ML20059A509
| ML20059A509 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 10/22/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9310260352 | |
| Download: ML20059A509 (5) | |
Text
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GENucI::t Energy Genwn!Ekstnc Compary l
175 Curtoer kenue. Sarr Jo:,e. CA 95125 1
October 22,1993 Docket No.52-001 Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submittal Supporting Accelerated ABWR Schedule - Justification of Number of Cycles / Events Specified in Table 3.9-1
Dear Chet:
Attached is the justification of the number of cycles / events specified in Table 3.91. Our detailed review determined that al1 of the Table 3.9-1 entries are justified with the exception of Events 6 and 14 (See attached markup of Table 3.9-1). We concluded that both of these i
events should be increased by a factor of 1.5.
During the review, we found that note 7 of Amendment 31 was inadvertently omitted from Amendment 32. Thus, a markup of this note has been provided on the attached Table 3.91.
Please provide copies of this transmittal to Jim Brammer and Dave Terao.
Sincerely,
- )0 i
a k Fox Advanced Reactor Programs cc:
Alan Beard (GE) l Norman Fletcher (DOE)
Maryann lierzog (GE) avm
()Y i
t 9310260352 931022-r n ^ ^ Ad, o
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BASIS FOR NUMBER OF CYCLES SPECIFIED IN TABLE 3.9-1 1.
BOLTUP:
Refuel every 18 months
40 cycles Extra cycles for fabrication & assembly
5 cycles Total = 45 cycles
- 2. HYDROSTATIC TEST:
Two test cycles for each boltup
= 90 cycles 3.
STARTUP: Based on BWR operating experience data, the expected number of cycles in 60 years is 205.
i For conservatism, 260 cycles is used in the design.
4.
DAILY & WEEKLY REDUCTION TO 50% POWER:
6 times / week X 50 weeks X 60 years = 18,000 cycles 5.
CONTROL ROD PATTERN CHANGE:
10 times / year X 60 years = 600 cycles 6.
LOSS OF FEEDWATER HEATERS:
2 times / year X 60 years = 120 cycles
- 7. SCRAM: Based on BWR operating experience, events 7a & 7b, typically occurred less than 75 times in 40 years.
Therefore, the specified number of design cycles for events 7a & 7b is conservative for 60 years.
Event 7a = 125 cycles Event 7b = 139 cycles Based on BWR operating experience, the number of cycles for event 7c were determined.
Event 7c =
8 cycles 8.
REDUCTION TO 0% POWER. HOT STANDBY. SHUTDOWN:
No. of cycles = Startup cycles - Event 7c cycles = 252 cycles
- 9. REFUELING SHUTDOWN w/ HEAD SPRAY AND UNBOLT:
No. of cycles = No. of Event 1 cycles = 45 cycles b
d BASIS FOR NUMBER OF CYCLES SPECIFIED IN TABLE 3.9-1 (Continued) 10.&11. SCRAM & IMPROPER OR SUDDEN START OF RECIRC PUMPS:
These are Level C events, with an event probability of less than once in 100 years of reactor operation.
Therefore, with a 60 year design life, these events are assumed to occur one time.
12.&l3. HEE:The design requirements for the number of cycles and events is specified in SSAR Section 3.7.3.2.
- 14. TURBINE STOP VALVE FULL CLOSURE (TSVC):
Turbine stop valve closure occurs during Event 7a turbine generator trips, and during Turbine Stop valve testing.
No. of cycles = 65 cycles Turb. Gen. trip + 4 Tests /yr.X 60 yr.
= 305 cycles, For design, use 330 cycles
- 15. SAFETY / RELIEF VALVE ACTUATION:
Occurs during Events 7a & 7b, therefore 264 events are specified. BWR 6's are designed for 1600 single valve actuation events. For 60 years, this number is extrapolated to 2400 single valve actuation events.
This would result in a total number of valve actuation events =2400 + 264 = 2664 events. For conservatism, this total was rounded up to 2800 events.
SRV (All or Automatic Depressurization System) events =
264 SRV (One or Two adjacent) events
= 2536 SRV Total number of events
= 2800 i
SRV Reactor Building Vibratory cycles on RPV E Internals:
2800 events x 7 cycles / event = 19600 cycles f
l
23A6100 Rev. 2 ABWR stand:rdsr.ury An:tysis nip:rt Table 3.9-1 Plant Events A. Plant Operating Events ASME Code Service No. of i
2 Limit Cycles / Events 1.
Boltup A
45 2
l 2.
Hydrostatic Test (two test cycles for each boltup cycle)
Testing 90 3.
Startup (56*C/hr Heatup Rate)3 A
260 2
4.
Daily and Weekly Reduction to 50% Power A
18,000 2
5.
Control Rod Pattern Change A
600 6.
Loss of Feedwater Heaters B
-es- ) 20 7.
Scram:
125
- b. Loss of Feedwater Flow, Loss of Auxiliary Power B
139
- c. Turbine Bypass, Single Safety or Relief Valve Blowdown D
8 8.
Reduction to 0% Power, Hot Standby, Shutdown (56*C/hr A
252 Cooldown Rate)3 2
9.
Refueling Shutdown with Head Spray and Unbolt A
45
- 10. Scram:
4 a.
Reactor Overpressure with Delayed Scram (Anticipated C
1 Transient Without Scram, ATWS) 4
- b. Automatic Blowdown C
1 d
- 11. Improper or Sudden Start of Recirculation Pump with Cold C
1 Bottcm Head or Hot Standby-Drain Shut Off-Pump Restart See next page for footnotes 3.9-86 Mechanical Systems and Components - Amendment 32
e-
~
23A6t00 Rev. 2 ABWR scad:rdS&ty An: lysis Rip:rf Table 3.9-1 Plant Events (Continued) 5 B. Dynamic Loading Events ASME Code Service No. of Limit 1 2
Cycles / Events
- 12. Safe Shutdown Earthquake (SSE) Event at Rated Power B6 2 Events 7 Operating Conditions 10 Cycles / event
\\8 1 Cycle d
- 13. Safe Shutdown Earthquake (SSE) at Rated Power Operating D
Conditions 3'LO
- 14. Turbine Stop Valve Full Closure (TSVC)\\ 9 B
-220 Events During Event 7a and Testing 3 Cycles /
Event
- 15. Safety / Relief Valve (SRV) Actuation (One, Two Adjacent, All or B
264 EventsW Automatic Depressurization System) During Event 7a and 7b
- 16. Loss-of-Coolant Accident (LOCA)
Small Break LOCA (SBL) or D8 14 4
i Intermediate Break LOCA (IBL) or De 3
8 4
Large Break LOCA (LBL)
D 1
1 1hese ASME Code Service Limits apply to ASME Code Class 1,2 and 3 components, component supports and Class CS structures. Different limits apply to Class MC and CC containment vessels and components, as discussed in Section 3.8.
2 Some events apply to reactor pressure vessel (RPV) only. The number of events / cycles applies to RPV as an example.
3 Bulk average vessel coolant temperature change in any one-hour period.
4 The annual encounter probability of a single event is <10-2 for a Level C event and <10 for a Level 4
D event (Subsection 3.9.3.1.1.5).
5 Table 3.9-2 shows the evaluation basis combination of these dynamic loadings.
l 6 The effects of displacement-limited, seismic anchor motions (SAM) due to SSE shall be evaluated for safety-related ASME Code Class 1,2, and 3 components and component supports to ensure their functionality during and following an SSE. The SAM effects shall include relative displacements of piping between building floors and slabs, at equipment nozzles, at piping penetrations and at connections of small diameter piping to large diameter piping. See Table 3.9-2 l
and Note 7 of Table 3.9-2 for stress limits to be used to evaluate the SAM effects.
8 // Appendix F or other appropriate requirements of the ASME Code are used to determine the service Level D limits, as described in Subsection 3.9.1.4.
9 4 Applicable to main steam piping system only.
104 The number of Reactor Building vibratory load cycles on the reactor vessel and internal components is 19,600 cycles of varying amplitude during the 264 events of SRV actuation.The number of Reactor Building vibratory load cycles on the piping systems inside the containment is 2536 events of single SRV actuation, with 3 stress cycles per event and 264 events of SRV actuation of all valves or the Automatic Depressurization System valves, with 3 stress cycles per event.
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Mechanical Systems and Components - Amendment 32 3 9-87
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