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At least 235 records · Page 13

Monolithic ATF MiniFuel Sample Capsules Ready for HFIR Insertion

Several advanced nuclear fuel concepts are being considered throughout the nuclear industry. Implementation of these concepts could increase accident tolerance or enhance performance beyond the capabilities of the UO₂-based nuclear fuel currently in use. Qualification and deployment of any new fuel requires rigorous irradiation testing to demonstrate performance under representative normal and off-normal operating conditions. The traditional approach for qualifying new fuels requires exhaustive execution of many integral fuel tests. However, due to the long timeframe for executing these integral tests and the limited number of available materials test reactors, this approach is becoming impractical. To accelerate fuel qualification, Oak Ridge National Laboratory developed the MiniFuel irradiation vehicle for use in conducting accelerated separate effects irradiation testing of a wide range of nuclear fuel materials in the High Flux Isotope Reactor (HFIR). The first MiniFuel irradiations performed in the facility tested sol gel–derived uranium nitride kernels and tristructural isotropic (TRISO)-coated particle fuels. This report describes the preparation and assembly of the first set of monolithic MiniFuel irradiations conducted to support accident-tolerant fuel (ATF) development and accelerated fuel qualification. Two irradiation targets containing a variety of UO₂ and U₃Si₂ disk fuel specimens were fabricated and assembled for irradiation to burnups of 8-10 and 28-40 MWd/kg U. The target irradiation temperature is 450-550°C. Irradiation of U₃Si₂ will provide new data regarding the irradiation performance of a candidate ATF to complement current ATF-1 integral experiments being performed in the Advanced Test Reactor. UO₂ samples were included as a reference so that the results from the MiniFuel experiments can be compared with the extensive UO₂ fuel performance database. The monolithic MiniFuel capsules were successfully assembled, welded, and tested per HFIR requirements and are ready for insertion into the reactor. Pictures of the assembly process are included in this report. The experiment is planned for insertion into the HFIR during cycle 487 in April 2020.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Concept Descriptions for the VTR Rabbit System and Driver Fuel Test Assemblies

Two of the experiment vehicles being developed for the Versatile Test Reactor (VTR) are presented here. The first is a rabbit system that will enable rapid insertion of small test capsules into the high fast flux of the VTR core for relatively short durations. The rabbit concept development includes the construction/demonstration of a near-full-scale system in a deep-water pool to demonstrate functionality, development of a concept of operations and initial procedures, and validation of thermal-hydraulic modeling. In addition, modeling efforts are underway to simulate the thermal and neutronic environment of a rabbit capsule. The second type of experiment vehicle presented here is a driver fuel test assembly for inserting fuel and materials tests into the core by replacing a driver fuel assembly. Here, a novel design for dismountable test assemblies is proposed for the VTR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Irradiation and PIE of alloys X-750 and XM-19 (EPRI Phase III)

The Nuclear Science User Facilities (NSUF) based at Idaho National Laboratory (INL), along with the Electric Power Research Institute (EPRI), formed an agreement to test representative alloys used as reactor structural materials as a pilot program to establish guidelines for future NSUF research programs. This report contains results from the portion of this program established as Phase III (of three phases), entailing irradiation and post-irradiation examination of select alloys typical of boiling water reactor (BWR) internal structural materials. Phases I and II are the subject of separate reports and represent baseline material test results and irradiation experiment design, respectively. The intent of this Phase III research program is to determine properties for the materials of interest after being irradiated at the Advanced Test Reactor (ATR) to three different target fast (E>1MeV) fluences: 5.0 x 10¹⁹ n/cm², 2.0 x 10²⁰ n/cm², and 1.0 x 10²¹ n/cm². These correspond to irradiation damage levels (displacements per atom [dpa]) of approximately 0.08, 0.30, and 1.4 dpa, which represent comparable levels to (a) a previous study which looked at X-750 irradiated to ~1 x 10¹⁹ n/cm², comparable to the lowest fluence; (b) approximately a medium level of fluence for BWR components; and (c) extended life (60 – 80 years) for BWR components. The materials chosen for this research are the nickel-based alloy X-750 and austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co., and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research Center (GE-GRC), and parts were provided to INL for use in this pilot project.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Irradiation and PIE of alloys X-750 and XM-19 (EPRI Phase III)

The Nuclear Science User Facilities (NSUF) based at Idaho National Laboratory (INL), along with the Electric Power Research Institute (EPRI), formed an agreement to test representative alloys used as reactor structural materials as a pilot program to establish guidelines for future NSUF research programs. This report contains results from the portion of this program established as Phase III (of three phases), entailing irradiation and post-irradiation examination of select alloys typical of boiling water reactor (BWR) internal structural materials. Phases I and II are the subject of separate reports and represent baseline material test results and irradiation experiment design, respectively. The intent of this Phase III research program is to determine properties for the materials of interest after being irradiated at the Advanced Test Reactor (ATR) to three different target fast (E>1MeV) fluences: 5.0 x 10¹⁹ n/cm², 2.0 x 10²⁰ n/cm², and 1.0 x 10²¹ n/cm². These correspond to irradiation damage levels (displacements per atom [dpa]) of approximately 0.08, 0.30, and 1.4 dpa, which represent comparable levels to (a) a previous study which looked at X-750 irradiated to ~1 x 10¹⁹ n/cm², comparable to the lowest fluence; (b) approximately a medium level of fluence for BWR components; and (c) extended life (60 – 80 years) for BWR components. The materials chosen for this research are the nickel-based alloy X-750 and austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co., and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research Center (GE-GRC), and parts were provided to INL for use in this pilot project.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Fusion materials irradiations at MaRIE's fission fusion facility

Los Alamos National Laboratory's proposed signature facility, MaRIE, will provide scientists and engineers with new capabilities for modeling, synthesizing, examining, and testing materials of the future that will enhance the USA's energy security and national security. In the area of fusion power, the development of new structural alloys with better tolerance to the harsh radiation environments expected in fusion reactors will lead to improved safety and lower operating costs. The Fission and Fusion Materials Facility (F{sup 3}), one of three pillars of the proposed MaRIE facility, will offer researchers unprecedented access to a neutron radiation environment so that the effects of radiation damage on materials can be measured in-situ, during irradiation. The calculated radiation damage conditions within the F{sup 3} match, in many respects, that of a fusion reactor first wall, making it well suited for testing fusion materials. Here we report in particular on two important characteristics of the radiation environment with relevancy to radiation damage: the primary knock-on atom spectrum and the impact of the pulse structure of the proton beam on temporal characteristics of the atomic displacement rate. With respect to both of these, analyses show the F{sup 3} has conditions that are consistent with those of a steady-state fusion reactor first wall.

43 PARTICLE ACCELERATORS↗

Evaluation of the NSUF Reactor Activation and Damage (RAD) Calculator damage component

The Nuclear Science User Facilities (NSUF) provide researchers access to nuclear research facilities and expertise all over the United States and at the Belgian Center for Nuclear Research (SCK/CEN). One special capability of the NSUF is providing access to nuclear research and test reactors often used for conducting material testing experiments. These services are provided at no cost to the researcher through a competitive, peer-reviewed process. With so many proposals being submitted simultaneously, an NSUF information management resource called the Combined Materials Experiment Toolkit (CoMET) was implemented to speed up the review process and give users access to the information and tools necessary for designing meaningful experiments and writing the best proposals possible. One of CoMET’s tools is the Reactor Activation and Damage (RAD) Calculator. For those with little or no background in it, neutron irradiation in research and test reactors can be a challenging subject to include in a proposal. That is why the RAD Calculator was created to assist users with providing estimates of irradiation and post-irradiation conditions, enabling them to select the best neutron irradiation and post-irradiation examination facilities for reaching their experiment goals. The RAD Calculator is a general scoping tool that estimates radiation damage in terms of DPA, in addition to the resulting radioactivity of the material post-irradiation. This report evaluates the neutron damage component of the RAD Calculator. The evaluation of the activation component of the calculator is covered in report INL/EXT-20-58080.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

A Variable Load LVDT-Based Creep Test Rig for Use in ATR Loop 2A

Understanding the creep behavior of materials in a radiation environment is essential in evaluating safety issues associated with nuclear power plant operation, especially for plants applying for life extension. In US materials testing reactors (MTRs), this understanding has been achieved through a “cook and look” approach (where specimens are irradiated under load for some period of time and then removed from an MTR for post-irradiation examination). A number of “cook and look” cycles can be required for each evaluated material, which makes this approach expensive. In addition, examinations are limited to end-state conditions, which may or may not be sufficient to reveal all phenomena of interest (because examinations are not performed at prototypic conditions relative to temperature, pressure, stress, etc.).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Quantitative multi-modal NDT data analysis

A single NDT technique is often not adequate to provide assessments about the integrity of test objects with the required coverage or accuracy. In such situations, it is often resorted to multi-modal testing, where complementary and overlapping information from different NDT techniques are combined for a more comprehensive evaluation. Multi-modal material and defect characterization is an interesting task which involves several diverse fields of research, including signal and image processing, statistics and data mining. The fusion of different modalities may improve quantitative nondestructive evaluation by effectively exploiting the augmented set of multi-sensor information about the material. It is the redundant information in particular, whose quantification is expected to lead to increased reliability and robustness of the inspection results. There are different systematic approaches to data fusion, each with its specific advantages and drawbacks. In our contribution, these will be discussed in the context of nondestructive materials testing. A practical study adopting a high-level scheme for the fusion of Eddy Current, GMR and Thermography measurements on a reference metallic specimen with built-in grooves will be presented. Results show that fusion is able to outperform the best single sensor regarding detection specificity, while retaining the same level of sensitivity.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Materials for Advanced Ultra-Supercritical (AUSC) Steam Turbines – AUSC Component Demonstration

The work by the United States Department of Energy (U.S. DOE)/Ohio Coal Development Office (OCDO) Advanced Ultra-Supercritical (AUSC) Steam Boiler and Turbine Materials Consortia from 2001 through September 2015 was primarily focused on lab scale and pilot scale materials testing. The field testing included air-cooled probes and steam-cooled “loops” that were installed into existing utility boilers to gain exposure of these materials to realistic conditions of high temperature and corrosion due to the constituents in the coal ash. Successful research and development resulted in identification of metallic alloy materials and fabrication processes suited for power generation applications with metal temperatures up to approximately 1472°F (800°C). These materials or alloys have shown, in extensive laboratory tests and shop fabrication studies, to have excellent applicability for high-efficiency low CO2 transformational power generation technologies. However, as valuable as these material field tests have been for obtaining information, their scale is significantly below that required to minimize the risk associated with a power company building a multi-billion-dollar AUSC power plant. To decrease the identified risk barriers to full-scale implementation of these advanced materials, the U.S. DOE/OCDO AUSC Steam Boiler and Turbine Materials Consortia identified the key areas of the technology that need to be tested at a larger scale. Based upon the recommendations and outcome of a Consortia-sponsored workshop with the U.S.’s leading utilities, a Component Test (ComTest) Program for AUSC was proposed. The AUSC ComTest program would define materials performance requirements, plan for overall advanced system integration, design critical component tests, fabricate components for testing from advanced materials, and carry out the tests. A Pre-FEED evaluation of the test facility including the evaluation of three potential sites has been completed and previously reported. This report describes the Front-End Engineering Design (FEED), Detailed Engineering and the Environmental Assessment of the ComTest project. A significant amount of work was accomplished during Phase 1 to identity the components, as well as the component size, that would be manufactured from advanced alloys such as INCONEL 740H or HAYNES 282 alloys. Pathways to supply these components for ComTest have been identified as well as any further development that would be required. Every project of this nature has challenges and this project was no exception. The greatest challenge has been the project cost. Largely due to the infrastructure and balance of plant equipment upgrades required for Youngstown Thermal, the estimated project cost for the “base case” project was greater than originally budgeted. The project team spent a significant amount of time exploring opportunities for reducing the project cost. These options included: • Youngstown Thermal was exploring building a Combined Heat and Power (CHP) unit at their facility. The Youngstown Thermal staff, supported by Thermal Engineering Group, worked with the project team to determine whether it was economically desirable to supply ComTest with steam and other utilities, which would be supplied by the CHP project. • The project team also evaluated opportunities to optimize the project scope to reduce cost, such as reducing project scope by eliminating the Superheater Outlet Header test. • Southern Company had offered two sites as possible host locations for ComTest. Both sites had infrastructure in place such as steam from an existing unit, natural gas, cooling water and ample electrical power that would be required to be added at the Youngstown Thermal site. Utilizing this existing infrastructure could reduce the project cost. An additional project challenge was the departure of Babcock and Wilcox from the project in September 2016. This project setback was overcome by GE Power’s willingness to assume the design of the ComTest superheater. Since the project could not proceed without the superheater, this was a very positive project development. The merger of the GE Power and Alstom Power steam turbine divisions created synergies leading to the conclusion that instead of building a small-scale turbine, a more practical and useful endeavor would be to validate the manufacturing technology and supplier readiness for fabricating and machining full scale steam turbine components. Consequently, in March 2017, GE Power decided that a steam turbine test would not be necessary as part of the ComTest demonstration. Removing the operational testing of the steam turbine from ComTest further reduced the total project cost. Additionally, there was an opportunity to move ComTest from Youngstown Thermal to one of the Southern Company sites and operate ComTest at supercritical steam conditions and test the steam turbine main stop/control valve at a temperature and pressure that would be representative of the conditions of a high pressure steam turbine inlet valve. The steam could also possibly be returned to the host boiler so there would be no economic penalty of using this steam for ComTest. Relocating the project could further reduce the project cost due to the infrastructure already in place, including high pressure natural gas and cooling water, and the potential for heat integration, allowing a reduction of operating costs. Due to these potential cost reductions, it was decided to move the project away from Youngstown Thermal in March 2017. As integrating ComTest at the Southern Company site was explored, an issue was raised regarding the integration of the ComTest cycle with the plant steam cycle. Steam would be supplied to ComTest at nominally ~3865 psia and 775°F. The steam would be heated with natural gas to 1400°F for the valve test and then cooled to 1000°F so that the steam would be returned to the host boiler at the same conditions as their high pressure steam turbine inlet. However, there were concerns raised that the addition of the energy (natural gas) to the steam could trigger a New Source Review on the unit. Southern Company did not want ComTest to cause this review, so condensing all of the steam exiting ComTest and returning it to the host steam cycle as condensate was explored to insure there would be no implication that ComTest was adding energy to the host plant’s steam cycle. Ultimately, Southern Company decided that they wanted to “decouple” ComTest from the operating unit, making ComTest self-supporting, similar to the Youngstown Thermal “base case.” Without the steam turbine, the project team reached a consensus that operating the ComTest Superheater and steam turbine valves at the typical pressure of a package boiler (~1000 psig) would not generate enough useful technical information to justify the cost of the test. Therefore, it was decided in the December 2017 project quarterly meeting to cancel the operational phase of ComTest. Throughout the engineering phase of ComTest, manufacturing and supply chain challenges were identified. When GE Power recommended dropping the steam turbine test from ComTest, they identified many important manufacturing issues that should be evaluated and demonstrated. Other challenges had been raised by the GE Power group designing the superheater as well as those engineering components such as pressure relief and steam turbine bypass valves. In the December 2017 meeting, the project team decided that it was more beneficial to the AUSC technology to pursue these manufacturing and supply chain issues as Phase 2 of the ComTest project. This report is organized in the following manner: • Section 2 discusses the ComTest project background and organization. • Section 3 discussed project management. • Section 4 discusses the design of the ComTest AUSC components – the Superheater, steam turbine and Superheater Outlet Header. • Section 5 discusses the implementation of ComTest at Youngstown Thermal and the Southern Company site. • Section 6 discusses the development of the AUSC Supply Chain • Section 7 discusses the recommended plan for Phase 2 of ComTest

01 COAL, LIGNITE, AND PEAT↗

Wire Arc Additive Manufacturing of Multi-material Parts Using H-13 Tool Steel

H-13 tools steel’s relative hardness, resistance to thermal fatigue, and high tolerance of thermal shock, make it very desirable for use in forging, pressing, casting, and extrusion processes[1]. H-13 is however typically quite expensive when compared to many other steel compositions. It is therefore desirable to encase a lower cost steel composition with H-13. This configuration allows for the benefits of H-13 tooling to be realized, at a considerably reduced cost. To demonstrate the viability of this concept, a casting tool was fabricated using a Wire Arc Additive Manufacturing (WAAM) process. Initial materials testing was conducted on a multi-material wall which consisted of one section of H-13 tool steel, and one section of 410NiNmo. These tests included scanning electron microscopy, energy dispersive X-ray spectroscopy, and hardness testing. This testing generated favorable results and subsequently a casting tool was fabricated for evaluation. Utilization of a Multi-Material WAAM process to fabricate large tooling in this manner could yield significant improvements in material cost, manufacturing agility, and supply chain complexity.

Lambert, Nathan [ORNL] (ORCID:0009000324134206)↗

Application of NSUF Capabilities Towards Understanding the Emulation of High Dose Neutron Irradiations with Ion Beams

Development of new materials for current nuclear power plants (NPPs) or materials for new NPPs requires extensive research, development, and deployment programs. Central to these efforts, particularly for core internals, are studies on the microstructural and mechanical changes due to neutron irradiation. These radiation effects studies are typically completed using materials test reactors followed by expensive and time-consuming post-irradiation examinations (PIE). This sluggish approach has severely limited the nuclear materials field to develop new materials or verify materials for new reactor design use. It is also very difficult to achieve the high levels of damage proposed for future reactors. A logical approach to accelerating the field is the application of ion beams to simulate neutron damage, especially for high dose applications. Ion beams have been used for decades to study radiation effects but not to emulate high dose reactor irradiations. In this regard, new (or renewed) methods for ion beam irradiations, new developments in materials characterization, and greater understanding on the governing mechanisms controlling observed radiation effects have led to a need for further studies to refine the technique. A primary challenge is the ability to determine the ion irradiation conditions that produce microstructures and mechanical properties relevant to in-core neutron irradiation conditions. A large, multi-disciplinary, multi-laboratory program deemed the High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation has been formed to assess and overcome the challenges associated with emulating high dose neutron irradiations using ion beams. Central to this program is the use of dual-beam ion irradiations and corresponding neutron irradiations within the BOR-60 fast reactor. These irradiations have produced hundreds of specimens for microstructural and mechanical characterization. A significant issue regarding the samples is the radioactivity of the neutron irradiated specimens as well as the need to characterize both irradiation types using the same equipment to maintain consistency and repeatability. The use of the Nuclear Science User Facilities (NSUF) has been central towards overcoming the mentioned issues and thereby enabling robust PIE in both the ion and neutron irradiated specimens of interest. In particular, the electron microscopes housed within the Low Activation Materials Development and Analysis (LAMDA) facility at Oak Ridge National Laboratory (ORNL) - an NSUF partner facility - has been instrumental in gathering high quality data for analysis. This paper will briefly highlight the PIE efforts using the NSUF capabilities within LAMDA.

36 MATERIALS SCIENCE↗

Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

The proposed work addresses the key research need for the development of constitutive models and overall failure models for graphite and high temperature structural materials, with the long-term goal being to maximize the design life of the Next Generation Nuclear Plant (NGNP). To this end, the capability of a Continuum Damage Mechanics (CDM) model, which has been used successfully for modeling fracture of virgin graphite, will be extended as a predictive and design tool for the core components of the very high- temperature reactor (VHTR). Specifically, irradiation and environmental effects pertinent to the VHTR will be incorporated into the model to allow fracture of graphite and ceramic components under in-reactor conditions to be modeled explicitly using the finite element method. The model uses a combined stress-based and fracture mechanics-based failure criterion, so it can simulate both the initiation and propagation of cracks. Modern imaging techniques, such as x-ray computed tomography and digital image correlation, will be used during material testing to help define the baseline material damage parameters. Monte Carlo analysis will be performed to address inherent variations in material properties, the aim being to reduce the arbitrariness and uncertainties associated with the current statistical approach. The results can potentially contribute to the current development of American Society of Mechanical Engineers (ASME) codes for the design and construction of VHTR core components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

High-Temperature Active Magnetic Bearing Development for Supercritical CO 2 Machinery Applications

Hermetic machinery utilizing gas bearings for MW-scale supercritical CO 2 (sCO 2 ) machinery applications can have significantly lower power loss and enable improved cycle efficiency compared to conventional machinery with oil-lubricated bearings. Active magnetic bearings (AMBs) are another option anticipated to have similar power loss and load capacity to gas bearings as well as offering larger mechanical tolerances, the ability to tune properties, and high reliability due to lack of mechanical wear. AMBs also have proven commercial experience at MW-scale, though environments for high-temperature sCO 2 power cycle machinery conditions are novel. Besides the potential impact of AMBs for sCO 2 turbomachinery, the technology also offers promising benefits for steam and gas turbines for power generation, compressors and expanders for industrial heat and power, and in other oil and gas and space applications. The goals of this project were to conceptual design an AMB and perform material testing. Conceptual designs for radial and thrust AMBs were produced based on a hermetically-sealed sCO 2 machinery waste-heat recovery (WHR) application for sizing and loads, and choosing a target design temperature of 540°C useful for high-temperature sCO 2 turbines for concentrating solar power (CSP) applications. Conceptual designs were initially developed for multiple radial and thrust AMBs with spreadsheet-based calculations before selecting one of each to develop further using higher-fidelity design methods for magnetic and structural performance. It was found that the radial AMB at 540°C was feasible, but the thrust AMB needed to be limited to 315°C for high-speed operation. The decision for a reduced-temperature thrust AMB was the result of several significant conclusions: 1) Hiperco 50A, originally chosen for good magnetic performance at high temperature, had insufficient strength for high-speed operation, so it was replaced with 17-4 PH. 2) The reduced magnetic performance from 17-4 PH yielded a larger bearing size, reducing the strength margin. 3) This ultimately led to a creative design implementing a more-compact E-core topology, compared to the original (conventional) C-core, and an integral shaft-disk with Hirth joint connection. These conceptual designs are unique for the size and temperature in CO 2 , relevant for MW-scale CSP applications. Long term, high temperature test data was generated for several materials, filling a void in the current body of literature. Corrosion and magnetic performance measurements were produced for PM materials (Alnico 5-7C, Alnico 9C, and SmCo) with and without nickel-coating for environments of high-temperature CO 2 up to 550°C at atmospheric pressure and 450°C at 103 bar for up to 6,000 hours. Corrosion measurements were also produced for Hiperco 50, a SM material relevant for AMB laminations, with and without C5 coating. Comparisons were also made for 450°C and 550°C, atmospheric pressure air exposures up to 5,000 hours. Results generally show that coatings can be effective at improving oxidation resistance of the bare materials, and Alnicos generally outperformed SmCo after high-temperature exposure. In addition to technical feasibility demonstrated by the design, economic feasibility was demonstrated by updating the TEA from the reference machine, re-evaluating it with CAPEX and OPEX to reflect estimated changes from process-lubricated bearings to AMBs. AMBs were shown to be comparable in performance to process-lubricated bearings, still showing a notable improvement over conventional machinery architecture with oil-lubricated bearings.

42 ENGINEERING↗

Improvement in precision, accuracy, and efficiency in sstandardizing the characterization of granular materials

Useful prediction of the kinematics, dynamics, and chemistry of a system relies on precision and accuracy in the quantification of component properties, operating mechanisms, and collected data. In an attempt to emphasize, rather than gloss over, the benefit of proper characterization to fundamental investigations of multiphase systems incorporating solid particles, a set of procedures were developed and implemented for the purpose of providing a revised methodology having the desirable attributes of reduced uncertainty, expanded relevance and detail, and higher throughput. Better, faster, cheaper characterization of multiphase systems result. Methodologies are presented to characterize particle size, shape, size distribution, density (particle, skeletal and bulk), minimum fluidization velocity, void fraction, particle porosity, and assignment within the Geldart Classification. A novel form of the Ergun equation was used to determine the bulk void fractions and particle density. Accuracy of properties-characterization methodology was validated on materials of known properties prior to testing materials of unknown properties. Several of the standard present-day techniques were scrutinized and improved upon where appropriate. Validity, accuracy, and repeatability were assessed for the procedures presented and deemed higher than present-day techniques. A database of over seventy materials has been developed to assist in model validation efforts and future desig

42 ENGINEERING↗

A simplified approach for creep damage and remaining life calculation for materials surveillance in advanced reactors

High temperature reactor materials will experience a combination of mechanical degradation caused by creep and fatigue and environmental degradation caused by neutron irradiation and, in some cases, exposure to corrosive coolants. A materials surveillance program is one option for ensuring the safe, reliable operation of key reactor components under these conditions. Such a program would monitor the degradation in key material properties over time, using this data to predict how changes in the material properties affect components performance. This could then be used to make plant operational decisions and support eventual plant license extensions. Past work de-scribes such a material surveillance program based on passively actuated mechanical test articles that impose creep-fatigue type loading on test materials driven only by changes in temperature, for example those experienced by components under standard operating cycles. This report focuses on two aspects of the proposed material surveillance program: how to design the test articles to mimic the mechanical response of the corresponding component and how to use ex-situ test data to deter-mine the amount of damaged experienced by the test article (and hence the corresponding compo-nent) in service. Specifically, this report derives and validated simplified methods for both tasks, replacing earlier, more complicated approaches based on simulating the test article response and finding the best specimen design or current damage via complex numerical optimization. The sim-plified methods for each task developed here can be implemented in spreadsheet software and are simple enough for practical use in future operating plants. This report derives the methods (one for sizing and two options for damage inference), verifies the simplified approaches versus more so-phisticated methods, and compares the results of applying the simplified methods to previous re-sults using the complex numerical optimization approach. The report also provides a worked sam-pled problem applying the simplified techniques to a realistic high temperature reactor component.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Forensic Analysis of Faulted NSTX-U Inner Poloidal Field Coil

The failure of the NSTX-U PF1A upper coil suspended plasma operations during the project's first operational campaign after commissioning. Initial nondestructive testing and examination of the failed coil was followed by extensive nondestructive radiography. Destructive testing facilitated visual borescope/videoscope inspection through the cooling path, vacuum testing, and electrical testing of every conductor segment. One conductor cooling path visually evidenced a void through the sidewall of the cooling path in a layer-to-layer region. The identified void in the conductor cooling path was not proximal to a braze or joggle. Electrical testing indicated that the voided conductor segment was a member of a group of 14 conductor segments that evidenced low-resistance connectivity. Four braze joints and two lead segments were subjected to 400-lbf/in 2 hydrostatic pressure testing followed by helium leak testing and evidenced that there were no detectable leaks in the tested cooling path segments. The faulted region was proximal to the center of the coil pack and evidenced historical electrical pitting and molten conductor debris. Samples of the epoxy resin insulation system were extracted and analyzed with dynamic mechanical analysis and differential scanning calorimetry techniques. Epoxy resin insulation samples were subjected to immersion testing per ASTM D570. Metallurgical samples were extracted from the coil pack conductors and subjected to hardness testing and grain structure analysis. The base material testing results indicated no abnormalities. In conclusion, the coil incipient failure was determined to most likely have resulted from conductive material located between conductor and insulation layers.

25 ENERGY STORAGE↗