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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Scoping Study of Effects on the Advanced Test Reactor’s Center Flux Trap Due to Various Loadings in the H and Inner A Test Positions

This study determined the effects experienced by the In-Pile Tube, of the Advanced Test Reactor’s center flux trap, due to various loading configurations of the H and Inner A positions. The effects of concern included: the axial profile of the linear heat generation rate of each fuel rodlet, the axial flux shape, and spectrum, in the fuel, and cladding, of each rodlet, and the total fission heating rate of the entire In-Pile Tube. Additionally, the reactivity worth of each configuration, relative to an ATR configuration with water filling all concerning test positions, was determined. The test positions were filled with commonly used test specimens, backup test specimens, and homogeneous substances. The study was performed using the modeling software MCNP6.2.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Sensitivity of ATF Experiments in the Center Flux Trap of the Advanced Test Reactor to Adjacent Experiments

Irradiation experiments conducted in the Idaho National Laboratory’s Advanced Test Reactor are typically assumed to have little effect on one another. This assumption does not hold true for certain experiments in close proximity. To evaluate the impacts on safety and programmatic parameters of experiments in the center flux trap, the contents of the adjacent H and inner-A positions were modeled with a range of possible irradiation targets. First, neutron flux maps with experiments in those positions were compared against a baseline configuration. Next, several safety and programmatic parameters for a generic accident-tolerant fuel test train were calculated. It was shown that these parameters can exhibit considerable sensitivity to the contents of the H and A experiment positions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

The Accident Tolerant Fuels’ Neutronics Scoping Study (ATF-2Pert)

This summary discusses an ongoing scoping study to better understand the neutronic characteristics surrounding the center flux trap of the advanced test reactor. This summary details the need for the study, the means by which it was carried out, the study's results, and how the results benefit the accident tolerant fuels program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

As-run physics analysis for the EPRI-2 experiment for cycle 153B

The purpose of this ECAR is to document the as-run physics analysis and source term in Curies and decay heat rate (Watts) for the shipment of the EPRI-2 experiment located in the Center Flux Trap (CFT). EPRI-2 was irradiated during cycle 153B. The experiment irradiation ended on April 13, 2013. The as-run heat rates and flux/fluence results were calculated using the MCNP ATR full core model. The heat rates and flux are calculated based on an average center lobe power of 30.8 MW. The fluence is based on the operating time of 13.45 days. ORIGEN2 Version 2.2 is used to determine the decay heat rate, gamma spectrum, and source term for the EPRI-2 experiment. The source term analysis was performed as requested by the project staff to demonstrate compliance with shipping requirements in the GE-2000 cask following irradiation as well as provide an estimated source term to support Post-Irradiation Examination as needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

As-Run Physics Analysis for the EPRI-1 Experiment

The purpose of this ECAR is to document the as-run physics analysis for the Electric Power Research Institute (EPRI)-1 experiment located in the Center Flux Trap (CFT). EPRI-1 was irradiated during cycle 157C. The experiment irradiation ended on February 15, 2015. The as-run heat rates, DPA, and flux/fluence results were calculated using the Monte Carlo N-Particle (MCNP) Advanced Test Reactor (ATR) full core model. The heat rates, DPA, and flux are calculated based on an average center lobe power of 21.1 MW. The fluence and DPA is based on the operating time of 5.4 days. ORIGEN2 Version 2.2 was used to calculate the isotopic concentration in grams for the EPRI-1 experiment. The source term analysis was performed as requested by the project staff. The results of this analysis may be used to support Post-Irradiation Examination (PIE)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

ATF-2 Ramp Conceptual Design Report

The following report documents the conceptual design for the ATF-2 Ramp experiment. The experiment aims to fill the in-pile irradiation testing gap to conduct integral ramp testing, which was created by the closure of R2, Osiris, and Halden test reactors. The concept involves the simultaneous ramping of three fuel pins using a power axial locator mechanism in the Loop-2A testing facility in the center flux trap of the advanced test reactor (ATR). The three fuel pins are ramped in individual coolant channels containing a prototypic PWR environment. Rod failure will be detected using a fuel rod elongation sensor attached to the fuel pin upper end cap. An axial stack of concentric hafnium and zirconium shrouds will be used to shape the flux around the test pins to create different power levels in each pin and to ensure the peak power location of each pin remains in the center of the test rod. Monte Carlo simulations are used to demonstrate the viability of this design concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Depletion benchmark for a high-assay low-enriched uranium fuel experiment in the advanced test reactor

Reactor physics depletion benchmarks for high-assay low-enriched uranium (HALEU) fuel are limited in number. In particular, there is limited data for HALEU benchmarks for U-10Mo (uranium-10% molybdenum) plate fuel that is being developed for use in the United States’ high performance research reactors including the Advanced Test Reactor (ATR), Advanced Test Reactor Critical Facility (ATR-C), High Flux Isotope Reactor (HFIR), Massachusetts Institute of Technology Reactor (MITR), University of Missouri Research Reactor (MURR), National Bureau of Standards Reactor (NBSR). These six reactors currently operate with highly enriched uranium dispersed fuel in an aluminum matrix. In support of conversion to a HALEU fuel, qualification of U-10Mo formed into a monolithic foil is being performed. Fuel qualification involves irradiating fuel specimens in the ATR. The irradiation tests provide an opportunity to benchmark depletion capabilities of reactor physics codes in support of the ATR operation, as well as develop benchmarks that can be used by other institutions to benchmark other reactor physics codes. This paper documents the development of a benchmark model of the irradiation of the ATR Full-size plate In center flux trap Position 7 (AFIP-7) experiment using the depletion codes MC21 and Advanced Dimensional Depletion for Engineering of Reactors (ADDER).

Nielsen, Joseph W. [Idaho National Laboratory (INL↗
DOE OSTI2023

Accident Tolerant Fuel Test (ATF-2c) Irradiation Test Report

A new fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2C has completed its first cycle of prototypic steady state irradiation. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and fast neutron flux histories of each of the 24 test pins has been calculated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Accident Tolerant Fuel Test 2B (ATF-2B) Irradiation Test Report

A fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2 has completed 7 cycles of prototypic steady state irradiation irradiating 41 test pins to burnups as high as 30 MWd/kgU. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and fast neutron flux histories of each of the 41 test pins has been calculated using a combined MCNP, and ORIGIN methodology to provide accurate and reliable power histories on a per pin basis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Depletion Benchmark of the AFIP-7 Experiment in the Advanced Test Reactor

Reactor physics depletion benchmarks for low-enriched uranium fuel are limited in number. In particular, there is very limited data for LEU benchmarks for U-10Mo (Uranium-10% Molybdenum) plate fuel developed for use in U.S. high-performance research reactors (USHPRR). USHPRR includes the Advanced Test Reactor (ATR), Advanced Test Reactor Critical Facility (ATR-C), High Flux Isotope Reactor (HFIR), University of Missouri Research Reactor (MURR), Massachusetts Institute of Technology Reactor (MITR), and National Bureau of Standards Reactor (NBSR) at the National Institute of Science and Technology. These reactors are fueled with high-enriched uranium dispersed fuel in a silicon/aluminum matrix. In support of conversion to a HALEU fuel, qualification of U-10Mo formed into a monolithic foil is being performed. Fuel qualification involves irradiated fueled specimens in the ATR. The irradiation tests provide an opportunity to benchmark depletion capabilities of reactor physics codes in support of the ATR operation, as well as develop benchmarks that can be used by other institutions to benchmark other reactor physics codes. This report documents the development of a benchmark model of the irradiation of the ATR Full -size plate In center flux trap Position 7 (AFIP-7) experiment.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Accident tolerant fuel test 2c (ATF-2C) irradiation test report

A fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2C has completed its four cycles of prototypic steady state irradiation. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and burnup histories of each of the 18 fueled test pins has been calculated.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

ANS Winter 2022: ATF-2C Physics Safety and Scoping Analysis

ATF-2C comprises four tiers with six rodlets each. The experiment test train (TT) is situated in Loop 2A in the center flux trap (CFT) of the ATR, in which typical pressurized water reactor (PWR) conditions shall be approximated. Tier 1 (bottom) of the TT tests Framatome silicon carbide (SiC) cladding, containing molybdenum heater rods. Low-enriched uranium (LEU) Framatome rodlets with chrome-coated M5 cladding have been modeled in their place for the physics safety analysis to allow the substitution of this tier with fuel if needed. Tier 2 tests LEU Framatome rodlets with M5 cladding, which had irradiated in ATR in prior ATF experiments. Tier 3 tests General Atomics SiC composite cladding, also containing molybdenum heater rods. Tier 4/5/6 (top) tests the Japan Atomic Energy Agency (JAEA)/Mitsubishi Heavy Industries (MHI) fueled rodlets. This upper tier tests LEU rodlets with chrome-coated MDA cladding—several with temperature or pressure instrumentation. To flatten the axial power profile, lower portion of Tier 4/5/6 is surrounded by a hafnium (Hf) shroud, and the bottom two UO2 pellets are of natural enrichment. The ATF-2C TT is to be irradiated in the CFT of ATR (Fig. 1) for three or four 60-day cycles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Effect of Water Chemistry on Crack Growth Rate in Neutron Irradiated X-750 and XM-19

The Advanced Test Reactor (ATR) National Science User Facility (NSUF) was used to study the effect of neutron damage on crack growth rates (CGR) under simulated boiling water reactor (BWR) water chemistry conditions using a fast neutron spectrum to accelerate the damage process. Nickel based Alloy X-750 and XM-19 were irradiated in the ATR center flux trap (CFT), with this experiment marking the first civilian project to utilize this position within the reactor. 0.4T-CT specimens were used to study crack growth rates for two sample fluences (~1.9 x 1020 n/cm2 and ~9.5 x 1020 n/cm2, E > 1MeV)) in both normal water chemistry (NWC) at 2.5 ppm O2 and hydrogen water chemistry (HWC) at ~80 ppb H2. CGR testing was conducted at a temperature of 288 °C. A decrease in CGR was observed for both alloys with the average ratio (CGR in NWC/CGR in HWC) being a factor of 281 for the XM-19 samples and a factor of 27.8 for the X-750 samples. Observed CGRs in this study showed little difference compared to non-irradiated samples (the non-irradiated XM-19 was 20% cold worked), suggesting limited sensitivity of these susceptible alloys to irradiation effects. Detailed results of CGR response to varying environmental electrochemical potential (ECP) and water purity will be presented in the context of crack growth mitigation in different water chemistries and the change in material behavior relative to non-irradiated samples of the material.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effect of Water Chemistry on Crack Growth Rates in Neutron Irradiated X-750 and XM-19

The Advanced Test Reactor (ATR) National Science User Facility (NSUF) was used to study the effect of neutron damage on crack growth rates (CGR) under simulated boiling water reactor (BWR) water chemistry conditions using a fast neutron spectrum to accelerate the damage process. Nickel based Alloy X-750 and XM-19 were irradiated in the ATR center flux trap (CFT), with this experiment marking the first civilian project to utilize this position within the reactor. 0.4T-CT specimens were used to study crack growth rates for two sample fluences (~1.9 x 1020 n/cm2 and ~9.5 x 1020 n/cm2, E > 1MeV) in both normal water chemistry (NWC) at 2.5 ppm O2 and hydrogen water chemistry (HWC) at ~80 ppb H2. A decrease in CGR was observed with HWC for both alloys with the average ratio (CGR in NWC/CGR in HWC) being a factor of 281 for the XM-19 samples and a factor of 27.8 for the X-750 samples. Observed CGRs in this study showed little difference compared to non-irradiated samples (the non-irradiated XM-19 was 20% cold worked), suggesting limited sensitivity of these susceptible alloys to irradiation effects. The CGR response to varying environmental electrochemical potential (ECP) and water purity will be presented.

36 MATERIALS SCIENCE↗
DOE OSTI2023

ATF-2 Sensitivity M&C 2023 (Slides)

Irradiation experiments in the Idaho National Laboratory's Advanced Test Reactor are typically assumed to have little effect on one another. This assumption does not hold true for certain experiments in close proximity. To evaluate the impacts on safety and programmatic parameters of experiments in the center flux trap, the contents of the adjacent H and inner-A positions are modeled with a range of possible irradiation targets. First, neutron flux maps with experiments in those positions are compared to those of a water-filled configuration. Then, several safety and programmatic parameters for a generic accident-tolerant fuel test train are calculated. It is shown that these parameters can exhibit considerable sensitivity to the contents of the H and A experiment positions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Conceptual Neutronics Scoping for Ramp Testing in the Advanced Test Reactor

With the closer of the Halden Boiling Water Reactor in 2018, the global capacity for in-reactor power ramp testing has been lost. As such tests provide valuable data for understanding pellet-cladding interaction (PCI) phenomena, finding new facilities for ramp testing is of interest to the United States Accident-Tolerant Fuel (ATF) Program. Several options at Idaho National Laboratory (INL) are being evaluated for adding ramp test capabilities. In the Transient Reactor Test Facility (TREAT), one option is testing inside a Transient Water Irradiation System in TREAT (TWIST) capsule. In the Advanced Test Reactor (ATR), one option under consideration is testing in an I-Loop toward the outer edge of the core. Another is testing in Loop-2A, located in the Center Flux Trap (CFT) of the ATR. In this paper, we examine the conceptual scoping for the Loop-2A experiment known as ATF-2Ramp.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Conceptual Neutronics Scoping for Ramp Testing in the Advanced Test Reactor

With the closure of the Halden Boiling Water Reactor in 2018, the global capacity for in-reactor power ramp testing has been lost. As such tests provide valuable data for understanding pellet-cladding interaction (PCI) phenomena, finding new facilities for ramp testing is of interest to the United States Accident-Tolerant Fuel (ATF) Program. Several options at Idaho National Laboratory (INL) are being evaluated for adding ramp test capabilities. In the Transient Reactor Test Facility (TREAT), one option is testing inside a Transient Water Irradiation System in TREAT (TWIST) capsule. In the Advanced Test Reactor (ATR), one option under consideration is testing in an I-Loop toward the outer edge of the core. Another is testing in Loop-2A, located in the Center Flux Trap (CFT) of the ATR. In this paper, we examine the conceptual scoping for the Loop-2A experiment known as ATF-2Ramp.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

ANS Winter 2024 Summary: Optimizing the ATF-2Ramp Power Profile

When the Halden Boiling Water Reactor closed down in 2018, a need to restore the capability for in-reactor power ramp testing arose. Such testing is valuable for studying pellet-clad interaction phenomena in nuclear fuels. The data from these studies is of great interest to a number of research programs, including the accident-tolerant fuel (ATF) program at Idaho National Laboratory (INL). In 2022, Woolstenhulme et al. proposed several power ramp testing ideas using facilities at INL, including irradiation in the Transient Reactor Test Facility (better known as TREAT) and the Advanced Test Reactor (ATR) [1]. Worrall et al. [2] and Labossiere-Hickman et al. [3] subsequently performed feasibility studies for the ATR testing options in 2023. This summary further investigates the three-pin trefoil design (Fig. 1) for the proposed ATF-2Ramp Experiment discussed in Labossiere-Hickman et al. [3]. ATF-2Ramp is designed to operate in the center flux trap (CFT) of the ATR during a powered axial locator mechanism (PALM) cycle: a short, variable-powered cycle with an asymmetric power distribution. Previously, it was shown that tailoring the thickness of the hafnium (Hf) neutron shields (“mini-shrouds”) surrounding each pin offered a degree of control sufficient to achieve the programmatic linear heat generation rate (LHGR) targets for ATF-2Ramp during the high-power period of a PALM cycle. New work involves shortening the experiment test train for consistency with the fuel pins in ATF-2D [4] and then shaping the axial power profile of the three test pins.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗