ML17227A789
| ML17227A789 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/28/1993 |
| From: | Chou C SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17227A785 | List: |
| References | |
| EMF-92-148, NUDOCS 9303310069 | |
| Download: ML17227A789 (64) | |
Text
SIEMENS EMF-92-148 Siemens Power Corporation St. Lucie Unit 1 Small Break LOCA Analysis February 1993 it[tp Siemens Power Corporation 5'U '.".'"".f "'",talon 93033100b9'303i9 PDR ADOCK 05000335 P
1 g
Siemens Power Corporation EMF-92-148 Issue Date:
p 93 ST. LUCIE UNIT 1 SMALLBREAK LOCA ANALYSIS Prepared by C. Y. Cho, Engineer PWR Reload. Analysis PWR Nuclear Engineering February 1993
Siemens Power Corporatkxt's warranties and representations concerning the subject mater of this document are those set fonh kt the Agreement between SIemens Power Collxxadon and the Customer pwsuant to which this document Is issued Accordktgfy, except as othelwlae expfeaaiy provided kl Such Agreement, neither Siemens Power Corporation nor any person acing on its behalf makes any warranty or representation, expressed or Implied, with respect to the accuracy, completeness, or usefLNness of the information contained in this document, or that the use of any Information, apparatus, method or process dfsdosed In this document wINnot infringe privately owned rights; or assumes any IIabiNIfeswith respect to the use of any Information, apparatus, method or process disclosed In this document.
Tllo klformatlon contained herein Is fofthe sole use of the Customer In order to avoid impairment of rights of SIemens Power Corporation in patents or Inventkes which may be included in the information contained in this doctsnent. the recipient. by its acceptance ofthis document. agrees not to publish or make pubic use (in the patent use of the term) of such information until so authonxed in wntktg by Siemens Power Corporation or until after six (6) months foaowktg termination or expfra5on of the aforesaid Agreement and any extension thereof, urges 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 46
- Pagei TABLE OF CONTENTS Section 1.0 iNTRODUCTiON 2.0
SUMMARY
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Pacae 1-1 2-1 3.0 PLANT AND MODEL DESCRIPT1ONS 3.1 Plant Description 3.2 Model Description 4.0 RESULTS AND ANALYSlS 4.1 Break Spectrum Calculations 4.2 Sensitivity Calculations 5.0 CONCLUSlONS
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3-1 3-1 3-1 4-1 4-1 4-2 5-1 6IO REFERENCES o
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6-1
EMF-92-148 Page ii LIST OF TABLES Table 3.1 St. Lucie Unit 1 SBLOCA System Analysis Parameters 3.2 Core and "Fuel Design Parameters 4.1 Calculated Event Times for Break Spectrum Caiculations..
4.2 Break Spectrum SBLOCA Results 4.3 Calculated Event Times for RCP Trip Delay Sensitivity Calculations..
4.4 RCP Trip Delay Sensitivity SBLOCA Results Pacae 3-4 3-6 4-4 4
5 4-6
EMF.92-1 48 Page iii LIST OF FIGURES
~Fi ure 4,1 Primary and secondary system pressures for the 0.1 ft break....
Pacae 4-8 4.5 Total HPSI flow rate for the 0.1 ft break Collapsed core liquid level for the 0.1 ft break 4.2 Break flow rate for the 0.1 ft break 4.3 Total SIT flow rate for the 0,1 ft break......,
4-9 4-10 4-11 4-12 4.6 4.7 4.8 4.9 Total primary system mass for the 0.1 ft break Hot rod temperature response for the 0.1 ft break Primary and secondly system pressures for the 0,05 ft break Break flow rate for the 0.05 ft break................
4-13 4-14 4-15 4-16 4.10 Total SIT flow rate for the 0.05 ft break..
4.11 Total HPSI flow rate for the 0.05 ft break 4.12 Collapsed core liquid level for the 0.05 ft break 4.13 4.14 Total primary system mass for the 0.05 ft break Hot rod temperature response for the 0,05 ft break 4.16 4.17 4.18 Break flow rate for the 0.1 5 ft break Total SIT flow rate for the 0.1 5 ft break Total HPSI flow rate for the 0.15 ft break 4.19 Collapsed core liquid level for the 0,15 ft break 4.20 Total primary system mass for the 0.15 ft break 4.15 Primary and secondary system pressures for the 0,15 ft break 4-17 4-18 4-19 4-20 4-21 4-23 4-24 4-25 4-26 4-27
EMF-92-1 48 Page iv UST OF FlGURES
~Fi ure 4.21 Pacae Hot rod temperature response for the 0.15 ft break...................
4-28 4.22 4-29 4-30 4-31 4.23 Break flow rate for the 04 ft break 4.24 Total SIT flow rate for the 0.2 ft break 4.25 Total HPSI flow rate for the 0.2 ft break
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4-32 Primary and secondary system pressures for the 0.2 ft break 4.26 4.27 Collapsed core liquid level for the 0,2 ft break Total primary system mass for the 0,2 ft break.....,,
4-33 4-34 4.28 4.29 4.30 4.31 Hot rod temperature response for the 0.2 ft break.....................
4 - 35 t
Effect of delayed RCP trip on the primary system pressure for the 0.1 ft break 4 - 36 Effect of delayed RCP trip on the break flow rate for the 0.1 ft break......
4-37 Effect of delayed RCP trip on the total primary system mass for the 0.1 ft break 4 - 38 4.32 Hot rod temperature response for the 0,1 ft break with delayed RCP trip 4-39 4.34 4.35 4,36 Effect of delayed RCP trip on the primary system pressure for the 0.15 ft break 4 - 40 Effect of delayed RCP trip on the break flow rate for the 0.15 ft break.....
4-41 Effect of delayed RCP trip on the total primary system mass for the 0.15 ft break4 - 42 Hot rod temperature response for the 0.15 ft break with delayed RCP trip...
4 - 43
EMF-92-1 46 Page 1 -
1 1.0 INTROOUCTION This report summarizes the results of the small break losswf~oolant accident (SBLOCA) calculaUons for St. Lucie Unit 1. The SBLOCA analysis was performed to support the changes in the Technical SpecNcaUons due to the increased steam generator (SG) tube plugging from 15% to 25%. Tba 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.
The cladding temperature transient is terminated at a time when the core geometry is sal amenable to cooling. The local fuel rod cladding oxidation shall not exceed 17%.
4, The core temperature is reduced and decay heat is removed for an extended period of time, as required by the Iongdived radioacUvity 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 teg 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.
In addition, sensithrity calculaUons were performed for the 0.1 ft and 0.15 ft breaks to evaluate the impact of delayed trip of one RCP in each coolant loop.
Section 20 ofthis report presents a summary ofthe SBLOCA analysis.
SecUon 3.0 contains brief descriptions ofthe St. Lucie Unit 1 plant and the analytical models. The SBLOCA analysis results are presented In SecUon 4.0. Conclusions and references are presented in SecUons 5.0 and 6,0, respecUvely.
EMF-92-148 Page 2-1 2.0
SUMMARY
The SBLOCA break spectrum calculations identNed the 0.1 ft break to be the limiting break.
The peak cladding temperature (PCT) forthis break was calculated to be 1672 F with a maximum local cladding oxidation of 1.65%. The results of the delayed RCP trip sensitivity calculations are bounded by the break spectrum calculations.
The analysis described in this report supports full power operation of the plant at 2754 MWt (2700 MWt plus 2% uncertainty) with an average steam generator tube plugging level of up to 25% and a maximum asymmetry of 7%. The steam generator tube plugging asymmetry was not specifically evaluated in this report.
Previous asymmetric sensitivity analyses for a similar plant indicate that the results from asymmetric tube plugging conditions would be bounded by the results from the symmetric tube plugging conditions. Furthermore, since the total primary system mass inventory ls the dominant variable in SBLOCA, the small variation in mass distribution between the two loops willnot have a signNcant impact on the PCT. The analysis also supports 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. Lucie Unit 1
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EMF-92-1 48 Page 3-1 3.0 PLANT AND MODEL DESCRIPTIONS
-1 A brief description of the St. Lucie Unit 1 plant is summarized in Section 3.1. The corresponding ANF-RELAP model is described in Section 3.2 St. Lucie Unit 1 is a Combustion Engineering (CE) designed two'-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 three high pressure safety injection (HPSI) pumps, four safety injection tanks (SIT), and two low pressure safety injection (LPSI) pumps.
3.2 Model Descri tion The SBLOCA evaluation model (Ref. 1) consists of three computer codes.
The appropriate conservatism, prescribed by Appendix K of 10 CFR 50, were incorporated.
The RODEX2 code (Ref. 2) was used to determine the initial fuel stored energy and gap conditions for the initialization of the system blowdown and hot rod response calculations.
The SPC version of RELAP5/MOD2 (ANF-RELAP) (Ref. 3) was used to model the primary system and secondary side ofthe 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 TQODEE2 code was used to simulate the behavior of the hot rod during the entire transient.
The code uses nodal fluid flow rate, steam temperatures above the mixture level, and mixture level boundary conditions from thn ANj'- Ill,".LOP system calculation.
EMF-92-1 48 Page 3-2 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. The two cold legs in the intact loop were lumped into a single loop. The intact and broken cold legs in the broken loop 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-specNc curves for St. Lucie Unit 1. These curves are consistent with those in the Cycle 12 Groundrules (Ref. 4).
The reactor core was modeled as a single radial region with 15 axial nodes.
A finer mesh in.the core upper region was uUlized, providing adequate detail of the boundary condiUons which were transferred to the TOODEE2 code.
The heat generation rate in the ANF-RELAP reactor core model was determined from reactor kinetics equations with acUnlde and decay heating ast prescribed by Appendix K.
Single failure criterion was saUsfied by assuming the loss of one diesel generator, which resulted in the disabling of one HPSI pump and one motor driven auxiliary feedwater pump.
In addiUon, one HPSI pump was assumed to be off line for service, therefore leaving only one active HPSI pump.
InitiaUon of the. HpSI 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 steam generator level while the turbine4riven pump is actuated at 600 seconds after break initiation.
In addition to the HPSI pump, credit was also taken for a charging pump.
SpecNcally, 40 gpm of water from a charging pump was included. Ofthe 40 gpm, half of the flowwent into the intact loop while half went to the broken loop.
In the break spectrum calculations, the cold legs in the intact loop were lumped as a single loop.
Hence the pumps in the intact loop were modeled as a single pump. The pumps in the broken
EMF-92-1 48 Page 3-3 loop were modeled individually.
In the break spectrum calculations, all RCPs were tripped immediately followingreactor scram.
In sensitivity calculations, the base ANF-RELAP nodalization was modNed such that all four cold legs were modeled explicitly. 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. Lucie Unit 1.
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 at100% 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 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.
EMF-92-148 Page 3-4 Table 3.1.
St. Lucie 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, ft SIT Volume, ft (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, ftAbf 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.
4522 1500 2020 1090 21 4,7 110 2'I 21 (259o) 6.00e6 435 5,955e6 272 32,495 886 1063.2 101,900 1887 1600 30.0
EMF-92-1 48 Page 3-5 Table 3.1.
St. Lucie Unit 1 SBLOCA System Analysis Parameters (Cont.)
HPSI Fiow Rate Versus RCS Pressure RCS Pressure (psia) 1129.0 1125.0 1115.0 1015.0 81 5,0 61 5.0 31 5.0 165.6 159.9 135.5 92.4 28.7 14.7 Total HPSI Flow (Ibm/sec) 0.
3.59 9.12 28.73 47.79 61.05 76.66 82.98 84.26 86.06 88.76 89.36 Total LPSI Flow5 (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.
1/2 of the total flow is distributed to the lumped loop while 1/4 of the flow is distributed to each of the cold legs in the broken loop.
EMF-92-1 48 Page 3-6 Table X2 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, %
OA40 0,384 0,377 0.580 8.180 136,7 63,19 3.90
EMF-92-148 Page 4-1 4.0 RESULTS AND ANALYSIS 4.1 Break S ectrum Calculations SBLOCA break spectrum calculations were performed for break sizes of 0.05 ft, 0.1 ft, 0.15 fthm, and 0.2 ff. The calculated event times for each break are summarized ln Tabte 4.1, The results from TOODEE2 hot rod response calculations are presented in Table 4,2.
Figures 4.1 through 4.7 summarize the key results for the 0,1 ff 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 press~re decreased immediately after break initiation. 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 230 seconds, liquid was expelled from the loop seal piping, allowing steam to flowdirectly to the break, which caused the primary pressure to decrease more rapidly.
The primary pressure increased slightfy after 1260 seconds, corresponding to the time of SIT flow initiation.
The addition of SIT fluid into the core increased the steaming rate and hence raised the primary pressure.
The break flowrate, shown in Figure 4.2, followed the same trend as the primary system pressure for the flrst 260 seconds of the transient.
The loop seals cleared at approximately 260 seconds, at which time the flow out the break transitloned from liquid to steam, causing a rapid decrease in the break flowrate. The minor oscillations around 1260 seconds were caused by the initiation of SIT flow, shown in Figure 4.3. At 1500 seconds, the break flow rate was only slightly larger than the total HPSI flow, shown in Figure 4,4.
For St. Lucie Unit 1, the setpoint for SIT initiation is 214.7 psia Figure 4.3 shows that the SIT flow initiated at approximately 1260 seconds and terminated at 1300 seconds, supplying a large burst of cooling water. The smaller HPSI system initiated flow at approximately 60 seconds, as
EMF-92-148 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 260 seconds as a result of loop seal clearing, which forced more liquid into the reactor vessel.
The core level then continued to decrease until 1260 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 1260 seconds.
The additional inventory at 1260 seconds was from SIT flow, The PCT for the 0.1 ft break calculation was 1672 F with a maximum local cladding oxidation of 1,65%.
The PCT occurred at 1244 seconds, approximately 20 seconds before the initiation of SIT flow. Rod burst was predicted In this case.
The burst of the-hot rod expedited the timing of the PCT. 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 calculation and are therefore, not discussed here.
The responses of the 0.15 ft and 0.2 ft break calculations are shown in Figures 4.15 through 4.28.
The PCTs for these two hreaks were 1606 F and 1520 F, 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 flowalone was enough to provide core cooling. The calculated PCT for the 0.05 ft break was 141 9 F. The responses of the 0.05 ft break calculation are shown in Figures 4.8 through 4.14.
4.2 Sensit 'alculations Two sensitivity calculations were 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
EMF-92-1 48 Page 4-3 were tripped at 120 seconds after reactor scram and the remaining two RCPs were tripped at the time of minimum primary inventory.
The trip of the pump in the broken loop cold leg and one pump in the intact loop was assumed to be delayed.
Calculations were performed for the 0.1 ft break and the 0.15 ft break.
The calculated event times for each sensitivity break calculation are summarized in Table 4.3.
The results from the corresponding TQQDEE2 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 4.29.
Figure 4,29 showed 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 break flow rate is shown in Figure 4.30.
The delayed RCP trip case had a slightly higher break flow rate.
This was expected since the pump was operating in the broken cold leg.
The effect of RCP trip on the total primary system mass ls shown in Figure 4.31.
In the first 1000 seconds, the primary system mass for the delayed trip case was higher than the early trip case.
Eventually the primary system mass for the delayed trip case dropped below the early trip case, at which time SlT flow initiated in the early trip case.
The PCT for the 0.1 ft break with delayed RCP trip was 1614 F with a maximun local cladding oxidation of 0.79%. The hot rod temperature response from the TOODEE2 calculation is shown in Figure 4.32. The PCT for the 0.1 ft break vyith early RCP trip was 1672 F. Thus, the overall effect of delayed RCP trip was a lower calculated PCT for the 0.1 ft break.
The RCP trip delay had similar effects on the calculated parameters for the 0.15 ft break, as illustrated in Figures 4.33 through 4.35, The hot rod temperature response from the TOODEE2 calculation is shown in Figure 4.36.
The PCT for the 0.1 5 ft break with delayed RCP trip was 1532 F with a maximum local cladding oxidation of 0.28%.
The PCT for the 0.15 ft break with early RCP trip was 1606 F. Thus, the overall effect of delayed RCP trip was also a reduction in PCT for the 0,15 ft break.
EMF-92-1 48 Page 4-4 Table 4.1. Calculated Event Times for Break Spectrum Calculations ME sec Break Size 0.05 5 0.1 ff 0.15 ff 0.2 ff Break initiation 0.0 0.0 0.0 0.0 Reactor trip/ RCP trip SIAS + 30 sec delay HPSI initiation Motor driven aux. feed Turbine driven aux. feed Loop seal clearing (Lumped loop)
Loop seal clearing (Broken loop, intact leg) 70.0 71.0 58.3 59,0 54.2 55.0 317.
190.
20,0 11.1 8.6 7.5 51.7 316,.
- 160, Loop seal clearing (Broken loop, broken leg)
Break uncovered Minimum primary system mass SIT flow initiation Time of PCT SlT flowtermination End of calculation 600.
1757.
1956.
260.
180.
400.
- 280, 1221.
721.
1266.
721.
1244.
726.
1303.
750.
- 1500, 900.
140.
506.
506.
511.
537.
~ 7M.
EMF-92-1 48 Page 4-5 Table 4.2.
Break Spectrum SBLOCA Results Break Size 0.05 5 0.1 5 0.1 5 ff 0.2 ff Hot Rod Burst
- Time (sec)
- Elevation (ft)
- Channel Blockage Fraction NA NA NA 1243.
NA 9.97 NA 0,33 NA NA NA NA Peak Clad Temperature
- Temperature
('F)
- Time (sec)
- Elevation (ft) 1419.
167 1606.
1520.
1956.
726.
511.
10.47 9.97 9,97 9.97 Metal-Water Reaction
- Local Maximum (%)
- Elevation of Local Max. (ft)
- Hot Pin Total (%)
- Core Maximum OA6 10.47 0.04
<1%
1.65 0.49 0.25 9.97 9,97 9.97
. 0.1 9 0.08 0.05
<1 %
<1 %
<1 %
EMF-92-148 Page 4-6 Table 4.3. Calculated Event Times for RCP Trip Delay Sensitivity Calculations TIME sec 0.1 ft Break Size 0.15 fP Break tnitiation Reactor trip RCP trip (2 early)
SIAS + 30 sec detay HPSI Initiation Loop seal clearing (Intact loop, leg A)
Loop seal clearing (Intact loop, leg B)
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 StT flowtermtnatton End of calculation 0,0 131.
49.3 50.0 600.
1500.
1500.
1521.
151 6.
1700.
0.0 8.6 129.
43.7 44.0 300.
300.
31 7.
891.
891.
891.
961.
1000.
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EMF-92-148 Page 4-7 Table 4.4.
RCP Trip Delay Sensitivity SBLQCA Results 0.1 ft Sreak Size 0.15 tt Hot Rod Burst Peak Clad Temperature
- Temperature
('F)
- Time (sec)
- Elevation (ft) 1614.
1516.
9.97 1532.
895.
9.97 Metal-Water Reaction
- Local Maximum (%)
- Elevation of Local Max. (ft)
- Hot Pin Total (%)
- Core Maximum 0.79 9.97 0,11 (1%
0.28 9.97 0,04
<1%
2400
~ 0 r~ <90.0 ~ Primary
~
~ Secondary
)
~.'i
". QQO.Q S.
M N0 1200.0 8
, 800.0 0
400.0 0.0 0.0 200.0 400.0 800.0 800.0 1000.0 1200.0 1400.0 1800.0 Time (sec)
Figure 4.1. Primary and secondary system pressures for the 0.1 ft break.
.400 0 0I 0) 1600.0 I
4$
1200.0 0'00.0 Nd 400.0 0.0 0.0 200.0 400.0 600.0 800.0 1000.0 1200.0 1400.0 1600.0 Time {sec)
Figure 4.8. Break flow rate for the 0.1 ft break.
1200.0 1000.0 400.0 N
200.0 0.0
-200.0 0.0 200.0 400.0 600.0 800.0 1000.0 1200.0 1400.0 1800.0 Time (sec)
Figure 4.3. Total SIT flow rate for the 0.1 ft break.
'7.5
- 76. 0
~,
50.0 87.5 25.0 12.5 0.0 0.0 200.0 400.0 SOO.O 800.0 1000.0 1200.0 1400.0 1SOO.O Time (sec)
Figure 4.4. Total HPSI flow rate for the 0.1 ft break.
- 12. 0
~
I I~
~\\
~ t,w t
~ I 0 ~
t t ~
fl i
~-
I
\\ ~
i tQ)'
l g; l'J
'0.0 8.0 6.0 t
i 4 0 2.0 0.0 0.0
'00.0 400.0 800.0 800.0 1000.0 1200.0 1400.0 Time (sec)
Figure 4.5. Collapsed core liquid level for the 0.1 ft break.
5.0 4.0 P 0 I'
g g 4 2.0 1.0 0.0 0.0 200.0 400.0 600.0 800.0 1000.0 1200.0 Time (sec)
Figure 4.8. Total primary system mass for the 0.1 ft break.
1400.0 1800.0
l.
PCT NOOC INODL 21 AT 9.97 fT., PCT -
1672.0 DI:Q'. I RUPTINED NODE SSK AS PCT NOIX.
150.0 300.0 450.0 600.0 750.0 900.0 1050.0 1200.0 1350.0 1500.0 TINE SECONOS Figure 4.7. Hot rod temperature response for the 0.1 ft break.
2400.0
- iCO.O ~ Primary
~
~ Secondary
~'F00.0
~ 4200.0
~-
oi p4r V.~
gno'g, 400.0 0.0 0.0 250.0 500.0 750.0 1000.0 1250.0 1500.0 1750.0 2000.0 2250.0 2500.0 Time (sec)
Figure 4.8. Primary and secondary system pressures for the 0.05 ft break.
1200.0 1000.0 800.0 SOO.O 400.0 RQQ.O 0.0 0.0 250.0 500.0 750.0 1000.0 1250.0 1500.0 1750.0 2000.0 2250.0 2500.0 Time (sec)
Figure 4.9. Break floor rate for the 0.05 ft break.
1.2 Q.a O
Q
~ ~
~ g i
Q.Q
.~.Iq 1.2 0.0 250.0 500.0 750.0 1000.0 1250.0 1500.0 1750.0 2000.0 2250.0 2500.0 Time (sec)
Figure 4.10. Total SIT flow rate for the 0.05 ft break.
Al EO 4M W 0)
87.5 75.0 OS 62.5 8
50.0 0
ld QC 37.5 lO to 25.0 69 12.5 0
0.0 250.0 500.0 750.0 1000.0 1250.0 1500.0 1750.0 2000.0 2250.0 2500.0 Time (sec)
Figure 4.11. Total HPSI flow rate for the 0.05 ft break.
12.0 10.0 8.0 g.p 2.0 0.0 250.0 500.0 750.0 1000.0 1250.0 1500.0 1V50.0 2000.0 2250.0 2500.0 Time (sec)
Figure 4.12. Collapsed core liquid level for the 0.05 ft break.
C) g5 0 40.0 35.0 30.0 Al NN a$
25.0 20.0 15.0 10 '
0.0 250.0 500.0 750.0 1000.0 1250.0 1500.0 1750.0 2000.0 2250.0 2500.0 Time (sec)
Figure 4.13. Total primary system mass for the 0.05 ft break.
C3 l.
PCT NODE lNODE 23 AT 10.47 fT., POT "
1419.3 DES.)
NO RUPTURED NODE.
(a) g C9 C)
I 4l co f-s CC CL lLl 0 p C9 O
0:
C3 CI R3.0 250.0 500.0 750.0 1000.0 1250.0 1500.0 1750.0 2000.0 2250.0 2500.0 T I t1E SECONDS Figure 4.14. Hot rod temperature response for the 0.05 ft2 break.
2400.0 2000.0 ~ Primary
~
~ Secondary 1SOO-0 I
0 1200.0 A
8 SOO.O 0
~~a
~~ ~~
$90.0 0.0 0.0 100.0 200.0 300.0 400.0 500.0 800.0 700.0 800.0 900.0 Time (sec)
Figure 4.15. Primary and secondary system pressures for the 0.15 ft break.
3500.0 3000.0 U9 2500.0 E
2000.0 8
1600.0 0
1OOO.O
@90 0 0.0 0.0 100.0 200.0 300.0 400.0 500.0 Soo.o 700.0 800.0 900.0 Time (sec)
Figure 4.18. Break flow rate for the 0.15 ft break.
1750.0 1500.0 OS gg 1250.0 8
1000.0 I
0$
750.0 o
500.0 4$
260.0 0.0 0.0 100.0 200.0 300.0 400.0 600.0 600.0 700.0 800.0 900.0 Time (sec)
Figure 4.17. Total SIT flow rate for the 0.15 ft break.
87.5 75.0 O
gg 62.5 Q
8 S) 50.0 I
d 37.5 0
25.0 a$
12.5 0.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 700.0 800.0 900.0 Time (sec)
Figure 4.18. Total HPSI flow rate for the 0.15 ft break.
CD 4
M EO IQ Ul CD
~
~
12.0 10.0 8.0 6.0 8
$.0 2.0 0.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 700.0 Time (sec)
Figure 4.19. Collapsed core liquid level for the 0.15 ft break.
800.0 900.0
3.0 8
N Nd 2.0 1 ~ 0 0.0 0.0 100.0 200.0 300.0 '00.0 500.0 800.0 700.0 Time (sec)
Figure 4.80. Total primary system mass for the 0.15 ft break.
800.0 900.0
1.
PCT NOOE (NOOE 21 AT 9.97 FT.i PCT -
1606.1 OEGf'.)
RO RUPT11REO HOOE.
90.0 180.0 270.0 360.0 450.0 540.0 630.0 TIt1E SECONDS 720.0 810.0 900.0 Figure 4.21. Hot rod temperature response for the 0.15 ft break.
2400.0 2000.0 ~ Primary
~
~ Secondary 1600-0 0I O
1200.0 8
600.0 0
~~ ~
~
400.0 0.0 200.0 600.0 100.0 0.0 300.0 400.0 Time (sec)
Figure 4.22. Primary and secondary system pressures for the 0.2 ft break.
i-. i000 0 00 8
3000.0 0
C$
o 2000.0 0$
1000.0 0.0 0.0 100.0 200.0 300.0 400.0 500.0 Time (sec)
Figure 4.23. Break floor rate for the 0.2 ft break.
2500.0 0
N 2000.0 8
8 Q
1600.0 I
a$
1000.0 0
500.0 td 0.0
-600.0 0.0 200.0 100.0 300.0 400.0 500.0 Time (sec)
Figure 4.24. Total SIT flow rate for the 0.2 ft break.
240.0 290.0 g ji
~)
10
~
~
180.0 I
a$
120.0 l4 0
80.0 N
td 40.0 0.0 0.0 100.0 200.0 300.0 400.0 500.0 Time (sec)
Figure 4.25. Total HPSI flow rate for the 0.2 ft break.
800.0
8.0 6.0 0
$.0 2.0 0.0 0.0 100.0 200.0 300.0 400.0 Time {see)
Figure 4.28. Collapsed core liquid level for the 0.2 ft break.
5.0 4.0 3.0
~%
1.0 0.0 0.0 ROO.O 500.0 300.0 400.0 Time (sec)
Figure 4.27. Total primary system mass for the 0.2 ft break.
SOO.O
l.
PCT UOOE lNOOE 21 AT 9.97 fT., PCT -
1519.7 DEGAS.)
NO RUPTURED NODE.
70.0 140.0 210.0 2BO.O 350.0 020.0 TtI1E SECONOS 490.0 560.0 630.0 700.0 Figure 4.2S. Hot rod temperature response for the 0.2 ft break.
2400.0 2000.0 c
c Early RCP Trip
~
~ Delayed RCP Trip 18OO.O r)
S too.o
\\ a
'M 800.0 Q
400.0
~~O 0.0 0.0 200.0 400.0 800.0 800.0 1000.0 1200.0 1400.0 1800.0 1800.0 Time (sec)
Figure 4.2S. Effect of delayed RCP trip on the primary system pressure for the 0.1 ft break.
'Y)00.0
"--'000.0 ~ Early RCP Trip
~
~ Delayed RCP Trip Q'>
N'.,5600.0 f,,
~I O..
~ a.
C$
>.899. 'i 3 ~
0 8OO.O N
Old joo.o 0.0 l I
~
0.0 200.0 400.0 800.0 800.0 1000.0 1200.0 1400.0 1600.0 1800.0 Time (sec)
Figure 4.30. Effect of delayed PCP trip on the break flow rate for the 0.1 ft break.
5 0 ~ Early RCP Trip
~
~ Delayed RCP Trip 4.0 3.0 8
2.0 1.0 r
~
0.0 0.0 200.0 400.0 800.0 800.0 1000.0 1200.0 1400.0 1600.0 1800.0 Time (sec)
Figure 4.31. Effect of delayed RCP trip on the total primary system mass for the 0.1 ft break.
1.
PCT NXK (NOOE 21 AT 9.97 FT., PCT -
1614.1 OE6f'. l 170.0 340.0 510.0 680.0 850.0 1020.0 1190.0 1360.0 1530.0 1700.0 TIt1E SECONDS Figure 4.32. Hot rod temperature response for the 0.1 ft break with delayed RCP trip.
2400.0 2000.0 0
0 Early RCP Trip
~
~ Delayed RCP Trip
~W M
1600.0
- Ii c.
M q) 1200.0 O'
800 0 r r AGO.O 0.0 0.0 100.0 200.0 800.0 400.0 500.0 800.0 700.0 800.0 900.0 1000.0 Time (sec)
Figure 4.33. Effect of. delayed RCP trip on the piimary system pressure for the 0.15 ft break.
3500.0 3000.0 ~ Early RCP Trip
~
~ Delayed RCP Trip
(-590.0 OOOO.O
.1 500.0 6A<.9 0.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0
'700.0 800.0 900.0 1000.0 Time (sec)
Figure $.34. Effect of delayed RCP trip on the break flow rate for the 0.15 ft break.
5 0 ~ Early RCP Trip
~
~ Delayed RCP Trip 4.0 3.0 8
43'A 6$
2.0 1.0
~
~
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 4.35. Effect of delayed RCP trip on the total primary system mass for the 0.15 ft break.
1.
PCT NODE Owe 21 AZ O.W rr., 1'Cr -
1SSa.a OECr.1 NO RS'lNEO NOOE.
1oo.o m.o zoo.o caa.a 'aa.o sao.o
~00.0 Boo.o T ICE SECONDS Figure $.38. Hot rod temperature response for the 0.15 ft break with delayed RCP trip.
900.0 1000.0
EMF-92-148 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. Lucie Unit 1 at a nominal power level of 2700 MWt and 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 14x1 4 fuel within the limits stated above assures that the NRC acceptance criteria for SBLOCA (10 CFR 50A6(b)) willbe met with the existing emergency core cooling system.
l
'I
EMF-92-1 48 Page 6-1
6.0 REFERENCES
1.
XN-NF<2<9 P Revision 1
Su lement 1 "Exxon Nuclear Company Evaluation Model-EXEM PWR Small Break Model," May 1992.
2.
XN-NF41-58 P A
Revision 1
and Su lements 14 "RODEX2: Fuel Rod Thermal Mechanical Response Evaluation Model," Exxon Nuclear Company, February 1983.
3, NUREG CR~12 GG-2396 Volumes 1 and 2 "RELAP5/MO02 Code Manual," August 1985.
4.
Letter fram J. A. Hendechuh (FPL) to J. L Holm (SPC), NF-92-733 "St. Lucie Unit t Cycfe 12 Groundrules", October 16, 1992.
EMF-92-1 48 Issue Date:
2-5-93 ST. LUCIE UNIT 1 SMALLBREAK LOCA ANALYSIS OIstrlbution C. Y. Chou R. A. Copeland K. M. Ouggan R. C. Gottula J. S. Holm T. R. Llndqulst L A. Nielsen L D. O'Dell P. Salim S. E. Spangler B. D. Stltt C. J. Volmer R. I. Wescott FPL(11)/J. L Holm Document Control (5)