ML17318A746
| ML17318A746 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/09/1980 |
| From: | INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | |
| Shared Package | |
| ML17318A747 | List: |
| References | |
| NUDOCS 8005130239 | |
| Download: ML17318A746 (4) | |
Text
a A
TABLE i.i-5 REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE 8
- SPECIMEN
- 1. Capsule T
- 2. Capsule X
- 3. Capsule Y
- 4. Capsule 0
- 5. Capsule S
- 6. Capsules V, M, REHOVAL INTERVAL 1.25 EFPY 3
EFPY 5
EFPY 9
~
~
a ~
~ A~
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~ o REACTOR COOLANT SYSTEM BASES The reactor vessel materials have been tested to determine their initial RTN
., the results of these tests are shown in Table 8 3/4.4-1.
Reactor opNNtion and resu'ltant fast neutron (E)1 Nev) irradiation will cause an increase in the RTN Therefore, an adjusted reference temperature, based upon the'iHuence and copper content of the material in question, can be predicted using Figures 8 3/4.4-1 and 8 3/4.4-2.
he heatup and cooldown limit curves, Figures 3.4-2 and 3.4-3, include predicted adjustments for this shift in RTR at the end of 12 RFPY, as well as adjustments for possible errors'he pressure and temperature sensing instruments.
The actual shift in RT~
T of the vessel material will be established periodica'lly during operatii by removing and evaIuating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.
The heatup and cooldown curves must be recalculated when the dRT, determiined from the surveillance capsule is different from the calnfated aRTRRT for the equivalent capsule radiation exposure.
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 require-ments of Appendix G to 10 CFR 50.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirenents 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.
3/4.4.10 STRUCTURAL INTEGRITY The required inspection programs for the Reactor Coolant System components ensure that the structural integrity of these components will be maintained at an acceptable level throug'n'out the life of the plant.
To the extent applicable, the inspection program for the Reactor Coolant System components is in compliance with Section XI of D. C.
COOK-UNIT 1
8 3/4 4-11
TAB S
C7 O
A OO I
CAPSULE 60 CAPSULES S
V,M~Z 1e CAPSULE T
2+
CAPSULE Y
3e CAPSULE X
CAPSULE U
1
-EFPY 3
EFPY 5
EFPY 9
REACTOR COOLANT SYSTEM BASES The actual shift in RTN T of the vessel material will be established periodically during operatiI by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance speci-mens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor ves'sel.
The heatup and cooldown curves must be recalculated when the aRT determined from the surveillance capsule is different from the calculaIh aRTRgT for the equivalent capsule radiation exposure.
e 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 frequen~ies for removing and testing 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, 3/4.4.10 STRUCTURAL INTEGRITY The inspection and testing programs for ASME Code Class 1,
2 and 3
components ensure that the structural integri ty of these components will be maintained at an acceptable level throughout the life of the plant.
To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.
0.
C.
COOK - UNIT 2 8 3/4 4-10