ML19325E811

Proposed Tech Specs Re Heatup,Cooldown & Inservice Leak & Hydrostatic Testing Pressure/Temp Limits Based on 15 EFPY of Reactor Operations

ML19325E811
Person / Time
Site:	Crystal River
Issue date:	10/31/1989
From:	FLORIDA POWER CORP.
To:
Shared Package
ML19292J545	List: ML19325E810 ML19325E811
References
NUDOCS 8911090036
Download: ML19325E811 (9)

Text

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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE Lili[IS 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, 3.4-3, and 3.4-4 during heatup, cooldown, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 50'F in any one hour period,

b. For the temperature ranges specified below, the cooldown rates should be as specified:

s

i. T > 280'F s 50'F in any 1/2 hour period i ii. 150*F < T s 280'F s 25'F in any 1/2 hour period iii. T $ 150*F s 10'F in any 1/2 hour period and

c. A maximum temperature change of less than or equal to 5*F in any  ;

one hour period during hydrostatic testing operations above system  !

design pressure. ,

i APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits 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 i 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 RCS T,vg and pressure to less than 200*F and  !

500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. j i

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i CRYSTAL RIVER UNIT 3 3/4 4-24 Amendment No.

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8911090036 891031 PDR ADOCK 05000302 P PDC

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Figuros3.4-2 l e REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS  !

FOR HEATUP FOR FIRST 15 EFPY l 2500- .

2500- The regions of acceptable operation are  !

below and to the right of the limit i l

2400- curve. Margins are included for the pressure differential between point of m.

system pressure measurement and the ,

pressure on the reactor vessel region .,

2200- controlling the limit curve. Margins --

h ,

of 25 psig and 10"F are included for ,' ,

possible instrument error. l 2000 ,'

i 1900- 4 m.-

i, 2000- 4. ,

n l

1700-1600 l 1500- 1 ._ \

,- ~

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g g 1400 l 9 >

1300 i l

1- 1200- F _

l 1100- /

u ., '

I g 1000 '

Applicable for heatup rates of 5 50*F

/

in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period j

l h

4 900 i

S 800' / I 8 E I F DATA POINTS (+):

" 700 / POINT TEMP PRESS

/ (deg F) (psig)

(! ' A 70 278 600 '

B 136 278 ,

500 C C 160 294

" D 185 329

( E 210 384 i 400 ,

F 235 467 f c

- 0 260 586 l

I 300 o i-f'r 0 -

H 290 800 E I 310 996 200 J 330 1217 ,

K 350 1510 100 L 335 1790

, i l l M 385 2250 i 0 .

i a i 50 100 150 200 250 300 350 400 450 500 550 600 INDICATED ACS INLET TEMPERATURE (deg F)

CRYSTAL RIVER UNIT 3 3/4 4-26 AMENDMENT NO.

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Figura 3.4-3 "

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REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS [

FOR COOLDOWN FOR FIRST 15 EFPY  ;

F Ii :

2600 The regions of acceptable operation are '

2500- below and to the right of the limit l curve. Margins are included for the  !

1 2400- pressure differential between point of system pressure measurement and the  ! ,

i l 2300- pressure on the reactor vessel region controlling the limit curve. Margins  ;,,

^

2200- of 25 psig and 10*F are included for l )

possible instrument error. l ,

2100- /

i  ! i 2000- 1. When the Decay Heat Removal Sptem

/

3ggg, is operating with no RC pumps l ,

operating, the . indicated DHR system f return temperature to the to the 1800- reactor vessel shall be used. ,I j

2. A maximum step temperature change 1700- j of 25'F is allowable when removing all RC pumps from operation with l 1600- the DHR system operating. 'Ihe step  ;

temperature change is defined as the '

1500- RC temp, Tc,(prior to stopping all J w In RC pumps) minus the DHR return l 1400- temp, T., (after stopping all RC l  :

Pumps). j 1300-l' l 1200- l

, l i L

1100

[

Applicable for cooldown rates of:  ;

1000 l T>280*F $50*F in any 1/2 hour period L p j c 280aF2T2150*F 525'F in any 1/2 hour period I p! 900 150* FAT 510*F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period i i

.. H 800 / q l E / DATA POINTb (+):

M 700 / POINT TEMP PRESS

/ F (deg F) (psig) 600 f A 70 260

/ B 160 260 500 J_ C 185 313 D 210 367 400 , / E 235 491 n F 260 649 0 290 952

=

300 "r E H 330 1461 I 391 2250 200 A  ;

100 0 i i i a a i 50 100 150 200 250 300 350 400 450 500 550 600 INDICATED RCS INLET TEMPERATURE (deg F)

CRYSTAL RIVER UNIT 3 3/4 4-27 AMENDMENT NO.

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' Figure 3.4-4 o PEACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR INSERVICE LEAK & HYDROSTATIC TESTS FOR FIRST 15 EFPY 2600 2500- The regions of acceptable operation are .

below and to the right of the limit ----

l 2400- curve. Margins are included for the l pressure differential between point of l 2300- system pressure measurement and the T pressure on the reactor vessel region .

c ntr lling the limit curve. Margins j 2200-of 25 psig and 10aF are included for l 2100 Possible instrument error. l 2000- For cooldown, Notes 1&2 on Figure  ;

. 3.4 3 are applicable l 1900- l

1. _ 1600 ,

( l .

1700- '

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1600- l 1500- i l 5  !

l y i400- , l -

(- .9 T 1300 / t r b 1200 l

' r

/ App!! cable for heatup rates of $ S&F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period and for cooldown 1100

? rates of:

1 p h1000 ,

T>280*F <5&F in any 1/2 hour period u ,

r

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280aF2T2150*F $25aF in any 1/2 hour period h 800 aC f

r r 150aF2T 510aF in any I hour period S 800 l

f 8 8 E ---

/ DATA POINTS (+):

1- M 700 f POINT TEMP PFIESS L  ? " (deg F) (psig) l 600 . ( ,

A 70 391 391 B 160

• 500 , _,. C 192 488

, 5' m D 215 480 400 ' "

E 217 587 a S F 235 666 300 0 260 823 H 290 1107 1 310 1366 200 J 330 1659 K 350 2042 100 369 2500

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0 i , i i i i i i i 50 100 150 200 250 300 350 400 450 500 550 600 INDICATED ACS INLET TEMPERATUAE (deg F)

CRYSTAL RIVER UNIT 3 3/4 4-28 AMENDMENT NO.

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E' g 1 CRYSTAL RIVER - UNIT 3 B 3/4 4-7 Amendment No.

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.i CRYSTAL RIVER - UNIT 3 8 3/4 4-8 Amendment No.

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g BASES TABLE 4-1 f REACTOR VESSEL TOUGHNESS E RT y ADJUSTED IWT FOR

!3 MATERIAL CU NI RT TRANS UPPER SHELF 21 FULL POWER YEARS COMPONENT TYPE W/0 W/0 NDT/F FT-LB 9 1/4 t *F-- 9 3/4 T. *f a b

Nozzle Belt SA-508 CL 2 . 10 .72 +10 183 117 101

• Upper Shell SA-5338 . 20 .54 +20 88 162 126

" ** Upper Shel'i SA-5338 . 20 . 54 +20 90 162 126 Lower Shell SA-5338 . 12 . 58 -20 119 87 66 Lower Shell SA-533B . 12 . 58 +45 88 142 121

• • • Surveillance Weld .30 ---

+43 63 Upper Long Weld .29 . 55 (+20)**** 66**** 199 153-Upper long Weld . 29 . 55 (+20)**** 66**** 199 153 Upper Circus Weld . 18 . 63

(+20)**** 66**** NA 133 (60%)

Upper Circum Weld .26 . 61 (+20)**** 66**** 187 NA (40%)

• Middle Circum Weld . 35 . 59 /,+20)**** 66**** 222 168 R' (I00%)

tower Long Weld . 29 . 55 (+20)**** 66**** 199 153-i- (100%)

Lower Circum Weld . 31 . 59 (+20)**** 66**** 69 65 (100%)

Dut 1st Nonle Weld . 19 ---

(+20)**** 66****

Middle Circum A typical weld .41 . 10 +90 211 180 (1005)

• Surveillance Base Metal A

• • Surveillance Base Metal B

• • • Surveillance Weld

• • • • Estimated Value N

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U REACTOR COOLANT SYSTEM BASES  ;

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer vall of the vessel becomes the <

controlling location. The thermal gradients established during heatup produce l tensile stresses at the outer wall of the vessel. These stresses are additivo y to the pressure induced tensile stresses which are already present. The thermal  !

induced stresses at the outer wall of the vessel are tensile and are dependent l on both the rate of heatup and the time along the heatup ramp; therefore, a lower '

bound curve similar to that described for the heatup of the inner wall cannot i

be defined. Consequently, for the cases in which the outer wall of the vessel

becomes the stress controlling location, each heatup rate of interest must be l anelyzed on an individual basis.

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 50'F per hour. During cooldown, similar l ,

types of thermal stress occur. Thus, the cooldown limit curve Figure 3.4 3, is  !

also a composite curve which was prepared based upon the same type analysis as l the heatup curve with the exception that the controlling location is always the  ;

inside wall where the cooldown thermal gradients tend to produce tensile stresses '

while producing comprehensive stresses at the outside wall.

During the 0,. t several years of service life, the most limiting Reactor Coolant System regions are the closure head region (due to mechanical loads resulting from bolt pre load) and the reactor vessel outlet nozzles. Nozzle sensitivity

, is caused by the high local stresses at the inside corner of the nozzle which  !

l can be two to three times the membrane stresses of the shell. After the first '

several years of neutron radiation exposure, the beltline region of the reactor i vessel becomes the most limiting region due to material irradiation. ,

for the service period for which the limit curves are established, the pressure / temperature limits were obtained through a point-by-point comparison '

l of the limits imposed by the closure head region, outlet nozzles, and the most I sensitive material in the beltline region. 1,.e lowest pressure calculated for ,

these three regions becomes the maximum allowable pressure for the fluid temperature used in the calculation. The calculated pressure / temperature curves are adjusted by 25 PSI and 10*F for possible instrument errors. The pressure  ;

limit is also adjusted for the pressure differential between the point of '

pressure measurement and the limiting component for all combinations of reactor coolant pump operations.

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l l CRYSTAL RIVER - UNIT 3 B 3/4 4-10 Amendment No.

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, Irradiation damage to the beltline region can be quantified by determining the

! decrease in the temperature at which the metal changes from ductile to brittle A fracture (ARTwo7). The unirradiated transverse impact properties of the beltline region have been determined for those materials for which sufficient amounts of materials were available and are listed on Table 41. The adjusted reference temperature (ARTwo7) and the unirradiated reference temperature. (The assumed unirradiated RTwot of the closure head region and of the outlet nozzle steel forgingswas60'F.) The adjusted RTwo7s of the beltline region materials at the end of the twenty first. full power year are listed on Table 41 for the one- l quarter and three quarter wall thickness of the vessel wall.

l The actual shift to RTwor of the beltline region material will be established periodically during operation by removing and evaluating the 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 the radius are essentially identical, the measured transition shift for a sample can be a) plied with confidence to the adjacent section of the reactor vessel. The lim't curves must be recalculated when the RTwor determined from the surveillance capsule is different from the calculated RTwor for the equivalent capsule radiation exposure. The pressure and temperature limits shown on Figure 3.4-4 for inservice leak and hydrostatic l testing, have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

The limitations imposed on pressurizer heatup and cooldown and spray water tem)erature differential are provided to assure that the pressurizer is operated wit 11n the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

3/4 4.10 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components, except steam generator tubes, 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.

l The internals vent valves are provided to relieve the pressure generated by L steaming in the core following a LOCA so that the core remains sufficiently I covered. inspection and manual actuation of the internals vent v61ves 1) ensure OPERABILITY, 2) ensure that the valves are not stuck open during normal operation, and 3) demonstrate that the valves are fully open at the forces assumed in the safety analysis.

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CRYSTAL RIVER - UNIT 3 8 3/4 4-11 Amendment No.