Intrinsic effects of Cr-layered accident-tolerant fuel cladding surface on reflood heat transfer

Author(s):  
Doyoung Shin ◽  
Sung Joong Kim
Keyword(s):  
Author(s):  
Pablo E. Araya Go´mez ◽  
Miles Greiner

A two-dimensional computational model of a spent 7×7 Boiling Water Reactor assembly in a horizontal support basket was developed using the Fluent computational fluid dynamics package. Heat transfer simulations were performed to predict the maximum cladding temperature for assembly heat generation rates between 100 and 600W, uniform basket wall temperatures of 25 and 400°C, and with helium and nitrogen backfill gases. Different sets of simulations modeled conduction/radiation and natural convection/radiation transport across the gas filled regions to assess the importance of different transport processes. Simulations that included natural convection exhibited measurably lower cladding temperatures than those that did not only for nitrogen, at the lower basket wall temperature, and within an intermediate range of heat generation rates. Outside these conditions and for helium, conduction and radiation transport are sufficiently large so that natural convection has no measurable effect. Finally, the maximum cladding temperature is more sensitive to the assumed value of the fuel cladding emissivities when nitrogen is the backfill gas than when helium is used.


2016 ◽  
Vol 4 ◽  
pp. 8 ◽  
Author(s):  
Vojtěch Caha ◽  
Jakub Krejčí

The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF) is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature). The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.


2019 ◽  
Vol 129 ◽  
pp. 375-389 ◽  
Author(s):  
Jun-young Kang ◽  
Tong Kyun Kim ◽  
Gi Cheol Lee ◽  
HangJin Jo ◽  
Moo Hwan Kim ◽  
...  

Author(s):  
Pablo E. Araya ◽  
Miles Greiner

The current work is a scoping study to determine which heat transfer effects are significant in the fuel/backfill gas region of spent nuclear fuel transport casks. A two-dimensional finite volume mesh that accurately models the geometry of a 7×7 Boiling Water Reactor (BWR) assembly with its channel in a square isothermal enclosure is constructed. The peak cladding temperature is determined using computational fluid dynamics (CFD) simulations for a range of enclosure temperatures, fuel heat generation rates, cladding surface emissivities, and for both nitrogen and helium backfill gases. This work quantifies both the effect of buoyancy induced gas motion in the fuel/backfill gas region and the conditions when it does not significantly affect heat transfer. Future cask design simulations that neglect gas motion will require less computational resources than ones that do not. This work also quantifies the sensitivity of the maximum cladding temperature to fuel cladding emissivity. This helps quantify the uncertainty of temperature predictions if the emissivity is not known. The current CFD technique must be experimentally benchmarked before it may be used with confidence to predict peak cladding temperatures in transport casks. This work indicates that the thermal resistance between a BWR assembly’s channel and the basket walls may be modeled analytically. This will reduce the effort required for benchmark experiments because they will not need to include the channel.


2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Olugbenga O. Noah ◽  
Johan F. Slabber ◽  
Josua P. Meyer

The ability of coated particles of enriched uranium dioxide fuel encased in graphite to discontinue nuclear fission reaction without human action in the case of complete loss of cooling is a vital safety measure over traditional nuclear fuel. As a possible solution toward enhancing the safety of light water reactors (LWRs), it is envisaged that the fuel, in the form of loose, coated particles in a helium atmosphere, can be used inside the cladding tubes of the fuel elements. This study is therefore a first step toward understanding the heat-transfer characteristics under natural convective conditions within the fuel cladding tubes of such a revolutionary new fuel design. The coated particle fuels are treated as a bed, from which the heat is transferred to the cladding tube and the gas movement occurs due to natural convection. A basic unit cell model was used where a single unit of the packed bed was analyzed and taken as representative of the entire bed. The model is a combination of both analytical and numerical methods accounting for the thermophysical properties of sphere particles, the interstitial gas effect, gas temperature, contact interface between particles, particle size, and particle temperature distribution used in this study to investigate the heat-transfer effect. The experimental setup was a packed bed heated from below with gas circulation due to natural convection. This allows for the development of an appropriate, conservative thermal energy balance that can be used in determining the heat-transfer characteristics in homogeneous porous media. Success in this method, when validated with suitable correlation, such as Gunn, suggests that the heat-transfer phenomenon/characteristics in the fuel cladding tube of the new design can be evaluated using this approach for design purpose.


Author(s):  
Bing Ren ◽  
Fujun Gan ◽  
Yu Dang ◽  
Libing Zhu

Corrosion products on fuel cladding surface have a significant impact on reactor operation. These types of deposits are defined as Corrosion Residual Unidentified Deposit (CRUD) and are consist of a porous matrix of nickel and iron based oxides deposited on the fuel cladding surface. It is well known that crud deposits may cause potential Crud Induced Localized Corrosion (CILC) risk and Crud Induced Power Shift (CIPS) risk. The paper presents a Computational Fluid Dynamic (CFD) method of predicting the crud effect on the thermal hydraulic performance. The effect of the crud roughness is mainly considered in the simulation, the flow near the wall of the crud is solved by modifying wall function in the prism layer. The simulation object is a span of typical 17×17 rod bundle with a mid grid in PWR, all the structures including grid straps, springs, dimples, mixing vanes and welding spots are included. Thicknesses of grid and fuel cladding are considered in order to precisely simulate the fluid-solid conjugate heat transfer. The crud is set to be covered on the full span downstream of the grid. The simulation is based on the CILC risk pre-analysis and the computed information in the mostly likely crud deposit position is used as boundary condition. Based on the simulation results, the crud effects on the flow characteristics including vortex structures, circulation, turbulent intensity and second flow intensity and the heat transfer characteristics including rod temperature, fluid temperature and heat transfer coefficient are discussed in detail.


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