ML20083N209
| ML20083N209 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 12/20/1982 |
| From: | Jensen S, Krajicek J, Stout R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17213B024 | List: |
| References | |
| XN-NF-82-98, NUDOCS 8302010588 | |
| Download: ML20083N209 (47) | |
Text
XN-NF-82-98 Issue Date: 12/20/82 ST. LUCIE UNIT 1 LOCA ANALYSIS USING THE EXEM/PWR ECCS MODEL 1'
Prepared by: )'
UJ. E.grajicek '
/ z//3/sz NSSS Vystems Analysis Concur: f ups /2f/f' f's'
S. E. A nsen, Manag'er ~
Approved: 7hM R. B. Stout, M6 nager
/J// y/F t
, Licensing & Safety Engineering Approved: b .'i b 2c J J G. A. 50 K tianAger Fuel Engineering & Technical Services 9f E(ON NUCLEAR COMPANY,Inc.
8302010588 830120 DR ADOCK 05000335 PDR
NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY _
This technical report was tierived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear fabricated reloarl fuel or other technical services provided by Exxon Nuclear for liaht water power reactors and it is true and correct to the best of Exxon Nuclear's kne , ledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstratiort of compi?ance with the USN RC's regulations.
Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf:
A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any informatiort apparatus, method, or process disclosed in this document will not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for darrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.
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XN- NF- F00, 766 l
i XN-NF-82-98 TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
AND
SUMMARY
........................... 1 2.0 BREAK SPECTRUM ANALYSIS ............................ 3 2.1 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION ................,.......... 3 2.2 THERMAL ANALYSIS'.............................. 4 2.2.1 Method of Analysis ..................... 4 2.2.2 Large Break LOCA Analysis Modeling ..... 5 2.3 BREAK SPECTRUM RESULTS ........................ 6
3.0 CONCLUSION
S ........................................ 37
4.0 REFERENCES
......................................... -38
if XN-NF-82-98 LIST OF TABLES Table No. Page 2.1 St. Lucie Unit 1 Large Break Events ................ 9 2.2 St. Lucie Unit 1 Large Break Results ............... 10 2.3 St. Lucie Unit 1 PWR Data .......................... 11 2.4 St. Lucie Unit 1 Dry Containment Data .............. 13
iii XN-NF-82-98 LIST OF FIGURES Figure No. Page 2.1 RELAP4-EM Blowdown System Nodalization for St. Lucie Unit 1 ................................... 15 2.2 Reflood Nodalization for Guillotine Breaks for St. Lucie Unit 1 ............................... 16 2.3 Reflood Nodalization for Split Breaks for St'. Lucie Unit 1 ................................... 17 2.4 Blowdown System Pressure, 0.4 DECLG Break .......... 18 2.5 Blowdown Total Break Junction Flow Rate, 0.4 DECLG Break .................................... 19 2.6 Blowdown Pressurizer Surge Line Flow Rate, 0.4 DECLG Break .................................... 20 2.7 Double Intact Loop Accumulator Flow Rate, 0.4 DECLG Break .................................... 21 2.8 Single Intact Loop Accumulator Flow Rate, 0.4 DECLG Break .................................... 22 2.9 Blowdown Average Channel Inlet Flow Rate, i
0.4 DECLG Break .................................... 23 2.10 Blowdown Average Channel Outlet Flow Rate, 0.4 DECLG Break .................................... 24 i
2.11 Blowdown Hot Channel Inlet Flow Rate, 0.4 DECLG Break .................................... 25 l 2.12 Blowdown Hot Channel Outlet Flow Rate, O.4 DECLG Break .................................... 26 1 2.13 Blowdown Hot Rod Cladding Surface Temperature, Node 25, 0.4 DECLG Break ........................... 27 l
2.14 Blowdown Hot Rod Volumetric Average Temperature, 1 Node 25, 0.4 DECLG Break ........................... 28 l
I i
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1 iv XN-NF-82-98 i
LIST OF FIGURES (Cont.)
l ,
I 4 Figure No. Page 2.15 Hot Rod Blowdown Heat Transfer Coefficient, Node 25, 0.4 DECLG Break ............................ 29 4
- 2.16 Hot Rod Blowdown Depth of Zirconium - Water ,
Reaction, Node 25, 0.4 DECLG Break .................. 30 2.17 Containment Backpressure Versus Time, ;
! 0.4 DECLG Break ..................................... 31 7 2.18 Normalized Power Versus Time, 0.4 DECLG Break ....... 32 !
2.19 Reflood Core Flooding Rate, 0.4 DECLG Break ......... 33 ,
2.20 Reflood Downcomer Mixture Level, 0.4 DECLG Break .... 34 4
2.21 Reflood Core Mixture Level, 0.4 DECLG Break ......... 35 ,
2.22 T00DEE2 Calculated Cladding Surface Temperature, i 0.4 DECLG Break ..................................... 36 I
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v XN-NF-82-98 The author wishes to express his appreciation to the individyals listed below for their efforts in performing various phases of the St. Lucie Unit 1 LOCA analyses as well as their comments and suggestions.
S. E. Jensen D. C. Kolesar D. R. Swope
1 XN-NF-82-98
1.0 INTRODUCTION
AND
SUMMARY
This report provides results of a LOCA ECCS analysis supporting St. Lucie Unit I reactor operation at 2700 MWt with Exxon Nuclear Company (ENC) supplied fuel. Contained within the report are results of (1) a LOCA break spectrum analysis, and (2) a limiting break recalculation using the GAPEX stored energy model and increased radial peaking. The break spectrum
.i analysis uses the RODEX2 stored energy model. All analyses were perforred using the EXEM/PWR ECCS evaluation model(1,2). For the limiting break (0.4 DECLG case) with beginning-of-life (BOL) stored energy (GAPEX value) and end-of-cycle (EOC) fission gas release, the calculated peak clad temperature was 20590F with the axial power peak at 70 percent of the core height and a peak linear heat generation rate of 15.30 kw/f t (102% of ~ 15.00 kw/f t). The limiting break analysis was performed in conformance to Appendix K of 10 CFR 50, yield results which satisfy the NRC criteria specified in 10 CFR 50.46.
The analysis applies only for Cycle 6 operation of St. Lucie Unit 1.
The LOCA break spectrum calculations included guillotine break con-figurations for double-ended cold leg pipe breaks (DECLG) with discharge coefficients of 1.0, 0.6 and 0.4. The split configuration breaks of the cold leg pipe were also calculated with a break area equal to twice the cross-sectional pipe area (DECLS,10.01 f 2t ), then with break areas of 6.01 and 0.8 square feet. The break spectrum analysis was performed for a core composed of ENC fuel at nominal beginning-of-life (BOL) conditions. The results of the spectrum analysis identified the limiting break to be the 0.4 DECLG case.
2 XN-NF-82-98 A detailed discussion of the break spectrum results is provided in Section 2.0. All of the calculations in the break spectrum were performed at a core power of 2754 MWt, which is 102 percent of rated power.
The break spe:trum calculations were performed with a version of the RODEX2 code supplying the initial stored energy. The NRC is currently reviewing the RODEX2 code, with approval not expected in time for St. Lucie Unit 1 Cycle 6 startup. Therefore, ENC repeated the limiting break analysis using the GAPEX stored energy model previously approved by NRC. A combination of BOL maximum stored energy and E0L maximum fission gas release was used to bound Cycle 6 operation.
i
3 XN-NF-82-98 2.0 BREAK SPECTRUM ANALYSIS 2.1 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION The analyses for large breaks specified by 10 CFR 50.46(3),
" Acceptance Criteria for Ew.ergency Core. Cooling Systems for Light Water Power Reactors," is presented in this section. The results of the loss of coolant accident analysis are shown in Tables 2.1 and 2.2, and indicate compliance with the Acceptance Criteria. The analytical techniques used are in compliance with Appendix K of 10 CFR 50, and are as described in XN-75-41, Volumes I and II, and supplements (l); ENC EXEM/PWR model is described in XN-NF-82-20(P) and supplements (2). The detailed system models are as given in the example problem report for a Combustion Engineering 2x4' PWR which is.
Supplement 3 of XN-NF-82-20(P).
For the purpose of loss-of-coolant accident (LOCA) analyses, a LOCA is defined as a hypothetical rupture of the Reactor Primary Coolant System piping, up to and including the double-ended rupture of the largest pipe in the Reactor Coolant System or of any line connected to that system up to the first closed valve.
Should a major break occur, depressurization of the Reactor Coolant System results in a pressure decrease in the pressurizer. A reactor trip signal occurs when the pressurizer lower pressure trip setpoint is reached.
l l Reactor trip and scram were conservatively neglected for the large break analyses. A Safety Injection System signal is actuated when the appropriate setpoint (high containment pressure) is reached. These countermeasures will l limit the consequences of the accident in two ways:
l
4 XN-NF-82-98
- 1. Reactor trip and borated water injection complements void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.
- 2. Injection of borated water provides heat transfer from the reactor core and prevents excessive clad temperatures.
At the beginning of the blowdown phase, the entire Reactor Coolant System contains subcooled liquid which transfers heat from the core by forced convection cooling. After the break develops, the time to departure from nucleate boiling (DNB) is calculated consistent with Appendix K of 10 CFR 50(3). Post-DNB core heat transfer (both transition and film boiling occurring) is also calculated in accordance with Appendix K of 10 CFR 50. As the core becomes uncovered, both turbulent and laminar force'd convection to steam are considered as core heat transfer me~chanisms.
When the Reactor Cool-it System pressure f alls below 230 psia, the accumulators begin to inject borated water. The conservative assumption is made that accumulator ECC water bypasses the core and goes out through the break until the termination of bypass. This conservatism is consistent with Appendix K of 10 CFR 50.
2.2 THERMAL ANALYSIS 2.2.1 Method of Analysis For breaks greater than 0.8 f t , 2the RELAP4-EM code in EXEM/PWR is used to calculate the transient depressurization of the Reactor Coolant System as well as to describe the mass and enthalpy of flow out of the break. A specialized calculation (RELAP4-EM/ HOT CHANNEL) is used to
5 XN-NF-82-98 i
calculate cladding temperatures using time dependent boundary conditions in the upper and lower plenum volumes from the basic blowdown analysis. Beyond the point of refill to the bottom of the core, a specialized calculation (REFLEX) is applied to determine the reflooding rate and system conditions.
Af ter end-of-bypass (E0BY), the program T00DEE2 is used to calculate peak clad temperatures.
2.2.2 Large Break LOCA Analysis Modeling The St. Lucie Unit 1 nuclear power plant is a 2x4 Combustion Engineering pressurized water reactor with a dry containment, The reactor
~
coolant system is nodalized into control volumes representing reasonably homogeneous regions, interconnected by flow paths or " junctions" as described in Supplement 3 of XN-NF-82-20(2). The nodalization in Figure 2.1 differs from the example problem nodalization in the broken loop cold leg region. The number of broken loop cold leg volumes have been reduced.
Five percent of the steam generator tubes were assumed to be l
l uniformly plugged. The unbroken loop was assumed symmetrical and modeled the same as the broken Icop except for the break nodalization and the pres-surizer. Pump performance curves characteristic of the St. Lucie Unit 1 pumps were used in the analysis. System input parameters are given in Table l
2.3.
The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and with decay heating as required by Appendix K of 10 CFR 50. The axial power profile l used for the break spectrum analysis is a top skewed curve with the power peak j above the core midplane.
l
6 XN-NF-82-98 The values for the primary coolant system core inlet temperatures and the steam generator secondary side pressure were set for the St. Lucie Unit 1 plant based upon information provided by the utility. The values of the core inlet temperature and the steam generator secondary side pressure are 5490F and 870 psig, respectively.
The containment backpressure for the analysis of the postulated LOCA was evaluated in accordance with the discussion presented in XN-75-41, Supplement 5, Section 4.6. A containment analysis was performed using the computer code CONTEMPT-LT, Version 22, modified as described in Supplement 5, Revision 1 of XN-75-41(1). The condensing heat transfer coefficient is modeled in accordance with Branch Technical Position CSB 6-1,
" Minimum Containment Pressure Model for PWR ECCS Performance Evaluation"(4).
4 The containment. parameters used in the containment analysis to determine the ECCS backpressure are presented in Table 2.4.
The reflood nodalization for the guillotine and split breaks in the break spectrum are shown in Figures 2.2 and 2.3. These nodalizations include a leakage path between the upper plenum and the upper downcomer.
2.3 BREAK SPECTRUM RESULTS Using the EXEM/PWR codes, transient system behavior is determined by solving the governing conservation equations for mass, energy, and momentum. Energy transport, flow rates, and heat transfer are determined from appropriate correlations. Table 2.1 presents the timing and sequence of events as determined for the large break guillotine configuration with discharge coefficients of 1.0, 0.6 and 0.4 and the split break configuration with break areas of 10.01, 6.01, and 0.8 square feet.
7 XN-NF-82-98' The blowdown calculations for the break spectrum were initialized with the ENC fuel performance code RODEX2. Pending NRC approval of RODEX2, the limiting break in the spectrum was reanalyzed using GAPEX for fuel rod stored energy initialization. The fuel initialization using GAPEX encom-passes the maximum fuel rod stored energy (80L) and end-of-cycle (E0C) fission gas release for Cycle 6. The input radial peaking factor was also increased to bound measurement uncertainties while. the peak liner heat generation rate was maintained at 15 kw/f t. The limiting break results which have been initialized with GAPEX and include the increased radial peaking factor are identified in Tables 2.1 and 2.2.
In general, the transient events occur slower for smaller discharge coefficients or break sizes. Table 2.2 presents the peak clad temperatures and maximum metal-water reaction results for the above spectrum of break cases. This range of break sizes was determined to include the limiting case for peak clad temperature.
The analysis of the loss-of-coolant accident is performed at 102%~
of 2700 MWt (2754 MWt). The core power and other parameters used in the i
analyses are given in Table 2.3. Since there is usually margin between the value of the peak linear power density used in this' analysis and-the value expected in operation, a lower peak clad temperature would be obtained by using the peak linear power density expected during operation.
For the result discussed below, the hot spot is defined to be the location of maximum peak clad temperature. This location is given in Table 2.2 for each break size analyzed.
i
s 8 XN-NF-82-98 Figures 2.4 througn 2.22 present the results of the revised analysis for the limiting break (0.4 DECLG). Unless otherwise noted, zero 4
time corresponds to the time of break initiation.
.The maximum peak cladding temperature of 20590F was ~ calculated for the double-ended cold leg guillotine break configuration (CD = 0.4) with a.
total linear heat generation rate of 15.30 kw/ft for ENC fuel (102% of 15.00 kw/ft). The maximum local metal-water reaction is less than 5% and the core-wide reaction is less than 1%, all well below the limit's set by the criteria -
of 10 CFR 50.46.
ENC has performed numerous analyses and sensitivity studies on PWR I
systems using the ENC ECCS evaluation model. These studies have demonstrated the adequacy of the system nodalization used. In addition, these studies have shown that for transient conditions similar to those calculated for the St. Lucie Unit 1 reactor during the LOCA, the reactor coolant inlet pipe or cold leg is the worst break location.
NSSS vendor analyres have shown large , breaks to be more limiting than small breaks for St. Lucie Unit 1 ECCS analyses.
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Table 2.1 St. Lucie Unit 1 Large Break Events Event Time (seconds)
DECLG* DECLG DECLG DECLG 1.0 DECLS 0.6 DECLS 0.08 DECLS (CD=0.4) (CD =0.4) (CD =0.6) (CD=1.0) (10.01 ft2) (6.01 ft2) (0.8 ft2)
Start 0. O. O. O. O. O. O.
Initiate Break 0.05 0.05 0.05 0.05 0.05 0.05 0.05 Safety Injection Signal 1.20 1.20 0.98 0.82 0.86 0.91 5.20 Pressurizer Empties 8.95 8.95 8.85 8.80 9.10 9.25 9.25 .
Accumulator Injection, Intact Loops 22.58 22.65 18.50 16.65 16.65 17.33 99.70 End-of-Bypass 28.01 27.03 24.27 21.98 21.21 21.96 103.53 Safety Injection Flow, SIS 31.20 31.20 30.98 30.98 30.82 30.86 35.20 Start of Reflood 45.58 44.42 42.07 39.78 38.87 39.61 120.57 Peak Clad Temperature Reached 161.5 156.5 159.1 161.6 156.0 153.1- 420.5 5
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11 XN-NF-82-98 Table 2.3 St. Lucie Unit 1 PWR Data Primary Heat Output, MWt 2700*
Primary Coolant Flow, Ibm /hr 1.394 x 108 Primary Coolant Volume, ft3 19,214**
Operating Pressure, psia 2250 Inlet Coolant Temperature, OF 549 Reactor Vessel Volume, ft3 4402 Pressurizer Volume, Total, ft3 1500 1
Pressurizer Volume, Liquid, ft3 800 Accumulator Volume, Total, ft3 (one of four) 2020 Accumulator Volume, Liquid, ft3 1090 i
Accumulator Pressure, psia 230 Steam Generator Heat Transfer Area, ft2 (one of two) 74,722 Steam Generator Secondary Flow, Ibm /hr 5.899 x 106 Steam Generator Secondary Pressure, psia 885 Reactor Coolant Pump Head, ft 280 Reactor Coolant Pump Speed, rpm 886 Moment of inertia, Ibm-ft / 2rad 101,900 Cold Leg Pipe, I.D., in. 30 Hot Leg Pipe, I.D., in. - 42 Pump Suction Pipe, l.D., in. 30
- Frimary Heat Output used in RELAP4-EM Model - 1.02 x 2700 = 2754 MWt.
, ** Includes total accumulator and pressurizer volume.
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Fuel Assembly Rod Diameter, in* .440 l , .,
Fuel Assembly Rod Pitch, in* .580 s
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Fuel Assembly Pitch, in* 8.180 i Fueled (Core) Height, in* 136.7 -
':- Fuel Heat Transfer Area, ft2 50,117 . - -
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- . Fuel Total Flow Area, ft2 53.19 ' --
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, Steam Generator Tube Plugging (Assumed Uniform) 5% . .
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13 XN-NF-82-98 Table 2.4 St. Lucie Unit 1 Dry Containment Data Containment Physical and Thermal Parameters Net Free Volume 2.511 x 106 ft3 Enclosure Building Temperature 380F Initiation Time for:
Soray Flow 30.0 sec Fan Coolers 30.0 sec Containment Initial Conditions:
Temperature 900F Pressure 14.6 psia Relative Humidity 100%
Containment Spray Water:
Temperature 550F Flow Rate (Total, 2 pumps) 6750 gpm Fan Air Cooler Capacity (Total, 4 coolers) l Vapor Temperature (OF) Capacity (Btu /hr) 60 0.
120 5.00 x 107 180 1.06 x 108 220 1.67 x 108 264 3.00 x 108 Thermal Conduc.tivity and Volumetric Heat Capacity Thermal Volumetric Conductivity Heat Capacity I
Materials (Btu /hr-ft OF) (Btu /ft 3.oF)
Steel 25.9 53.6 Structural Concrete 1.0 34.2 Stainless Steel 9.8 54.0 Galvanizing 64.0 40.6
Table 2.4 St. Lucie Unit 1 Dry Containment Data (Cont.)
Containment Passive Heat Sinks Surface Area Description Material Thickness Ft2
- 1. Containment Shell Steel .1171 ft 86,700
- 2. Miscellaneous Concrete Concrete 1.5 ft 87,751
- 3. Floor Slab Concrete 20 ft 12,682
- 4. Galvanized Steel Zinc 0.00059 ft 130,000 Steel 0.014 ft
- 5. Carbon Steel Steel 0.031 25,000
- 6. Stainless Steel Steel 0.038 ft 22,300
- 7. Miscel<laneous Steel Steel 0.063 ft 40,000
- 8. Miscellaneous Steel Steel 0.021 ft 41,700
- 9. Miscellaneous Steel Steel 0.177 ft 7,000
- 10. Imbedded Steel Steel 0.071 ft 18,000 Concrete 7.0 ft
4 15 XN-NF-82-98 3
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3.0 CONCLUSION
S For breaks up to and including the double-ended severance of a reactor coolant pipe, the St. Lucie Unit 1 Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10 CFR 50.46 with the Cycle 6 core, with the results described herein and for ENC reload fuel. That is:
- 1. .The calculated peak fuel element clad temperature does not exceed the 22000F limit.
- 2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1% of the total amount of zircaloy in the reactor.
- 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 limits of 17% are not exceeded during or after quenching.
- 4. The system long term cooling capabilities provided for previous cores remain applicable for ENC fuel.
These Acceptance Criteria are satisfied if the St. Lucie Unit I reactor is operated at 2700 MWt within the maximum LHGR of 15.00 kw/ft.
l
i 38 XN-NF-82-98
4.0 REFERENCES
- 1. Exxon Nuclear Company, " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model," XN-75-41:
- a. Volume I, July 1975
- b. Volume II, August 1975
- c. Volume III, Revision 2, August 1975
~d. Supplement 1, August 1975
- e. Supplement 2, August 1975
- f. Supplement 3, August 1975
- g. Supplement 4, August 1975
- h. Supplement 5, Revision 5, October 1975
- i. Supplement 6, October 1975
- j. Supplement 7, November 1975
- 2. Exxon Nuclear Company, " Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," XN-NF-82-20(P), February 1982.
- a. Supplement 1, March 1982
- b. Supplement 2, March 1982
- c. Supplement 3, July 1982
- 3. 10 CFR 50.46 and Appendix K of 10 CFR 50, " Acceptance Criteria for Emergency Core Cooling System for Light Water Cooled Nuclear Power Reactors," Federal Register, Volume 39, Number 3, January 4, 1974.
- 4. U. S. Nuclear Regulatory Commission, " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation," Branch Technical Position CSB 6-1.
l XN-NF-82-98 Issue Date:
12/20/82 f ST. LUCIE UNIT 1 LOCA ANALYSIS USING j THE EXEM/PWR ECCS MODEL Distribution D. J. Braun J. C. Chandler R. E. Collingham G. C. Cooke T. J. Helbling J. S. Holm S. E. Jensen W, V. Kayser D. C. Kolesar J. E. Krajicek W. T. Nutt G. F. Owsley G. A. Sofer - Letter Only R. B. Stout - Letter Only D. R. Swope T. Tahvili G. N. Ward FPL/T. J. Helbling (25)
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