ML17212A709

Application for Amend to License DPR-67 Submitted as Response to NRC 810728 Info Request & Proposed Amends to Tech Specs Re Boration Control,Moderator Temp Coefficient, Reactor Coolant Pumps & Boron Dilution & Addition

ML17212A709
Person / Time
Site:	Saint Lucie
Issue date:	09/04/1981
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Clark R Office of Nuclear Reactor Regulation
Shared Package
ML17212A710	List: ML17212A709 ML17212A711
References
L-81-388, NUDOCS 8109100207
Download: ML17212A709 (52)

Text

REGULATOR NFORMATION DISTRIBUTION S TKM (RIDS)

ACCESSION NBR:8109100207 DOC ~ DATE': 61/09/04 NOTARIZED I NO OOCKEll'" 0 FACILt;50-335 St, LUcie PlantE Unit 1< Florida Power 8 Light Co. 05000335 AUTH',NAME'UTHOR AFFILIATION UHRIGgR,E, F l or,i da- Power 8 Light Co, REC IP ~ NAMEl RECIP IENT AF F ILS ATION-CLARiX"PR ~ AD Operating, Reactors Branch 3

SUBJECT:

Application for amend to License DPR-67 submitted as response to NRC 810728 info request 8 proposed amends to Te'ch Specs re boration controlimoderator temp coefficienti reactor coolant pumps 8 boron dilution 8, CODE;: AOOIS,. COPIES RECEEVED:l.iTR addition.'ISTRIBUTION

+ENCL + SIZE'::

Gener al Distr ibution for after>> Issuance~ of Operating Lii cense 'lITLEt:-

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RECIPlKNT COPIES RECIPIENT COPIES ID CODE/NAMEI LTTR ENCL>> IO CODE/NAME LlTTRi ENCLI ACT'ION:" ORB 03 BC! 04" 13 13 INTKRNALl, D/DIRPHU4l FACOB 1 DIRi DI V OF LIC 1 1 I8 Ei 06'R 2 2 OELD 11 1 0 ASSESS'R 10. 1 0 RAD ASMT BR 1 L 01 1 1 09 16 16 LPDR 03 KXTERNALi: ACRS NRC NTIS PDR 02i 1 1

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FLORIDA POWER & LIGHT COMPANY September 4, 1981 L-81-388 c

Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Robert A. Clark, Chief Operating Reactors Branch $/3 I S~P Og 1981 Qi%a +~~95

Subject:

St. Lucie Unit 1 Docket No. 50-335 Stretch Power Proposed Amendment

References:

1. Letter, R. A. Clark to R. E. Uhrig, 7/28/81

2. Letter, R. E. Uhrig to D. G. Eisenhut, L-80-381, 11/10/80

3. Letter, R. E. Uhrig to D. G. Eisenhut, L-81-306 7/23/81

Dear Mr. Clark:

In response to the information request of your Reference 1 letter, we have enclosed responses to your ten (10) questions in Attachment 1 to this letter.

In order to clarify the relationship of our Reference 3 submittal (Shutdown Margin and MTC changes) to our Reference 2 submittal (Stretch Power) we have described the proposed amendment to Stretch below and have enclosed all the pertinent amended Technical Sepcification pages in Attachment 2 to this letter.

Pa es 3/0 l-l R 3/0 1-2 R 3/0 1-5 R 3/0 0-1 R B 3/0 l-l The requirements for shutdown margin were increased, and a shutdown margin calculation change was added. The requirements for part loop operation were simplified and the shutdown margin requirements decreased slightly. The requirement for the moderator temperature coefficient (MTC) at rated thermal power was changed.

The proposed amendment to Stretch has been previously reviewed and approved by the St. Lucie Facility Review Group and the Florida Power

• Light Company Nuclear Review Board. Specifically the new requirements for shutdown margin and MTC are bounded in all the other analyses which use the more conservative values of 0.3% Ijhk/k and -2.5 x 10-< hk/k/OF, respectively.

Further we were able to simplify the requirements for part loop operation 8109100207 Bi 0904 PDR ADOCK 05000335 POR PEOPLE... SERVING PEOPLE

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• because the required reactor coolant pump (RCP) trip causes the full loop and part loop events to behave with no significant differences in results.

T We have enclosed the safety evaluations for the Excess Load (EL) and the Steam Generator Tube Rupture (SGTR) events in Attachment 3 to this. letter.

These events along with the Steam Line Rupture (SLB) event (submitted through Reference 3) were reanalyzed for Cycle 5 to include the effect of NRC mandated TMI-2 related operational and design changes, i.e. automatic initiation of auxilliary feedwater flow and manual, trip of all four RCP s. Other analyses are not significantly affected by these changes. These three event safety evaluations (SLB, EL and SGTR) should replace those submitted through Reference 2. No new Technical Specification changes to Stretch, other than those in Attachment 2 to this letter, arise as a result of the reanalysis of these events. Also the responses to questions on SLB and SGTR (Questions 7,8, and 9) in Attachment 1 to this letter are based on these revised analyses in Attachment 3 and Reference 3.

Very t yours Robert E. Uhrig Vice President Advanced Systems 2 Technology cc: Mr. J. P. O'Reilly, Director, Region II Mr. Harold F. Reis, Esquire

ATTACHMENTI guestion 1 4

The inverse boron'worth values listed in Table 7.1.1-1 are increased for all modes of operation. Increased inverse boron worth means that more boron must be diluted for a given change in reactivity, which is less conservative. Oescribe the bases for and justify the new values of inverse boron worth for each mode of operation.

~Res ense The new inverse boron worths reported in Table 7.1.1-1 are based on explicit diffusion theory ca1culations of reactivity which span the

'power levels and temperature range allowed within each operating mode.

These inverse boron worths are consistent with the critical boron con-centrati'ons shown in Table 7.1,1-1. Although the i'nverse boron worths have increased when compared to the Reference Cycle values, the new values reported in Tab'le 7,1.1-1 are sttll l.ower than the explicit Cycle 4 cal-culated values. Since the new values bound the explicit calculated values, their use jn the, Cycle 4 boron dilution event is justified.

Ouestion 2 1

1 The refueling shutdown margin listed in Table 7.1.1-1 has been changed from 9.45Ã subcritical to 6.28Ã subcritial, which reduces the dilution time to reach criticality . What is the boron concentration that corresponds with the new shutdown margin? Compare this with the previous refueling boron concentration.

~Res ense The critical boron concentration for Cycle 4 is 1280 PPM, in comparison to the reference cycle value of 1200 PPH. The initial boron concentra-tion for both Cycle 4 and the reference cycle is the minimum required Technical Specification boron concentration of 1720 PPM.

~ I uestion 3 The results of the boron dilution events shown in Table 7.1.1-2 list the time to lose prescribed shutdown margin for each mode. Please be aware that SPR Section 15.4.6 specifies minimum times from when an alarm makes the operator aware of an unplanned dilution event as acceptance criteria. What alarms makes the operator aware of boron dilution in each mode? What are the setpoints, time delays, and errors associated with detection and alarm systems, and how are these accounted for in the time for the operator to react to a boron dilution event?

Response

The indicators that are available to the operator for determining if an unplanned dilution is in progress are: 1) the startup flux channels,

2) the low level alarm on the Volume Control Tank, 3) the boronometer and 4 periodic sampling. Depending on the mode of operation and on the rate of dilution, one or all of these indicators would alert the operator that an inadvertent dilution is in progress.

The least amount of time to lose'rescribed shutdown margin is in Mode

5. The primary indicator in Mode 5 is. the startup'lux channels. Two startup flux channels are requi red to be operable in Mode 5 by the Technical Specifications. Procedures will be developed which will require the operator to:

a) Observe the count rate upon entering Mode 5, b) Periodically check that the count rate has not increased (the interval is dependent on the number of charging pumps in operation and the liquid volume .in the RCS),

c) Take corrective action whenever the count rate exceeds a prescribed value (i.e., effectively an alarm limit)

These actions are sufficient because in Node 5 the boron concentration is normally higher than required by Technical Specications. This higher concentration results from not diluting from the higher required concentrations for Nodes 4 and 6.

It should also be noted that past experience at St. Lucie has verified the quality of operator training and operator action during a boron dilution event, LER 335-80-71 reported a boron dilution at power which was correctly controlled by the St. Lucie operators.

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uestion 4 The parameters shown in Table 7.1.4-4 are stated to maximize the calculated peak RCS pressure for a loss of load event. However, the initial pressure of 2200 psia is lower than the value previously utilized (2250 psia) to maximize the RCS peak pressure. Provide further discussion on why a lower initial pressure is conservative, or evaluate the effects of a higher initial pressure on the cal-culated peak pressure.

~Res ense The use of the lowest initial RCS pressure is conservative since this delays the time of High Pressurizer Pressure (HPP) trip.

Delaying the time of HPP trip maximizes the rate of pressure in-crease at the time of trip and ther eby maximizes the pressure over--

shoot after reactor trip. This results in. the peak RCS pressure during theevent. Therefore, the lowest RCS pressure of 2200 psia allowed by the Technical Specification was conservatively assumed to determine. the peak pressure during the Loss of Load ev'ent;

guestion 5 The Loss of Coolant Flow analysis has several areas which are not fully addressed and may be non-conservative. Please discuss the following: 1) The initial core power is at 100% rather than 102Ã as required by SRP Section 15.3.1; 2) The assumed scram characteristics do not discuss are if the most reactive rod is held out of the core; provided to justify the pump coastdown curve.

3) t<o bases

~Res onse

1) Reference 1 documents C-E's statistical combination of uncertainty methodology. The methods and initial conditions used in the Loss of Flow event are consisteqt with /hose reported in geferencq .I,.

In particular, the uncertainty in initial power levei is included as .,a-t'erm in the total uncertainty. Therefore, an initial power level of 100 was assumed in the Loss of Flow event analysis.

2) The,:;,-scram worth used in the analysis was calculated with the most",reactive rod held out of the core.

3) The'-pump coastdown curve used in the Loss of Flow event is calculated using the code COAST (Reference 2). This coastdown curve is'identical to the one .used and accepted by the NRC in the FSAR and previous reload safety analysis..

References

1. CEH-12(F)-'P, "Statistical Combination of Uncertainties, Part.3,"

March 1980.

2. CENPD-98, "COAST Code Description," May 1973.

(}uestion 6 1 t The Loss of Non-Emergency AC Power event utilizes the same DNB analysis used for the Loss of Coolant Flow transient (7.2.2).

The items in question 5 must be satisfactorily resolved before the analysis for Loss of AC Power will be considered valid. 1n

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addition, the value of 1.15 used for the doppler coefficient multiplier must be justified as conservative considering the previous value of 0.85 used in the FSAR.

~Res ense A doppler coeff'icient multiplier of 1.15 was used in the analysis since this results in a slower power rampdown following reactor trip. This increases the residual heat that must be removed during plant cooldown and increases the steam releases. Higher steam releases during the cooldown increases the site boundary doses. Thus, it is conservative to use a doppler coefficient multiplier of 1.15.

uestion 7 provide justification for the values of the initial core coolant temperature and pressure to show that they are conservative for the Steam Line Break analysis. Also, discuss the basis for the initial, core flow rates assumed and the delayed neutron fraction.

~Res ense The maximum initial core coolant temperature allowed by the Technical Specification was used in the analysis. This causes the greatest coolant temperature decrease during the event, which results in the maximum positive reactivity insertion due to moderator feedback.. The greatest amount of positive reactivity insertion enhances the potential for Return-to-Criticality (R-T-C) and Return-to-Power (R-T-P).

The SLB event initiated with the. maximum initial RCS pressure delays the initiation of Safety Injection Actuation Signal (SIAS).

This results in the least amount of negative reactivity added to the core due to boron injected,via the, High Pressure Safety Injection (HPSI) pumps, The smaller amount of negative reactivity inserted enhances the potential for R>>T-C and R-T-P.

0 The maximum value for the delayed neutron fraction at end of cycle was assumed in the analysis. The maximum value increases the subcritical multiplication and thus enhances the potential for R-T-P.

The initial core mass flow rate assumed in the analysis is consistent with'he minimum guaranteed Technical Specification vessel flow rate of 370,000 GPN.

uestion 8 No DNB analysis was performed despite the rapid system depressurization.

What are the minimum DNB ratios calculated?

~Res ense The minimum DNBR during the transient was calculated using the MacBeth rod cluster correlation (Reference 1) with the Lee non-uniform heat flux correction factor (Reference 2). The minimum. transient DNBR for the HFP SLB event occurs at 145 seconds and is equal to 1.27.

References 1.. R. V. MacBeth, "An appraisal of Forced Convection Burn-Out Data",

Proc. Instn. Mech. Engrs., 1965-66, Vol. 180, Pt. 3C, pp. 37-50.

2. D. H. Lee, "An Experimental Investigation of Forced Convection Burnout in High Pressure Mater; Part IV, Large Diameter Tubes at About 1600 psia", AEBl-R 479, November, 1966.

uestion 9

',The Steam Generator Tube Rupture Event shows.a rapid drop in RCS pressure and temperature at about 600 seconds in Figures 7.3.3-3 and 7.3.3-4.

Please provide figures with finer detail in this region (approximately 550 650 seconds) and evaluate the chances of and effects of steam bubble

'o formation in the vessel head or hot legs. The effects of steam bubble formation on the"radiological evaluations should also be considered.

~Res ense As requested, Figures 1 and 2 present in finer detail the RCS pressur e and temperature from 550 seconds to 650 seconds.

The reference, prepared in response to previous NRC questions on upper head voiding, confirms that the model being- used in this analysis adequate1y addresses the effects of steam bubbl'e formation in the vessel upper head and hot'egs during a Steam Generator Tube Rupture event. In addition, the

. the. reference contains an evaluation of the radiological dose due to steam bubble formation.

Letter from Robert E. Uhrig to Darrell G. Eisenhut,

'eference:

"St. Lucie Unit 1 Docket No. 50-335 Natural

'irculation Cooldown", L-81-43, February 9, 1981;

guestion 10 The Seized Rotor analysis does not include a calculated DNB, MDNBR

{accounting for statistical uncertainties with the new C-E methodology) or a peak clad temperature as required by SRP Section 15.3.3 Please provide this information and confirm that the most reactive rod was assumed stuck out of the core.

~Res onse The minimum ONBR for a Seized Rotor event initiated from Technical Specification DNB Limiting Conditions for Operation is 1.025~ As stated in Section 7.3.4, the predicted number of fuel pin failures is not based on a single HDNBR value but is ca]'c61ated through a distribution of the fraction of pins with a par'ticular ONBR as a function of DNBR. This distribution is then,:,convoluted with a probability of burnout vs. DNBR to obtain the amount, of fuel failure.

The scram worth used in analyzing this. event'~was calculated assuming that the most reactive rod is stuck out of the,-.core.

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ATTACHMENT2 ATTACHMENT3 7.1.3 Excess Load Event The Excess Load Event was reanalyzed to determine that the DNBR and CTM design limits are not exceeded during Cycle 5.

The analyses included the effects of manually tripping the RCP's on SIAS due to low pressurizer pressure and the initiation of auxiliary feedwater flow 180 seconds after reactor trip.

The High Power Level and .Thermal Margin/Low Pressure {TM/LP) trips provide primary protection to prevent exceeding the DNBR limit during the full power Excess Load event. Additional protection is provided by other trip signals

.including high rate of change of power, low steam generator water level, and low steam generator pressure. The approach to the. CTM limits is terminated by either the Axial Flux Offset trip, the DNB related trip or the High Power Level trip. In, this analysis, credit is, taken only for the action of the High Power Level trip in the determination of the minimum transient DNBR and maximum CTM. For the zero power Excess Load transient, protection is provided by the Variable High Power Level trip to prevent violation of the DNBR and CTM liririts.

As presented in the FSAR, the most limiting load increase events at full power and hot .Rem power conditions"occur'or the. complete opening of the.

steam dump and bypass 'valves. Of these two events, the full power case is the'more limiting {i.e., approaches closer to the acceptable DNBR and CTM limits) case.

For conservatism in the analyses, auxiliary feedwater flow rate corresponding to 15.3% of full power main feedwater flow (i.e., 7.66K of full power main feedwater flow per generator) was assumed. The addition of the auxiliary feedwater to each steam gener ator was initiated at 180 seconds, after reactor trip. The addition of auxiliary feedwater enhances the cooldown of the RCS and the potential for a return-to-power {R-T-P) or criticality arising from reactivity feedback mechanisms.

>The Excess Load event at full power was initiated at the conditions given in Table 7.1.3-.1. A, Moderator Temperature Coefficient of -2.5x10-" ap/oF was assumed in the analysis. This MTC, in conjunction with the decreasing coolant inlet temperature, enhances the rate of increase in the core heat flux at the tfme of reactor trip. 5 minimum Fuel Temperature Coefficient (FTC), corresponding to beginning of cycle conditions with an uncertainty of 155, was used in the analysis since this FTC results in the least amount of negative reactivity addition to mitigate the transient increase in core heat flux. The minimum CEA worth assumed to be available for shutdown at the tjme of reactor trip for full power operation is 4.3Xap. The analysis conservatively assumed that the worth of boron injected by the safety injection system is -1.0Ãap per 105 PPM. The pressurizer pressure control system was assumed to be inoperable because this minimizes the RCS pressure during the event and therefore reduces the ca1culated DNBR. All other control systems were assumed to be in manual mode of operation and have no significant impact on the results for this event.

The Full Power Excess Load event results in a High Power Level trip at 8.4 seconds. The minimum DNBR calculated for the event at the conditions speci-

. fied in Table 7.1.3-1 is 1.29 compared to the design limit of 1.23. The maximum.

local linear heat generation rate for the event is 18.3 KW/ft.

For the Excess Load event initiated from HFP conditions, SIAS is generated 54.0 seconds. Upon generation of an SIAS, the RCP's are manually tripped by

't the'perator. The coastdown of the pumps decreases the rate of decay heat removal and maintains the RCS coolant temperatures and pressure at higher values.

Auxiliary feedwater flow is delivered to both steam g'enerators at 188.4 seconds.

The subcooled feedwater flow causes an additional cooldown of the RCS. The decreasing RCS temperatures, in combination with a negative MTC, result in positive reactivity insertion which enables the core to approach criticality.

The negative reactivity inserted by the CEAs and the boron injected via the High Pressure Safety Injection (HPSI) pumps,.however, is sufficient to maintain the core in a subcritical condition.

Table 7.1.3 -2 presents the sequence of events for an Excess Load event initiated at HFP conditions. Figures 7.1.3-1 to 7-1.3-5 show the NSSS response for power, heat flux, RCS temepratures, RCS pressure, and steam generator pressure during this event.

The Zero Power Excess Load event was initiated at the conditions given in Table ?.1.3-3. The h)TC and FTC values assumed in the analysis are the same as for the full power case for the reasons previously given. .

The minimum CEA shutdown worth available is conservatively assumed to be -4.3Ãap.

The results of the analysis show that a Variable High Power trip occurs at 44.6 seconds. The minimum DNBR calculated during the event is 3..15 and the peak linear heat generation rate is 11.59 KW/ft.

For the ZP'xcess Load event, an SIAS signal on low pressurizer pressure is generated at 73.7 seconds. At 224.6 seconds auxiliary feedwater

'flow is delivered to both steam generators. The additional positive reactivity resulting from the enhanced cooldown of the RCS is mitigated by the negative reactivity inserted due to the CEAs and the boron injected via the HPSI pumps. The negative reactivity added is sufficient to maintain the core subcritical at all times after auxiliary feedwater flow is initiated.

The .sequence of events for the zero power case is presented in Table 7.1.3-4. Figures 7.1.3-6 to 7.1.3-10 show the NSSS response for core power, core heat flux, RCS temperature, RCS pressure and steam generator pressure.

For the full and zero power Excess Load events initiated by a full opening of the steam dump and bypass valves, the DNBR and CTH limits are not exceeded. In addition, the core remains subcritical following automatic initiation of the auxiliary feedwater flow and manual tripping of the RCP's on SIAS due to low pressurizer pressure. The reactivity transient during a HFP and HZP Excess Load event is less limiting than the corresponding Steam Line Rupture events.

~ Tab1e 7.1.3-1 KEY PARAMETERS ASSUMED FOR FULL POWER EXCESS LOAD EVENT ANALYSIS Parameter Units ~Cele 3 Initial Core Power Level MWt 2754 Core Inlet Temperature OF 551 Reactor Coolant System Pressure psia 2200 Core Mass Flow Rate xlO ibm/hr 133.7 Moderator Temperature Coefficient x10 hp/ F -2.5 CEA Worth Available at Trip -4.3 Doppler Multiplier e85 Inverse Boron Worth PPM/Cap 105 Auxiliary Feedwater Flow Rate ibm/sec 125.4/S.G.

High Power Level Trip Setpoint X.of Full Power 112 Low S.G. Water Level Trip Setpoint 29.9 Reference. Cycle is FSAR. Full Power. Excess Load results were not presented in FSAR, therefore no comparison is made.

Table 7.1.3-2 SEQUENCE OF EVENTS FOR THE EXCESS LOAD EVENT AT FULL POWER TO CALCULATE MINIMUM DNBR Time (sec) Event Set oint or Value 0.0 Complete Opening of Steam Dump and Bypass Valves at Full Power Hi g h Power Tri p Si gna1 Generated 112K of full power 8.8 Trip Breakers Open 9.3 CEAs Begin to Drop Into Core 9.3 Maximum Power; 114.4X of full power Maximum Linear Heat Generation Rate Occurs 18.3 KW(ft 10.0 Minimum DNBR Occurs 1.29 54.0 Safety Injection Actuation Signal 1578 psia Generated; Manual Trip of RCP's 54.1 Pressurizer Empties 69.3 Rampdown of Main Feedwater Flow 5X of full main =

Completed feedwater flow 72.5 Main Steam Isolation Signal 578, psia 73.3 Low Steam Generator Level Trip 29.9 ft Setpoint Reached 13P. 5 Isolation of Main Feedwater Flow to Both Steam Generators 188.4 Auxiliary Feedwater Flow Oelivered 125.4 lb/sec to each to Both Steam Generators steam generator 600.0 Operator Terminates Auxiliary Feedwater Flow to Both Steam Generators

t t KEY PARAMETERS ASSUMED FOR HOT STANDBY EXCESS LOAD EVENT ANALYSIS Parameter Units ~Cele 5 Initial Core Power Level MWt Core Inlet Temperature 0F 532 Reactor Coolant System Pressure psia 2200 6 137.0 Core Mass Flow Rate x10 1 bm/hr Moderator Temperature Coefficient x10 hp/ F -2.5 CEA Worth Available at Trip -4.3 Doppler Multiplier Xhp"'PM/Sap .85 Inverse Boron Worth 100 Variable High Power Trip'Setpoint 5 of-'-.full Power 40 Low S.G. Mater Level-Trip Setpoint ft 29.9 Auxiliary Feedwater Flow Rate ibm/sec 125.4/S.G.

Reference Cycle is FSAR..

'Table'7;l;3-'4 SEQUENCE OF EVENTS FOR EXCESS LOAD EVENT AT HOT STANDBY CONDITIONS TO CALCULATE MINIMUM DNBR Time (sec) Set oint or Value 0.0 Steam Dump and Bypass Valves Open to Maximum Flow Capacity 45.6'vent 44.6 45.0

.Variable High Generated Trip Breakers Power Open Trip Signal 40K of full power

'5.5 CEAs Begin to Drop in the Core Maximum Power; 41.09Ã of. full power Maximum Linear Heat Generation Rate Occurs 11.59 KM/ft.

"vv Ig 46.1 Minimum DNBR Occurs (CE-.2) ' 3.15 67.7 Pressurizer Empties 0

71.1 Main Steam Isolation Signal 578 psia Generated 73.7 Safety Injection Actuation Signal 1578 psia Generated; Manual Trip of Reactor Coolant Pumps 131.1 Isolation of Main Feedwater Flow to Both Steam Generators 224.6 Auxiliary Feedwater Flow Delivered 125.4 lb/sec to each to Both Steam Generators steam generator 600.0 Operator Terminates Auxiliary Feedwater Flow to Both Steam Generators

12G FIJLL POWER iGO CD LIJ

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ul 40 1GG 200 3GG 400 SGO TINE~ SECONDS FLORIDA EXCESS LOAD INCIDENT FIGURE POWER 5 LIGHT CO St. Lucie Plant CORE POMER VS TINE 7.1.5-1 Unit 1

120 I FULL POMER 1GO I

OC 80 UJ 40 20 1GO 200 300 40G 6GO TINE . SECONDS FLORIDA LIGHT COe EXCESS LOAD INCIDENT FIGURE POWER 8 St. Lucie Plant HEAT FLUX YS TINE 7I 1I 3-2 Unit 1

'GG FULL POWER TOUT TAVG CY.'00 Z:

300 = AVERAGE CORE COOLANT TENPERATURE TAYG CORE OUTLET TENPERATURE OUT TIN CORE INLET TENPERATURE 100 0 100 200 300 4GO SO0 S00 TINE SECONDS FLORIDA POWER LIGHT CO> EXCESS LOAD INCIDENT FIGURE 8t St. Lucie P1ant TPIPERATURE YS TIYiE ~

7.1,3-3 Unit 1

2400 FULL POWER 2GGG 1600 12GG SGO 1GO 200 300 400 SGO BGG TIME . SECONDS FLORIDA EXCESS LOAD INCIDENT FIGURE PONER 5 LIGHT CO<

St. Lucie Plant NAIN STEAN PRESSURE VS TINE 7.1,3-0 Unit I

>2GO FULL POMER

+(y.1 p

>" the PLHGR SAFDL is not approached. The Thermal Margin/Low Pressure trip, with conservative coefficients which account for the limiting radial and axial peaks, maximum inlet temper'ature, RCS pressure, core power, and conservative CEA scram characteristics, would be the primary RPS trip intervening during the course of the tran-sient. However, to maximize the coo'lant transported from the primary to secondary and thus the radioactive steam releases to the atmosphere, the analysis was performed assuming that the reactor trip is not initiated un-til the minimum setpoint (floor) of the Thermal Marqin/Low Pressure trip (i.e., Low Pressuri er Pressure Trip) is reached. This prolongs the steam releases to the atmosphere and thus maximizes, the site boundary doses. The Steam Generator Tube Rupture'as analyzed for a power level of 2754 M!(t (102/ of 2700 Ml<t)..The'results will be applicable to 2560 Milt since the higher operating power leads to more conservative site boundary doses. The analysis assumes operation of 3 High Pressure Safety Injection pumps. This assumption leads to faster refilling of the pressurizer, therefore resulting in higher RCS pressure and thus, increasing the primary to secondary leak. The methodology followed is consistent with the methods previously used and approved by NRC. These methods are documented in Reference 3. Table 3 ' 3 '-1 shows the key parameters assumed in the analysis of the event. The sequence of events for the SGTR event with manual trip of RCPs is presented in Table 3.2.3.3-2. The analysis conservatively assumed that at 1800 seconds, the operator 'nitiates cooldown by using the Atmospheric Dump Valves (ADV). The analysis did not credit the use of steam dump and bypass system to the condenser. The use of atmospheric dump valves results in a substantial increase in the calculated site boundary dose since the ADV partition factor is .1 compared to .0005 for the condenser air ejectors. Figures 3.2.3.3-1 through 3.2.3.3-5 present the transient behavior of core power, heat flux, RCS pressure, RCS temperatures, and steam generator pressure. I-131 activity release is based on the Tech Spec allowed primary to secondary leak rate of 1 GPM and on the steam flow required to cool the plant to condi-tions where the shutdown cool,ing system can be initiated. This release is calculated as the product of-:st'earn flow, the time dependent steam activity and the decontamination factors applicable to each release pathway. The 0 to 2 hour I-131 site;boundary dose, is calculated from: AI-131 + BP x > x DDSE (REM) CFI-131 where: AI-131 activity I-131 released .to site boundary, Ci, BR breathing rate, m /sec, x/Q dispersion coefficient, sec/m , CFI-131 I-131 dose conversion factor, Rem/Ci. . In determining the whole body dose,'he major assumption made is that all noble gases leaked through the ruptured tube will be released to the atmosphere. Therefore, the whole body dose is proportional to the total primary to secondary leak and is calculated using the following equation. i<hole Body Dose = [ .25 (K + )] * .25 E g L *A RCS g ,where: E average energy release by gamma decay, Y E average energy release by beta decay, total primary to secondary mass transport RCS noble gas activity of primary coolant g/(( di".)ii'r'n c'o<'t l i < i<'nt. ~ ~ The results of the analysis are that 81540 lbs. of primary coolant are transported to the steam generator secondary. side. Based on this mass transport and values in Table 3.2.3.3-3, the 0-2 Hr site boundary doses calculated are: Thyroid (DEQ I-ll ): 0.32 REN Whole Body (DEQ Xe-133): 0.08 REN The reactor protective system (i.e., TN/LP trip) intervenes to protect the core from exceeding the DHBR limit. The do'ses resulting from the activity released as a consequence of h double-ended rupture of one steam generator tube, assuming the maximum allowable Tech Spec activity for the primary concentration at a core power of 2754 NIlt, are significantly below the guidelines of 10CFR100. Thus, the results do not e'xceed acceptance criteria. TABLE 7.3.3-"1 KEY PARAMETERS ASSUMED IH THE STEAN GENERATOR TUBE RUPTURE EYEtlT KEY TRANSIENT RELATED PARAMETERS: Parameter Units FSAR ~Cele 5 Power 2611 2754 MTC xl0 ap/'F -2.5 -2.5 Doppler Coefficient 1.15 1.15 Multiplier Scram 1/orth 4.55 -4.0. 544 in RCS Pressure psia 2300 2300 Initial Core Mass Flow Rate x10. lb/hr 117,.5 133.9 (548oF, 2200 psia) Initial Secondary Pressure Dsi a 841 9O2. 0 g~ ~ TABLE 7.'3; 3-2 SEQUENCE OF EVENTS FOR THE STEAM GENERATOR TUBE RUPTURE EVENT WITH RCP COASTDOllN ON SIAS Time (sec) Event Setpoint or Value 0.0 Tube Rupture Occurs 577.2 Low Pressurizer Pressure Trip 1853 psia Signal Generated 577.4 Dump Valves Open 578. 6 CEAs Begin to Drop Into Core 579.1 Bypass Valves Open 584.8 Maximum Steam Generator Pressure 949 psia 587.4 Pressurizer Empties 588.0 Safety Injection Actuation Signal Generated; RCPs Manually Tripped I 1578 psia 1395.4 Minimum RCS Pressure 1034 psia 1800.0 Operator Isolates Damaged Steam Generator and Begins Cooldown to 325'F 7859 Operator Initiates Shutdown Cooling (TAV F) TABLE 7. 3. 3-3 ASSUMPTIONS FOR THE RADIOLOGICAL EVALUATION FOR THE STEAM GEhERATOR TUBE RUPTURE Parameter uni ts C cle 5 Value Reactor Coolant System Maximum yCi/gm 1.0 Allowable Concentration (DEQ I-131) Steam Generator Maximum Allowable uCi/gm Concentration (DEQ I-131)1 Reactor Coolant System Maximum pCi/gm 100/E Allowable Concentration of'oble Gases (DEQ Xe-133)1 Atmospheric Dump Valve Partition Factor Condenser Air Ejector Partition Factor .0005 Atmospheric Dispersion Coefficient sec/m 8.55x10 . Breathing Rate m /sec 3.47x10 Dose Conversion Factor (I-131) REM/Ci 1.48xlO Tech Spec limits. P 0-2 hour accident condition for St. Lucie Unit 1. ,c7 C 110 99 8" 77 55 UJ gq CD 32 22 0 200 900 600 800 1000 1200 1000 1600 1800 TIl'lE, SECONDS FLORIDA )TEAt"l GEI",ERECTOR TUBE FAILURE EVENT Figure ~ POV/ER 6 LICL<T CO. St. Lvcie PIont ~ CORE POYiER vs TIk'IE 7.3.3-1 ~1 ~ 110 ag 77 66 55 22 I I I 0. 0 200 400 600 800 1000 120Q 1000 160Q 1800 TINE, SECONDS STEAM GENERATOR TU BE FAILLE,iE E'LtEiilT FlgVf C F l.OR IDA t'O'"'L'."", l.t.C i !T CC. CORf AVFiliXGF flEAT FI Ui'i vs TIA'iE 7 ~3 &3 2 5t. Lv-te f'loci( ~ ~ ~ ~ ~ ~ 2403 220D 2003 1800 1603 D . 1403 1203 1000 .0 200 000 600 800 1000 1200 1400 1600 . 1300 TINE, SECONDS Ficure FLORIDA S) EAM GEIJERATOR TURNE FAILURE EVEi'lT POVCER 5 LlGt<T CQ. 7 3 &3 3 St. Lucio Plont REACTOR COOLAl'JT SYSTEM:l PRESSURE vs TIi~,E 65J TOtjTLET 603 TAYERAGE TINLET 550 500 CD I I I I 05.0 0 200 400 600 800 . 1000 . 1200 1000 1600 1800 TIME, SECONDS STEAI';1 GENERATOR TU~uE FAILURE EVE:~T FIgv;c: Fl ORIDA POWER 8 LIGHT CO. PEACTPR CPOl.ANT S'(STEMMA TEMPERATURE vs TITLE 7.3.3-4 5t, Lvc i c P I a i: I 950 900 850 800 750 c 700 650 600 550 I I I I I I 500 0 200 000 600 800 1000 1200 1000 1600 1800 TINE, SECONDS Figur c 7.3.3- Q f v,~ ) I <