ML17215A429

Marked-up Proposed Tech Specs Reflecting Changes Required to Commence Cycle 2 Operation

ML17215A429
Person / Time
Site:	Saint Lucie
Issue date:	06/04/1984
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17215A430	List: ML17215A429 ML17215A431 ML17301A108
References
NUDOCS 8406110289
Download: ML17215A429 (67)

Text

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE DNBR 2.1.1.1 The combination of THERMAL POWER, pressurizer pressure, and maximum cold leg coolant temperature shall not exceed the limits shown on Figure 2. 1-1.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the combination of THERMAL POWER, pressurizer pressure and maximum cold leg coolant temperature has exceeded the limits shown on Figure 2. 1-1, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7. 1.

PEAK LINEAR HEAT RATE

2. 1.1.2 The peak linear heat rate of the fuel shall be maintained less than or equal to MkW/ft (value corresponding to centerline fuel melt).

22. o APPLICABILITY: MODES 1 and 2.

ACTION:

22. 0 Whenever the peak linear heat rate of the fuel has exceeded . kW/ft (value corresponding to centerline fuel melt), be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7. 1.

REACTOR COOLANT SYSTEM PRESSURE

2. 1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant'System pressure within its limit within 1 hour, and comply with the requirements of Specification 6. 7. 1.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded '2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7. 1.

ST. LUCIE - UNIT 2 2-1 840b04 840bi10289 05000389 PDR *DOCK I p

n rrt UNACCEPTABLE OPERATION

~ ~

no '"'>> -i -U NACCEPT L OPER N 660 I ... I >>

~

':

I ~ ~ ~ " ~ ~

'-

--I i

=l>> 5

. >> (b VESSEL FLOW LESS ME UREMENT

. UNCERTAIN>>TIES.~ 370. PM

.>>

640 LIMITS CONTAIN NO ALLOWANCE ->>

FOR INSTRUMENT ERROR OA A

~ - FLUCTUATIONS 0

VALID FOR AXIALSHAPES AND 0

". INTEGAATED ROD RADIALPEAKING >>

9 FACTORS LESS THAN OA EQUAL TO O THOSE ON FIGURE 8 2.1-1 620 REACTOR OPERATION LIMITEDTO LESS:" rlu V

. THAN 660'F BY ACTUATION OF THE

,... PZ~ZI

.:: gm>ml SECONDARY SAFETY VALVES m

" -~2'Ch'". ~

I

gr 600 Cv

,... mppp..

r:.'::

'.!- .'::- rg>gi ~ ~ i

'u tugr ". ~ ~

< pj:..':.

ACCEPTABLE OPERATION Q ll ZHm I App ~ I

Z

~ ~

'"

1.40 1.60 1.80 2.00 FRACTION OF RATED THERMALPOWER Rgure 2.1-1 Reactor core thermal margin safety limit lines Four reactor coolant pumps operating

\ ~ ~

!

i el

~ ~ ~ ~ e ~

~ I ~ ~ ~

eli!

'

j'l:

~ e

~~~ ~ ~ ~

... ff!f ~~

~

I

~

~ ~

~ ~

~~

~ ~ ~ ~ ~~ ~ ~~ ~~~ ~~~ ~ .::e:I ill Cfl Al~ Ce

~ ~

il.

RATION ~

D }1 I"

i "l'~ i II":

~

Illf ~~ ~

i,

~

~ ~ ~

IINCCEPTABLE e

~ ~

~ ~

~

)

~ I'

~

I OPK RAT IQI ~ ~

I .:I it

~ e ~

~

I Il ff: ';i 8

I l: }

~

I ~ l ~

g~

I. C+

~ ~ ~ ~ ~~ ~~ O I ~ ~ I ~ 5 I

il

'I ~ II

~

.~

~~

I':

SEO g i I ,.. .}tf f ~~

'.:.

g g+

P- ~ ~

I

.I

~ ~ ~

~ ~

LIIIITS CNITAIII NO ALLONlICK I)

~ ~ ~~

~

)

FOR IISTROKNT ERRN OR ;I l IO oQ p

FLUCTNITINS I ~

'e i}' :i

~t fe ~~ ~~

520 f~ ~

0

)I f I ~

~ eje

$ gpg

~ e~

NLIO FOR ANNAL SHAPES AN p rt 9 IIITEGRATEO ROO RADIAL PEAKING .I

'I }:' ~ ~

I 8 FACTORS LESS TINN OR EgNL TO TINSE Nl FIINRE B 2'.I-I I pic) ~

~

~ ~~

l4 K I

500 I

REACTOR OPERATINI LIIIITEO TO I.KSS QXIDQ I

TIWI 580"F SY ACTNTINI OF YllE M

'0 CO Q4 g ~ ~ ~ I ~

SECOND SAFETV VALVES .I C QX

}~

I ~~

eief I myA 'f.[ fe)

~ ~

I QUD ~

1 ACCEPTNI.E )s f OPERATION il il I

I  ! }

I O.ao 0.60 0.80 ).00 I.20 1.40 }.60 2. 00 2.20 FRACTINI QF RATEO TIlEiMAL POMER

TABLE 2.2"1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIHITS I

M f77 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES C:

1. Hanual Reactor Trip Not Applicable i Not Appl cable

2. Variable Power Level - High Four Reactor Coolant Pumps < 9.61X above THERMAL POWER, < 9.6]X above THERHAL POWER, and Operating with a minimum setpoint of a minimum setpoint of 15K'of 15K of RATEO THERMAL POWER, RATEO THERMAL POWER and a maximum and a maximum of < 107.0X of of < 107.0X of RATEO THERMAL POWER.

RATEO THERMAL POWER.

3. Pressurizer Pressure - High < 2370 psia < 2374 psia

4. Thermal Hargin/Low Pressure I

Four Reactor Coolant Pumps Trip setpoint adjusted to not Trip setpoint adjusted to not Operating exceed the limit lines of exceed the limit lines of Figures 2.2-3 and 2.2-4. Figures 2.2-3 and 2.2-4.

Hinimum value of 1900 psia. Hinimum value of 1900 psia.

- /. I g

5. Containment Pressure High < 4. 0 psig < ~psig

6. Steam Generator Pressure - Low > 626.0 psia (2) > 621.0 psia (2)

7. Steam Generator Pressure < 120,0 psid < 132.0 psid Difference - High (Logic in TH/LP Trip Unit)

8. Steam Generator Level - Low > 39.5X (3) > 39.1X (3)

TABLE 2.2-1 Continued REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS I

C:

M fT7 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES I

9. Local Power Oensity - High~ Trip setpoint adjusted to Trip setpoint adjus'ted to not exceed the limit lines not exceed the limit lines of Figures 2.2-1 and 2. 2-2. of Figures 2.2"1 and 2.2-2.

10. Loss of Component Cooling Water > 636 gpm"" > 636 gpm to Reactor Coolant Pumps-Low ll. Reactor Protection System Logic Not Applicable Not Applicable

12. Reactor Trip Breakers Not Applicable Not Applicable

13. Rate of Change of Power - High < 2.49 decades per minute < 2.49 decades per minute 14., Reactor Coolant Flow - Low > 95.4X of design Reactor > 94.9X I

of design Reactor Coolant flow with four Coolant flow with four pumps operating" pumps operating".

15. Loss of Load (Turbine) > 800 psig > 800 psig Hydraulic Fluid Pressure - Low t" Oesign reactor coolant flow with four pumps operating 10-minute time delay after relay actuation.

is,949,o~

gpm.

I C

O III lAO I II '

'll ~

I I I

~

~ ~

~ ~ \ ler ~ ~ ~ ~ ts

I:::ll'I 1)I! I.-, I I' I ~ .I e!lt s ~ ~ ~ ~ ~ ~

C I~ ~ ~

'I ill-':

~ ~ >> ~I t>>

~

z ~ ~

~ ~ I ~ ~ ~ ~

~ ~ ~ ~

e

~ ~ ~

I>> ~ <<

~ ~ ~

rt

=I !I I I--'j i!!s

~ ~

\ ~ ~

~

~ ~

ill! tell ~ ~ ~ I~~ ilia ~ ~ ~

'r

-l

~ ~

IeI I I ~

.6A SI + 1. 10 I ~ ~

I

~

r>>

~ ~ ~

s ~

IIt

~

't ~ It 'II-~

t ~ ~ ~ >> >>

l \

I ~

I~

~

~ >>ret

~

I ~I'

~

I ~

I >>I I <<s '!~ ~ ~

I

~

'e 4<<

>>

t e II

~

~

I I  :;I! };I::-t!le": WHERE: Al x QONa F j'.i I I I:I: AND I~ ~ ~

~ I

~

~

~

si'- ~,

~

P VA = 2061 x QONB + 15.85 x TIN n

~ ~ I

' ' 0

~ / ~

I I~ ~

~ ~ , ~

I. I,>>) ill: '

I ' I

~

~

\

11' i ~

~ sl, I Iel i<<r- r

le ~ ~ ~

~ ~ Il ~

~ I

~

>>I. ~

isl s! r ~ l ft I~t I>>

ttf

~

'I I ~

~

s~ I Ir I I

~ >>

e I I II I ~ ~ ~ 1'l' ~ ~

~ ~

illi !It> !1 t~ ~

~ ~ t

~ ~ ~

fN eilI ~ ~ I C OR EIN LET TEMP. 0F

~ e A) = 0.6 AS I+ 0.9

~ ~ ~ I 1.10 II

~ ~ ~ ~

~

I ~

s I

~ ~

I I 5

~ ~

'il I ~ ~ I

~ ~ ~ >>

ri]>>

~

.00

-O.B -0.4 -0.2 0.0 0.2 4 O.B AXIALSHAPE INDEX, Yl Figure 2.2-3 Thermal marglnllnw pressure trip setpolnt Pert 1 {Y> versus A>)

1.70 1.60 var 1.50 ) l

~ I 1.40 I I i

i

1. 30

! I i

~

1 1.20 1.10 I

I'

. ~

1.00

-0.6 -0.4 -0.2 0;0 0.2 0.4 0.6 AXIAL SHAPE INDEX, .Yl FIGURE 2.2-3 THERNL NRGIN/LOW PRESSURE TRIP SETPOINT .

PART 1 O"1 ~ "sus Al ST. LUCIE-UNIT 2 2-9

>bc's ecch'c.e rcecjte c'e ccc'.< 4<e 4(l~'~ P~qa ~

WHERE: + x QR~ = Q N AND vAR

= 2061 x QoNs + 15.85 x TIN - 9000 1.2 1.0 O.S 1.00 O.S5 0.85 0,8

0. 70 OAO 0.2 0.0 0.2 0.4 0.6 0.8 1.0 FRACTION OF RATED THERMAL POWER Figure 2.24 Thermal margin/Iow pressure trip setpolnt Part 2 (Fraction of RATED THERMAL POWER versus QR>)

ST. LUCIE - UNIT 2 2-10

WHERE: Al x QRl QDNB AND P= 1400 x QD B

+ 17.85 x T(- 9410 1.2 1.0 0.95 I

a ~ l I 0 Q 85 I r 0.8 QR1 0.6 ~

~

~ ~

~

~

0.4 0.2

~ .l.

0.0 0.0 0.2 .0.4 '.6 O.S 1.0 FRACTION OF RATED THERMAL POWER FIGURE 2.2-4 THERMAl. MARGIN/LOW PRESSURE TRIP SETPOINT PART 2 FRACTION OF RATED THERMAL POWER VERSUS QRl ST. LUCIE - UNIT 2 2-10

2. 1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perfoJ ation which would result in the release of fission products to the reactor coolant. Over heating, of the fuel is prevented by maintaining the steady-state peak linear heat rate below the level at which centerline fuel melting will occur. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the CE-1 correlation. The CE-1 DNB correlation has been developed to predict the DNB heat flux and the location of DNB for axially uniform and non-uniform heat flux distributions. .The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. a,n acct kcbte l'g~,'.

~

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to

.value corresponds to a 95K probability at a 95K confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating This conditions.

The curves of Figure 2. 1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four

[ g reactor Coolant Pumps operating for which the minimum DNBR is no less than radial peaks shown in for the family of axial shapes and corresponding

, Figure 2. 1-1. The limits in Figure 2.1-1'ere calculated for reactor y coolant B inlet temperatures less than or equal to 580'F. The dashed line at 580 F coolant inlet temperature is not a safety limit; however, operation above 580 F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature.

Reactor oper ation at THERMAL POWER levels higher than 112K of RATED THERMAL POWER is 'prohibited by the high power level trip setpoint specified in Table 2.2-1.

~

The area of safe operation is below and to the left of these lines.

The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure.

The Thermal Margin/Low Pressure and Local Power Density Trip Systems, in conjunction with Limiting Conditions for Operation, the Variable Overpower Trip and the Power Dependent Insertion- Limits, assure that the Specified Acceptable Fuel Design Limits on DNB and Fuel Centerline Melt are not exceeded during normal operation and design basis Anticipated Operation Occurrences.

ST. LUCIE - UNIT 2 B 2-1

I n

C m

2.

T 1.8 F, = 1.69 z

O I

D 1.6 co CC I- 1.4 F, = 1.85 th A

K 1.2 O 1.0 T

F = 1.64 X 0.8 O

0.6 N

0.4 K

0 z 0.2 0.0 PERCENT OF ACTIVE CORE LENGTH FROM BOTTOM Figure B 2.1-1 Axial power distribution for thermal margin safety limits

2.0 1.6 F = 1.79 R

1.4

= 1.77 F = 1.67 FT 1.2 T 1.0 0.0- F T

= 1.62 a4J 0.6 I 0.4 0.2 0.0 25 50 75 100 PERCEflT OF ACTlVE CORE LENGTH FROH 00TTQ4 Figure 0 2.1-1 power distribution for thermal margin safety limits

'xial I

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Variable Power Level-Hi h A Reactor trip on Variable Overpower is provided to protect the. reactor core during rapid positive reactivity addition excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal'Margin/Low Pressure Trip.

The Variable Power Level High trip setpoint is operator adjustable and can be set no higher than 9.62K above the indicated THERMAL POWER level.

Operator action is required to increase the trip setpoint as THERMAL POWER The trip setpoint is automatically decreased as THERMAL POWER is'ncreased.

decreases. The trip setpoint has a maximum value of 107.0X of RATED THERMAL POWER and a minimum setpoint of 15.0X of RATED THERMAL POWER. Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMAL POWER level at which a trip would be actuated is 112K of RATED THERMAL POWER, which is the value used in the safety analyses.

Pressurizer Pressure-Hi h The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at less than or equal to 2375 psia which is below the nominal lift setting 2500 psia of the pressurizer safety valves and its operation minimizes the undesirable operation of the pressurizer safety valves.

Thermal Mar in/Low Pressure The Thermal Margin/Low Pressure trip is provided to prevent operatign when the DNBR is less than&28i %he. ~ccrp4.4lc w<<i~~~ bMB< /i'~i'F'.

The trip is initiated whenever the Reactor Coolant System pressure signal drops below either 1900 psia or' computed value as described below, whichever is higher. The computed value is a function of the higher of bT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function. In addition, CEA group sequencing in accordance with Specifica-tions 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.

The Thermal Margin/Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error. A safety margin is provided which includes: an allowance of 2.0X of RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 3.04F to compensate for potential temperature measurement uncertainty; and a further allowance of 91.0 psia to compensate for pressure measurement error and time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit. The 91.0 psia al.lowance is made up of a 25.0 psia pressure measurement allowance and a 66.0 psia time delay allowance.

ST. LUCIE - UNIT 2 B 2-4

0 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T LESS THAN OR E UAL TO 2OO F LIMITING CONDITION FOR OPERATION 9

3.1.1.2 The GMUTOOWN MARGIN shall be greater than or equal to g.OX delta k/k.

/

APPLICABILITY: MODE 5.

ACTION:

9 With the SHUTDOWN MARGIN less than>.(C delta k/k, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS

4. 1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to g.OX delta k/k:.

3 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable CEA(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s) is inoperable. If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).

b. At least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> s by consideration of the following factors:

l. Reactor coolant system boron concentration,

2. CEA position,

3. Reactor coolant system average temperature, Fuel burnup based on gross thermal energy generation,

5. Xenon concentration,and

6. Samarium concentration.

C. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, when the Reactor Coolant System is drained below the hot leg centerline, by consideration of the factors in 4. 1. 1.2b. and by verifying at least two charging pumps are rendered inoperable by racking out their motor circuit breakers.

ST. LUCIE - UNIT 2 3/4 1"3

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths and one associated heat tracing circuit shall be OPERABLE:

ao Two flow paths from the boric acid makeup tanks via either a boric acid makeup pump or a gravity feed connection, and a charging pump to the Reactor Coolant System, and

b. The flow path from the refueling water tank via a charging pump to the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one of the above required boron injection. flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least Q ~.OX delta k/k at 200'F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS

4. 1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path from the boric acid makeup tanks is above the temperature limit line shown on Figure 3. 1-1.

b. At least once per 31 days by verifying that each valve (manual, power-operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in .its correct position.

c. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on an SIAS test signal.

d. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a delivers at least 40 gpm to the Reactor Coolant System.

ST. LUCIE " UNIT 2 C

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION

3. 1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES I, 2, 3 and a.

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in>at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least N.lC delta k/k at 200 F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS

4. 1.2.4. 1 At least two'harging pumps shal 1 be demonstrated OPERABLE by verifying that each pump develops a flow rate of greater than or egual to 40 gpm when tested pursuant to Specification 4.0.5.

4.1.2.4.2 At least once per 18 months verify that each charging pump starts automatically on an SIAS test signal.

ST. LUCIE - UNIT 2 3/4 1-10

REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 At least the boric acid'makeup pump(s) in the boron injection flow path(s) required OPERABLE pursuant to Specification 3. 1.2.2a shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump(s) in Specification 3. 1.2.2a is OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one boric acid makeup pump required for the boron injection flow path(s) pursuant to Specification 3. 1.2.2a inoperable, restore the boric acid makeup pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY withj.n the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at leastMOX delta k/k at 200 F; restore the above required boric acid makeup pump(s) to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS

4. 1.2.6 The above required boric acid makeup pump(s) .shall be demonstrated OPERABLE by verifying, that on recirculation-flow,. the pump(s) develop a discharge pressure of greater than or equal to 90 psig when tested pursuant to Specification 4.0.5.

ST. LUCIE - UNIT 2 3/4 1-72

REACTIVITY iONTROL SYSTEMS BORATED WATFR SOURCES OPERATING LIMITING CONDITION FOR OPERATION

'3.1.2.8 Each of the following borated water sources shall be OPERABLE:

aO At least one boric acid makeup tank and at least one associated heat tracing circuit'per tank with the contents of the tank in accordance with Figure 3. 1-1, and I

b. The refueling water tank with:

l. A minimum. contained borated water volume of 417,100 gallons,

2. A boron concentration of between 1720 and 2100 ppm of boron, and

3. A solution temperature between 55'F and 1004F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a0 With the above required boric acid makeup tank inoperable, restore the tank to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next.Q hours and borated to a SHUTDOWN MARGIN equivalent to at least ~X delta k/k at 200'F; restore the above required boric acid makeup tank to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.1.2.8 Each borated water source shall be demonstrated OPERABLE:

ao At lea'st once per 7 days by:

l. Verifying'the boron concentration in the water, C

2. Verifying- the contained bor ated water volume of the water source, and

3. Verify'ing the boric acid makeup tank solution temperature.

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWT temperature when the outside air temperature is outside the range of 55'F and 100 F.

I ST. LUCIE - UNIT 2 3/4 1-14

0 REACTIVITY CONTROL SYSTEMS 3/4. 1:3 MOVABLE CONTROL ASSEMBLIES CEA POSITION LIMITING CONDITION FOR OPERATION 3.1.3.1 The CEA Block Circuit and all full-length (shutdown and regulating)

CEAs which are inserted in the core, shall be OPERABLE with each CEA of a given group positioned within 7.0 inches (indicated position) of all other CEAs in its group.

APPLICABILITY: MODES 1" and 2".

ACTION:

aO With one or more full-length CEAs inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in at least HOT STANPBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the CEA Block Circuit inoperable, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

1. With one CEA position indicator per group inoperable take action per Specification 3. 1.3.2, or

2. With the group overlap and/or sequencipg interlocks inoperable maintain CEA grou~s 1, 2, 3, ~and&Fully withdrawn and the CEA's in group~to less than 15K.insertion and place and maintain CEA drive system in either the "Manual" or "Off" position, or

3. 'Be in at least HOT STANDBY.

c With more than one full-length CEA inoperable or misaligned .from any other CEA in its group by more than 15 inches (indicated position),

be in at least HOT STANDBY within 6 hours.

d. With one full-length CEA misaligned from any other CEA in its group by more than 15 inches, operation in MODES 1 and 2 may continue, rovided,that ~ ~

%he nis~lij~ed CA is maioli'i~e erik i'n,lb

~

i'fh h ti e Cy

~

~

~

~

inc."i'-5 oP. %he ofhei- 6'As in

~

s4o~n .'i -i Pt e .I 1o i4 group in accordancp See Special Test Exceptions 3. 10.2, 3. 10.4, and 3.10.5.

us ST. LUCIE - UNIT 2 3/4 1-18

Ai $ h One ftA Ij-lCESdl >'4 CGA m>aatigs< 4 4O N aug Ofher CPA I'WE't+a 0 rOOp g< ~pt e 4ha~ I 5 Erick> bc.yowtl 4 "c <<+4 c'owl>424'4 a4dsww in I=2 jure po~<r o 707ER o F'4c.J >~<>w<1 Pone~ p>iaw 0-o N.l,.ftyler>>

(O~q4.4i~q AC<~4'll C.. I dg.<J g. Z, 4p4E

+)fan)l E: +44 CE A +y ypdmg a))de p Qptmp + >pgz+ <p+ ypdmgI+iLmg

~~/

1 lKCit~. Y'C'g Vlssg REACTIVITY CONTRDL SYSTEMS ACTION'Continued)

2. OeoiareitCinoperahle and the SMUTOQWM MARGIE requirement of

'

~ Specification 3.1.1.1.' After declaring the CEA inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3. 1.3.6 provided:"

a) Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the remainder of the CEAs in the group with the inoperable CEA shall be aligned to within 7.0 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3. 1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1. 3. 6 during subsequent operation.

b) The SHUTDOWN MARGIN requirement of Specification 3.1.1. 1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With one or more ful;-length CEA(s) misaligned from any other CEAs (

in its group by more than 7.0 inches but less than or equal to 15 inches, operation in MODES 1 and 2 may continue, provided that within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the misaligned CEA(s) is either:

1. Restored to OPERABLE status within its above specified alignment- ..

requirements, or

2. Declared inoperable and the SHUTDOWN, MARGIN requirement of Specification 3.1. 1. 1 is satisfied. After declaring the CEA inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3. 1.3.6 provided:

a) Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the remainder of the CEAs in the group with .

the irioperable CEA shall be aligned to within 7.0 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3. 1-2; the THERMAL POWER level shal'l be restricted pursuant to Specification 3 1. 3. 6 during subsequent operation.

~

b) The SHUTDOWN MARGIN requirement of Specification 3.1.1. 1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With one full-length CEA inoperable due to causes other than /

addressed by ACTION a., above, and inserted beyond the Long Term Steady State Insertion Limits but within its above specified alignment requirements, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3. 1.3.6.

With one full-length CEA inoperable due to causes other than addressed by ACTION a., above, but within its above specified align-ment requirements and either fully withdrawn or within the Long Term Steady State Insertion Limits if in full-length CEA group+'P operation in MODES 1 and 2 may continue.

If the pre-misalignment ASI was more 'negative than -0.15, reduce power to < 70K of RATED THERMAL POWER or 70K of the THERMAL POWER level prior to the mis" alignment, whichever is less, prior to completing ACTION gt.2. a) and g.2. b).

e e ST. LUCIE - UNIT 2 3/4 1-19

4'eo j~~iQ Figur .1-la Allowable Time to Reali n CEA vs. Initial F fthm

'I'

,--II e

t-

~

1 ~

Yfj "7

1t.t t.':.: ee t the je.4 t *+ ~

I ~

~

• I 4 et I F 14'"

4 et I

'I 71';tt 4

~ 'I t4 \ 4 1=

~

f4 4 'I ~

~ I\

4 4

1

~ ~ ~

70 (XS , 1 . 70) 41T

~ I e

,I,e

~ 4

-; '8- 4

~ ~ ~ ~ I 44

~

j t..'e. ttf j ~ 44 1~ =

~

Mt4 ~ I ~

1 I~

I ~

I" et' ~ 4 I'lf f1' ttfe '

• 44 ~ ~ ~

If I 1 It I

~

et ~ it 1 cate

'7 7

te 7

~

4 ~ 11. 1 t\ ~" el r 'e

~ I

~ ~

4 e.f, 4 I ttj) tiet41

~ 4

~

'1 tltf I

~

f jt ~+'~ 1 1,'jj

~ *t r-- et

~ -4

~ r

+t 1

~

I ~

Yil e

1 t tet I

jj I

1 ~

~ ' ~ 1 ~ ~ .44 4

I I 'L ~ ~

I 2"'..

~ I jlfj I'

~

'4 4 ~ 4 Ij:1

• 1 f4 I

t 0 t,s 1 *

~ ~

7. e 1 ~ ~ ~ ~

~ ~

41 I" j at Full to Real inute 4

.

Time Power

.~L I'fe v 4

~ 1 1

REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION gI7 3.

from a fully Athdrawn position, shall be less than or equal to ~

1.3.4 The individual full-length (shutdown and regulating) CEA drop time, seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90K insertion position with:

a. T avg greater than or equal to 5154F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With the drop time of any, full-length CEA determined to exceed the above limit:

1. If in or MODE l or 2, be in at least HOT STANDBY within 6 hours,

2. If in MODE 3, 4, or 5, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the CEA drop times within limits but determined at less than full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.

SURVEILLANCE RE UIREMENTS

4. l. 3.4 The CEA drop time of full-length CEAs shall be demonstrated through measurement prior to reactor criticality:

a. For all CEAs following each removal and installation of .the reactor vessel head,

b. For specifically affected individuals CEAs following any main-tenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and C. At least once per 18 months:

ST. LUCIE " UNIT 2 3/4 1-24

J (BANK 6-38", POWER = 100%)

I C

O m

C z LU 0.80 ~ (BANK 5.8", POWER = 80%)

t4 O 0.60 ~ (BANK 5-82". POWER = 60%)

O POWER 0+ENOENT INSERTION LIMIT

u. 0.40 O + ANK 4-54". POWER = 36%)

Z SHORT TERM 3 0 STEADY STATE INSERTION LIMIT-b a: 0.20 L LONG TERM STEADY STATE ~ (BANK 3-27". POWER = 12%)

INSERTI N LIMIT 0.00 (BANK 3-54", PO ER = 0%)

6 GROUPS 0 27 55 82 109 137 0 27 55 82 109 137 0 27 55 82 1 9 137 5 3 0 27 55 82 109 137 0 27 55 82 109 137 CEA INSERTION (INCHES)

~- 'I Figure 3.1-2 CEA Insertion limits vs THERMAL POWER with four reactor coolant pumps operating

~%

c

@o 0 IA 2 o a o a w c C c o c e C9 g

~

g 0 II c op Sg c 0

@Q an a~ Q, Q~ c D N

I o a Qu g

I 8 0.80 c I I jv m

g U (a

a Cl 0 c

=1 )II 0 0.70 "I C3 V O

'-

I 5 c 0.60 I) g'

@I  :"l aW 0.50 I- gl a

> 040 ",I C9 C) 0.30

.I 0.20 4.ONO TEAM-~ -SIIOAT TEAM-STEAOY STEAOY STATE STATE IN S E ATION 0.10 INSEATION LIMIT LIMIT b

Gnoups 0 20 40 60 80 100 0 20 40 60 80 100 0 20 40 60 80 100 (136)(108.8)(81.6) (54.4) (2l.2) (8) (136) (108.8) (81.6) (54.4) (27.2) (0) (136) (108. 8) (81.6) (51.4) (27. 2) (0) 2 0 20 40 60 80 100 0 20 40 60 80 100 (136)(108 8)(81.6) (51.4) (21.2) (0) (136)(108.8)(81.6) (54.4) (27.2) (0)

% CEA INSEATION IINCIIES CEA WITIIOAAWN)

Figure 3.1-2 CEA Insertion Limits vs. Ti(ERNL POWER with Four Reactor Coolant Pumps Operating

1.2 I jj! ! l ' jt't"t tlat, ~

II Ijji turf ttrj'L t

'.llj l tjt !!jj jj Lt li!l III

.'-ilt ptj I ~

.!L 'I I t~

I

~

j I ~ I ~ l t I ~ I~

,'I'I  !!It :tll iftl tl l'. UNACCEPTABLE l:j sts js t l '.

I,'! I. I

~ ~

j ~ ~ ~ , ~ ~ ~ l I ~I ~ ~ ~

I I ~

t~ ~I OPERATION I ~

~ ~

Itl REGION

't's s It ~ ~ ~ ~ ~ ~  !

1.0 II It I 'I tt Ilsl Il l'lt'.:I,s

~ ~

'jsj  ::li ~ ~

~

st js

~

~ I~

~

~ ~

I~

!II

~ ~ ~  :::, -0.1 5,0 925) i: ':I (0.15, 0.925) ~ ~ I ~

~ sl ~ ~ ~ ~ I I I CL

~ I

~ II ~ I I~ s!  !~ t ~ ~

l tsar sl I .I

~ ~ I

~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~

~ I ~ I

~ ~ ~

I 'Lt

~ !I' ~ ~ ~

I

~ ~ ~ ~ ~ ' ~

~

s ,Is ~

s

~

~

I

o. 0.8 I~ ~ ~ ~ ~ ~ I I~ ~ I ~ ~ I ~ st ~

l ~ ~ ~

t ~I

! l.:

~ ~ ~

II t~

~ I! s

~ ~

~ ~ I I'I

~ I s ~ ~ ~ ~ ~ ~ ~

Itsti s

~

s

~

~

I ~ s I~ ~ ~ ~

I~ ~ j ~ ~

I' ~ ~ I ~ ~ ~ ~ I 0.70) I.' ~ ~ ~ I (0.3, 0 .70)

\ ~ I~

I Is  !.'ll I

~ ~

I

~ ~

I

~ ~ ~

~ ~

~ ~ ~

~ ~ s ~

It t ~ ~

~ ~ ~

0.5

.. t~ ~

A CEPTABLE PERATION I!' s ~

i i

REGION '-"!!

X ~ ~ ~ ~

~ ~ st

~

~

stI

~ ~ ~ ~ ~

t ~

I I '

~

~ ~ I~

~ ~ I ~ ~

~ ~ ~ ~ ~ II

~ ~ I~ I I~

~ ~ ~ ~ st ~

~ ~ t ~

\ ~

! I: t!: I ~ ~ ~

~

~ ~ I s

~ ~ ~ ~

~ ~

' liI

~ ~

~

~ ~

s ~

0.2 I~I

~ ~

~ ~ ~

~ ~

• I

~ ~ ss ~

0.0 C.4 A.2 0.0 0. 0.4 0.6 PERIPHERAL AXIALSHAPE INDE Rgure 3.2-2 AXIALSHAPE INDEX vs fraction of maximum allowable po er level per Specification 4.2.1.3 ST. LUCIE - UNIT 2 3/4 2-4

1.1

'PERATION" 1.0 REGION w 0.9 0.8 ~y

~ 1' ~ ~$ ~

0.7 CD 0.6 0.5 0.4

-0.6 -0.4 '-0.2 '.0.0 '.2 0.4 0.6 PERIPHERAL AXIAL SHAPE INDEX

.. FIGURE 3.2-2 AXIAL SHAPE INDEX VS FRACTION OF NXMN ALLOWABLE POWER LEVEL PER SPECIFICATION 4,2.1.3 ST. LUCIE - UNIT 2 3/4 2-4

1.2 UNACCEPTASLE OPERATION REBLION .'.".

-:::--i:.:- i-.:.:-t-w:::::t-: ~:-t':=::::-:

l1.60. 1

'1.0 F, LIMITCURVE, FLIMITCURVE 0  :,: l1.68, 0.80)::

I'.-

~

I I I 0.8

. ~

0 0.6

=j I :jt ACC EPTAB LE OPE RATION REGION z

0 4 ~ \

0.4 0.2 0.0 1.62 1.64 MEASURED F,, F Figure 3.2 3 1.66 1.68 1.70 l

1.72 Allowable comblnetlons of thermel power end F,, FRg

Fl 3.2-3 A(lMKKC@B 1 QT 1 NS OF TIEL PNER ND Fp I Fxy

~o ~ ~ oo )oo 5

~ << ~ ~ ~

~

/ ~ ~ ~ olo o

~ ~ ~ oo .E. 'oo. ~~

~ ~

~~

1.0

~

~ ~o

~o ~ ~~ ~o ~ ~ ~ t 5 ~o ~ ~ ~o 85, 8) 0.8 ~ ~

a.e 1.65 1.70 1.75 1.&0 ~ 1.85

%hSNED F, F

ee POWER DISTRIBUTION LIMITS 3/4.2.'2 TOTAL PLANAR RADIAL PEAKING FACTORS - F LIMITING CONDITION FOR'.OPERATION 3.2.2 The APPLICABILITY:

calculated value of MODE 1".

Fx shall be limited to ( ~

l .7z" ACTION:

With F > ~, within t.7>-

6 hours either:

ae Reduce THERMAL POWER to bring the combination of THERMAL POWER and Fx to within the limits of Figure, 3.2-3 and withdraw the full length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3. 1.3.6; or

b. Be in HOT STANDBY.

SURVEILLANCE RE UIREMENTS e

4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F T. xy shall be calculated by the expression F T = F (1+T ) when xy xy is calculated with a non-full core power distribution analysisq code and

~

F shall be calculated as' T = F when calculations are performed with a full xy xy core power distribution analysis code. F shall be determined to be within xy its limit at the following intervals:

a. Prior to operation above 70K of RATED THERMAL POWER after each fuel loading,

b. At. least once per 31 days of accumulated operation in MODE 1, and

c. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the AZIMUTHAL POWER TILT (T )

q is > '0.03.

See Special Test Exception 3. 10.2.

ST. LUCIE " UNIT 2 3/4 2-7

P POWER DISTRIBUTION LIMITS TOTAL INTEGRATED RADIAL PEAKING FACTOR - F LIMITING CONDITION FOR OPERATION 3.2.3 The calculated value of Fr, T shall be limited to < ~l,aa APPLICABILITY: MODE 1".

ACTION:

With FrT > ~,l. TO within 6 hours either:

a. Be in at least HOT'TANDBY, or

b. Reduce THERMAL POWER to bring the combination of THERMAL POWER and F

r to within the limits of Figure 3.2-3 and withdraw the full-length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3. 1.3.6. The THERMAL POWER limit determined from Figure 3.2-3 shall then be used to establish a revised upper THERMAL POWER level limit on Figure 3.2-4 (truncate Figure 3.2-4 at the .

allowable fractioh of RATED THERMAL POWER determined by Figure 3.2-3) and subsequent operation shall be maintained within the reduced acceptable operation region of Figure 3.2-4.

SURVEILLANCE RE UIREMENTS 4.2.3. 1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F r shall be calculated by the expression

~

F T =

F r (1+Tq ) ~~w h 1 1 1 1 Fh 11 11 analysis code and shall be calculated as F -"F when r

calculations are performed with a full core power distribution analysis code.

Fr shall be determined to be within its limit at the following intervals.

a. Prior to operation above 70K of RATED THERMAL POWER after each fuel loading.

b. At least once per 31 days of accumulated operation in MODE 1, and

c. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the AZIMUTHAL POWER TILT (T ) is > 0.03.

See Special Test Exception 3.10.2."

ST. LUCIE - UNIT 2 3/4 2"9

DNBR PENALTY PENALTY MULT TO BE UNDLE DNBR WITH GRID SPACING 'ED GWd/MTU TY PENALTY TO MEASURED F 0-10. 0 0.5 1.5 1. 013

10. 0-20. 0 .0 2.0

20. 0- 2. 0 3.0 1. 026

. -40.0 3.5 4.5 1. 038

40. 0" 50. 0 5.5 ~

6.5 1. 055 ST. LUCIE - UNIT 2 3/4 2-u

12 .

I I ~ \

I ~ ~ I I

1.0

~ ~

0.15, 1.00) .15, 1.00)

I ~

I UNACCEPTAB ~

UNACCEPTABLE I I OPERATION ~ ~ ~

r

~ ~

'PERATION REGION REGION r ~

r I 0a.

~

0.8 'I'I I I03 075)'-

.(-0.3. 0.75) I~ ~ r ' '

cC

~ ~ ~

ACC PTABLE xI OPERATION O +REGION 0.6 ~ I 0 r r ~ ~ ~ ~

R ~ ~

O

-

r r

rr ~ ~ ~ ~

r r

~ ~ ~

~ I ~ ~ ~ ~ ~ ~

0.2

~ e ~ ~

-:-- ~

)

r ~

I

~ r ~ ~ = ~ ~

5 0.0 A.2 0.0 0.2 0.4 0.6 PERIPHERAL AXIALSHAPE INDEX ( t)

Rgure 3.2Q AXIALSHAPE INDEX operatinII limits with four reactor coolant pumps operating ST. LUCIE - UNIT 2 3/4 2-12

1.1 iR DN 1.0 0.9 P- 0.8 ACCEPTABLE OPERATION h 0.7 0.6 0.5 P ~

0.4

-0.6 -0.4 -'0;2 .

0;.0 '.2 0.4 0.6 PERIPHERAL AXIAL SHAPE INDEX Yl FIGURE 3.2-4 AXIAL SHAPE INDEX OPERATING LIMITS MITH FOUR REACTOR COOLANT PUMPS OPERATING J

ST. LUCIE - UNIT 2 3(4 2-12

TABLE 3.2-2 DNB MARGIN LIMITS FOUR REACTOR COOLANT PUMPS PARAMETER OPERATING S91 Cold Leg Temperature (Narrow Range) 535 F* < T < &% F Pressure izer Pressure 2225 psia"" < P PZR-< 2350 psia" 9t'9,os Reactor Coolant Flow Rate > &79-~ gPm AXIAL SHAPE INDEX Figure 3.2-4 Applicable only if power level > 70K RATED THERMAL POWER.

Limit not applicable during either a'HERMAL POWER ramp increase in excess of 5X of RATED THERMAL POWER or a THERMAL POWER step increase of greater than lOX of RATED THERMAL POWER.

ST. LUCIE - UNIT 2 3/4 2-15

TABLE 3.3-2 REACTOR PROTECTIVE INSTRUHENTATION RESPONSE TIHES FUNCTIONAL UNIT RESPONSE TIHE

1. Hanual Reactor Trip Not Applicable

2. Variable Power Level - High < 0.40 second"'""

3. Pressurizer Pressure - High < 1 15 seconds

4. Thermal Hargin/Low Pressure < 0.90 second""

- l )S

5. Containment Pressure High < 4-.SS seconds

6. Steam Generator Pressure Low < 1.15 seconds

7. Steam Generator Pressure Oifference - High < 1.15 seconds

8. Steam Generator Level - Low < 1.15 seconds

9. Local Power Density - High < 0.40 second"'"*

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES r

ALLOWABLE FUNCTIONAL UNIT TRIP VALUES SETPOINT'ot

1. SAFETY INJECTION (SIAS)

a. Manual (Trip Buttons) Applicable Not Applicable

b. Containment Pressure - High < ~I.v psig <

'/.8'b 4-.+0 psig

c. Pressurizer Pressure - Low > 1736 psia > 1728 psia

d. Automatic Actuation Logic Not Applicable Not Applicable

2. CONTAINMENT SPRAY (CSAS)

a. Manual (Trip Buttons) Not Applicable Not Applicable High-High 9'Yo N. PO

b. Containment Pressure < +H& psig < 9-.40 psig

c. Automatic Actuation Logic Not Applicable Not Applicable

3. CONTAINMENT ISOLATION (CIAS)

a. Manual CIAS (Trip Buttons) Not Applicable Not Applicable

b. Safety In)ection (SIAS) Not Applicable Not Applicable c.

d.

Containment Pressure Containment Radiation

- High

- High psig < ~

yea psig

< 10 R/hr < 10 R/hr

e. Automatic Actuation Logic Not Applicable Not Applicable

4. MAIN STEAM LINE ISOLATION

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Steam Generator Pressure -. Low > 600 psia > 567 psia

c. Containment Pressure - High <

'I.7 5'sig .

< ~g.so psig

d. Automatic Actuation Logic Not Applicable Not Applicable

TABLE 3.3-5 Continued ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

2. Pressurizer Pressure-Low a ~ Safety Injection (ECCS) 3P 0%/20 Paw

b. Containment Isolation < 21.75"/11.75~"

C. Shield Building Ventilation System < 26.0"/10.0""

d. Containment Fan Coolers < 24.15*/11.15""

e. Charging Flow < 330.00"/180.00""

3. Containment Pressure-Hi h

a. Safety Injection (ECCS) 3P Pk/20 Pkk

b. Containment Isolation < 21.75"/11.75""

c. Shield Building Ventilation. System < 26.0"/10.0""

d. Containment Fan Coolers < 24. 15"/ll. 15""

e. Feedwater Isolation M*iAY**

f. Main Steam Isolation < 6. 75"/6. 75**

4. Containment Pressure--Hi h-Hi h

25. 65"/11. 15**

~ ~

a. ~ Containment Spray/Iodine Removal <

~

5. Containment Radiation-Hi h

a. Containment Isolati'on < 26. 75"/16 75""

~

b. Shield Building Ventilation System < 32.75"/16.75""

6. Steam Generator Pressure-Low 5 /S

a. Feedwater Isolation $ -.35"*

b. Main Steam Isolation < 6.75/6.75""

7. Refuelin Water Stora e Tank-Low

a. Containment Sump Recirculation < 111. 15 "/101. 15""

8. 4.16 kV Emer enc Bus Undervolta e Loss of Volta e

a. Loss of Power (4. 16 kV) < 14 b Loss of Power (480 V) < 14

9. 4. 16 kV Emer enc Bus Undervolta e De raded Volta e)

a. Loss of Power (4. 16 kV) < 12

b. Loss of Power (480 V) < 22 ST. LUCIE - UNIT 2 3/4 3"20

TABLE 3.3-5 Continued ENGINEERED SAFETY FEATURES RESPONSE TIMES e

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

10. Steam Generator Level-Low

a. Auxiliary Feedwater < 120*/120""

b. Feedwater Isolation ( M iDP*

11. Feedwater Header hP

a. Auxi 1 iary Feedwater 120%/120AA

b. Feedwater Isol ati on

12. Steam Generator hP

a. Auxi 1 i ary Feedwater < 120"/120""

b. Feedwater Isolation < 5.35"/5.35""

NOTE: Response time for Motor-Driven and Steam-Driven Auxiliary Feedwater Pumps on all AFAS signal starts < 120.0 TABLE NOTATION Diesel generator starting and sequence loading delays included. Response time limit includes movement of valves and attainment of pump or blower discharge pressure..

Diesel generator starting and sequence loading delays not included. Offsite power available. Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

C Containment Isolation response time is applicable to the valves specified in Specification 3.6.3.

ST. LUCIE - UNIT 2 3/4 3-21

REACTOR COOLANT SYSTEM 3/4.4.3

~ ~ PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a minimum water level of greater than or equal to 27K indicated level and a maximum water level of less than or equal to 85K indicated level and at least two groups of pressurizer heaters capable of being powered from 1E buses each having a nominal capacity of at least 150 kW.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one group of the above required pressurizer heaters inoperable, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURYEILLANCE RE UIREMENTS 4.4.3.1 The pressurizer water volume shall be determined 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 />.

4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified to be at least 150 kW at least once per 92 days.

4.4.3.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at- least once per 18 months by verifying that on an Engineered Safety Features Actuation test signal concurrent with a loss of offsite power:

a. the pressurizer heaters are automatically shed from the emergency power sources, and

b. the pressurizer heaters can be reconnected .to their respective buses manually from the control room.

ST. LUCIE " UNIT 2 3/4 4-9

.,'.3/4.7 PLANT SYSTEMS ~

-'/4. 7. 1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION "3.7.1.1 All main steam line code safety valves shall be OPERABLE.<ri+t~H4 e

.- APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE statu~ or the Power Level-High trip setpoint is reduced per Table 3.7-Q, otherwise, 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7;1.1 AC4lhnN 5~~~~~4 +PE ~~8 ~ E., ~z I 4 li A Sc l /hf 5 7

~ r ydY >pww s I Bra a5 54o~w ih /~4 /g / 7 g g'g ~dc'0'l hi'Fh

~

~sebi'ow 2D uC eke A5C E 8n)'(<~ a~A Pre>>

I C', iq~v e4,A <.

-

t'e

',.ST. LUCIE - UNIT 2 3/4 7-1.

TABLE 3.7-1 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBE LIFT SETTING +IX " Rk ED CAPACITY""

Line No. 1 Line No. 2

a. 8201 .8205 1000 psia 744,210 lb/hr

b. 8202 8206 1000 psia 744,210 lb/hr

c. 8203 8207 1000 psia 744,210 lb/hr

d. 8204 8208 1000 psia 744,210 lb/hr

e. 8209 8213 1040 psia 774,000 lb/hr 8210 8214 y 1040 psia 774,000 lb/hr

g. 8211 8215 '4040 psia 774,000 lb/hr

h. 8212 8216 1040 psia 774,000 lb/hr I

The of the lift setting pressure shall cor'respond to ambient conditions valve at nominal operating temperature and pressure.

Capacity is rated at

/

lift setting +3K accumulation.

ST. LUCIE - UNIT 2 3/4 7-2

TNLK 3.7-1 NXINN ALLOMNLK POWER LEVEL-IllGH TRIP SETPOIN NITH IHOPERNLE ST ET E R NG OPERATI I BOT I STE GEH TORS Haxfeum Allocable Pmer Nax)~ Nuiiber of Inoperable Safety Level-Iligh Trip Setpolnt Valves on An 0 erat3n Steaa Generator Percent of RATEO TIIERNL POMER 93.2 79.8 66.5

TABLE 3.7-2 MAXIMUM A OWABLE LINEAR POWER LEVEL-HIGH TRIP SETP STEAM LINE AFETY VALVES DURING OPERATION WITH B r

T WITH INOPERABLE STEAM GENERATORS Maximum Allowable Linear Power Maximum Number of In Safety .Level-High Trip Setpoint Valves on An 0 eratin terable Steam Generator Percent of RATED THERMAL POWER 107.0 96.0

82. 0

68. 0

55. 0 ST. LUCIE - UNIT 2 3/4 7-3

TABLE l.7-0 STEN LINE SAFETY VALVES PER lOOP I

g VALVE NN8ER l.lFT SETTING 4 IX ORIFICE SIZE I

6 Header h Beadar 8

a. 8201 8205 1000 psia 16 in.

b. 8202 8206 1000 psia 16 in.

c. 8203 8207 1000 psia 16 in.

N 820l 8208 1000 psia 16 in.

e. 8209 8213 1040 psia 16 in.

8210 8214 1040 psia 16 in.

g. 8211 8215 1040 psia 16 in.

h. 8212 8216 1040 psia 16 in.

PLANT SYSTEMS MAIN FEEDWATER LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION

3. 7. l. 6 Each main feedwater line isolation valve shall be OPERABLE.

APPLICABILITY: MODE% 1, 2, 3, and 4.

ACTION:

MODE 1 With one main feedwater line isolation valve inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, 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 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

MODES 2, 3- With one main feedwater line isolation valve inoperable, and 4 subseqent operation in MODE 2, 3, or 4 may proceed provided:

a. The isolation valve is maintained closed.

b. The provisions of Specification 3.0.4 are not applicable.

Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.7.1.6 Each main feedwater line isolation valve shall be demonstrated OPERABLE by:

'a 0 Part-stroke exercising the valve at least once per 92 days, and

b. full closure within ~

5; J$

seconds on any closure actuation

~

'erifying signal while in HOT STANDBY with T > 515'F during each reactor shutdown except that verification 9 full closure within seconds need not be determined more often than once per 92 days.

ST. LUCIE - UNIT 2 3/4 7-10

3/4. 1 REACTIVITY CONTROL SYSTEMS BASES 3/4. 1 . 1 BORAT ION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T . The most restrictive occurs at EOL, with T at no load operating temperature, and .is avg'ondition avg associated with a postulated steam line break accident and resulting uncon-trolled RCS coo'adown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 5.0X delta k/k is required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. At earlier times in core life, the minimum SHUTDOWN MARGIN required for the most restric" tive conditions is less than 5. OX hk/k. With Tavg less than or equal to 2004F, the reactivity transients resulting from any postulated accident are minimal and a g4 delta k/k SHUTDOWN MARGIN provides adequate protection.

3/4. 1. 1. 3 BORON DILUTION A minimum flow rate of at least 3000 gpm provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 gpm will circulate an equivalent Reactor Coolant System volume of 10,931 cubic feet in approximately 26 minutes. The reactivity change rate associated with boron concentration reductions will therefore be within the capability of operator recognition and control.

3/4. 1. 1.4 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the accident and transient analysis remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since this coe'fficient changes s'lowly due principally to the reduction in RCS boron concentration associated with fuel burnup. The confirmation that the measured MTC value is within its limit provides assurances that the coef-ficient will be maintained within acceptable values throughout each fuel cycle.

ST. LUCIE - UNIT 2 B 3/4 1-1

REACTIVITY CONTROL SYSTEMS BASES 3/4. 1. 1. 5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 515 F. This limitation is required to ensure (1) the moderator temperature coefficient is

.within its analyzed temperature range, (2) the protective instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor pressure vessel is above its minimum RTNDT temperature.

3/4. 1. 2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid makeup pumps, (5) associated heat tracing systems, and (6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 2004F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed fai lure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

. The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions o~.OX delta k/k after xenon decay and cooldown to 200'F. The maximum expected boration capability requirement occurs at EOL from full power equi librium xenon conditions and requires boric acid solution from the boric acid makeup tanks in the allowable concentrations and volumes of Specification 3. 1. 2. 8 or 72,000 gallons of 1720 ppm - 2100 ppm borated water from the refueling water tank.

1ttith the RCS temperature below 200'F one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.

The boron capability required below 200 F is based upon providing a %o delta k/k SHUTDOWN MARGIN after xenon decay and cooldown from 200'F to 140'F.

This condition requires either 4,150 gallons of 1720 ppm - 2100 ppm borated water from the refueling water tank or boric acid solution from the boric acid makeup tanks in accordance with the requirements of Specification 3. 1. 2. 7.

The contained water volume limits includes allowance for water not available because of discharge line location and other physical characteristics.

The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

The '1imits on contained water volume and boron concentration of the RWT also ensure a pH value of between 7.0 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

ST. LUCIE - UNIT 2 B 3/4 1-2

REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued Overpower margin is provided to protect the core in the event of a large misalignment () 15 inches) of a CEA. However, this misalignment would cause distortion of the core power distribution. This distribution may, in tur n,

. have a significant effect on (1) the available SHUTDOWN MARGIN, (2) the time-dependent long-term power distributions relative to those used in generating LCOs and LSSS setpoints, and (3) the ejected CEA worth, used in the safety analysis. Therefore, the ACTION statement associated with the large misalignment of a CEA requires a prompt realignment of the misaligned CEA.

The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA. Conformance with these alignment requirements bring the core, within a short period of time, to a configuration consistent with that assumed in generating LCO and LSSS setpoints. However, extended operation with'CEAs significantly inserted in the core may lead to per turhations in (1) local burnup, (2) peaking factors, and (3) available shutdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LCO ~

and LSSS setpoints determination. Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.

The requirement to reduce power in certain time limits depending upon the previous F is to eliminate a potential nonconservatism for situations when a CEA has be5n declared inoperable. A worst-case analysis has shown that a DNBR SAFDL violation may occur during th'e second hour after the CEA misalignment if this requirement is not met.

'

P.lt neilet'we. This potential DNBQSAFD violation is eliminated by~~I'wta HB aaghhl ~ aplpa4!n Pove r once e M power reductions ."'these reductions will be nece5da'Fy eviated CEA has been declared inoperable. This time allowed to bg Pure continued operation at a reduced power level can be permitted for the following reasons:

1. The margin calculations which support the Technical Specifications are based on a steady-state radial peak of F

) ao

2. When the actual F = ~,

l 1v

~

significant additional margin exists.

T

3. This additional margin can be credited to offset the increase in Fr with time that can occur following a CEA misalignment.

4. . This increase in F r is

. caused by xenon redistribution.

5. The present analysis can support allowing a misalignment to exist for up to Pf minutes without correction, if the initial Fr < 1.5P'.

(a h ST. LUCIE - UNIT 2 B 3/4 1-4

POWER DISTRIBUTION LIMITS BASES assumptions used in establishing the Linear Heat Rate, Thermal Margin/Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid.

An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

The requirement that the measured value of Tq be mutiplied by the calculated values of F r and 'xy to determine F r and F is applicable on y F,

when F and F are calculated with a non-full core power distribution analysis Xy P code. When monitoring a reactor core power distribution, F or Fx with a full xy core power distribution analysis code the azimuthal tilt accounted for as part of the radial power distribution used to calculate is explicitly Fx and Fr.

The Surveillance Requirements for verifying that T F T and T are F, within their limits provide. assurance that the actual values of Fx, xy' Fr and T do not exceed the assumed values, Verifying F and F after each fueel loading prior to exceeding 75K of RATED THERMAL POWER provides additional assurance that the core was proper ly loaded.

3/4. 2. 5 DNB PARAMETERS The limits on, the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and safety analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain throughout each analyzed transient.

4 N OCr e p 4C. 4 I C ~ i~ i ~~

ice The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following }oad changes and other expected transient operation. The 18-month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12-hour basis.

ST. LUCIE - UNIT 2 B 3/4 2-2

TABLE B 3/4.2-1 PE LTY TO BE APPLIED FTO ACCOUNT R ROD BOW E FECTS ON DNBR A DNBR PENALTY PENALTY MULTIPLIER TO BE BURNUP OF BUNDLE DNBR ITH GRID SPACING GMd/MTU) PENALTY, 4) APPI IED TO MEASURED F 0-10. 0 0~5 l. l. 013

10. 0-20. 0 ,i"l. 0 2,.0 1. 017

20. 0-30. 0 2.0 3.0 1. 026

30. 0" 40. 0 3.5 4.5 1. 038 40.0-50.0 5.5 6.5 l. 055 ST. LUCIE - UNIT 2 B 3/4 2-3

%4.t$ ~~te G PCQ C. tQ 'AC f lMCC.Q, t.vve. Poll ~iwg page 3/ .7 PLANT SYSTEMS BASES 3/4.7.1 TUR INE CYCLE 3/4. 7.'l. 1 SAF Y VALVES iozg The OPERABIL Y of the ain steam line code safety valve ensures that its design pressure psi9 during the most severe anticipated system operational transient. The maximum relieving capacity is ssociated with a turbine trip from 100K +TED THERMAL POWER coincident wi an assumed loss of condenser heat sink (i.e> no steam bypass to the conde ser).

The

/

specified valve li t settings and relieving capacities are in accordance with the requirements of Section III of tPe ASIiE Boiler and Pressure Vessel Code, 1971 Edition. Thk total relieving capacity for all valves on all of the steam lines is 12,384,000>lbs/hr which is l10.0X of the total secondary steam flow of 11,172,000 lbs/hr at lOOX RATED THERMAL POWER. A minimum of one OPERABLE safety valve per steam generator ensures that sufficient relieving capacity is available for removing decay heat.,

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION regqirements on the basis of the reduction in secondary system steam flow and THERMA/ POWER required by the reduced reactor trip settings of the Power Level-Hi' channels. The reactor trip setpoint reductions are derived on the folio ing bases:

For two loop op ."ation SP = ( 8

) x 110.0 where: -I SP reduced reactor trip setpoint in pe ent of RATED THERMAL POWER. This is a ratio of the avai la le relieving capacity over the total steam flow at rated pow r.

total number of secondary safety valves or one steam generator.

/

The number of inoperable secondary safety v ives on the.

steam generator with the greater number of i operable valves.

110. 0 the ratio of the total relieving capacity of al sixteen (16) secondary safety valves (12,384,000 lbs/hr at 1071

/psia, maximum set pressure plus 3X accumulation) over the secondary steam flow at 100M Rated Thermal Load (11,172,000 lbs/hr).

ST. LUCIE " UNIT 2 B 3/4 7-1

/4.7 PLANT SYSTEMS BASES 3/4.7. 1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within its design pressure of 1025 psig during the most severe anticipated system opera-tional transient. The maximum relieving capacity is associated with a turbine trip from 100$ RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of. Section III of the ASME Boiler and Pressure Code, 1971 Edition and ASME Code for Pumps and Valves, Class II.

The total relieving capacity ior all valves on all of the steam lines is 12.38 x 10 lbs/hr which is 102.8 percent the total secondary steam flow of 12.04 x 10 lbs/hr at 100~> RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for removing decay heat.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reacto'r trip settings of the Power Level-High channels. The reactor trip setpoint reductions are derived on the following bases:

For two loop operation:

S (X) - (Y)(V)

( 06. )

X where:

SP reduced reactor trip setpoint in percent of RATED THECAL POWER maximum number of'.inoperable safety valves per steam line 106. 5 Power Level-High Trip Setpoi,nt for two loop operation X e Total relteviog capacity of all aa[ety vatvea per steam line in lbs/hour (6.192 x 10 lbs/hr.)

Maximum relieving capaiity of any one safety valve, in lbs/hour (7.74 x 10 lbs/hr.)

ST. LUCIE - UNIT 2 B3/4 7-1

5.0 DESIGN FEATURES

5. 1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1.

LOW POPULATION ZONE

5. l. 2 The low population zone shall be as shown in Figure 5. 1-1.

5. 2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel building of cylindrical shape, with a dome roof and having the following design features:

a. Hominal inside diameter = 140 feet.

b. Nominal inside height = 232 feet.

c. Net free volume = ~z.sa x

C 10'ubic feet.

d. Nominal thickness of vessel walls = 2 inches.

e. Nominal thickness of vessel dome = 1 inch.

f. Hominal thickness of vessel bottom = 2 inches.

5. 2. 1. 2 SHIELD BUILDING

a. Min'imum annular space = 4 feet.

b. Annulus nominal volume = 543,000 cubic feet.

c. Nominal 'outside height (measured from top of foundation mat to the top of the dome) = 228.5 feet.

d. Nominal inside diameter = 148 feet.

e. Cylinder wall. minimum thickness = 3 feet.

f. Dome minimum thickness = 2.5 feet.

g. Dome inside radius = 112 feet.

DESIGN PRESSURE AND TEMPERATURE

5. 2. 2 The steel reactor containment building is designed and shall be maintained for a maximum internal pressure of 44 psig and a temperature of 264 F.

ST. LUCIE " UNIT 2 5-1

5.3 REACTOR CORE worn o.nd poison r od 'loca~~<<~. All 4~i

~

~ ~

co 4 e 4 t. C.

~

FUEL ASSEMBLIES ot assembly containing 236 fuel ~

5.3. 1 The reactor core shall contain 217 fuel assemblies with each fuel clad with Zircaloy-4. Each fu 1 rod s a have a nominal active fuel length of 136.7 inches and contain 1

l enrichment of 2.73 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.70 weight percent U-235.

CONTROL ELEMENT ASSEMBLIES gl 5.3.2 The reactor core shall contain M full-length control element assemblies and no part-length control element assemblies.

5.4 REACTOR COOLANT SYSTEM 5.4. 1 The Reactor Coolant System is designed and shall be maintai'ned:

'a ~ In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and C. For a temperature of 650~F, except for the pressurizer which is 700 F.

ST. LUCIE - UNIT 2 5" 3