Flow Testing and Analysis of the FSP-1 Experiment

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade R.; Weiss, Aaron; Howard, Trevor

doi:10.1115/nuclrf2017-3639

Cited by 1 publication

(3 citation statements)

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“…Computational and analytical models are used to evaluate the performance of the experiment. [4][5][6][7][8][9] In general, one-dimensional (1-D) codes and analytical formulas are employed at this stage. Upon determining feasibility, the experiment moves into the design phase, in which the hardware is detailed and the computational predictions are refined.…”

Section: Methodsmentioning

confidence: 99%

“…The FSP-1 experiment is a FSP experiment designed to investigate the performance of the fuel with a plate that is more representative of the final plate geometry. 8,9 This reduces the influence of edge peaking on the interior of the plate and allows for the evaluation of the fabrication process of the various configurations. The mock fuel plates in FSP-1 were fabricated with depleted uranium (DU) to emulate the performance of monolithic fuel plates.…”

Section: Ivd Fsp-1 Experimentsmentioning

confidence: 99%

“…The FSP-1 flow test campaign was used to progressively select the appropriate orifice size for the experiment outlet to achieve the targeted flow rate for the experiment. 8,9 Figure 21 shows the progression of system characteristic coolant velocity, from the initial SME reasonable value- Fig. 13.…”

Section: Vib Fsp-1 Flow Test Experimentsmentioning

confidence: 99%

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

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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, ...

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Section: Methodsmentioning

confidence: 99%

Section: Ivd Fsp-1 Experimentsmentioning

confidence: 99%