Modeling the influence of interaction layer formation on thermal conductivity of U–Mo dispersion fuel
Not Available
Burkes, Douglas E.↗
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Leveraging MARVEL and SPHERE to Demonstrate NEAMS Thermal Hydraulics Codes
The Single Primary Heat Extraction and Removal Emulator (SPHERE) and Microreactor Applications Research Validation and Evaluation (MARVEL) were utilized to demonstrate thermal hydraulics codes of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. A recent long-duration test of a sodium heat pipe conducted at the SPHERE facility was utilized for the validation of the heat pipe code Sockeye. The first 10 hours of operation featured a startup of the heat pipe from room temperature; this portion was modeled with Sockeye and compared to the experimental data. Measurements included several thermocouple measurements across the length of the heat pipe surface. Two of Sockeye’s heat pipe models were applied to experiment, as well as several models of the cooling at the condenser end of the heat pipe. Sockeye results gave a decent match to the experimental data, with notable discrepancy in the condenser section, where experimental data indicated a significant inactive length of the heat pipe. The MARVEL microreactor was modeled through a coupled simulation of System Analysis Module (SAM) and SubChannel Module (SCM). This multiscale coupled simulation was performed as an exercise to demonstrate the compatibility and applicability of these codes for thermal-hydraulic analysis of microreactors. The coupling between the SCM and SAM codes for multiscale modeling of MARVEL, was achieved through a domain overlapping approach. In this example a transient SAM simulation provided boundary conditions to SCM, and SCM provided pressure drop to SAM.
97 - MATHEMATICS AND COMPUTING↗
Enhanced micro-pocket fission detector for high temperature reactor evaluations - 259
A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat a l'energie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 deg. C. The MPFD technology utilizes small, multi-purpose, robust, in-core fission chambers and can be configured with a thermocouple. The small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. In addition, the accomplishments of this project have attracted independent funding from other Department of Energy Office of Nuclear Energy (DOE-NE) programs for MTR irradiations of the MPFD technology. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Thermal-Hydraulic-Mechanical Modeling and Simulation of Sodium-Potassium–Cooled MARVEL Microreactor Core
The U.S. Department of Energy's Microreactor Program, with Idaho National Laboratory's development of a nuclear microreactor applications test bed named MARVEL, aims to support R&D for the deployment of small, transportable reactors across civilian, industrial, and defense sectors. The MARVEL microreactor, an 85-kWth thermal fission reactor, incorporates TRIGA nuclear fuel and a sodium-potassium eutectic as its primary coolant, designed for safety and efficiency, with natural circulation eliminating the risk of critical heat flux conditions. The reliance on natural circulation for primary cooling means the reactor avoids using fuel spacers to minimize core pressure drop, which could disrupt the primary coolant's natural flow. However, the reactor core’s tight P/D ratio of 1.05, in the absence of fuel spacers, could pose a risk of fuel rod contact and increased peak cladding temperatures. To ensure the reactor safety, this study conducted computational modeling and simulations to investigate the reactor's thermal-hydraulic-mechanical characteristics, including the reactor core heat transfer coefficients, the potential for rod-to-rod contact, and assessed its impact on peak cladding temperature and overall reactor safety. The computational analyses of the MARVEL microreactor core revealed that the thermal deformation of fuel rods under worst-case scenario may lead to the fuel rod contact, but the peak cladding temperatures will remain significantly lower than the safety criteria, ensuring the safety operation reactor without fuel spacers under normal operating conditions.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
A model to predict thermal conductivity of irradiated U–Mo dispersion fuel
Not Available
Burkes, Douglas E.↗
Demonstrate Viability of Accelerated Fuel Qualification Approaches
This presentation summarizes the research under LDRD grogram, Demonstrate Viability of Accelerated Fuel Qualification Approaches. It is demonstrated that the fabrication process of uranium carbide fuel has been developed and neutronic, thermal and hydraulic analysis have been conducted for the Fission Accelerated Steady-state Testing (FAST). This program provides basis of expansion of FAST as an efficient platform to broad nuclear fuels.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
The Advanced Test Reactor
The Advanced Test Reactor (ATR) is currently the highest-powered research reactor in the world with a maximum thermal power of 250 MW. The ATR began operation in 1967 to support fuels and materials irradiation testing programs in the United States and current plans (2019) are to operate the ATR for two more decades. Since 2007, the ATR has been designated a National Scientific User Facility (NSUF) to provide more access for non-government researchers to use the ATR for academic irradiation research. This article is a brief description of the ATR history, construction, and experiment configurations.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Preliminary Scaling Analysis of Reactivity Initiated Accidents
The life cycle for new fuel development requires extensive modeling and testing of the fuel concept from benchtop scale experiments to full scale assemblies in prototypical conditions. As a result, the timeline to develop new fuel types can extend up to 20 years and hinders the ability to develop a diverse assortment of fuel types. The licensing of recent reactor designs such as Westinghouse's AP-600/1000 designs utilized a scaling methodology to perform nuclear safety code validation on experimental test data from scaled down thermal hydraulic facilities. The advantage of using a scaling methodology was a significant cost decrease in the testing program because it avoided needing to build full scale prototypes and focused on improving safety code's ability to predict test data. This study investigates how to apply a scaling hierarchy for fuel performance during a Reactivity Initiated Accident that could be implemented with scaling methods such as Hierarchical Two Tiered Scaling (H2TS), Fractional Scaling Analysis (FSA), and Dynamical System Scaling (DSS) .
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Computational and Experimental Examination of Simulated Core Damage and Relocation Dynamics of a BWR Fuel Assembly - 16013
The Japan Atomic Energy Agency (JAEA) plans a large-scale test to evaluate damage and relocation behavior of Boiling Water Reactor (BWR) core materials consisting of fuel rods (zirconia as fuel simulant), channel boxes, control blade assemblies as well as lower core-support structures. Its purpose is to contribute to the understanding of the core material relocation behavior in the event of severe accidents with the BWR design conditions for which existing experimental database is quite limited. Prior to large-scale testing anticipated at JAEA in the future, JAEA desires preliminary investigations to examine melting test pieces, being composed of a mixture of zirconia and metals by a non-transferred thermal plasma heating system. The purpose of such tests is to verify the materials and test piece will be heated by plasma as a primary heat source to the target temperature of ca. 2900 K and to collect data about the relocation behavior of the high-temperature materials. This paper describes the first in a series of tests demonstrating the effective use of thermal plasma technology to recreate the dynamics of a typical fuel core bundle channel assembly section. Results from preliminary computational simulations of simplified fuel bundle assemblies are presented illustrating the effectiveness of a 150 kW, non transferred (NTR) configuration of a thermal plasma argon jet to heat the assembly within a highly-insulated containment vessel. An experimental test program using the computational analyses as a basis and a Phoenix Solutions Co (PSC) commercial model PT150, 150 kW plasma torch is described and the test results are presented illustrating the time versus temperature history of the fuel bundle assembly melting dynamics. Details on plasma operation are discussed including: power requirements, current flow, voltage levels achieved, argon gas flow rates, processing times and standoff distances of the torch discharge from the top of the fuel bundle assembly (achieved with remote control of the axial (vertical) translation of the torch within the containment processing chamber). Photos taken from the top of the processing chamber are presented illustrating the various phases of melting characteristics of the zircaloy-clad, zirconium oxide filled fuel tubes, all within a stainless steel tube bundle channel centered within an insulated zirconia crucible. Based on lessons learned from this initial test campaign, a brief discussion of the requirements and features of a follow-on Phase II test program is provided. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report
A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chamber and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A project meeting was held at KSU to discuss the roles and responsibilities between INL and KSU for development of the HT MPFDs. Provide input to various irradiation programs for installation of the MPFD technology in irradiation tests. As documented in this report, FY15 funding has allowed the project to meet year one planned accomplishments to develop a HT MPFD that offers US MTR users enhanced capabilities for real-time measurement of flux and temperature with a single detector. In addition, the accomplishments of this project have attracted funding from other Department of Energy Office of Nuclear Energy (DOE-NE) programs for additional applications. The work in those programs will build on current activities completed in this NEETASI HT MPFD project, but the MPFD will be specifically tailored to meet their program needs.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Test Report for DPP-1 Regulatory Compliance Testing. Volume 1 - Main Report
Consolidated Nuclear Security, LLC (CNS), the management and operating contractor for the Y-12 National Security Complex (Y-12) for the National Nuclear Security Administration (NNSA), is currently responsible for the regulatory testing of a new Type B fissile material shipping package called the Defense Programs Package (DPP)-1. The DPP-1 is a drum packaging with an inner liner and a removable lid. Both the drum body and the lid are filled with an impact-limiting, thermal-insulating material that protects the inner containment vessel (CV) during normal conditions of transport (NCT) and hypothetical accident conditions (HAC) as defined by 10 US code of Federal Regulations (CFR) 71. In fiscal year 2021, the Oak Ridge National Laboratory (ORNL) Package Testing Program (PTP), located at the National Transportation Research Center (NTRC), conducted Type B performance tests as prescribed in Title 10 Code of Federal Regulations Part 71 (10 CFR 71) on seven DPP-1 prototype units. This test report documents test unit (TU) preparation, pretest condition, conditioning of each TU, NCT and HAC testing, and posttest measurements and observations made of the damage resulting from these tests.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Local Doppler broadening effects from thermal neutrons during transient testing
Doppler broadening refers to the changing of a nuclear cross-section as a response to kinetic energy changes from interaction on a single-nucleus level. Interactions between irradiated particles causes a change in vibration of the nucleus, which is commonly correlated to the overall material's temperature, affecting how irradiate materials interact with additional incident particles. This phenomenon poses unique problems when specifically addressing transient conditions in the Transient Reactor Test Facility (TREAT) at Idaho National Laboratories (INL) because the facility is mainly concerned with in the thermal energy range (velocities beneath 2200 m/s) interacting with a plethora of alloyed metals inserted into the experimental test train slot of the facility. This work investigates the past validation of codes for on-the-fly (OTF) Doppler broadening programs and cross references them with experiments to determine the thermal range and limitations of the TREAT facility for transient conditions. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Ceramography and Thermal Modelling of Irradiated TRISO Particles using BISON
This study presents a comprehensive analysis of irradiated TRISO fuel particles from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program conducted at the Advanced Test Reactor (ATR) as part of my internship. Failure Analysis using high-resolution microscopy was performed to characterize mechanisms in TRISO particle layers, including the kernel, buffer, inner pyrolytic carbon (IPyC), silicon carbide (SiC), and outer pyrolytic carbon (OPyC) coatings. Failure modes were systematically classified. Thermal conductivity properties of each TRISO layer were evaluated through experimental measurements and computational models for use in BISON finite element code. Experimental thermal conductivity values ranged from 0.5 W/m·K for the buffer layer to 168 W/m·K for SiC, while BISON simulations provided averaged values of 0.57 W/m·K (buffer), 4.0 W/m·K (PyC layers), and 15.95 W/m·K (SiC). Temperature-dependent correlations from PARFUME were implemented for kernel thermal properties.
BISON↗
Nevada National Security Site Environmental Report 2018
This report presents the National Nuclear Security Administration Nevada Field Office environmental protection and monitoring programs data and the compliance status for calendar year 2018 at the Nevada National Security Site and at its two support facilities, the North Las Vegas Facility and the Remote Sensing Laboratory-Nellis. This report also addresses environmental restoration projects conducted by the Environmental Management Nevada Program Office at the Tonopah Test Range.
54 ENVIRONMENTAL SCIENCES↗
THERMAL ANNEALING OF IRRADIATED PREHYDRIDED ZIRCALOY-4 MATERIALS
Hydrogen-assisted irradiation growth is an important issue for reliability and performance of nuclear materials in nuclear fuel cladding and fuel assembly components. The mechanisms of irradiation growth are not entirely clear and the role of hydrogen and hydrides are less so. There is evidence that breakaway irradiation growth is strongly influenced by the formation and dynamics of dislocation loops that vary with alloy composition, heat treatment and hydrogen content. Thermal annealing of eight Zircaloy-4 specimens irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory as part of the joint EPRI Nuclear Fuels Industry Research (NFIR) and Fuel Reliability Program (FRP) AsTeR irradiation growth study were performed by hot-stage transmission electron microscopy (TEM), differential scanning calorimetry (DSC), and dilatometry via NSUF RTE 17-1093. Two fluence levels (6.7 and 12.3 dpa) and three hydrogen contents (~10, 40, 120 ppm) were investigated. TEM with a hot stage was used for the quantification of "a" and "c" dislocation loop population densities and size distributions at a range of temperatures from 25 to 800oC. DSC and dilatometry were performed by dynamically heating and cooling from 30 to 750ºC at rates of 5, 10, and 20 oC/min to determine the behavior of dislocation loop annealing. Dynamic temperature DSC tests were able to detect the temperature ranges and enthalpies of "a" and "c" loop annihilation, hydride dissolution and precipitation, secondary particle precipitation, and a?a + ß phase transformations upon heating and cooling. Dilatometry tests showed that all of the test showed initial linear expansion followed by a decrease in expansion rate from about 400 to 600 ºC associated with annihilation of "a" and "c" loops and showed evidence for the thermal stabilizing effect of hydrogen on dislocation loops. The results are discussed in the context of the larger AsTeR program that will include higher exposure (~20 dpa) and different alloys when they become available in 2021.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Thermal Hydraulics Validation Activities in the ART-GCR Campaign
The Advanced Reactor Technologies - Gas Cooled Reactor Campaign (ART-GCR) is part of the United States Department of Energy - Office of Nuclear Energy's Advanced Reactor Technologies program aimed at developing a High Temperature Gas-cooled Reactor (HTGR) which will offer enhancements in safety and efficiency. The Design, Methods, and Validation arm of the ART-GCR program provides experience and advanced tools for HTGR design and analysis. The Methods and Validation arm is developing data to validate predictions of decay heat removal through an extensive experimental campaign the Natural Convection Shutdown Heat Removal Test Facility at Argonne National Laboratory. In-core thermal hydraulics validation is being facilitated through the OECD/NEA's Thermal Hydraulic Code Validation Benchmark for High Temperature Gas-Cooled Reactors using HTTF Data. That benchmark contains code-to-code and code-to-data comparisons for 3 sets of HTGR thermal hydraulics phenomena. This ongoing benchmark activity has 16 participants from 9 countries developing leading insights into HTGR thermal hydraulics code validation. The ART-GCR campaign is also working on the development of state-of-the-art MOOSE-based models of the High Temperature Engineering Test Reactor (HTTR). Through the OECD/NEA's HTTR Loss of Forced Cooling (LOFC) program, the ART-GCR campaign is developing a multiphysics transient code validation benchmark based on 3 HTTR experiments that will provide an opportunity for international collaboration on multiphysics code validation.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Simulations of the High Temperature Test Facility using SAM
Under the support from the U.S. Department of Energy Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, an effort is being pursued to support the modeling and simulation needs of high-temperature gas-cooled reactor (HTGR) technology development. There is a particular need for advanced modeling and simulation methods and tools to predict thermal-fluid behavior in the nuclear reactor primary system during safety-related transients. This report focuses on one such activity related to HTGR: developing a model of the High Temperature Test Facility (HTTF) at Oregon State University using the system-level code SAM, and using the model to understand thermal response behavior in the facility. Note this activity is coordinated with the DOE-NE’s Advanced Reactor Technology Gas-Cooled Reactor Program
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Response Bias of Electrical Cable Coatings at FIRE Conditions (REBECCA-FIRE)
This report presents the results of a series of cable fire-retardant coating tests sponsored by the US Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research and performed at Sandia National Laboratories in conjunction with the National Institute of Standards and Technology (NIST). The goal of the tests was to assess the effects of three commercially available fire-retardant cable coating materials on cable thermal and electrical response behavior under fire-exposure conditions. The specific test objectives were to assess, under severe radiant heating conditions, how the coating materials impacted (1) cable thermal response and (2) electrical integrity behavior. The tests were not explicitly designed to assess the impact of the coatings on cable flammability, although some insights relative to the burning behavior of the coating materials themselves and cable ignition times were gained. NIST is currently investigating these attributes under the Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIEFIRE) program (NUREG/CR-7010). The cables used in construction of the test articles were all seven-conductor 12AWG (American wire gage) control or power type copper conductor electrical cables. Two cable insulation types were represented, a polyethylene thermoplastic material and a cross-linked polyethylene thermoset material. Both cable types used have been tested extensively in recent NRC-sponsored experimental programs involving both circuit failure modes and effects testing and fire growth testing. The test articles included uncoated cables and cables coated with one of three fire-retardant coating materials: Carboline Intumastic 285, Flamemastic F-77, and Vimasco 3i. Test configurations included single lengths of cables, bundles of seven cables, and bundles of ten cables. The tests show that, under certain conditions, the fire-retardant coatings provide a substantial benefit relative to delays in cable heating, ignition and electrical failure times. However, as has been seen in prior test programs, the performance varied substantially among the coating products. The current tests also show that the benefit gained by the coatings was heavily dependent on the thermal mass of the coated cable system. Low thermal mass systems, such as the single lengths of coated cable, saw essentially no net benefit from application of the coatings. Intermediate mass systems, represented by the seven-cable bundles, saw some benefit from application of the coatings, but the benefit was inconsistent, and some cables in the bundles saw essentially no delay in thermal response or time to failure. For the larger thermal mass systems, represented by the ten-cable bundles, the benefit of the coatings was both more pronounced and more consistent with all coatings providing a measurable benefit.
36 MATERIALS SCIENCE↗