Results for “REACTOR OPERATION” — Page 36

DOE OSTI2023

On-the-Fly Energy Condensation for Whole-Core Multiphysics Simulations

As the nuclear energy industry considers ways to achieve improved economics in the current fleet of light-water reactors (LWRs), one possible approach is to operate each cycle longer. This causes a greater portion of the fuel to be burned and reduces the frequency of outages, which ultimately reduces the cost to operate the reactor. However, this also leads to higher burnup fuels than have traditionally been allowed in these reactors. There are concerns about integrity of high-burnup (HBu) fuel, especially during accident conditions such as loss-of-coolant accidents (LOCAs), as shown by Capps et al. To investigate these concerns, advanced modeling and simulation capabilities are under development to determine the susceptibility of HBu fuel to fuel fragmentation, relocation, and dispersion (FFRD). Improvements have previously been made to fuel performance capabilities to model these phenomena more accurately; multiphysics simulations have also been conducted to determine the power and burnup histories of the HBu fuel, which are needed as inputs to the fuel performance calculations. Most recently, new statistical approaches have been developed to identify a subset of fuel rods that have greater FFRD susceptibility, reducing the total number of fuel performance simulations required.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

WEC Irradiation Can Thermal Testing Report

With renewed interest for the next generation of small modular reactors and space-based nuclear reactors, metal hydrides have been a focus of research for moderating materials to reduce the need for highly enriched fuel to maintain small core sizes. Yttrium hydride (YH) is an attractive candidate for these metal hydrides due to its ability to retain hydrogen at elevated temperatures as well as the low neutron absorption cross section of yttrium metal. Given the necessity for reactors to operate at elevated temperatures for extended periods of time, understanding hydrogen retention in moderator channels in simulated operating thermal conditions is paramount in predicting reactor performance and the long-term stability of YH. Additionally, coating strategies on the moderator encapsulating will be explored to ascertain if it aids in retention of hydrogen within the capsule and moderator.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

DEVELOPMENT OF AN APPROACH FOR DOWNBLENDING AND DISPOSITION OF IRRADIATED GRAPHITE FUEL

The IGR reactor, a pulse-type test experimental reactor, reached full power in 1961. The reactor contains a center axial experimental ampoule that is subjected to a large pulse of neutrons (neutron flux maximum density of 7×1016 n/cm2·s, peak power 10GW) when the reactor is operated. The fuel form for the IGR reactor is graphite cubes and rods impregnated with highly enriched (90%) uranyl-dinitrate (UO2(NO3)2·6H2O). This fuel is eligible for return to Russia for reprocessing under the Russian Research Reactor Fuel Return program. However due to some technical and licensing issues it was decided to immobilize this fuel in Kazakhstan. The disposition approach thus selected for the spent HEU fuel involves down-blending of the fuel followed by cementation. Development of the process for down-blending and subsequent cementation requires a process that satisfies many criteria, including fissile content, compressive strength, permeability, and containment of radioactivity, among others. One key goal of the final product is to meet the criteria for termination of safeguards. This paper will describe the overall challenge associated with disposition of irradiated HEU graphite spent fuel, the identification of the objectives of the disposition process, and the constraints associated with development of the process and the waste form.

Gaines, Kris↗

DOE OSTI2023

Replacement of Legacy Analytical Codes at the Advanced Test Reactor

For each operating cycle of the Advanced Test Reactor (ATR) at Idaho National Laboratory, a Core Safety Assurance Package (CSAP) is necessary to demonstrate compliance with the safety basis approved by the United State Department of Energy (DOE). Certain computer codes are used in CSAP development, most of them developed in-house. This work describes replacement of a large set of these codes and updates previous work. Replacement of legacy codes is necessary due to computer hardware failure but also has generally improved user-friendliness.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Replacement of Legacy Analytical Codes at the Advanced Test Reactor

For each operating cycle of the Advanced Test Reactor (ATR) at Idaho National Laboratory, a Core Safety Assurance Package (CSAP) is necessary to demonstrate compliance with the safety basis approved by the United State Department of Energy (DOE). Certain computer codes are used in CSAP development, most of them developed in-house. This work describes replacement of a large set of these codes and updates previous work. Replacement of legacy codes is necessary due to computer hardware failure but also has generally improved user-friendliness.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2026

MARVEL Reactor Fuel Performance Report

The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Development of human machine interface (HMI) for digital control rooms in nuclear power plants - 119

The analog control systems of traditional Nuclear Power Plants (NPPs) are being replaced by digital systems due to the fact that analog Instrumentation and Control (I and C) systems are becoming obsolete and such technology is no longer supported by manufacturers. The purpose of this study is to jointly evaluate the performance of an operator with a Human Machine Interface (HMI) in a NPP digital control room ensuring the efficiency and the safety of the NPP. The validity and the accuracy of the operators' actions are the primary goals that have to be preserved through the emergence of the HMI based in digital systems. The transition from analog to digital control systems leads to a more advanced support of the NPP; Video Display Units (VDUs) are used for monitoring the state of the NPP while alarm systems help for directing attention. The operators, in order to monitor information from the different VDUs, have to switch attention between different screens (alarms systems and control systems). Therefore, a well-established HMI is a significant challenge to the overall operation of the NPP, that actually helps the operators to learn from the available information in detecting possible anomalies and making the right decisions, to ensure safety, and in parallel to diminish the human error. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

MARVEL Reactor Fuel Performance Report (Rev.2)

The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Production and Chemical Analysis of NaCl-KCl-UCl 3 Salts

Thermochemical and thermophysical properties of several molten salt systems of interest to molten salt reactor (MSR) developers are being measured at Argonne National Laboratory. These properties are needed by developers to design, license, and operate their reactors. Thermochemical properties being measured at Argonne include thermal transitions, phase behavior, and heat capacity. Thermophysical properties being measured at Argonne include liquid density, surface tension, volumetric thermal expansion of the liquid phase, thermal diffusivity, thermal conductivity, and viscosity. Properties of the ternary NaCl-KCl-UCl 3 system that may be used as a fuel bearing salt are being measured. Low-melting eutectic salt compositions are of interest to developers owing to the thermal efficiency. Using salts with low melting temperatures allow reactors to be operated at a lower temperature to improve operation and decrease safety concerns. Confirming low melting behaviors are maintained during composition changes that occur during fission is also an important factor. It is essential to any molten salt reactor developer utilizing a fuel bearing salt to have a complete understanding of the behavior of that fuel salt at a variety of compositions near the eutectic composition they plan to use. This kind of stability is indicated on a phase diagram by low liquidus slopes at compositions bordering the eutectic. Therefore, a thermodynamic model was constructed to determine what compositions of the ternary NaCl, KCl, and UCl 3 system should be studied to support MSR developers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

EPRI Guidance for Transitioning from Operation to Decommissioning at Nuclear Power Plants - 17438

A wide range of key activities are necessary after permanent shutdown of a nuclear power plant and before active dismantlement of the plant can begin. This period between permanent shutdown and the start of active dismantlement is typically referred to as the transition period (or the post-operational phase). In some cases, these transition activities may be prescribed by regulation or performed in the absence of regulations, while in other cases they may be more practically driven or even optional. In either case, it is often unclear what activities are needed and when/how those activities should be conducted to most efficiently transition from operational to decommissioning status while maintaining radiological safety and the protection of the public. The Electric Power Research Institute has prepared a report that presents experiences and guidance for transitioning from an operational to decommissioning status based on US and international decommissioning regulations and plant experiences. Regulations and experiences in a number of countries were considered, including US, Germany, France, Spain, and Switzerland. Best practices and lessons learned related to transitioning to decommissioning are identified, and guidance is presented for developing a plan to transition from normal operations to decommissioning for several scenarios, including: - Transition to DECON (immediate dismantling) for a planned shutdown; - Transition to SAFSTOR (deferred dismantling) for a planned shutdown; - Transition to DECON (immediate dismantling) for an unplanned shutdown; - Transition to SAFSTOR (deferred dismantling) for an unplanned shutdown. Further, the EPRI report contains guidance for currently operating plants that do not anticipate permanent shutdown in the near future to help minimize the cost and impact of an unexpected permanent shutdown if such a scenario were to occur. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Supporting the industry by developing a design guidance for computer-based procedures for field workers - 274

The paper-based procedures currently used for nearly all activities in the commercial nuclear power industry have a long history of ensuring safe operation of the plants. However, there is potential to greatly increase efficiency and safety by improving how the human interacts with the procedures, which can be achieved through the use of computer-based procedures (CBPs). A CBP system offers a vast variety of improvements, such as context driven job aids, integrated human performance tools and dynamic step presentation. As a step toward the goal of improving procedure use performance, the U.S. Department of Energy Light Water Reactor Sustainability Program researchers, together with the nuclear industry, have been investigating the feasibility of replacing current paper-based procedures with CBPs. The main purpose of the CBP research for nuclear field workers conducted at the Idaho National Laboratory was to provide design guidance to the nuclear industry to be used by both utilities and vendors. After studying existing design guidance for CBP systems, the researchers concluded that the majority of the existing guidance is intended for control room CBP systems, and does not necessarily address the challenges of designing CBP systems for instructions carried out in the field. Further, the guidance is often presented on a high level, which leaves the designer to interpret what is meant by the guidance and how to specifically implement it. The authors developed a design guidance to provide guidance specifically tailored to instructions that are carried out in the field based. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

NASA NTRS1975

Performance of a catalytic reactor at simulated gas turbine combustor operating conditions

The performance of a catalytic reactor 12 cm in diameter and 17 cm long was evaluated at simulated gas turbine combustor operating conditions using premixed propane and air. Inlet temperatures of 600 and 800 K, pressures of 3 and 6 atm, and reference velocities of 9 to 30 m/s were tested. Data were taken for equivalence ratios as high as 0.43. The operating range was limited on the low-temperature side by very poor efficiency; the minimum exit temperature for good performance ranged from 1400 to 1600 K depending on inlet conditions. As exit temperatures were raised above this minimum, emissions of unburned hydrocarbons decreased, carbon monoxide emissions became generally less than 1 g CO/kg fuel, and nitrogen oxides were less than about 0.1 g NO2/kg fuel.

Anderson, D. N.↗

DOE OSTI2017

Toward biologically inspired self healing digital embedded devices: BIO-SymPle - 388

One of the more interesting concepts being postulated for next gen power plants, is the notion of highly resilient and survivable I and C systems. While resilience and survivability are 'cross system attributes', resilience and robustness will rely on highly robust and fault resilient devices as a foundation. This paper presents a new self-healing digital I and C architecture inspired by the operation of two biological concepts: human immune system and embryonic stem cells. The purpose of this research is look at different bio-inspired paradigms for organizing system functions so that they become inherently resilient due to device organization and behaviors. Our current proposed approach (called BIO-SymPle) is organized in two levels: critical service layer, which is providing the intended service of critical application, and healing layer, that supervises the operation of all repair mechanisms. To facilitate coordination, organized behavior among cells we employ network-based topology structures which is comprised of four active cells, four pre-generated redundant cells, and one embryonic stem cell. Collectively, the ensemble represents the building block that executes the local functionality based on the expression for DNA genetic codes stored inside each cell. The four bio-operational active cells are protected by their neighboring spare cells that imitate how white blood cells are distributed around the antigens in their locations. Ultimately, the proposed multi-layer architecture is able to cover a significant number of coincident faults and we believe architectures like BIO-SymPLe can positively impact the next generation digital I and C system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Determination of a Generic Human Error Probability Distribution, Part 2: A Dynamic SPAR-H Example Application

Humans are an integral part of a nuclear power plant (NPP) from its construction, to its everyday maintenance and operation. In the nuclear industry there are many approaches to quantifying component reliability, which is called probability risk assessment (PRA). In PRA there are two considerations given to the quantification, static and dynamic. Static is quantified before or after an operation; and dynamic is time dependent, in that the probability changes as real time data are provided. In PRA, the importance of the components is calculated, but what of the importance of the human component? There are several methods for human reliability analysis (HRA), which include but are not limited to: THERP, ATHEANA, CREAM, SPAR-H, ASEP, and SLIM. Each of these methods aims to provide a more accurate and objective human error probability (HEP) quantification. Of the methods listed above, ASEP, THERP, and SPAR-H are all inextricably linked. The Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) method was generated to better generalize the Technique for Human Error Rate Prediction (THERP) and Accident Sequence Precursor (ASEP). THERP is mapped to specific scenario templates, and ASEP is a simplified screening method, while SPAR-H employs generalizations to describe the spectrum of human behavior while maintaining the same theoretical underpinnings. SPAR-H is a well-known and accepted HRA method that uses performance shaping factors (PSFs) to classify inputs to the human component. PSFs capture a wide variety of input data, including plant status, crew dynamics, task description, and psychological aspects of the human operator. Current efforts have been focused to further generalize SPAR-H in to a dynamic framework. These methods can fit into dynamic NPP simulations such as RAVEN / RELAP-7. In order to do this, a dynamic methodology is applied to model the human as would be applied to the components. This method is applied in part one in the accompanying paper.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Fuel Flow Modeling under Thermal Fluidics Considerations within Operational Domains of Molten Salt Reactors

Molten Salt Reactors (MSRs) offer significant versatility supported by their design flexibility and safety advantages due to high temperatures, design customization capabilities, strong negative reactivity feedback effects, and salt chemistry characteristics. Different control schedules and different mass flow rates in the MSR primary systems impact neutronics and thermal hydraulics coupling. The result is two-fold. First, it leads to adaptability in various applications when the customization of a MSR unit is desired. Second, it calls for the quantification of dynamics and safety characteristics to assess the operational domain and make sure reactor stability and inherent safety characteristics are maintained. This paper provides an in-depth transient analysis of the operational implications of MSRs having heat-generating fuel and heat-transporting salt mixed forming liquid fuel salts contained within primary systems. A coupling model was developed to simulate fuel flow in molten salt reactors. A zero-dimensional reactor kinetics model was used with a one-dimensional heat transfer model in the reactor core. The temperature reactivity feedbacks of the fuel salt resulting from the thermal-hydraulics model were used in the reactor kinetics model to complete the coupled code. The model was applied to simulate the Molten Salt Breeder Reactor (MSBR) response at steady state and transients. The developed model was shown to be a suitable tool for the dynamic analysis of molten salt reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Feature selection based on PCA and the transient identification in nuclear power plants - 289

For the sake of enhancing safety and achieving more economic benefits, it is very important to timely and correctly identify the transients during the operation of nuclear power plants. There are thousands of monitoring signals in a nuclear power plant. It is difficult to complete the calculations timely if all the related signals are chosen for an algorithm of transient identification. Moreover, the irrelevant state variables may also do harm to the correctness of the identification process. Thus, one of the key tasks for transient identification is to figure out effective methods to extract the features from state variables. In this paper, we will address an efficient method for feature selection and transient identification based on principal component analysis (PCA). The effectiveness of the method are illustrated by the tests based on the data from the simulator designed for the High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM), which is under construction in China and the construction will be completed around 2017. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Regulatory Framework for Permanent Shut-down and Decommissioning of Nuclear Power Plant in Korea - 18101

Kori Unit 1, the first commercial NPP in Korea which was commissioned in 1978, was put on permanent shutdown for decommissioning on June 2017, after 40 years of operation. For permanent shutdown of the Kori Unit 1, an operation change permit was submitted in 2016 prior to the expiration since the NSA (Nuclear Safety Act) recognizes permanent shutdown as one of the operation stage. In preparation for the decommissioning of Kori Unit 1, the NSA related to decommissioning was revised and promulgated in 2015. The lower statues including Enforcement Decree and Regulation of the NSA, etc. were promptly revised to reflect decommissioning related requirements and procedures accordingly. For new NPPs, preliminary decommissioning plan is submitted and is being reviewed in the course of license review. Operating NPPs are also planning to submit the decommissioning plan by 2018 and receive approvals from Nuclear Safety and Security Commission. In case that an operator intends to decommission a nuclear installation, the operator shall apply for decommissioning and receive an approval from the NSSC. As discussed, framework for permanent shutdown and decommissioning of nuclear facilities have been well formatted by successful revision of NSA in Korea. Further detailed technical standards and regulatory guides on decommissioning will be developed to facilitate the relevant activities and ensure the nuclear safety. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Feasibility Assessment of Using Electrical Impedance Tomography for Damage Localization in Graphite Microreactor Components

Microreactors have great potential to decrease capital costs and construction timelines, reducing the barriers to implementing advanced nuclear reactor technologies. However, the lower power output of these microreactors introduces economic challenges if the operation and maintenance costs cannot be reduced sufficiently. Many microreactor concepts use graphite materials for in-core moderator and structural purposes, which will require periodic inspection or, ideally, in situ structural health monitoring. This work describes an initial evaluation of the feasibility of leveraging graphite’s semiconducting properties to perform electrical impedance tomography (EIT) for defect localization. First, a study was performed to identify the best methods for bonding electrical contacts to minimize the effects of contact resistance that interfere with impedance measurements of the graphite. After determining the best bonding approach, electrodes were attached to multiple graphite components with varying geometries (e.g., simple block and more representative microreactor hexagonal block). In parallel, finite element analysis approaches were developed and implemented to analyze the impact of defects and contact resistances and to inform an iterative inverse model for reconstructing the conductivity distribution. The results of this study show that defect localization in graphite components is possible using EIT if a sufficient number of electrodes (to improve spatial resolution) can be bonded using a technique with very low contact resistance and those contact resistances remain stable during reactor operation. If the reactor core and vessel design do not allow electrodes to be bonded during reactor operation, then it may be possible to detach/reattach electrodes between reactor operation and outages if a suitable mechanical connection that offers repeatable contact resistance can be identified. Alternatively, future work could focus on identifying contacts that can survive the harsh reactor operating environment so that contacts do not need to be removed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗