Multiscaled Experimental Investigations of Corrosion and Precipitation Processes After Loss-of-Coolant Accidents in Pressurized Water Reactors

The “Whole Assembly Seed and Blanket” (WASB) design, which utilizes mostly thorium in the blanket, consists of 84 seed and 109 blanket assemblies which may be backfitted into existing Pressurized Water Reactors (PWRs). Since the seed assemblies produce significantly more power than the blanket assemblies, a preliminary safety analysis of this design has been performed. Three accidents/transients (Large Break Loss of Coolant Accident (LBLOCA), Complete Loss of Primary Flow (LOPF) and Loss of Off-site Power (LOSP)), have been analyzed for both the WASB design and a typical all UO2 design for a typical 4-Loop Westinghouse PWR plant. LBLOCA results show that the peak cladding temperature (PCT) for the WASB design is approximately 260 K higher than that for a typical PWR design. However, this higher PCT for the WASB design is still about 200 K lower than the present regulatory safety limit. The response of the WASB and all UO2 core for LOPF and LOSP transients are very similar, and no post-DNB type rapid cladding temperature rise was observed in either of the two calculations.

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The Loss-of-Coolant Accident Analysis in Pressurized Water Reactors

Nuclear Science and Engineering ◽

10.13182/nse69-a19715 ◽

1969 ◽

Vol 36 (2) ◽

pp. 159-188 ◽

Cited By ~ 8

Author(s):

Amir N. Nahavandi

Keyword(s):

Pressurized Water Reactors ◽

Accident Analysis ◽

Pressurized Water ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Water Reactors

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Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

Nuclear Engineering and Design ◽

10.1016/0029-5493(89)90003-4 ◽

1989 ◽

Vol 110 (3) ◽

pp. 313-328 ◽

Cited By ~ 3

Author(s):

C.D. Fletcher ◽

R.A. Callow

Keyword(s):

Pressurized Water Reactors ◽

Pressurized Water ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Water Reactors

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Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2016.06.006 ◽

2016 ◽

Vol 305 ◽

pp. 489-502 ◽

Cited By ~ 1

Author(s):

André Seeliger ◽

Sören Alt ◽

Wolfgang Kästner ◽

Stefan Renger ◽

Holger Kryk ◽

...

Keyword(s):

Fluid Dynamic ◽

Pressurized Water Reactors ◽

Dynamic Effects ◽

Pressurized Water ◽

Loss Of Coolant ◽

Zinc Corrosion ◽

Water Reactors

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Experimental Studies on Critical Heat Flux in Tight Lattice Rod Bundles

Volume 2B: Thermal Hydraulics ◽

10.1115/icone22-30915 ◽

2014 ◽

Author(s):

Xuemei Lang ◽

Houjun Gong ◽

Lei Zhou ◽

Feng Xie ◽

Ye Liu

Keyword(s):

Heat Flux ◽

Critical Heat Flux ◽

Experimental Studies ◽

Helix Angle ◽

Pressurized Water Reactors ◽

Experimental Investigations ◽

Rod Bundles ◽

Pressurized Water ◽

Water Reactors ◽

Tight Lattice

The tight fuel lattice of pressurized water reactors (PWR) is helped to reduce the volume ratio of water-uranium, to increase the conversion ratio, to decrease the volume of core. It is especially useful for very high burnup and high volume power flux. The design of tight-lattice pressurized water reactors requires the knowledge of critical heat flux (CHF) in tight rod bundles. The tight hexagonal 19-rod bundles is used in this test. There are 4 wires wrapped in outside wall of each rod to support and locate. Experimental investigations on CHF behavior in the two kind bundles of helix angle 3° and 5° were performed. The CHF data points have been obtained in a range of parameters: pressure 8.0–16.6 MPa, mass flux 164.6–3283.0 kg/m2s and bundle exit steam quality −0.315 to 0.747. It is found that the CHF value of helix angle 5° bundle was more higher than that of helix angle 3° bundle in the same T/H condition. The effect of different parameters on CHF in the tight rod bundle is similar to that in the open literature. The CHF correlations of helix angle 5° bundle was obtained based on the test data.

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