ML17212A423

Proposed Tech Spec Pages B2-5,3/4 3-4,B3/4 1-1,3/4 1-1,3/4 1-2,3/4 4-1,5-4 & Tables 2.2-1 (Page 2-4),3.3-4 (Page 3/4 3-14),2.2-1 (Page 2-5) 3.3.3 (Page 3/4 3-12),7.3.2-1 7.3.2-2,7.3.2-3 & 7.2.2-4 Re Reactor Coolant Pump Trip

ML17212A423
Person / Time
Site:	Saint Lucie
Issue date:	07/23/1981
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17212A422	List: ML17212A420 ML17212A423
References
NUDOCS 8107280409
Download: ML17212A423 (37)

Text

0 0Q)

~s ) TABLE 2.2-1

>lVj UN REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT- LIMITS no'w,'

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES QIN O~,'O 1. Manual Reactor Trip Not Applicable Not Applicable OM OQIQ 2. Power Level - High (1) au+

Four Reactor Coolant Pumps Operating < 9.61$ above THERMAL POWER, < 9.61'X above THERMAL POWER, and with a minimum setpoint of 15$ a minimum setpoint of 15$ of RATED of RATED THERMAL POWER, and a THERMAL POWER and a maximum of maximum of < 107.0g, of RATED < 107.0$ of RATEO THERMAL POWER.

THERMAL POWER.

3. Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps Operating > 95'f design reactor coolant > 951, of design reactor coolant f'low 7low with 4 pumps operating* with 4 pumps operating*

4. Pressurizer Pressure - High < 2400 psia < 2400 psia

5. Containment Pressure - High < 3.3 psig < 3.3 psig

6. Steam Generator Pressure - Low (2) > 600 psia > 600 ps>a

7. Steam Generator Water Level -Low > 37.0$ Water Level - each steam

, > 37.0$ Water Level - each generator steam generator

8. Local Power Density - High (3) Trip setpoint adjusted to not Trip setpoint adjusted to not exceed the limit lines of exceed the limit lines of Figures Figures 2.2-1 and 2.2-2 2.2-1 and 2.2-2.

• Design reactor coolant flow with 4 pumps operating is 370,000 gpm.

I PO GO I

00

) ~ ~

TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. SAFETY INJECTION (SIAS)

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Containment Pressure - High < 5 psig 5 psig

c. Pressurizer Pressure - Low > 1600 psia > 1600 psia

2. CONTAINMENT SPRAY (CSAS)

a. Manual (Trip Buttons)

Not Applicable Not Appl i cable

b. Containment Pressure High-High < 10 psig < 10 psig

3. CONTAI NMEHT ISOLATION (CIS)

a. Manual (Trip Buttons) Hot Applicable Hot Applicable

b. Containment Pressure High < 5 psig < 5 psig

c. Containment Radiation - High < 10 R/hr < 10 R/hr

d. SIAS (SEE FUNCTIONAL UNIT 1 ABOVE)-

4. MAIN STEAM LINE ISOLATION (MSIS)

a. Manual (Trip Buttons) Hot Applicable Not Applicable

b. Steam Generator Pressure - Low > 585 psig > 585 psig

5. CONTAINMENT SUMP RECIRCULATION (RAS)

a. 'anual RAS (Trip Buttons) Not Applicable Not Applicable

b. Refueling Water Tank - Low 48 inches above tank bottom 48 inches above tank bottom

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR COOLANT FLOW-LOW Continued reactor coolant pumps are taken out of service. The low-flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above 1.30 under normal operation and expected transients. For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip setpoints, the Power Level-High trip setpoints, and the Thermal Margin/Low Presure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two- or three-pump position. Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below 1.30 during normal operational transients and anticipated transients when only two or three reactor coolant pumps are operating.

Pressurizer Pressure-Hi h The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides . reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. Th'.s trip's setpo'.nt is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.

Containment Pressure-High The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. 'The setting of 600 psia is sufficiently below the full-load operating point of 800 psig so as not ST. LUCIE - Unit 1 B2-5 7-23-81

TABLE 2.2-1 Continued REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

9. Thermal Margin/Low Pressure (1)

Four Reactor Coolant Pumps Operating Trip setpoint adjusted to not Trip setpoint adjusted to not 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.

10. Loss of Turbine - Hydraulic > 800 psig > 800 psig Fluid Pressure - Low (3) ll. Rate of Change of Power - High (4) < 2.49 decades per minute < 2.49 decades per minute TABLE NOTATION (1) Trip may be bypassed below 1$ of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > lg of RATED THERMAL POWER (2) Trip may be manually bypassed below 685 psig; bypass shall be automtically removed at or above 685 psig.

(3) Trip may be bypassed below 15$ of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15$ of RATED THERMAL POWER.

(4) Trip may be bypassed below 10 5 and above 15$ of RATL'D THERMAL POWER.

TABLE 3.3-3 Continued TABLE NOTATION (a) Trip function may be bypassed in this MODE when pressurizer pressure is < 1725 psia; bypass shall be automtically removed when pressurizer pressure is > 1725 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be bypassed in this MODE below 685 psig; bypass shall be automatically removed at or above 685 psig.

The provisions of Specification 3.0.4 are not applicable.

ACTION STATEMENTS ACTION 8 With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-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 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 />.

ACTION 9 With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisified:

a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of intial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped, condition.

b. Within one hour, all functional units receiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped, as applicable) as that required by a. above for the inoperable channel.

c. The Minimum Channels OPERABLE requirement is met; however, one additonal channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition.

ST LUCI E - UNIT 1 3/4 3-12 7-23-81

0 TABLE 3.2.-1 Continued TABLE NOTATION

• With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

g The provisions of Specification 3.0.4 are not applicable.

(a) Trip may be bypassed below 1$, of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > lg,. of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 685 psig; bypass shall be automatically removed at or above 685 psig.

(c) Trip may be bypassed below 155 of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15$ of RATED THERMAL POWER.

(d) Trip may be bypassed below 10 5 and above 15$ of RATED THERMAL POWER; bypass shall be automtically removed when THERMAL POWER is > 10 "5 or <

15$ of RATED THERMAL POWER.

(e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

(f) There shall be at least two decades of overlap between the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Monitoring Channels.

ACTION STATEMENTS ACTION 1 With number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requi rement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in 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/or open the protective system trip breakers.

ACTION 2 With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may poceed provided the following conditions are satisfied:

a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY; however, the inoperable channel shal 1 then be either restored to OPERABLE status or placed in the tripped condition.

ST. LUCIE - UNIT 1 3/4 3-4 7-23-81

0 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION 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 subcri ti cal from all operating conditions, 2) the reactivity transi ents associ ated wi th postul ated accident condi tions are control lab l e wi thin 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 Tav ~

The most restrictive condition occurs at EOL, with Tav at no load operating temperature, and is associ ated wi th a postul ated steam line break accident and resul ting uncontrol 1 ed RCS cool down. In the analysi s of thi s accident, a minimum SHUTDOWN MARGIN of 5. Og, d.k/k i s requi red to control the react i vi ty I transient. Accordingly, the SHUTDOWN MARGIN required by Specification 3.1.1.1 is based upon this limiting condition and is consistent with FSAR accident analysis assumptions. For earlier periods during the fuel cycle, this 'value is conservative. With Tav < 200'F, the reactivity transients resulting from any postulated accident aPe minimal and a 1$ dk/k shutdown margin provides adequate protection.

3/4.1.1.3 BORON DILUTION AND ADDITION A minimum fl ow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration changes in the Reactor Coolant System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of '11,400 cubic feet in approximately 26 mi nutes. The reactivi ty change rate assoicated with boron concentration changes will be within the capability for operator recognition and control.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT MTC The limiting values assumed for the MTC used in the accident and transient analyses were + 0.5 x 10 4h k/k/'F for THERMAL POWER levels < 7% of RATED THERMAL POWER, + 0.2 x 10 4 hk/k/'F for THERMAL POWER levels >70%, of RATED THERMAL and -2.2 x 10 4 hk/k/'F at RATED THERMAL POWER. Therefore, these limiting values are included in this specification. Determination of MTC at the specified conditions ensures that the maximum positive and/or negative values of the MTC will not exceed the limiting values.

ST. LUCIE - UNIT 1 B 3/4 1-1 7-23-81

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARG I N Tav > 200 F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shal 1 be > 5.0$ bk/k.

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

ACTION:

Wi th the SHUTDOWN MARGIN < 5. l@ dk/k, immediately initiate and continue boration at > 40 gpm of 1720 ppm boron or equivalent until the required SHUTDOWN MARGTN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be > 5.0$ ~k/k:

a. Within one hour 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., When in MODES 1 or 2g, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that CEA group withdrawal is within the Power Dependent Insertion Limits of Specification 3.2.3.6.

c When in MODE 2gg, at least once during CEA withdrawal and at least once per hour thereafter until the reactor is critical.

d. Prior to initial operation above 5$ RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the CEA groups at the Power Dependent Inserti on Limits of Speci fication 3.1.3.6.

See Special Test Exception 3.10.1.

With Keff > 1.0 With Keff < 1.0.

ST LUCIE - UNIT 1 3/4 1-1 7-23-81

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued e.. When in MODES 3 or 4, 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 consideration of the following factors:

1. Reactor coolant system boron concentration,

2. CEA position "

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1.0$ < k/k at least once per 31 Effective Full .Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.l.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

• For Modes 3 and 4, during calcuation of shutdown margin with all CEA's verified fully inserted, the single CEA with the highest reactivity worth need not be assumed to be stuck in the fully withdrawn position.

ST. LUCIE - UNIT 1 3/4 1-2 7-23-81

3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.4.1 Four reactor coolant pumps shall be in operation.

APPLICABILITY: As noted below, but excluding MODE 6.

ACTION:

MODES 1 and 2:

With less than four reactor coolant pumps in operation, be ip 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 />.

MODES 3, 4 and 5:

Operation may proceed provided (1) at least one reactor coolant loop is in operation with an associated reactor coolant pump or shutdown cooling pump and (2) the SHUTOOMN MARGIN requirmnent of Specification 3.1.1.1 is increased to and maintained at > 5.0S ak/k during operation in NOOE 3 when less than four reactor coolant pumps are in operation. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENT 4.4.1 The Flow Dependent Selector Switch shall be determined to be in the 4 pump position within 15 minutes prior to making the reactor critical 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 Al 1 reactor cool ant pumps and shutdown cool i ng pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, provided no operations are permitted which coul d cause dilution of the reactor coolant system boron concentration.

- UNIT 7-23-81 ST. LUCIE 1 3/4 4-1

DESIGN FEATURES 5.2.1. 2 SHIELD BUILDING

a. Minimum annular space = 4 feet.

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

c. Nominal outside height (measured from top of foundation base to the top of the dome) = 230.5 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 containment vessel i s designed and shal 1 be maintained for a maximum internal pressure of 44 psig and a temperature of 264'F.

PE NETRATIONS 5.2.3 Penetrations through the containment structure are designed and shall be maintained in accordance with the original design provisions contained in Sections 3.8.2.1. 10 and 6.2.4 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 176 fuel rods clad with Zi rcol oy-4. Each fuel rod shall have a naoinal active fuel length of 136.7 inches and contain a maximum total weight of 2250 grams uranium. The initial core loading shall have a maximum enrichment of 2.83 weight percent U-235. Reload fuel shall be similar- in physical design to the initial core loading.

5.3.2 Except, for special test as authorized by the NRC, all fuel assemblies

.nder control element assemblies shall be sleeved with a sleeve design previously approved by the NRC.

ST. LUCI E - UNIT 1 5-4 7-23-81

Two-Loop-2754 MHt The Two Loop-2754 MWt case was initiated at the conditions listed in Table 7.3.2-1. The Moderator Temperature Coefficient (MTC) of reactivity assumed in- the analysis cor responds to end of life, since this negative MTC results in the greatest positive reactivity addition during the RCS cooldown caused by the Steam Line Rupture. Since the reactivity change associated with moderator feedback varies significantly over the moderator density covered in the analysis, a curve of reactivity insertion versus density rather than a single value of MTC is assumed in the analysis. The moderator cooldown curve given in Figure 7.3.2-1 was conservatively calculated assuming that on reactor scram, the highest worth Control Element Assembly is stuck in the fully withdrawn position.

The reactivity defect associated with fuel temperature decrease is also based on end of life Doppler defect. The Doppler defect based on an end of life Fuel Temperature Coefficient (FTC), in conjunction with the decreasing fuel temperatures, causes the greatest positive reactivity insertion during the Steam Line Rupture event. The uncertainty on the FTC assumed in the analysis is given in Table 7.3.2-1. The 8 fraction assumed is the maximum absolute value including uncertainties for end of life conditions. This too is conservative since it maximizes subcritical multiplication and thus, enhances the potential for Return-to-Power (R-T-P).

The minimum CEA worth assumed to be available for shutdown at the time of reactor trip at the maximum allowed power level is 6.74Xap. This available scram worth was calculated for the stuck rod which produced the moderator cooldown curve in Figure 7.3.2-1.

The analysis conservatively assumed that, on Safety Injection Actuation Signal, one Hiqh Pressure Safety Injection pump fails to sta'rt.

The analysis also assumed a cons'ervatively low value of boron reactivity worth of -1.0Ã hp per 105 PPM.

The conservative assumptions on fee'dwater flow were discussed previously.

The feedwater flow and enthalpy as a function of time are presented in Figures 7.3.2-2 and 7.3.2-3 respectively.

Table 7.3.2-2 presents the sequence of events for the full power case initiated at the conditions given in Table 7.3.2-1. The reactivity in-sertion as a function of time is presented in Figure 7.3.2-4. The response of the NSSS during this event is given in Figures <,3.2-5 through 7.3.2-9.

The results of the analyses show that SIAS is actuated at 15 seconds, at which time the Reactor Coolant Pumps are manually tripped by the operator. The manual trip of RCP's slows down the rate of primary heat removal and thus delays the time when the affected steam generator blows dry. The affected steam generator blows dry at 154 seconds and termi-nates the cooldown of the RCS. The peak reactivity attained prior to delivery of auxiliary feedwater flow is -.205Ãap at 169 seconds. A peak post trip power of l4.0~,consisting of 12.2'X decay heat and 1.8'X fission power, is produced at 72.8 seconds.. The continued production of decay heat from the fuel after termination of blowdown, causes the reactor coolant temperatures to increase. This in turn reduces the magnitude of the positive moderator reactivity inserted and thus the total reactivity becomes more negative.

7-23-81

The delivery of auxiliary feedwater flow to the affected steam generator at 180 seconds initiates a further cooldown of the RCS which results in more positive reactivity insertion. However, continued addition of negative reactivity by the boron injected via the HPSI pump prevents the positive reactivity insertion of the moderator from causing a return to . criticality.'he core remains subcritical throughout

'he remainder of the transient.

The Steam Line Rupture event from HFP conditions with automatic initiation of auxiliary feedwater and manual trip of RCP's on SIAS shows that the core does not produce significant instantaneous fission power. Since there is no significant Return-to-Power, heat fluxes wi-ll not he exceeded.

it can be concluded that critical Two Loo -No Load The two Loop-No Load case was initiated at the conditions qiven in Table 7.3.2-3. The moderator cooldown curve is given in Fiqure 7.3.2-10.

The cooldown curve corresponds to an end of life NTC. An end of life FTC was also used for the reasons previously discussed in connection with the Two Loop-2754 MWt case.

The minimum CEA shutdown worth available is conservatively assumed to be the minimum required Technical Specification limit of 5.0 hp. A maximum inverse boron worth of 80 PPH/Rap was conservatively assumed for the safety injection during the no load case. The feedwater flow and the enthalpy used in the analysis are presented in Figures 7.3.2-11 and 7.3.2-12.respectively.

>Ihen all CEA's are verified to be fully inserted (during modes 3 and 4) i is not necessary to assure one (the one of highest reactivity worth) is stuck in a fully withdrawn position when calculating shutdown margin.

Table 7.3.2-4 presents the sequence of events for the Two Loop-No Load case initiated from the conditions given in Table 7.3.2-3. The reactivity insertion as a function of time is presented in Figure 7.3.2-13. The response- of key NSSS parameters during -this event are given in Figures 7.3.2-'l4

'hrough 7.3.2-18.

The results of the analysis show that SIAS is actuated at 12.0 seconds, at which time the RCP's are manually tripped by the operator. The affected steam generator blows dry at 180.0 seconds. Auxiliary feed-water flow is initiated at 180.1 seconds which continues the cooldown of the RCS. Thus, the total core reactivity is critical'for a shor t period of time. The peak reactivity attained is only +.140K ap at 401.0 seconds. The addition of boron via High Pressure Safety Injection mitiqates the reactivity transient. The core power level remains less than 1% of rated power at all times during the event. Since there is no significant R-T-P for the Two Loop-No Load case, it can be concluded that the critical heat flux will not be exceeded during this event.

TABLE'7;3.2-1 KEY PARAMETERS ASSUMED IN THE STEAM LINE RUPTURE ANALYSIS 2-LOOP-2754 MWT Reference Parameter ts C cIe* ~Ccie 5 Initial Core Power 2754 2754 Initial Core Inlet oF 551 551 Temperature Initial RCS Pressure psla 2300 2300 Initial Core Mass x10 ibm/hr '33.8 133.8 Flow Rate Initial Steam Generator ps1a '909 900 Pressure Minimum CEA Worth W.25 -6.74 Available at Trip

~

. . Doppler. Multiplier 1.15 1.15 Moderator Cooldown Cap vs. density See Figure See Figur e Curve 7.3.2-1 7.3.2-1 Inverse Boron PPM/Xnp 105 105 Worth uncertainties)'ni Effective Full Power MTC x10 ap/ F -25 <<2e2 g Fraction (including, .0060 .0060

• (Note: The last complete blowdown. analysis for this event was performed for the Cycle 4 stretch power submittal.)

TABL'E 7.3;2-2 SEQUENCE OF EVENTS FOR STEAM LINE RUPTURE EVENT WITH AUTOMATIC INITIATION OF AUXILIARY FEEDWATER AND MANUAL TRIP OF REACTOR COOLANT PUMP 2-LOOP-2754 MWT Time sec) Event Set oint or Value 0.0 Steam Line Rupture Occurs 2.3 Low Steam Generator Pressure Trip 578 psia Signal Occurs 2.3 Main Steam Isolation Signal generated 578 psia

.Qo2 Main Steam Isolation Valves Begin to Close 3.2 ,Trip Breakers Open 3.7 CEAs Begin to Drop into Core

'9.2 Main Steam Isolation Valves Completely Closed 14.6 Pressurizer Empties:

15.0 Safety Injection Actuation Signal 1578.0 psia 15.0 Reactor Coolant Pumps Manually Tripped 45.0 High Pressure Safety Injection Pumps Start 63.3 Hain Feedwater Isolation 72.8 Peak Post. Trip Power 14.05 Affected Steam Generator Blows 154.5'69.0 Dry Peak Reactivity, Prior to -.205 'ap Auxiliary Feedwater Flow 180.0 Auxiliary Feedwater Flow Initiated 253.6 ibm/sec to Ruptured Steam Generator 600.0 Operator Isolates Ruptured Steam Generator and Terminates Auxiliary Feedwater Flow 7-23-81

Is ss

'TABLE'7;3;2-3 KEY PARAMETERS ASSUMED IN THE STEAM LINE RUPTURE ANALYSIS 2-LOOP, NO LOAD Reference Parameter Units ~Ccl e* ~Cele 5 Initial Core Power MWt 1.0 1.0 Initial Core Inlet 'oF 532 532 Temperature Initial RCS Pressure psia 2300 2300 Initial

'00 Steam Generator psia 900 Pressure Minimum CEA Worth -4.3 -5.0 Availab1 e at Trip Doppler Multiplier 1.15 0;85 Moderator Cooldown Cap vs. density See Figure See Figure Curve 7.3.2-10 7.3.2-10 Inverse Boron 1<orth PPM/lap 100 80 Effective Full Power MTC . 10 ap/ F -28 2%2 g Fracti.on (including '0060

.0060 uncertainty)

.*Cycle 4 stretch power submittal 7-23-81

TABLE 7;3.2-4 SEQUENCE OF EVENTS FOR STEAM LiNE RUPTURE EVENT WITH AUTOMATIC INITIATION OF AUXILIARY FEEDWATER AND MANUAL TRIP OF REACTOR COOLANT PUMPS 2-LOOP, NO LOAD Time sec Event Set oint or Value 0.0 Steam Line Rupture Occurs

'I 3.8. Low Steam Generator Pressure Trip Signal 578.0 psia Occurs 3.8 Main Steam Isolation, Signal Generated 578.0 psia 4.7 Main Steam Isolation Valves Begin to Close 4.7 Trip Breakers Open 5.2 CEAs Begin to Drop Into Core 10.2 Pressurizer Empties 10.7 Main Steam Isolation Valves Completely Closed Safety. Injection Actuation Signal 1578.0 psia 12.0'2.0 Reactor Coolant Pumps Manually Tripped 42.0 High Pressure Safety Injection Pumps Start 86.0 Hain Feedwater Isolation 180.0 Affected Steam Generator Blows Dry 180.1 Auxiliary Feedwater Flow Initiated to 253.6 ibm/sec Ruptured Steam Generator 401.0 Peak Reactivity +e140Xap 520.0 Return to Power c15 of rated power 600.0 Operator Isolates Ruptured Steam Generator and Terminates Auxiliary Feedwater Flow 7-23-81

2 LOOP-FULL PONER CYCLE II CYCLE 5 0

QO 50 55 60 65 DENSITY; LBi'l/CUBIC FT 7-23-81 FLORIDA STEAM LINE BREAK EVENT POWER 8. LIGHT Co. 7,3,2-1 St. Lucie Pfant MODERATOR REACTIVIT'( FEEDBACK vs unnco <Tno net,c:TT'4 .

1800 1600 2 LOOP-FULL POHER 1I400 1200 1000 800 600 000 AFFECTED SG 200

.0 UNAFFECTED SG

-200 0 100 '00 1100

'00 500 600' TINE, SECONDS LORLDA POWER 8 LIGHT CO. STEAM LINE 8REAK EVENT 5t. Lvcie Plant 713 I 2-2 Unit 1 FEED>YATER FLOVl vs TIME 7-23-81

500 2 LOOP FULL POWER 000 300

~ . 200 100

,0 0 100 200 300 <100 500 600 TINE, SECONDS FLORIDA POV/ER 8 LIGHT CO. STEAM LINE 8REAK EVENT St. Lucie Plant 7,3,2--"

Unit 3.

FEEDV/ATER EHTHALPY vs TIME 7-23-81

2 LOOP-FULL PONER rtODERATOR DOPPLER 0- 0 BOROI'l t j UJ TOTO.L I

SCRAM RODS 0 100 200 00 F00 500 . 600 TINE, SECONDS 7-23-81 F LORIDA POKIER 8 LIGHT CO. STEAhl LINE 8REAK EVENT St. Lvcie Plant 713I2 nt. 4f 774/TTv i<<TTMF

120 2-LOOP FOLL POWER 110 100 88 P 70 U

. 60

~ 50 g 00 30

~

20 10 0

0 100 200 300 000 500 600 TIl1E, SECOI'JDS "LOR[DA POV/ER 8 LIGHT CO. STEAM LINE BREAK QUENT St. Lucie Plant CORE POVlER v~ TIME 7,3,2-5 Unit 1 7-23-81

0 2 LOCP-FULL PONER 100 I

CD 80 CD U

CD 60 C; eo UJ 20' 0 100 200 =

300 000 500 600 TIi~iEr SECONDS FLORlDA POWER 8 LlGHT CO. STEAM LINE BREAK EVENT i<<TTMP'-23-81 St. Lucie Plant ff '.a ~ i not.- Aifao Z t.~ i-ivory Ct [ IV 7,3,2-6

2 LOOP-FULL POMER 600 O

500 TP.VERAGE I

TOUTLET 900 TINLET C = rco 5

~

~ 20.0 100 0 100 200 300 000 500 600

~

TIME, SECONDS 7-23-81 FLORIDA

(

POWER 8 LIGHT CO. STEAM LINE 8REAK EVENT 9 Lucie Plant REACTOR COOLANT SYSTE h TEMPERATURES vs TIME Unit 1

2900 2 LOOP-FULL POWER 2000 16CO 1200 8 800 0

200 300 900 500 600 TINE, SECONDS 7-23-81 FLORIDA POWER 8 LIGHT CO. STEAh) LINE BREAK EVENT St. Lucie Pldnk REACTOR COOLANT, SYSTEM PRESSURE vs TIl'hE 7 3 2-8

+1000 2 LOOP-FULL POWER

+8CO UNAFFECTED SG I

- +600 g +000

+200 AFFECTED SG 0

.ZOO 0 100 200 300 L}00 600 TIf'lE, SECONDS 7-23-81 FLORIDA POWER 8 LIGHT gO STEAM LINE 8REAK EVENT St. Lucie Plant 7,5,2-9 Unit l STEAM GENERATOR PRESSURES vs TIME

5.0 2 LOOP- NO LOAD 3.0 CYCLE 0

~ I CYCLE 5 CD 2.0 l

8 1.0 50.0 5S. O t,'o. o 65. 0 I

NODERATOR DENSITY, LBN/FT 7-23-81 FLORIDA STEAM LINE RUPTURE Q(ENT Figure PO%'ER 8 LIGHT CO.

St. Lucie PIont MODERATOR R'ACTIVITYvs MOD"RATOR DENSITY 7.3.2-10 i S ~

~ ~

~ I I> ~

500 2 LOOP NO LOAD 000 C3 Vl' 500 C)

~

AFFECTED STEAN 6EHERATOR

w. 200.

~ ~

F00 UNAFFECTE D STEAM GENERATOR .

0 0 100 2GO 300 400 500 .600 TIME, SECONDS 7-23-81

. FLORiDA STEAM LINE RUPTURE EVENT Figvre POVIER 8 LlGHT CO.

St. Lucie Plcnt FEEDVlATER FLOVl vs TIME 7.3. 2-11 Unit 1

~ f~

( cl 100 2 LOOP NO lOAD 90 80 70

>-" 60 40 30

.20 p ~

10 1

0 0 200 300 . 400 500 . 6CO

. TIME, SECONDS 7-23-81 FLOa[OA STEAhl LTiilE R~JPTURE EVEiMT Figure POV<'ER c LIGHT CO.

St. Lucia Plant FEEDS/ATER ENTHALPY vs TIME 7.3. 2-~ 2 Unit l

LOOP - NO LOAD MODERATOR DOPPLER

TOTAL BORON SCRAM ROD V/ORTH 100 200 300 400 500 600 TINE, SECQi~JDS 7-23-81 FLORIO'OVIER STEAM LIIXE RUPTUR- EVENT Figure 8, LIGHT CO, St, Lucic Pion) REACTIVITYvs TIME 7.3. 2-1 Unit l

0 0

~ l

~

}00 90 2 LOOP-NO LQAD 80

. <70 60 CO

. 50

,o 40 30 20.

10 0

0 100 200 300 400 500 600 TXME; SECONDS 7-23-81 FLORIOA STEAM LIt<E RUPTURE EVENT Figure POPOVER 8 LIGHT CO.

St. Lucia Plant CORE POVJER vs TItHE 7.3. 2-1 Unit 1

, 120 2 LOOP-HO LOAD I

100 90 80 X ~o 70 60 40

- 30 20 10 00 100 200 3CO 400 TIME, SECONDS 7-23-81 FLORtDA SEAM LIibE RUP URE 'VEJ')T Figure PAYER 8, l.lGHT CO, S<. Lvcic Plant -CORE AiJFRAG'EAT FLUX vs TIME 7.3. 2-15 ';

ilail )

2500 2 LOOP

- f'LO LOAD 2000 a- 1500

~

~>coo I

500 ~ ~

CD 0

~

100 200 300 400 . 500 600 TIME, SECONDS 7-23-81 F LORtDA STEAM LINE RL'PTVRE EVENT Figyre POViER 8 LlGHT CO, St, Lucic Plant REACTOR COOLANT SYS Icl'~'l PRESSURE vs TIME 7.3. 2-1(

Unit 1

600 2 I OOP-NO LOAO c 500 400 ~O N ~

TOUT 300 TAVE

~~

~ ~ ~i~e ~ ~

~ ~ ~ ~+

~ zoo C)

~ 100 0

0 100 200 300 400 500 600 TINE, SECONDS 7-23-81 FLO<loc STEAM.LIME RUPTURE EVENT Fig re PO"/ER 8, LlGHT CO, St, Lucie Plant REACTOR COOLANT SYS it:M TEMPERATURES vs Ti.'AE 7.3. 2-}

1000

2. LOOP-NO LOAD 800 600 UNAFFECTED SC 400

~

I 200 AFFECTED S6.

-100 0 200 300 460 TIME, SECONDS 7-23-81 FLOR)DA STEAM LINE RUPTURE EVEiNT POV/ER 6 LfGHT CO.

St. Lucia Plant STEAM GEi~,'ERATOR PRESSURES vs TIME Unit l

~ 4