ML17325B292
| ML17325B292 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/25/1989 |
| From: | INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | |
| Shared Package | |
| ML17325B291 | List: |
| References | |
| NUDOCS 8911060267 | |
| Download: ML17325B292 (14) | |
Text
REACTOR COOLANT SYSTEM 3 4.4.9 PRESSURE TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
a, A maximum heatup of 60 F in any one hour period.
0 b.
A maximum cooldown of 100 F in any one hour period.
0 0
c.
A maximum temperature of less than or equal to 5 F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times.
ACTION:
'I With any of the above limits exceeded, restore the temperature and/or pressure within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and 0
ave pressure to less than 200 F and 500 psig, respectively, within the foYxowing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE UIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup,
- cooldown, and inservice leak and hydrostatic testing operations.
4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-5.
The results of these examinations shall be used to update Figures 3 '-2 and 3.4-3.
8911060267 891025 PDR ADOCK 05000316 P
PNU COOK NUCLEAR PLANT - UNIT 2 3/4 4-24 AMENDMENT NO ~
O QJ 2600 2400 2200 2000 1800 1600 1400 1200 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR FIRST 12 EFFECTIVE FULL POWER YEARS. ( MARGINS OF 60 PSIG AND 10 F ARE INCLUDED FOR POSSIBLE INSTRUMENT ERROR.)
- LEAKTEST LIMIT NATERIAL PROPERTY BASIS BASE NETAL CU ~ 0.15K NI 0.57K.
INITIALRTNPT ~ 58 F
12 EFPY RTNPT (I/4T) 178 F
3/4T) 150 F
UNACCEPTABLE OPERATION ACCEPTABLE OPERATION 1000 800 PRESSURE"TEMPERATURE LIMITFOR HEATUP RATES UP TO 60oF/HR RITICALITY LIMIT 600 400 200 O
50 100 150 '00 250 300 350 400 AVERAGE. REACTOR COOLANT SYSTEM TEMPERATURE (deg F)
FIGURE 3.4-2 REACTOR COOLANT SYSTEH PRESSURE - TEkFERAllSE LIHlTS VERSUS 60 dagW/W RATE CRITICALITYerr ue HYDROSTATIC TEST LIMIT 450
S rj
~
+h F\\
1 I
O I1j Q
M Q
UlI-U)
M I-M O
g 0D K
OI-D tuK 2600 2400 2200 2000 1800 1600 1400 1200 1000 800 600 400 200 HATERIAL PROPERTY BASIS BASE NETAL Cu ~ 0.15K Nl 0.57K INITIAL RTNDT 58 -F 12 EFPY RTNDT (I/4T) 178 F
(3/4T)
~ 150 F
UNACCEPTABLE OP E RATION PRESSURE-TEMPERATURE LIMITS COOLDO RATE F
W IH 0
20 -.
40 -.
60:
400 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE FOR FIRST 12 EFFECTIVE FULL POWER YEARS. ( MARGINS OF 60 PSIG AND 10 F ARE INCLUDED FOR POSSIBLE INSTRUMENT ERROR.)
ACCEPTABLE OPERATION a
O 50 100 150 200 250 300 350 400 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE (deg F)
FIGURE 3.4-3 REACTOR COOLANT SYSTEI4 PRESSURE - TEl&ERATINE LIMITS VERSUS COOLDOMN RATES 450
E g
44 tg)
',j.
ail C~.,
E
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 for design purposes are provided in Section 4.1.4 of the FSAR.'uring 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 tough-ness 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 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 on 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 RTNDT,'
The results of these tests are shown in Table B 3/4.4-1.
Reactor operation and resultant fast neutron (E ) 1 MeV) irradiation will cause an increase in the RT Therefore, an adjusted reference tem-perature must be predicted 7n accordance with Regulatory Guide 1.99, Revision 2.
This prediction is based on the fluence and a chemistry factor determined from one of two Positions presented in the Regulatory Guide.
'Position (1) determines the chemistry factor from the copper and nickel content of the material.
Position (2) utilizes surveillance data sets which relate the shift in reference temperature of surveillance specimens to the fluence.
The selection of Position (1) or (2) is made based on the availability of credible surveillance
- data, and the results achieved in applying the two Positions.
COOK NUCLEAR PLANT - UNIT 2 B 3/4 4-6 AMENDMENT NO.
INTENTIONALLY LEFT BLANK COOK NUCLEAR PLANT - UNIT 2 B 3/4 4-7 AMENDMENT NO.
0 1
'LI
'1
INTENTIONALLY LEFT BLANK COOK NUCLEAR PLANT - UNIT 2 B 3/4 4-8 AMENDMENT NO,
P 1
I
't P.
I %
TABLE B 3 4.4-1 REACTOR VESSEL TOUGHNESS 50 FT-LB 35 MIL USE COMPONENT CL. HD.
DOME CL. HD.
SEG.
CL. HD.
SEG.
CL.
HEAD FLG.
VESSEL FLANGE INLET NOZZLE INLET NOZZLE INLET NOZZLE INLET NOZZLE OUTLET NOZZLE OUTLET NOZZLE OUTLET NOZZLE OUTLET NOZZLE UPPER SHELL UPPER SHELL UPPER SHELL INTER SHELL INTER SHELL LOWER SHELL LOWER SHELL BOT. HD.
SEG.
BOT. HD.
SEG.
BOT. HD.
SEG.
INTER. & LOWER SHELL LONG. and GIRTH WELD SEAM CODE NO.
B0048-2 B9883-2 A5189-2 4437-V-1 4436-V-2 269T-2 270T-1 269T-1 270T-2 271T-1 271T-2 272T-1 272T-2 C5518-2 C5521-1 C5518-1 C5556-2 C5521-2 C5540-2 C5592-1 C5823-2 A4957-3 B0019-18 (HT S3986 Linde 124 Flux Lot No. 0934)
MATERIAL TYPE A533B CL. 1 A533B CL. 1 A533B CL. 1 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 SAW CU NA NA NA NA NA NA NA NA NA NA NA NA NA
.12
~ 14
~ 12
.15
.14
.11
.14 NA NA NA
.06 NI 0.64 0.66 0.63
. 0.70 0.70 0.85 0.91 NA NA 0.80 0.80 NA NA 0.61 0.59 0.57 0.57 0.58 0.64 0.59 0.57 0.51 0.61 0.97
-20
-20 10
-20 30
-20
-20
-10
-10 0
0
-10 0
10 0
10 0
10
-20
-20
-10
-10
-50
-40 Temp(a)
~P 30
-3 72 5
15
-15
-3 NA NA 12
-15 NA NA 88 93 66 118(b) 98(b) 35(b) 25(b) 45 20 0
25
-20
-20 12
-20 30
-20
-20
-10
-10 0
0
-10 0
28 33 10 58(b) 38(b)
-20(b)
-'20(b)
-10
-10
-50
-35(b)
MWD
~ft-1b 148 143.5 140.5 239 161 201.5 239.5 NA NA
>179 181 NA NA 107.5 112
> 82.5 109.5 111.5 113 107 129 149 177 NA NMWD(a)
~ft-1b 96 93 91 155 105 131 156 NA 117. 5 NA 70 73 90(b) 86(b) 110(b) 103(b) 84 97 115 97(b) a) Estimated per NRC Standard Review Plan b) Actual values NA - Not available or not applicable, as appropriate COOK NUCLEAR PLANT - UNIT 2 MWD - Ma]or Working Direction NMWD - Normal to MWD B 3/4 4-9 AMENDMENT NO.
INTENTIONALLY LEFT BLANK COOK NUCLEAR PLANT - UNIT 2 B 3/4 4-9a AMENDMENT NO.
l
'E
REACTOR COOLANT SYSTEM BASES The actual shift in the reference temperature of surveillance specimens and neutron fluence is established periodically by removing and evaluating reactor vessel material irradiation surveillance specimens and dosimetry installed near the inside wall of the reactor vessel in the core area, The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT at the end of 12 EFPY, as well as adjustments for possible errors in the pressure and temperature sensing instruments.
The 12 EFPY heatup and cooldown curves were developed based on the following:
1, The projected fluence values established by specimen analysis.
2, Intermediate shell plate C5556-2 being the limiting material as determined by Position 1 of Regulatory Guide 1.99, Revision 2, with a copper and nickel content of 0.15% and 0.57't, respectively.
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 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.
The OPERABILITY of two 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 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 152 F.
Either PORV or RHR safety valve has 0
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 0
to 50 F above the RCS cold leg temperatures or (2) the start of a charging pump and its injection into a water solid RCS.
3 4.4.10 STRUCTURAL INTEGRITY The inspection and testing programs for ASME Code Class 1,
2 and 3
components ensure that the structural integrity 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.
COOK NUCLEAR PLANT - UNIT 2 B 3/4 4-10 AMENDMENT NO.