Contributions from the ACRR in-pile experiments to the understanding of key phenomena influencing unprotected loss of flow accident simulations in LMFBRs

Royl, P.; Breitung, W.; Fischer, E.A.; Schumacher, G.; Gauntt, Randall O.; Wright, Steven A.

doi:10.1016/0029-5493(87)90088-4

Cited by 6 publications

(3 citation statements)

References 5 publications

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“…Condition (4) rules out instantaneous explosive vaporization upon contact, 6) Condition (5) rules out sodium reentry, 13) and Condition (6) eliminates significant coolant entrapment as the fuel displaces the coolant. 14) These general characteristics are consistent with experimental observations [10][11][12][13] and are especially important in ruling out energetic recriticality events, i.e., eliminating the potential for reaching super prompt critical conditions from fuel compaction mechanisms driven by fuel coolant interactions between dispersed fuel and liquid sodium near the top and bottom of the active core. Condition (7) prevents significant in-core direct fuel coolant contact, i.e., the fuel and the coolant within the core is largely separated by solid cladding.…”

Section: Subassembly Blockage Potentialsupporting

confidence: 71%

“…9) Posttest examinations of numerous inpile fuel disruption tests for both fresh fuel and pre-irradiated fuel have shown extensive mixing of solidified steel within solidified fuel remains. [10][11][12] The general character of the fuel remains from all the loss-of-flow simulation tests resembles that of a porous mass of fuel with fairly large voids and with considerable entrainment of globules of steel. The resulting vaporization of entrapped steel drove fuel debris to the ends of the fuel region, i.e.…”

Section: The Role and Application Of General Behavior Princeiplesmentioning

confidence: 99%

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Assessment of the FBR Core Disruptive Accident (CDA): The Role and Application of General Behavior Principles (GBPs).

Fauske¹,

Koyama²,

Kubo

2002

J. Nucl. Sci. Technol.

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General behavior principles (GBPs) first introduced in 1976 (H. K. Fauske, the Role of Core-Disruptive Accidents in Design and Licensing of LMFBRs, Nuclear Safety, Vol. 17, No. 5, 550-567, 1976) are applied to assess the core disruptive accident (CDA) outcome for a large (3,500 MWt) high power density oxide-fueled sodium-cooled fast breeder reactor (FBR). Non-energetic termination in the so-called Initiating Phase is emphasized and is made possible by considering departures from the traditional core design including the Subassembly Inner Duct and Limited Blanket Removal concepts. This early accident scenario termination eliminates potential concerns raised related to recriticality events and also facilitates the potential for in-vessel fuel debris coolability for the large FBR with a compact reactor vessel.

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Section: Subassembly Blockage Potentialsupporting

confidence: 71%

Section: The Role and Application Of General Behavior Princeiplesmentioning

confidence: 99%

Assessment of the FBR Core Disruptive Accident (CDA): The Role and Application of General Behavior Principles (GBPs).

Fauske¹,

Koyama²,

Kubo

2002

J. Nucl. Sci. Technol.

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“…Pulse width is a critical parameter as it affects the timing of pellet-to-cladding heat transfer, and the resulting temperature-dependent mechanical properties, along with affecting strain rates during the mechanical interaction. Recent simulations suggest that peak cladding hoop stress can be more than doubled in a pulse 30 ms versus a slightly wider pulse 90 ms FWHM (Folsom, et al, 2016 (Cabrillat, et al, 2003), Japanese Nuclear Safety Research Reactor (NSRR) (Fuketa, et al, 2006), the United States (US) Annual Core Research Reactor (ACRR) (Royl, et al, 1987), Impulse Graphite Reactor (IGR) in Kazakhstan (Payot, et al, 2017), and the US Power Burst Facility (PBF) (Petti, et al, 1994 and forthcoming LWRenvironment capsules/loops for TREAT (Woolstenhulme, et al, 2016b). While CABRI's innate FWHM is approximately 10 ms, it can be "stretched" by continuing to insert reactivity after the initial step-up to achieve 20 -80 ms (Clamens, et al, 2016).…”

Section: Transient Testing To Support Lwr Accident Simulationsmentioning

confidence: 99%

Narrowing transient testing pulse widths to enhance LWR RIA experiment design in the TREAT facility

Bess

Woolstenhulme

Davis

et al. 2019

Annals of Nuclear Energy

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The Transient Reactor Test (TREAT) Facility provided thousands of transient irradiations and plays a crucial role in nuclear-heated safety research. The facility's flexible design and multi-mission nature saw historic experiments for numerous reactor fuels and transient types but was never specifically adapted to address very-brief pulse transients akin to postulated Light Water Reactor (LWR) Reactivity Initiated Accidents (RIA). Since the behaviors of fuel under these conditions depends strongly on energy input duration and resulting cladding time/temperature response under pellet cladding mechanical interactions, this three-year project was conceived to investigate new pulse-narrowing capabilities. Kinetic models showed incremental improvements for minor facility enhancements including increased reactivity step insertions (to initiate the power pulse) and slightly-increased transient rod drive speed for pulse termination ("clipping"). Replacing peak fuel assemblies, repositioning non-transient control rods to hold down the "hot side" of the core, and balancing against required excess reactivity needs, the limiting fuel assembly power can be reduced by ~20 %. This is a remarkable discovery of latent capability in TREAT, not only in enabling the subject capabilities for reduced pulse widths, but also significantly "uprating" the core's transient energy capacity. Incremental improvements can likely enhance TREAT's capability into BWR-relevant missions; briefer PWR pulse shapes can only be achieved with an advanced clipping system. Helium-3 ( 3 He) was found to offer the greatest benefits for clipping design and overall performance. A unique concept showed great promise for enabling a 3 He-based system with a reasonable cost; this design concept will likely become the focal point of future work.

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Development of Advanced Instrumentation for Transient Testing

Jensen

Fleming

2019

Nuclear Technology

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Contributions from the ACRR in-pile experiments to the understanding of key phenomena influencing unprotected loss of flow accident simulations in LMFBRs

Cited by 6 publications

References 5 publications

Assessment of the FBR Core Disruptive Accident (CDA): The Role and Application of General Behavior Principles (GBPs).

Assessment of the FBR Core Disruptive Accident (CDA): The Role and Application of General Behavior Principles (GBPs).

Narrowing transient testing pulse widths to enhance LWR RIA experiment design in the TREAT facility

Development of Advanced Instrumentation for Transient Testing

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