ML17256A417
| ML17256A417 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/08/1982 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| Shared Package | |
| ML17256A416 | List: |
| References | |
| NUDOCS 8212140413 | |
| Download: ML17256A417 (12) | |
Text
3.1.2 Heatu
'and Cooldown Limit Curves for Hormal 0 eration 3 '.2.1 The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.1-1 and 3.1-2 for the first, 21.0 effective full 3 '.2.2 3 '.2 '
Basis:
power years.
a.
Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.
The heatup and cooldown rates shall not exceed 60oF/hr and 100oF/hr, respectively.
Limit lines for cooldown rates between those presented may be obtained by interpolation.
b.
Figures 3.1-1 and 3.1-2 define limits to assure prevention of non-ductile failure only.
The limit lines shown in Figures 3.1-1 and 3.1-2 shall be recalculated periodically using methods discussed in the Basis Section.
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator vessel is below 70oF.
The pressurizer heatup and cooldown rates shall not exceed 100oF/hr nd 200oF/hr, respectively.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320oF.
Fracture Tou hness Pro erties The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the Summer 1965 Section III of the ASME Boiler and Pressure Vessel
- Code, Reference (1),
and ASTN E185, Reference
(?),
and in accordance with additional reactor vessel requirements.
These properties t
8212140413 821208 PDR ADOCK 05000244 p
PDR 3.1-5 Amendment No. ~i~
Proposed
~ g
are then evaluated in accordance with Appendix G of the 1972 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel
- Code, Reference (3) and the calculation methods described in Reference (4).
Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT at the end of 21 '
effective full power years (EFPY).
The 21.0 EFPY period is chosen such that the limiting RTNDT at the 1/4 T location in the core region is higher than the RTNDT of the limiting unirradiated material.
This service period assures that all components in the Reactor Coolant, System will be operated conservatively in accordance with Code recommendations.
The highest RTNDT of the core region material is determined by adding the radiation induced RT for the applicable time period to the original RTNDT shown in Reference (5).
The fast neutron (E
1 Mev) fluence at 1/4 thickness and 3/4 thickness vessel locations is given as a function of full power service life in Reference (5).
Using the applicable fluence at the end of the 21 '
KFPY period for 1/4 thickness and the copper content of the material in question, the RTND> is more conservative. than the value obtained from the second capsule of radiation surveillance program.
Values of RTNDT determined in this manner will be used until more results from the material surveillance
- program, when evaluated according to ASTM E185, are available.
The next capsule will be removed at approxi-mately 17 KFPY (see Technical Specification 4.3.1).
The heatup and cooldown curves must be recalculated 3.1-6 Amendment No. gg Proposed
H when the RTNDT determined from the surveillance capsule is greater than the calculated RT D
for the equivalent capsule radiation exposures Heatu and'Cooldown Limit Curves Allowable pressure temperature relationships for various heatup and cooldown rates are calculated using methods derived from Non-Mandatory Appendix G in Section III of the ASME Boiler and Pressure Vessel Code and discussed in detail in Reference (4).
I The approach specifies that the allowable total stress intensity factor (KI) at any time during heatup or cooldown cannot be greater than that shown in the KIR curve for the metal temperature at that
- 3. l-7 Amendment No. gg Proposed
assumed reference flaw.
During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID.
This condition is, of course, not true for, the steady state situation.
It follows that the T induced during cooldown results in a calculated higher KIR for finite cooldown rates than for steady state under certain conditions.
Because operation control is on coolant temperature and cooldown rate may vary during the cooldown transient, the limit curves shown in Figure 3.1-2 represent a
composite curve consisting of the more conservative values calculated for steady state and the specific cooling rate shown.
Details of these calculations are provided in Reference (4).
'emperature requirement for the steam generator corresponds with the measured NDT for the shell of the steam generator.
A temperature difference of 320oF between the pressurizer and reactor coolant system maintains thermal stresses within the pressurizer spray nozzle below design limits.
(1)
ASME Boiler and Pressure Vessel Code Section III (Summer 1965)
(2)
ASTM E185 Surveillance Tests on Structural Materials in Nuclear Reactors (3)
ASME Boiler and Pressure Vessel
- Code,Section III, Summer 1972 Addenda (note Code Case 1514)
(4)
W.
S. Hazelton, S. L. Anderson, and S.
E. Yanichko,
, WCAP-7924, "Basis for Heatup and Cooldown Limit Curves" (5)
Analysis of Capsule T from the Rochester Gas and Electric Corporation R.
F.. Ginna Nuclear Plant Reactor Vessel Radiation Surveillance Program (WCA>-10086)
- 3. 1-12 Amendment No- ~
Proposed
2600 2400 2200 2000
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TERROR = 10 F, PERROR= 60 PSI
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~ig~>e Reactor Coolant System Cooldown Limitations Applicable for 21.0 Effective Full Power Years.
TERROR 10 F,
PERROR = 60 PSI
0
4.3 REACTOR 'COOLANT"SYSTEM Applicabilit Applies to surveillance of the reactor coolant system and its components.
To ensure operability of the reactor coolant system and its components.
S ecifications:
4.3.1 4.3.1.1 Reactor Vessel Material Surveillance Testing "The reactor vessel material surveillance testing program is designed to meet the requirements of Appendix H
to 10 CFR Part 50.
This program consists of the metal-lurgical specimens receiving the following test:
- tensile, charpy impact and the VOL test.
These tests of the Radiation Capsule Specimens shall be performed as follows:
~Ca sule D
Time Removed for Testin (Removed in 1971)
(Removed in 1974)
(Removed in 1980) 17 EFPY at nearest refueling N
Standby Standby 4.3.1.2 The report. of the Reactor Vessel Material Surveillance shall be written as a
Summary Technical Report as required by Appendix H to 10 CFR Part 50.
4.3.?
Pressurizer 4 ~ 3
~ 2 ~ 1 The pressurizer water level shall be verified to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during power operation and hot shutdown.
4'-1 Amendment, No. ~,
Proposed
Attachment B
The analysis of the reactor vessel material contoured in the third reactor vessel material surveillance capsule is reported in the enclosed report entitled "Analysis of Capsule T from the Rochester Gas and electric Corporation R.
F.. Ginna Nuclear Plant Reactor Vessel Radiation Surveillance Program (NCAP-10086)".
The capsule was removed from the Ginna pressure vessel in the Spring of 1980 after 6.8 full power years of plant operation.
The analysis demonstrates that the plant heatup and cooldown curves currently being used for plant operation are appropriate for use to 21 effective full power years of operation.
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