ML20151U887

Limiting Break K(Z) Loca/Eccs Analysis

ML20151U887
Person / Time
Site:	Cook
Issue date:	11/30/1985
From:	Ades M, Copeland R, Holm J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17324A587	List: ML20151U887
References
XN-NF-85-115, NUDOCS 8602110135
Download: ML20151U887 (56)

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XN-NF-85-ll5 l

D.C. COOK UNIT 1 LIMITING BREAK KLZ;l LOCA/ECCS ANALYSIS 1

1 NOVEMBER 1985 l

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XN-NF-85-115 Issue Date: 11/1/85 D. C. COOK UNIT 1 LIMITING BREAK K(Z) LOCA/ECCS ANALYSIS Contributors: G.R. Sawtelle (Energy Incorporated)

R.O. Hentzen (Energy Incorporated)

_ Prepared by:

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, M. J.7 des Model Development

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Concur: [, .  % ic!fi[f3' J,f5. Holm, Managet '

PGR Safety Analysis Concur: ,, df gr-PWR Reload Licensir.g Approved by: # [, ,.,,w fo/3 //ff H. E. Williamson, Manager Licens,ing & Safety Engineering Concur: \W iMhh' 16 !3/[S J. '{i. Mdrgan, Mdnager Customer Services Engineering

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Approved by:

G. L. Ritter,~ Manager

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Fuel Engineering ~& Technical Services gf 1

ERON~ \UC_ A9 COV 3A\Y \C.

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i NUCLEAR REGUt.ATORY COMMIS$3ON DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THis DOCUMENT PLEASE READ CAREFULLY This technical report was derived 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 facilitete safety ana!yses by licensees of the USNRC which utilize Exxon Nuclear fabricated reload fuel or other tect.. .al services provided by Exxon Nuclear for licht water power rea: tors and it is true and correct to the best of Exxon Nucteer's knowledge, information, and belief. The informaticn contained herein may be used by the USNRC .

In its review of this report, and by licensees or appicants before the -

USNRC which are customers of Exxon Nudeer in their demonstration of comoliance with the USN RC's regulations.

Without derogating from the foregoirg neither Exxon Nuclear nor any person acting nn 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 information, apparatus, method, or process dittosed in this document will not infnnge privately owned rights i

or i

B. Assumes any liabilities with resosct to the use of, or for darrages resulting from the use of, any information, ap-paratus, method, or process drsclosed in this document, h

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TABLE OF CONTENTS p

. Section Page

? !1.0 INTR 000CTION'................................................ 1.

2. 0. . K ( Z ) LOCA ANALY S I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -5 p

2.'1' LOCA An'alysis Mode 1.........................................

5 2.2.-Results..................................................... 6

- 3 . 0 ' : R E F E R E NC E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47

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1 Section Page '

1.1 D.C. Cook Unit 1 LOCA-ECCS Analysis Results................. 3 2.1 D.C. Cook Unit 1 System Data................................ 8

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2.2 Fuel Design Parameters...................................... 9 2.3 D.C. Cook Unit 1 LOCA-ECCS Analysis Results, Event Times.... 10 i

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f, iii XN-NF-85-11'5 LIST OF FIGURES

Figure. Page 1.1- Comparison of Power Distributions Analyzed to Limits........ 4 2.1 Blowdown System Nodalization for 0.C. Cook Unit 1........ .. 11-

.2.2: Axial Peaking Factor versus Relative. Height, 1.0 DECLS Break, 80C......................................................... :12 2.3 -Axial Peaking Factor versus Relative Height,1.0 DECLS Break, E0C......................................................... 13 2.4. Downcomer Flow Rate,'1.0 DECLS Break........................ 14 2.5- Upper Pl_enum Pressure, 1.0 DECLS Break...................... 15 t

2.6 Average Core Inlet Flow,~1.0-0ECLS Break.................... 16 '

2.7 Av'erage Core Outlet Flow, 1.0 DECLS Break................... 17 k 2.8 ' Total Break Flow', 1.0 DECLS Break........................... 18 y

2.9 Break Junction-Enthalpy, 1.0 DECLS.......................... 19 2.10 Flow from Intact Loop Accumulator, 1.0 DECLS-Break.......... 20 2.11 Containment Back Pressure,:1.0 DECLS Break.................. 21

'2.12 Hot Channel Heat Transfer Coefficient, 1.0 DECLS Break, B0C................... .....................................

22 2.13 Clad Surface Temperature, 1.0 DECLS Break, B0C.............. 23 2.14 Depth of' Metal-Water Reaction,~1.0 DECLS Break, B0C......... 24

-2.15 Hot Channel Average Fuel Temperature, 1.0 DECLS Break, 800.. 25 2.16 Hot Assembly Inlet Flow, 1.0 DECLS' Break, 80C............... 26 2.17 Hot A'ss'embly. Outlet Flow, 1.0 DECLS Break, 80C.............. 27 2.18 Hot Channel Heat' Transfer Coefficient, 1.0 DECI.S, Break, E0C......................................................... 28 2.19 Clad Surface Temperature, 1.0 DECLS Break, E0C.............. 29-2.20 Depth of Metal-Water Reaction, 1.0 DECLS Break, E0C......... 30 2.21 Hot Channel Average Fuel Temperature, 1.0 DECLS Break,

'E0C..........................................~............... 31 2.22 Hot Assembly Inlet Flow, 1.0 DECLS Break, E0C............... 32-2.23HotAssemblyOutletFlow,1.0bECLS-Brcak,E0C.............. 33-2.24 Normalized Power, 1.0 DECLS Break, 80C...................... 34

iv XN-NF-85-115 LIST OF FIGURES (Cont.)

Figure Page 2.25 Normalized Power, 1.0 DECLS Break, E0C...................... 35 2.26 Reflood Core Mixture Level, 1.0 DECLS Break, B0C. . . . . . . . . . . . 36 2.27 Reflood Downcomer Mixture Level, 1.0 DECLS Break, B0C....... 37 2.28 Reflood Upper Plenum Pressure, 1.0 DECLS Break, B0C......... 38 2.29 Core Flooding Rate, 1.0 DECLS Break, B0C. . . . . . . . . . . . . . . . . . . . 39 2.30 Reflood Core Mixture Level, 1.0 DECLS Break, E0C............ 40 2.31 Reflood Downcomer Mixture Level, 1.0 DECLS Break, E0C....... 41 2.32 Reflood Upper Plenum Pressure,1.0 DECLS Break, E0C. . . . . . . . . 42 2.33 Core Flooding Rate, 1.0 DECLS Break , E0C. . . . . . . . . . . . . . . . . . . . 43 2.34 T00DEE2 Cladding Temperature vs Time, 1.0 DECLS Break, B0C......................................................... 44 2.35 T00DEE2 Cladding Temperature vs Time,1.0 DECLS Break, E0C......................................................... 45 2.36 Hot Channel Factor Normalized Envelope for Fg = 2.04, 46 K(Z) Function...............................................

1 XN-NF-85-115 l i

1.0 INTRODUCTION

AND SUfeiA').Y This document presents analytical results for a postulated large break loss-of-coolant accident (LOCA) for the D.C. Cook Unit I reactor operating with ENC fuel. The analysis was performed to determine the axial dependence of the linear heat generation rate (LHGR) limits for D.'C. Cook Unit 1 (i.e., the K(Z) curve). The analyses assume a reactor operating power of 3315 MWt (3250 MWt plus 2% power uncertainty), and use of Exxon Nuclear Company's (ENC's) 15x15 fuel. The calculations were made for the double-ended cold leg split break with a discharge coefficient of 1.0 (1.0 DECLS), identified in previcus k analyses as the most limiting break (1.2).

The LOCA analyses were performed for'a full core of ENC fuel using the models f

outlined in section 2.1. The maximum allowable linear heat generation rate (including the 1.02 factor for power uncertainty) is 14.3 kW/f t, corres-ponding to a maximum total power peaking factor of 2.04 (FT g), and nuclear enthalpyriseof1.51(F{H)-

)

The present LOCA ECCS analyses were performed for Beginning-of-Cycle (60C) fuel (2,000 mwd /MTM) and exposed fuel at End-of-Cycle (EOC) with a con-servatively low peak average rod burnup of 9,000 mwd /MTM to maximize peak stored energy. A cosine axial power shape was used at the BOC exposure and a power shape representative 'of, or conservative with respect to, the anti-cipated power shapes at the E0C exposure was used. These power shapes are I

shown in Figure 1.1 and compared to the Fg(Z) limit. ,

) The calculational basis and results of the present analysis are summarized in Table 1.1. The maximum calculated PCT is equal to 20550F, and occurs at 77 seconds from the start of the transient at a location 6.00 feet from the bottom of the active core, with a total metal-water reaction less than one percent.

2 XN-NF-85-115 As in the previous analysis,(2) it was assumed that one of the LPSI pumps had failed. An earlier sensitivity study (6) showed that the peak clad temperature (PCT) increased when a conservative estimate of maximum LPSI flow was assumed.

A sensitivity analysis using the maximum LPSI flow assumption was therefore performed with the current power distribution and models. The analysis results indicate that the maximum calculated PCT is increased less than 650F.

Inclusion of the APCT value in the PCT results calculated in the present analysis would still support the axial dependence shown. in Figure 1.1 The results of the analyses show that within the limits established, the D.C. Cook Unit I nuclear reactor satisfies the criteria specified by 10 CFR 50.46(16) for operation at the rated system power level. The criteria are as follows:

(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 percent of the total amount of zircaloy in the reactor.

(3) The cladding temperature transi'ent 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.

= _-

Table 1.1 D.C. Cook Unit 1 LOCA-ECCS Analysis Results* - K(Z)

BOC 2000 MWD /MTM Peak E0C 9000 MWD /MTM Peak Analysis Results _ Average Rod Exposure Average __ Rod Exposure Peak Clad Temperature (PCT), OF 2055 1854 Time of PCT, sec. 77 200 Peak Clad Temperature Location, ft. 6.00 11.00 Local Zr/H 2O Reaction (max.), %** 4.2 2.40 Local Zr/H 2 O Location, ft. from bottom 6.00 11.00 Total 112 Generation, % of. Total Zr Reacted 1.0 1.0 Hot Rod Burst Time, sec. 50 '57.6 Ilot Rod Burst Location, f t. 6.00 9.75 w Peak Power Location, ft. 6.00 9.75 Calcul_ational Basis License Core Power, MWt 3250 3250 Power Used for Analysis, MWt*** 3315 3315 Peak Linear Power for Analysis, kW/ft*** 14.3 13.7 Total' Peaking Factor, FTg 2.04 1.95 EnthalpyRise. Nuclear,F{n 1.51 1.51 $!

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• Results calculated by using the single LPSI pump failure assumption. Sensitivity studies [.

indicate an increase of up to 650F for the assumption of maximum LPSI flow. 3;

• • Computer value at 400 seconds

• • • Including 1.02 factor for power uncertainties

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5 Figure 1.1 Comparison of Power Distributions Analyzed to Limits

l 5 XN-NF-85-115 2.0 K(Z) LOCA ANALYSIS This report provides the results of a LOCA/ECCS analysis performed for the O.C. Cook Unit I reactor operating with ENC 15x15 fuel. The purpose of this analysis was to define the axial dependence of the LOCA limit. The analytical techniques used are in compliance with Appendix K of.10 CFR 50, and are described in the ENC WREM models(7), and the Emergency Core Cooling System '

Evaluation Model Updates: WREM-II(8), WREM-IIA (9), EXEM/PWR(3), and the FCTF ,

reflood correlations (10,17),

A LOCA break spectrum analysis was performed for D.C. Cook Unit 1, with results reported in XN-NF-76-51(15). The limiting LOCA break was determined to be an equivallnt double-ended split break of the cold leg (1.0 DECLS).

2.1 LOCA Analysis Model The Exxon . Nuclear Company EXEM/PhR ECCS evaluation model(3) was used to i

perform the analyses. This model consists of the following computer codes:

RODEX2(4) code for initial red stored energy and internal fuel rod gas l t inventory; RELAP4-EM(ll) for the system blowdown and hot channel blowdown l

calculations; ICECON(12) for the computation of ice condenser containment backpressure; REFLEX (3,5,13) for computation of system reflood; and T00-DEE2(3,5,14) 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 references 10 and 17.

d The O.C. Cook Unit I nuclear reactor it a four-loop Westinghouse pressurirzed water reactor with an ice condenser contairment. The reactor coolant system is nodalized into control volumes representing reasonably homogeneous re-gions, interconnected by flow-paths or " junctions." The system nodalization is as depicted in Figure 2.1. The pump performance characteristic curves are

6- XN-NF-85-ll5 supplied by the NSSS vendor. 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 analysis of the loss-of-coolant accident is performed at 102 percent of rated power. The fuel design parameters are shown in Table 2.2.

2.2 Results The LOCA/ECCS anal *ysis 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 unchanged. The K(Z) function is shown in Figure 2.36.

LOCA/ECCS calculations were performed at BOC and EOC conditions to bound the power distributions anticipated to occur. The BOC chopped-cosine axial power distribution (Figure 2.2) was analyzed by using the FAH and Fg power peaking factors that correspond to the Technical Specification limit. The EOC 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 relative to the axial power shape anticipated at E0C conditions 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 results in a higher stored energy than awould be anticipated to occur in' conjunction with the axial power shape utilized. The axial power shapes used in the analyses at 80C and E0C are shown in Figure 1.1 and compared to the Fg(Z) limit.

7 XN-NF-85-115 Table 2.3 presents the timing and sequence 0. events as determined for the split break with a discharge coefficient of 1.0. Comparison of the system

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results with the previous LOCA/ECCS analysis shows very slight change in the event times. Figures 2.4 through 2.10 present plotted results for system blowdown analysis. Unless otherwise noted on the figures, time zero corresponds to the time of break initiation. Figure 2.11 presents calculated containment backpressure time history. 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.

t The maximum peak cladding temperature (PCT) calculr.ted for the 1.0 DECLS break at BOC is 20550F (Figure 2.34). The maximum local metal-water reaction in this case is 4.2% after 400 seconds, and the total core metal-water reaction is less than 1%. The PCT location is at an elevation of 6.00 feet from the bottom of active core. For ENC fuel at E0C, the PCT is 18540F (Figure 2.35),

occurring at 11.00 feet elevation relative to the bottom of the active core.

The local metal-water reaction is 2.4%, with a total metal-water reaction of less than 1%.

As in the previous analysis (2), it was assumed that one of the LPSI pumps had failed. An earlier sensitivity study showed that the peak clad temperature increased when a conservative est; mate of the maximum LPSI flow was assumed.

A sensitivity analysis using the maximum LPSI flow assumption was therefore performed with the current power shapes and models. The results indicated that the calculated PCTs increase 650F and 220F at BOC and EOC respectively when a conservative estimate of maximum LPSI flow is assumed. Inclusion of these APCT values in the PCT results calculated in the present analysis would still support the K(Z) shown in Figure 2.36.

8 XN-NF-85-115 Table 2.1 Donald C. Cook Unit 1 System Data Primary Heat Output, MWt 3250-*.

Primary Coolant Flow, Ibm /hr 135.6 x 106 Primary Coolant' Volume, f t3 11,890.

Operating Pressure, psia 2250.

Inlet Coolant Temperature, OF 536.3 Reactor Vessel Volume, ft3 4945.

Pressurizer Volume, Total, ft3 1800.

Pressurizer Volume, Liquid, ft3 1080.

Accumulator Volume, Total, ft3 (each of four) 1350.

Accumulator Volume, Liquid, ft3 929.

Accumulator Pressure, psia 636.

Steam Generator Heat Transfer Area, ft2 51,500.

Steam Generator Secondary Flow, lbm/hr 3.53 x 106 Steam Generator Secondary Pressure, psia 758.

Reactor Coolant Pump Head, ft 277.

Reactor Coolant Pump Speed, rpm 1190.

Moment of Inertia, ibm-ft /2 rad 82,000.

Cold Leg Pipe, 1.0., in 27.5 Hot Leg Pipe, I.D., in 29.0 Pump Suction Pipe', I.D., in 31.0 Fuel Assembly Rod Diameter, in** 0.424 Fuel Assembly Rod Pitch, in** 0.563 Fuel Assembly Pitch, in** 8.466 Active Core Height, in** 144.0 Fuel Heat Transfer Area, ft2 52,200 Fuel Total Flow Area, ft2 50.91

• Primary heat output used in RELAP4-EM Model = 1.02 x 3250 = 3315 MWt

• • ENC fuel parameters.

9 XN-NF-85-115 -

l s Table 2.2 Fuel Design Parameters Cladding, 0.D., in 0.424 Cladding, 1.0., in 0 364 Cladding Thickness, in 0.030 Pellet 0.0., in 0.3565 Diametral Gap, in 0.0075 Pellet Density, %TD 94.0

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Active Fuel Length, in 144'.0 i

Rod Pitch, in- 0.563 nW l

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10 XN-NF-85-115 Table 2.3 D.C. Cook Unit 1 LOCA/ECCS Analysis Results, Event Times Event Time (sec)

. Start 0.00 Break Initiation .05 Safety Injection Signal .65 Accumulator Injection, Broken Loop 2.0 Accumulaior Injection, Intact Loop 15.5 End-of-Bypass 22.9 Safety Injection Flow 25.7 Accumulator Empties, Broken Loop 36.2 ,

Start of Reflood 39.6 Accumulator Empties, Intact Loop 50.7 Peak Clad Temperature Reached -

B0C (Chopped cosine axial power peaking) 76.7 E0C (Upskewed axial power peaking) 200.1

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3.0 REFERENCES

! 1. XN-NF-81-07, "LOCA/ECCS Reanalysis for D.C. Cook Unit 1 Using the ENC WREM-IIA PWR ECCS Evaluation Model," Exxon Nuclear Company, Inc.,

Richland, WA, February 1981.

2. XN-NF-83-61, "D.C. Cook Unit 1 LOCA/ECCS Analysis for Extended Expo-sure," Exxon Nuclear Company, Inc., Richland, WA, August 1983.

3. 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 1984, Exxon Nuclear Company, Inc., Richland, WA.

4. XN-NF-81-58(A), Revision 2 & Supps.1 and 2, "RODEX2: Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, Inc.,

Richland, WA, March 1984.

5. XN-NF-82-07(P)(A), Revision 1, " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Inc., Richland, WA, August 1982.

6. Letter, H.G. Shaw (ENC) to H.L. Sobel ( AEP), HGS:062:82, dated Feb-ruary 23, 1982.

7. XN-75-41, " Exxon Nuclear Company UREM-Based Generic PWR ECCS Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA, July 1975, and Supplements and Revisions thereto.

8. 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, Exxon Nuclear Company, Inc., Richland, WA.

9. XN-NF-78-30(A), " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-IIA," Exxon Nuclear Company, Inc.,

Richland, WA, May 1979.

10. XN-NF-85-16(P), Volume II, "PWR 17x17 Fuel Cooling Test Program:

Reflood, Quench, Carryover, and Heat Transfer Correlations," Exxon Nuclear Company, Inc., Richland, WA, May 1985.

11. Letter, T.A. Ippolito (USNRC) to W.S. Nechodom (ENC), "SER for ENC RELAP4-EM Update," dated March 1979.

12. XN-CC-39, Revision 1,'"ICECON: A Computer. Program Used to Calculate Containment Backpressure for LOCA Analysis (Including Ice Condenser Plants)," Exxon Nuclear Company, Inc., Richland, WA, November 1977.

l i _ _._ _.______ _______.___.._..___.m_-

48 XN-NF-85-115

13. XN-NF-CC-52(P), Volume II, " REFLEX and REFLEX /UPI PWR Reflood Computer Program - User's Manual," Exxon Nuclear Company, Inc., Richland, WA, February 1985.

14. G.N. Lauben, "T00DEE2: A Two-Dimensional Time Dependent Fuel Element Thermal Analysis Pregram," NRC Report NUREG-75/057, May 1975.

15. XN-NF-76-51, " Donald C. Cook Unit 1 LOCA Analysis Using the ENC WREM-Based PWR ECCS Evaluation Model (ENC-WREM-II)," Exxon Nuclear Company, Inc., Richland, WA, October 1976.

16. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10CFR50.46 and Appendix K of 10CFR50.

17. XN-NF-85-105(P), " Scaling of FCTF-Based Reflood Heat Transfer Cor-relation for Other Bundle Designs," Exxon Nuclear Company, Inc.,

Richland, WA, October 1985.

XN-NF-85-115 Issue Date: 11/1/85 D. C. COOK UNIT 1 LIMITING BREAK K(Z) LOCA/ECCS ANALYSIS Distribution M. J. Ades R. A. Copeland J. S. Holm S. E. Jensen W. V. Kayser H. G. Shaw T. Tahvili H. E. Williamson American Electric Power /HG Shaw (6)

Document Control (5) u . .. _ _ . _ _ _ _ _ _ _ _