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At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Results from a scaled reactor cavity cooling system with water at steady state

We present a summary of steady-state experiments performed with a scaled, water-cooled Reactor Cavity Cooling System (RCCS) at the Univ. of Wisconsin - Madison. The RCCS concept is used for passive decay heat removal in the Next Generation Nuclear Plant (NGNP) design and was based on open literature of the GA-MHTGR, HTR-10 and AVR reactor. The RCCS is a 1/4 scale model of the full scale prototype system, with a 7.6 m structure housing, a 5 m tall test section, and 1,200 liter water storage tank. Radiant heaters impose a heat flux onto a three riser tube test section, representing a 5 deg. radial sector of the actual 360 deg. RCCS design. The maximum heat flux and power levels are 25 kW/m{sup 2} and 42.5 kW, and can be configured for variable, axial, or radial power profiles to simulate prototypic conditions. Experimental results yielded measurements of local surface temperatures, internal water temperatures, volumetric flow rates, and pressure drop along the test section and into the water storage tank. The majority of the tests achieved a steady state condition while remaining single-phase. A selected number of experiments were allowed to reach saturation and subsequently two-phase flow. RELAP5 simulations with the experimental data have been refined during test facility development and separate effects validation of the experimental facility. This test series represents the completion of our steady-state testing, with future experiments investigating normal and off-normal accident scenarios with two-phase flow effects. The ultimate goal of the project is to combine experimental data from UW - Madison, UI, ANL, and Texas A and M, with system model simulations to ascertain the feasibility of the RCCS as a successful long-term heat removal system during accident scenarios for the NGNP. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Design and development of a high-temperature sodium compatibility testing facility

The use of advanced alloys within sodium-cooled fast reactors (SFRs) has been identified as a means of increasing plant efficiency and reducing construction costs. In particular, alloys such as NF-616, NF-709 and HT-UPS are promising because they exhibit greater strength than traditional structural materials such as 316-SS. However, almost nothing is known about the sodium compatibility of these new alloys. Therefore, research taking place at the Univ. of Wisconsin-Madison is focused on studying the effects of sodium corrosion on these materials under prototypic SFR operating conditions (600 [ deg. C], V Na=10 [m/s], C 0{approx} 1 [wppm]). This paper focuses on the design and construction of the testing facility with an emphasis on moving magnet pumps (MMPs). Corrosion data from a preliminary 500 [hr] natural convection test will also be presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to the Baffle Former Plates. The FaST is designed to remove the Baffle Former Plates from the Core Barrel. The VRS further volume reduces segmented components using multiple configurations of the 38i and horizontal reciprocating saws. After the successful removal and volume reduction of the Internals, the RV will be segmented using a 'First in the US' thermal cutting process through a co-operative effort with Siempelkamp NIS Ingenieurgesellschaft mbH using their experience at the Stade NPP and Karlsruhe in Germany. SNS mobilized in the fall of 2011 to commence execution of the project in order to complete the RVI segmentation, removal and packaging activities for the first unit (Unit 2) by end of the 2012/beginning 2013 and then mobilize to the second unit, Unit 1. Parallel to the completion of the segmentation of the reactor vessel internals at Unit 1, SNS will segment the Unit 2 pressure vessel and at completion move to Unit 1. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Using Calibrated Water Data for Preliminary Validation of the SRT Code for Advanced Reactors

Recent interest and corresponding progress worldwide regarding advanced nuclear reactors has renewed focus on their performance and related safety assessments. Specifically, the U.S. Nuclear Regulatory Commission has emphasized the importance of mechanistic approaches to source term analysis for advanced reactor licensing applications, which attempt to realistically account radionuclide transport and retention phenomena. Further model development is required due to the numerous and complex physical and chemical phenomena associated with mechanistic source term analyses. Reflecting the need for modeling advancement, Argonne National Laboratory developed a mechanistic source term analysis tool for sodium fast reactors. The Simplified Radionuclide Transport (SRT) code describes fuel pin failure (for simulating the initial condition at the point of fuel pin breach), bubble scrubbing, deposition, leakage and following environmental impact. In the current work, a validation study of the SRT bubble scrubbing model is performed using a water-loop experiment performed at the University of Wisconsin-Madison. Through the analysis, the approach and fundamental bubble scrubbing models in SRT, which examine the removal of aerosols within the bubble as it is transported through a pool, have been widely evaluated. The results of the assessment demonstrate a high level of agreement in the regions of greater aerosol size. In the parameter range of minimal aerosol removal, the simulation slightly underpredicts the experiment results; however, considering the scale of plots and huge uncertainties inherently included, the deviation can be judged to be minor and would produce a conservative result. In addition, uncertainty analysis has been further refined to reflect the experimental distribution of parameters including aerosol sizes, which induces a span of performance for each representative aerosol size. Based upon the initial validation results along with uncertainty effects, SRT is expected to provide meaningful insights for the analysis of bubble scrubbing. Future sodium-loop tests will provide further validation basis.

Kam, Dong Hoon↗
DOE OSTI2016

Water-cooled Reactor Cavity Cooling System Flow Analysis

A research team at University of Wisconsin - Madison designed and constructed a 1/4 scaled experimental facility to study natural circulation cooling in a water based reactor cavity cooling system (WRCCS) for decay heat removal in an advanced high temperature reactor. The facility is capable of natural circulation operation scaled for simulated decay heat removal (up to 45 kW input power, which is equivalent to 1.70 MW at full scale) and pressurized up to 2 bar pressure. The WRCCS facility has been used to study natural circulation flow behavior from transients due to evaporation of the water inventory from decay heat removal. The natural circulation flow is observed to be oscillatory with the mean flow increasing as the pressure head in the system drops due to evaporation of the water inventory during two-phase operation. A first principles model was developed to understand the behavioral limits and trends. Through the use of void fraction measurements and the model, we find that the mass flow rate in the system increases due to small increases in the time-averaged void fraction over the course of a four-hour boil-off experiment. This increase in void fraction causes an increase in the buoyancy head and mass flow. The model also determines the operational limits on the mass flow rate under two-phase operation that can be verified with extended tests in the WRCCS facility. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The Current State of Radiation Dosimetry in Radionuclide Therapy

Radionuclide therapy combines the specificity of cancer cell targeting with the well-known cytotoxic effects of ionizing radiation. It has the potential to simultaneously eradicate both the primary tumor site as well as the metastatic extent of the disease throughout the body. Despite these advantages, a major limitation of radionuclide therapy is that conventional internal dosimetry methods have failed to accurately predict treatment outcomes in patients. It has been proposed that this shortcoming is due to the fact that therapeutic dosimetry is typically performed using methods developed for diagnostic imaging that do not consider patient specific anatomy, pharmacokinetics and dose heterogeneity. Even worse, dosimetry methods are often completely neglected during treatment planning and dose prescription is based on Body Surface Area/Weight (BSA/BW) and platelet counts. The most widely used internal dosimetry software is OLINDA/EXM (Organ Level Internal Dose Assessment for Exponential Modeling). OLINDA/EXM calculates absorbed doses using the formulism developed by the Medical Internal Radiation Dose (MIRD) committee of the Society of Nuclear Medicine. In its current implementation, OLINDA/EXM requires the use of S factors, which are derived in phantoms that do not accurately represent the actual patient. In addition, OLINDA/EXM assumes that the activity distribution in each organ and tumor is uniform, when it has been well established that activity distributions are often highly heterogeneous. Furthermore, the tumor is not modeled in the phantom, which introduces additional uncertainties in the calculation. Due to these limitations, there is a clear need for improved dosimetry methods for therapeutic nuclear medicine that utilize patient-specific imaging information. A variety of efforts are currently underway to develop such methods. There are two commercial software packages on the market for voxel level internal dosimetry: STRATOS (Philips Technologies, Aachen, Germany) and VoxelDose (DosiSoft, Cachan, France), but both packages are not widely used in clinical practice. In addition, several academic platforms have been developed. Similar to these offerings, RAPID is a robust, fast and multimodal platform. However, RAPID has been designed with tremendous flexibility to perform accurate dosimetry for any radiopharmaceutical in the pre-clinical and clinical setting in order to aid in drug development and to assist in radionuclide therapy treatment planning. This presentation will overview existing dosimetry platforms while focusing on the Monte Carlo-based platform developed at the University of Wisconsin known as RAPID (Radiopharmaceutical Assessment Platform for Internal Dosimetry). RAPID is a robust and universal platform that can provide accurate 3D internal dosimetry for any radiopharmaceutical during both the pre-clinical and clinical stages of drug development as well as clinical treatment planning in TRT. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

A Comparison of the U.S. Nuclear Regulatory Commission Technical Evaluation for the Zion and La Crosse License Termination Plans with a Focus on Release Criteria

Zion Nuclear Power Station (ZNPS) and the La Crosse Boiling Water Reactor (LACBWR) sites are both undergoing decommissioning. Although these two sites are different in terms of size, design, and operational history, the approach for defining release criteria for decommissioning at each site is similar. ZNPS consists of two (Units 1 and 2) Pressurized Water Reactors (PWR), each approximately 1100 MWe. ZNPS is located near the city of Zion in northeast Illinois on the west shore of Lake Michigan. The two units were operated from 1973 to 1998. LACBWR was an AEC Demonstration Reactor (50 MWe) which operated from 1967 to 1987. The LACBWR site is in Vernon County, Wisconsin and is shared with an operational fossil plant. EnergySolutions is managing the decommissioning of both these sites under the separate entities ZionSolutions and LaCrosseSolutions. The U.S. Nuclear Regulatory Commission (NRC) transferred the Zion Facility Operating License from Exelon Generation Company, LLC (Exelon) to ZionSolutions on September 1, 2010. A similar license transfer from Dairyland Power Cooperative to LaCrosseSolutions is planned for the second quarter of 2016. On December 19, 2014, ZionSolutions submitted the ZNPS License Termination Plan (LTP). It is expected that the LTP for the La Crosse BWR will be submitted to the NRC in early 2016. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Microstructural Characterization of In-reactor Corrosion Tested Hastelloy N{sup R} and 316 Stainless Steel

The success of the Molten Salt Reactor Experiment (MSRE) at the Oak Ridge National Laboratory (ORNL) in the period 1950's-70's has rekindled the interested in molten salt-cooled reactors. The most recent form of this type of reactor, the Fluoride salt-cooled High-temperature nuclear Reactors (FHRs) is emerging as a leading reactor concept for the next generation nuclear reactors because it offers, among other benefits, a high degree of passive safety, high thermal efficiency, and low spent fuel. Since 2011, two three-year Integrated Research Projects (IRPs) were awarded by the U. S. Department of Energy (DoE) to the Massachusetts Institute of Technology (MIT) in partnership with the University of California, Berkeley and the University of Wisconsin-Madison to develop a 'path forward' for the development of the FHR. One primary challenge in these projects is to evaluate the corrosion and compatibility of structural alloys in the potential primary coolant of molten Li{sub 2}BeF{sub 4} (FLiBe) salt at 700 deg. C in a neutron irradiation environment. In support of structural materials development for the FHR, the static in-reactor corrosion tests of two promising alloys, nickel-based Hastelloy N{sup R} and 316 stainless steel, in {sup 7}Li-enriched molten FLiBe have been conducted in both metal-lined and graphite crucibles in the MIT nuclear research reactor in 2013. This work focuses on the microstructural characterization of the in-reactor corrosion tested alloys for understanding of their performance in high-temperature molten FLiBe salt. Our preliminary results indicate significant thermal effects, graphite effects, and irradiation effects on the alloys' performance in high-temperature molten FLiBe salt, particularly for the in-reactor corrosion tests. The neutron irradiation induces tritium generation in molten salt which could combine with fluorine ions to form corrosive HF to dramatically attack alloys. Additionally, the neutron irradiation introduces structural defects into alloys which probably alter alloys' corrosion resistivity to high temperature molten FLiBe salt. The effect of graphite on the alloys' corrosion in a nuclear reactor is also possibly different from the out-of-reactor corrosion. All these issues will be better understood by characterizing the microstructure of in-reactor corrosion tested alloy samples in our near future plans. The experimental data in this study provides valuable input for evaluating structural materials performance in the FHR. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Impact of Spatial Discretization on Reactivity Biases in Depleted TRIGA Fuel

Ongoing work at Kansas State University (KSU) and the University of Wisconsin involves the determination of fuel compositions in TRIGA reactors. Modeling of depletion requires a number of approximations to be made, which often includes selection of an appropriate spatial discretization for tracking material evolution in time. Incorporation of temperature feedback also requires such spatial zoning. Any approximation made leads to computational biases, i.e., differences between computed and accurate values of some parameter of interest. The objective of this study is to understand the potential magnitude of biases introduced by the selected spatial discretization. Specifically, biases in the reactivity of a square-pitch unit cell representative of a typical TRIGA fuel element were studied as a function of burnup for different radial and axial discretizations. Temperature effects were also studied by introducing a fixed, radial temperature profile. The results indicate a very fine radial discretization (> 10 radial regions) may be needed to eliminate biases (of statistical significance) from temperature feedback effects, while an axial discretization of {approx_equal}7 regions may be sufficient to eliminate biases due to evolution of the fuel composition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Analysis on the Heat Transfer Phenomena Inside the Scaled-Down VHTR RCCS Riser Experiment

Very High Temperature gas-cooled Reactor (VHTR) that is one of GEN-IV reactors has been developed for the purpose of hydrogen production. VHTR is characterized by high operating temperature and stepped-up inherent safety given by passive safety systems such as Reactor Cavity Cooling System (RCCS). In case the active heat removal system losses its capability, RCCS significantly contributes to the inherent safety of VHTR by removing the decay heat transferred from the reactor core to the reactor vessel. Therefore, the performance of RCCS is the key parameter to ensure the safety of the reactor. Several computation and experiment researches on the performance of RCCS have been conducted. There are two types of RCCS depending on its coolant; one is the air-cooled type and the other is water-cooled type. This study focused on the former type. Due to the excessive height of the prototype RCCS, experiment facilities were constructed in reduced scale. Korea Atomic Energy Research Institute (KAERI) constructed the quarter-scale RCCS experiment facility, NACEF and Argonne National Laboratory (ANL) has constructed half-scale experiment facility, NSTF. University of Wisconsin (UW) also has the quarter-scale experiment facility and Texas A and M University has conducted separate effect test on the jet effect at the upper plenum of RCCS. From the reduced scale experiment result, the performance of RCCS such as a heat removal rate of RCCS and a peak temperature of reactor vessel were evaluated. When the reduced-scale result is applied to the prototype, however, a distortion could be brought about due to the reduced height of RCCS in the experiments. Especially, if the heat transfer regime inside the riser which roles as a heat exchanger to discharge the decay heat into the atmosphere is not identical, the riser wall temperature and the consequent reactor vessel temperature of the prototype could be incorrectly predicted. Thus, the convective heat transfer phenomena inside the riser duct both in the reduced scale and in the prototype is necessary to be investigated. In this study, the preliminary analysis on the heat transfer phenomena inside the RCCS riser was conducted. The subjects of the analysis were the prototype RCCS riser of PMR 200, and its scaled-down experiment facilities, NACEF. First, the geometry, flow and thermal conditions of the riser in PMR200 RCCS and NACEF were introduced with a scaling method. And then, the heat transfer regimes and the heat transfer coefficient corresponding to each condition were predicted using correlation and CFD calculation. Finally, planned activities including the experiment was introduced. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Development of Accident Tolerant Fuel Options for Near Term Applications: Fuel Performance Modeling under Transient / Severe Accidents by MELCOR: PART I: Benchmark for SURRY: Short Term Station Black Out (STSBO)

Currently, existing LWR fuel technology is near its inherent limit of performance without improvement in fuel materials. As a result of the great earthquake of east Japan, along with the catastrophic tsunami, the Fukushima Daiichi boiling water plant was severely damaged. The light water reactors (LWRs) fuel gave rise to a serious discussion in the United States. In order to provide the development of Accident Tolerant Fuel (ATF) options for present day applications, MIT is leading a DOE project to search for enhanced accident tolerant fuels. Group members include MIT, UW-Madison, TAMU, AREVA, and Westinghouse. As part of the work, fuel performance modeling under transient / severe accidents: MELCOR is being used to analyze different kinds of materials under several limited accident conditions. While various research groups are obtaining the parameters of the experimental ATF materials, benchmarks for SURRY are being done by use of MELCOR in the University of Wisconsin-Madison. We chose version 1.8.6 for the edition of materials, and INL provided the uncertainty analysis between version 1.8.6 and version 2.1.0. Comparison will be done between potential ATF materials, which include FeCrAl, SiC, Cr, ZrSi, TiAlC, TiSiC and so on. Furthermore, zircaloy is used as the verification. Among all the parameters which influence the materials' performance, the temperature history and hydrogen generation will be the key factors in the future. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of a Small Liquid Sodium Purification Loop

Generation IV nuclear reactors provide management solutions to address spent fuel problems. The most researched Generation IV reactor is the sodium cooled fast reactor (SFR). Improved engineering is has allowed for systematic improvements in both the safety and efficiency of power production. However, before a SFR can be brought back online in the Untied States, constructive research must be pursued on numerous scales to ensure the capabilities of aforementioned reactors. While larger research facilities such as the one at Argonne National Laboratory or the University of Wisconsin - Madison address some of the larger scale studies surrounding the SFR, Fort Lewis College (FLC) is aiming to provide small scale studies on liquid sodium as a thermal hydraulic cooling system. The sodium diagnostic loop at FLC is constructed to conduct research that includes: analyzing pumping characteristics, thermal stratification, freeze plug analysis, and corrosion studies. The FLC loop is attached to an inert glovebox so these experiments can be conducted with an additional layer of safety. The objective of this sodium research apparatus is to combine the best features of each instrument commonly associated with sodium research loops in a cost effective manner while still obtaining high fidelity data. Furthermore, an innovative design of the loop maximizes the volume of available research space inside of the glovebox by mounting the diagnostic loop adjacent to the glovebox.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Progress in Development of Fluoride-salt-cooled High-Temperature Reactors (FHRs)

Fluoride-salt-cooled High-temperature Reactors (FHRs) are a new class of reactors that use graphite-matrix coated-particle fuel with a clean liquid salt coolant. The largest university FHR program is an Integrated Research Project (IRP) sponsored by the U.S. Department of Energy and others that is led by the Massachusetts Institute of Technology (MIT) and includes the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The year 2016 was the fifth year of this program. We report herein some of the major results of work done in 2016. The base-line FHR concept studied in this IRP uses (1) graphite-matrix coated-particle fuel with a pebble bed geometry and (2) a clean liquid salt coolant-flibe (a mixture of {sup 7}LiF and BeF{sub 2}). The combination of a very high-temperature fuel with failure temperatures above 1650 deg. C, a coolant boiling point above 1400 deg. C, and a coolant that traps most fission products implies the capability to assure no major fuel failure under severe accident conditions. There are a family of FHR concepts using different fuel geometries and different fluoride salts. The power conversion system is a Nuclear Air-Brayton Combined Cycle (NACC) with peak power provided by natural gas or a Firebrick Resistance Heated Energy Storage (FIRES) system. This enables a base-load reactor to provide variable electricity to the grid that substantially increases revenue relative to base-load nuclear reactors. Most of the results reported herein are applicable to most FHR concepts as well as liquid-fueled molten salt reactors (MSRs)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Diffusion bonded compact heat exchanger in Inconel 740H for high temperature and high-pressure applications

It is desired to develop heat exchangers able to be operated at high temperatures (750˚C) and pressures (3636 psi) for use in advanced Brayton power cycle. Conventional materials have not had the high strength at temperature required for these applications previously, however a fairly new material Inconel 740H developed by Special Metals for the ultra-supercritical water fossil industry has shown good performance under these conditions. In further efforts to capitalize on these new high strength high temperature alloys CompRex has developed a diffusion bonding procedure capable of alloying compact high effectiveness heat exchangers that can perform under these high temperature and high-pressure conditions. Initial tests of the heat exchanger are being conducted at the University of Wisconsin using supercritical CO2 working fluid. The bonding process has been adjusted such that the strength of the bonded block has achieved over 90% yield and tensile strength of the 740H base material values. Grain size before and after bonding were measured to determine the extent of grain growth and grain diffusion across the interface. A discussion of the heat exchanger design, etching and initial tensile tests on sample bonds will be discussed.

14 SOLAR ENERGY↗
DOE OSTI2024

High Fidelity Simulations of Air-Cooled Reactor Cavity Cooling System

Nuclear energy is increasingly acknowledged as pivotal in the global shift towards cleaner energy solutions. Advanced nuclear technologies, including High Temperature Gas-cooled Reactors (HTGRs), stand out as appealing options among Generation IV reactors due to their high temperature heat output and potential for cogeneration. HTGR designs incorporate passive safety systems, such as the Reactor Cavity Cooling System (RCCS), which utilize natural principles to manage heat dissipation from the reactor pressure vessel (RPV) during accidents or routine shutdowns. Regulatory bodies require thorough validation of safety systems like the RCCS to ensure they meet specified standards. Consequently, there is a pressing need within the industry for advanced simulation tools capable of assessing these systems’ performance accurately. There is a knowladge gap in the literature concerning high-fidelity data for the RCCS, which motivates the focus of this study. This research focuses on a specific RCCS designed for the Modular High-Temperature Gas Reactor developed by General Atomics (GA- MHTGR). Experimental studies on a scaled version of the air-cooled RCCS of GA-MHTGR were conducted by the University of Wisconsin-Madison. This work contributes to a broader initiative aimed at establishing a numerical benchmark based on the UW-Madison experiments. As first step we performed high fidelity simulations of the experimental facility setup, to analyze flow physics in such systems and validate NekRS and the MOOSE heat transfer and radiation modules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

High Fidelity Simulations of Air-Cooled Reactor Cavity Cooling System

High Temperature Gas Reactor (HTGR) designs incorporate passive safety systems (e.g., the Reactor Cavity Cooling System [RCCS]) that utilize natural principles to manage heat dissipation from the reactor pressure vessel (RPV) during accidents or routine shutdowns. The industry community is experiencing a pressing need for advanced simulation tools that can accurately assess the performance of these types of systems. In the literature, a knowledge gap exists concerning high-fidelity data for the RCCS, and this gap is one of the areas of focus of the present study. This research focuses on a specific RCCS designed for General Atomics' Modular High-Temperature Gas Reactor (GA-MHTGR). Experimental studies on a scaled version (can be seen in Figure 1) of the air-cooled RCCS used in GA-MHTGR were conducted by the University of Wisconsin-Madison (UW-Madison). This work contributes to a broader initiative aimed at establishing a numerical benchmark based on the UW-Madison experiments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Integrated Traction Drive Thermal Management: Keystone Project 3

The presentation represents a report of accomplishments during FY 2024 and is prepared for 2024 DOE VTO Annual Merit Review. The project goal is identifying pathways enabling high-performance, compact, and reliable integrated electric drives and facilitation of achieving DOE 2025 target of 33 kW/L system power density for an electric traction drive through support of other EDT consortium members in design and development of thermal management systems for their respective integrated drive concepts. Specific objectives are: 1) research and evaluation of motor-integrated power electronics packaging technologies and thermal management approaches; 2) explore novel materials and manufacturing options (3D printing with thermally enhanced polymers or ceramic materials) for key thermal management system components; 3) develop thermal management system and its sub-components to enable integrated electric drive DOE power density targets in collaboration with project partners; 4) support activities of DOE's Electric Drive Technologies (EDT) consortium members, Oak Ridge National Laboratory (ORNL), University of Wisconsin and Georgia Tech research teams in thermal management component design and thermal modeling of their integrated traction drives.

ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2024

Electric Motor Thermal Management

The poster reports the accomplishments of EDT-Electric Motor Thermal Management consortium project for 2024 VTO Annual Merit Review (AMR). The overall aims of the project are to support research enabling compact, reliable, low-cost, and efficient electric machines aligned with roadmap research areas; to collaborate with ORNL, Ames, and SNL to provide motor thermal analysis support, reliability evaluation, and material measurements on related motor research at national laboratories; and collaborate with university partners including Georgia Institute of Technology and University of Wisconsin Madison to support university-led motor thermal management research efforts.

ADVANCED PROPULSION SYSTEMS↗