Restart of Transient Fuels Testing at the Annular Core Research Reactor.
Abstract not provided.
Martin, William Joseph↗
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Conceptual Phase Neutronic and Thermal Analyses for the LEU Conversion of the Transient Reactor Test Facility (TREAT)
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Connaway, Heather M.↗
Temperature Limited Transient Calculations for the Transient Reactor Test Facility (TREAT) Using MCNP and the Point Kinetics Code TREKIN
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Kontogeorgakos, Dimitrios C.↗
Optimization of a Fast Neutron Scintillator for Real-Time Pulse Shape Discrimination in the Transient Reactor Test Facility (TREAT) Hodoscope
We present a multi-channel, fast neutron/gamma ray detector array system that utilizes ZnS(Ag) scintillator detectors. The system employs field programmable gate arrays (FPGAs) to do real-time all digital neutron/gamma ray discrimination with pulse height and time histograms to allow count rates in excess of 1,000,000 pulses per second per channel. The system detector number is scalable in blocks of 16 channels.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Optimization of a Monte Carlo Model of the Transient Reactor Test Facility
The ultimate goal of modeling and simulation is to obtain reasonable answers to problems that don’t have representations which can be easily evaluated while minimizing the amount of computational resources. With the advances during the last twenty years of large scale computing centers, researchers have had the ability to create a multitude of tools to minimize the number of approximations necessary when modeling a system. The tremendous power of these centers requires the user to possess an immense amount of knowledge to optimize the models for accuracy and efficiency.This paper seeks to evaluate the KENO model of TREAT to optimize calculational efforts.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Optimization of a Monte Carlo Model of the Transient Reactor Test Facility
The ultimate goal of modeling and simulation is to obtain reasonable answers to problems that don't have representations which can be easily evaluated while minimizing the amount of computational resources. With the advances during the last twenty years of large scale computing centers, researchers have had the ability to create a multitude of tools to minimize the number of approximations necessary when modeling a system. The tremendous power of these centers requires the user to possess an immense amount of knowledge to optimize the models for accuracy and efficiency. Research reactors require that the neutron flux be well characterized so that the flux in a given location or the power delivered to the target in an experiment center can be accurately determined.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
RIA Testing in TREAT
The Transient Reactor Test facility (TREAT) was constructed in the late 1950’s. The reactors flexible design and multi-mission nature provided thousands of transient irradiations for numerous reactor fuels and transient types before being placed in standby in 1994. The reactor resumed operations in 2017 to reclaim its crucial role in nuclear-heated safety research. A principal goal of restarting the reactor was to perform transient testing of Accident Tolerant Fuels (ATF), supporting their safe adoption by the commercial Light Water Reactor (LWR) fleet. In support of the Accident Tolerant Fuels mission, multiple capsules have been designed and are currently being designed to test fuel specimens under varying conditions. The Minimal Activation Retrievable Capsule Holder (MARCH) system was designed to support high throughput testing by using a modular approach where an outer containment structure will act as the primary safety containment. The MARCH system is designed to house various modules for testing different fuel systems in different environments. The near-term tests for the ATF concepts will use the Static Environment Rodlet Transient Test Apparatus (SERTTA), aka MARCH-SERTTA. MARCH-SERTTA is capable of testing a single ATF rodlet 10 cm in length in a static pool of water with initial conditions up to 200°C and 500 psi. The purpose of the MARCH-SERTTA device is to perform integral tests evaluating the Pellet Clad Mechanical Interaction (PCMI) performance of LWR fuel concepts in design basis Reactivity Initiated Accidents (RIA). This will be accomplished by testing a variety of ATF fuel concepts under varying initial thermal-hydraulic conditions and varying prompt energy pulses. The flexibility of TREAT and the MARCH-SERTTA capsule enables testing of a unique, non-fueled, rod specimen under similar conditions as the fueled specimens to perform separate-effect tests. This non-fueled rod will be used to study the thermal-hydraulic behavior (specifically determine the critical heat flux) during these fast prompt transients.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Nuclide inventory characterization of EBR-II MOX fuel test pins
Characterization of 3 mixed oxide (MOX) fuel pins from the Experimental Breeder Reactor II (EBR-II) SPA-2/-2B experiment program was performed in support of the Advanced Reactor Experiments for Sodium Fast Reactor Fuels (ARES) joint project between Idaho National Laboratory (INL) and the Japan Atomic Energy Agency (JAEA). The characterization efforts were performed to enable transient testing in the Transient Reactor Test (TREAT) facility at INL and will include post irradiation examination (PIE) and other analyses. MOX fuel pins UW02009, 02011, and 06024 were historically irradiated from 1989 to 1994 for a total of 822.42 effective full power days (EFPDs) and peak burnup of approximately 14.3 at.% (atomic percent of heavy metal atoms, ∼134.4 GWd/t). They had since been placed in storage until a future use was identified. Characterization of the fission products and activation of each fuel pin was necessary to enable removal from storage and subsequent irradiation testing or PIE activities. The initial fuel pin geometry and composition were established using historical documentation. Irradiation history details were utilized to compute irradiated and decayed fuel pin masses, atomic densities, and activities. A pellet-wise axial neutron flux was also computed to assist in evaluation of the axial distribution of {sup 235}U and {sup 239}Pu, per pellet, within each of the test fuel pins. The results computed herein supported experiment design for advanced high-burnup MOX fuel designs under slow transient overpower (MOXTOP) conditions, which was Phase I of the ARES project. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan
The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allow for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
The Fuel Motion Monitoring System at TREAT - Current Status and Future Plans
An important component of the United States Nuclear Fuel Safety Transient Testing Program, the Fuel Motion Monitoring System (FMMS) at Idaho National Laboratory's Transient Reactor Test Facility (TREAT) is fast-neutron hodoscope capable of imaging the location, movement, and relocation of nuclear fuel experiments under simulated transient accident conditions. The FMMS was refurbished in parallel with the TREAT restart program starting in 2014, restoring 96-channels of fast-neutron detection. Since returning to operation in 2017 the FMMS has supported many fuel safety experiments supporting accident tolerant fuel development, light-water reactor safety, space thermal nuclear propulsion fuel development, and advanced reactor research and development. In Phase 2 of the FMMS restoration. work is now under way to expand the FMMS' field-of-view by adding an additional 96 channels of fast neutron detectors to the system's hodoscope, along with an expanded data acquisition system and associated transient timing electronics. An overview of the FMSS system and its fast-neutron detectors will be presented along with examples of current FMMS imaging performance and associated information.
46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
GAIN Fuel Safety Research Workshop
The Gateway for Accelerated Innovation in Nuclear (GAIN) Fuel Safety Research Workshop was held in Idaho Falls, Idaho, at the Idaho National Laboratory (INL) meeting center in the Energy Innovation Laboratory (EIL). The purpose of the workshop was to gather fuel safety research needs from the broad range of industrial nuclear technology users and developers. The needs will be analyzed relative to the national laboratory capabilities to help inform program planning and funding. The workshop provided a forum to learn about transient testing from a research and regulatory perspective and tour participation of the Transient Reactor Test Facility (TREAT), Advanced Test Reactor (ATR), and Materials and Fuels Complex (MFC).
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Feasibility of Power Ramp Testing in the Advanced Test Reactor
For decades, the Halden Boiling Water Reactor (HBWR) in Norway was an international resource for assessing nuclear fuels and materials behavior and its unexpected shutdown in 2018 represented a significant loss in experimental capability for prototypical irradiation testing. In the aftermath of the closure, a study was performed to assess capability gaps related to the Accident Tolerant Fuels (ATF) program. It was concluded that the primary capability gaps left by the closure of the HBWR were the loss of prototypic in-pile Light Water Reactor (LWR) loops that provide operational transient testing and in-pile Loss-of-Coolant Accident (LOCA) testing capabilities. The study also concluded that the Advanced Test Reactor (ATR) and Transient Reactor Test Facility (TREAT) likely have the necessary key capabilities to absorb the breadth of the HBWR mission gaps related to the ATF program, but additional investments in experimental infrastructure were needed. One such investment is the design and installation of additional pressurized water loops in the medium-I positions of the ATR, hereafter referred to as I-Loops, to support power ramp testing and testing in Boiling Water Reactor (BWR) conditions. This paper gives an overview of the planned ATR I-Loops and assesses the feasibility of performing power ramp testing inside such a loop.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Results of the CHF-SERTTA In-Pile Transient Boiling Experiments at TREAT
To advance the current understanding of heat transfer between nuclear fuel rod cladding and the surrounding coolant under transient scenarios involving critical heat flux (CHF), the CHF-Static Environment Rodlet Transient Test Apparatus (CHFSERTTA), an in-pile experiment, was developed for and irradiated in the Transient Reactor Test (TREAT) facility at Idaho National Laboratory. The CHF-SERTTA experiment consists of a hollow, borated stainless steel rod submerged in a static water pool. The rod, heated via the (n, a) reaction in boron-10, was designed for investigating in-pile transient heat transfer between the rod surface and the coolant, avoiding the complexity of including the heat transfer between the fuel, gas gap, and cladding in typical light-water reactor (LWR) fuel rods. The instrumentation package for CHF-SERTTA focused on capturing those data that best describe the heat transfer behaviour between the rod and coolant. This paper outlines the CHFSERTTA experiment design and presents the initial results of the experiment campaign.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Structure and Process of Managing the UO2 Dry In-Pile Fracture Test Irradiation Experiment
The management and structure of a novel research and development (R&D) project can be hard to navigate. This is especially true when it involves nuclear fuel. Familiarity with institutional safety and procedural requirements, a well defined scope, and accurate cost estimates can all help ensure successful project management from start to finish. The Dry In-Pile Fracture Test (DRIFT) experiments performed at the Transient Test Reactor Facility are an excellent example of managing an R&D project within its scope and budget constraints. The objective of the DRIFT test was to develop fracture propagation data in a manner consistent with light-water reactors to validate and improve the fracture propagation models of uranium oxide fuels in the MOOSE (Multiphysics Object-Oriented Simulation Environment)-BISON-Marmot code framework [1].
99 GENERAL AND MISCELLANEOUS↗
Westinghouse Accident Tolerant Fuel Program
The Westinghouse accident tolerant fuel (ATF) program is aimed at providing a significant increase in fuel robustness able to increase its tolerance to beyond design basis accidents at an economically attractive price to nuclear power providers. The Westinghouse ATF program, which began in 2004, advanced significantly during the current first phase (1A and 1B) begun in 2012 of this Department of Energy (DOE) program. Phase 1B will culminate in the manufacture of fuel rods that will undergo a 6 year exposure at steady state conditions in the Advanced Test Reactor and the Halden test reactor (Phase 2) and in transient tests in TREAT and Halden. The goal is to load lead test rods in a commercial reactor in 2022. In this paper, the current status of the ATF development and future irradiation plan is discussed in detail. Westinghouse has picked the most viable concepts that will provide maximum accident tolerance as well as economics, and those that will make the most sense for utility adoption. All three options appear to be feasible from a technical and economic point of view. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Mo-100 Disk Structural Integrity Tests II
This report is a continuation of the work performed in the report, Mo-100 Disk Structural Integrity Tests I [1]. The NorthStar pressed and sintered 29 mm diameter, 0.5 mm thick Mo-100 disks will be subjected to large temperature gradients when in-beam during production. These large temperature gradients cause high stress and deformation within the disks that could prove detrimental to the target assembly’s cooling performance. Disk fracture is also a possibility that must be understood. Therefore, disk structural integrity tests were performed to understand the thermal induced behavior of these disks. In order to replicate the in-beam conditions, the Mo-100 disks were subjected to large temperature gradients using a 30kW induction heater and coil. A series of transient tests were conducted in a similar fashion as the previous work, as well as an attempt to create steady state conditions for more accurate temperature gradients. The transient thermal test setup and results are described in this report.
36 MATERIALS SCIENCE↗
Biodiesel Impact on Engine Lubricant Dilution During Active Regeneration of Aftertreatment Systems
Experiments were conducted with ultra low sulfur diesel (ULSD) and 20% biodiesel blends (B20) to compare lube oil dilution levels and lubricant properties for systems using late in-cylinder fuel injection for aftertreatment regeneration. Lube oil dilution was measured by gas chromatography (GC) following ASTM method D3524 to measure diesel content, by Fourier transform infrared (FTIR) spectrometry following a modified ASTM method D7371 to measure biodiesel content, and by a newly developed back-flush GC method that simultaneously measures both diesel and biodiesel. Heavy-duty (HD) engine testing was conducted on a 2008 6.7L Cummins ISB equipped with a diesel oxidation catalyst (DOC) and diesel particle filter (DPF). Stage one of engine testing consisted of 10 consecutive repeats of a forced DPF regeneration event. This continuous operation with late in-cylinder fuel injection served as a method to accelerate lube-oil dilution. Stage two consisted of 16 hours of normal engine operation over a transient test cycle, which created an opportunity for any accumulated fuel in the oil sump to evaporate. Light duty (LD) vehicle testing was conducted on a 2010 VW Jetta equipped with DOC, DPF and a NOx storage catalyst (NSC). Vehicle testing comprised approximately 4,000 miles of operation on a mileage-accumulation dynamometer (MAD) using the U.S. Environmental Protection Agency's Highway Fuel Economy Cycle because of the relatively low engine oil and exhaust temperatures, and high DPF regeneration frequency of this cycle relative to other cycles examined. Comparison of the lube oil dilution analysis methods suggests that D3524 does not measure dilution by biodiesel. The new back-flush GC method provided analysis for both diesel and biodiesel, in a shorter time and with lower detection limit. Thus all lube oil dilution results in this paper are based on this method. Analysis of the HD lube-oil samples showed only 1.5% to 1.6% fuel dilution for both fuels during continuous operation under DPF regeneration events. During the second stage of HD testing, the ULSD lube-oil dilution levels fell from 1.5% to 0.8%, while for B20, lube-oil dilution levels fell from 1.6% to 1.0%, but the fuel in the oil was 36% biodiesel. For the LD vehicle tests, the frequency of DPF regeneration events was observed to be the same for both ULSD and B20. No significant difference between the two fuels' estimated soot loading was detected by the engine control unit (ECU), although a 23% slower rate of increase in differential pressure across DPF was observed with B20. It appears that the ECU estimated soot loading is based on the engine map, not taking advantage of the lower engine-out particulate matter from the use of biodiesel. After 4,000 miles of LD vehicle operation with ULSD, fuel dilution in the lube-oil samples showed total dilution levels of 4.1% diesel. After 4,000 miles of operation with B20, total fuel in oil dilution levels were 6.7% consisting of 3.6% diesel fuel and 3.1% biodiesel. Extrapolation to the 10,000-mile oil drain interval with B20 suggests that the total fuel content in the oil could reach 12%, compared to 5% for operation on ULSD. Analysis of the oil samples also included measurement of total acid number, total base number, viscosity, soot, metals and wear scar; however, little difference in these parameters was noted.
09 BIOMASS FUELS↗
Advanced Reactor Fuels Irradiation Experiment Design Objectives
This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗