ML17321A725
| ML17321A725 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 07/18/1985 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | |
| Shared Package | |
| ML17321A724 | List: |
| References | |
| NUDOCS 8507230268 | |
| Download: ML17321A725 (11) | |
Text
ATTACHMENT NO.
1 TO AEP:NRC:0894E TECHNICAL SPECIFICATION CHANGE DESCRIPTIONS Change No.
1 Unit No. 1; Page 3/4 4-27; Figure 3.4-2 Page 3/4 4-28; Figure 3.4-3 Page B 3/4 4-6; Section 3/4 4.9 Page B 3/4 4-10; Table B 3/4.4-1 Page B 3/4 4-11; Section 3/4 4.9 Figure 3.4-2 and Figure 3.4-3 are the heatup and cooldown curves for Unit No. 1.
These curves are composite curves prepared based upon the most limiting value of the predicted adjusted reference temperature at the end.
of 12 EFPY.
The revisions to the curves reflect the analyses done as a
result of the capsule (Y) testing by Southwest Research Institute and the material data for the reactor vessel beltline region weld chemistry noted in our letter AEP:NRC:0894C dated July 3, 1985.
The assumptions and limitations used in the development of the curves are noted in the revised bases.
These revised curves supersede the curves submitted for Unit No.
1 in our earlier submittal No. AEP:NRC:0894 dated February 14, 1985.
The bases for Pressure/Temperature limits were re-written to reflect the current analyses and to read similarly to the Unit No.
2 bases for consistency.
The proposed change constitutes an additional limitation, restriction or control not presently included in the Technical Specifications and complies with changes in the federal regulations.
We believe that the results of the change are clearly within all acceptable criteria since they are the direct result of an evaluation using methods previously accepted by the NRC.
Therefore we believe the proposed changes do not involve a significant hazards consideration as defined by 10 CFR 50.92.
8507230268 850718 ADOCK 05000S15 I
P PDR
ATTACHMENT 2 AEP."NRC:0894E
n n00 2800 2600 2400 2200 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS AP-PLICABLE FOR FIRST 12 EFFECTIVE FULL POWER YEARS.
(MARGINS OF 60 PSIG AND 104F ARE IN-CLUDED FOR POSSIBLE INSTRUMENT ERROR.)
LEAK TEST LIMIT CD M
M De Oe s
QP 4J CO CO u
cj O
O 0
4JV EQ 4l CC 2000 1800 1600 1400 1200 1000 800 600 400 12 EFPY RT NDT 1/4T)
= 2344F 3/4T)
= 117'F PRESSURE-TEMPERATURE LIMIT FOR HEATUP RATES UP TO 60 F/HR MATERIAL PROPERTY BASIS WELD METAL CU = 0.31%,
P= 0.017%
INITIALRT
= 04F NDT UNACCEPTABLE OPERATION ACCEPTABLE OPERATION CRITICALITY LIMIT 200 60 100 150 200 250 300 350 400 450 50Q R0 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE
( F)
Figure 3.4-2 REACTOR COOLANT SYSTEM PRESSURE - TEMPERATURE LIMITS VERSUS 604F/HOUR RATE CRITICALITY AND HYDROSTATIC TEST LIMIT
0
O00Ã 2800 2600 2400 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS AP-PLICABLE FOR FIRST 12 EFFECTIVE FULL POWER YEARS.
(MARGINS OF 60 PSIG AND 10 F ARE IN-CLUDED FOR POSSIBLE INSTRUMENT ERROR.)
2200 2000 MATERIAL PROPERTY BASIS 1800 1600 1400 1200 s'lELD METAL CU INITIALRT NDT 12 EFPY RT NDT 0.31%,
P = 0.017%
PoF (1/4T)
= 234 F
(3/4T)
= 117oF UNACCEP OPERAT TABLE ION ACCEPTABLE OPERATION JJ C
Cg 0
O CJ 0
4JV EQ Cl CC 1000 800 600 400 PRESS URE TEMPERATURE LIMITS COOLDOWN RATE oF/HR 0 F/Hr I
200 60 F/Hr Ir: ".
100 150 200 100 F/Hr
(
250 300
~
~.il 350 400 450 500 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE
( F)
Figure 3.4-3 REACTOR COOLANT SYSTEM PRESSURE - TEMPERATURE,.LIMITS VERSUS COOLDOWN RATES
REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to with-stand 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 or 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 severalfactors 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 tough-ness is dependent upon irradiation and temperature and therefore the fluence profile thzough the reactor vessel wall, the rate of heatup and also the r'ate 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 0
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 pre-pared based upon the most limiting value of the predicted adjusted reference temperature at the end of 12 EFPY.
The reactor vessel materials have been tested to.determine their initial RT
- 'the results of these tests are shown in Table B 3/4.4-1.
NDT 4
Reactor operation and resultant fast neutzon (E >
1 Mev) irradiation wall cause an increase in the RT Therefore, an adjusted reference tem-..
- perature, based upon the fluence and copper content of the material in
- question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2.
The heatup and cooldown.limit cu res of Ficures 3.4-2 and 3.4-3 include ore-dicted adjustments for this shift in RT 't the end of 12 EFPY, as well NDT as adjustments or possible errors in tne pressure and temperatuze sensing instruments.
D. C.
COOK - UNIT 1 B 3/4 =4-6 Amendment'No.
AEACTOtl. VESSEL TOUG1INESS COHI'OttENT MELD +
ttAZ CORf COHP CODE 14GV
, 15GV HATERIAL CU TYPE P
NOTT F
MfLD 0.27 0.023
"-70 itA2
.-60 50 FT-LD/35 HIL TfHP F LONHfBAHS ttA Ot
-40 HA HIH. UPPER S11ELF RTNDT FT-Lli.
i lBB~R,:
Wl
-52 HA
".114**
HA 118 NA SURV ESTIHATfD 60 F DR 100 FT-LD TfHP, MttICttfVER.IS LESS) fSTIHATED 77 FT-LD/54 HIL TEHP FOR I.ONGITUDIHAL DATA) fSTIHATED 65 PERCENT OF LONGITUDINAL SIIELF)
PlloanttLE HATERIAL FOR SURVEILLANCE PAOGAAH ACCOttDING TO f185 FOR PURPOSES OF CALCULATING SUBSEQUENT HEAT UP AND COOLDOWN CURVES HELD MATERIAL CHEMISTRY OF 0.31%
CU AND 0.017%,
P SHALL BE USED.
~ I 4-
~
REACTOR COOXANT hVSVm BASES The shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accord-ance with ASTM E185-73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. 'ince the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the transi-tion shift for a sample can be applied with confidence to the ad)acent section of the reactor vessel.
The heatup and cooldown curves must be recalculated when the b RT determined from the surveillance capsule is different from the calculated ~ RT for the equivalent capsule NDT 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 pro vided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50
'he number of reactor vessel irradiation surveillance
'spe'cimens and the frequencie's 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
'he 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 "<<o PORVs, one PORV and the RHR safety valve, or an RCS vent opening of greater than or equal to 2 square inches ensures that the RCS will be protected from pressure transfents which could exceed the 1fefts of Appendfx G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 188'F.
Either PORV or RHR safety valve has adequate relieving capabf1fty to protect the RCS frea overpressurfzatfon when the transient fs limited to either (1)'he start of an idle RCP with the secondary ~ater tempera-ture of the steaa generator less than or equal to 50~F above the RCS cold leg temperatures or (2) the start of a chargfrig puap and its fn$ectfon into a water solid RCS.
D, C COOK-UNZT 1 B 3/44 11 Amendment No.
0 eL r
+4