ML032020335
| ML032020335 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 07/18/2003 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| Download: ML032020335 (4) | |
Text
3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM Reactor Coolant System Heatup Limitations Without Margins for Instrumentation Error Applicable for 32 EFPY of Operation Limiting Material: Axial Weld 13253112008, Cu = 0.21%, Ni = 0.873%6, and Lower Shell Plate 5440743, Cu = 0.14%, N1 = 0.50%
Initial ART: 58 Deg. F, Limiting ART Values at 32 EFPY: 1/4T = 199 Deg. F, 314T = 143 Deg. F 2500-l lLeak Test Limit 2250 _ ___. _ _
Un0accepta le
__Acceptab 2000
_ Operation Operatlo Heatup Rate Criticality 1750 60 Deg. FIHr Limit At en2
= 1500 En E 1250 S!
a)
I 1000 0
0 0
15 750 43 a
500 250 0
0-50 loC 150 200 250 300 350 400 450 Average Reactor Coolant System Temperature (Deg. F) 500 550 FIGURE 3.4-2 REACTOR COOLANT SYSTEM PRESSURE - TEMPERATURE LIMITS VERSUS 60 F/HR RATE, CRITICALITY LIMIT, BOLTUP LIMIT, AND LEAK TEST LIMIT COOK NUCLEAR PLANT-UNIT 1 Page 314 4-27 AMENDMENT 99, 467, M, 278
3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM Reactor Coolant System Cooldown Limitations Without Margins for Instrumentation Error Applicable for 32 EFPY of Operation Umiting Material: Axial Weld 13253112008, Cu = 0.21%h, Ni = 0.873%, and Lower Shell Plate B4407-3, Cu = 0.14%, Ni = 0.50%
Initial ART: 58 Deg.-F, Limiting ART Values at 32 EFPY: 114T = 199 Deg. F, 314T = 143 Deg. F 2500 2250 2000 a 1750 raa-o 1500 2
E t
1250 Cu C
o 1000 U
h.0 s
750 It 500 250 0
0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperature (Deg. F)
FIGURE 3.4-3 REACTOR COOLANT SYSTEM PRESSURE -
TEMPERATURE LIMITS VERSUS COOLDOWN RATES COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-28 AMENDMENT 99, 467, 2-73, 278
314 BASES 3/4.3 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.1.4 of the FSAR.
During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
An ID or OD one-quarter thickness surface flaw is postulated at the location in the vessel which is found to be the limiting case. There are several factors which influence the postulated location. The thermal induced bending stress during heatup is compressive on the inner surface while tensile on the outer surface of the vessel wall.
During cooldown the bending stress profile is reversed. In addition, the material toughness is dependent upon irradiation and temperature and therefore the fluence profile through the reactor vessel wall, the rate of heatup and also the rate of cooldown influence the postulated flaw location.
The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60'F per hour. The cooldown limit curves of Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature for the controlling materials at the end of 32 EFPY.
Reactor operation and resultant fast neutron (E > 1 Mev) irradiation will cause an increase in the RTNDT.
Therefore, an adjusted reference temperature, based upon the fluence, and the copper and nickel content of the material must be predicted. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include the adjusted RTNwu at the end of 32 EFPY, as well as adjustments for possible errors in the pressure and temperature sensing instruments.
COOK NUCLEAR PLANT-UNIT 1 Pam B 3/4 4-6 AMENDMENT 99, 467, 273, 278
3/4 BASES 3/4.3 REACTOR COOLANT SYSTEM The 32 EFPY heatup and cooldown curves were developed based on the following:
- 1.
The limiting materials being axial weld 13253/12008, with a copper and nickel content of 0.21 % and 0.873%, respectively, and lower shell plate B4407-3, with a copper and nickel content of 0.14 % and 0.50% respectively.
- 2.
The fluence values contained in Table 6-14 of WCAP-12483, Revision 1, "Analysis of Capsule U from the American Electric Power Company D. C. Cook Unit 1 Reactor Vessel Radiation Surveillance Program," dated December 2002.
- 3.
Figure 1, NRC Regulatory Guide 1.99, Revision 2 The shift in RTNrT of the reactor vessel material has been established by removing and evaluating the material surveillance capsules installed near the inside wall of the reactor vessel in accordance with the removal schedule in Table 4.4-5. Per this schedule, Capsule U is the last capsule to be removed until Capsule S is to be removed after 32 EFPY (EOL). Capsule V, W, and Z will remain in the reactor vessel, and will be removed to address industry reactor embrittlement concerns, if required.
The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
The OPERABILITY of two PORVs, or of one PORV and the RHR safety valve, ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 1521F.
Either PORV or RHR safety valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 500F above the RCS cold leg temperatures or (2) the start of a charging pump and its injection into a water solid RCS. Therefore, any one of the three blocked open PORVs constituted an acceptable RCS vent to preclude APPLICABILITY of Specification 3.4.9.3.
COOK NUCLEAR PLANT-UNIT 1 Page B 3/44-7 AMENDMENT 88, 467,4-76, 27.3, 278 COOK NUCLEAR PLANT-UNIT 1 Page B 3/4 4-7 AMENDMENT 99, 467,476, W3, 278