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DOE OSTI2022

Sensitivity analyses of a homogeneous model and a RZ model of the MYRRHA reactor in its critical configuration

Two homogenized models of the MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) reactor in its critical configuration are presented in this study. Both models will be verified by comparing the main reactor parameters (K{sub eff}, β{sub eff} and Λ) with those of the heterogeneous model and a sensitivity analysis will be performed. A sensitivity analysis of some important parameters related to reactivity, namely the vacuum coefficient, the Doppler and the power peaking factors, is also presented. The study of the vacuum coefficient has been carried out by analyzing two possible scenarios: 50% vacuum in the coolant (lead bismuth eutectic) and 100%, obtaining a list of the 10 reactions and nuclides that most affect the k{sub eff} value. With respect to the Doppler coefficient, it will be seen that a 300 degree increase in fuel temperature (starting at 800 K) results in a reduction of the total reactivity of the system by 134 pcm.. Finally, the map of the power peaking factors will be shown, as well as a sensitivity analysis. For making the sensitivity calculations Serpent-2 reactor physics Monte Carlo code has been used and JEFF-3.3 library has been chosen in order to continue with the OECD/NEA WPEC SG46 work evaluating the nuclear data of this library. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Evaluation of Oak Ridge National Laboratory Health Physics Research Reactor Operation Data for Critical Benchmark Creation

The Oak Ridge National Laboratory (ORNL) Health Physics Research Reactor (HPRR) was a research reactor designed and built at ORNL in 1961. The critical assembly used a highly enriched uranium and molybdenum alloy as the fuel and could be operated in steady-state or burst modes. The HPRR has recently been the object of an investigation to create a criticality benchmark. Such benchmarks are very important, as they are used primarily to show the accuracy of newly developed modeling codes and to help experimental validation and reactor licensing. The evaluated experiments considered in this paper were carried out between 1974 and 1986 from various HPRR activities such as steady-state subcritical, steady-state critical, and burst prompt super-critical operations of the reactor for dosimetry, irradiation, or training purposes. By using the HPRR experimental logbook information and the as-built drawings of the critical assembly, a highly detailed model of the HPRR was created with SCALE 6.2.4/KENO-VI, and a first version of a critical benchmark of the HPRR was developed following the International Criticality Safety Benchmark Evaluation Project (ICSBEP) guidelines for thorough description and uncertainty/sensitivity quantification. Unfortunately, in most of the evaluated experiments, the obtained difference between calculated and experimental k eff is around 1,000 pcm, corresponding to a relative error of approximately 1%, beyond the quality standards of the ICSBEP recommending a relative error below 0.1%. Moreover, the derived experimental uncertainty is high, around 4% relative, mainly due to the U-Mo fuel density uncertainty, but also from numerous other factors. For these reasons, the creation of a valuable critical benchmark from HPRR operation data is thus far compromised. In this paper, the different steps of the experiments’ evaluation are summarized, and the reasons for the experimental/calculation discrepancies and potential ways to solve them are explored. This paper also aims to remind us always to exercise considerable care when performing experimental work, and to record all the data possible for potential future uses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Consistent Multigroup Theory Enabling Accurate Course-Group Simulation of Gen-IV Reactors

The objective of this proposal is the development of a consistent multi-group theory that accurately accounts for the energy-angle coupling associated with collapsed-group cross sections. This will allow for coarse-group transport and diffusion theory calculations that exhibit continuous energy accuracy and implicitly treat cross- section resonances. This is of particular importance when considering the highly heterogeneous and optically thin reactor designs within the Next Generation Nuclear Plant (NGNP) framework. In such reactors, ignoring the influence of anisotropy in the angular flux on the collapsed cross section, especially at the interface between core and reflector near which control rods are located, results in inaccurate estimates of the rod worth, a serious safety concern. The scope of this project will include the development and verification of a new multi-group theory enabling high-fidelity transport and diffusion calculations in coarse groups, as well as a methodology for the implementation of this method in existing codes. This will allow for a higher accuracy solution of reactor problems while using fewer groups and will reduce the computational expense. The proposed research represents a fundamental advancement in the understanding and improvement of multi- group theory for reactor analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Light Water Reactor Sustainability Program: Grizzly Year-End Progress Report

The Grizzly software application is being developed under the Light Water Reactor Sustainability (LWRS) program to address aging and material degradation issues that could potentially become an obstacle to life extension of nuclear power plants beyond 60 years of operation. Grizzly is based on INL’s MOOSE multiphysics simulation environment, and can simultaneously solve a variety of tightly coupled physics equations, and is thus a very powerful and flexible tool with a wide range of potential applications. Grizzly, the development of which was begun during fiscal year (FY) 2012, is intended to address degradation in a variety of critical structures. The reactor pressure vessel (RPV) was chosen for an initial application of this software. Because it fulfills the critical roles of housing the reactor core and providing a barrier to the release of coolant, the RPV is clearly one of the most safety-critical components of a nuclear power plant. In addition, because of its cost, size and location in the plant, replacement of this component would be prohibitively expensive, so failure of the RPV to meet acceptance criteria would likely result in the shutting down of a nuclear power plant. The current practice used to perform engineering evaluations of the susceptibility of RPVs to fracture is to use the ASME Master Fracture Toughness Curve (ASME Code Case N-631 Section III). This is used in conjunction with empirically based models that describe the evolution of this curve due to embrittlement in terms of a transition temperature shift. These models are based on an extensive database of surveillance coupons that have been irradiated in operating nuclear power plants, but this data is limited to the lifetime of the current reactor fleet. This is an important limitation when considering life extension beyond 60 years. The currently available data cannot be extrapolated with confidence further out in time because there is a potential for additional damage mechanisms (i.e. late blooming phases) to become active later in life beyond the current operational experience. To develop a tool that can eventually serve a role in decision-making, it is clear that research and development must be perfomed at multiple scales. At the engineering scale, a multiphysics analysis code that can capture the thermomechanical response of the RPV under accident conditions, including detailed fracture mechanics evaluations of flaws with arbitrary geometry and orientation, is needed to assess whether the fracture toughness, as defined by the master curve, including the effects of embrittlement, is exceeded. At the atomistic scale, the fundamental mechanisms of degradation need to be understood, including the effects of that degradation on the relevant material properties. In addition, there is a need to better understand the mechanisms leading to the transition from ductile to brittle fracture through improved continuum mechanics modeling at the fracture coupon scale. Work is currently being conducted at all of these levels with the goal of creating a usable engineering tool informed by lower length-scale modeling. This report summarizes progress made in these efforts during FY 2013.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Remote Reactor Monitoring

The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

End-User Evaluation of SHARP for Analysis of Sodium-Cooled Fast Reactors

The DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is developing the SHARP toolkit to support growing reactor analysis needs in other programs. Although SHARP includes physics modules—especially the computational fluid dynamics code Nek5000 and the structural mechanics tools in Diablo—that have an existing user-base in other fields, most of the analysis and validation efforts for nuclear reactor applications have been performed by members of SHARP’s own development team. The SHARP development team includes engineers and computational scientists from a diverse set of backgrounds, including nuclear engineering. However, before the SHARP tools can be adopted by the Advanced Reactor Technology (ART) program to support the analysis of sodium-cooled fast reactors (SFRs), it would be beneficial for an outside set of analysts and engineers to begin using SHARP and providing feedback to both programs. The authors of this report were charged with learning to use SHARP, applying SHARP to SFR analyses of interest to ART, evaluating the tools, and providing a set of recommendations and requirements to guide future development efforts by the NEAMS program. The authors are engineers (nuclear, mechanical, and civil) by training; all have experience in modeling and simulation for nuclear energy applications; all have either supported ART directly or report to a supervisor who does. This report is a formal summary of the findings of this team of engineers—the potential end-users of SHARP for ART activities. In an informal way, this effort also opened up communication lines between the SHARP developers and ART engineers, which will continue beyond the scope of this effort.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Advanced Small Modular Reactor Economics Status Report

This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic and nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial operation. Advanced fuel materials and fabrication costs have large uncertainties based on complexities of operation, such as contact-handled fuel fabrication versus remote handling, or commodity availability. Thus, this analytical work makes a good faith effort to quantify uncertainties and provide qualifiers, caveats, and explanations for the sources of these uncertainties. The overall result is that this work assembles the necessary information and establishes the foundation for future analyses using more precise data as nuclear technology advances.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Advanced Fast Reactor - 100 (AFR-100) Report for the Technical Review Panel

This report is written to provide an overview of the Advanced Fast Reactor-100 in the requested format for a DOE technical review panel. This report was prepared with information that is responsive to the DOE Request for Information, DE-SOL-0003674 Advanced Reactor Concepts, dated February 27, 2012 from DOE’s Office of Nuclear Energy, Office of Nuclear Reactor Technologies. The document consists of two main sections. The first section is a summary of the AFR-100 design including the innovations that are incorporated into the design. The second section contains a series of tables that respond to the various questions requested of the reactor design team from the subject DOE RFI.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

FASTER Test Reactor Preconceptual Design Report

The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization (M 3 ).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Tritium Mitigation/Control for Advanced Reactor System

A tritium removal facility, which is similar to the design used for tritium recovery in fusion reactors, is proposed in this study for fluoride-salt-cooled high-temperature reactors (FHRs) to result in a two-loop FHR design with the elimination of an intermediate loop. Using this approach, an economic benefit can potentially be obtained by removing the intermediate loop, while the safety concern of tritium release can be mitigated. In addition, an intermediate heat exchanger (IHX) that can yield a similar tritium permeation rate to the production rate of 1.9 Ci/day in a 1,000 MWe PWR needs to be designed to prevent the residual tritium that is not captured in the tritium removal system from escaping into the power cycle and ultimately the environment. The main focus of this study is to aid the mitigation of tritium permeation issue from the FHR primary side to significantly reduce the concentration of tritium in the secondary side and the process heat application side (if applicable). The goal of the research is to propose a baseline FHR system without the intermediate loop. The specific objectives to accomplish the goals are: To estimate tritium permeation behavior in FHRs; To design a tritium removal system for FHRs; To meet the same tritium permeation level in FHRs as the tritium production rate of 1.9 Ci/day in 1,000 MWe PWRs; To demonstrate economic benefits of the proposed FHR system via comparing with the three-loop FHR system. The objectives were accomplished by designing tritium removal facilities, developing a tritium analysis code, and conducting an economic analysis. In the fusion reactor community, tritium extraction has been widely investigated and researched. Borrowing the experiences from the fusion reactor community, a tritium control and mitigation system was proposed. Based on mass transport theories, a tritium analysis code was developed, and the tritium behaviors were analyzed using the developed code. Tritium removal facilities were designed and laboratory-scale experiments were proposed for the validation of the proposed tritium removal facilities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Dynamic System Models for Informing Licensing and Safeguards Investigations of Molten Salt Reactors

Fluid-fueled molten salt nuclear reactors (MSRs) have recently gained significant interest. As with all reactors, modeling and simulation are key factors for advanced reactor design and licensing and will be required for the deployment of MSRs. However, there are significant gaps between simulation capabilities and system behavior for MSRs. This paper presents a system model of an MSR based on the Molten Salt Demonstration Reactor. The model includes important physics specific to MSRs, such as fission product and tritium transport, as well as reactivity feedback.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Assembly and Delivery of Rabbit Capsules for Irradiation of Reinforced Radiation Resistant SiC-SiC Composites in the High Flux Isotope Reactor

The development of reinforced radiation resistant composites presents a critical challenge for in-vessel component application in advanced high temperature reactors. In its pursuit of this, Physical Optics Corporation (POC) is interested in developing a silicon carbide (SiC)-based ceramic matrix composite (CMC) with improved thermal conductivity in a high neutron radiation environment. The US Department of Energy (DOE) is supporting research efforts to develop Advanced Reactor Technology. Ultimately, the results of this project will determine the viability of using SiC matrix composites in hot structures in advanced high temperature reactors. The first objective of this project is to irradiate torsion and bend bar specimens in the High Flux Isotope Reactor (HFIR) and develop an initial data set to evaluate the evolution of composite mechanical strength, dimensional changes, and thermal conductivity. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of three rabbit capsules to the HFIR. Rabbits of both torsion and bend bar configurations have been assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Alternate Design Concepts for the High Flux Isotope Reactor using Low-Enriched Uranium Fuel Systems: Scoping Analysis

The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is a multi-mission research reactor dedicated mainly to isotope production, neutron scattering experiments, material irradiation, testing and activation. In the framework of non-proliferation policies, the international community is currently aiming to reduce or even eliminate, when and where possible, the use of Highly Enriched Uranium (HEU: 235 U/U ≥ 20 wt. %) in civilian applications. In that context, most research reactors worldwide still operating with HEU fuel are actively engaged in an effort to convert to Low Enriched Uranium fuel system (LEU: 235 U/U < 20 wt. %). Within the US, the High Performance Research Reactor (USHPRR) fleet (which includes HFIR) is expected to convert to LEU fuel using the UMo “monolithic” fuel system currently under development. ORNL is responsible for the development of a LEU design for HFIR. In the past years, Argonne National Laboratory (ANL) offered technical assistance to the ORNL LEU design team whenever deemed appropriate. Due to their complex nature, acceptable HFIR LEU designs identified so far, all based on the UMo monolithic fuel system, would require a substantial fabrication development effort before being proven commercially viable. In order to mitigate technical risks and reduce cost/schedule uncertainties, it appears appropriate to study alternate design solutions and alternate fuel systems as possible backup options for the HFIR conversion to LEU fuel. In that context, ANL has been exploring alternate design concepts for HFIR. Complementing the ORNL design activities, the goal is to find the best design solution for HFIR in term of cost, performance and safety. The present report summarizes the findings regarding two types of alternate conceptual design: U 3 Si 2 designs and un-contoured UMo design. Based on available tools and methods, preliminary analysis show that both concepts could probably meet performance and steady-state safety requirements but element fabrication would remain complex and would require a non-trivial fabrication development effort.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 OSTI2019

Integrated Approach to Fluoride High Temperature Reactor Technology and Licensing Challenges (FHR-IRP)

This document constitutes progress made on this project by the Georgia Tech led Integrated Approach to Fluoride High Temperature Reactor Technology and Licensing Challenges (FHR-IRP). This is the first draft of the project final report. Chapter 3 is focused on Tritium management. Tritium management in Fluoride-salt High-temperature Reactors (FHRs) presents a major operational concern and a significant challenge. Tritium is generated via multiple pathways including various interactions in fuel and structural materials, as well as coolant (primarily interaction with lithium and beryllium), leading to a complex evolution and spatial distribution of tritium generation rate. FHR generates orders of magnitude more tritium than in LWRs, and has potential for a significant release, via permeation through structural materials (due to high temperature), and through heat exchanger(s). Therefore, accurate estimate is necessary. For that purpose relevant tritium production pathways were identified. A methodology to obtain space-time dependent tritium generation rate was developed and implemented, employing SCALE and MCNP codes. Detailed models of the reference Advanced High Temperature Reactor (AHTR) design were developed. A multi-parametric metrics was introduced to capture relevant tritium production indicators, including the instantaneous tritium production rate, total tritium production (neglecting subsequent decay), and accumulated tritium production (accounting for the decay). The obtained results will inform tritium management, including the sizing of the components, and to gain insight on the impact of the core design parameters on tritium generation rates. Research effort on Liquid Salt Impurity Removal, Redox and Corrosion Control is discussed in Chapter 4. Section 4.1 is an introduction chapter. Based on the current available knowledge and operation experience on previous MSRs and other types of nuclear reactors, the report identifies the major impurities that are related to either material compatibility, salt properties or radioactive source terms, and the potential contamination sources of these impurities, their transport in the primary coolant loop and available cleanup methodologies for a particular group of impurity. In section 4.2, the available purification methods were discussed, including the liquid/liquid extraction using liquid Bi-Li which has been applied in MSBR and electrochemical purification which is developed based on Pyroprocessing. In this section, the new multi-stage and multi-salt-component was developed and discussed. Section 4.3 discussed the redox potential control for salt chemistry and corrosion mitigation, including the control theory and the soluble/soluble buffer control method. The section also developed method on how to analyze soluble/soluble redox reaction in a molten salt system and designed an electrochemical sensor for redox potential control. Corrosion models were also reported in this section. Section 4.4 focused on how to measure the fundamental properties in a molten salt. As an example, the section focused on La fluoride in the molten salt. The section developed methods on how analyze electrochemical measurement data to obtain the fundamental properties including the redox potential, diffusion coefficient and exchange current density. All these fundamental information is important for salt purification system development. Section 4.5 focused on measurements of corrosion products in the salts. Because of the different behaviors between Cr and La in the salt, new analysis method for Cr fluorides in the salt was developed. Some results for Fe fluoride in the salt and corrosion measurements were also reported. Section 4.6 demonstrated the separation of oxide from the salt. This section also measured the solubility of La 2 O 3 in the salt as function of temperature. Section 4.7 focused on the separation of La and Ce from the molten salt. These two elements were selected because La was found to be separated by using inert electrode, while for separation of Ce, reactive electrode must be selected. Section 4.8 demonstrated the separation iodine ions from the molten salt. Two kinds of working electrode materials (metal W and graphite) were investigated. Section 4.9 focused on the fundamental data of fission products and actinides in liquid Bi. The chapter reviewed the available data, then based a CALPHAD model, the available data were broadened to the concentration-dependent cases.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

PTT System Performance Evaluation in 3-D Imaging of Calibrated Defects: Pulsed Thermal Tomography Nondestructive Examination of Additively Manufactured Reactor Materials and Components

Research activity in this project is aimed at developing pulsed thermal tomography (PTT) architecture and algorithms for in-service non-destructive examination (NDE) of additively manufactured (AM) reactor components and materials. Recent advances in AM technology could allow cost-efficient fabrication of parts for operating reactors. However, because of stringent safety requirements, long-term performance of AM reactor components needs to be investigated before AM is widely accepted. Because of the complex shapes and significant surface roughness of components due to layer-by-layer welding process in AM, conventional methods might not be useful for in-service NDE of AM components. PTT is ideally suited for in-service NDE of AM metallic components because the method is non-contact, and offers high resolution #-D imaging of material flaws. The objective of this report is to provide initial evaluate of PTT performance sin detecting calibrated flaws in reactor structural materials. Preliminary COMSOL models were developed to conduct super computing simulations of PTT. In the experimental studies, high strength Stainless Steel 316 and Inconel 718 alloys were considered, as well as lower grade Stainless Steel 304, Nickel 200, and Hastelloy C276. Specimens investigated in this report consisted of approximately 1/4in-thick plates made out of these alloys using conventional manufacturing methods. The calibrated defects were created in the form of flat bottom holes (FBH) drilled in metallic plates. The diameters of FBH's varied from 1mm to 8mm, and their depths below the plate flat surface varied between 1mm and 6mm. The size of the smallest FBH was limited to 1mm because conventional mechanic drills were used for creating the holes. PTT imaging results have shown that 1mm-diameter FBH located 1mm and 2mm below the surface were detectable. Larger size FBH were detectable at greater depth. For example, 6mm-diameter FBH could be detected at 8mm depth. Image contrast varied slightly between the specimens, with the best reconstructions obtained in SS316 and C276 plates. In addition, a 2/3in-thick Inconel 718 nozzle plate produced with additive manufacturing method was imaged with PTT. It was shown that PTT can scan through the plate in approximately 20s. Several modes of 3-D data visualization were explored, including using ImageJ and MATLAB software packages.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

RAZORBACK - A Research Reactor Transient Analysis Code (Version 1.0 - Volume 2: Software Design and Development Report)

RAZORBACK is a research reactor transient analysis computer code designed to simulate the operation of a research reactor (such as Sandia National Laboratories' Annular Core Research Reactor (ACRR)). The code provides a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This Software Design and Development Report describes how the code is designed to implement the functional requirements established in the Software Requirements Specification document for RAZORBACK. The description presents the physical model equations used in the design, and how these equations are set up to be solved numerically. This report also provides an evaluation of how the software design meets each of the functional requirements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗