Module 12   Two Phase Fluid Flow And Heat Transfer 2010 June Nrc
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Module 12 Two Phase Fluid Flow And Heat Transfer 2010 June Nrc

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This provides an overview of two phase systems including equations used to predict flow and heat transfer.

This provides an overview of two phase systems including equations used to predict flow and heat transfer.

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Module 12 Two Phase Fluid Flow And Heat Transfer 2010 June Nrc Presentation Transcript

  • 1. Fundamentals of Nuclear Engineering Module 12: Two Phase Heat Transfer and Fluid Flow Joseph S. Miller, P.E. 1
  • 2. 2
  • 3. Objectives: Previous Lectures described fluid flow, heat transfer and reactivity in single phase systems. This lecture: 1. Describe two-phase Systems 2. Describe important thermal-hydraulic concepts important to a BWR 3. Describe two-phase flow equations 4. Describe two phase heat transfer rates from fuel to coolant and Boiling Transition 5. Describe steady state core temperature profiles 6. Describe fluid flow, and pressure drops in two phase systems 7. Describe behavior of system during accident 3
  • 4. 1. Two Phase Fluid Systems 4
  • 5. Two-Phase Flow Systems in Nuclear Engineering • Heat Exchangers • Piping Systems in Balance of Plant and Reheat of Feedwater • PWR Steam Generator • BWR Reactor 5
  • 6. 6
  • 7. BWR Flow Paths 7
  • 8. PWR Steam Generator 8
  • 9. 2. Important Thermal Hydraulic Stages for BWR • Start-up and Steady State Operation • Operational Transients • Loss of Coolant Accidents Each of these stages require many different analytical techniques for predicting heat transfer and fluid flow in a BWR. 9
  • 10. Start-up and Steady State Operation • Core Thermal Power • Power-Flow Map • Control Rod Positioning • Feedwater Temperature Control – Amount of Subcooling – more power • Core Response to Recirculation Flow Changes - BWR 10
  • 11. Core Thermal Power • Core thermal power by energy balance and use of instrumentation. • Inlet subcooling • Quality in channel • Void fraction in the fuel channel • Fluid flow and pressure drop • Core orifices • Core bypass flow • Enrichment distribution in fuel 11
  • 12. Operational Transients • Based on FSAR Chapter 15 requirements for initiating events • Nuclear reactor system pressure increases by reactor trip, MSIV isolation, etc. • Positive reactivity insertion by moderator temperature increase as in loss of feedwater heating. 12
  • 13. Loss of Coolant Accidents • Large breaks • Small breaks • Intermediate breaks 13
  • 14. 3. Two Phase Flow Equations 14
  • 15. General Terminologies • Two-phase flow – Simultaneous flow of any two phases (liquid-gas/vapor) of a single substance – Examples: reactor fuel channels, steam generators, kettle on a hot stove – Also referred to as “Single-component two-phase flow” • Two-component flow – Simultaneous flow of liquid and gas of two substances – Examples: oil-gas pipelines, beer, soft drink, steam-water-air flow at discharge of safety valve. – Also referred to as “Two-component, two-phase flow” 15
  • 16. Terminology Unique to Two Phase • In static (non-flowing system) steam quality: χ is defined: χ = (mass of steam) / (total mass of steam + liquid) • In static (non-flowing system) void fraction: α is defined: α = (volume of steam in mixture) / (total volume of steam + liquid) • Void fraction can be expressed in terms of steam quality and specific volumes (from Steam Tables) as follows: α = χvg / ((1- χ )vf + χvg ) = 1 / {1 + [(1 – χ)/ χ] vf / vg } • Where: vg is specific volume of steam in ft3/lb-m vf is specific volume of liquid in ft3/lb-m vfg = vg - vf is difference in specific volumes Good source for fluid properties: http://webbook.nist.gov/chemistry/fluid/ 16
  • 17. Terminology Unique to Two Phase (cont.) • In static (non-flowing system) steam quality: χ is defined: χ = (mass of steam) / (total mass of steam + liquid) χ = (v - vf ) / vfg Where v is the specific volume of the two-phase mixture 17
  • 18. Quality and P- v-T Diagram 18
  • 19. P- v-T Diagram Definitions 1000 psi Line • “I” state represents initial state at some subcooled (compressed liquid) value at a lower temperature • “IJ” line represents the heat up of the fluid from a lower temperature subcooled state “I” to saturated liquid state “J” at saturated temperature of 544 ͦ F at 1000 psi F(See steam tables on page 83) • “JK” line represents the continued heat of the two-phase mixture at 1000 psi. The mixture stays at 544 ͦ F while it is two-phase, but the specific volume continues to increase. From this time the quality changes from 0 to 1 (Saturated liquid to saturated steam). • “KL” line represents the continued heat –up from saturated steam to super heated steam as temperature and specific volume increase. 19
  • 20. Calculate Quality 20
  • 21. Void Fraction • Ratio of Vapor flow area to total flow area • Depends strongly on pressure, mass flux, and quality • Applied to calculate the acceleration pressure drop in steady-state homogeneous code • Large number of correlations proposed • Solved from conservation equations in two-fluid reactor safety codes 21
  • 22. Steam Rises Faster in Channel Than Liquid • Because of lower density (buoyancy) steam will rise up vertical channel faster than surrounding liquid • Slip ratio: S, is the ratio of steam velocity to liquid velocity S = Vg / Vf where: Vg is steam velocity in ft. / sec. Vf is liquid velocity in ft. / sec. • Slip ratio modifies static definitions of α (void fraction) and χ (steam quality) in flowing two phase system 22
  • 23. Relative Velocities Are Different • If total mass flow rate is: W (lb-m/sec) • Steam flow rate is: χW • Liquid flow rate is: (1- χ)W • Phase volumetric velocity of steam is: Vg = vg W χ / Ag -where: Ag is relative cross sectional area of steam in two phase column • Phase volumetric velocity of liquid is: Vf = vf W(1- χ) / Af 23
  • 24. Definition of Slip in Terms of Steam Quality and Void Fraction • Slip: S = Vg / Vf • Slip defined in terms of steam quality by combining: Vg = vg W χ / Ag Vf = vf W(1- χ) / Af • This yields: S = (χ / 1- χ)(Af / Ag)(vg / vf) • Noting that in small slice of column, ratio of steam to total mixture is: α = Ag / Ag+ Af , rearranging this: (Af / Ag) = (1 – α )/ α • Slip can then be expressed: S = (χ / 1- χ)[(1 – α )/ α](vg / vf) 24
  • 25. Definition of Void Fraction in Terms of Steam Quality and Slip • Slip equation can be rearranged to define void fraction: α in terms of steam quality and Slip: 1 α= (1 − χ )  v f  1+ [ ] S χ  vg   • When S = 1: steam and liquid move at exact same speed Effect of slip: • Slip decreases void fraction α(χ) below that which exists in situation of no slip between steam and liquid 25
  • 26. Calculate Void Fraction with Slip 26
  • 27. Void Fraction Sensitivities: • Sensitivity of Void fraction α to Mixture Quality and Slip can be seen by computing α(χ) for a spectrum of pressure and assumed Slip values • S = 1 implies homogeneous steam/water flow (moving together) 27
  • 28. Example Slip Ratio for BWR Fuel Channel 28
  • 29. What Does Slip Mean to the Core Fluid Flow? Low S value (S = 1) implies: • Voids and water travel about same speed Higher S value (S = 2,3) implies: • Steam carries out higher enthalpy water thus heat is removed faster • Voids swept out of channel faster which is benefit for neutron economy • Predicting Slip from first principles is not easy • Designers rely upon tests and scaling-up from previous 29
  • 30. 4. Two Phase Heat Transfer Rates from Fuel to Coolant and Boiling Transition 30
  • 31. Two Phase Heat Transfer Regimes 31
  • 32. Six Distinctive Boiling Regions: • Single phase, forced convection • Nucleate boiling • Critical heat flux • Transition boiling • Minimum film boiling • Film boiling Methods and experimental correlations exist to describe each region 32
  • 33. Boiling Curve 33
  • 34. Definitions for Transition points • Onset of nucleate boiling. Transition point between single-phase and boiling heat transfer • Onset of net vapor generation. Transition point between single-phase and two-phase flow (mainly for pressure- drop calculations) • Saturation point. Boiling initiation point in an equilibrium system. • Critical heat flux point. Transition point between nucleate boiling and transition/film boiling. • Minimum film-boiling point. Transition point between transition boiling and film boiling. 34
  • 35. Flow Patterns • Distribution of phases inside a confined area • Depend strongly on liquid and vapor velocities • Channel geometry • Surface Heating 35
  • 36. Flow Patterns in Horizontal Flow 36
  • 37. Flow Patterns in Vertical Flow 37
  • 38. Vertical Heated Channels 38
  • 39. Single Phase Forced Convection Heating • Dittus-Boelter correlation was previously described for PWR steady state heat removal • Dittus-Boelter correlation is appropriate for subcooled region of BWR fuel channel (before boiling starts) hfilm = (k / Dh) 0.023 Pr0.4 Re0.8 • Above subcooled region – different hfilm model would apply • ALSO: When modeling cooling on tube: hfilm = (k / Dh) 0.023 Pr0.3 Re0.8 • Example problem: heat transfer in subcooled region • Assume: 1000 psia, Tin = 515°F, Vf = 6.8 ft./sec. • Use standard dimensions of GE 8x8 Fuel Bundle • Calculate hfilm 39
  • 40. Example hfilm Calculation for 8x8 BWR Fuel 40
  • 41. Nucleate Boiling • Thom correlation is one commonly used for evaluating nucleate boiling: • qTHOM = 0.05358 x exp(P/630) x (Tc – Tsat(P))2 • qTHOM is the heat transfer rate in BTU/sec.ft.2 • P is pressure in psia • Tc is clad surface temperature in °F • Tsat is saturation temperature for pressure: P 41
  • 42. Critical Heat Flux This is an example BWR Fuel Bundle CHF correlation developed by EPRI 42
  • 43. Transition Boiling Something Like This Would Be For LOCA 43
  • 44. 5. Steady State Core Temperature Profiles 44
  • 45. BWR Axial Heat Transfer 45
  • 46. BWR Axial Heat Transfer • Recall: Axial, radial distribution derived earlier (same as in PWR) • Φ(r,z) = ΦoJo(2.405r/R)Cos(πz/H) • Again assume linear power density in individual rod given by: q(z) = qoCos(πz/H) qo = (πRc2) Ef ∑f Φo Jo(2.405r/R) • Energy balance along single rod in BWR must now reflect heating subcooled water up to saturation point below: HBOIL • Above HBOIL: boiling heat transfer 46
  • 47. BWR Axial Heat Transfer • As subcooled water enters heated channel • Temperature rises until boiling point Tsat(P) z at HBOIL: q( z ) T ( z , P ) = Tin ( P ) + ∫ − H eff WC p ( P ) dz 2 z q( z ) h( z , P ) = hin ( P ) + ∫ − H eff W dz 2 • Above HBOIL further heat addition only increases steam content, not temperature • Enthalpy rise: h(z,P) = hsat(P) + χ(z)hfg(P) z q( z ) where : χ ( z ) = ∫ dz H BOIL Wh fg ( P) 47
  • 48. Simulation of Uniform Linear Power Density 48
  • 49. Simulation of Cosine Linear Power Density 49
  • 50. Simulation of Cosine Linear Power Density 50
  • 51. 6. Fluid Flow and Pressure Drops in Two-Phase Systems 51
  • 52. Simulation of Variable Recirculation Flow •Previous lecture noted BWR capability to vary recirculation flow to raise/lower power 52
  • 53. Two Phase Flow Pressure Drop 53
  • 54. Pressure Drop in Two Phase System • Recall: For single phase flow system in channel, pressure drop in psia can be calculated:  fL  ρυ 2  ρυi 2   D  2(144) g + ∑  K i 2(144) g  ∆Pfriction =     h i   • In two phase system: pressure drop is larger • Experimental tests have lead to a simple working relationship between single phase and two phase pressure drops. • Following homogeneous two phase pressure drop has been developed for steady state flow conditions: ∆P2φ R= ∆P φ 1 • -where: ΔP2Φ is calculated assuming all liquid flow at total mass flow rate 54
  • 55. Martinelli-Nelson Friction Multiplier • This is classical approach. Advanced approaches exist • If equivalent single phase pressure drop is known • Homogeneous two phase pressure drop is: ∆P2φ = R × ∆P φ 1 • Where: 55
  • 56. 56
  • 57. BWR Fuel Bundle Geometry 57
  • 58. BWR Fuel Channel Pressure Drop (Pressure Drops Due to Grid Spacers, Inlet/Outlet Geometry Would Need to be Added!) This calculation uses steam quality and bundle geometry from previous example: 58
  • 59. Example Acceleration Pressure Drop Calculation 59
  • 60. 2-Phase Expansion and Contraction Losses • Recall in treatment of single phase pressure drops: ΔP = KρVin2 / 2 • Situation for two-phase flow is more complicated • Corresponding pressure drops for two-phase flow are larger. • Higher void fractions result in larger pressure drops 60
  • 61. Approximate Pressure Drop Across a BWR Channel Acceleration - 1.5 psi Gravity - 2.5 psi Friction - 6 psi Local (form) - 12 psi Total - 22 psi Compare to Approximate Pressure Drop Across a PWR Channel Gravity - 4.4 psi Friction - 6.25 psi Local (form) - 9 psi Total - 19.65 psi 61
  • 62. 7. Behavior of System During Accident 62
  • 63. BWR Break Water Level 63
  • 64. 64
  • 65. 65
  • 66. 66
  • 67. 67
  • 68. 68
  • 69. 69
  • 70. 70
  • 71. Summary • Heat transfer in BWR fuel channels can be evaluated using approaches based on convective heat transfer based experimental data. • Heat flux models exist for all heat transfer regimes. These are complicated correlations based on experiments. • Pressure drops due to two phase flow are greater than those found for single phase flow. • Fluid Flow Transient and LOCA Situations are evaluated using Large Computer Programs such as RELAP5, TRAC and TRACE 71
  • 72. Important Links • Some good course material two-phase flow to review - http://www2.et.lut.fi/ttd/studies.html • Basic Nuclear Energy - http://www.nrc.gov/reading-rm/basic- ref/students.html • Basic BWR - http://www.nrc.gov/reading- rm/basic-ref/teachers/03.pdf 72
  • 73. References 1. Wallis G.B., One-dimensional two-phase flow, McGraw-Hill Book Company, New York (1969) 2. Lahey R.T., Moody F.J., The thermal-hydraulics of a boiling water nuclear reactor, 2nd Ed., American Nuclear Society (1993). 3. N. Todreas and M Kazimi, ”Nuclear Systems I – Thermal Hydraulic Fundamentals”,Taylor & Francis”, (1993). 4. N. Todreas and M Kazimi, ”Nuclear Systems II – Elements of Thermal Hydraulic Design”,Hemishere Publishing Corporation, (1990). 5. Bird, R.B., Steward, W.E., and Lightfoot, E.N., ”Transport Pheomena”, New York: Wiley, (1960) 6. ASME Steam Tables 7. L.S. Tong & Joel Wiesman, ”Thermal Anhalu=ysis of Pressurized Water Reactors”, third edition, American Nuclear Society, La Grange Park, Il, (1996) 73
  • 74. References (cont) 8. M.M. El-Wakil, “Nuclear Energy Conversion”, American Nuclear Society, La Grange Park, Il, Third Printing, (January 1982) 74
  • 75. Extra Information Steam Tables 75
  • 76. 76