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Title: Reducing Uncertainty in Hydrodynamic Modeling of ATR Experiments Via Flow Testing, Validation, and Optimization

Abstract

The development, characterization, and qualification testing of nuclear fuel at Idaho National Laboratory’s Advanced Test Reactor (ATR) requires extensive design and analysis activities prior to the insertion of an irradiation experiment in-pile. Significant effort is made in the design and development phase of all in-pile experiments to ensure that the maximum feasible impacts of all necessary experimental requirements are satisfied. The advancement of fuel, cladding, and in-reactor materials technology in recent years has introduced complexities associated with the design and construct of in-pile experiments necessitating deeper understanding of boundary conditions and increasingly comprehensive observations resulting from the experiment. Each unique experiment must be assessed for neutronics response, thermal/hydraulic/hydrodynamic performance, and structural integrity. This is accomplished either analytically, computationally, or experimentally, or some combination thereof, prior to insertion into the ATR. The various effects are interrelated to various degrees, such as the case with the experiment temperature affecting the thermal cross section of the fuel or the increased temperature of the experiment’s materials reducing the mechanical strength of the assemblies. Additionally, the feedback between the experiment’s response to a reactor transient could alter the neutron flux profile of the reactor during the transient. Each experiment must therefore undergo a barrage ofmore » analyses to assure the ATR operational safety review committee that the insertion and irradiation of the experiment will not detrimentally affect the safe operational envelope of the reactor. In many cases, the nuclear fuel being tested can be double-encapsulated to ensure safety margins are adequately addressed, whereas failed fuel would be encased in a protective capsule. In other cases, the experiments can be inserted in a self-contained loop that passes through the reactor core, remaining isolated from the primary coolant. In the case of research reactor fuel, however, the fuel plates must be tested in direct contact with the reactor coolant, and being fuel designed for high neutron fluxes, they are inherently power-dense plates. The combination of plate geometry, high-power density, and direct contact with primary coolant creates a scenario where the neutronic/thermomechanic/hydrodynamic characteristics of the fuel plates are tightly coupled, necessitating as complete characterization as possible to support the safety and programmatic assessments, thus enabling a successful experiment. This paper explores the efforts of the U.S. High-Performance Research Reactor program to thermomechanically/hydromechanically characterize the program’s wide variety of experiments, which range from stacks of miniplate capsules to full-sized, geometrically representative curved plates. Special attention is given to instances where the combination of experimental characterization and analytical assessment has reduced uncertainties of the safety margins, allowing experiments to be irradiated that would otherwise not have passed the rigorous qualification process for irradiation in the ATR. In some cases, the combined processes have exposed flow and heat transfer characteristics that would have been missed using historical methods, which allows for more accurate and representative postirradiation assessments.« less

Authors:
ORCiD logo [1];  [2];  [3]; ORCiD logo [1]; ORCiD logo [4];  [1]; ORCiD logo [1]; ORCiD logo [1]; ORCiD logo [1]; ORCiD logo [1];  [1]; ORCiD logo [1];  [3];  [3];  [5]
+ Show Author Affiliations
  1. Idaho National Laboratory (INL), Idaho Falls, ID (United States)
  2. Idaho National Laboratory (INL), Idaho Falls, ID (United States); Oregon State Univ., Corvallis, OR (United States)
  3. Oregon State Univ., Corvallis, OR (United States)
  4. Idaho National Laboratory, Nuclear Science and Technology, Experiment Analysis, 2525 Fremont Avenue, Idaho Falls, Idaho 83415
  5. Idaho National Lab. (INL), Idaho Falls, ID (United States)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1556996
Report Number(s):
INL/JOU-17-43172-Rev000
Journal ID: ISSN 0029-5450
Grant/Contract Number:  
AC07-05ID14517
Resource Type:
Accepted Manuscript
Journal Name:
Nuclear Technology
Additional Journal Information:
Journal Volume: 201; Journal Issue: 3; Journal ID: ISSN 0029-5450
Publisher:
Taylor & Francis - formerly American Nuclear Society (ANS)
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; special issue; Nuclear Technology; ATR 50th Anniversary; hydraulic testing; Nuclear Fuels Characterization; Fluid-Structure Interaction

Citation Formats

Jones, Warren F., Marcum, Wade R., Weiss, A. W., Jensen, Colby B., Hawkes, Grant L., Murray, Paul E., Crawford, Douglas S., Herter, Justin W., Kennedy, John C., Woolstenhulme, Nicolas E., Wiest, James D., Chapman, Daniel B., Howard, T. K., Latimer, G. D., and Phillips, Ann Marie. Reducing Uncertainty in Hydrodynamic Modeling of ATR Experiments Via Flow Testing, Validation, and Optimization. United States: N. p., 2018. Web. doi:10.1080/00295450.2017.1407907.
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Jones, Warren F., Marcum, Wade R., Weiss, A. W., Jensen, Colby B., Hawkes, Grant L., Murray, Paul E., Crawford, Douglas S., Herter, Justin W., Kennedy, John C., Woolstenhulme, Nicolas E., Wiest, James D., Chapman, Daniel B., Howard, T. K., Latimer, G. D., & Phillips, Ann Marie. Reducing Uncertainty in Hydrodynamic Modeling of ATR Experiments Via Flow Testing, Validation, and Optimization. United States. https://doi.org/10.1080/00295450.2017.1407907
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Jones, Warren F., Marcum, Wade R., Weiss, A. W., Jensen, Colby B., Hawkes, Grant L., Murray, Paul E., Crawford, Douglas S., Herter, Justin W., Kennedy, John C., Woolstenhulme, Nicolas E., Wiest, James D., Chapman, Daniel B., Howard, T. K., Latimer, G. D., and Phillips, Ann Marie. Thu . "Reducing Uncertainty in Hydrodynamic Modeling of ATR Experiments Via Flow Testing, Validation, and Optimization". United States. https://doi.org/10.1080/00295450.2017.1407907. https://www.osti.gov/servlets/purl/1556996.
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@article{osti_1556996,
title = {Reducing Uncertainty in Hydrodynamic Modeling of ATR Experiments Via Flow Testing, Validation, and Optimization},
author = {Jones, Warren F. and Marcum, Wade R. and Weiss, A. W. and Jensen, Colby B. and Hawkes, Grant L. and Murray, Paul E. and Crawford, Douglas S. and Herter, Justin W. and Kennedy, John C. and Woolstenhulme, Nicolas E. and Wiest, James D. and Chapman, Daniel B. and Howard, T. K. and Latimer, G. D. and Phillips, Ann Marie},
abstractNote = {The development, characterization, and qualification testing of nuclear fuel at Idaho National Laboratory’s Advanced Test Reactor (ATR) requires extensive design and analysis activities prior to the insertion of an irradiation experiment in-pile. Significant effort is made in the design and development phase of all in-pile experiments to ensure that the maximum feasible impacts of all necessary experimental requirements are satisfied. The advancement of fuel, cladding, and in-reactor materials technology in recent years has introduced complexities associated with the design and construct of in-pile experiments necessitating deeper understanding of boundary conditions and increasingly comprehensive observations resulting from the experiment. Each unique experiment must be assessed for neutronics response, thermal/hydraulic/hydrodynamic performance, and structural integrity. This is accomplished either analytically, computationally, or experimentally, or some combination thereof, prior to insertion into the ATR. The various effects are interrelated to various degrees, such as the case with the experiment temperature affecting the thermal cross section of the fuel or the increased temperature of the experiment’s materials reducing the mechanical strength of the assemblies. Additionally, the feedback between the experiment’s response to a reactor transient could alter the neutron flux profile of the reactor during the transient. Each experiment must therefore undergo a barrage of analyses to assure the ATR operational safety review committee that the insertion and irradiation of the experiment will not detrimentally affect the safe operational envelope of the reactor. In many cases, the nuclear fuel being tested can be double-encapsulated to ensure safety margins are adequately addressed, whereas failed fuel would be encased in a protective capsule. In other cases, the experiments can be inserted in a self-contained loop that passes through the reactor core, remaining isolated from the primary coolant. In the case of research reactor fuel, however, the fuel plates must be tested in direct contact with the reactor coolant, and being fuel designed for high neutron fluxes, they are inherently power-dense plates. The combination of plate geometry, high-power density, and direct contact with primary coolant creates a scenario where the neutronic/thermomechanic/hydrodynamic characteristics of the fuel plates are tightly coupled, necessitating as complete characterization as possible to support the safety and programmatic assessments, thus enabling a successful experiment. This paper explores the efforts of the U.S. High-Performance Research Reactor program to thermomechanically/hydromechanically characterize the program’s wide variety of experiments, which range from stacks of miniplate capsules to full-sized, geometrically representative curved plates. Special attention is given to instances where the combination of experimental characterization and analytical assessment has reduced uncertainties of the safety margins, allowing experiments to be irradiated that would otherwise not have passed the rigorous qualification process for irradiation in the ATR. In some cases, the combined processes have exposed flow and heat transfer characteristics that would have been missed using historical methods, which allows for more accurate and representative postirradiation assessments.},
doi = {10.1080/00295450.2017.1407907},
journal = {Nuclear Technology},
number = 3,
volume = 201,
place = {United States},
year = {2018},
month = {1}
}
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https://doi.org/10.1080/00295450.2017.1407907
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