ML17227A788

Large Break Loca/Eccs Analysis w/25 + or - 7% SG Tube Plugging.

ML17227A788
Person / Time
Site:	Saint Lucie
Issue date:	02/28/1993
From:	Salim P SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17227A785	List: ML17227A784 ML17227A786 ML17227A787 ML17227A788 ML17227A789
References
EMF-92-176, NUDOCS 9303310064
Download: ML17227A788 (52)

Text

SIEMENS EMF-92-1 76 St. Lucie Unit 1 Large Break LOCA/ECCS Analysis With 25~7% SGTP February 1993 Siemens Power Corporation Nuclear Division 93OSSZ006e 9303X9 PDR ADOCK 05000335 P PDR

EMF-92-1 76 Issue Date:

St. Lucle Unit 1 Large Break LOCA/ECCS Analysis With 25+?% SGTP Prepared by:

P >gJ~

P. Salim, Engineer PWR Reload Analysis PWR Nuclear Engineering Contributor: K. M. Duggan, Engineer February 1993

/skm

CU 0 S I 0 0 C OARDI 0 CO A DUS 0 IS QOCUQQQ' R CA SIemens Power Corporation's warranties and representations concerning the subject matter of this document are those set forth In the Agreement between Semens Power Corporation and the Customer pursuant to which this document Is krsued. Accordingly, except as otherwise expressly provided in such Agreement, neither Semens Power Corporation nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the Information contained in this document, or that the use of any Information, apparatus, method or process dlsdosed ln this document wN not Infringe privately owned rights; or assumes any liabilities with respect to the use of any information, apparatus, method or process disdosed In this document.

The information contained herein ls for the sole use of the Customer.

In order to avoid impairment of rights of Siemens Power Corporation in patents or inventions which may be induded in the information contained in this document, the red pient, by its acceptance of this document, agrees not to publish or make public use gn the patent use of the term) of such Information until so authorized in writing by Siemens Power Corporation or until after sbt (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document.

EMF-92-1 76 Page i Table of Contents Section ~acae 1 e0 INTRODUCTION e v ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 1 2.0

SUMMARY

Of RESULTS................,,...........,........... 2-1 3.0 LARGE BREAK LOCA/ECCS ANALYSIS ............., 3-1 3.1 Description of LBLOCA Transient............,, 3-1 3.2 Description of Analytical Models ................. ~ 3-2 3.3 Plant Description and Summary of Analysis Parameters..... . ~...... 3-3

......... ~............

~

3.4 LBLOCA Results 3A

4.0 CONCLUSION

S,...,,... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

5.0 REFERENCES

.....,,............ .,... ~ 5-1

EMF-92-1 76 Page ii List of Tables Table Pacae 2.1 Summary of LBLOCA Results for 0.8 DECLG Cases ...... 2-3 3.1 St. Lucie Unit 1 System Analysis Parameters...........,

3.2 Core and Fuel Design Parameters . 3-8 3.3 Event Times for 0.8 DECLG Break Limiting Case (EOC Shape) . ~ ~ ~ ~ ~ ~ ~ 39

EMF-92-1 76 Page iii List of Fi ures

~FI ure Pacae 2.1 Allowable Peak Linear Meat Rate vs Burnup .

3.1 Normalized Power vs Time .......,...,,............... ~ . ~ . ~.... ~ ~ ~ . 3-10 3.2 Single Intact Loop Safety Injection Tank Flow Rate vs Time .............. 3-11 3.3 Double Intact Loop Safety Injection Tank Flow Rate vs Time 3-12 3.4 Broken Loop Safety Injection Tank Flow Rate vs Time 3-13 3.5 Single Intact Loop HPSI Plus LPSI Flow Rate vs Time ... 3-1 4 3.6 Double intact Loop HPSI Plus LPSI Flow Rate vs Time 3-1 5 3.7 Broken Loop HPSI Flow Rate vs Time...... 3-1 6 3.8 Broken Loop LPSI Flow Rate vs Time ........... ~..... ~..........,

~ 3-1 7 3.9 Upper Plenum Pressure vs Time....... 3-1 8 3.10 Total Break Flow Rate vs Time ........,,... ~, 3-19 3.11 Average Core Inlet Flow Rate vs Time for EOC Shape (HC Run) .......... 3-20 3.12 Average Core Outlet Flow Rate vs Time for EOC Shape (HC Run) ........... 3-21 3.13 Hot Assembly Inlet Flow Rate vs Time for EOC Shape (HC Run)............. 3-22 3.14 Hot Assembly Outlet Flow Rate vs Time for EOC Shape (HC Run) ~ ~ ~ ~ ~ ~ ~ ~ ~ 3 23 3.15 Hot Assembly Flow Rate from Middle Core Volume to Upper Core Volume vs Time for EOC Shape (HC Run) . ~ ..,, . ~ ~ ~ ..,...,..... ~ . ~ ~ ~ ~ ~ 3-24 3.16

~...........,....

Hot Assembly Upper Core Volume Fluid Quality vs Time for EOC Shape (MC Run) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3 25 3.17 Hot Assembly Upper Core Volume Fluid Temperature vs Time for EOC Shape (HC Run) .... 3-26 3.18 PCT Node Fuel Average Temperature vs Time for EOC Shape (HC Run) . ~.... 3-27

EMF-92-1 76 Page iv

~tC i d)

~Fi ure Pacae 3.19 PCT Node Cladding Surface Temperature vs Time for EOC Shape (HC Run) ....., .................................. 3-28 3.20 PCT Node Heat Transfer Coefficient vs Time for EOC Shape (HC Run) . ~ ..:. 3-29 3.21 PCT Node Heat Flux vs Time for EOC Shape (HC Run) 3-30 3.22 PCT Node Cladding Surface Temperature vs Time for EOC Shape (TOODEE2 Run) . ~.... ~.......... 3-31 3.23 Downcomer Mixture Level vs Time for EOC Shape.............. 3-32 3.24 Effective Reflood Rate vs Time for EOC Shape ......, ............ 3-33 3.25 Core Mixture Level vs Time for EOC Shape,.... 3-34 3.26 Containment Pressure vs Time for EOC Shape,........,.... ~ ~.... 3-35

EMF-92-1 76 Page 1-1 1.0 INTRODUCTlON This report documents the results of a large break LOCA/ECCS analysis. that was performed for St. Lucie Unit 1. The large break LOCA event required re-analysis due to increased steam generator tube plugging (SGTP). The analysis was performed with an average steam generator tube plugging of 25%, with a tube plugging asymmetry of + 7%. The analysis was performed at an increased radial peaking factor of 1.75, even though the maximum allowed radial peaking factor (Fr) in the Technical Specifications remains unchanged at 1.70. The purpose of incorporating the increased radial peaking factor was to support potential future increases in the Technical Specification radial peaking factor. The analysis also addresses primary coolant temperature coastdown at full power Endwf-Cycle conditions with a maximum reduction in primary coolant temperature of 26'F. The Cycle 10 fuel design remains bounded by Reference 1 and was not re-analyzed as part of this analysis.

EMF-92-1 76 Page 2-1 2.0

SUMMARY

OF RESULTS The large break LOCA (LBLOCA) analysis was performed at the previously determined limiting break size, which was a 0.8 Double-Ended-Cold-Leg-Guillotine (DECLG) break< >. System and hot channel blowdown calculations were performed with steam generator tube plugging of 25

~7% to reconfirm the 0.8 DECLG break as the limiting break size. Calculations for 0.6, 0.8, and 1.0 DECLG break sizes reconfirmed the 0.8 DECLG break as the limiting break size.

Two cases were analyzed at the limiting break size to support an axially and exposure independent LHR limit of 15 kW/ft. The first case combined the maximum fuel stored energy, which occurs at a hot rod average burnup of 1.8 MWd/kg, with a bounding axial power shape which conservatively represents axial shapes that may exist between Beginning-of-Cycle (BOC) and Middle-of-Cycle (MOC) The second case combined the fuel stored energy at MOC with a

~

bounding axial shape which conservatively represents axial shapes that may exist between MOC and Endwf-Cycle (EOC). This approach conservatively bounds the possible combinations of fuel rod stored energy and axial power shapes in the St. Lucie Unit 1 plant. The results for these two cases are summarized in Table 2.1. The peak cladding temperature was calculated to be 191 2'F for the case with the EOC axial shape.

Calculations were performed in Reference 2 to support an End-of-Cycle full power primary coolant system T~e coastdown of 26'F. The PCT for that case was demonstrated to be significantly lower than the limiting case due to a significant decrease in fuel stored energy at End+f-Cycle. Although, a specific calculation was not performed in this analysis for an End-of-Cycle T~e coastdown of 26'F, the PCT for this case would also be significantly lower than the limiting PCT reported above due to a significantly lower fuel stored energy at Endwf-Cycle.

Therefore, an Endwf-Cycle T~e coastdown of 26'F is supported for an increased steam generator tube plugging of 25% with an asymmetry of +7%.

The Cycle 10 fuel remains bounded by the analysis performed in Reference 1. The Cycle 10 fuel will be third cycle fuel in Cycle 12 and will operate at significantly lower power levels than fresh fuel. The lower power level will more than offset any adverse effects of increased steam

EMF-92-1 76 Page 2-2 generator tube plugging and asymmetry. The Technical Specification Fr remains at 1.7 for the Cycle 10 fuel.

EMF-92-176 Page 2-3 Table 2.1 Summary of LBLOCA Results for 0.8 DECLG Cases BOC Stored MOC Stored Energy, MOC Energy, EOC Axial Shape Axial Shape Parameter L = 0.77 L = 0.85 Maximum LHR (kW/ft) 15 15 Hot Rod Burst

- Time (sec) 41.79 44.10

- Elevation (ft) 8.97 9.97

- Channel Blockage Fraction 0.399 0.434 Peak Cladding Temperature

- Temperature ('F) 1852 1912

- Time (sec) 56.15 73.03

- Elevation (ft) 8.97 10.47 Metal-Water Reaction

- Local Maximum (%) 2.12* 2.09*

- Elevation of Local Maximum (ft) 8.97 9.97

- Hot Pin Average (%) 0.39* 0.34*

- Core Wide Maximum (%) <(1 0>> ( <1.0*

't 200 seconds.

UNACCEPTABLE OPERAT ION ACCEPTAHLE OPERAT ION ill CYCLE L IFE 0 Tl Q

e lO FIGURE 2.1 ALLOWABLEPEAK LINEAR HEAT RATE VS BURNUP IQ 0)

~,

EMF-92-1 76 Page 3-1 3.0 LARGE BREAK LOCA/ECCS ANALYSIS Section 3.1 of this report provides a description of the postulated large break loss~f-coolant transient. Section 3.2 describes the analytical models used in the analysis. Section 3.3 provides a description of the St. Lucie Unit 1 plant and a summary of plant and fuel parameters used in the LOCA analysis. Section 3.4 describes the results of the LOCA analysis including justification to support a full power primary coolant temperature coastdown at endwf-cycle.

3.1 Descri tion of LBLOCA Transient A loss-of~oolant accident (LOCA) is defined as the rupture of the Reactor Coolant System (RCS) primary piping up to and including a double-ended guillotine break. The limiting break occurs on the pump discharge side of a cold leg pipe. The LOCA is assumed to result from an earthquake and is co-incident with the loss'-offsite power. Primary coolant pump coastdown occurs co-incident with the loss-of-offsite power. Following the break, depressurization of the reactor coolant system, including the pressurizer, occurs. A reactor trip signal occurs when the pressurizer low pressure trip setpolnt is reached. Reactor trip and scram are conservatively neglected in the LOCA analysis. Early in the blowdown, the reactor core experiences flow reversal and stagnation which causes the fuel rods to pass through critical heat flux (CHF).

Following CHF, the fuel rods dissipate heat through the transition and film boiling modes of heat tiansfer. Rewet is precluded during blowdown by Appendix K of 10 CFR 50.

A Safety Injection System (SIS) signal is actuated when the appropriate setpoint (high containment pressure) is reached. Due to loss-of-offsite power, a time delay for startup of diesel generators and SIS pumps is assumed. Once the time delay criteria is met and the system pressure falls below the shutoff head of the High Pressure Safety Injection (HPSI) and Low Pressure Safety Injection (LPSI) pumps, SIS flow is injected into the cold legs. When the system pressure falls below the Safety Injection Tank (SIT) pressure, flow from the SITs is injected into the cold legs. Single failure criteria is met by assuming that one LPSI pump is not available for operation. Flow from the Emergency Core Cooling System (ECCS) is assumed to bypass the core and flow to the break until the end-of-bypass (EOBY) is predicted to occur (sustained downflow in the downcomer), Following EOBY, ECCS flow flills the downcomer and lower plenum

EMF-92-176 Page 3-2 until the liquid level reaches the bottom of the active core (beginningwf-core-recovery or BOCREC time). During the refill period, heat is transferred from the hottest fuel rod to surrounding rods by radiation heat transfer.

The reflood period begins at BOCREC time. ECCS fluid fills the downcomer and provides the driving head to move coolant through the core. As the mixture level moves up the core, steam is generated. Steam binding occurs as the steam flows through the intact and broken loop steam generators and pumps. The pumps are assumed to have a locked rotor (per Appendix K of 10 CFR 50) which tends to reduce the reflood rate. The fuel rods are eventually cooled and quenched by radiation and convective heat transfer as the quench front moves up the core. The reflood heat transfer rate is predicted through experimentally determined heat transfer and carry-over rate fraction correlations.

The purpose of the LBLOCAanalysis is to demonstrate that the criteria stated in 10 CFR 50.46(b) are met. The criteria are:

1) The calculated peak fuel element cladding temperature does not exceed the 2200 'F limit.

2) The amount of fuel element cladding which reacts chemically with water or steam does not exceed 1% of the total amount of zircaloy in the core.

3) The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling. The hot fuel rod cladding oxidation limit of 17% is not exceeded during or after quenching.

4) The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.

3.2 Descri tlon of Ana ical Models (3)

The SPC EXEM/PWR evaluation model was used to perform the analysis. This evaluation model consists of the following computer codes:

(4)

1) RODEX2 for computation of initial fuel stored energy, fission gas release, and gap

, conductance;

2) RELAP4-EM for the system and hot channel blowdown calculations; 0

EMF-92-1 76 Page 3-3

3) CONTEMPT/LT-22 as modified in accordance with NRC Branch Technical Position CSB 6-1 for computation of containment back pressure;

4) REFLEX for computation of system reflood; and

5) TOODEE2 for the calculation of fuel rod heatup during the refill and reflood portions of the LOCA transient.

The quench time, quench velocity, and carryover rate fraction (CRF) correlations in REFLEX, and the heat transfer correlations in TOODEE2 are based on SPC's Fuel Cooling Test Facility (FCTF) data.

The governing conservation equations for mass, energy, and momentum transfer are used along with appropriate correlations consistent with Appendix K of 10 CFR 50. The reactor core in RELAP4 is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback, and with actinide and decay heating as required by Appendix K. Appropriate conservatisms specified by Appendix K of 10 CFR 50 are incorporated in all of the EXEM/PWR models.

3.3 Plant Descri tion and Summa of Anal sis Parameters The St, Lucie Unit 1 plant is a Combustion Engineering (CE) designed pressurized water reactor which has two hot leg,pipes, two U-tube steam generators, and four cold leg pipes with one recirculation pump in each cold leg. The plant utilizes a large dry containment. The reactor coolant system was nodalized into control volumes representing reasonably homogeneous regions, interconnected by flow paths or "junctions". The two cold legs connected to the intact loop steam generator were assumed to be symmetrical and were modeled as one intact cold leg with appropriately scaled input. The model considers four SITs, a pressurizer, and two steam generators with both primary and secondary sides of the steam generators modeled. ECCS flow from the HPSI and LPSI pumps was modeled using filljunctions at the SIT lines, with conservative flow rates given as a function of system back-pressure. The primary pump performance curves are characteristic of CE pumps. The reactor core was modeled radially with an average core and a hot assembly as parallel flow channels, each with three axial nodes. A steam generator tube plugging level of 25% was assumed with an asymmetric steam generator tube plugging of ~7%.

U

EMF-92-1 76 Page 3-4 The break was conservatively assumed to have occurred in the most highly plugged loop since this results in more steam binding during reflood and a higher peak cladding temperature.

Values for system parameters used in the analysis are shown in Table 3.1. Core and fuel design parameters are shown in Table 3,2. The primary coolant loop flow split was recalculated in this analysis due to the increase in assumed steam generator tube plugging to 25%. The calculation of loop flow rates used a Technical SpecNcation loop flow rate of 355,000 gpm. The minimum measured flow rate allowed for plant operation is the Technical Specification loop flow rate plus measurement uncertainty. Therefore, the flow rate used in the analysis is conservatively low relative to allowed plant operation.

3,4 LBLOCA Results System and hot channel blowdown calculations were performed at the increased steam generator tube plugging- of 25 ~7% to reconfirm the 0.8 DECLG break size as the limiting break.

Calculations were performed for DECLG break sizes of 0.6, 0.8, and 1.0. The fuel and cladding temperatures at EOBY confirmed that the 0.8 DECLG break size remains the limiting break size.

Two cases were analyzed at the limiting break size to support an axially and exposure independent LHR limit of 15 kW/ft, The first case combined the maximum fuel stored energy, which occurs at a hot rod average burnup of 1.8 MWd/kg, with a bounding axial power shape which conservatively represents axial shapes that may exist between BOC and MOC (peaked highest in the core at a relative core height of 0.77). The second case combined the fuel stored energy at MOC with a bounding axial shape which conservatively represents axial shapes that may exist between MOC and EOC (peaked highest in the core at a relative height of 0.85). Axial shapes at 100% power with ASls from 0.0 to <.2 were reviewed to determine the shape peaked highest in the core. Rod positions included ARO, rods at their 100% power PDIL, and rods at their 90% PDIL The limiting axial shapes occurred at the ARO condition. This approach conservatively bounds the possible combinations of fuel rod stored energy and axial power shapes in the St. Lucie Unit 1 plant.

EMF-92-1 76 Page 3-5 The results for the two axial shape cases analyzed are summarized in Table 2.1. The peak cladding temperature was calculated to be 1852'F for the case with the MOC shape and 1912'F N

for the case with the EOC axial shape. The maximum local cladding oxidation was calculated to be less than 3%, which is much less than the allowed local oxidation of 17%. Core wide oxidation was determined to be much less than the allowable 1%.

Plots of various parameters for the EOC shape limiting case are shown in Figures 3.1 through 3.26. Event times for the limNng EOC shape case are shown in Table 3.3.

An Endwf-Cycle full power primary coolant temperature coastdown with a maximum reduction in primary coolant temperature of 26'F is also supported for the plant. The justification for the support is based on the primary coolant temperature coastdown calculations from the Cycle 11 LBLOCA analysis, The calculations used fuel stored energy at an EOC burnup combined with a bounding axial shape representative of EOC, peaked at a relative core height of 0.85. The Cycle 11 primary coolant temperature coastdown analysis predicted a significantly lower PCT than the PCT calculated for the limiting case. A lower PCT was predicted due to a lower fuel stored energy at EOC. A similar effect will occur for the case with increased steam generator tube plugging. The limNng case would bound the primary coolant temperature coastdown case.

Therefore, a re-analysis of the T~~ coastdown case was not necessary. The Cycle 11 analysis justifies a reduction in the average temperature of 26'F at 100 percent power and at EOC for ST.

Lucie Unit 1, with 25 ~7% SGTP.

The Cycle 10 fuel design was analyzed in Reference 1 for a steam generator tube plugging level of 15%, The Cycle 10 fuel will be in its third cycle during Cycle 12 and will be operating at a significantly lower power level than fresh fuel. The'ower power level will more than offset any adverse effects of increased steam generator tube plugging to 25 ~7%. The Cycle 10 fuel is therefore bounded by the Reference 1 analysis for the LBLOCA event. The Technical Specification Fr limit for Cycle 10 fuel remains at 1.70.

EMF-92-1 76 Page 3-6 Table 3.1 St. Lucle Unit 1 System Analysis Parameters Primary heat output, MWt 2700>>

Primary coolant flow rate, Ibm/hr 1.339x1 0 ** (355,000 gpm)

Primary coolant system volume, ft 18,897***

Operating pressure, psia Cold leg coolant temperature (hottest loop), 'F Reactor'vessel volume, ft 4522 Pressurizer volume (total), fts 1500 Pressurizer volume (liquid), ft SIT Volume (total), ft (one of four) 2020 SIT Volume (liquid), fts 1090 SIT pressure, psia 230 Steam generator tube plugging 32% - 18% split Steam generator secondary pressure, psla 790 Steam generator feedwater temperature, 'F Reactor coolant pump rated head, ft 272 Reactor coolant pump rated torque, ft-Ibf 32,495 Reactor coolant pump rated speed, rpm 886 Reactor coolant pump moment of Inertia, Ibm-ft 101,900 Primary heat output used in RELAP4-EM model = 1,02 x 2700 = 2754 MWt Technical Specifications minimum loop flow rate

• • • Includes total SIT, pressurizer volume and 25% SGTP

EMF-92-1 76 Page 3-7 Table 3.1 St. Lucie Unit 1 System Analysis Parameters (Continued)

Containment volume, fthm 2.511x10 Containment temperature, 'F SIS Fluid temperature, 'F HPSI delay time, sec LPSI delay time, sec 30 Containment fan coolers initiation time, sec 30 Containment sprays initiation time, sec 30

EMF-92-1 76 Page 3-8 Table 3.2 Core and Fuel Design Parameters Cladding O. D ln. 0.440 Cladding I. D., in. 0.384 Fuel O. D., in. 0.377 Fuel rod pitch, in. 0.580 Fuel assembly pitch, in. 8.180 Active fuel length, in. 136.7 Core flow area, ft 53.19 Core bypass flow, % 3.90

EMF-92-1 76 Page 3-9 Table 3.3 Event Times for 0.8 DECLG Break Limiting Case (EOC Shape)

Timing of Event Event sec Start 0.0 Break is fully open 0.05 Safety Injection signal 0.89 SIT flow begins in broken loop 11.54 SIT flow begins in single intact loop 16.07 SIT flow begins in double intact loop 16.08 End-of-bypass (EOBY) 21.11 SIS flow begins 30.89 Beginning-of-core-recovery (BOCREC) 38.35 Cladding rupture 45.11 SIT flow ends in broken loop 61.60 SIT flow ends in single intact loop 65.30 SIT flow ends in double intact loop 65.30 Peak cladding temperature is reached 73.03

12 16 28 32 TINE (SEC)

FIGURE 3.1 NORHALIZEO POWER VS TINE

+a Ko

~ ID C) 4.

6 C)

Cl

~~8 Vc 43 Ua

~ C) lA 60 80 100 120 140 160 TINE (SEC)

FIGURE 3.2 SINGLE INTACT LOOP SAFETY INJECTION TANK FLOW RATE VS TINE foal Ql 0 (Q ~

np I

Q)

60 80 100 120 100 160 Tlt1E (SEC)

FIGURE 3.3 DOUBLE INTACT LOOP SAFETY INJECTION TANK FLOW RATE VS TINE

~ ~

SS a IS wI

CD C3 foal Vl 63 le CL

+O ON 4

0 C3 o

f-e C3 fJ %

CI) 40 60 80 l00 l20 140 t60 TINE (SEC)

FIGURE 3.5 SINGLE INTACT LOOP llPSI PLUS LPSI FLOW RATE VS TIME

C)

CD C3 43 Vl

-I 63 Cd CC Zo O<

4 fL C) 08 C3 CC l

ED Kl C3 Cl 80 l00 l 20 ll0 l60 TINE (SECl FIGURE 3.6 OOUBLE INTACT LOOP I)PSI PLUS LPSI FLOW RATE VS TINE

60 80 100 120 140 160 Tlt1E (SEC)

FIGURE 3.7 BROKEN LOOP IIPSI FLOW BATE VS TINE

I~

,~

~ ~

~ l II el s t f l I

D D

(C 5

lL Vl CA bJ

+ D 0- Pu 0

~8 43 0

0 8

l2 16 28 32 TINE (SEC)

FIGURE 3.9 UPPER PLENUH PRESSURE VS TINE

l2 l6 28 32 TINE (SEC)

FIGURE 3.10 TOTAL DBEAK FLOM BATE VS TIME

C)

~O C3 Og 4J Vl K

Kl Mg 43 lL lL C)

C3 (9 o

~O a=8 I

l2 l6 26 32 TIHE (SEC) fTl FIGURE 3.II AVEMGE CORE INI.ET FI.OW MTE VS TIHE FOR FOC SIIAI'f (IIC RUN) Q (p lO IQ I

Ib N 00

~ .

II) l2 l6 28 32 TIHE (SEC)

FIGURE 3.l2 AVERAGE CORE OUTLET FLOW RATE VS TIHE FOR EOC SIIAI'E (IIC RUN)

QO 4J CA Vl fO>n I

C5 Pl 70 l2 l6 28 32 TIHE (SEC)

FIGURE 3.13 IlOT ASSEHBLY INLET FLOW RATE VS TIHE FOR EOC SIIAPE (flC RUN)

Ci P4 VO l2 l6 26 32 TIHE (SEC) m FIGURE 3.14 IIOT, ASSEHBLY OUTLET FLOW RATE VS TIHE FOR EOC SIIAPE (IIC RUN)

'll I

e np I

IO V Gl Ol

fal f-e CL

<o C)

C) 0 C3

>o I

63 43 CA tg Cl x

7Q l2 l6 TIHE (SEC)

FIGURE 3. 15 HOT ASSEMBLY fLOW RATE FROM MIDDLE CORE VOLUME TO UPPER CORE VOLUME VS TIME FOR EOC SHAPE (HC RUN)

CI C3 4

~ ED 43 tA lA CC f-s O<

X. o 43 Q

0 A

CI CI Cg l2 l6 28 TIHE (SEC)

FIGURE 3.16 HOT ASSEHBLY UPPER CORE VOLUHE FLUID (UALITY VS TIHE FOR EOC SINPE (IIC RUN)

4

~D

~R LL f-s

>o 0

g 43 CI l D 8 m

K LJ Ul 8 C) x,'L QD Q D D

PQ l2 l6 28 32 TIHE (SEC)

FIGURE 3.17 KOT ASSEHBLY UPPER CORE VOLUHE FLUID TEHPERATURE VS TINE FOR EOC SIIAPE (IIC RUN)

CI 12 16 28 32 TIHE (SEC)

FIGURE 3.18 PCT NODE FUEL AVERAGE TEHPERATURE VS TIHE FOR EOC SIIAPE (IIC RUN) m 0

CD CO I

fO V Ol

Id 0

CC Qo Id o Id Id Oo

~IS CA fd C3 C)

O oo IL co o

PcQ l2 l6 28 32 TIHE (SEC)

ITl FIGURE 3.19 PCT NODE CLADDING SURFACE TEHPERATURE VS TIHE FOR EOC SIIAPE (IIC RUN)

Q o (D IQ I

lO V M 0)

12 l6 28 32 TIHE (SEC)

FIGURE 3t20 PCT NODE IIEAT TRANSFER COEFFICIENT VS TINE FOR EOC SIIAPE (IIC RUN)

l2 l6 28 32 TINE (SEC)

FIGURE 3.21 PCT NODE IIEAT FLUX VS TINE FOR EOC SIIAPE (tlC RUN)

II I II II a

I I I I I I ~ I I Il I II I I I I I I I Il I II I I I ~

~

ii ~ll III

20 30 40 50 60 70 80 90 TINE FIF'TER BREfIK (SEC) lTl CD (Q

ll CD FIGURE 3.23 DOWHCOHER HIXTURE I.EVEL VS TIHE FOR EOC SIIAPE

6) V fQ 0)

C3 Ll lA Ll f-o tC Cl CI CI 4

Ll

~n)

Ll C3 Ll ta Ll 70 10 20 00 50 60 70 80 90 TINE RF'TER BRERK (SECl FIGURE 3.24 EFFECTIVE REFLOOD RATE VS TINE FOR EOC SllAPE

30 40 50 60 70 80 TINE Bf TER BREAK tSECl FIGURE 3.25 COBE HIXTURE LEVEL VS TIHE FOR EOC SIIAPE

200 250 300 350 400 TINE RF'TER BREAK (SEC)

FIGURE 3.26 CONTAINMENT PRESSURE VS TIME FOR EOC SNAPE

EMF-92-1 76 Page 4-1 4.0 CONCl.uSIONS A revised LOCA/ECCS analysis has been performed to support an increased asymmetric steam generator tube plugging level of 25 ~7%, The analysis supports operation of the St. Lucie Unit 1 plant at a nominal full power of 2700 MWt with an average steam generator tube plugging level of 25%. The analysis supports a maximum LHR of 15 kW/ft that is independent of core height and exposure, It also supports a radial peaking factor of 1.75. This analysis also supports a primary coolant temperature coastdown at EOC and at full power with a maximum reduction in primary coolant temperature of 26'F, This analysis will support future cycles unless any future changes in fuel design, Technical Specifications, or plant operation indicate that a re-analysis is required.

The extended exposure LOCA/ECCS analysis performed in Reference 5 continues to support an assembly average exposure of up to 52,500 MWd/MTU, with a corresponding peak rod average exposure of 56,800 MWd/MTU. The new fuel design does not alter the conclusions of Reference 5 in that reduced fuel stored energy at Endwf-Life (EOL) dominates any adverse effect of increased rod pressure at EOL on the peak cladding temperature (PCT). In addition, third cycle fuel does not operate near the Technical Specification peaking limits which are used in the analysis.

Operation of the St. Lucie Unit 1 plant with SPC 14x14 fuel at or below the LHR limit shown in Figure 2.1 assures that the NRC acceptance criteria [10 CFR 50.46(b)] for Loss-of-Coolant Accident pipe breaks up to and including the double~nded severance of a reactor coolant pipe will be met with the emergency core cooling system for the St. Lucie Unit 1 plant.

EMF-92-1 76 Page 5-1

5.0 REFERENCES

1. St. Lucie Unit1 Revised LOCA-ECCSAnal siswith15%Steam Generator Tube Plu in Break S ectrum and Ex osure Results, XN-NF~117, Supp. 1, Exxon Nuclear Company, Richland, WA 99352, December 1985.

St. Lucie Unit1 Lar e Break LOC ECCS Anal sis SNP-91-151, Siemens Nuclear Power Corporation, Richland, WA 99352, September 1991.

3. Dennis M. Crutchfield (USNRC Asst. Director Division of PWR Licensing-B), "Safety Evaluation of Exxon Nuclear Company's Large Break ECCS Evaluation Model EXEM/PWR and Acceptance for Referencing of Related Licensing Topical Reports," dated July 8, 1986.

4. RODEX2: Fuel Rod Thermal Mechanical Res onse Evaluation Model, XN-NF-81-58(P)(A),

Revision 2, Supplements 1 and 2 dated March 1984 and Supplements 3 and 4 dated June 1990, Exxon Nuclear Company, Richland, WA 99352.

5. St. Lucie Unit1 LOC ECCS Extended Ex osureAnal sis, ANF<7-148,Advanced Nuclear Fuels Corporation, Richland, WA 99352, November 1987.

EMF-92-1 76 Issue Date:

2/5/93 ST. LUCIE UNIT 1 LARGE BREAK LOCA/ECCS ANALYSIS WITH 25+7% SGTP Distribution C. Y. Chou R. A. Copeland K. M. Duggan R. C. Gottula J. S. Holm S, E. Jensen T. R. Lindquist L A, Nielsen L D. O'Dell P, Salim S, E. Spangler B. D. Stitt R. W. Twitchell C. J. Volmer T. A. Wells R. I. Wescott FPL(11)/J. L Holm Document Control (5)

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow ENCLOSURE 3 EMF-92-148, "ST. LUCIE UNIT 1 SMALL BREAK LOCA ANALY8ZS": Siemens Power Corporation, February 5, 1993.

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