ML16343A191
| ML16343A191 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/24/2016 |
| From: | Indiana Michigan Power Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16336A263 | List:
|
| References | |
| AEP-NRC-2016-42 | |
| Download: ML16343A191 (4) | |
Text
IINNDDIIAANNAA AANNDD M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
23 Table:
14.1-3 Page:
1 of 4 UNIT 1
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/oF)
Moderator Density (K/gm/cc)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output (MWt)
Reactor Vessel Coolant Flow (GPM)1 Vessel Average Temp.
(oF)
Pressurizer Pressure (PSIA)
Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition TWINKLE FACTRAN THINC Refer to Section 14.1.1 Min2 W-3/WRB-1 See Section 14.1.1 No 0
146,432 547 2033 Uncontrolled RCCA Bank Withdrawal at Power3 LOFTRAN
+5
.54 Min and Max4 WRB-1 Yes 3327 1996 333 339,100 575.45 564.58 549.93 2100 1 The non-LOCA analyses are based on a Thermal Design Flow (TDF) of 83,200 gpm/loop and a Minimum Measured Flow (MMF) of 84,775 gpm/loop. However, subsequent evaluations were performed to show that the following higher flows are also supported: 88,500 gpm/loop (TDF) and 90,725 gpm/loop (MMF).
2 Minimum Doppler power defect (pcm/%power) = -9.55 + 0.00104Q where Q is in MWt.
3 Multiple power levels, Tavg, and reactivity feedback cases were examined.
4 Maximum Doppler power defect (pcm/%power) = -19.4 + 0.002Q where Q is in MWt.
5 Value used in the DNB analysis is performed at the MUR power uprate program maximum Tavg of 575.4°F.
IINNDDIIAANNAA AANNDD M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
23 Table:
14.1-3 Page:
2 of 4 UNIT 1
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/oF)
Moderator Density (K/gm/cc)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output (MWt)
Reactor Vessel Coolant Flow (GPM)1 Vessel Average Temp.
(oF)
Pressurizer Pressure (PSIA)
RCCA Misalignment6 LOFTRAN THINC N/A7 NA NA WRB-1 Yes 3250 339,100 576.3 (5) 2100 Uncontrolled Boron Dilution N/A N/A N/A N/A N/A N/A 3425 0
N/A N/A N/A Loss of Forced Reactor Coolant Flow (6)
LOFTRAN FACTRAN THINC
+5 N/A Max WRB-1 Yes 3270 339,100 576.3 (5) 2100 Locked Rotor (Peak Pressure)
LOFTRAN
+5 N/A Max N/A N/A 3335 332,800 581.4 2317 Locked Rotor (Peak Clad Temp)
LOFTRAN FACTRAN
+5 N/A Max N/A N/A 3335 332,800 581.4 2033 6 An uprated core power of 3315 MWt (NSSS power of 3327 MWt) is supported via an evaluation that addresses a reduction in power uncertainty from 2% to approximately 0.3%.
7 N/A - Not Applicable
IINNDDIIAANNAA AANNDD M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
23 Table:
14.1-3 Page:
3 of 4 UNIT 1
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/oF)
Moderator Density (K/gm/cc)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output (MWt)
Reactor Vessel Coolant Flow (GPM)1 Vessel Average Temp.
(oF)
Pressurizer Pressure (PSIA)
Locked Rotor (Rods-in-DNB) (6)
LOFTRAN FACTRAN THINC
+5 N/A Max WRB-1 Yes 3270 339,100 576.3 2100 Loss of Electrical Load and/or Turbine Trip8 LOFTRAN
+5
.54 Max and Min WRB-1 Yes No 3327 3327 339,100 332,800 576.3 (5) 581.4 2100 2033 Loss of Normal Feedwater9 LOFTRAN
+5 N/A Max N/A N/A 3409 332,800 548.9 2317 Excessive Heat Removal(9) Due to Feedwater System Malfunction LOFTRAN N/A
.54 Min WRB-1 Yes 3327 0
339,100 576.3 (5) 547 2100 Excess Load Increase(9)
Incident LOFTRAN N/A 0 and.54 Max and Min WRB-1 Yes 3425 366,400 578.7 2100 8 Minimum and Maximum reactivity feedback cases were examined 9 Values presented were used in the rerating analysis. Subsequent evaluations support the 30% SGTP parameters presented as Cases 1 and 2 of Table 14.1-1
IINNDDIIAANNAA AANNDD M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
23 Table:
14.1-3 Page:
4 of 4 UNIT 1
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/oF)
Moderator Density (K/gm/cc)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output (MWt)
Reactor Vessel Coolant Flow (GPM)1 Vessel Average Temp.
(oF)
Pressurizer Pressure (PSIA)
Loss of Offsite Power (LOOP)(9) to the Station Auxiliaries LOFTRAN
+5 N/A Max N/A N/A 3409 332,800 548.9 2033 Rupture of a Steam Pipe LOFTRAN THINC See Figure 14.2.5-1 N/A See Figure 14.2.5-2 W-3 N/A 0
332,800 547 2100 Rupture of a Control Rod Drive Mechanism Housing TWINKLE FACTRAN See Section 14.2.6 N/A Min N/A N/A 3335 0
332,800 146,432 581.4 547 2033