ULNRC-05938, Enclosure 1: Callaway Plant Unit 1 - Responses to RAI Set 17 and Amendment 17 to the Callaway LRA
| ML12349A180 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 12/13/2012 |
| From: | Ameren Missouri |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML123490272 | List: |
| References | |
| ULNRC-05938 | |
| Download: ML12349A180 (76) | |
Text
ULNRC-05938 December 13, 2012 Page 1 of 24 CALLAWAY PLANT UNIT 1 LICENSE RENEWAL APPLICATION REQUEST FOR ADDITIONAL INFORMATION (RAI) Set #17 RESPONSES
ULNRC-05938 December 13, 2012 Page 2 of 24 RAI 4.3-2a
Background:
In response to Part (a) to request for additional information (RAI) 4.3-2 dated October 11, 2012, Union Electric Company d/b/a Ameren Missouri (the applicant) provided the short-term and long-term weighting factors. However, the applicant did not provide short-term and long-term occurrences (i.e., short-term average rate of accumulation and total average rate of accumulation, respectively) of each transient as requested in Part (a) of RAI 4.3-2.
Issue:
Since the applicant used the 60-year transient projections to support the disposition of the time-limited aging analyses (TLAAs) evaluated in license renewal application (LRA) Sections 4.7.2 and 4.7.7, the staff requires additional information to determine whether the long-term and short-term weighting factors and the associated transient occurrences for these weighting factors used in the projection methodology is appropriate and conservative.
Request:
Provide the short-term average rate of accumulation and total average rate of accumulation for each transient listed in LRA Table 4.3-2 that supports the calculation of the 60-year projections.
Callaway Response Table 1 (refer to table following this response) provides the short-term average rate of accumulation and total average rate of accumulation for each transient listed in LRA Table 4.3-2.
There are some differences between the information shown in Table 1 and LRA Table 4.3-2.
- 1. LRA Table 4.3-2 shows a date of 1983 versus 1984. The date of 1983 includes the transients that occurred during startup testing prior to the start of commercial operation.
However the rate calculations assume that all transients occurred after the start of commercial operation in 1984.
- 2. There are differences between some of the projections in Table 1 and those in LRA Table 4.3-2. The LRA Table 4.3-2 projections were generated with fatigue management program software and the exact long-term and short-term average rates of accumulation used in the calculation vary slightly between transients. However the largest difference is 3 events which demonstrates good agreement between Table 1 and LRA Table 4.3-2. In order to be consistent with the described methodology, LRA Table 4.3-2 projections have been revised as shown by Amendment 17 in Enclosure 2 to be consistent with Table 1.
The changes do not affect any of the TLAA dispositions.
- 3. The baseline analysis of Callaway plant cycles and fatigue usage noted in LRA reference 16 was revised to correct a non-conservative assumption where all pre-1995 reactor trips were classified as Upset Condition Transient # 4a (reactor trip from full power, without cooldown). A review of the manual records determined some of the Upset Condition Transient # 4a events to be reclassified to Upset Condition Transient #4b (reactor trip from full power, with cooldown, without safety injection) or Upset Condition Transient # 4c (reactor trip from full power, with cooldown, with safety injection). LRA Table 4.3-2 has been revised as shown by Amendment 17 in Enclosure 2 to include the updated reactor trip count for Upset Condition Transient # 4a, Transient # 4b, and Transient # 4c. The
ULNRC-05938 December 13, 2012 Page 3 of 24 revised transient count does not affect the dispositions of the EAF calculations which are based on the 60 year cycle projections.
With the exception of the reactor coolant loop (RCL) leak-before-break (LBB) analysis, the transients used by the TLAAs evaluated in LRA Sections 4.7.2 and 4.7.7 were provided in Ameren letter ULNRC-05903, dated September 6, 2012 (ADAMS Access # ML12251A130 and ML12251A131), in response to RAI 4.7.2-1, RAI 4.7.2-3, and RAI 4.7.7-1. Tables 2 & 3 (refer to tables following this response) revise the transients provided to include RCL LBB transients and the percent accumulation. None of the transient projections which utilized the short term and long term weighting factors exceed 50% of the design value. This large amount of margin supports the conclusion that these TLAAs can be dispositioned in accordance with 10 CFR 54.21(c)(1)(i).
In addition, the Fatigue Monitoring program, described in LRA Appendix B, Section B3.1, will ensure that actual plant experience remains bounded by the thermal and pressure transient numbers and severities analyzed in the design calculations, or that corrective actions maintain the design and license basis.
Corresponding Amendment Changes Refer to the Enclosure 2 Summary Table "Amendment 17, LRA Changes for a description of LRA changes with this response.
ULNRC-05938 December 13, 2012 Page 4 of 24 RAI 4.3-2a Table 1 Accumulation Rates that Support the 60-Year Projections Transient Description Long Term (1984 - 2011)
Short Term (2002 -2011)
Projection (2011-2044)
Total (1984-2044)
Limiting Number
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
Normal Condition Transients 1a. Plant heatup at 100ºF/hr Pressurizer Heatup at 100°F/hr 29 1.07 10 1.11 36 1.10 65 1.08 200 1b. Plant cooldown at 100ºF/hr Pressurizer Cooldown at 200°F/hr 29 1.07 10 1.11 36 1.10 65 1.08 200 2a. Unit loading at 5%
of full power per min 178 6.59 7
0.78 74 2.23 252 4.20 11200 2b. Unit unloading at 5% of full power per min 184 6.81 13 1.44 92 2.79 276 4.60 13200 3a. Step increase 10% of full power 24 0.89 11 1.22 38 1.14 62 1.03 2000 3b. Step decrease 10% of full power 20 0.74 8
0.89 28 0.85 48 0.80 2000
- 4. Large step decrease with steam dump 6
0.22 2
0.22 7
0.22 13 0.22 200 5a. Steady state fluctuations, Initial fluctuations Not Counted 1.5 E5 5b. Steady state fluctuations, Random fluctuations 3.00 E6
ULNRC-05938 December 13, 2012 Page 5 of 24 Transient Description Long Term (1984 - 2011)
Short Term (2002 -2011)
Projection (2011-2044)
Total (1984-2044)
Limiting Number
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
0.44 52 1.57 186 3.10 SG B 132 4.89 1
0.11 43 1.31 175 2.92 SG C 134 4.96 3
0.33 49 1.49 183 3.05 SG D 134 4.96 2
0.22 46 1.41 180 3.00 7a. Loop out of service, Normal loop shutdown Not Licensed for N-1 loop operations 0
0 80 7b. Loop out of service, Normal loop startup 70 8a. Unit loading between 0-15% of full power 93 3.44 20 2.22 83 2.53 176 2.93 500 8b. Unit unloading between 0-15% of full power 92 3.41 20 2.22 83 2.52 175 2.92 500
- 9. Boron concentration equalization Not Counted 26400
- 10. RCP startup and shutdown 3800 RCP A, Start 139 5.15 13 1.44 78 2.37 217 3.62 RCP A, Stop 138 5.11 12 1.33 75 2.28 213 3.55 RCP B, Start 142 5.26 13 1.44 79 2.40 221 3.68 RCP B, Stop 142 5.26 13 1.44 79 2.40 221 3.68 RCP C, Start 139 5.15 14 1.56 81 2.45 220 3.67 RCP C, Stop 139 5.15 14 1.56 81 2.45 220 3.67 RCP D, Start 122 4.52 13 1.44 73 2.21 195 3.25 RCP D, Stop 122 4.52 13 1.44 73 2.21 195 3.25
ULNRC-05938 December 13, 2012 Page 6 of 24 Transient Description Long Term (1984 - 2011)
Short Term (2002 -2011)
Projection (2011-2044)
Total (1984-2044)
Limiting Number
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- 11. Reduced temperature return to power Callaway does not load follow 2000
- 12. Refueling 17 0.63 6
0.67 22 0.66 39 0.65 80
- 13. Turbine roll test Startup test 7
0.12 10
- 14. Primary side leak test 8
0.30 0
0.00 2
0.07 10 0.17 50
- 15. Secondary side leak test 1
0.04 1
0.11 3
0.09 4
0.07 80 16a. Feedwater heaters out of service:
One heater out of service 56 2.07 12 1.33 50 1.52 106 1.77 120 16b. Feedwater heaters out of service:
One bank out of service 12 0.44 0
0.00 4
0.11 16 0.27 120
- 17. RPV bolting/unbolting 20 0.74 7
0.78 25 0.77 45 0.75 57 Upset Condition Transients
- 1. Loss of load (without immediate reactor trip)
No historical events 1
0.02 80
- 2. Loss of Power (with natural circulation in the RCS)
Projection is set equal to the number of events used in the EAF analyses.
1 0.02 40
- 3. Partial loss of flow (loss of one pump)
No historical events 1
0.02 80 4a. Reactor trip from full power, Without cooldown.
61 2.26 2
0.22 24 0.73 85 1.42 230
ULNRC-05938 December 13, 2012 Page 7 of 24 Transient Description Long Term (1984 - 2011)
Short Term (2002 -2011)
Projection (2011-2044)
Total (1984-2044)
Limiting Number
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr) 4b. Reactor trip from full power, With cooldown, without safety injection 4
0.15 0
0.00 1
0.04 5
0.08 160 4c. Reactor trip from full power, With cooldown, with safety injection Projection is set equal to the number of events used in the EAF analyses.
1 0.02 10 4d. Reactor trip from full power, With no inadvertent cooldown
- emergency overspeed Not Counted 20
- 5. Inadvertent RCS depressurization 2
0.07 1
0.11 3
0.10 5
0.08 20 5a. Inadvertent RCS depressurization due to inadvertent auxiliary spray 1
0.02 10
- 6. Inadvertent startup of an inactive RCS loop No historical events 1
0.02 10
- 7. Control rod drop 1
0.02 80
- 8. Inadvertent safety injection actuation Projection is set equal to the number of events used in the EAF analyses.
2 0.03 60
- 9. Operating Basis Earthquake No historical events 1
0.02 20 events
- 10. Excessive Feedwater Flow 1
0.02 30
- 11. RCS Cold Over-pressurization 1
0.02 10
ULNRC-05938 December 13, 2012 Page 8 of 24 Transient Description Long Term (1984 - 2011)
Short Term (2002 -2011)
Projection (2011-2044)
Total (1984-2044)
Limiting Number
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
Test Condition Transients
- 1. Primary side hydrostatic test Startup tests 1
0.02 5
- 2. Secondary side hydrostatic test 1
0.02 10
- 3. Tube leakage test No historical events 1
0.02 800
- 4. Cold hydrostatic test Not Counted 10 Auxiliary Transients
- 1. Normal charging and letdown shutoff and return to service 36 Alternate path No historical events 1
0.02 Normal path 4
0.15 0
0.00 1
0.04 5
0.08
- 2. Letdown flow shutoff with prompt return to service 120 Alternate path 1
0.04 1
0.11 3
0.09 4
0.07 Normal path 11 0.41 3
0.33 12 0.35 23 0.38
- 3. Letdown flow shutoff with delayed return to service 12 Alternate path 2
0.07 2
0.22 6
0.19 8
0.13 Normal path 12 0.44 2
0.22 9
0.28 21 0.35
ULNRC-05938 December 13, 2012 Page 9 of 24 Transient Description Long Term (1984 - 2011)
Short Term (2002 -2011)
Projection (2011-2044)
Total (1984-2044)
Limiting Number
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- 4. Charging flow shutoff with prompt return to service 12 Alternate path No historical events 1
0.02 Normal path 12 0.44 3
0.33 12 0.36 24 0.40
- 5. Charging flow shutoff with delayed return to service 12 Alternate path 1
0.04 1
0.11 3
0.09 4
0.07 Normal path 9
0.33 1
0.11 5
0.17 14 0.23
- 6. Charging flow decrease & return to normal Not Counted 14400
- 7. Charging flow increase & return to normal 14400
- 8. Letdown flow decrease & return to normal 1200
- 9. Letdown flow increase & return 14400
- 10. Load follow boration 24000
- 11. Accumulator actuation, accident operation No historical events 1
0.02 21
- 12. Inadvertent accumulator blowdown 1
0.02 4
- 13. RHR operation -
plant cooldown 29 1.07 10 1.11 36 1.10 65 1.08 200
ULNRC-05938 December 13, 2012 Page 10 of 24 Transient Description Long Term (1984 - 2011)
Short Term (2002 -2011)
Projection (2011-2044)
Total (1984-2044)
Limiting Number
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- 14. High head safety injection 110 Loop A 3
0.11 0
0.00 1
0.03 4
0.07 Loop B No historical events 1
0.02 Loop C 1
0.02 Loop D 1
0.02
- 15. [Seal] Injection flow temperature change 29 1.07 10 1.11 36 1.10 65 1.08 180
- 16. Elevated Seal Water Injection Temperature 55 2.04 19 2.11 69 2.09 124 2.07 200
- 17. Loss of seal injection flow 3
0.11 1
0.11 4
0.11 7
0.12 40
- 18. Elevated CCW injection temperature 30 1.11 11 1.22 39 1.19 69 1.15 150
- 19. CCW - Seasonal temperature change Not Counted 40
- 20. Loss of CCW flow 35 1.30 14 1.56 49 1.49 84 1.40 200
- 21. Normal PORV activation 8
0.30 3
0.33 11 0.32 19 0.32 600
- 22. Pressurizer safety valve operation 11 0.41 9
1.00 28 0.85 39 0.65 83
- 23. Low Pressure Safety Injection No historical events 1
0.02 1
ULNRC-05938 December 13, 2012 Page 11 of 24 Transient Description Long Term (1984 - 2011)
Short Term (2002 -2011)
Projection (2011-2044)
Total (1984-2044)
Limiting Number
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- of Events Rate (event/yr)
- 24. Cold shutdown depressurization Not Counted 1
- 25. Reactor vessel/pressurizer vent 1
- 26. Zero load 200
- 27. Pressurization 400
- 28. Excess letdown heat exchanger operation Linear extrapolation 270 4.50 650
ULNRC-05938 December 13, 2012 Page 12 of 24 RAI 4.3-2a Table 2 Transients Used in the Fatigue Crack Growth Analyses Transient Description (Trans. # corresponds to the # in LRA Table 4.3-2 RAI 4.7.2-1:
RV Inlet Nozzle Flaw RAI 4.7.2-3: SWOL Repairs From LRA Table 4.3-2 Percent Accumulation Spray Nozzle Safety/Relief Nozzles Surge Nozzle Limiting Value Baseline Projection Normal 1a. Plant heatup at 100ºF/hr Pressurizer Heatup at 100°F/hr 200 200 200 200 200 29 65 32.50%
1b. Plant cooldown at 100ºF/hr Pressurizer Cooldown at 200°F/hr 200 200 200 200 29 65 32.50%
2a. Unit loading at 5% of full power per min 13,200 18,300 13,200 13,200 11,200 178 251 2.24%
2b. Unit unloading at 5%
of full power per min 18,300 13,200 13,200 13,200 184 276 2.09%
3a. Step increase 10% of full power 2,000 2,000 2,000 2,000 24 61 3.05%
3b. Step decrease 10% of full power 2,000 2,000 2,000 20 48 2.40%
- 4. Large step decrease with steam dump 200 200 6,350 200 200 6
13 6.50%
5a. Steady state fluctuations, Initial fluctuations 1.5E5 3.15E6 1E6 Transient does not result in the accumulation of fatigue usage.
5b. Steady state fluctuations, Random fluctuations 3.00E6
134 / 134 186 / 175 /
183 / 180 9.30% / 8.75%
9.15% / 9.00%
7a. Loop out of service, Normal loop shutdown 80 80 0
0 0%
ULNRC-05938 December 13, 2012 Page 13 of 24 Transient Description (Trans. # corresponds to the # in LRA Table 4.3-2 RAI 4.7.2-1:
RV Inlet Nozzle Flaw RAI 4.7.2-3: SWOL Repairs From LRA Table 4.3-2 Percent Accumulation Spray Nozzle Safety/Relief Nozzles Surge Nozzle Limiting Value Baseline Projection 7b. Loop out of service, Normal loop startup 70 70 0
0 0%
- 9. Boron concentration equalization 26,400 26,400 26,400 This is a load following transient, so the design number will not be approached in 60 yrs.
- 12. Refueling 80 80 17 39 48.75%
- 13. Turbine roll test 20 20 20 10 7
7 70.00%
- 14. Primary side leak test 200 200 200 200 50 8
10 20.00%
- 15. Secondary side leak test 200 80 80 80 1
4 5.00%
Upset
- 1. Loss of load (without immediate reactor trip) 80 80 6,130 80 80 0
1 1.25%
- 2. Loss of Power (with natural circulation in the RCS) 40 40 2,050 40 40 1
1 2.50%
- 3. Partial loss of flow (loss of one pump) 80 80 80 80 80 0
1 1.25%
4a. Reactor trip from full power, Without cooldown.
230 230 230 230 6166 8592 36.96%
4b. Reactor trip from full power, With cooldown, without safety injection 160 160 160 160 40 61 3.75%
4c. Reactor trip from full power, With cooldown, with safety injection 10 10 10 10 10 21 20.00%
- 5. Inadvertent RCS depressurization 20 20 20 2
5 25.00%
5a. Inadvertent RCS depressurization due to inadvertent auxiliary spray 10 570 10 0
1 10.00%
- 6. Inadvertent startup of an inactive RCS loop 10 10 10 0
1 10.00%
- 7. Control rod drop 80 80 80 80 0
1 1.25%
ULNRC-05938 December 13, 2012 Page 14 of 24 Transient Description (Trans. # corresponds to the # in LRA Table 4.3-2 RAI 4.7.2-1:
RV Inlet Nozzle Flaw RAI 4.7.2-3: SWOL Repairs From LRA Table 4.3-2 Percent Accumulation Spray Nozzle Safety/Relief Nozzles Surge Nozzle Limiting Value Baseline Projection
- 8. Inadvertent safety injection actuation 60 80 60 60 2
2 3.33%
- 9. Operating Basis Earthquake (20 earthquakes of 10 cycles each) 400 cycles 400 cycles 400 cycles 20 events 0 events 1 event 5.00%
- 10. Excessive Feedwater Flow 30 30 30 0
1 3.33%
Test
- 1. Primary side hydrostatic test 10 10 10 5
1 1
20.00%
Auxiliary
- 26. Zero Load 200 200 29*
65*
32.50%
Value set equal to the number of cooldowns.
ULNRC-05938 December 13, 2012 Page 15 of 24 RAI 4.3-2a Table 3 Transients Used in the Leak-Before-Break Analyses Transient Description (Tran. # corresponds to the # in LRA Table 4.3-2.
Reactor Coolant Loop Accumulator Lines Residual Heat Removal Lines LRA Table 4.3-2 Percent Accumulation Limiting Value Baseline Projection Normal 1a. Plant heatup at 100ºF/hr Pressurizer Heatup at 100°F/hr 200 200 29 65 32.50%
1b. Plant cooldown at 100ºF/hr Pressurizer Cooldown at 200°F/hr 200 200 29 65 32.50%
2a. Unit loading at 5% of full power per min 18,300 13,200 13,200 11,200 178 251 2.24%
2b. Unit unloading at 5% of full power per min 18,300 13,200 13,200 13,200 184 276 2.09%
3a. Step increase 10% of full power 2000 2,000 2,000 2,000 24 61 3.05%
3b. Step decrease 10% of full power 2000 2,000 2,000 2,000 20 48 2.40%
- 4. Large step decrease with steam dump 200 200 200 6
13 6.50%
5a. Steady state fluctuations, Initial fluctuations 1E6 3.2E6 3.2E6 1E6 Transient does not result in the accumulation of fatigue usage.
5b. Steady state fluctuations, Random fluctuations
134 / 134 186 / 175 /
183 / 180 9.30% / 8.75%
9.15% / 9.00%
8a. Unit loading between 0-15%
of full power 500 500 93 176 35.20%
8b. Unit unloading between 0-15% of full power 500 500 92 175 35.00%
- 12. Refueling 80 80 17 39 48.75%
- 13. Turbine roll test 10 20 20 10 7
7 70.00%
- 14. Primary Side Leak Test 50 50 8
10 20.00%
ULNRC-05938 December 13, 2012 Page 16 of 24 Transient Description (Tran. # corresponds to the # in LRA Table 4.3-2.
Reactor Coolant Loop Accumulator Lines Residual Heat Removal Lines LRA Table 4.3-2 Percent Accumulation Limiting Value Baseline Projection Upset
- 1. Loss of load (without immediate reactor trip) 80 80 80 0
1 1.25%
- 2. Loss of Power (with natural circulation in the RCS) 40 40 40 1
1 2.50%
- 3. Partial loss of flow (loss of one pump) 80 160*
80 0
1 1.25%
4a. Reactor trip from full power, without cooldown.
400 230 230 6166 8592 36.96%
4b. Reactor trip from full power, with cooldown, without safety injection 160 160 160 40 61 3.75%
4c. Reactor trip from full power, with cooldown, with safety injection 10 10 10 21 20.00%
- 5. Inadvertent RCS depressurization 20 20 20 2
5 25.00%
- 7. Control rod drop 80 80 80 0
1 1.25%
- 8. Inadvertent safety injection actuation 60 60 2
2 3.33%
Test
- 1. Primary Side Hydrostatic Test 5
5 1
1 20.00%
- 4. Cold Hydrostatic Test 10 10 Manufacturer test Auxiliary Trans
- 11. Accumulator actuation, accident operation 21 21 0
1 4.76%
- 12. Inadvertent accumulator blowdown 4
4 0
1 25.00%
- 13. RHR operation - plant cooldown 200 200 29 62 31.00%
- 14. High head safety injection (Loop A / B / C / D) 110 110 3 / 0 / 0 / 0 4 / 1 / 1 / 1 3.64% / 0.91%
0.91% / 0.91%
80 cycles from dead loop and 80 cycles from active loop
ULNRC-05938 December 13, 2012 Page 17 of 24 RAI 4.3-3a
Background:
In response to Part (a) of RAI 4.3-3 dated October 11, 2012, the applicant stated it experienced 22 events of Normal Transient #16a, [f]eedwater heaters out of service: One heater out of service, between 2000 and 2011. In its response to Part (b) of RAI 4.3-3, the applicant stated that in the past 9 years there have been 12 events of Normal Transient #16a and that a weighting factor of 1 for the long term and 3 for the short term was used. In addition, it was stated that the short-term period consists of the preceding 9 years.
Issue:
Based on the information provided, the staff noted that there were approximately 10 events of Normal Transient #16a that occurred from 2000 through 2001. Furthermore, LRA Table 4.3-2 states the 60-year projection is 106 cycles and the design-limiting value is 120 cycles. It is not clear to the staff whether using the occurrences from the preceding 9 years is conservative and represents the future trend for Normal Transient #16a, since the short-term occurrence would be 22 if a short-term period of 11 years was used. Since the applicant used the 60-year transient projections to support the disposition of the TLAAs evaluated in LRA Sections 4.7.2 and 4.7.7, the staff requires this information to determine if the projection for Transient #16a is appropriate and conservative.
Request:
a) Provide the yearly occurrence for Normal Transient #16a since the plant startup to 2011 to show the trend of this transient.
b) Justify that the 60-year projection for Normal Transient #16a, using a short-term period of the preceding 9 years (2002-2011), is conservative considering that 10 events occurred from 2000 through 2001.
Callaway Response a) Table 4 (refer to table following this response) provides the yearly occurrence for Normal Transient #16a since the plant startup to 2011.
b) The reason for the high frequency of the Normal Transient #16a (one feedwater heater out of service) transient between 2000 and 2001 is due to spurious isolation of the 3B low pressure heater due to the Hi-Hi level alarm. In response to this condition, Callaway replaced the stage 3 feedwater heaters in Refuel 11 (Spring 2001). While the replacement feedwater heater did not completely resolve the condition, it did change the accumulation rate. During Refuel 18 (Fall 2011) the float chamber was replaced with a new design. There have been no Hi-Hi level alarms since installation of the new float chamber design.
The use of a 9 year short term period is acceptable for Normal Transient #16a because:
- 1. The transient is monitored by the Fatigue Monitoring Program which will detect any changes in the accumulation rate in the future and incorporate the change into the projection. This will ensure that the design number of events is not exceeded.
ULNRC-05938 December 13, 2012 Page 18 of 24
- 2. The projection will be 120 events in 60 years if the additional 2 years are included in the Short Term period results. This is not a drastic increase in the projected number of events, i.e. the projected number of events will not approach the design number (120 events) within the next three fuel cycles (the corrective action limit of the Fatigue Monitoring program).
- 3. The 60 year projection for Normal Transient #16a does not affect the TLAAs evaluated in LRA Sections 4.7.2 and 4.7.7. As shown in the Tables 2 and 3 in the RAI 4.3-2a response, Normal Transient #16a is not used in these analyses; therefore the projection does not affect the disposition of these TLAAs.
Corresponding Amendment Changes No changes to the License Renewal Application (LRA) are needed as a result of this response.
ULNRC-05938 December 13, 2012 Page 19 of 24 RAI 4.3-3a Table 4 Baseline and Projection Summary of Transient 16a Transient Description Long Term Accumulation (1984 - 2011)
Short Term Accumulation (2002 - 2011)
Future Accumulation (2011 - 2043) 60 Year Projection (1983 - 2043)
- of Events LTR
(#/yr)
- of Events STR
(#/yr)
- of Events Rate
(#/yr)
- of Events Rate
(#/yr) 16a. Feedwater heaters out of service:
One heater out of service 56 2.07 12 1.33 50 1.52 106 1.77 Transient Description Long Term Accumulation (1984 - 2011)
Short Term Accumulation (2000 - 2011)
Future Accumulation (2011 - 2043) 60 Year Projection (1983 - 2043)
- of Events LTR
(#/yr)
- of Events STR
(#/yr)
- of Events Rate
(#/yr)
- of Events Rate
(#/yr) 16a. Feedwater heaters out of service:
One heater out of service 56 2.07 22 2.0 64 1.52 120 1.77
ULNRC-05938 December 13, 2012 Page 20 of 24 RAI 4.3-15a
Background:
In response to Part (b)(i) of RAI 4.3-15 dated October 11, 2012, the applicant stated that the letdown heat exchanger tubesheet cumulative usage factor (CUF) contribution by the [l]etdown flow decrease and return to normal transient would increase from 0.169 [0.144+0.018+0.007] to 0.2535 [1.5x(0.144+0.018+0.007)]. In addition, it is stated that this will increase the CUF from 0.910 to 0.995 and the response also provided the CUF contribution for each transient pair.
Issue:
Based on staffs review of Table 4 in the applicants response to RAI 4.3-15, the CUF contribution from the 3rd transient pair #6 [l]etdown flow decrease and return to normal (490°F to 290°F) and
- 6 [l]etdown flow decrease and return to normal (140°F to 380°F) would increase to 0.224
[0.144/1800)x2800]. Thus, the CUF contribution by the [l]etdown flow decrease and return to normal transient would increase from 0.169 to 0.2615 [0.224+1.5x(0.018+0.007)], as opposed to the applicants conclusion of 0.2535.
The staff noted that this will increase the CUF from 0.910 to over 1.00. The staff noted that the CUF value remains less than 1.0, which is the ASME Code Section III CUF requirement. The staff noted that the applicant must demonstrate that the re-calculated CUF value will be less than 1.0 through the period of extended operation to support the TLAA disposition in accordance with 10 CFR 54.21(c)(1)(ii).
Request:
Justify that the re-calculated CUF value is less than 1.0 assuming 3,000 cycles of letdown flow decrease and return to normal transient will occur through the period of extended operation.
Callaway Response The fatigue table of the letdown heat exchanger tubesheet contains a conservatism which can accommodate an increase of the CUF contribution to 0.224. The transient pairing of letdown flow increase and return to normal at (405°F to 279°F) & (290°F to 405°F) includes 24000 events. The transient pairing only needs to include 22000 events since 2000 (290°F to 405°F) events are paired with the Letdown flow decrease and return to normal transient. When the number of Letdown flow decrease and return to normal events is increased to 3000, this number of paired events would be decreased to 21000 events and the CUF would decrease to 0.992.
This is demonstrated in Table 5 (refer to table following this response). LRA Section 4.3.8 has been revised as shown by Amendment 17 in Enclosure 2 to correct the projected CUF for the letdown heat exchanger tubesheet.
Corresponding Amendment Changes Refer to the Enclosure 2 Summary Table "Amendment 17, LRA Changes for a description of LRA changes with this response.
ULNRC-05938 December 13, 2012 Page 21 of 24 RAI 4.3-15a Table 5 Tubesheet Fatigue Table for Letdown Heat Exchanger Transient Pairs(1)
Salt (psi)
Spec. #
(n)
Allow.
(N)
U
- 2 Charging flow shut-off and prompt return
- 2 Charging flow shut-off and prompt return 100488 100 1400 0.071
- 1 Letdown flow shut-off and prompt return
- 6 Letdown flow decrease and return to normal (140°F to 380°F) 62672 200 11000 0.018
- 6 Letdown flow decrease and return to normal (490°F to 290°F)
- 6 Letdown flow decrease and return to normal (140°F to 380°F) 60170 1800(2) 12500 0.144(2) 2800(3) 0.224(3)
- 6 Letdown flow decrease and return to normal (490°F to 290°F)
- 1 Letdown flow shut-off and prompt return 58734 200 16000 0.013
- 4 Charging flow decrease and return to normal
- 4 Charging flow decrease and return to normal 45074 24000 47000 0.511
- 6 Letdown flow decrease and return to normal (290°F to 140°F)
- 7 Letdown flow increase and return to normal (290°F to 405°F) 38514 2000(2) 110000 0.018(2) 3000(3) 0.027(3)
- 7 Letdown flow increase and return to normal (405°F to 279°F)
- 7 Letdown flow increase and return to normal (290°F to 405°F) 33466 24000(2) 300000 0.080(2) 22000(4) 0.073(4) 21000(3) 0.070(3)
- 7 Letdown flow increase and return to normal (405°F to 279°F)
- 6 Letdown flow decrease and return to normal (380°F to 490°F) 34529 2000(2) 300000 0.007(2) 3000(3) 0.010(3)
- 7 Letdown flow increase and return to normal (279°F to 198°F)
- 7 Letdown flow increase and return to normal (198°F to 290°F) 22816 24000 1000000 0.024
- 5 Charging flow increase and return to normal
- 5 Charging flow increase and return to normal 20566 24000 1000000 0.024 Total U 0.910(2) 0.903(4) 0.992(3) 1 Transient numbers correspond to those identified in LRA Section 4.3.8, Fatigue Analyses of Class 2 Heat Exchangers.
2 Assumes 2000 Letdown flow decrease and return to normal events.
3 Assuming 3000 Letdown flow decrease and return to normal events.
4 This value corrects an error in the original design report describe in the text of this RAI response.
ULNRC-05938 December 13, 2012 Page 22 of 24 RAI 4.7.9-5a
Background:
In its letter dated September 20, 2012, the applicant responded to RAI 4.7.9-5, which addresses whether the applicants TLAA for steam generator tube wear adequately considers potential effects of flow rate calculation errors or flow correction factor errors on steam generator tube wear.
In its response, the applicant stated that its analysis includes 2 percent for measurement uncertainties. The applicant also stated that to ensure that the plant remains within this uncertainty, plant instrumentation undergoes periodic surveillance and calibration and the uncertainty is incorporated into the setpoints. The applicant further stated that critical plant instrumentation whose failure could result in a violation of this uncertainty is incorporated into the plants technical specifications and/or procedures.
Issue:
In its review of the applicants response, the staff noted that the applicant did not clearly address how the 2 percent measurement uncertainty bounds the feedwater flow rate measurements. In addition, the staff seeks clarification regarding the inspection method and frequency used for the periodic surveillance of the feedwater flow meters.
The staff further needs confirmation as to whether Regulatory Issue Summary (RIS) 2007-24, NRC Staff Position on Use of the Westinghouse CROSSFLOW Ultrasonic Flow Meter for Power Uprate or Power Recovery, dated September 27, 2007 (ADAMS Accession No. ML063450261),
is an applicable concern for the applicants flow measurement.
Request:
a) Clarify what parameter in the applicants analysis involves the 2 percent measurement uncertainty (e.g., reactor thermal power or feedwater flow rate). As part of the response, describe how the applicants TLAA considers this uncertainty, in order to confirm that the uncertainty does not affect the assumptions or technical bases of the TLAA.
b) Clarify what inspection method and frequency are used in the applicants periodic surveillance for the feedwater flow meters. In addition, clarify why these inspections can ensure that aging degradation of the feedwater flow meters does not increase the uncertainty above the specified uncertainty range.
c) Clarify whether the RIS 2007-24 is an applicable concern to the applicants TLAA.
Specifically, clarify whether the applicant uses the CROSSFLOW ultrasonic flow meter to calibrate the feedwater flow meters or to determine feedwater flow rates.
Callaway Response a) This represents the two percent for measurement uncertainties on the reactor thermal power analysis specified for the thermal hydraulic analysis. The tube wear TLAA actually selects an uprated power level as the bounding case for input into the analysis. The uprated power level is 3864 MWth versus the nominal power level of 3579 MWth or 3651 MWth with a 2%
uncertainty incorporated.
ULNRC-05938 December 13, 2012 Page 23 of 24 b) Each feedwater flow venturi is visually inspected and cleaned as necessary at least once per 18 months in accordance with FSAR 16.4.9.1.1. The inspections address the potential fouling of the feedwater venturi which could bias the results from the precision heat balance in a non-conservative manner. This frequency has been shown by experience to be acceptable for ensuring the continued performance of the feedwater venturi which specifies flow coefficient accuracy of +/-0.25% maximum. This accuracy is used in secondary calorimetric calculations performed at the start of each operating cycle to obtain a precision measurement of RCS flow to satisfy Technical Specifications 3.3.1.2 and 3.4.1.4 requirements, which ensures the plant is within the 2% measurement uncertainty.
c) RIS 2007-24 is not applicable to Callaway. Callaway does not have the CROSSFLOW ultrasonic flow meter installed.
Corresponding Amendment Changes No changes to the License Renewal Application (LRA) are needed as a result of this response.
ULNRC-05938 December 13, 2012 Page 24 of 24 RAI 4.7.9-2a
Background:
In its letter dated September 20, 2012, the applicant responded to RAI 4.7.9-2, which addresses whether the estimated wear rate of the applicants TLAA for steam generator tube wear is in agreement with the applicants inspection results.
The applicants response indicates that the total number of the steam generator tubes is 23,488, including one tube that was plugged pre-service. In its response, the applicant also stated that the anti-vibration bar (AVB) wear is the most active wear mechanism as observed by the steam generator tube inspections, which are described in the letter dated May 17, 2012. The applicant further state that only 1 percent (232 tubes) of all the steam generator tubes show indications of AVB wear.
Issue:
In contrast with the total number (23,488) of the steam generator tubes described in the applicants response, Table 1 in the applicants letter dated May 17, 2012, indicates that the total number of the steam generator tubes is 22,144, including the tube plugged pre-service.
Therefore, the staff needs clarification of the total number of steam generator tubes.
In addition, the applicants response does not address the number of the steam generator tubes that show wear indications due to tube support plate wear. The staff needed to confirm whether a total of 258 steam generator tubes have wear indications due to AVB wear and tube support plate wear as described in the applicants inspection report dated May 17, 2012.
Request:
Clarify the total number of steam generator tubes. In addition, confirm whether a total of 258 steam generator tubes have wear indications due to AVB wear and tube support plate wear as described in the applicants inspection report dated May 17, 2012.
Callaway Response There are 5,872 tubes in each steam generator for a total of 23,488 tubes. The number provided in the inspection report dated May 17, 2012 (ML12139A275) is a summary of the number of tubes that were eddy current tested.
Table 2 of the inspection report dated May 17, 2012 (ML12139A275) correctly presents the number of tube support plate indications. There are a total of 258 steam generator tubes that show wear indications due to AVB wear and/or tube support plate wear.
Corresponding Amendment Changes No changes to the License Renewal Application (LRA) are needed as a result of this response.
ULNRC-05938 December 13, 2012 Page 1 of 39 Amendment 17, LRA Changes Summary Table Affected LRA Section LRA Page Table 4.3-2 4.3-6 through 18 Table 4.3-3 4.3-19 through 24 Section 4.3.4 4.3-31 through 35 Table 4.3-6 4.3-35 through 37 Table 4.3-7 4.3-36 through 38 Section 4.3.8 4.3-42 Section 4.8 4.8-1 and 4.8-2 Section A3.2.3 A-27 Section B3.1 B-127 through B-130
ULNRC-05938 December 13, 2012 Page 2 of 39 Callaway Plant Unit 1 Page 4.3-6 License Renewal Application Amendment 17 Table 4.3-2 Transient Accumulations and Projections Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments Normal Condition Transients 1a. Plant heatup at 100ºF/hr Pressurizer Heatup at 100°F/hr 200 200 29 65(2) 1b. Plant cooldown at 100ºF/hr Pressurizer Cooldown at 200°F/hr 200 200 29 65(2) 2a. Unit loading at 5% of full power per min 13200 11200 178 252251(2)
The limiting value is from the RPV and CETNA fatigue analyses. 2000 transients are allocated to normal transient 11, Reduced Temperature Return to Power.
2b. Unit unloading at 5% of full power per min 13200 13200 184 276(2) 3a. Step increase of 10% of full power 2000 2000 24 6261 3b. Step decrease of 10% of full power 2000 2000 20 48
- 4. Large step decrease with steam dump 200 200 6
13(2)
ULNRC-05938 December 13, 2012 Page 3 of 39 Callaway Plant Unit 1 Page 4.3-7 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments 5a. Steady state fluctuations, Initial fluctuations 1.5 E5 1 E6 Not Counted (NC)
NC The limiting value is from the reactor coolant loop (RCL) leak-before-break analysis described in Section 4.7.7.
These fluctuations are assumed to occur only during the first 20 full power months of operation, therefore they do not need to be counted.
5b. Steady state fluctuations, Random fluctuations 3.0 E6 NC NC The limiting value is from the RCL leak-before-break analysis described in Section 4.7.7.
Transient does not result in the accumulation of fatigue usage.
134 / 134 186 / 175 /
183 / 180 -
7a. Loop out of service, Normal loop shutdown 80 80 0
0(2)
Callaway is not licensed for N-1 loop operation.
7b. Loop out of service, Normal loop startup 70 70 0
0(2) 8a. Unit loading between 0 15% of full power 500 500 93 176 8b. Unit unloading between 0 15% of full power 500 500 92 175
- 9. Boron concentration equalization 26400 26400 NC NC The number is based on 2 load changes per day (1 loading & 1 unloading) for 40 years with a 90% capacity factor. Callaway does not load follow and will not approach the limit.
ULNRC-05938 December 13, 2012 Page 4 of 39 Callaway Plant Unit 1 Page 4.3-8 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments 10a1. Reactor coolant pump startup and shutdown cold condition, RCS venting 800 3800 RCP A 139 RCP B 142 RCP C 139 RCP D 122 RCP A 217216 RCP B 221220 RCP C 220219 RCP D 195194 The limiting value is from the RSG fatigue analyses described in Section 4.3.2.3.
The RCP startup and shutdown transients are monitored as a summation of all RCP startup and shutdown.
Monitoring this group of transients is acceptable because the fatigue analyses treat these transients in this manner.
10a2. Reactor coolant pump startup and shutdown cold condition, RCS heatup, cooldown 200 10b. Reactor coolant pump startup and shutdown pump restart condition, Hot functional, stops, starts 500 10c. Reactor coolant pump startup and shutdown hot condition, Transients and misc.
2500
- 11. Reduced temperature return to power 2000 2000 0
0(2)
Maneuver is not utilized because Callaway does not load follow.
- 12. Refueling 80 80 17 39(2)
- 13. Turbine roll test 20 10 7
7(2)
The limiting value is from the RCL leak before break analysis described in Section 4.7.7.
Test performed during initial startup and no more tests are expected.
- 14. Primary side leak test 200 50 8
10(2)
The limiting value is from the RCL leak before break analysis described in Section 4.7.7.
Baseline result assumes 1 event for each opening of the primary system prior to 1995 in addition to the documented event.
ULNRC-05938 December 13, 2012 Page 5 of 39 Callaway Plant Unit 1 Page 4.3-9 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments
- 15. Secondary side leak test 80 80 1
4 16a. Feedwater heaters out of service:
One heater out of service 120 120 56 106 Baseline The manually counted baseline results for 1983 to 2000 are judged to be conservative based on a review of instrumentation data available from 2000-2011.
16b. Feedwater heaters out of service:
One bank out of service 120 120 12 16
Baseline results assume transient coincides with refueling, plus pre-start test, and an additional RF17 event.
Upset Condition Transients
- 1. Loss of load (without immediate reactor trip) 80 80 0
1(2)
- 2. Loss of power (with natural circulation in the RCS) 40 40 1
1(2)
The 60 year projection is kept at 1 event because this is the value analyzed in the EAF calculations. The justification is provided in Section 4.3.4.
- 3. Partial loss of flow (loss of one pump) 80 80 0
1(2) 4a. Reactor trip from full power, without cooldown.
230 230 6166 8592(2) 4b. Reactor trip from full power, with cooldown, without safety injection 160 160 40 51(2)
ULNRC-05938 December 13, 2012 Page 6 of 39 Callaway Plant Unit 1 Page 4.3-10 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments 4c. Reactor trip from full power, with cooldown, with safety injection 10 10 10 1(2)
-The 60 year projection is kept at 1 event because this is the value analyzed in the EAF calculations. The justification is provided in Section 4.3.4.
4d. Reactor trip from full power, with no inadvertent cooldown - emergency overspeed 20 20 NC NC Included as part of upset transient 4a, Reactor Trip from Full Power, without Cooldown.
- 5. Inadvertent RCS depressurization 20 20 2
5(2) 5a. Inadvertent RCS depressurization due to inadvertent auxiliary spray 10 0
1(2)
The limiting value is from the Pressurizer, Class 1 Piping, Surge line, and NUREG/CR-6260 fatigue analyses.
- 6. Inadvertent startup of an inactive RCS loop 10 10 0
1(2)
- 7. Control rod drop 80 80 0
1(2)
- 8. Inadvertent safety injection actuation 60 60 2
2(2)
The 60 year projection is kept at 2 events because this is the value analyzed in the EAF calculations. The justification is provided in Section 4.3.4.
- 9. Operating basis earthquake (20 earthquakes of 10 cycles each) 200 cycles 20 events 0
1(2)
- 10. Excessive feedwater flow 30 30 0
1(2)
- 11. RCS cold overpressurization 10 10 0
1
ULNRC-05938 December 13, 2012 Page 7 of 39 Callaway Plant Unit 1 Page 4.3-11 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments Test Condition Transients
- 1. Primary side hydrostatic test 10 5
1 1(2)
The limiting value is from the RCL leak before break analysis described in Section 4.7.7.
Test performed during initial startup and no more tests are expected.
- 2. Secondary side hydrostatic test 10 10 1
1 Test performed during initial startup and no more tests are expected.
- 3. Tube leakage test 800 800 0
1 Tests performed at 200, 400, 600, and 800 psig
- 4. Cold hydrostatic test 10 NC NC The limiting value is from the RCL leak before break analysis described in Section 4.7.7.
Manufacturer test, this transient is not required to be monitored for fatigue by article NB-3226.e of the ASME code.
ULNRC-05938 December 13, 2012 Page 8 of 39 Callaway Plant Unit 1 Page 4.3-12 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments Auxiliary Transients
- 1. Normal charging and letdown shutoff and return to service (Alt / Normal) 36 0 / 4 1 / 56 This transient is applicable to Class 1 piping, NUREG/CR-6260 locations described in Section 4.3.4, and the Class 2 heat exchanger analyses described in Section 4.3.8.
The charging lines include a reduction to 60% of the design number based on alternating between the normal and alternate charging paths.
The alternate charging line is now used to prevent transients in the normal charging line.
- 2. Letdown flow shutoff with prompt return to service (Alt / Normal) 120 1 / 11 4 / 23
- 3. Letdown flow shutoff with delayed return to service (Alt / Normal) 12 2 / 12 8 / 2124
- 4. Charging flow shutoff with prompt return to service (Alt / Normal) 12 0 / 12 1 / 2427
- 5. Charging flow shutoff with delayed return to service (Alt / Normal) 12 1 / 9 4 / 1415
- 6. Charging flow decrease and return to normal 14400 NC NC This transient is applicable to NUREG/CR-6260 locations described in Section 4.3.4, Class 1 CVCS piping, and the Class 2 heat exchanger analyses described in Section 4.3.8.
The number is based on 2 load changes per day (1 loading & 1 unloading) for 40 years and 80% capacity factor. The charging lines include an additional 60%
reduction based on rotating between the normal and alternate charging paths. Callaway does not load follow and will not approach the limit during a 60 year plant life as described in Section 4.3.8.
- 7. Charging flow increase and return to normal 14400 NC NC
ULNRC-05938 December 13, 2012 Page 9 of 39 Callaway Plant Unit 1 Page 4.3-13 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments
- 8. Letdown flow decrease and return to normal 1200 NC NC This transient is applicable to NUREG/CR-6260 locations described in Section 4.3.4, Class 1 CVCS piping, and Class 2 heat exchanger analyses described in Section 4.3.8.
The transient is not a normal operating transient, but is included for conservatism and is not a significant contributor to fatigue as described in Section 4.3.8. The charging lines include an additional 60% reduction based on rotating between the normal and alternate charging paths.
ULNRC-05938 December 13, 2012 Page 10 of 39 Callaway Plant Unit 1 Page 4.3-14 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments
- 9. Letdown flow increase and return 14400 NC NC This transient is applicable to NUREG/CR-6260 locations described in Section 4.3.4, Class 1 piping, and the Class 2 heat exchanger analyses described in Section 4.3.8.
The charging lines include an additional 60% reduction based on rotating between the normal and alternate charging paths. Callaway does not load follow and will not approach the limit during a 60 year plant life as described in Section 4.3.8.
- 10. Load follow boration 24000 NC NC The transient is specified for the Class 2 heat exchanger. The number is based on 2 load changes per day (1 loading & 1 unloading) for 40 years and 80%
capacity factor. Callaway does not load follow and will not approach the limit during a 60 year plant life as described in Section 4.3.8.
ULNRC-05938 December 13, 2012 Page 11 of 39 Callaway Plant Unit 1 Page 4.3-15 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments
- 11. Accumulator actuation, accident operation 21 0
1 Transients are from the Class 1 valve fatigue analyses and accumulator leak before break analysis.
- 12. Inadvertent accumulator blowdown 4
0 1
- 13. RHR operation - plant cooldown 200 29 6562 Transients are from the Class 1 valve fatigue analyses and accumulator leak before break analysis. Baseline assumes a RHR event coincides with plant cooldown.
- 14. High head safety injection (Loop A / B / C / D) 110 3 / 0 / 0 / 0 4 / 1 / 1 / 1 Transients are from the Class 1 valve fatigue analyses and accumulator leak before break analysis.
- 15. Seal injection flow temperature change 180 29 6563 This transient is only applicable to the RCPs fatigue analysis described in Section 4.3.2.1.
Baseline assumes event coincides with plant cooldown.
- 16. Elevated seal water injection temperature 200 55 124121 This transient is only applicable to the RCPs described in Section 4.3.2.1.
Baseline assumes event coincides with plant cooldown as specified plus one event per summer to account for seasonal temperature changes.
- 17. Loss of seal injection flow 40 3
7 This transient is only applicable to the RCPs described in Section 4.3.2.1.
ULNRC-05938 December 13, 2012 Page 12 of 39 Callaway Plant Unit 1 Page 4.3-16 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments
An administration action limit of 150 cycles is imposed as described in Section 4.3.2.1.
Baseline assumes event coincides with plant cooldown in addition to the documented events.
Justification for not counting described in Section 4.3.2.1.
- 20. Loss of CCW flow 200 35 8481 This transient is only applicable to the RCPs and Class 2 heat exchangers described in Sections 4.3.2.1 and 4.3.8.
Baseline assumes event coincides with plant cooldown in addition to the documented events.
- 21. Normal PORV activation 600 8
19 The limiting value is from the PORV fatigue analysis.
- 22. Pressurizer safety valve (PSV) operation 83 11 3937 The limiting value is from the pressurizer safety valve fatigue analysis.
No operational transients identified. Baseline result based on test events.
ULNRC-05938 December 13, 2012 Page 13 of 39 Callaway Plant Unit 1 Page 4.3-17 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments
- 23. Low pressure safety injection 1
0 1
The limiting value is from the RHR pump discharge check valve fatigue analysis.
Faulted condition, e.g. LOCA, which is combined with high head safety injection, transient #14, in the fatigue analysis. Per the Code, it is not required to be included in the fatigue analysis.
- 24. Cold shutdown depressurization 1
NC NC The limiting value is from the PSV fatigue analysis.
Has a negligible contribution to fatigue.
- 25. Reactor vessel/pressurizer vent 1
NC NC The limiting value is from the PSV fatigue analysis.
Has a negligible contribution to fatigue.
- 26. Zero load 200 NC NC(2)
The limiting value is from the fatigue crack growth analysis performed in support of the pressurizer SWOL described in Section 4.7.2 and NUREG/CR-6260 locations described in Section 4.3.4.
Cycle counting for the 200 heatups and cooldowns will capture the 200 zero load states.
ULNRC-05938 December 13, 2012 Page 14 of 39 Callaway Plant Unit 1 Page 4.3-18 License Renewal Application Amendment 17 Transient Description FSAR Design Cycles Design Limiting Value(1)
Baseline (1983 -
2011)
Projected Events for 60 Years Comments
- 27. Pressurization NS 400 NC NC The transient is specified for the Class 2 heat exchangers but can be satisfactorily managed by counting other transients as described in Section 4.3.8.
- 28. Excess letdown heat exchanger operation NS 650 122 270 The transient is specified at 100 events for the Class 2 heat exchangers but was analyzed for 650 events as described in Section 4.3.8.
1 Design limiting value identified through a review of the design and licensing basis analyses.
2 These values will also be incorporated into the cycle counting action limits to ensure that the EAF results for the hot leg surge nozzle shown in Section 4.3.4 are not exceeded. This was necessary because of limitations in the CBF algorithm to account for insurge and outsurge transients.
ULNRC-05938 December 13, 2012 Page 15 of 39 Callaway Plant Unit 1 Page 4.3-19 License Renewal Application Amendment 17 Table 4.3-3 ASME Class 1 Fatigue Analyses Under the Fatigue Monitoring Program Component CUF Reactor Pressure Vessel Components Inlet Nozzles Support 0.0795 0.0306 Outlet Nozzles Support 0.1078 0.0205 Head Flanges 0.0155 Vessel Flange 0.0196 Studs Studs Installed in Holes with Damaged Threads 0.4780 0.75 CRDM Housing 0.1093 Bottom Head to Shell Junction 0.0070 Bottom-Mounted Instrument Tubes 0.3184 Vessel Wall Transition 0.0105 Core Support Lugs 0.0617 Head Adapter Plug 0.0036 Core Exit Thermocouple Nozzle Assembly Head Port Adapter 0.123 Clamp 0.21 Upper Nozzle Housing 0.37 Drive Sleeve 0.02 Clamp Bolt 0.92 Reactor Coolant Pump Pressure Boundary Components Weir Plate 0.440
ULNRC-05938 December 13, 2012 Page 16 of 39 Callaway Plant Unit 1 Page 4.3-20 License Renewal Application Amendment 17 Table 4.3-3 ASME Class 1 Fatigue Analyses Under the Fatigue Monitoring Program Component CUF Casing at Discharge Nozzle Juncture 0.915 Bolting ring 0.086 Main closure bolts (studs) 0.45 Casing at Large Support Feet Junctures 0.083 Thermal Barrier Flange at Component Cooling Water Connection (Flange Holes) 0.9334 Pressurizer Components(1)
Spray Nozzle 0.411 Upper Head 0.928 Surge Nozzle 0.963(1) 0.034(2) (Dry) /
0.032(2) (Wet)
Safety and Relief Nozzle 0.169 Support Skirt and Flange 0.734(1) 0.7284(2,3) (Dry)
Lower Head 0.112(1) 0.098(2) (Dry) /
0.0190.022(2) (Wet)
Heater Penetration (Heater Well) 0.128(1) 1.441(2) (Dry) /
0.562(2) (Wet) 0.8146(2,3) (Dry) /
0.0103(2,3) (Wet)
Seismic Support Lug 0.444 Shell at Support Lug 0.992
ULNRC-05938 December 13, 2012 Page 17 of 39 Callaway Plant Unit 1 Page 4.3-21 License Renewal Application Amendment 17 Table 4.3-3 ASME Class 1 Fatigue Analyses Under the Fatigue Monitoring Program Component CUF Trunnion Buildup 0.567 Instrument Nozzle 0.236(14) 0.746(2) (Dry) /
0.0190.439(2) (Wet) 0.0103(2,3) (Wet)
Manway Bolt 0.915 Manway Pad 0.141 Valve Support Bracket 0.118 Class 1 Valves Pressurizer Safety Valves 0.018 PORV 0.139 PORV Solenoid 0.68 Loop SI / RHR Check Valves 0.17 RHR and SI System Loop 2 & 3 Recirculation Supply Header Check Valves 0.17 RHR Pumps to RCS Cold Leg Check Valves 0.1647 RCS Cold Leg SI Accumulator Check Valves 0.26 SI Accumulator Outlet Upstream Check Valves 0.26 RHR Pump Suction Isolation Valves 0.64 RCS Hot Leg to RHR Pump Isolation Valves 0.64
ULNRC-05938 December 13, 2012 Page 18 of 39 Callaway Plant Unit 1 Page 4.3-22 License Renewal Application Amendment 17 Table 4.3-3 ASME Class 1 Fatigue Analyses Under the Fatigue Monitoring Program Component CUF Class 1 Piping Pressurizer Surge Line 0.099 Includes the effects of thermal stratification.
Spray/Aux. Spray Loops 1 & 2 0.84 Pressurizer Safety and Relief Valve Piping 0.975 Pressurizer Relief Valve Piping 0.970 Hot Leg 0.95 Crossover Leg 0.50 Cold Leg 0.37 Drain Line Loop 2 0.95 Normal Letdown/Drain Line Loop 3 0.95 Drain Line Loop 1 0.01 Excess Letdown/Drain Line Loop 4 0.191 Normal/Alternate Charging - Loops 1 & 4 0.93 Seal Water Injection Loop 1 0.066 Seal Water Injection Loop 2 0.066 Seal Water Injection Loop 3 0.114 Seal Water Injection Loop 4 0.067 RHR Loops 1 & 4 Suction Line 0.296 Loop 1 Hot Leg Safety Injection Line 0.661
ULNRC-05938 December 13, 2012 Page 19 of 39 Callaway Plant Unit 1 Page 4.3-23 License Renewal Application Amendment 17 Table 4.3-3 ASME Class 1 Fatigue Analyses Under the Fatigue Monitoring Program Component CUF Loop 4 Hot Leg Safety Injection Line 0.110 Accumulator Lines - Loops 1, 2, 3, & 4 0.980 SI Hot Leg Loops 2 & 3 0.090 Boron Injection Header 0.773 Boron Injection Header Lines - Loops 1, 2, 3, & 4 0.930 Class 1 Piping Nozzles and Thermowells Pressurizer Spray Nozzle-Cold Leg Loop 1 & 2 0.84 Pressurizer Surge Nozzle-Hot Leg Loop 4 0.30 Includes the effects of thermal stratification.
Drain Nozzle-Crossover Leg Loop 1 & 2 0.70 Cold Leg Thermowells 0.025 Hot Leg Thermowells 0.017 Mid-Loop Level Tap from Hot Leg Loops 1 & 4 0.327 Hot Leg RTD Scoop Nozzles 0.65 Cold Leg RTD Nozzles 0.15 Excess Letdown-Crossover Leg Loop 4 0.804 Normal Letdown-Crossover Leg Loop 3 0.10 Normal/ Alternate Charging Nozzle-Cold Leg Loop 1 & 4 0.95(5) 0.90(6)
SIS Nozzle-Hot Leg Loops 2 & 3 0.10 Boron Injection Header Nozzles-Cold Leg Loops 1, 2, 3, & 4 0.999
ULNRC-05938 December 13, 2012 Page 20 of 39 Callaway Plant Unit 1 Page 4.3-24 License Renewal Application Amendment 17 Table 4.3-3 ASME Class 1 Fatigue Analyses Under the Fatigue Monitoring Program Component CUF Accumulator Nozzle-Cold Leg Loops 1, 2, 3, & 4 0.95 RHR Nozzle-Hot Leg Loops 1 & 4 0.81 Pressurizer Thermowells 0.0 Spray Line Thermowells 0.021 Surge Line Thermowells 0.020 1
Original fatigue usage used as for common basis comparison in the EAF screening evaluation. It is provided in Table 4.3-7. The identification of CUFs as Wet or Dry is meant to identify those CUFs which are affected by environmentally assisted fatigue (EAF) discussed in Section 4.3.4.
2 Includes the results from evaluation of revised insurge-outsurge transients and further refined analysis. Based on an elastic analysis which includes the revised insurge-outsurge transients. The wet surfaces are used in the EAF screening shown in Table 4.3-7 to provide a common basis comparison with the other CUFs in the PZR Lower Head thermal zone, i.e. all CUFs are based on an elastic analysis.
3 Based on an elastic-plastic analysis which includes the revised insurge-outsurge transients.
4 CUF excludes the revised insurge-outsurge transients. It provides a common basis comparison for the instrument nozzle in the Pressurizer Upper Head thermal zone.
5 CUF used in the EAF screening in Table 4.3-7 to provide a common basis comparison with other CUFs in the CVCS - Charging thermal zone, i.e. all CUFs assume a nominal letdown flow (75 gpm).
6 Includes the results from the evaluation of high letdown flow.
ULNRC-05938 December 13, 2012 Page 21 of 39 Callaway Plant Unit 1 Page 4.3-31 License Renewal Application Amendment 17 4.3.4 Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)
The NRC concluded that effects of the reactor coolant environment might need to be included in the calculated fatigue life of components, and opened three generic safety issues to address this question, all finally closed to a single Generic Safety Issue 190. Subsequent research and studies refined the methods, which no longer use the interim fatigue curves of NUREG/CR-5999 but calculate an environmental fatigue effect multiplier Fen, which depends on material type, temperature, strain rate, and dissolved oxygen; and for carbon and low-alloy steel, sulfur content.
NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants states that The applicant's consideration of the effects of coolant environment on component fatigue life for license renewal is an area of review, noting the staff recommendation that the samples in NUREG/CR-6260 should be evaluated considering environmental effects for license renewal.
The GSI-190 review requirements are therefore imposed by the Standard Review Plan and do not depend on the individual plant licensing basis. Callaway addressed GSI-190 review requirements by assessing the environmental effect on fatigue at the NUREG/CR-6260 locations for the newer-vintage Westinghouse Plant.
NUREG/CR-6260 identifies seven sample locations for newer vintage Westinghouse plants which need to consider the effects of reactor coolant environment on component fatigue life for license renewal:
- 1. Reactor Vessel Lower Head to Shell Juncture
- 2. Reactor Vessel Primary Coolant Inlet Nozzle
- 3. Reactor Vessel Primary Coolant Outlet Nozzle
- 4. Hot Leg Surge Nozzle
- 5. Charging Nozzles
- 6. Safety Injection Nozzles
- 7. Residual Heat Removal Line Inlet Transition Table 4.3-6, Summary of Fatigue Usage Factors at NUREG/CR 6260 Sample Locations is includes a summary of environmentally-assisted fatigue (EAF) of the NUREG/CR-6260 locations. The Fen relationships are calculated from NUREG/CR-6583 for carbon and low-alloy steels and from NUREG/CR-5704 for stainless steels, as appropriate for the material at each of these locations.
The NUREG/CR-6260 locations in Table 4.3-6, Summary of Fatigue Usage Factors at NUREG/CR 6260 Sample Locations with an EAF CUF (Uen) below 1.0, when using the design basis CUF and the maximum Fen, require no further analysis. Three of the NUREG/CR-6260 locations, (1) RPV lower head to shell juncture, (2) RPV inlet nozzle, and (3) RPV outlet nozzles meet this criterion (Reference 4). All three locations are low alloy steel locations.
ULNRC-05938 December 13, 2012 Page 22 of 39 Callaway Plant Unit 1 Page 4.3-32 License Renewal Application Amendment 17 The maximum Fen for low alloy steel assumes the dissolved oxygen level to be less than 0.05 ppm, which corresponds to a low oxygen environment. This is consistent with the Callaway primary chemistry program, which maintains RCS hydrogen level at 25 to 50 cc/kg. A minimum hydrogen concentration will ensure the RCS is free of oxygen. Sulfur content is assumed to be at the maximum concentration in the NUREG.
The remaining NUREG/CR-6260 locations were reevaluated with a refined fatigue analysis using NB-3200 methods in a 3-D finite element analysis model using the design number of transients to reduce the CUF values. After reanalysis the RHR inlet transition was the only location to pass the UenEAF CUF criterion of 1.0 (Reference 5).
Two options are available to further reduce the UenEAF CUFs for the charging system nozzles, safety injection nozzles, and hot leg surge line nozzle: (1) calculate a strain rate dependent Fen; and (2) calculate CUF based on the 60 year projected numbers of transient events or both.
Revision of Fen Based on Strain-Rate The strain-rate dependent Fen values are calculated for the significant load set pairs in the fatigue analyses. Load set pairs that produce no significant stress range or fatigue contribution were assigned the maximum Fen for the material. The integrated strain rate method described in MRP-47, Guidelines for Addressing Fatigue Environmental Effects in a License Renewal Application, was used to calculate Fen values for individual load pairs that produce significant stress ranges. Dissolved oxygen of less than 0.05 ppm is assumed, which corresponds to a low oxygen environment. This is conservative since lower dissolved oxygen concentrations yield higher Fen values for stainless steel. Sulfur content is only applicable to low alloy steel locations.
Revision of CUF Based on 60-Year Projections of Transients However, multiplying the revised CUF by the weighted average Fen value computed above still results in UenEAF CUFs greater than 1.0 for the charging system nozzles, safety injection nozzles, and hot leg surge line nozzle after conservatism has been removed. In order to demonstrate that monitoring fatigue in these locations is a sufficient form of aging management, the UenEAF CUF was calculated based on the numbers of transients projected to 60 years in Table 4.3-2, Transient Accumulations and Projections. If the transient is not projected, then the full number of design basis events is used. There were threetwo transients that were not analyzed at the 60 year projection. The threetwo exceptions are the reactor trip from full power, with cooldown, with safety injection (type C reactor trip), inadvertent safety injection, and loss of power events, which were analyzed as the events to-date. The only UenEAF CUF calculation that could be affected by the use of the type C reactor trip, inadvertent safety injection, and loss of power transients to-date is associated with the safety injection nozzle, e.g. affected greater than the order of magnitude. This is addressed below.
The projected normal and alternate charging nozzles UenEAF CUFs are 0.57 and 0.53 based on SBF usage factors of 0.092 and 0.078, and Fen of 6.22 and 6.75 (Reference 6). The SBF usage
ULNRC-05938 December 13, 2012 Page 23 of 39 Callaway Plant Unit 1 Page 4.3-33 License Renewal Application Amendment 17 factors were generated with computer software that was benchmarked against NB-3200 methods consistent with RIS 2008-30 as discussed in Section 4.3.1.1, Fatigue Monitoring Methods.
The projected safety injection nozzle UenEAF CUF is 0.74 based on the usage factor of 0.11 and Fen of 6.5 (Reference 7). Even though this location is analyzed for the numbers of type C reactor trip, inadvertent safety injection, and loss of power events to-date, it is monitored with CBF; therefore UenEAF CUF will be updated as additional events occur.
The projected hot leg surge line nozzle UenEAF CUF is 0.765 based on the usage factor of 0.076 and Fen of 10.10 (Reference 8).
All of the locations specified in NUREG/CR-6260 for newer vintage Westinghouse plants listed in Table 4.3-6, Summary of Fatigue Usage Factors at NUREG/CR 6260 Sample Locations will be monitored by the Fatigue Monitoring program, described in Appendix B3.1. Most of the locations will be monitored using CBF or SBF. The hot leg surge nozzle will be monitored by incorporating the 60 year cycle projections into the cycle counting action limits to ensure that the results for the hot leg surge nozzle presented in this section are not exceeded. Therefore, the effects of the reactor coolant environment on fatigue usage factors will be managed for the period of extended operation. These TLAAs are dispositioned in accordance with 10 CFR 54.21(c)(1)(iii).
Disposition: Aging Management, 10 CFR 54.21(c)(1)(iii)
Evaluation of Limiting Sentinel Locations for Environmental Assisted Fatigue In order to assure that the limiting plant-specific EAF locations are identified as sentinel locations, Callaway performed a systematic review of all wetted, RCPB components with a Class 1 fatigue analysis [Ref. 17]. This was done either to show that the NUREG/CR-6260 locations are bounding or to incorporate EAF into the licensing basis for those more limiting components. The screening used EPRI Technical Report 1024995 Environmentally-Assisted Fatigue Screening, Process and Technical Basis for Identifying EAF Limiting Locations,
[Ref. 18].
The CUF for wetted, RCPB locations were categorized based on the strain-rate of the dominant transient. The strain-rate classification was determined with a qualitative assessment based on experience and not a quantitative stress analysis. The estimated strain rate was used to calculate an estimated Fen. The estimated Fen value was also calculated assuming a low DO environment; the maximum fluid/metal temperature; and the maximum sulfur concentration, and is based on the methods in NUREG/CR-5704 for austenitic stainless steels and Ni-Cr-Fe steels, and NUREG/CR-6583 for carbon and low alloy steels, and NUREG/CR-6909 for Ni-Cr-Fe steels. This estimated Fen was then averaged with the maximum Fen for that material type to
ULNRC-05938 December 13, 2012 Page 24 of 39 Callaway Plant Unit 1 Page 4.3-34 License Renewal Application Amendment 17 calculate the average Fen. The average Fen and the design basis CUFs were used to calculate the estimated UenEAF CUF.
These estimated UenEAF CUFs were then organized according to their system, thermal zone, and material type. A thermal zone is defined as a collection of piping and/or vessel components which undergo essentially the same group of thermal and pressure transients during plant operations. The maximum UenEAF CUF for each thermal zone and material was selected as a sentinel location. In addition, if the next highest UenEAF CUF with the same thermal zone and material is within 50% of the maximum, additional locations were identified as sentinel locations.
This initial list was reduced further using EPRI Technical Report 1024995 [Ref. 18].
x One Thermal Zone can bound another Thermal Zone in a System:
Both the CUF and Fen values for one sentinel location in one thermal zone are each higher than the CUF and Fen values for the sentinel locations in other thermal zones.
x One material in a Thermal Zone can bound other materials in the same Thermal Zone:
This circumstance could be achieved if within the same thermal zone, both the CUF and Fen values for one sentinel location composed of one material are each higher than the CUF and Fen values for the sentinel locations composed for all other materials.
x One material in a Thermal Zone can bound other materials in another Thermal Zone:
This circumstance combines the guidelines of the two listed above and must satisfy both criteria listed.
x A location with Uen < 0.8EAF CUF < 1.0 may be removed from the sentinel location list:
If the sentinel location UenEAF CUF for the projected number of design cycles is low (e.g.,
EAF CUF < 0.25e.g., Uen < 0.8), that sentinel location may be removed from the final list due to the small likelihood that it will be the leading sentinel location in a system. If, however, the sentinel location UenEAF CUF for the projected number of design cycles is fairly high (e.g., EAF CUF > 0.8e.g., Uen > 0.8), the possibility exists that it could remain the sentinel location for its group and should be included in the monitoring program that ensures that it does not exceed a value of 1.0.
x A location that can be shown to be bounded by another location on a common basis stress evaluation may be removed from the sentinel location list:
This judgment relies upon the comparison of transients in terms of severity and/or number of occurrences between locations in the same or different thermal zones. For example, it may be possible to demonstrate that:
o The bounding location resides in two different thermal zones and experiences more numerous and more severe transients than the bounded location that only experiences transients from one of those thermal zones. An example of this is the charging nozzle which experiences transients in the CVCS and RCS thermal zones and will bound the fatigue behavior of the other CVCS locations. This is appropriate because all locations are designed to the same CVCS transients
ULNRC-05938 December 13, 2012 Page 25 of 39 Callaway Plant Unit 1 Page 4.3-35 License Renewal Application Amendment 17 (e.g. loss of charging and loss of letdown). This is conservative because the charging nozzles CVCS transients have a reflood component where the nozzle temperature experiences a step increase from the CVCS transient temperature to the RCS Cold Leg. For example, during a Loss of Letdown event, the charging nozzle will cooldown from normal charging temperature (500°F) to the VCT temperature (70°F). This is followed by an abrupt increase in temperature from 70°F to RCS cold leg temperature (560°F) as the charging flow stops to maintain allowable pressurizer level and RCS cold leg flow refloods the nozzle. No other CVCS component experiences this reflood shock from the cold leg. Thus, the charging nozzles will bound the fatigue behavior of the other CVCS locations that flow to or from the RCS cold legs and can eliminate these other locations from the list of sentinel locations. The auxiliary spray piping also participates in two thermal zones (CVCS transients plus the main spray piping transients) and is designated as a sentinel location.
o For two locations in the same thermal zone which are designed to identical transients (both in terms of number and severity), the location with the higher value of Uen can be used to manage both locations. An example is the pressurizer heater well (SS)
(CUF = 0.562, Fen = 13.117, Uen = 7.372) and the pressurizer (lower) instrument nozzle (SS) (CUF = 0.439, Fen = 13.117, Uen = 5.758). The higher Uen of the former location identifies the bounding location. This is appropriate because both components are designed to the same design basis transients including insurge/outsurge. This is conservative because the components will experience the same transients in terms of number and severity over the life of the plant and the relative ranking of Uen will not change as fatigue accumulates in each component.
Table 4.3-7 identifies the final locations, including the NUREG/CR-6260 locations, that will be used as sentinel locations used during the period of extended operation to manage the EAF aging mechanism during the period of extended operation. The plant specific sentinel locations are incorporated into Table 4.3-6 in order to demonstrate that the Uen will remain below 1.0.
Those non-NUREG/CR-6260plant specific locations with an UenEAF CUF greater than 1.0 will be evaluated further using the same methods as those used to remove conservatisms for the NUREG/CR-6260 locations described above.
The results of these final analyses will be incorporated into the Fatigue Monitoring program either by either counting the transients assumed, or computing fatigue usage through the CBF/SBF capabilities of the program incorporate the stress intensities into a CBF ability of the program. As an alternative, the Fatigue Monitoring program may implement SBFs of certain locations in order to ensure the component does not exceed an UenEAF CUF of 1.0. Any use of SBF will be implemented in compliance with RIS 2008-30. Therefore, the effects of the reactor coolant environment on the non-NUREG/CR-6260plant specific locations will be managed for the period of extended operation. These TLAAs are dispositioned in accordance with 10 CFR 54.21(c)(1)(iii).
Disposition: Aging Management, 10 CFR 54.21(c)(1)(iii)
ULNRC-05938 December 13, 2012 Page 26 of 39 Callaway Plant Unit 1 Page 4.3-35 License Renewal Application Amendment 17 Table 4.3-6 Summary of Fatigue Usage Factors at NUREG/CR 6260 Sample Locations Licensing Basis Uen for the Sentinel Locations Location Material CUF Fen UenEAF CUF CUF-Fen Basis Further Eval.
Req.
NUREG/CR-6260 Locations RPV Bottom Head to Shell Junction SA 533, Grade B, Class 1, Low Alloy Steel 0.0070 2.45 0.01715 Design basis CUF NUREG/CR-6583 maximum Fen No RPV Inlet Nozzle SA 508, Class 2, Low Alloy Steel 0.0795 2.45 0.195 Design basis CUF NUREG/CR-6583 maximum Fen No RPV Outlet Nozzle SA 508, Class 2, Low Alloy Steel 0.1078 2.45 0.264 Design basis CUF NUREG/CR-6583 maximum Fen No Hot Leg Surge Line Nozzle SA 182, Type 316, Stainless Steel 0.07572 10.097 0.7646 CUF re-evaluated with NB-3200 elastic methods based on 60 year cycle projections, NUREG/CR-5704 strain-rate dependent Fen No Charging System Nozzle
[Normal and Alternate]
SA 182 Type 316, Stainless Steel 0.0919 /
0.0782 6.22 /
6.75 0.5715 /
0.5273 CUF re-evaluated with benchmarked SBF and 60 year cycle projections, NUREG/CR-5704 strain-rate dependent Fen No
ULNRC-05938 December 13, 2012 Page 27 of 39 Callaway Plant Unit 1 Page 4.3-36 License Renewal Application Amendment 17 Location Material CUF Fen UenEAF CUF CUF-Fen Basis Further Eval.
Req.
Safety Injection Nozzle [Boron Injection Header nozzles]
SA 182 Type 316, Stainless Steel 0.1135 6.495 0.7374 CUF re-evaluated with NB-3200 elastic methods based on 60 year cycle projections, NUREG/CR-5704 strain-rate dependent Fen No Residual Heat Removal Inlet Nozzle
[RHR nozzle-hot-leg]
SA 182 Type 316, Stainless Steel 0.0234 15.35 0.3591 CUF re-evaluated with NB-3200 elastic methods based on design cycles, NUREG/CR-5704 maximum Fen No Plant Specific Limiting Locations CETNA Upper Nozzle Housing Stainless Steel 0.37 13.117 4.853 Design basis CUF Estimated Fen based on NUREG/CR-5704 Yes RPV Bottom Head Instrument Tubes Ni-Cr-Fe 0.3184 13.117 4.176 Design basis CUF Estimated Fen based on NUREG/CR-5704 Yes PZR Heater Penetration (Heater Well)
Stainless Steel 0.0103 13.117 0.135 Design basis CUF based on NB-3200 elastic-plastic methods Estimated Fen based on NUREG/CR-5704 No Pressurizer Spray Nozzle Stainless Steel 0.411 9.013 3.704 Design basis CUF Estimated Fen based on NUREG/CR-5704 Yes Pressurizer Safety Valve Piping Stainless Steel 0.788 11.486 9.051 CUF re-evaluated with CBF and 60 year cycle projections, Estimated Fen based on NUREG/CR-5704 Yes
ULNRC-05938 December 13, 2012 Page 28 of 39 Callaway Plant Unit 1 Page 4.3-37 License Renewal Application Amendment 17 Location Material CUF Fen UenEAF CUF CUF-Fen Basis Further Eval.
Req.
Auxiliary Spray Piping Stainless Steel 0.72 10.350 7.452 Design basis CUF Estimated Fen based on NUREG/CR-5704 Yes RCP Casing/Discharge Nozzle Junction Stainless Steel 0.915 13.117 12.002 Design basis CUF Estimated Fen based on NUREG/CR-5704 Yes Accumulator Nozzles Stainless Steel 0.54 10.350 5.589 CUF re-evaluated with CBF and 60 year cycle projections, Estimated Fen based on NUREG/CR-5704 Yes Hot Leg SIS Nozzles, Loops 2 and 3 Stainless Steel 0.09 10.350 0.9315 CUF re-evaluated with CBF and 60 year cycle projections, Estimated Fen based on NUREG/CR-5704 No RSG Tubesheet (Continuous Region)
Low Alloy Steel 0.428 2.455 1.051 Design basis CUF NUREG/CR-6583 maximum Fen Yes RSG Tube-to-Tubesheet Connection Ni-Cr-Fe 0.068 13.117 0.892 Design basis CUF Estimated Fen based on NUREG/CR-5704 No
ULNRC-05938 December 13, 2012 Page 29 of 39 Callaway Plant Unit 1 Page 4.3-36 License Renewal Application Amendment 17 Table 4.3-7: Sentinel Locations for EAF Monitoring System Thermal Zone Material Component NUREG
/CR-6260 Design CUF Avg.
Fen Est. Uen EAF CUF Reactor Pressure Vessel RPV Nozzles LAS
- 1. RPV Outlet Nozzle Y
0.1078 2.455 0.265
- 2. RPV Inlet Nozzle Y
0.0795 2.455 0.195 RPV Upper Head SS
- 3. CETNA Upper Nozzle Housing N
0.37 13.117 4.853 RPV Bottom Head LAS
- 4. Bottom Head-to-Shell Junction Y
0.007 2.455 0.017 Ni-Cr-Fe
- 5. Bottom Head Instrument Tubes N
0.3184 13.117 4.093 4.176 1.303 Pressurizer PZR Lower Head SS
- 6. Pressurizer Heater Penetration N
0.562 13.117 7.372 LAS
- 7. Pressurizer Shell at Support Lug N
0.992 2.455 2.435
- 8. Pressurizer Surge Nozzle N
0.963 2.455 2.364
- 9. Pressurizer Lower Head/Support Skirt N
0.734 2.455 1.802 PZR Spray SS
- 7. Pressurizer Spray Nozzle N
0.411 9.013 3.704 PZR SRV/PORV SS
- 8. Safety and Relief Valve Piping N
0.975 11.486 11.199
- 12. Power Operated Relief Valve Solenoid N
0.68 11.486 7.811
ULNRC-05938 December 13, 2012 Page 30 of 39 Callaway Plant Unit 1 Page 4.3-37 License Renewal Application Amendment 17 System Thermal Zone Material Component NUREG
/CR-6260 Design CUF Avg.
Fen Est. Uen EAF CUF Surge Piping Surge Line SS
- 9. Hot Leg Surge Nozzle Y
0.3 11.486 3.446 CVCS Charging SS
- 10. Normal Charging Nozzles, Loop 1 Y
0.90 0.95 7.240 10.350 6.516 9.833
- 11. Alternate Charging Nozzles, Loop 4 Y
0.90 0.95 7.240 10.350 6.516 9.833 Auxiliary Spray SS
- 12. Auxiliary Spray Piping N
0.72 5.970 10.350 4.298 7.452 RCS RCS Cold Leg SS
- 13. RCP Casing/Discharge Nozzle Junction N
0.915 9.628 13.117 8.810 12.002 RHR RHR Inlet (Suction)
- 14. RHR Nozzles, Hot Leg Loops 1 & 4 Y
- 15. BIT Nozzles (All Loops)
Y 0.999 7.811 10.350 7.803 10.340 Accumulator SS
- 16. Accumulator Nozzles (All Loops)
N 0.95 7.811 10.350 7.420 9.833 SIS SS
- 17. Hot Leg SIS Nozzles, Loops 2 & 3 N
0.1 10.350 1.035
ULNRC-05938 December 13, 2012 Page 31 of 39 Callaway Plant Unit 1 Page 4.3-38 License Renewal Application Amendment 17 System Thermal Zone Material Component NUREG
/CR-6260 Design CUF Avg.
Fen Est. Uen EAF CUF Steam Generator Tubesheet LAS
- 18. RSG Tubesheet (Continuous Region)
N 0.428 2.455 1.051 Ni-Cr-Fe
- 19. RSG Tube-to-Tubesheet Connection N
0.068 13.117 0.892
ULNRC-05938 December 13, 2012 Page 32 of 39 Callaway Plant Unit 1 Page 4.3-42 License Renewal Application Amendment 17 (Section 4.3.8)
Letdown Heat Exchanger The fatigue analysis for the letdown heat exchanger indicated a maximum CUF of 1.84 for the flange. This CUF is the result of a recent reanalysis to account for operation with a letdown flow of 140 gpm. The CVCS design specification identifies the nominal letdown flow of 75 gpm with maximum flow of 120 gpm. Callaway operated from 1993 to 2011 at the maximum letdown flow, but has returned to the nominal value of 75 gpm. The analysis with an increased letdown flow was to account for this period. The CUF is driven mainly by transient 4, Charging flow step decreased and return to normal. This is a load following transient. Callaway does not practice load following operation and the number expected to be experienced is a small fraction of the number of assumed transients. The assumed number of this transient was dropped by an order of magnitude, which is about equal to 3 transients a month for 60 years and is more consistent with Callaways operation, and the CUF dropped to 0.894.
The fatigue analyses of the other components in the Callaway letdown heat exchanger included the tubesheet, tube side nozzles, and the studs. These components have CUFs of 0.910, 0.843, and 0.635. The fatigue analyses included transients 1, 2, 4, 5, 6, 7, 8, 9, and 10.
Transients 1, 2, 8, 9, and 10 will be monitored by the Fatigue Monitoring program (B3.1). The remaining transients, Transients 4, 5, 6, and 7, are not monitored. Transients 4, 5, and 7 assume 24,000 events and are load following events. They are based on 2 load changes per day (1 loading and 1 unloading) for 40 years with an 80 percent capacity factor. Callaway does not load follow and this high limiting number of events will not be approached during a 60-year plant life. Transient 6, letdown flow step decrease and return to normal, is not a normal operating event with the plant at power; however, this transient was included for conservatism.
It was assumed to occur approximately once a week for 40 years. If this assumption is extended through the period of extended operation, then 3,000 events will be assumed to occur and the CUF will increase to 0.9950.992, 0.880, and 0.696.
Therefore these CUFs are projected through the period of extended operation and the TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(ii).
Disposition: Projection, 10 CFR 54.21(c)(1)(ii)
ULNRC-05938 December 13, 2012 Page 33 of 39 Callaway Plant Unit 1 Page 4.8-1 License Renewal Application Amendment 17
4.8 REFERENCES
- 1. Westinghouse Report WCAP-15400-NP. Analysis of Capsule X from the Ameren UE Callaway Unit 1 Reactor Vessel Surveillance Program. Rev. 0. June 2000. Westinghouse Non-Proprietary Class 3.
- 2. Callaway PTLR. Callaway Plant Pressure and Temperature Limits Report. Rev. 5.
Released 11. December 2006.
- 3. Westinghouse Report. WCAP-17168-NP. Callaway Unit 1 Time-Limited Aging Analysis on Reactor Vessel Integrity. Rev. 0. September 2010. Westinghouse Non-Proprietary Class
- 3.
Rev. 0. Structural Integrity Associates, Inc. San Jose, California. 19 August 2010.
- 5. SIA Calculation 0901271.315. Residual Heat Removal (RHR) Inlet Nozzle Environmentally-Assisted Fatigue Analysis Calculation. Rev. 0. Structural Integrity Associates, Inc. San Jose, California. 11 August 2010.
Analysis Using 60-Year of Operation Using Stress Based-Fatigue (SBF) Results from the Baseline Evaluation. Rev. 0. Structural Integrity Associates, Inc. San Jose, California.
27 October 2011.
- 7. SIA Calculation 0901271.331. Safety Injection (BIT) Nozzle Environmentally-Assisted Fatigue (EAF) Analysis Using 60-Year Projected Numbers of Cycles. Rev 0. Structural Integrity Associates, Inc. San Jose, California. 16 September 2011..
- 8. SIA Calculation 0901271.330. Hot Leg Surge Nozzle Environmentally-Assisted Fatigue (EAF) Analysis Using 60-Year Projected Numbers of Cycles. Rev. 0. Structural Integrity Associates, Inc. San Jose, California. 15 September 2011.
- 9. Precision Surveillance Corporation Document No. CA-N1042-500. Final Report of the 25th Year IWL Inspection. Rev. 0. 16 September 2010. Supplemented by Callaway CAR 201009644.
- 10. Ameren Missouri Letter ULNRC-5100. Docket Number 50-483, Union Electric Company Callaway Plant, Transmittal of Inservice Inspection Summary Report for Refuel 13, and WCAP-16280-P, Flaw Evaluation Handbook For Callaway Unit 1 Reactor Vessel Inlet Nozzle Safe-End Weld Region, May 2004. 13 December 2004. (ADAMS Accession No ML043650441).
ULNRC-05938 December 13, 2012 Page 34 of 39 Callaway Plant Unit 1 Page 4.8-2 License Renewal Application Amendment 17
- 11. Ameren Missouri Calculation BB-183. Evaluation of Reactor Vessel Cladding Indication Inside Bottom Head During Refuel 13. Rev. 1.
- 12. Westinghouse Topical Report WCAP-15666-A. Extension of Reactor Coolant Pump Motor Flywheel Examination. Rev. 1. October 2003.
- 13. Ameren Missouri Letter ULNRC-05553. Graessle, Luke H. Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility Operating License NPF-30 Follow-Up Information Regarding 10 CFR 50.55a Request: Proposed Alternative to ASME Section XI Requirements for Replacement of Class 3 Buried Piping (TAC No. MD6792). Fulton, MO.
9 October 2008. (ADAMS Accession No ML082900027).
- 14. Ameren Missouri Letter ULNRC-05542. Graessle, Luke H. Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility Operating License NPF-30 Additional Information Regarding 10 CFR 50.55a Request: Proposed Alternative to ASME Section XI Requirements for Replacement of Class 3 Buried Piping (TAC No MD6792). Fulton, MO.
15 September 2008. (ADAMS Accession No ML082630798).
- 15. SIA Report FP-CALL-310. Benchmarking of Charging Nozzle Stress-Based Fatigue. Rev.
- 0. San Jose, California: Structural Integrity Associates. 22 June 2011.
- 16. SIA Report FP-CALL-304. Baseline Analysis of Callaway Plant Cycles and Fatigue Usage -
Startup through 1/31/2011. Rev. 21. San Jose, California: Structural Integrity Associates.
12 October 201213 October 2011
- 17. SIA Report FP-CALL-307. Environmentally-Assisted Fatigue Screening. Rev. 24. San Jose, California: Structural Integrity Associates. 30 April 201228 November 2012.
- 18. EPRI Technical Report 1024995. Environmentally-Assisted Fatigue Screening, Process and Technical Basis for Identifying EAF Limiting Locations.
- 19. WOG Topical Report WCAP-15338-A. A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants. Westinghouse Electric Company LLC.
October 2002.
ULNRC-05938 December 13, 2012 Page 35 of 39 Callaway Plant Unit 1 Page A-27 License Renewal Application Amendment 17 Appendix A Final Safety Analysis Report Supplement A3.2.3 Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)
All of the locations specified in NUREG/CR-6260 for newer vintage Westinghouse plants will be monitored by the Fatigue Monitoring program, described in Section A2.1. If any of the analyzed CUF values for these locations exceeds the fatigue design limit, the analyses may be revised using actual plant transients experienced. Callaway has completed an evaluation to identify any additional plant-specific bounding EAF locations. The supporting environmental factors, Fen, calculations will be performed with NUREG/CR-6909 or NUREG/CR-6583 for carbon and low alloy steels, NUREG/CR-6909 or NUREG/CR-5704 for austenitic stainless steels, and NUREG/CR-6909 for nickel alloys. The effects of the reactor coolant environment on fatigue usage factors in the NUREG/CR-6260 and plant-specific bounding EAF locations will be managed for the period of extended operation. The supporting environmental factors, Fen, calculations will be performed with NUREG/CR-6909 or NUREG/CR-6583 for carbon and low alloy steels, NUREG/CR-6909 or NUREG/CR-5704 for austenitic stainless steels, and NUREG/CR-6909 for nickel alloys. These TLAAs are dispositioned in accordance with 10 CFR 54.21(c)(1)(iii).
ULNRC-05938 December 13, 2012 Page 36 of 39 Callaway Plant Unit 1 Page B-127 License Renewal Application Amendment 17 Appendix B AGING MANAGEMENT PROGRAMS B3.1 FATIGUE MONITORING Program Description The Fatigue Monitoring program manages fatigue cracking caused by anticipated cyclic strains in metal components of the reactor coolant pressure boundary. The program ensures that actual plant experience remains bounded by the thermal and pressure transient numbers and severities analyzed in the design calculations, or that corrective actions maintain the design and licensing basis.
The Fatigue Monitoring program tracks fatigue by one of the following methods:
- 1) The Cycle Counting (CC) monitoring method tracks transient event cycles affecting the location to ensure that the numbers of transient events analyzed by the fatigue analyses are not exceeded. This method does not calculate cumulative usage factors (CUFs).
- 2) The Cycle-Based Fatigue (CBF) monitoring method utilizes the CC results and stress intensity ranges generated with the ASME III methods that use three dimensional six component stress-tensor methods to perform CUF calculations for a given location. The fatigue accumulation is tracked to determine approach to the ASME allowable fatigue limit of 1.0.
- 3) The Stress-Based Fatigue (SBF) monitoring method computes a "real time" stress history for a given component from data collected from plant instruments to calculate transient pressure and temperature, and the corresponding stress history at the critical location in the component. The stress history is analyzed to identify stress cycles, and then a CUF is computed. The CUF will be calculated using a three dimensional, six component stress tensor method meeting ASME III NB-3200 requirements, or a method will be benchmarked consistent with the NRC Regulatory Issue Summary RIS 2008-30.
The Fatigue Monitoring program requires periodic reviews of the plant instrumentation and operator logs to ensure that the fatigue critical thermal and pressure transients have not exceeded design transient severity or analyzed number, and to ensure that usage factors will not exceed the allowable value of 1.0 without corrective actions.
The Fatigue Monitoring program will be enhanced to include the effects of the reactor coolant environment on component fatigue life for a set of sample reactor coolant system locations. The set includes fatigue monitoring of the NUREG/CR-6260 sample locations for a newer-vintage Westinghouse Plant and plant-specific bounding environmentally assisted fatigue (EAF) locations. The supporting environmental factors, F(en),
calculations will be performed with NUREG/CR-6909 or NUREG/CR-6583 for carbon
ULNRC-05938 December 13, 2012 Page 37 of 39 Callaway Plant Unit 1 Page B-128 License Renewal Application Amendment 17 and low alloy steels, NUREG/CR-6909 or NUREG/CR-5704 for austenitic stainless steels, and NUREG/CR-6909 for nickel alloys.
NUREG-1801 Consistency The Fatigue Monitoring program is an existing program that, following enhancement, will be consistent with NUREG-1801,Section X.M1, Fatigue Monitoring.
Exceptions to NUREG-1801 None Enhancements Prior to the period of extended operation, the following enhancements will be implemented in the following program elements:
Scope of the Program - Element 1 Procedures will be enhanced to include fatigue usage calculations that consider the effects of the reactor water environment for a set of sample reactor coolant system locations. The set includes the NUREG/CR-6260 sample locations for a newer-vintage Westinghouse Plant, and plant-specific bounding EAF locations.
Procedures will be enhanced to ensure the fatigue crack growth analyses, which support the leak-before-break analyses, ASME Section XI evaluations, and the HELB break selection criterion remain valid by counting the transients used in the analyses.
Preventive Actions - Element 2 Procedures will be enhanced to require the review of the temperature and pressure transient data from the operator logs and plant instrumentation to ensure actual transient severity is bounded by the design and to include environmental effects where applicable.
If a transient occurs which exceeds the design transient definition the event is documented in the Corrective Action Program and corrective actions are taken.
Parameters Monitored or Inspected - Element 3 Procedures will be enhanced to include additional transients that contribute significantly to fatigue usage. These additional transients were identified by evaluation of ASME Section III fatigue and fatigue crack growth analyses.
Procedures will be enhanced to include additional locations which receive more detailed monitoring. These locations were identified by evaluation of ASME Section III fatigue analyses and the locations evaluated for effects of the reactor coolant environment. The monitoring methods will be benchmarked consistent with the NRC RIS 2008-30.
ULNRC-05938 December 13, 2012 Page 38 of 39 Callaway Plant Unit 1 Page B-130 License Renewal Application Amendment 17 Monitoring and Trending - Element 5 Procedures will be enhanced to project the transient count and fatigue accumulation of monitored components into the future.
Acceptance Criteria - Element 6 Procedures will be enhanced to include additional cycle count and fatigue usage action limits, which permit completion of corrective actions if the design limits are expected to be exceeded within the next three fuel cycles. The fatigue results associated with the NUREG/CR-6260 sample locations for a newer vintage Westinghouse plant and plant-specific bounding EAF locations will account for environmental effects on fatigue. The cycle count action limits for the hot leg surge nozzle will incorporate the 60 year cycle projections used in the hot leg surge nozzle EAF analysis.
Corrective Actions - Element 7 Procedures will be enhanced to include appropriate corrective actions to be invoked if a component approaches a cycle count or CUF action limit or if an experienced transient exceeds the design transient definition. If an action limit is reached, corrective actions include fatigue reanalysis, repair, or replacement. When a cycle counting action limit is reached, action will be taken to ensure that the analytical bases of the HELB locations are maintained. Re-analysis of a fatigue crack growth analysis must be consistent with or reconciled to the originally submitted analysis and receive the same level of regulatory review as the original analysis.
Operating Experience The following discussion of operating experience provides objective evidence that the Fatigue Monitoring program will be effective in ensuring that intended functions are maintained consistent with the current licensing basis for the period of extended operation.
- 1. In response to NRC Bulletin 88-11, Pressurizer Surge Line Thermal Stratification, Westinghouse performed a plant-specific evaluation of Callaway pressurizer surge line. It was concluded that thermal stratification does not affect the integrity of the pressurizer surge line. Callaway responses to NRC Bulletin 88-11 describe the inspections, analyses, and procedural revisions made to ensure that thermal stratification does not affect the integrity of the pressurizer surge line. There have been no signs of damage from surge line movement.
- 2. NRC Regulatory Issue Summary RIS 2008-30, Fatigue Analysis Of Nuclear Power Plant Components informed licensees of analysis methodology (Greens function)
ULNRC-05938 December 13, 2012 Page 39 of 39 Callaway Plant Unit 1 Page B-130 License Renewal Application Amendment 17 used to demonstrate compliance with the ASME Code fatigue acceptance criteria could be non-conservative if not correctly applied. Ameren Missouri is committed to using a three dimensional, six component stress tensor method meeting ASME III NB-3200 requirements, or benchmarking the chosen method. This benchmarking has been performed for the normal and alternate charging nozzle to in order to implement SBF at that location. Any additional locations which will be monitored with SBF must meet the ASME III NB-3200 requirements or be benchmarked.
3 An error was identified in the previous SBF transfer function for the normal and alternate charging nozzles. The SBF transfer function incorrectly included thermal sleeves for the nozzles and therefore would calculate less fatigue than the nozzles would actually accumulate. The extent of this condition is limited only to the normal and alternate charging nozzle SBF models. The transfer function has been updated to exclude thermal sleeves. The SBF transfer functions for the normal and alternate charging nozzles were also benchmarked in accordance with NRC RIS 2008-30.
- 4. The CVCS design specification identifies the nominal letdown flow of 75 gpm with maximum flow of 120 gpm. Callaway operated from 1993 to 2011 at the maximum letdown flow, but has returned to the nominal value of 75 gpm. The effects of this increased flow rate have been evaluated. To account for the increase in fatigue, Callaway reduced the assumed number of load following transients to be more consistent with its operation as a base load plant. Also, starting in Refuel Outage 17, Callaway has switched from the normal charging flow path to the alternate charging flow path in order to spread fatigue over the two paths.
The operating experience of the Fatigue Monitoring program did not identify an adverse trend in performance. Occurrences that would be identified under the Fatigue Monitoring program will be evaluated to ensure there is no significant impact to safe operation of the plant, and corrective actions will be taken to prevent recurrence.
Guidance for re-evaluation, repair, or replacement is provided for locations where aging is found. There is confidence that the continued implementation of the Fatigue Monitoring program will effectively identify aging prior to loss of intended function.
Conclusion The continued implementation of the Fatigue Monitoring program, following enhancement, provides reasonable assurance that aging effects will be managed such that the systems and components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
ULNRC-05938 December 13, 2012 Page 1 of 13 ENCLOSURE 3 Callaway Revision to RAI Set #9 Responses (Reference ULNRC-05915 dated October 11, 2012) x RAI 4.3-2 response is updated to reflect the revised baseline calculation.
x RAIs 4.3-20 & -21 responses are updated to reflect the revised EAF screening calculation.
ULNRC-05938 December 13, 2012 Page 2 of 13 RAI 4.3-2
Background:
LRA Section 4.3.1.2 states the baseline cycle counting results were projected to a 60-year operating life based on the actual accumulation history since the start of plant life. A rate of future cycle accumulation is computed for each transient and the cycle projections are based on a long term rate and a short term rate of cycle accumulation. In addition, these accumulation rates are then combined based on a weighting factor of one for the long term and three for the short term.
Issue:
Since the applicant used the 60-year transient projections to support the disposition of the time-limited aging analyses (TLAAs) evaluated in LRA Sections 4.7.2 and 4.7.7, the staff requires additional information to determine whether the long-term and short-term weighting factors and the associated transient occurrences for these weighting factors used in the projection methodology is appropriate and conservative.
Request:
a) Identify the transients in LRA Table 4.3-2 in which the long-term and short-term weighting factors, described in LRA Section 4.3.1.2, are applicable and provide the short-term and long-term occurrence for each transient.
b) If any design transient in LRA Table 4.3-2 used a different 60-year projection methodology, other than the one discussed in LRA Section 4.3.1.2, describe and justify that this alternative 60-year projection methodology is conservative.
Callaway Response a) All transients in LRA Table 4.3-2, except those described below in part b of this RAI response, use a weighting factor of 1 for the long term and 3 for the short term. The short-term period is the preceding 9 years (2002 through 2011) for all transients.
b) The below transients used a different 60 year projection methodology than that discussed in LRA Section 4.3.1.2. The transients are grouped according to the method used.
The following transients have a 60 year projection that is set equal to the baseline because the tests were performed during initial startup and no more tests are expected.
x Normal Transient #13, Turbine roll test x
Test Transient #1, Primary side hydrostatic test x
Test Transient #2, Secondary side hydrostatic test The following transients have a 60 year projection that is set equal to the baseline because this is the value analyzed in the EAF calculations. The justification for the use of the baseline value is provided in LRA Section 4.3.4.
x Upset Transient #2, Loss of Power (with natural circulation in the RCS) x Upset Transient #4c, Reactor trip from full power, with cooldown, with safety injection x
Upset Transient #8, Inadvertent safety injection actuation
ULNRC-05938 December 13, 2012 Page 3 of 13 For the following transients there have been no historical events recorded on which to base an accumulation rate. Therefore, the projected number of events was conservatively set equal to 1 event.
x Upset Transient #1, Loss of load (without immediate reactor trip) x Upset Transient #3, Partial loss of flow (loss of one pump) x Upset Transient #4b, Reactor trip from full power, with cooldown, without safety injection x
Upset Transient #4c, Reactor trip from full power, with cooldown, with safety injection x
Upset Transient #5a, Inadvertent RCS depressurization due to inadvertent auxiliary spray x
Upset Transient #6, Inadvertent startup of an inactive RCS loop x
Upset Transient #7, Control rod drop x
Upset Transient #9, Operating Basis Earthquake x
Upset Transient #10, Excessive Feedwater Flow x
Upset Transient #11, RCS Cold Overpressurization x
Test Transient #3. Tube leakage test x
Auxiliary Transient #1, Normal charging and letdown shutoff and return to service (Alt) x Auxiliary Transient #4, Charging flow shutoff with prompt return to service (Alt) x Auxiliary Transient #11, Accumulator actuation, accident operation x
Auxiliary Transient #12, Inadvertent accumulator blowdown x
Auxiliary Transient #14, High head safety injection (Loop B / C / D) x Auxiliary Transient #23, Low Pressure Safety Injection The following transients use a unique method to develop the 60 year projection.
x Normal Transient #7a/b, Loop out of service, Normal loop shutdown/startup: The 60 year projection is 0 because Callaway is not licensed for N-1 loop operation.
x Normal Transient #11, Reduced temperature return to power: Maneuver is not utilized because Callaway does not load follow.
x Auxiliary Transient #28, Excess letdown heat exchanger operation: Based on a linear projection of the available data. The projection is explained in LRA Section 4.3.8.
Corresponding Amendment Changes No changes to the License Renewal Application (LRA) are needed as a result of this response.
Refer to the Enclosure 2 Summary Table "Amendment 17, LRA Changes for a description of LRA changes with this response.
ULNRC-05938 December 13, 2012 Page 4 of 13 RAI 4.3-20
Background:
LRA Section 4.3.4, as amended by letter dated May 3, 2012, states that the CUF for wetted reactor coolant pressure boundary (RCPB) locations were categorized based on the strain rate of the dominant transient, which was determined with a qualitative assessment based on experience and not a quantitative stress analysis. In addition, this estimated strain rate was used to calculate an estimated environmental fatigue effect multiplier (Fen.)
It further states that this estimated Fen was then averaged with the maximum Fen for that material type to calculate the average Fen, which was used with the design basis CUFs to calculate the estimated EAF CUF. The estimated Fen value was based on NUREG/CR-5704 for austenitic stainless steels, NUREG/CR-6583 for carbon and low alloy steels and NUREG/CR-6909 for Ni-Cr-Fe steels. These estimated EAF CUFs were then organized according to their system, thermal zone and material type.
LRA Section 4.3.4, as amended by letter dated May 3, 2012, defines a thermal zone as a collection of piping and/or vessel components which undergo essentially the same group of thermal and pressure transients during plant operations.
Issue:
Since the estimated strain rate was determined based on a qualitative assessment, judgment of the appropriate strain rate must be made based on knowledge of, at a minimum, the transient, system and/or thermal zone in question. However, the applicant did not provide the details of how this qualitative assessment was performed for its plant nor was it justified that this approach was appropriate or conservative for the Callaway plant.
Since the estimated EAF CUFs were organized according to their system, thermal zone and material type, the staff noted that to have a meaningful comparison of the EAF CUFs, it is important that the CUFs were assessed similarly (e.g., amount of rigor in calculating CUF) and used the same fatigue curves in ASME Code,Section III, Appendix I.
In addition, since the LRA states that NUREG/CR-6909 was used for Ni-Cr-Fe steels, it is not clear whether the fatigue curve in Appendix A, Figure A.3, of NUREG/CR-6909 was used when using the Fen expression for Ni-Cr-Fe steels.
Request:
a) Describe the qualitative assessment that was used to categorize the CUF for wetted RCPB locations based on the strain rate of the dominant transient and provide the basis that this assessment is appropriate or conservative for the Callaway plant. As part of this description and justification, specifically include the criteria of the qualitative assessment to determine the appropriate strain rate to use in the categorization of EAF CUF.
b) Provide the reason that the method for calculating the average Fen (i.e., average of the estimated Fen and the maximum Fen for the material type) is appropriate and conservative for the plant-specific conditions.
c) Confirm that the EAF CUFs that were organized according to their system, thermal zone and material type were assessed similarly (e.g., amount of rigor in calculating CUF) and used the same fatigue curves in ASME Code Section III Appendix I to provide a meaningful comparison. If not, provide the basis for ranking or comparing the EAF CUFs to one another
ULNRC-05938 December 13, 2012 Page 5 of 13 to provide an appropriate method for screening and determining a "sentinel" location.
d) Since NUREG/CR-6909 was used for Ni-Cr-Fe steels, confirm that the fatigue curve in Appendix A, Figure A.3, of NUREG/CR-6909 was used for determining the Fen of Ni-Cr-Fe steels and EAF CUFs for Ni-Cr-Fe components. If not, provide the basis for not considering Figure A.3 of NUREG/CR-6909 for screening and determining a "sentinel" location.
Callaway Response a) A qualitative estimate of the strain rate for the controlling fatigue transient(s) was determined, based on experience with the corresponding Callaway plant system. These categorizations of strain rate were guided by transient descriptions described in system design documents. Each component was identified with one of eight possible f categories as shown in Table 9 (refer to table following this response).
Based on experience with fatigue analyses of many plant components, a strain-rate category was selected to represent each component. This selection was based on the identified transients that would govern the CUF at the given component, and ranking them with respect to how quickly the maximum and minimum stress states are established. For instance, if the majority of CUF for this component would generally be derived from very large temperature step changes, the High strain rate category was used. Conversely, components governed by hours-long ramps of temperature and pressure would be assigned to the V. Slow category.
b) The Average Fen (average of the estimated Fen and maximum Fen) is appropriate and conservative on the following basis:
This two-part average Fen is based on experience with performing detailed Fen analyses; in general, the effective Fen from a detailed analysis is similar to the Fen value computed for just the controlling transient pairs, but only slightly offset due to contributions from the less-significant fatigue pairs. Choosing an average Fen halfway between the estimated Fen and the maximum Fen provides reasonable assurance that the average Fen is higher than the effective Fen. This procedure is appropriate for the purpose of screening components for potential further evaluation. In addition, the resulting estimated Uen values are compared to a conservative threshold value of 0.8 for inclusion in the ranking of sentinel locations.
c) Based on the ASME Code years of analysis, it is assumed that the same fatigue curves for each material were used for the analyses relied upon for the screening process.
The level of analytical rigor has not been specifically reviewed. However, the CUF values used in the screening were developed from the same generation of analyses and are expected to have been performed using the same level of rigor (i.e. elastically-determined stresses with the same transient list and severity using the same ASME,Section III, Appendix I fatigue curve). No elastic-plastic evaluations, which would skew the ranking results, were found for these components. (An example of the difference between elastic and elastic-plastic analytical techniques is that for the same location in the pressurizer inner heater well, the elastic CUF = 0.562 while the elastic-plastic CUF = 0.0103).
ULNRC-05938 December 13, 2012 Page 6 of 13 Based on analytical experience and engineering judgment, the relative design report CUF values of the components indicate that any transient lumping used in the various analyses have not skewed the screening and ranking results.
Thus, a consistent technical basis appropriate for providing a meaningful comparison has been used to perform the EAF screening and identify appropriate sentinel locations.
d) The EAF screening was revised not to use the equations in NUREG/CR-6909 for the Fen for Ni-Cr-Fe. Instead the revised calculation uses NUREG/CR-5704 to compute Fen values for Ni-Cr-Fe material while using ASME Code determined CUF values.
Previously a correction factor was applied to the Fen computed in NUREG/CR-6909 when applied to a CUF computed from the ASME fatigue curve instead of the fatigue curve resident in NURE/CR-6909. However this factor could vary from less than 1 for low-cycle fatigue contributors to nearly 3 for high-cycle contributors. So NUREG/CR-5704 was used to compute Fen values for Ni-Cr-Fe material. The NRC staff has presented guidance which has found this approach acceptable.
This approach is only for the EAF screening. Further refinement of these factors can be achieved in future analysis using the methods in NUREG/CR-6909 for Ni-Cr-Fe material when the fatigue tables are available for use with individual components.
The design report CUF values (ASME Section III fatigue curve) and the Ni-Cr-Fe equation from NUREG/CR-6909 for computing Fen were used to compute the estimated Uen used for comparisons. Thus, the fatigue curve in Appendix A, Figure A.3, of NUREG/CR-6909 was not used for these comparisons and rankings. This is acceptable because the three components identified as Ni-Cr-Fe material that were evaluated had very low CUF values using the ASME Section III Appendix I fatigue curve (maximum CUF = 0.068). If adjusted for the fatigue curve in Appendix A, Figure A.3, of NUREG/CR-6909, they would not have been high enough to change the ranking results.
Corresponding Amendment Changes No changes to the License Renewal Application (LRA) are needed as a result of this response.
Refer to the Enclosure 2 Summary Table "Amendment 17, LRA Changes for a description of LRA changes with this response.
ULNRC-05938 December 13, 2012 Page 7 of 13 Table 9: Strain Rate Categories Strain Rate Category est. g
[%/sec]
Extreme 5.0 V. High
~ 1.3 High
~ 0.33 Mid-High
~ 0.087 Medium
~ 0.023 Low-Mid
~ 0.0059 Slow
~ 0.0015 V. Slow 0.0004
ULNRC-05938 December 13, 2012 Page 8 of 13 RAI 4.3-21
Background:
LRA Section 4.3.4, as amended by letter dated May 3, 2012, indicates that an initial screening list may have been reduced by using one of the following methods:
x One Thermal Zone can bound another Thermal Zone in a System x
One material in a Thermal Zone can bound other materials in the same Thermal Zone x
One material in a Thermal Zone can bound other materials in another Thermal Zone LRA Section 4.3.4, as amended by letter dated May 3, 2012, defines a thermal zone as a collection of piping and/or vessel components which undergo essentially the same group of thermal and pressure transients during plant operations.
Issue:
Since the initial screening list may have been reduced by using one of the methods described above, which is based on the CUF, Fen, thermal zone and material type, the staff noted that in order to have a meaningful comparison to screen EAF CUFs it is important that the CUFs were assessed similarly (i.e., amount of rigor in calculating CUF) and used the same fatigue curves in ASME Code,Section III, Appendix I. In addition, the applicant did not provide specific examples of how/when these methods were used to reduce the initial screening list; therefore, it is not clear how each method was applied and a basis was not provided to support that these methods are appropriate and conservative.
In addition, since the LRA states that NUREG/CR-6909 was used for Ni-Cr-Fe steels, it is not clear whether the fatigue curve in Appendix A, Figure A.3, of NUREG/CR-6909 was used when using the Fen expression for Ni-Cr-Fe steels.
Specifically for the "One Thermal Zone can bound another Thermal Zone in a System" method, it is not clear that a higher CUF and Fen in one thermal zone bounds a lower CUF and Fen from a different thermal zone. Factors such as the following should be considered, at a minimum:
(1) the CUF values having been assessed similarly in both thermal zones (i.e., amount of rigor in calculating CUF), (2) use of the same fatigue curves in ASME Code,Section III, Appendix I when comparing CUF, and (3) the thermal zone being considered "bounding" should experience thermal and pressure transients that are more severe compared to the other thermal zone.
Specifically for the "One material in a Thermal Zone can bound other materials in the same Thermal Zone" method, it is not clear that a higher CUF and Fen for one material in one thermal zone will always bound a lower CUF and Fen of all other materials in the same thermal zone.
Factors such as the following should be considered, at a minimum: (1) the CUF values having been assessed similarly for all materials (i.e., amount of rigor in calculating CUF) and (2) the material properties of the components in question, which can affect CUF and Fen.
Finally, the "One material in a Thermal Zone can bound other materials in another Thermal Zone" method is a combination of the two methods described above. Thus, the staff's concern
ULNRC-05938 December 13, 2012 Page 9 of 13 about those methods is also applicable.
Request:
For the "One Thermal Zone can bound another Thermal Zone in a System" method:
a) Provide two examples of when this method was used to reduce the initial screening list, including a reason that this method was appropriate and conservative for each situation.
b) Provide the reason that a higher CUF and Fen in one thermal zone will bound a lower CUF and Fen from a different thermal zone. As part of the justification, specifically address any factors or criteria that are applicable when implementing this method.
c) If the following factors and criteria were not included, as part of the response above, provide the reason that they are not appropriate and do not need to be considered: (1) CUF values being assessed similarly in both thermal zones (i.e., amount of rigor in calculating CUF), (2) use of the same fatigue curves in ASME Code,Section III, Appendix I when comparing CUF, and (3) the thermal zone being considered "bounding" should experience thermal and pressure transients that are more severe compared to the other thermal zone.
For the "One material in a Thermal Zone can bound other materials in the same Thermal Zone" method:
d) Provide two examples of when this method was used to reduce the initial screening list, including a reason that this method was appropriate and conservative for each situation.
e) Provide the reason that a higher CUF and Fen for one material in one thermal zone will bound a lower CUF and Fen of all other materials in the same thermal zone. As part of the justification, specifically address any factors or criteria that are applicable when implementing this method.
f) If the following factors and criteria were not included, as part of the response above, provide the reason that they are not appropriate and do not need to be considered: (1) CUF values being assessed similarly for all materials (i.e., amount of rigor in calculating CUF) and (2) the material properties of the components in question, which can affect CUF and Fen.
For the "One material in a Thermal Zone can bound other materials in another Thermal Zone" method:
g) Provide two examples of when this method was used to reduce the initial screening list, including a reason that this method was appropriate and conservative for each situation.
h) Provide the reason that one material in a thermal zone can bound other materials in another thermal zone. As part of the justification, specifically address any factors or criteria that are applicable when implementing this method.
i) If the following factors and criteria were not included, as part of the response above, provide the reason that they are not appropriate and do not need to be considered: (1) CUF values being assessed similarly in both thermal zones for the different materials (i.e., amount of rigor in calculating CUF), (2) the material properties of the components in question, which can affect CUF and Fen, and (3) the thermal zone being considering "bounding" should experience thermal and pressure transients that are more severe compared to the other
ULNRC-05938 December 13, 2012 Page 10 of 13 thermal zone.
Callaway Response In all of the responses that follow, the term Uen is used to report computed values of Environmentally-Assisted Fatigue (EAF). These Uen computations were performed as the product of a design CUF and the average Fen computed as the average of the estimated Fen and the maximum Fen for the material type.
a) Example #1, The Pressurizer Safety and Relief Valve piping thermal zone bounds the Pressurizer Upper Head thermal zone:
Pressurizer Instrument Nozzle (SS) (CUF=0.236, Fen=11.48613.117, Uen=2.7113.096) in the Pressurizer Upper Head thermal zone in the Pressurizer component is bounded by the 3 and 6 Pressurizer Safety and Relief Valve piping (SS) (CUF=0.975, Fen= 11.486, Uen=11.199) in the Pressurizer SRV/PORV thermal zone in the same component (Pressurizer). In this example, both components are made of the same material.
This example is appropriate because the two components are welded to the same pressurizer vessel and are part of the same pressure boundary experiencing most of the same transients in terms of thermal and pressure severity and number of occurrences.
The example is conservative because the computed CUF of the SRV piping is over 4 3 times greater than the pressurizer instrument nozzles and the average Fen values are equivalent based on the similarities in the loading rates of the controlling transients. The SRV piping will experience higher Uen throughout component life based on the similar transient sets.
Example #2, The Tubesheet thermal zone bounds the RSG Primary Head thermal zone:
Replacement Steam Generator (RSG) Primary Manway Drain Tube (LAS) (CUF=0.391, Fen=2.455, Uen=0.960) in the RSG Primary Head thermal zone in the RSG component is bounded by the RSG Tubesheet (Continuous region) (LAS) (CUF=0.428, Fen=2.455, Uen=1.051) in the Tubesheet thermal zone in the same component (RSG). In this example, both components are made of the same material.
This example is appropriate because the two components are part of the same RSG primary pressure boundary experiencing most of the same transients in terms of thermal and pressure severity and number of occurrences.
The example is conservative because the computed CUF of the RSG Tubesheet (Continuous region) is greater than the RSG Primary Manway Drain Tube and the average Fen values are equivalent based on the similarity of the loading rates of the controlling transients. The RSG Tubesheet (Continuous region) will experience higher Uen throughout component life based on the similar transient sets.
b) A similar level of analytical rigor and basis (e.g., elastically-determined stresses with the same transient list, number of occurrences and severity using the same ASME,Section III,
ULNRC-05938 December 13, 2012 Page 11 of 13 Appendix I fatigue curve) is assumed. The computed CUFs will represent an appropriate basis for relative CUF ranking.
c) Items (1) and (2) of c) were included in the response to item b). Item (3): this concept was not a part of the example. However, determination of boundedness using this concept requires consideration both that using a more severe set of transients (i.e., larger Ts with shorter duration) than those used in the bounded thermal zone will likely cause larger CUF values, but that the possibly countervailing factor that a shorter duration strain rate may produce smaller Fen values. The product of these factors is the key attribute to be used for the boundedness determination.
d) Example #1, The SS material bounds the LAS material in the Pressurizer Upper Head thermal zone:
Pressurizer Upper Head/Upper Shell at Support Lug (LAS) (CUF=0.9280.992, Fen=2.455, Uen=2.2782.435) in the Pressurizer Upper Head thermal zone in the Pressurizer component is bounded by the Pressurizer Instrument Nozzle (SS) (CUF=0.236, Fen=
11.48613.117, Uen=2.7113.096) in the same Pressurizer Upper Head thermal zone in the same component (Pressurizer). In this example, the two components are made of different materials. The instrument nozzle is in turn bounded by the PSV and PORV. See response (g).
This example is appropriate because the two components are welded to the same pressurizer vessel and are part of the same pressure boundary experiencing most of the same transients in terms of thermal and pressure severity and number of occurrences. The example is conservative for the following reasons:
x The average Fen for the stainless steel material (Pressurizer Instrument Nozzle) is over 4.65.3 times greater than that for Low Alloy Steel (Pressurizer Upper Head/Upper Shell at Support Lug).
x The CUF of the stainless steel component is a factor of 0.25 0.24 of the low alloy steel component (this factor is not sufficiently low to overwhelm the 4.65.3 times higher Fen factor).
x Although the Uen values are approximately equivalent on a design transient basis, the Pressurizer Upper Head/Upper Shell at Support Lug is dominated by an unlikely event (inadvertent auxiliary spray event), which, if eliminated from the computation, reduces the design CUF significantly (order of magnitude), while the Pressurizer Instrument Nozzle is affected by transients that occur regularly (pressurizer heatup/cooldown and insurge-outsurge). The basis for minimizing the effect of the inadvertent auxiliary spray event is that the event has a design value of 10 occurrences, but has not actually occurred in the life of the plant and only 1 occurrence is conservatively forecast to occur in the period of time to 60 years.
This is the only example where one material in a Thermal Zone bounds another material in the same Thermal Zone.
Example #2, The LAS material bounds the Ni-Cr-Fe material in the RSG Tubesheet thermal zone:
ULNRC-05938 December 13, 2012 Page 12 of 13 Replacement Steam Generator (RSG) Tube-to-Tubesheet Connection (Ni-Cr-Fe)
(CUF=0.068, Fen=4.093, Uen=0.278) in the Tubesheet thermal zone in the RSG component is bounded by the RSG Tubesheet (Continuous region) (LAS) (CUF=0.428, Fen=2.455, Uen=1.051) in the same Tubesheet thermal zone in the same component (RSG). In this example, the two components are made of different materials.
This example is appropriate because the two components are part of the same RSG primary pressure boundary experiencing most of the same transients in terms of thermal and pressure severity and number of occurrences.
The example is conservative because the computed CUF of the bounding RSG Tubesheet (Continuous region) is over 6 times greater than the RSG Tube-to-Tubesheet Connection while its average Fen value is only 1.67 times that of the bounded Ni-Cr-Fe component, resulting in a 3.78 times larger Uen value.
e) A higher CUF and Fen for one material in a thermal zone is used to bound a lower CUF and Fen of all other materials in the same thermal zone is if the Uen for the former sentinel location would be more than double the Uen values of the other sentinel locations. This degree of margin always is expected to accommodate any differences in transients or environment experienced.
A similar level of analytical rigor and basis (e.g., elastically-determined stresses with the same transient list, number of occurrences and severity using the same set of ASME,Section III, Appendix I fatigue curves) is assumed. The computed CUFs will represent an appropriate basis for relative CUF ranking. The Fen values for each material are computed using the similar loading rates, temperatures and DO and are appropriate for comparison purposes.
f) These two factors (1) and (2) are addressed in e).
g) Example #1, The LAS material in the Pressurizer Upper Head thermal zone is bounded by the SS material in the Pressurizer Safety and Relief Valve piping thermal zone:
The Pressurizer Upper Head/Upper Shell at Support Lug (LAS) (CUF=0.9280.992, Fen=2.455, Uen=2.2782.435) in the Pressurizer Upper Head thermal zone in the Pressurizer component is bounded by the 3 and 6 Pressurizer Safety and Relief Valve piping (SS)
(CUF=0.975, Fen= 11.486, Uen=11.199) in the Pressurizer SRV/PORV thermal zone in the same component (Pressurizer).
This example is appropriate because the two components are either welded to or are an integral part of the same pressurizer vessel, thus being part of the same pressure boundary and experiencing most of the same transients in terms of thermal and pressure severity and number of occurrences.
The example is conservative because the CUF and Fen of the bounding component are both greater than those of the bounded component. This is the only example where one material in a Thermal Zone bounds another material in another Thermal Zone.
ULNRC-05938 December 13, 2012 Page 13 of 13 h) For different thermal zones in the same component or system, a similar level of analytical rigor and basis (e.g., elastically-determined stresses with the same transient list, number of occurrences and severity using the same set of ASME,Section III, Appendix I fatigue curves) was assumed. Since this level of analytical rigor and basis was comparable, the computed CUFs will represent an appropriate basis for relative CUF ranking. The Fen values for each material are computed using the appropriate loading rates, temperatures and DO and are appropriate for comparison purposes. Although not a part of the Example
- 1, for thermal zones in different components in different systems, unless the thermal and pressure transients of the bounding component are demonstrably more severe than those of the bounded component, the comparisons of CUF and Fen may not prove to be appropriate.
i)
Items (1) and (2) of i) were included in the response to item h). Item (3): this concept was not a part of the example. However, determination of boundedness using this concept requires consideration both that using a more severe set of transients (i.e., larger Ts with shorter duration) than those used in the bounded thermal zone will likely cause larger CUF values, but that the possibly countervailing factor that a shorter duration strain rate may produce smaller Fen values. The product of these factors is the key attribute to be used for the boundedness determination.
Corresponding Amendment Changes No changes to the License Renewal Application (LRA) are needed as a result of this response.
Refer to the Enclosure 2 Summary Table "Amendment 17, LRA Changes for a description of LRA changes with this response.