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# Nuc E 431 W Design Project Presentation

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Presentation for my 2010 senior core design project at The Pennsylvania State University with Westinghouse.

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### Nuc E 431 W Design Project Presentation

1. 1. This presentation will cover the following topics: This presentation will cover the following topics: 1. Introduction 2. 2 Loading Pattern Generation di G i 3. Safety Calculations 4. Operational Calculations 5. Thermal Hydraulics 5 Thermal‐Hydraulics 6. Conclusions
2. 2. Section 01
3. 3. Terminal Objective j • Become familiar with codes and methods used to  generate core loading patterns and perform reload  design analysis d i l i Enabling Objectives Enabling Objectives • Develop an acceptable reload core loading pattern • Perform safety and operational calculations on the Perform safety and operational calculations on the  designed LP along with thermal‐hydraulics analysis • Provide an oral presentation and a written report
4. 4. ANC: Advanced Nodal Code • Multidimensional nodal code (3D, 2D, 1D) • Licensed by the NRC in 1988 for PWR analysis • Calculates – Core reactivity – Assembly power and burnups A bl db – Rodwise power and burnups – Reactivity coefficients y – Core depletion – Control rod and fission product worths
5. 5. APA H code set used due to hexagonal geometry APA‐H code set used due to hexagonal geometry and consists of: • ALPHA‐H • PHOENIX‐H • ANC H ANC‐H These codes are the same in function as square geometry codes but modified to use hexagonal geometry.
6. 6. Differences from square geometry versions: Differences from square geometry versions: • Both the assembly and the core are modeled  in 1/6 and full core geometry in 1/6 and full core geometry • ANC‐H uses only one node per assembly as  compared to four nodes per assembly in ANC d f d bl i ANC Inputs and outputs are virtually the same
7. 7. VVER‐1000 • PWR Design – 3000 MWt g • Four‐Loop System • Hexagonal Fuel Assemblies Hexagonal Fuel Assemblies http://www.nukeworker.com/pictures/displayimage‐28‐37.html http://www.elemash.ru/en/production/Products/NFCP/VVER1000/
8. 8. • Inlet core temperature varies from 533.5 °F to  et co e te pe atu e a es o 533.5 to 553.1 °F from 0% to 100% power • Full Power Axial Offset (AO) band is ± 5% • Control rods vary from 0 to 175 steps withdrawn ( ) • Rod Insertion Limits (RILs) are a function of core  power • Westinghouse ZrB2 integral fuel burnable  absorbers (IFBA) are used. Possible configurations  are 0, 18, 24, 30, 36, and 48 rods per assembly.
9. 9. Section 02
10. 10. • Cycle Length Cycle Length • FΔH Peaking Factor • Moderator Temperature Coefficient (MTC) d C ffi i ( C) • Feed Inventory
11. 11. Parameter Limit Cycle Length Cycle Length ≥ 308 EFPD (11329 ≥ 308 EFPD (11329 MWd/MTU) ARO Peaking Factor (FΔH) ≤ 1.532 HZP MTC HZP MTC ≤ 0.00 pcm/ F ≤ 0.00 pcm/°F Feed Inventory ≤ 42 Feeds
12. 12. Customer plans to shut down cycle 4 at a cycle Customer plans to shut down cycle 4 at a cycle length of 308 EFPD. This value is used to  calculate the EOC burnup: l l h OC b
13. 13. The EOC of the core is identified as when the boron concentration is equal to 10 ppm. The E‐SUM output edit from cyc4_depl.0949.out confirms that  the designed loading pattern meets the limit of 308 EFPD which occurs at the  11329 MWd/MTU burnup step. 11329 MWd/MTU burnup step
14. 14. FΔH is literally defined as the normalized rise in enthalpy in a y py given subchannel. Since ANC‐H is a nodal based code based on the fuel assemblies and not the subchannels, ANC uses integrated rod power as the value for FΔH.
15. 15. A portion of the input from  03_anch_B1C4_depl.job is shown to the right. This input was also  g p used to determine cycle length.
16. 16. The maximum FΔH at each burnup step is included in the E‐SUM output edit. The limit of 1.532 must not be exceeded at any burnup step and is monitored at HFP ARO conditions.
17. 17. 1.540 1.530 1.520 1.510 F∆H 1.500 1.490 1.480 Actual Limit 1.470 1.460 0 2000 4000 6000 8000 10000 12000 Burnup [MWD/MTU] Burnup [MWD/MTU]
18. 18. The F of each assembly for a particular The FΔH of each assembly for a particular burnup step is shown in the C‐FDH output edit.
19. 19. MTC  change in core reactivity due to a change in MTC – change in core reactivity due to a change in moderator temperature (fuel temperature is held constant) and is checked at HZP for all burnup steps. constant) and is checked at HZP for all burnup steps A portion of the input from 03_anch_B1C4_depl.job is:
20. 20. The E SEQ output edit displays the MTC values The E‐SEQ output edit displays the MTC values for each burnup step.
21. 21. The calculation from ANC is verified for the most  limiting case (150 MWd/MTU burnup step). limiting case (150 MWd/MTU burnup step).
22. 22. 0 1700 ‐2 1500 1300 ‐4 pm]  1100 n Concentration [pp ‐6 MTC [pcm/°F]  900 ‐8 700 ‐10 10 Boron 500 MTC ‐12 300 MTC Limit ‐14 100 Boron Concentration Boron Concentration ‐16 ‐100 0 2000 4000 6000 8000 10000 12000 Burnup [MWD/MTU]
23. 23. Design Criteria Target Actual Cycle Length 308 EFPD 308.8 EFPD Maximum FΔH 1.532 1.514 Maximum MTC 0.00 pcm/°F ‐1.056 pcm/°F Feed Inventory 42 42
24. 24. Section 03
25. 25. Safety Calculations were performed using the Safety Calculations were performed using the Westinghouse Reactor Safety Analysis Checklist (RSAC) which covers: ( S C) hi h • Rodded FΔH • Shutdown Margin • Rod Ejection Accident Rod Ejection Accident
26. 26. Since most reactors are permitted to operate Since most reactors are permitted to operate at full power with some control rods inserted in the core, FΔH must also be checked with  i h l b h k d ih allowable control rods inserted. For this  particular scenario, the calculation was  performed with the lead control bank at its RIL. performed with the lead control bank at its RIL
27. 27. Input from roddedFDH.job p j Xenon was skewed for conservatism
28. 28. The rodded The rodded FΔH is displayed in the E‐SUM output is displayed in the E SUM output  edit from roddedFDH.0960.out.
29. 29. C‐FDH output edit from roddedFDH.0960.out
30. 30. Burnup [MWd/MTU] Δ Axial Offset (%) Rodded FΔH 150 5.61 1.499 500 5.30 1.496 1000 5.19 1.507 2000 5.16 1.518 3000 5.32 1.514 4000 5.39 1.510 5000 5.49 1.508 6000 5.66 1.504 7000 4.79 1.500 8000 5.86 1.494 9000 6.04 1.485 10000 6.22 1.477 11000 6.43 1.470 11329 6.49 6 49 1.467 1 467 11360 6.50 1.467
31. 31. 1.540 1.530 1.520 1.510 Fdh 1.500 1.490 1.480 Fdh Rodded Fdh Rodded Fdh 1.470 Fdh Limit 1.460 0 2000 4000 6000 8000 10000 12000 Burnup [MWD/MTU] Burnup [MWD/MTU]
32. 32. Shows that in any circumstances the operator will be able to safely shut down the core. Technically defined as the amount by which the core would  would be subcritical  (%Δρ) at hot shutdown conditions following a reactor trip, assuming the highest worth control rod is stuck out. Six cases in ANC: • K1 B K1 – Base Case at Burnup of Interest (BOC or EOC) C tB fI t t (BOC EOC) • K2 – Rods are Inserted to RILs • K3 – Over‐Power/ Over‐Temperature, Skew Power to     Top of Core  ( (worst conditions for trip) p) • K4 – Trip to Zero Power • K5 – Full Core at All Rods In (ARI)  • K6 – Worst Stuck Rod Out
33. 33. Calculation performed at both BOC and EOC Calculation performed at both BOC and EOC Total Power Defect‐ amount the core will  l f h ill increase in reactivity due to the trip to HZP Available SDM = Calculated SDM – Rod Worth Uncertainty – Voids
34. 34. E‐SUM output edit from sdownemBOC.0979.out
35. 35. E‐SUM output edit from sdownemEOC.1004.out
36. 36. Requirement BOC Worth (pcm) EOC Worth (pcm) Control Banks Power Defect 1943.7 3152.6 Void Effects 50 50 (1) Total Control Bank Requirement (1) Total Control Bank Requirement 1993.7 1993 7 3202.6 3202 6 Control Rod Worth (HZP) All rods inserted less most   6867 7677.3 reactive rod stuck out (2) Less 10% 6180.3 6909.6 Shutdown Margin Calculated Margin (2) – (1) 4186.6 3707 Required Shutdown Margin 1300 1300
37. 37. Purpose: Simulate the unlikely event of a single  p y g control rod being ejected from the core due to  failure in the control rod pressure housing. Total  peaking factor, F and %Δρ must be below limit  peaking factor FQ ,  and %Δρ must be below limit for each condition. Evaluated at Four Conditions: 1. BOC HFP 2. EOC HFP 3. BOC HZP 4. EOC HZP
38. 38. Input sample from rodejectionHFP.job Only control bank 10 is  j ejected from core at HFP. Since rod ejection is a  , fast transient, all  feedback effects are  frozen under an adiabatic  assumption.
39. 39. The E‐SUM output edit from  The E SUM output edit from rodejectionHFP.0963.out contains the total  peaking factor and eigenvalues.
40. 40. The rod ejection worth is calculated for each  The rod ejection worth is calculated for each case using the equation:
41. 41. Rod Ejection at HFP %Δρ (10%  FQ (13%  Case Eigenvalue dk/k %Δρ FQ uncertainty) uncertainty) BOC Full Core 1.000000 ‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐‐‐ BOC Bank 10 1.000128 0.000128 0.012799 0.014079 1.949 2.20237 EOC Full Core EOC Full Core 0.999200 0 999200 ‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐‐‐ EOC Bank 10 0.999377 0.000177 0.017713 0.019484 1.811 2.04643
42. 42. Approach is virtually same as for HFP with the exception being for HFP with the exception being the number of control rods ejected. Now four locations are ejected Now four locations are ejected individually.
43. 43. BOC Rod Ejection at HZP %Δρ (10%  FQ (13%  Case Eigenvalue dk/k %Δρ uncertainty) FQ uncertainty) Full Core 1.000001 ‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐‐‐‐ Bank 10 1.001415 0.001413 0.141300 0.158256 2.929 3.60267 Bank 9 1.002729 0.002724 0.272428 0.305120 4.922 6.05406 Bank 9  1.002328 0.002324 0.232429 0.260321 3.137 3.85851 (center) Bank 8 1.000479 0.000478 0.047789 0.053523 2.750 3.38250
44. 44. EOC Rod Ejection at HZP %Δρ (10%  FQ (13%  Case Eigenvalue dk/k %Δρ uncertainty) FQ uncertainty) Full Core 1.037299 ‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐‐‐‐ Bank 10 1.039348 0.001973 0.197337 0.221018 3.923 4.82529 Bank 9 1.040963 0.003526 0.352603 0.394915 6.408 7.88184 Bank 9  1.039740 0.002350 0.235046 0.263252 3.909 4.80807 (center) Bank 8 1.038825 0.001470 0.147005 0.164645 5.159 6.34557
45. 45. Rod Ejection Overview j Calculated Calculated Case (Bank) %Δρ Limit FQ Limit %Δρ FQ BOC HFP BOC HFP 0.014079 0.200 2.20237 5.8 EOC HFP 0.019484 0.200 2.04643 6.5 BOC HZP (10) 0.158256 0.860 3.60267 13.0 BOC HZP (9 ( 0.305120 0.860 6.05406 13.0 BOC HZP (9c) 0.260321 0.860 3.85851 13.0 BOC HZP (8) 0.053523 0.860 3.38250 13.0 EOC HZP (10) 0.001018 0.900 4.82529 21.0 EOC HZP (9) 0.394915 0.900 7.88184 21.0 EOC HZP (9c) EOC HZP (9c) 0.263252 0 263252 0.900 0 900 4.80807 4 80807 21.0 21 0 EOC HZP (8) 0.164645 0.900 6.34557 21.0
46. 46. Section 04
47. 47. Several Calculations must be performed before Several Calculations must be performed before the reactor can go back online after an outage: • BOC HZP Rodworths OC d h • Xenon Reactivity after Startup and Trip • Differential Boron Worth • Isothermal Temperature Coefficient Isothermal Temperature Coefficient • BOC HZP Critical Boron Concentration
48. 48. Rodworths of control banks are determined of control banks are determined  using the boron dilution method.  Input sample from rodworth.job E SUM edit from rodworth.0981.out E‐SUM edit from rodworth.0981.out
49. 49. Control Bank Worth Overview Control Banks Inserted CBC [ppm] Bank No. Bank Worth [ppm] ARO 1872 ‐‐‐‐‐‐‐‐‐‐‐‐ ‐‐‐‐‐‐‐‐‐‐‐‐‐ 10 1796 10 76 10 + 9 10 + 9 1656 9 140 10 + 9 + 8 1563 8 93 10 + 9 + 8 + 7 1480 7 83
50. 50. Reactivity worth of xenon is calculated in ANC H  Reactivity worth of xenon is calculated in ANC‐H for the following cases: • S Startup – BOC, MOC, EOC at 50% and 100% power • Trip – BOC, MOC, EOC at 50% and 100% power
51. 51. • Core is collapsed to 2‐D  for calculation • Xenon reactivity found  p over 100 hour period • No change in burnup  after startup after startup
52. 52. E‐SUM output edit from su_boc_fp.0983.out
53. 53. 0 0 500 ‐500 ‐500 500 BOC Full Power MOC Full Power ‐1000 ‐1000 ty [pcm] Reactivity [pcm] BOC Half Power MOC Half Power 1500 ‐1500 ‐1500 1500 Reactivit ‐2000 ‐2000 2500 ‐2500 ‐2500 2500 ‐3000 ‐3000 0 20 40 60 80 100 120 0 20 40 60 80 100 120 Time [hr] Time [hr] Time [hr] Time [hr] Reactivity after Startup
54. 54. 0 ‐500 EOC Full Power ‐1000 EOC Half Power Reactivity [pcm] ‐1500 ‐2000 ‐2500 ‐3000 0 20 40 60 80 100 120 Time [hr] Reactivity after Startup
55. 55. 0 0 ‐500 ‐500 ‐1000 ‐1000 ‐1500 ‐1500 Reactivity [pcm] Reactivity [pcm] ‐2000 ‐2000 ‐2500 ‐2500 ‐3000 ‐3000 MOC Full Power BOC Full Power ‐3500 ‐3500 MOC Half Power BOC Half Power BOC Half Power ‐4000 ‐4000 ‐4500 ‐4500 0 20 40 60 80 100 120 0 20 40 60 80 100 120 Time [hr] Time [hr] Time [hr] Time [hr] Reactivity after Trip
56. 56. 0 ‐500 ‐1000 ‐1500 pcm] ‐2000 Reactivity [p ‐2500 ‐3000 EOC Full Power ‐3500 EOC Half Power ‐4000 ‐4500 ‐5000 0 20 40 60 80 100 120 Time [hr] Reactivity after Trip
57. 57. Necessary to understand the Necessary to understand the reactivity effect of boron in the core under various conditions. d i di i Obtained by varying the boron concentration by ± 25 ppm throughout cycle. throughout cycle Input sample from dbw_HFP.job
58. 58. E‐SUM output edit from dbw_HFP.0998.out
59. 59. E‐SUM edit from dbw_hzp.0999.out
60. 60. ‐6.5 ‐7 m/ppm] erential Worth [pcm ‐7.5 ‐8 HZP Diffe HFP ‐8.5 ‐9 0 2000 4000 6000 8000 10000 12000 Burnup [MWd/MTU]
61. 61. The isothermal temperature coefficient (ITC) is The isothermal temperature coefficient (ITC) is used to confirm the validity of the MTC  prediction. ITC = MTC + DTC Most limiting case occurs at BOC HZP where the boron concentration is highest. the boron concentration is highest
62. 62. Input Sample from itc.job E‐SUM edit from itc.0997.out
63. 63. The value for ITC is not calculated in ANC H,  The value for ITC is not calculated in ANC‐H so it must be hand calculated:
64. 64. Confirmation of the BOC critical boron Confirmation of the BOC critical boron concentration at HZP is one of the final steps  required before startup can occur.  i db f E‐SUM output edit from hzp_cbc.1000.out
65. 65. Section 05
66. 66. Objective: perform realistic and conservative  Objective: perform realistic and conservative calculations to determine the departure from  nuclear boiling (DNBR) at full power and the  nuclear boiling (DNBR) at full power and the power level at which a boiling crisis occurs. Analysis performed using the COBRA‐IV PC code for the hot typical cell and the hot thimble cell
67. 67. • Applies numerical solutions to determine Applies numerical solutions to determine  thermal‐hydraulic parameters using  subchannel analysis method subchannel analysis method • Capable of determining flow and enthalpy  distribution at various axial and radial  distribution at various axial and radial locations • U Uses the Homogeneous Equilibrium Model  h H E ilib i M d l (HEM)
68. 68. COBRA IV used to calculate: COBRA‐IV used to calculate: • fuel, clad, and coolant temperature  distributions • flow quality and void fraction distributions • pressure drop • inter‐channel crossflow
69. 69. • Calculated as a function of elevation Calculated as a function of elevation • Typical and thimble cells calculated with  i l d hi bl ll l l d ih nominal and overpower cases directly  compared d
70. 70. Mass Flux for Hot Typical Channel 3.05 3 2.95 2.9 x (Mlb/hr/ft2) 2.85 2.8 Mass Flux 2.75 2.7 Nominal Case 2.65 Overpower Case Overpower Case 2.6 2.55 0 20 40 60 80 100 120 140 Axial Location (in) Axial Location (in)
71. 71. Mass Flux for Hot Thimble Channel 3 2.8 2.6 x (Mlb/hr/ft2) 2.4 Mass Flux 2.2 2 Nominal Case Overpower Case 1.8 1.6 0 20 40 60 80 100 120 140 160 Axial Location (in) Axial Location (in)
72. 72. Plotting coolant and cladding temperatures Plotting coolant and cladding temperatures illustrates different regions of the core that may undergo: d – Forced Convection – Nucleate Boiling – Saturated Boiling
73. 73. Hot Typical Cell Nominal  Hot Typical Cell Overpower  Temperatures Temperatures 750 750 Coolant Temperature Coolant Temperature Cladding Temperature Cladding Temperature 700 700 erature (F) Temperature (F) 650 650 Tempe 600 600 550 550 0 50 100 150 0 50 100 150 Axial Location (in) Axial Location (in)
74. 74. Hot Thimble Cell Nominal  Hot Typical Cell Overpower  Temperatures Temperatures 750 Coolant Temperature 750 Coolant Temperature Cladding Temperature Cladding Temperature 700 700 erature (F) Temperature (F) 650 650 Tempe 600 600 550 550 0 50 100 150 0 50 100 150 Axial Location (in) Axial Location (in)
75. 75. Since the onset of nucleate boiling can be Since the onset of nucleate boiling can be problematic for reactor kinetics, quality and void fraction are evaluated. id f i l d Void Fraction: percentage of volume in a  p y p channel occupied by vapor
76. 76. Hot Typical Cell Quality Hot Typical Cell Void Fraction 0.16 0.45 0.14 0.4 Nominal Void Fraction Nominal Void Fraction Nominal Quality 0.35 0.12 Overpower Void Fraction Overpower Quality 0.3 0.1 Void Fraction 0.25 Quality 0.08 0.2 Q 0.06 0.15 0.04 0.1 0.02 0.05 0 0 0 50 100 150 0 50 100 150 Axial Location (in) Axial Location (in)
77. 77. Hot Thimble Cell Quality Hot Thimble Cell Void Fraction 0.16 0.45 0.14 0.4 0.35 0.12 Nominal Void Fraction Nominal Quality 0.3 0.1 Void Fraction Overpower Void Fraction Overpower Quality 0.25 Quality 0.08 0.2 Q 0.06 0.15 0.04 0.1 0.02 0.05 0 0 0 50 100 150 0 50 100 150 Axial Location (in) Axial Location (in)
78. 78. Departure from Nucleate Boiling Ratio: ratio of  the heat flux needed to cause DNB to the  the heat flux needed to cause DNB to the actual heat flux of a fuel rod
79. 79. Minimum DNBR (MDNBR) limit is 1.17. Minimum DNBR (MDNBR) limit is 1 17 Power was increased to determine at what  i d d i h overpower the limit was reached Power  MDNBR Rod Channel Axial Location (in.) Cell Type 100% 3.37 2 2 107.2 Thimble 153% 1.174 11 31 135.8 Typical
80. 80. Typical Cell DNBR 25 20 Nominal Case Nominal Case Overpower Case Boiling Crisis 15 DNBR D 10 5 0 0 20 40 60 80 100 120 140 Axial Location (in) Axial Location (in)
81. 81. Thimble Cell DNBR 25 20 Nominal Case Nominal Case Overpower Case Boiling Crisis 15 DNBR D 10 5 0 0 20 40 60 80 100 120 140 Axial Location (in) Axial Location (in)
82. 82. Section 06
83. 83. Terminal Objective Terminal Objective • Successfully became familiar with codes (ANC  and COBRA‐IV) used to generate core loading  and COBRA IV) used to generate core loading patterns and perform reload design analysis
84. 84. Enabling Objectives Enabling Objectives • Successfully developed an acceptable core  reload pattern that met all limitations reload pattern that met all limitations • Safety, Operational, and Thermal‐Hydraulic  calculations were performed  l l i f d • Written report completed