ML20137Z465
| ML20137Z465 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 09/30/1985 |
| From: | Ades M, Holm J, Valentine P SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML111751118 | List: |
| References | |
| XN-NF-85-98, NUDOCS 8510080252 | |
| Download: ML20137Z465 (60) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _
XN-NF-85-98 KEWAUNEE LIMITING BREAK K LZ; LOCA/ECCS ANALYSIS SEPTEMBER 1985 RICHLAND,WA 99352 ERON \\UC_ AR COV 3A\\Yl\\C.
puooRBaMiBuoN85 l
l XN-NF-85-98 l
Issue Date: 9/26/85 I
l KEWAUNEE LIMITING BREAK K(Z)
LOCA/ECCS ANALYSIS Prepared by: /C.Y'-
9/2Pffr
'M. J.' A(es
/
Mod 1 Development rA Prepared by:
9 25-#5 f)R Safety Analysis J. Valentine PW Concurred by:
-)
C 'L[lfi O.'5. tim, Manager' PWR Safety Analysis Concurred by: th) e&rf-R.
A'.
Co pe l an d,'
PWR Reload Licensing Approved by:
M2 s'/r6~
H. E. Williamson, Manager Licensing & Safety Engineering Concurred by:
d/
d 9/45/e_S J. fi7 Morgan, Manager Proposals & Customer Services Engineering l
Approved by:
, f.
'l /2f/ff l
G. L. Ritter, Manager l
Fuel Engineering & Technical Services
/mah 1
ERON \\UC A9 COV 3A\\Y \\C.
CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Eaumon Nucleur Corneeny's warranose and representecono concoming the subient enetter of this docuenent are those set forth in the Agreement becomen Emmen Nuclear Company, Inc. and the Cueensner pursuant to whidi this document le issued. Accordingly, encoat se otherwnse expressly provided in such Agreement, norther EJunon Nucteer Company, Inc. nor any person anting on its behalf makes any warranty or repremontecon, expressed or imedied, with respost to the accurney, cornpietensee, or usefulness of the informeelen contsened in this document. or that the use of any informsoon, espereque, rnethod or procese dioelosed in this document will not infringe prweemly owned nghte; or assumes any liabilities with respect to the use of any informenon, esperatus, method or prosese diecioned in this decurnent.
The information contmaned herein is for the solo use of Customer.
In order to evoed ungestment of rights of Exxon Nuclear Company, Inc.
in poesnts or irwenoons which may be included in the informecon contemed ire this desument, the resspoent, by its acceptance of this document eyees not to pub 6ish or make public use (in the potent use of the term) of sude utformation unal so authorued in writmg by Exmon Nucteer Company, Inc.
or until after um (6) months following termineoon or expiration of the aforessed Ayeoment and any extensson thereof, unless otherweue expressly prtpridad in the Agreement. No rights or licensee in or to any potente are implied by the furmahing of this document.
xpe-pos.7ee
i XN-NF-85-98 TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
1 2.0 LIMITING BREAK LOCA ANALYSIS................................
5 2.1 LOCA Analysis Model.........................................
5 2.2 Results.....................................................
7
3.0 REFERENCES
47
11 XN-NF-85-98 LIST OF TABLES Section Page 1.1 Kewaunee LOCA-ECCS An alys i s Res ul ts......................... 3 2.1 K e wa un e e Sy s tem 0a t a........................................ 8 2.2 F ue l De s i g n P a r ame te r s...................................... 9 2.3 Kewaunee LOCA-ECCA Analys i s Resul ts, Event Times...........
10
iii XN-NF-85-98 LIST OF FIGURES Section Page 1.1 Comparison of Power Distributions Analyzed to Limits........
4 2.1 RELAP4/EM Blowdown System Nodalization for Kewaunee........
11 2.2 Axial Peaking Factor versus Rod Length, 0.4 DECLG Break, 80C........................................................
12 2.3 Axial Peaking Factor versus Rod Length, 0.4 DECLG Break, E0C........................................................
13 2.4 Downcomer Flow Rate, 0.4 DECLG Break.......................
14 2.5 Upper Plenum Pressure 0.4 DECLG Break.....................
15 2.6 Average Core Inlet Flow, 0.4 DECLG Break...................
16 2.7 Average Core Ou tlet F low, 0.4 DECLG Break..................
17 2.8 Total Break Flow, 0.4 DECLG Break..........................
18 2.9 Average Break Flow Enthalpy, 0.4 DECLG.....................
19 2.10 Flow from Intact Loop Accumulator, 0.4 DECLG Break.........
20 2.11 Containment Back Pressure, 0.4 DECLG Break................. 21 2.12 Hot Channel Heat Transfer Coefficient, 0.4 DECLG Break, B0C........................................................
22 2.13 Clad Surf ace Temperature, 0.4 DECLG Break, 80C............. 23 1
2.14 Depth of Metal-Water Reaction, 0.4 DECLG Break, 80C........ 24 2.15 Hot Channel Average Fuel Temperature, 0.4 DECLG Break, 80C. 25 2.16 Hot Assembly Inlet Flow, 0.4 DECLG Break, B0C.............. 26 2.17 Hot Assembly Outlet flow, 0.4 DECLG Break, 80C.............
27 2.18 Hot Channel Heat Transfer Coefficient, 0.4 DECLG Break, E0C........................................................
28 2.19 Clad Surf ace Temperature, 0.4 DECLG Break, E0C............. 29 2.20 Depth of Metal-Water Reaction, 0.4 DECLG Break, E0C........
30 2.21 Hot Channel Average Fuel Temperature, 0.4 DECLG Break, E0C........................................................
31 2.22 Hot Assembly inlet Flow, 0.4 DECLG Break, E0C..............
32 2.23 Hot Assembly Outlet F low, 0.4 DECLG Break, E0C............. 33 2.24 Normalized Power, 0.4 DECLG Break, B0C.....................
34
l iv XN-NF-85-98 LIST OF FIGURES (con't)
Section Page 2.25 Normal i zed Power, 0.4 DECL G Bre ak, E0C.....................
35 2.26 Reflood Core Mi xture Level, 0.4 DECLG Break, 80C...........
36 2.27 Reflood Downcomer Mixture Level, 0.4 DECLG Break, 80C......
37 2.28 Reflood Upper Plenum Pressure, 0.4 DECLG Break, 80C........
38 2.29 Core Flooding Rate 0.4 DECLG Break, 80C...................
39 2.30 Reflood Core Mixture Level, 0.4 DECLG Break, E0C........... 40 2.31 Reflood Downcomer Mixture Level, 0.4 DECLG Break, E0C......
41 2.32 Reflood Upper Plenum Pressure, 0.4 DECLG Break, E0C........ 42
- 2. 33 Core Flooding Rate, 0.4 DECLG Break, E0C................... 43 2.34 T000EE2 Cladding Temperature vs Time, 0.4 DECLG Break, 80C....,...................................................
44
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2.35 T000EE2 Cladding Temperature vs Time, 0.4 DECLG Break, E0C........................................................
45 2.36 H t Channel Factor Normalized Envelope for Fg = 2.28, F H = 1. 5 5, K ( Z ) Fu n c t i o n................~.................. 4 6
1 XN-NF-35-98
1.0 INTRODUCTION
AND
SUMMARY
This document presents analytical results for a postulated large break loss-of-coolant accident (LOCA) for the Kewaunee reactor operating with ENC fuel.
The analysis was performed to determine the axially dependent linear heat generation rate (LHGR) limits for Kewaunee (i.e.,
the K(Z) curve).
The analyses assume a reactor operating power of 1683 MWt (1650 MWt plus 2% power uncertainty), and use of Exxon Nuclear Company's (ENC's) fuel.
The calculations were made for the double-ended cold leg guillotine break with a discharge coefficient of 0.4 (0.4 DECLG), identified in the previous analyses as the most limiting break (1,2,3,4),
1 l
The LOCA analyses were performed for a full core of ENC fuel using the EXEM/PWR ECCS evaluation model(5), with the R0DEX2 computer model for evaluating the rod stored energy and fission gas release (6). The EXEM/PWR ECCS evaluation model includes the NRC fuel swelling and flow blockage model, NUREG-0630(7),
The analyses are applicable to a five percent (5%) average steam generator (SG) tube plugging.
The maximum allowable linear heat generation rate (including the 1.02 f actor for power uncertainty) is 14.76 kW/ft, corres-T ponding to a maximum total power peaking factor of 2.28 (F g), and nuclear enthalpyriseof1.55(FfH)(4)-
The present LOCA ECCS analyses were performed for Beginning-of-Cycle (B0C) fuel and exposed fuel at End-of-Cycle (EOC) with a conservatively low peak average rod burnup of 10,000 MWD /MTM to maximize peak stored energy. Power shapes representative of, or conservative with respect to, the most top peaked shapes anticipated at the exposure analyzed were used. These power shapes are shown in Figure 1.1 and compared to the Fg(Z) limit.
LOCA analyses using cosine-shaped axial power profile peakings were not performed, since cal-culations with the cosine power shape had been previously performed (4) and had been used as the basis for setting the maximum to+.al peaking at 2.28.
The calculational basis cod results of the present analysis are summarized in Table 1.1.
The maximum calculated PCT is equal to 18870F, and occurs at 59
2 XN-NF-85-98 seconds from the start of the transient at a location 7.63 feet from the bottom of the active core, with a total metal-water reaction less than one percent.
The 18870F PCT result includes a 720F temperature correction to allow for the use of NRC interim upper plenum injection model(8) as modified by West-inghouse(9).
The results of the analyses show that within the limits established, the Kewaunee nuclear reactor satisfies the criteria specified by 10 CFR 50.46(10) for operation at the rated system power level and with the steam generator tube plugging up to 5%.
For breaks up to and including the double-ended severance of a reactor cold leg coolant pipe, the Emergency Core Cooling System for the Kewaunee unit will T
meet the Acceptance Criteria as presented in 10 CFR 50.46, with the 2.28 F g and1.55FdH limits. The criteria are as follows:
(1) The calculated peak fuel element clad temperature does not exceed the 22000F 1imit.
(2) The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent 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 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.
I Table'l.1 Kewaunee LOCA-ECCS Analysis Results - K(Z)
EOC 10000 MWD /MTM Peak Analysis Results BOC Average Rod Exposure-Peak C1ad. Temperature-(PCT), OF***
1887 1816 APCT for UPI, OF
-72
-70 Time of PCT, sec.
59 150-Peak Clad Temperature Location, ft...
7.63 10.88 Local Zr/H O Reaction (max.), %*
2.6 2.5 2
Local Zr/H O Location, ft. from bottom 7.63 10.88 2
Total H2 Generation, % of Total Zr Reacted
<1.0
<l.0 Hot Rod Burst Time, sec.
42 44.7
~
Hot Rod Burst Location, ft.
7.63 10.125 Peak Power Location, ft.
7.63 10.25 Calculational Basis License Core Power, MWt 1650 1650 Power Used for Analysis, MWt**
1683 1683
^
Peak Linear Power for Analysis, kW/ft**
14.4 14.0 T
- Total Peaking Factor, F g 2.23 2.16 AH 1.79 1.55 Enthalpy Rise, Nuclear, F Steam Generator Tube Plugging (%)
5.00 5.00
?
Y Computer value at 380 seconds 3
Including 1.02 factor for power uncertainties h
Includes APCl for UPI
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5 XN-NF-85-98 2.0 LIMITING BREAK LOCA ANALYSIS This report provides the results of a LOCA-ECCS analysis performed for Kewaunee with total steam generator tube plugging up to 5%. The analytical techniques used are in compliance with Appendix K of 10 CFR 50, and are described in the ENC WREM models(ll), and the Emergency Core Cooling System Evaluation Model Updates: WREM-II(12), WREM-IIA (13) and EXEM/PWR(5,18),
A LOCA break spectrum analysis was performed for a similar Westinghouse two-loop plant, with results reported in XN-NF-78-46(1). The limiting LOCA break was determined to be a large double-ended guillotine break of the cold leg, with a discharge coefficient of 0.4 (0.4 DECLG).
2.1 LOCA Analysis Model The Exxon Nuclear Company EXEM/PWR ECCS evaluation model(5) was used to perform the analyses. This model consists of the following computer codes:
RODEX2(6) code for initial rod stored energy and internal fuel rod gas inventory; RELAP4-EM(14) for the system blowdown and hot channel blowdown
-calculations; CONTEMPT-LT/22 as modified in CSB 6-1(15) for computation of containment backpressure; REFLEX (5,7,16) for computation of system reflood; and T00DEE2(5,7,17) for the calculation of final fuel rod heatup. The quench and heat transfer coefficient models used in the reflood portion of the transient are based on the Fuel Cooling Test Facility (FCTF) test data and are reported in reference 18. The NRC upper plenum injection (UPI) interim model, developed by the NRC Staff (8) and modified by Westinghouse (9), was utilized.
The Kewaunee nuclear reactor is a two-loop Westinghouse pressurized water reactor with an upper plenum injection and dry containment.
The reactor coolant system is nodalized into control volumes representing reasonably homogeneous regions, interconnected by flow-paths or " junctions" as described in XN-NF-77-25(A)(15). The system nodalization is as depicted in Figure 2.1.
6 XN-NF-85-98 j
l The pump performance characteristic curves are supplied by the NSSS vendor.
1 Five percent of the steam generator tubes are assumed to be plugged in each generator.
The transient behavior was determined from the governing conservation equations for mass, energy, and momentum.
Energy transport, flow rates, and heat transfers are determined from appropriate correlations.
System input parameters are given in Table 2.1.
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 LOCA/ECCS analysis presented in this report supports the current K(Z) function developed by the NSSS vendor for the portion of the function defined by the large break LOCA. Where small break LOCA is limiting, the K(Z) curve is defined such that the Linear Heat Generation Rates (LHGRs) determined by the NSSS vendor analysis are un-changed. The K(Z) function is shown in Figure 2.36. The analysis of the loss-of-coolant accident is performed at 102 percent of rated power.
The fuel desi,gn parameters are shown in Table 2.2.
LOCA/ECCS calculations were performed at B0C and E0C conditions to bound the power distributions anticipated to occur. Two power shapes representative of the most top peaked anticipated at BOC and E0C conditions were chosen from a study of a number of different reactors and cycles.
The B0C axial power distribution (Figure 2.2) was analyzed in conjunction with a conservative in excess of the Technical Specification limit. This was done value for FAH in order to be able to analyze with a peak Fg at the desired Technical Specification limit.
The E0C axial power distribution (Figure 2.3) was conservatively increased in value in the top portion of the core and decreased at the bottom portion of the core in order to be analyzed with a peak Fg at the Technical Specification limit and with an FAH value equal to the Technical Specification limit. The E0C case was analyzed with a conservatively low rod burnup. The use of a low rod burnup resuits in a higher stored energy than would be anticipated to occur in conjunction with the axial power shape utilized.
LOCA analyses using cosine-shaped axial power peakings were not
7 XN-NF-85-98 performed since analyses with these shapes were previously reported. These power shapes are shown in Figure 1.1 and compared to the Fg(Z) limit.
2.2 Results Table 2.3 presents the timing and sequence of events as determined for the large guillotine break with a discharge coefficient of 0.4.
Comparison of these results with the previous LOCA-ECCS analysis for ENC fuel shows very slight change in the event times. Figures 2.4 through 2.10 present plotted results for system blowdown analysis (4). Unless otherwise noted on the figures, time zero corresponds to the time of break initiation. Figure 2.11 presents calculated containment backpressure time history (4). Figures 2.12 through 2.23 present results for the hot channel blowdown calculations.
Figures 2.24 and 2.25 show the normalized power calculation results.
The reflood calculation results are shown in Figures 2.26 through 2.33.
The maximum peak cladding temperature (PCT) calculated for the 0.4 DECLG break at BOC is 18870F (Figure 2.34).
This value includes a 720F temperature reduction associated with the use of the NRC interim upper plenum injection (UPI) model as modified by Westinghouse.
The maximum local metal-water reaction in this case is 2.6% after 380 seconds, and the total core metal-water reaction is less than 1%. The PCT location is at an elevation of 7.63 feet from the bottom of active core. For ENC fuel at E0C, the PCT is 18160F (Figure 2.35)- including -700F for UPI effect, occurring at 10.88 feet elevation relative to the bottom of the active core. The local metal-water reaction is 2.5%, with a total metal-water reaction of less than 1%. The peak cladding temperatures shown in Figures 2.34 and 2.35 do not include the UPI corrections.
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8 XN-NF-85-98 Table 2.1 Kewaunee System Data Primary Heat Output, MWt 1650*
Primary Coolant Flow, lbm/hr 6.82 x 107 Operating Pressure, psia 2,250 Inlet Coolant Temperature, OF 534 Reactor Vessel Volume, ft3 2406 Pressurizer Volume, Total, ft3 1000 Pressurizer Volume, Liquid, ft3 600 Accumulator Volume, Total, ft3 (each of two) 2000 Accumulator Volume, Liquid, ft3 1250 Accumulator Trip Point Pressure, psia 714.7 2
48,925**
Steam Generator Secondary Heat Transfer Area, ft Steam Generator Secondary Flow, lbm/hr 3.56 x 106 Steam Generator Secondary Pressure, psia 750 Reactor Coolant Pump Head, ft (Design) 277 Reactor Coolant Pump Speed, rpm (Design) 1190 2
Moment of Inertia, lbm-ft / rad 80,000 Cold Leg Pipe, I.D.,
in 27.5 Hot Leg Pipe, I.0.,
in 29 Pump Suction Pipe, I.D., in 31
- Primary Heat Output used in RELAP4-EM Model = 1.02 x 1650 = 1683 MWt.
- Includes 5% SG tube plugging.
9 XN-NF-85-98 J
Table 2.2 Fuel Design Parameters Cladding, 0.D., in.
0.424 Cladding, I.D.,
in.
0.364 Cladding Thickness, in.
0.030 Pellet 0.0., in.
0.3565 Diametral Gap, in.
0.0075 l
Pellet Density, % TD 94.0 Active Fuel Length, in.
144.0 Rod Pitch, in.
0.556
10 XN-NF-85-98 Table 2.3 Kewaunee LOCA-ECCS Analysis Results, Event Times Event Time (sec.)
Start 0.00 Break Initiation
.05 Safety Injection Signal
.65 Accumulator Injection, Broken Loop 4.8 Accumulator Injection, Intact Loop 8.8 End-of-Bypass 22.7 Safety Injection Flow 25.7 Start of Reflood 36.9 Accumulator Empties, Intact Loop 43.1 Peak Clad Temperature Reached -
BOC 59.0 E0C 150.0 i
mm
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v 47 XN-NF-85-98
3.0 REFERENCES
1.
XN-NF-78-46, "ECCS Large Break Spectrum Analysis for Prairie Island Unit 1 Using ENC WREM-IIA PWR Evaluation Model," November 1978.
2.
XN-NF-79-1, "ECCS Analysis for Kewaunee Using ENC WREM-IIA PWR Evalua-tion Model," January 1979.
3.
Kewaunee Nuclear Generating Plant Final Safety Analysis Report.
4.
XN-NF-84-31, Revision 1, "Kewaunee High Burnup Safety Analysis: Limiting Break LOCA and Radiological Consequences," October 1984.
5.
XN-NF-82-20(P), Revision 1, " Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," August 1982; Supplement 1. March 1982; Supplement 2, March 1982; Supplement 3, June 1985; and Supplement 4, July l
1984.
l 6.
XN-NF-81-58(A), Revision 2 & Supps.1 and 2, "RODEX2: Fuel Rod Thermal-Mechanical Response Evaluation Model," March 1984.
l 7.
XN-NF-82-07(P)(A), Revision 1, " Exxon Nuclear Company ECCS Cladding l
Swelling and Rupture Model," August 1982.
8.
U.S. Naclear Regulatory Commission, " Safety Evaluation Report on Interim ECCS Evaluation Model for Westinghouse Two-Loop Plants," Analysis Branch, Division of System Safety, Office of Nuclear Reactor Regulation, l
November 1977.
1 9.
Letter. L.0. Mayer to Director of Nuclear Reactor Regulation, February 24, 1978 (Docket No. 50-282 and 50-306).
- 10. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50.
- 11. XN-75-41, " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model," July 1975, and Supplements and Revisions thereto.
- 12. XN-76-27, " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-II " July 1976; Supplement 1, September 1976; and Supplement 2. November 1976.
- 13. XN-NF-78-30(A), " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-IIA," May 1979.
-~-- -
48 XN-NF-85-98
- 14. Letter, T.A.
Ippolito (USNRC) to W.S. Nechodom (ENC), "SER for ENC RELAP4-EM Update," March 1979.
- 15. XN-NF-77-25(A), " Exxon Nuclear Company ECCS Evaluation of a 2-Loop Westinghouse PWR with Dry Containment using the ENC WREM-II ECCS Model -
Large Break Example Problem," September 1978.
- 16. XN-NF-CC-52(P), Volume II, " REFLEX and REFLEX /UPI PWR Reflood Computer Program - User's Manual," February 1985.
17.
G.N. Lauben, "T000EE2: A Two-Dimensional Time Dependent Fuel Element Thermal Analysis Program," NRC Report NUREG-75/057, May 1975.
- 18. XN-NF-85-16(P), Volume II, "PWR 17x17 Fuel Cooling Test Program: Reflood Quench, Carryover, and Heat Transfer Correlations," May 1985.
J l
l
~
XN-NF-85-98 Issue Date: 9/26/85 KEWAUNEE LIMITING BREAK K(Z)
LOCA/ECCS ANALYSIS Distribution M. J. Ades J. C. Chandler R. A. Copeland L. J. Federico R. C. Gottula J. S. Holm W. V. Kayser J. M. Ross P. J. Valentine H. E. Williamson Wisconsin Public -Service /JM Ross (20)
Document Control (5)
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