Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

Brunett, Acacia J.; Fei, T.; Strons, Philip; Papadias, Dionissios D.; Hoffman, E. A.; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, A.E.

doi:10.2172/1224563

Cited by 2 publications

(10 citation statements)

References 8 publications

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“…TREAT was operated for about 35 years using the original HEU fuel. The estimated average annual energy production in the TREAT HEU core was 350,000 MJ, corresponding to ~1.1×10 22 fissions per year [27]. The average annual burnup per unit volume and core-average neutron fluence were estimated to be ~2.65×10 -2 KWH/cm 3 and ~2.21×10 18 neutrons/cm 2 (fast neutron fluence: ~4.13×10 17 neutrons/cm 2 ), respectively.…”

Section: Reactor Radiation Damage During a Long-term Operation Of Treatmentioning

confidence: 99%

Effect of reactor radiation on the thermal conductivity of TREAT fuel

Miao

Kontogeorgakos

et al. 2017

Journal of Nuclear Materials

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The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

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Section: Reactor Radiation Damage During a Long-term Operation Of Treatmentioning

confidence: 99%

Effect of reactor radiation on the thermal conductivity of TREAT fuel

Miao

Kontogeorgakos

et al. 2017

Journal of Nuclear Materials

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show abstract

“…Because of the large inlet plenum above the core, it is reasonably assumed that the flow is distributed equally among all assemblies. The majority of the flow that passes through the core, approximately 96%, is distributed to coolant passages formed by the chamfered corners of neighboring assemblies [4]; the remaining 4% passes along the small gaps between assembly sides. During the design basis accident, in which there is an accidental insertion of reactivity, cooling by the filtration/cooling system (F/CS) is assumed to be unavailable, and natural convection is also neglected.…”

Section: Coolingmentioning

confidence: 99%

“…Numerous safety analyses have been completed during the LEU design process [4] with the objective to perform best estimate, yet bounding, calculations so that these analyses will not need to be repeated as each LEU design phase is completed. Of the analyses completed in this conceptual phase for previous LEU designs, several are still considered bounding for the LEU conceptual design.…”

Section: Previously-performed Bounding Analysesmentioning

confidence: 99%

“…This includes the analysis of the dose rate at the HEPA filters due to tramp uranium fission products and heating of the radial reflector, both of which are discussed below. Additional analysis details can be found in the previous report [4].…”

Section: Previously-performed Bounding Analysesmentioning

confidence: 99%

“…Neutronic and thermal hydraulic simulations have been performed in order to assess the LEU conceptual design performance under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, the ancillary safety analyses which were performed for previous LEU designs [4] have been reviewed, and the calculations have been updated as needed. To support the fuel manufacturing studies being performed by INL and LANL, analysis of heat transfer from the fuel UO 2 to the fuel's graphite matrix and of changes in fuel conductivity with burnup have also been performed.…”

Section: Introductionmentioning

confidence: 99%

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