ML17228A690

SBLOCA Analysis

ML17228A690
Person / Time
Site:	Saint Lucie
Issue date:	06/06/1994
From:	Chou C SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17228A689	List: L-94-147, Forwards Rev 1 to Rept EMF-92-148, St Lucie Unit 1 SBLOCA Analysis & SBLOCA Data Relevant to Review of Util 930319 Proposed License Amend Request to Change RCS Design Flowrate.Requests That NRC Issue Subj Amend by 941115 ML17228A690
References
EMF-92-148, EMF-92-148-R01, EMF-92-148-R1, NUDOCS 9408260288
Download: ML17228A690 (55)

Text

SlE MENS EMF-92-1 48 Revision 1

Siemens Power Corporation - Nuclear Division St. Lucie Unit 1 Small Break LOCA Analysis May 1994 Siemens Power Corporation Nuclear Division 9408260288 940818 PDR ADQCK 05000335 P- -..=-

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment L-94-147 ENCLOSURE Copy of Siemens Power Corporation-Nuclear Division Report ENF-92-148, Revision 1. May 1994.

Siemens Power Coiporation - Nuclear Division EMF-92-148 Revision 1

Issue Date:

g/6/gq St. Lucie Unit 1 Small Break LOCA Analysis Prepared by:

C. Y. Chou, Engineer PWR Reload Analysis PWR Nuclear Engineering Analysis Contributor:

D. L. Caraher May 1994 lskm

CUSTOMER DISCLAIMER IMPORTANTNOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Siemens Power Corporation's warranties and representations concerning the subject matter of this document are those set forth in the Agreement between Siemens Power Corporation and the Customer pursuant to which this document is issued.

Accordingly, except as otherwise expressly provided in such Agreement, neither Siemens Power Corporation nor any person acting on its behalf makes any warranty or representation, ex-pressed or implied, with respect to the accuracy, completeness, or useful-ness of the information contained in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infringe privately owned rights; or assumes any liabilitieswith respect to the use of any information, apparatus, method or process disclosed in this document.

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

In order to avoid impairment ot the rights of Siemens Power Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use {in the patent use of the term) of such information until so authorized inwritingby Siemens Power Corporation or until six {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-92-148 Revision 1

Page ii TABLE F

NTENT

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1.0 INTRODUCTION

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2..0

SUMMARY

3.0 PLANT AND MODEL DESCRIPTIONS...................

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3-1 3.1 Plant Description 3.2 Model Description

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4.0 RESULTS AND ANALYSIS.......

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4-1 4.1 Break Spectrum Calculations 4.2 Sensitivity Calculations....

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4-1 4-2 5@0 CONCLUSIONS

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5-1

6.0 REFERENCES

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6 1

EMF-92-148 Revision 1

Page iii LI T F TABLE

~TI 3.1 St. Lucie Unit 1 SBLOCA System Analysis Parameters

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3-4 3.2 Core and Fuel Design Parameters................................

3-6 4.1 Calculated Event Times for Break Spectrum Calculations...............

4-5 4.2 Break Spectrum SBLOCA Analysis Results 4-6 4.3 Calculated Event Times for RCP Trip Delay Sensitivity Calculation.........

4-7 4.4 RCP Trip Delay Sensitivity SBLOCA Analysis Results..................

4-8 4.5 Asymmetric Tube Plugging Sensitivity SBLOCA Analysis Results..........

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EIVIF-92-148 Revision 1

Page iv LI T FFI RE

~FI ~r 4.1 Primary and Secondary System Pressures for the 0.1 ft Break..........

4-10 4.2 Break Flow Rate for the 0.1 ft Break..........

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2 4-11 4.3 Total SIT Flow Rate for the 0.1 ft Break...

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4-1 2 4 4 Total HPSI, LPSI, and Charging Pump Flow Rate for the 0.1 ft Break 4-1 3 4.5 Collapsed Core Liquid Level for the 0.1 ft Break.... ~..............

4-14 4.6 Total Primary System Mass for the 0.1 ft Break..... ~....... ~.....

4-15 4.7 Hot Rod Temperature

Response

for the 0.1 ft Break............

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4-1 6 4.8 Primary and Secondary System Pressures for the 0.15 ft Break.......

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4.9 Break Flow Rate for the 0.15 ft Break..........

4-17 4-18 4.10 Total SIT Flow Rate for the 0.15 ft Break............

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4-19 4.11 Total HPSI, LPSI, and Charging Pump Flow Rate for the 0.15 ft Break....

4-20 4.12 Collapsed Core Liquid Level for the 0.15 ft Break...............

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4-21 4.13 Total Primary System Mass for the 0.15 ft Break....

4-22 4.14 Hot Rod Temperature

Response

for the 0.15 ft Break...............

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4-23 4.15 Primary and Secondary System Pressures for the 0.2 ft Break..........

4-24 4.16 Break Flow Rate for the 0.2 ft Break...

4-25 4.17 Total SIT Flow Rate for the 0.2 ft Break 4-26 4.18 Total HPSI, LPSI, and Charging Pump Flow Rate for the 0.2 ft Break.....

4-27 4.19 Collapsed Core Liquid Level for the 0.2 ft Break.....

4.20 Total Primary System Mass for the 0.2 ft Break..

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EMF-92-148 Revision 1

Page v LI T FFI RE

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~Pe 4.21 Hot Rod Temperature Response for the 0.2 ft Break.................

4-30 4.22 Effect of Delayed RCP Trip on the Primary System Pressure for the A2 A 0.1 8 Break....

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0 4-31 4.23 Effect of Delayed RCP Trip on the Integrated Break Flow Rate for the 0.1 ft~

Breek................

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4 32 4.24 Effect of Delayed RCP Trip on the Total Primary System Mass for the 4+2 A 0.1 ft Break......................................

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4-33 4.25 Hot Rod Temperature

Response

for the 0.1 ft Break with Delayed RCP Trip 4-34

EMF-92-148 Revision 1

Page 1-'l St. Lucie Unit 1 Small Break LOCA Analysis

1.0 INTRODUCTION

This report summarizes the results of the small break loss-of-coolant accident (SBLOCA) calculations for St. Lucie Unit 1. The SBLOCA analysis was performed to support increased steam generator (SG) tube plugging up to 25%.

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

The calculated maximum fuel element cladding temperature does not exceed 2200 F.

2.

The amount of fuel element cladding which reacts chemically with water or steam does not exceed 1% of the Zircaloy within the heated length of the core.

3.

4.

The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling. The local fuel rod cladding oxidation shall not exceed 17%.

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.

The break was assumed to occur at the pump discharge section of a cold leg pipe. A cold leg pipe break is more limiting than a hot leg pipe break, based on general industry experience.

Break spectrum calculations were performed to determine the limiting break size.

The break spectrum consisted of four break sizes: 0.05 ft, 0.1 ft, 0.15 ft, and 0.2 ft

. The reactor coolant pumps (RCP) were tripped immediately following reactor scram in the base calculations.

Two sensitivity calculations were performed for the 0.1 ft break.

One to evaluate the impact of delayed trip of one RCP in each coolant loop, and one to evaluate the impact of having a 7% asymmetry in tube plugging (18'/0, 32'/0) rather than 25% plugging in each steam generator.

Section 2.0 of this report presents a summary of the SBLOCA analysis.

Section 3.0 contains brief descriptions of the St. Lucie Unit 1 plant and the analytical models.

The SBLOCA

EMF-92-1 48 Revision 1

Page 1-2 analysis results are presented in Section 4.0.

Conclusions and references are presented in Sections 5.0 and 6.0, respectively.

The content of this report {Revision 1) supersedes that in the original report. The SPC small break methodology was revised in response to NRC's questions.

In the revised methodology, the core is modeled with three radial regions, with each region having 14 axial nodes.

Thus, new break spectrum calculations and sensitivity calculations were performed to support the revised methodology and the results are summarized in this report.

EMF-92-148 Revision 1

Page 2-1 2.0

SUMMARY

The SBLOCA break spectrum calculations identified the 0.1 ft break to be the limiting break.

The peak cladding temperature (PCT) for this break was calculated to be 1846'F with a maximum local cladding oxidation of 2.25%.

The results of the delayed RCP trip sensitivity calculation are bounded by the limiting break calculation. The results of the asymmetric tube plugging calculations are also bounded by the limiting break calculation.

The analysis described in this report supports full power operation of the plant at 2754 N)Wt (2700 IVlWtplus 2% uncertainty) with an average steam generator tube plugging level of up to 25% and a maximum asymmetry of 7%, a maximum linear heat rate (LHR) of 15 kW/ft, and a radial peaking factor (Fr) of 1.75. The analysis demonstrated that the 10 CFR 50.46(b) criteria are satisfied for St. Lucia Unit 1.

EMF-92-148 Revision 1

Page 3-1 3.0 PLANT AND IVIODELDESCRIPTIONS A brief description of the St. Lucie Unit 1 plant is summarized in Section 3.1

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The corresponding ANF-RELAP model is described in Section 3.2 St. Lucie Unit 1 is a Combustion Engineering (CE) designed two-by-four loop PWR with two hot legs, four cold legs, and two vertical U-tube steam generators.

The reactor has a rated core power of 2700 MWt.

The reactor vessel contains a downcomer, upper and lower plenums, and a reactor core containing 217 fuel assemblies.

The hot legs connect the reactor vessel with the vertical U-tube steam generators.

Feedwater is injected into the downcomer of each steam generator.

There are three auxiliary feedwater pumps, two motor driven and one turbine driven. The emergency core cooling system (ECCS) contains two high pressure safety injection (HPSI) pumps, four safety injection tanks (SIT), and two low pressure safety injection {LPSI) pumps.

The SBLOCA evaluation model (Ref.

1 and

2) consists of three computer codes.

The appropriate conservatism, prescribed by Appendix K of 10 CFR 50, were incorporated.

The RODEX2 code was used to determine the initial fuel stored energy and gap conditions for the initialization of the system blowdown and hot rod response calculations.

2.

The SPC version of RELAP5/MOD2 (ANF-RELAP) was used to model the primary system and secondary side of the steam generators during the blowdown.

The governing conservation equations for mass,

energy, and momentum transfer are used along with appropriate correlations consistent with Appendix K of 10 CFR 50.

3.

The TOODEE2 code was used to simulate the behavior of the hot rod during the entire transient.

The code uses nodal fluid flow rate, steam temperatures

EMF-92-148 Revision 1

Page 3-2 above the mixture level, and mixture level boundary conditions from the ANF-RELAP system calculation.

The ANF-RELAP model of St. Lucie Unit 1 includes four safety injection tanks (SITs), a pressurizer, and two steam generators with both primary and secondary sides modeled.

All four cold legs were modeled individually. A steam generator tube plugging level of 25% was assumed.

The HPSI and LPSI pumps were modeled as time dependent junctions at the SIT lines, with flows given as a function of pressure.

The primary coolant pump performance curves were plant-specific curves for St. Lucie Unit 1.

The reactor core was modeled with three radial regions.

One region represents the hot assembly.

The power in this region corresponds to the power at the peaking limits. A second region represents the central 30% of the core less the hot assembly.

This region typically operates at about 1.2 times the average core power. A third region represents the remaining 70/o of the core.

In addition, each radial region consists of 14 uneven axial nodes.

A finer axial noding scheme in the core upper region was utilized, providing adequate detail of the boundary conditions which were transferred to the TOODEE2 code. The heat generation rate in the ANF-RELAP reactor core model was determined with actinide and decay heating as prescribed by Appendix K.

Single failure criterion was satisfied by assuming the toss of one diesel generator, which resulted in the disabling of one HPSI pump and one motor driven auxiliary feedwater pump.

initiation of the MPSI system was delayed 30 seconds beyond the time of safety injection actuation signal (SIAS). The 30-second delay represents the time required for diesel generator startup and switching. The disabling of a motor driven auxiliary feedwater pump would leave one motor driven pump and the turbine driven pump available.

The motor driven pump actuation setpoint is based on the narrow range steam generator level. The pump is turned on when the steam generator level reaches 18 /o. The turbine-driven pump is actuated at 600 seconds after break initiation.

EMF-92-148 Revision 1

Page 3-3 In addition to the HPSI pump, credit was also taken for a charging pump.

Specifically, 40 gpm of water from a charging pump was included. All of the charging flow was assumed to go to the broken loop, with one half of the flow going to the intact cold leg and the remaining half going to the broken cold leg.

In the pump trip sensitivity calculation, two of the RCPs, one in each coolant loop, were tripped at 120 seconds after reactor scram and the remaining two RCPs were tripped at the time of minimum primary inventory.

This assumption is consistent with the Emergency Operating Procedures for St. Lucia Unit 1.

In the asymmetric tube plugging sensitivity calculation the steam generator in the intact loop was assumed to have a tube plugging level of 18% and the steam generator in the broken loop was assumed to have a tube plugging level of 32%

In this analysis, the core average axial power distribution at EOC was used (peaked highest in the core at a relative height of 0.85). Axial shapes at 100% power with ASls from 0.0 to

-0.2 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 axial power shapes in the St. Lucie Unit 1 plant.

Table 3.1 contains a summary of the system parameters and Table 3.2 contains the core and fuel design parameters used in the SBLOCA analysis.

Only those parameters which are specifically required by the SPC SBLOCA methodology or Appendix K of 10CFR50 are biased in a conservative direction.

All other parameters are the nominal values.

EMF-92-1 48 Revision 1

Table 3.1 St. Lucia Unit 1 SBLOCA System Analysis Parameters Primary Heat Output, MWt Primary Coolant Flow, gpm Primary Coolant System Volume, ft Operating Pressure, psia Inlet Coolant Temperature, 'F Reactor Vessel Volume, ft Pressurizer Total Volume, fts SIT Volume, ft3 (each of four)

SIT Liquid Volume, ft SIT Pressure, psia SIT Fluid Temperature, F

Total Number of Tubes per Steam Generator Number of Tubes Plugged per Steam Generator Secondary Flow Rate / Steam Generator, Ibm/hr Steam Generator Secondary Pressure, psia Steam Generator Feedwater Temperature, 'F Steam Generator Safety Relief Valve Flow Rate, Ib/hr Reactor Coolant Pump Rated Head, ft Reactor Coolant Pump Rated Torque, ft-tbf Reactor Coolant Pump Rated Speed, rpm Initial Reactor Coolant Pump Speed, rpm Reactor Coolant Pump Moment of Inertia, Ibm-ft Sl Fluid Temperature, 'F Reactor Scram Low Pressure Setpoint, psia SIAS Activation Setpoint Pressure, psia HPSI Pump Delay Time on SIAS, sec 2700'55,000 10433 2250 549 45zz 1500 2020 1090 21 4.7 110

.8485 2121 t25 /o) 6.00e6 770 435 5.955e63 272 32,495 886 1063.24 101,900 100 1887~

1600'0.0

EMF-92-148 Revision 1

Page 3-5 Table 3.1 St. Lucie Unit 1 SBLOCA System Analysis Parameters (Continued)

HP I

n LP IFI wR Vr R

Pr r

RCS Pressure (psia) 1129.0 1125.0 1115.0 1015.0 81 5.0 615.0 315.0 1 65.6 159.9 135.5 92.4 28.7 14.7 Total HPSI Flow (Ibm/sec) 0.0 3.59 9.12 28.73 47.79 61.05 76.66 82.98 83.22 84.26 86.06 88.76 89.36 Total LPSI Flow (Ibm/sec) 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 132.60 270.73 408.86 546.99 546.99 Primary heat output used in ANF-RELAP model 1.02 x 2700 = 2754 MWt.

Includes pressurizer total volume and 25% SG tube plugging.

There are a total of eight steam generator safety relief valves, four of which open at a setpoint pressure and the other four open at a slightly higher pressure.

The valves are assumed to close at the same pressures as the opening pressures.

The rated flow corresponds to the flow at the highest setpoint pressure to minimize the flow out the valves.

Value used in ANF-RELAP for initialization.

These are nominal setpoint values.

Biasing of these parameters is not required by the SPC SBLOCA methodology and does not significantly affect the results of the transient..

1/4 of the flow is distributed to each of the cold legs.

With the failure of one diesel generator, only two low pressure injection valves willbe opened for LPSI flow. The above LPSI flow assumed all four valves were open.

This assumption,".as no im~act on the results since the LPSI pumps did not actuate in the limiting break (0.01 ft ) and in the larger breaks, LPSI flow came on after the time of PCT.

EMF-92-1 48 Revision 1

Page 3-6 Table 3.2 Core and Fuel Design Parameters Cladding O.D., in.

Cladding I.D., in.

Fuel O.D., in.

Fuel rod pitch, in.

Fuel assembly pitch, in.

Active fuel length, in.

Core flow area, ft Core bypass flow, /o 0.440 0.384 0.377 0.580 8.180 136.7 53.19 3.90

EMF-92-148 Revision 1

Page 4-1 4.0 RESULTS AND ANALySIS 4.1 Br I<

rum I

la i n SBLOCA break spectrum calculations were performed for break sizes of 0.05 ft, 0.1 ft, 0.15 ft, and 0.2 ft. The calculated event times for each break are summarized in Table 4.1.

The results from TOODEE2 hot rod response calculations are presented in Table 4.2.

Figures 4,'I through 4.7 summarize the key results for the 0.1 ft break calculation since this break produced the highest PCT in the break spectrum calculations.

The primary and secondary pressure responses are shown in Figure 4.1.

The primary pressure decreased immediately after breakinitiation. Reactor scram occurred when the primary pressure reached 1887 psia.

The secondary pressure increased rapidly after break initiation as the reactor scrammed and steam generator isolation took place.

The secondary pressure continued to increase until the steam generator safety valves opened, causing the secondary pressure to stabilize.

At approximately 280 seconds, liquid was expelled from the loop seal piping, allowing steam to flow directly to the break, which caused the primary pressure to decrease more rapidly. The primary pressure increased slightly around 1230 seconds.

The addition of SIT fluid into the core increased the steaming rate and hence raised the primary pressure.

The break flow rate, shown in Figure 4.2, followed the same trend as the primary system pressure for the first 280 seconds of the transient.

The loop seals cleared at approximately 280 seconds, at which time the flow out the break transitioned from liquid to steam, causing a rapid decrease in the break flow rate.

The minor oscillations around 1230 seconds were caused by the initiation of SIT flow, shown in Figure 4.3. At 1400 seconds, the break flow rate was only slightly larger than the total HPSI flow, shown in Figure 4.4.

For St. Lucia Unit 1, the setpoint for SIT initiation is 214.7 psia.

Figure 4.3 shows that the SIT flow initiated at 1224 seconds and terminated at 1255 seconds, supplying a large burst of cooling water.

The smaller HPSI system initiated flow at approximately 60 seconds, as

l EMF-92-148 Revision 1

Page 4-2 illustrated in Figure 4 4, delivering cooling water to the primary system during most of the transient.

The collapsed core liquid level is shown in Figure 4.5.

The level dropped immediately after the break and increased slightly around 280 seconds as a result of loop seal clearing, which forced more liquid into the reactor vessel.

The core level then continued to decrease until 1224 seconds, at which time the SIT flow initiated. The addition of SIT flow increased the liquid inventory in the reactor vessel. The total primary system mass, illustrated in Figure 4.6, showed a gradual decreasing trend until 1224 seconds.

The additional inventory at 1224 seconds was from SIT flow.

The PCT for the 0.1 ft break calculation was 1846 F with a maximum local cladding oxidation of 2.25%.

The PCT occurred at 1230 seconds.

Rod burst was predicted in this case.

The hot rod temperature response from the TOODEE2 calculation is shown in Figure 4.7.

The 0.15 ft and 0.2 ft break calculations displayed similar trends as the 0.1 ft break 2

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2 calculation and are therefore, not discussed here. The responses of the 0.15 ft and 0.2 ft 2

2 break calculations are shown in Figures 4.8 through 4.21.

The PCTs for these two breaks were 1610'F and 13004F, respectively.

For the 0.05 ft~ break, the system pressure was too high to allow SIT flow. However, due to the relatively small break size, the HPSI flow alone was enough to provide core cooling.

Core uncovery did not occur for this break size and therefore, there was no hot rod heatup.

4.2 n i ivi I

I i n A sensitivity calculation was performed to investigate the effect of delayed RCP trip on the PCT. To be consistent with the Emergency Operating Procedures for St. Lucie Unit 1, two RCPs were tripped early after reactor scram and the remaining two RCPs were tripped at the time of minimum primary inventory.

The early trip time was assumed to be 120 seconds.

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EMF-92-148 Revision 1.

Page 4-3 Pump trip in the broken cold leg of the broken loop and one of the legs in the intact loop was assumed to be delayed.

The calculated event times for the delayed RCP trip sensitivity calculation are summarized in Table 4.3.

The results from the corresponding TOODEE2 hot rod response calculations are presented in Table 4.4.

The effect of delayed RCP trip on the primary system pressure for the 0.1 ft break is shown in Figure 4.22.

Figure 4.22 shows that continued operation of two RCPs (delayed RCP trip case) resulted in a slower average rate of depressurization.

This trend was consistent throughout the transient.

The effect of RCP trip on the integrated break flow rate is shown in Figure 4.23. The delayed RCP trip case had a slightly lower break flow rate in the first 600 seconds and a slightly higher break flow rate after 600 seconds.

The effect of RCP trip on the total primary system mass is shown in Figure 4.24.

In the first 600 seconds, the primary system mass for the delayed trip case was higher than the early trip case.

After 600 seconds, the primary system mass for the delayed trip case dropped below the early trip case.

In both cases the primary system mass increased with the onset of SIT flow.

The PCT for the 0.1 ft break with delayed RCP trip was 1546 F with a maximum local cladding oxidation of 0.41%.

The hot rod temperature response from the TOODEE2 calculation is shown in Figure 4.25.

The PCT for the 0.1 ft~ break with early RCP trip was 1846'F. Thus, the overall effect of delayed RCP trip was a lower calculated PCT for the 0.1 ft2 break.

A sensitivity calculation was made to investigate the effect of an asymmetric distribution of plugged tubes.

Instead of assuming 25/o plugged tubes in each steam generator, the intact steam generator was assumed to have 18% of its tubes plugged and the broken loop steam generator was assumed to have 32% of its tubes plugged.

The calculated event times for the asymmetric plugging, sensitivity calculation were very similar to those presented in Table 4,1 for the symmetric plugging case.

The results from the

EMF-92-148 Revision 1

Page 4-4 TOODEE2 hot rod response calculation is presented in Table 4.5.

The TOODEE2 hot rod response calculation for the asymmetric plugging case yielded a PCT of 1811 'F with a maximum local cladding oxidation of 2.14%. The PCT for the 0.1 ft break with symmetric tube plugging was 18464F. Thus, the overall effect of a 7% asymmetry in tube plugging was a slightly lower calculated PCT for the 0.1 ft break.

EMF-92-148 Revision 1

Page 4-5 Table 4.1 Calculated Event Times for Break Spectrum Calculations EVENT TIIVIE Br akSiz KQM<2 9 >~<

9 ~'~t Break initiation 0.0 0.0 0.0 0.0 Reactor trip/ RCP trip SIAS + 30 sec delay 19.9 70.

11.1 8.6 59.

54.

7.6 52.

HPSI initiation Motor driven aux. feed Turbine driven aux. feed 70.

59.

54.

328. '20.

317.

600.

600.

600.

52.

31 6.

600.

Loop seal clearing (Intact loop, cold leg 1)

NA 280.

180.

140.

Loop seal clearing (Intact loop, cold leg 2)

Loop seal clearing (Broken loop, intact leg)

NA NA NA NA NA 140.

NA Loop seal clearing (Broken loop, broken leg) 600.

280.

180.

140.

Break uncovered Minimum primary system mass SIT flow initiation Time of PCT SIT flow termination End of calculation 600.

280.

180.

140.

1450.

1 224.

690.

51 5.

NA 1 224.

690.

51 5.

NA 1230.

698.

520.

NA 1 255 720.

600, 2337.

1400.

1000.

800.

EMF-92-1 48 Revision 1

Page 4-6 Table 4.2 Break Spectrum SBLOCA Analysis Results Hot Rod Burst Bre k iz gg~~2 ~~2 g

q s; g~2 ~pg 2

- Time (sec)

- Elevation (ft)

- Channel Blockage Fraction NA'A NA 1108.

NA 9.97 NA 0.46 NA NA NA NA Peak Clad Temperature

- Temperature ('F)

- Time (sec)

- Elevation (ft)

NA NA NA 1 846.

1 610.

1230.

698.

10.47 9.97 1300.

520.

9.97 lVletal-Water Reaction

- Local Maximum (%)

- Elevation of Local Max. (ft)

- Hot Pin Total (%)

- Core Maximum NA NA NA NA 2.25 OA3 9.97 9.97 0.29 0.08

<1%

<1%

0.05 9.97 0.01

<1%

Not applicable since no hot rod heatup was predicted.

EMF-92-148 Revision 1

Page 4-7 Table 4.3 Calculated Event Times for RCP Trip Delay Sensitivity Calculation TIME Break initiation Reactor trip SIAS + 30 sec delay HPSI initiation RCP trip (2 early)

Loop seal clearing (Intact loop, cold Ieg 1)

Loop seal clearing (Intact loop, cold leg 2)

Loop seal clearing (Broken loop, intact leg)

Loop seal clearing (Broken loop, broken leg)

Break uncovered Motor driven aux. feed Turbine driven aux. feed Minimum primary system mass RCP trip (2 delayed)

SIT flow initiation Time of PCT SIT flow termination End of calculation 0.0 49.3 50.0

'I 31.

320.

320.

320.

320.

320.

320.

600.

1 600.

'I 600.

1 600.

1617.

1690.

2000.

EMF-92-148 Revision 1

Page 4-8 Table 4 4 RCP Trip Delay Sensitivity SBLOCA Analysis Results Hot Rod Burst NA Peak Clad Temperature

- Temperature ('F)

- Time (sec)

- Elevation (ft) 1546.

1617.

10.47 Metal-Water Reaction

- Local Maximum (%)

- Elevation of Local Max. (ft)

- Hot Pin Total (%)

- Core Maximum 0.41 10.47 0.03 1%

EMF-92-148 Revision 1

Page 4-9 Table 4.5 Asymmetric Tube Plugging Sensitivity SBLOCA Analysis Results Hot Rod Burst

- Time (sec)

- Elevation (ft)

- Channel Blockage Fraction 1159.

9.97 0.46 Peak Clad Temperature

- Temperature ('F)

- Time (sec)

- Elevation (ft) 1811.

1279.

10 47 Metal-Water Reaction

- Local Maximum (%)

- Elevation of Local Max. (ft)

- Hot Pin Total (%)

- Core Maximum 2.14 9.97 0.28

<1%

EMF-92-148 Revision 1

Page 4-10

~

J Ce 2400.0 2000.0 ~ Primary

~

~ Secondary C4 1600.0 Q

1200.0 47 6oo.o 0

400.0 0.0 0.0 200.0 400.0 600.0 600.0 1000.0 1200.0 1400.0 1600

~ 0 Time (sec)

Figure 4.1 Primary and Secondary System Pressures for the 0.1 ft Break

EMF-92-148 Revision 1

Page 4-11 2400.0 2000.0 0

1600.0 8

Cl e$

1200.0 F4 0

600.0 400.0 0.0 0.0 200.0 400.0 600.0 600.0 1000.0 1200.0 1400.0 1600.0 Time (sec)

Figure 4.2 Break Flow Rate for the 0.1 ft Break

EMF-92-1 48 Revision 1

Page 4-12 1200.0 1000.0 O

800.0 (5

800.0 0

400.0 200.0 0.0 0.0 200.0 400,0 600.0 800.0 1000.0 1200.0 1400.0 1800.0 T1me (sec)

Figure 4.3 Total SIT Flow Rate for the 0.1 ft Break

EMF-92-1 48 Revision 1

Page 4-13 87.5 76.0 O

~N 82.5 8

6o.o C$

37.5 0

Ca 25.0 Ci 0.0 0.0 200

~ 0

~

400.0 800.0 800.0 1000.0 1200.0 1400.0 1800.0 Time (sec)

Figure 4A Total HPSI, LPSI, and Charging Pump Flow Rate for the 0.1 ft Break

EMF-92-1 48 Revision 1

Page 4-14 12.0 10.0 8.0 ID Cl 0.0 2.0 0.0 200.0 400.0 SOO.O

$ 00.0 1000,0 1200.0 1400.0 1800.0 Time (sec)

Figure 4.5 Collapsed Core Liquid Level for the 0.1 ft Break

~,)

L

~8,I EMF-92-148 Revision 1

Page 4-15 co 45.0 40.0 35.0 30.0 8

c$

25.0 20.0 15.0 10.0 0.0 200.0 400.0 800.0 800.0 1000.0 1200.0 1400 0 1800 0 Time (sec)

Figure 4.6 Total Primary System Mass for the 0.1 ft Break

EMF-92-148 Revision 1

Page 4-16 tCI WXE (MXX 23 IC 10 RT rI t PCI 1%5 9 KCfs 1 2o RLFILSfD ICXE Ill 21 lC 9,91 rI,l 150.0 300.0 450.0 QM,O 250.0 9M.O 1050.0 1200.0 1350.0 1500.0 TIDE - SECONOS Figure 4.7 Hot Rod Temperature Response for the 0.1 ft Break

EMF-92-148 Revision 1

Page 4-17 2400.0 2000.0 a

C4 1800.0 0

eN 1200 0 P

0 000.0

~~ ~~

400.0 0.0 0.0 100 ~ 0 200.0 SOO.O 400.0 500.0 000.0 700.0 Soo.o 900.0 1000

~ 0 Time (sec)

Figure 4.8 Primary and Secondary System Pressures for the 0.15 ft Break

EMF-92-148 Revision 1

Page 4-18 3600.0 3000.0 2600.0 8o 2OOO.O j

1600. 0 0

1000.0 600.0 0.0 0.0 100.0 200.0 300.0 400.0 600.0 600,0 700.0 800.0 900.0 1000.0 Time (sec)

Figure 4.9 Break Flnw Rate for the 0.15 ft Break

~

ii 4

I ~I EIVIF-92-1 48 Revision 1

Page 4-19 2800.0 2400.0

~N 2000.0 8

1800.0 1200.0 0

K N

800.0 400.0 0.0 0.0 100.0 200.0 300.0 400.0 500.0 800.0 700.0 800.0 900.0 1000.0 Time (sec)

Figure 4.10 Total SIT Flow Rate for the 0.15 ft Break

EMF-92-148 Revision 1

Page 4-20 120. 0 100.0 OI 80.0 0

a 8O.O 0

40.0 N

Ct 20.0 0.0 0.0 100,0 200.0 300.0 400.0 600.0 800.0

'loo ~ 0 800.0 000.0 1000.0 Time (sec)

Figure 4.11 Total HPSI, LPSI, and Charging Pump Flow Rate for the 0.15 ft Break

EMF-92-148 Revision 1

Page 4-21 12.0 10.0 8.0 Q

Q 8 0 4.0 2.0 0.0 100.0 200.0 300.0 400.0 500.0 800.0 700.0 800.0 000.0 1000.0 Time (sec)

Figure 4.12 Collapsed Core Liquid Level for the 0.15 ft Break

EMF-92-148 Revision 1

Page 4-22 "h

6.0 i.o 3.O 8

C$

2.0 1.0 0.0 0.0 100

~ 0 200.0 300.0 400.0 500.0 500.0 700 0 800

~ 0 900.0 1000

~ 0 Time (sec)

Figure 4.13 Totai Primary System Mass for the 0.'t5 ft Break

EMF-92-148 Revision 1

Page 4-23 I ~

PCI NXX:

ll6XX: 2l ill 9.97 PIo p PCI ~

l609o6 0CCf'o) l00.0 200.0 300.0 4$.0 500,0 600.0 7Ã.0 800.0 900.0 l000.0 TINE - SECONOS Figure 4.14 Hot Rod Temperature Response for the 0.15 ft Break

EMF-92-148 Revision 1

Page 4-24 2400.0 2000.0 4

~H a.

18OO.O 4

N Cl 1200.0 4

0 800.0 8

0 400.0 0.0 0.0 100.0 200.0 300.0 400.0 500.0 800.0 700.0 800.0 Time (sec)

Figure 4.15 Primary and Secondary System Pressures for the 0.2 ft Break

EMF-92-148 Revision 1

Page 4-25 4000.0 0

3000.0 N

8 4

2000.0 0

1000.0 0.0 0 ~ 0 100.0 200.0 300.0 400.0 500.0 000.0 700.0 800.0 Time (sec)

Figure 4.16 Break Flow Rate for the 0.2 ft Break

EMF-92-148 Revision 1

Page 4-26 2400.0 2000.0 O

i800.0 8

0 1200.0 CC 0

800.0 400.0

,0 0.0 100.0 200.0 300.0 400.0 500.0 800.0 700.0 800.0 Time {sec)

Figure 4.17 Total SIT Flow Rate for the 0.2 ft Break

la

~

t

~

EMF-92-148 Revision 1

Page 4-27 280.0 240.0 N

200.0 8

180.0 I5 120.0 0

Q N

80.0 40.0 0.0 0.0 100.0 200.0 800.0 400.0 500.0 800

~ 0 700.0 800.0 Time (sec)

Figure 4.18 Total HPSI, LPSI, and Charging Pump Flow Rate for the 0.2 ft Break

EMF-92-148 Revision 1

Page 4-28 12.0 10.0 8.0 I

Q 8.0 4.0 2.0 0.0 100.0 200.0 300.0 400.0 500.0 800.0 700.0 800.0 Time (sec)

Figure 4.'19 Collapsed Core Liquid Level for the 0.2 ft Break

EMF-92-148 Revision 1

Page 4-29 4.0 0

8 2.0 1.0 0.0 O.O

<00.0 2OO.O 000.0 iOO.O SOO.O epp.p VOO.O SOO.O Time (sec)

Figure 4.20 Total Primary System Mass for the 0.2 ft Break

FMF-92-148 Revision 1

Page 4-30 PCT NXK WXE 2l hT 9.07 FZ., KT 1%0.2 OCCF.I 4

Q CI wZ 8

H>>

~L L.o oo.o

>so.o ao.o m.o ee.o coo.o seo.o sa.o no.o ooo.o TING " SECONOS Figure 4.21 Hot Rod Temperature Response for the 0.2 ft Break

EMF-92-148 Revision 1

Page 4-31 2400

~ 0 2000.0 ~ Early RCP Trip

~

~ Delayed RCP Trip 18OO.O 1200.0 C4 8

800.0 400.0 0.0 0.0 200.0 400.0 800.0 800.0 1000.0 1200.0 1400.0 1800.0 1800.0 Time (sec)

Figure 4.22 Effect of Delayed RCP Trip on the Primary System Pressure for the 0.1 ft Break

EMF-92-148 Revision 1

Page 4-32 6.0 a

p Early RCP Trip

~

~ Delayed RCP Trip 4.0 0

~N 6

8.0 Cl D4 2.0 rEa C$

1.0 0.0 0.0 200.0 400.0 600.0 800.0 1000.0 1200.0 1400.0 1600.0 1800.0 Time (sec)

Figure 4.23 Effect of Delayed RCP Trip on the integrated Break Flow Rate for the 0.1 ft~ Break

EMF-92-148 Revision 1

Page 4-33 6.0 ~ Early RCP Trip

~

~ Delayed RCP Trip 4.0 3.0 8

cl 2.0 1.0

~

~

~

0.0 0.0 200.0 400.0 800.0 800.0 1000.0 1200.0 1400.0 1800.0 1800.0 Time (sec)

Figure 4.24 Effect of Delayed RCP Trip on the Total Primary System Mass for the 0.1 ft Break

EMF-92-148 Revision 1,

Page 4-34 r a FCl lOOC wm zr er ro.~r rr., vcr -

>s~s.e mr.)

glI ID I

mX

~R d

.0 200.0 CO,O 600.0 800.0 re.0 1 200.0 l%0.0 1600.0 1800.0 IXI+0 TIt1C - SECONOS Figure 4.25 Hot Rod Temperature Response for the 0.1 ft Break with Delayed RCP Trip

EMF-92-1 48 Revision 1

Page 5-1

5.0 CONCLUSION

S The SBLOCA analysis for St. Lucie Unit 1 identified the 0.1 ft~ break to be the limiting break.

The analysis supports operation of St. Lucia Unit 1 at a nominal power level of 2700 MWtand steam generator tube plugging of up to 25% with a maximum asymmetry of 7%.

The analysis supports a peak LHR of 15 kW/ft and a radial peaking factor of 1.75.

Operation of St. Lucie Unit 1 with SPC 14x14 fuel within the limits stated above assures that the NRC acceptance criteria for SBLOCA (10 CFR 50.46(b)) will be met with the existing emergency core cooling system.

EMF-92-148 Remsion 1

. Page 6-1

6.0 REFERENCES

2.

"Exxon Nuclear Company Evaluation acr map iSqz

<l<l~

N-NF-

-4 "Exxon Nuclear Company Evaluation Model-EXEM PWR Small Break Model," April 1989; F-Model-EXEM PWR Small Break Model,"

EMF-92-148 Revision 1

Issue Date: 6/6/94 St. Lucle Unit 1 Small Break LOCA Analysis

~DI frit~in

~Ri hland C. Y. Chou H. Chow T. P. Currie R. A. Copeland K. M. Duggan J. C. Hibbard R. C. Gottula S. E. Jensen P. Salim B. D. Stitt C. J. Volmer Document Control (2)

~BII a~ve R. I. Wescott FPL (4)

A rp 7i f1 P

4 fP ty