzed in this report. It was demonstrated that a surface source is more conservative than a uniform volumetric source for the treatment of neutron-activated corrosion products with respect to external package dose rates. An analysis of the maximum radionuclide loadings reported on spent resins identified the radionuclides 137 Cs, 60 Co, 134 Cs, 65 Zn, and 58 Co as the primary contributors to external package dose rate. For a resin cooled for 3.08 years, the external package dose rate was entirely produced by the reported 137 Cs and 60 Co inventory. The neutron sources from actinides found on spent resins or activated metals produced negligible dose rates and may be ignored in dose rate analyses. Effects of idealized waste material, source geometry, and spatial material/source distributions on external package dose rates were determined based on dose rate results for a simplified cask model under normal conditions of transport. Type 304 stainless steel, zirconium, and aluminum with adjusted mass densities based on a maximum content weight were analyzed for material modeling effects on external package dose rate. These materials produced identical external package dose rates within the statistical uncertainties of the dose rate estimates. Among four different source geometry configurations with homogeneous material of different mass densities, uniform volumetric source distribution, and the same source strength, the geometry configuration with lowest mass density (i.e., minimum self-shielding effect among the four cases) was most conservative. Spatial source distributions that better represent localized peak 60 Co activity values were more conservative than a uniform volumetric source distribution, assuming the same weight and total source strength per package. The increase in external dose rate caused by localized source peaks can be as much as the ratio between source peak activity density to average activity density, depending on the location of the activated metal with peak activity density. Therefore, the shielding analysis may be simplified if localized peak activities and the average activity per package can be measured/determined and documented at the time of cask loading. For simplicity, external package dose rates may be determined based on average source activity and a uniform volumetric source distribution. The dose rate results from that calculation model multiplied by the ratio of peak activity density to average activity density will produce maximum dose rate values for conservative estimates.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2009

Fuel-Cell Power Systems Incorporating Mg-Based H2 Generators

Two hydrogen generators based on reactions involving magnesium and steam have been proposed as means for generating the fuel (hydrogen gas) for such fuel-cell power systems as those to be used in the drive systems of advanced motor vehicles. The hydrogen generators would make it unnecessary to rely on any of the hydrogen storage systems developed thus far that are, variously, too expensive, too heavy, too bulky, and/or too unsafe to be practical. The two proposed hydrogen generators are denoted basic and advanced, respectively. In the basic hydrogen generator (see figure), steam at a temperature greater than or equals 330 C would be fed into a reactor charged with magnesium, wherein hydrogen would be released in the exothermic reaction Mg + H2O yields MgO + H2. The steam would be made in a flash boiler. To initiate the reaction, the boiler could be heated electrically by energy borrowed from a storage battery that would be recharged during normal operation of the associated fuel-cell subsystem. Once the reaction was underway, heat from the reaction would be fed to the boiler. If the boiler were made an integral part of the hydrogen-generator reactor vessel, then the problem of transfer of heat from the reactor to the boiler would be greatly simplified. A pump would be used to feed water from a storage tank to the boiler.

Kindler, Andrew↗
DOE OSTI2020

Initial development for the introduction of provisions for irradiation effects in Division 5 rules

Heat pipe microreactor concept involves the use of a solid core block to encapsulate the nuclear fuel and the passive extraction of heat from the core block to the heat exchanger block by advanced heat pipes. The core block is a key component of the heat pipe microreactor design. It integrates the functions of reactor vessel, core structural components and fuel cladding of traditional Gen IV advanced reactor designs into one single structure. The effects of irradiation on the structural failure modes are outside the scope of the rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B for metallic Class A components. Methods to account for the irradiation effect were investigated in order to develop design methods that can be applied to the core block design in the heat pipe microreactor. This report presents an initial development of an approach to account for the irradiation effects in the design rules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Reactor cell neutron dose for the molten salt breeder reactor conceptual design

The private sector’s interest in the active development of molten salt reactors has led to the need to develop and test advanced modeling and simulation tools to analyze various advanced reactor types under numerous conditions. This paper discusses the effort undertaken to model the Oak Ridge National Laboratory (ORNL) Molten Salt Breeder Reactor (MSBR) design using ORNL’s Shift Monte Carlo code. The MSBR model integrates a Monte Carlo N-Particle (MCNP) MSBR core model with an MCNP model that was generated from a CAD model of the external components and the reactor building, which was subsequently run in Shift. This paper focuses on development of the fully integrated model and its use in performing neutron transport calculations in the reactor cell area. This model is intended to aid in understanding radiological dose conditions during operation, as well as the iron dpa rates in the reactor vessel. The neutron biological dose rates and flux calculated in the reactor cell are much higher in the MSBR than in typical light-water reactors. The implications of these results and future work are also discussed in this paper.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Verification of numerical models for seismic fluid-structure interaction analysis of internal components in liquid-filled advanced reactors

Earthquake shaking of a liquid-filled advanced reactor induces fluid-structure interaction (FSI) between the reactor vessel (tank), its internal components, and the contained liquid. Seismic design, qualification, and risk assessment of such reactors must consider fluid-structure responses, their geometries and support conditions, and three-directional seismic inputs, all of which require numerical simulations. Numerical models used for analysis of safety-related nuclear equipment must be verified and validated. Here, this paper verifies numerical models of submerged components using analytical solutions. Seismic FSI analysis of the numerical models is performed using the Arbitrary Lagrangian-Eulerian (ALE) and Incompressible Computational Fluid Dynamics (ICFD) solvers in LS-DYNA. Prior analytical solutions are reworked in this paper, and calculation errors are identified and corrected. The solutions address frequencies of two concentric cylindrical pipes filled with liquid: the inner pipe is a submerged component, and the outer pipe contains the liquid. A unitless frequency coefficient is defined and presented for pipes of different materials, filled with different liquids, and with a range of dimensions for application to advanced reactors. The numerical models are verified here by comparing the lateral frequencies of submerged components with those calculated using the corrected analytical solutions. Recommendations for verification of numerical models of internal components in advanced reactors for seismic FSI analysis are provided. Although the unitless frequency coefficients and verification procedures are developed for application to advanced nuclear reactors, they are broadly applicable to FSI analysis of submerged components in liquid-filled vessels such as storage tanks, boilers, and steam generators.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Seismic attenuation system for a nuclear reactor

A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Seismic attenuation system for a nuclear reactor

A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Real-Time In-Pile Acoustic Measurement Infrastructure at the Advanced Test Reactor

The core of any nuclear reactor presents a particularly harsh environment for sensors and instrumentation. The reactor core also imposes challenging constraints on signal transmission from inside the reactor core to outside of the reactor vessel. In this paper, an acoustic measurement infrastructure installed at the Advanced Test Reactor (ATR), located at Idaho National Laboratory, is presented. The measurement infrastructure consists of ATR in-pile structural components, coolant, acoustic receivers, primary coolant pumps (PCPs), a data acquisition system, and signal-processing algorithms. Intrinsic and cyclic acoustic signals generated by the operation of the PCPs are collected and processed. Finally, the characteristics of the intrinsic signal can indicate the process state of the ATR (such as reactor startup, reactor criticality, reactor attaining maximum power, and reactor shutdown) during operation (i.e., real-time measurement).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Development of CTF Capability for Modeling Reactor Operating Cycles with Crud Growth

CTF is a thermal-hydraulics simulation tool designed for modeling both fluid and solid regions inside light water reactor vessels. It forms the thermal-hydraulic package of the Virtual Environment for Reactor Applications Core Simulator being developed by the Consortium for Advanced Simulation of Light water reactors. The project documented by this report has been undertaken to add a new crud-modeling capability to CTF. Achieving this goal required the completion of three specific tasks: (1) developing a method for running CTF simulations of reactor operating cycles , (2) developing and implementing a basic crud modeling “surrogate”, and (3) developing and implementing a method for inputting detailed core power distributions. These three tasks have been completed and have resulted in a new feature that allows users to run long-time-scale simulations with CTF as a collection of individual, steady-state points that comprise a larger transient. The user is able to drive inlet mass flow rate, inlet temperature, outlet pressure, and core power distributions to new values for each state in the simulation. The crud modeling tool is used at each of these state points to grow crud as a function of current thermal hydraulic conditions of the previously solved state. The developed crud modeling tool makes many simplifications that allow it to make a crude approximation of general crud behavior. It is envisioned that the infrastructure that has resulted from this project will be utilized in the future to facilitate coupling to more advanced crud chemistry codes. This document details work completed during this project and demonstrates the verification of this new capability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Application of Electrochemical Methods to Molten Salt Reactors: Draft TLR Documenting Assessment of Electrochemical Monitoring

Nuclear Regulatory Commission (NRC) is developing the regulatory framework and technical expertise to support regulatory review of advanced non-water reactors, including molten salt reactors (MSRs). In an MSR, it is essential that the salt chemistry be maintained in a desired range in terms of redox potential for reliable operation and mitigation of corrosion to structural materials in the reactor, particularly the reactor vessel and heater exchanger. Measuring the redox potential of the salt in the reactor would also allow for the material lifetimes to be predicted more accurately, and chemical issues to be diagnosed more quickly. In addition to the chemical composition analysis by ICP-MS, electrochemical methods including potentiometry and linear wave scanning (LSC) were also used in molten salt reactor experiment (MSRE) for redox potential monitoring purposes. Electrochemcial methods offers unique advantages such as quick turnaround in results and unique capability of in-line monitoring of redox potential, and are considered popular electroanalytical techniques that can be used to monitor redox potentials and salt chemistry including impurities. The last two decades have seen significant advances in science and engineering of electrochemical methods for potential application to molten salts. The primary goal of this report is to assist NRC in understanding the monitoring of the salt chemistry by electrochemical methods and provide NRC reviewers with necessary information and tools to support regulatory review of MSR designs. This reports consists of two majors parts—Part 1, chemical potential of molten salts and effects by fission process in MSR; Part 2, assessment of electrochemical methods for application to MSRs. The TRLs of the typical relevant electrochemical methods for molten salts were evaluated based on the DOE TRL guidelines and upon a review of the current status of the electrochemical methods for two typical salt systems, fluoride and chloride, for MSRs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Sample Testing Basket: Design and Capabilities Report

SFR designs all require integral components that can withstand prolonged submersion in the liquid sodium primary coolant as well as in the gas space of the reactor vessel. Materials are often desired to withstand these conditions for the lifetime of the reactor. The Sample Testing Basket (STB) was designed and fabricated to meet the anticipated need for material exposure testing in prototypic reactor coolant environments; in the liquid sodium primary coolant and the sodium-vapor-laden gas space. Developed at Argonne National Laboratory (ANL), the STB is a piece of equipment designed for use in the Mechanisms Engineering Test Loop (METL) to allow for sodium exposure testing of materials within the METL facility.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Process Intensification and Scale-Up of a Continuous Enzymatic Hydrolysis and Separation Process

Combining separate unit operations into one where the best of each part can be maximized is one of the benefits of process intensification. An example is the combination of lignocellulosic biomass enzymatic hydrolysis with the downstream solid-liquid separation step to produce clarified sugars ready for fermentation or catalytic upgrading. The productivity and endpoint yield of enzymatic hydrolysis both enjoy the benefits of reduced feedback inhibition through the continuous removal of sugars by incorporating separations into the reactor. Likewise, the efficiency of recovering clarified sugars from the enzymatic hydrolysis slurry can be enhanced by operating separations equipment at steady-state conditions simultaneously with the continuously fed hydrolysis process. A key parameter that enables greater processing capacity while also raising the risks of failure is the solids loading or concentration. Higher solids loading allows for smaller reactor vessels and results in clarified sugars of higher concentration; however, required pumping power increases, reactor agitation may become ineffective, and membrane flux suffers. Feedstock material attributes influenced by upstream pretreatment must also be scrutinized more carefully: dilute-acid pretreated and deacetylated-and-disc-refined feedstocks exhibit different characteristics that affect agitation and pumping. The authors invite you to further explore the process science enabling the scale-up of this technology from conceptual work at the bench to pilot-scale industrially-relevant equipment where the challenges and solutions of integration and process optimization are expounded upon.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗