ML17319B412

DC Cook Unit 2 LOCA ECCS Analysis Using Exem/Pwr Large Break Results.

ML17319B412
Person / Time
Site:	Cook
Issue date:	04/30/1982
From:	Braun D, Jensen S, Tahvili T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17319B410	List: ML17319B409 ML17319B411 ML17319B412
References
XN-NF-82-35, NUDOCS 8207080407
Download: ML17319B412 (118)

Text

XN-NF-82-35 Issue Date: O4/3p/B2 DONALD C. COOK UNIT 2 LOCA ECCS ANALYSIS USING EXEM/PWR LARGE BREAK RESULTS Prepared by:

a vl 1 NSSS Systems Analysis Prepared by:

r aun Fuel Response Analysis Reviewed by:

nsen, anage NSSS Systems Analysis (ECCS)

Concur: mar y/~~/p~

W. V. ayser, Manager Fue Response Analysis Concur:

G. F. ws ey, an er Reload Fuel Licensing Approved by:

organ, an er Licensing 8 Safety Engineering Approved by:

o er, ana er Fue Engineering 5 Technical Services gf E)f(ON NUGLEAR GOMPANV, Inc.

8207080407 8 PDR ADQCK 0 soooa~e P

PDR l

NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was rlerived 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 reload 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 knowledge, 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 demonstration of compliance with the USNRC's regulations.

Without derogating from the foregoing, 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 disclosed in this document will not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for danages resulting from the use of, any information, ap.

paratus, method, or process disclosed in this document.

XN- NF- FOO, 766

XN-NF-82-35 TABLE OF CONTENTS Section ~Pa e

1.0 INTRODUCTION

AND

SUMMARY

........................... 1 2.0 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOCA) SPECTRUM ANALYSIS ........................... 5 2.1 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 2.2 LOCA ANALYSIS MODEL ........................... 6 2 .3 RESULTS ....................................... 8 3.0 EXPOSURE SENSITIVITY ............................... 37

4.0 CONCLUSION

S ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ o 51

5.0 REFERENCES

......................................... 52

XN-NF-82-35 LIST OF TABLES Table No. ~Pa e 1.1 Large Break Spectrum Analysis Results for Donald C. Cook Unit 2 with ENC Fuel 1.2 Exposure Sensitivity Results for D. C. Cook Unit 2 XN-1 Reload ................................. 4 2.1 Donald C. Cook Unit 2 Large Break Events ........... 1O 2~ 2. D. C. Cook Unit 2 System Input Parameters .......... 11

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XN-NF-82-35 LIST OF FIGURES Fi ure No. ~Pa e 2.1 System Blowdown Nodalization for the Donald C. Cook Unit 2 PMR ........................... 12 2.2 Axial Peaking Factor versus Rod Length for Donald C. Cook Unit 2 ECCS Analysis ................. 13 2.3 Upper Plenum Pressure - D. C. Cook Unit 2; D ECLG (CD=1.0) 14 2.4 ,Pressurizer Pressure - D. C. Cook Unit 2; D ECLG (CD 1 0) e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 15

2. 5 Total Break Flow - D. C. Cook Unit 2; D ECLG (CD=1.0) 16 2.6 Average Core Inle't Flow - D. C. Cook Unit 2; D ECLG (CD=1.0) 17 2.7 Average Core Outlet Flow - D. C. Cook Unit 2; D ECLG (CD=1.0) 18 2.8 Downcomer Flow Rate - D. C. Cook Unit 2; D ECLG (CD 1 ~ 0) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 19 2.9 Pressurizer Surge Line Flow - D. C. Cook Unit 2.; DECLG (CD=1.0) .............................. 20 2.10 Flow from Intact Loop Accumulators-D. C. Cook Unit 2; DECLG (CD=1.0) ...........'low

~ ~ ~ ~ ~ ~ ~ 21 2.11 from Broken Loop Accumulator-D. C. Cook Unit 2; DECLG (CD=1.0) ............ ~ ~ ~ ~ ~ ~ ~ 22 2.12 Hot Channel Average Fuel Temperature PCT Node D. C. Cook Unit 2; DECLG (CD=1.0) ................... 23 2.13 Clad Surface Temperature PCT Node-D; C. Cook Unit 2; DECLG (CD=1.0) ................... 24

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1V XN-NF-82-35 LIST OF FIGURES (Cont.)

Fi ure No. ~Pa e 2.14 Depth of Metal-Water Reaction-D. C. Cook Unit 2; DECLG (Cp=l.O) .................. 25 2.15 Hot Channel Heat Transfer Coefficient PCT Node-D. C. Cook Unit 2; DECLG (Cp=1.0) .................. 26 2.16 Hot Assembly Inlet Flow-D. C. Cook Unit 2; DECLG (Cp=l.O) .................. 27 2.17 Hot Assembly Outlet Flow-D. C. Cook Unit 2; DECLG (Cp=1.0) ......-............ 28 2.18 ICECON Containment Backpressure-D. C. Cook Unit 2; DECLG (CD=1.0) .................. 29 2.19 Normalized Power - D. C. Cook Unit 2; DECLG (Cp=1.0) ..................................... 30 2.20 Core Flooding Rate - D. C. Cook Unit 2; DECLG (Cp=1.0) ..................................... 31 2.21 Reflood Downcomer Mixture Level-D. C. Cook Unit 2; DECLG (Cp=l.O) .................. 32 2.22 Reflood Core Mixture Level-D. C. Cook Unit 2; DECLG (Cp=l.O) .................. 33 2.23 Reflood Upper Plenum Pressure-D. C. Cook Unit 2; DECLG (Cp=l.O) .................. 34 2.24 Reflood Core Saturation Temperature-D. C. Cook Unit 2; DECLG (Cp=1.0) .................. 35 2.25 TOODEE2 Cladding Temperature vs Time-D. C. Cook Unit 2; DECLG (CD=1.0) .................. 36 3.1 RELAP4/Hot Channel Fuel Average Temperature at the PCT Node for 1.0 DECLG Break from Bound lng Fuel Rod History Analysis .................... ~ ~ ~ ~ ~ ~ 39

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XN-NF-82-35 LIST OF FIGURES (Cont. )

Fi ure No. ~Pa e 3.2 RELAP4/Hot Channel Clad Surface Temperature at the PCT Node for 1.0 DECLG Break from Bounding

'uel Rod History Analysis ......................... 40 3.3 RELAP4/Hot Channel Depth of Metal-Water Reaction at the PCT Node for 1.0 DECLG Break from Bounding Power History Analysis .............. 41 3.4 RELAP4/Hot Channel Heat Transfer Coefficient at the PCT Node for 1.0 OECLG Break from Bounding Fuel Rod History Analysis ................ 42 3.5 RELAP4/Hot Channel Hot Assembly Inlet Flow for 1.0 OECLG Break from Limiting Power U'

History Analysis A 1

.................................. 43 3.6 RELAP4/Hot Channel Hot Assembly Exit Flow for 1.0 DECLG Break from Bounding Power History Analys>s .......................................... 44 3.7 REFLEX Core Inlet Flooding Rate for 1.0 DECLG Break from Bounding Fuel Rod History Analysis ..... 45 3.8 REFLEX Downcomer Mixture Level for 1.0 DECLG Break from Bounding Fuel Rod History Analysis ..... 46 3.9 REFLEX Core Mixture Level for 1.0 OECLG Break from Bounding Fuel Rod History Analysis ..... 47 3.10 REFLEX Upper Plenum Pressure for 1.0 DECLG Break from Bounding Fuel Rod.History Analysis ..... 48 3.11 REFLEX Core Saturation Temperature for 1.0 DECLG Break from Bounding Fuel Rod History Analysis ..... 49 3.12 TOODEE/Heatup Peak Cladding Temperature Plot for Rupture Node and PCT Node for 1.0 DECLG Break from Bounding Fuel Rod History Analysis ........... 50

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XN-NF-82-35

1.0 INTRODUCTION

AND

SUMMARY

This document presents analytical results for a spectrum of postulated large break loss-of-coolant accidents (LOCA's), performed for the Donald C. Cook Unit 2 nuclear power plant, operating at 3425 MWt, and fueled by Exxon Nuclear Company (ENC). Also included are results from a bounding exposure calculation for the limiting break LOCA. The analyses, used ENC's pressurized water reactor (PWR) Emergency Core Cooling System (ECCS) evaluation model, EXEM/PWR(> ~2). The EXEM/PWR model has been developed in accordance with NRC 10 CFR 50 Appendix K(3) requirements. The results of the analyses show that within the limits established, the criteria speci-fied by the 10 CFR 50.46 are satisfied.

The break spectrum calculations include five large break LOCA's: three Double Ended Cold Leg Guillotine (DECLG), breaks of the'discharge pipe of the reactor coolant pump, with discharge coefficients (CD) of 1.0, 0.8 and 0.6, and two breaks of a split configuration, and break areas equal to 1.0 and 0.6 times the double ended pipe area of the pump discharge line (DECLS).

The limiting break from the spectrum analysis was shown to be the large guillotine break with a discharge coefficient of 1.0 (1.0 DECLG).

For this break, the Peak Cladding Temperature (PCT) is 2092oF, and occurs at 287 seconds into the accident at a location 9.38 feet from the bottom of the active core. The LOCA break spectrum was performed for the D. C.

Cook Unit 2 operating at 3425 MWt with 17xl7 ENC fuel at the beginning-of-life conditions. The maximum linear heat generation rate (LHGR) was 12.0 kW/ft (including 1.02 factor for power uncertainties), which corresponds

XN-NF-82-35 T

to a total peaking factor (Fq) of 2.1. Table 1.1 shows the calculated peak cladding temperatures and metal-water reaction results for the five different breaks. In all cases, the emergency core cooling system is shown to meet the Acceptance Criteria as presented in 10 CFR 50.46.

The exposure study examined the limiting large break LOCA at end-of-life fuel conditions (47.0 GWD/NTH peak rod burnup) in the maximum power fuel rod. End-of-life conditions yield the highest initi'al fuel temperatures and greatest fission gas release, and thus, bound the entire fuel history. Calculated results of the exposure study are given in Table 1.2 and show a PCT of 2111oF for the assumed LHGR of 12.0 kW/ft and Fq of 2.10. Thus, ENC results with the EXEN~PWR model as referenced show operation of the Donald C. Cook Unit 2 at 3425 NWt within the above limits over the entire fuel life will meet the NRC acceptance criteria of 10 CFR 50.46.

Table 1.1 Large Break Spectrum Analysis Results for Donald C. Cook Unit 2 with ENC Fuel T =

Fq 2.1 DECLG DECLG DECLG DECLS DECLS Anal sis Results ~C=1.0 ~C 0.8 ~C=0.6 A=8.25 ft2 A=4.95 ft2 Peak Clad Temperature (PCT) oF 2092 1925 1726 2026 1940 Time of PCT, sec. 287 221 165 248 232 Peak Clad Temperature location, ft. 9.38 9.12 8.87 9.12 9.12 Local Zr/H20 Reaction (max.), X* 5.83 3.42 1.55 4.79 3.52 Local Zr/H20 Location, ft. from bottom 9.38 9.12 9.12 9.12 9.12 Total H2 Generation, X of total Zr Reacted <1.0 1.0 < 1.0 < 1.0 < 1.0 Hot Rod Burst Time, sec. 68.37 74. 14 84.56 70.57 72.56 Hot Rod Burst Location, ft. 6.75 7.25 7.50 7.00 7.00 Calculation License Core Power, MMt 3425 Power Used for Analysis, NWt 3493.5 Peak Linear Power for Analysis, kW~ft. 12.0 T

Total Peaking Factor, Fq 2.10 N

Enthalpy Rise, Nuclear, F~H 1.55 Core Nuclear Conditions BOL

• Computed value at 400 seconds.

4'N-NF-82-35 Table 1.2 Exposure Sensitivity Results for D. C. Cook Unit 2 XN-1 Reload Total Peaking (Fq) 2.10 Peak Rod Burnup Maximum (GWD/MTM) 47.0 Maximum Local Zr/H20 Reaction (X) at 400 sec. 6.26 Peak Cladding Temperature (oF) 2111 Maximum Zr/H20 Reaction Location (ft) 9.38 Hot Rod Burst Time (sec.) 62.4 Hot Rod Burst Location (ft,) 7.00 Hot Rod Burst Temperature (oF) 1602 Rupture Pressure (psia) 593 Subchannel Flow Blockage (X) 49.4 Time of PCT (sec.) 286 PCT Location (ft.) 9.38' Core Wide Zr/H20 Reaction (/)

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XN-NF-82-35 2.0 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOCA) SPECTRUM ANALYSIS 2.1 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION The analysis for large breaks specified by 10 CFR 50.46(3)

"Acceptance Criteria for Emergency Core Cooling Systems for Light Water Power Reactors" is presented in this section. The results of the loss of coolant accident spectrum analysis are shown in Tables 1.1 and 1.2 and demonstrate 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(>), and EXEM/PWR Emergency Core Cooling System Evaluation Model updates as described in XN-NF-82-20(P) and supplements'(2) . Except as noted below, the detailed system models are as given in the example problem report XN-NF-82-20(P),

Supplement 2'(2).

For the purpose of LOCA analyses, a loss of coolant accident is defined as a rupture of the Reactor Primary Coolant System piping 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 low pressure trip setpoint is reached. Reactor trip and scram were conservatively neglected for the

l XN-NF-82-35 large break analyses. A Safety Injection System signal is actuated when the appropriate setpoint (high containment pressure) is reached. These countermeasures will limit the consequences of the accident in two ways:

1. Reactor trip and.borated water injection complements void formation in causing rapid reduction of power to a residual level correspqnding 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 is calculated, consistent with Appendix K of 10 CFR 50(3). Thereafter, the core heat transfer is unstable, with both transition and film boiling occurring. As the core becomes uncovered, both turbulent and laminar forced convection go steam are considered as core heat transfer mechanisms.

When the Reactor Coolant System pressure falls below 636 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 I

termination of bypass. This conservatism is again consistent with Appendix K of 10 CFR 50.

2.2 LOCA ANALYSIS MOOEL The spectrum analysis using the ENC EXEN/PWR, ECCS evaluation model includes the following computer codes: RODEX2(4) code for initial

XN-NF-82-35 stored energy; RELAP4-EN(5) for the system blowdown and hot channel blow-down calculations; ICECON( ) for the computation of the ice condenser containment backpressure; REFLEX(2~7) for computation of system reflood; and TOODEE2(2i8ig), with changes noted earlier, for calculation of the final fuel rod heatup. To eliminate discontinuities at fluid node boundaries, calculated temperatures from the hot channel run were smoothed before being inserted in the final TOODEE2 heatup calculation.

The Oonald C. Cook Unit 2 nuclear power plant is a 4-loop West-inghouse pressurized water reactor with ice condenser 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-82-20(P), Supplement 2(2). The system nodalization is depicted in Figure 2.1. The unbroken loops were assumed symmetrical and modeled as one intact loop with appropriately scaled input. Pump per-formance curves characteristic of a Westinghouse series 93A pump were used in the analysis. The transient behavior was determined from the governing conservation equations for mass, energy, and momentum. Energy transport, flow' ates, and heat transfer were determined from appropriate correla-tions. System input parameters are given in Table 2.2.

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 used for the analysis is shown 'in Figure 2.2, with an axial peaking factor of 1.355 (the 3 percent engineering uncertainty is included in the axial).

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XN-NF-82-35 b

The analysis of the loss-of-coolant accident is performed at 102 percent of rated power. The core power and other parameters used in the analyses are given in Table 2.2.

Containment backpressure for ice condenser containment is per-formed using the ICECON(6) computer code which has been approved for this purpose.

2.3 RESULTS Table 2.1 presents the timing and sequence of events as determined for the large break guillotine configuration with discharge coefficients of 1.0, 0.8, and 0.6 and the split break configuration with break areas of 8.25 and 4.95 square feet.

In general, the transient timing is slower for smaller discharge coefficients or break sizes. Table 1.1 presents the peak cladding tempera-tures and maximum metal-water reaction for the above spectrum of break cases. This range of break sizes was determined to include the limiting case for peak cladding temperatures, since both ENC and NSSS vendor calcu-lations have shown that other break locations and small breaks will not be limiting for this Westinghouse NSSS design. l Figures 2.3 through 2.25 present plotted results of the analysis for the limiting break (1.0 DECLG). Unless otherwise noted on the figures, time zero corresponds to the time of break initiation.

The maximum peak cladding temperature 2092oF was calculated for the equivalent double-ended cold-leg guillotine break configuration (CD=1.0)

T at a total linear heat generation rate of 12.0 kW/ft (Fq ='2.1) for ENC

I XN-NF-82-35 fuel. The maximum local metal-water reaction is 5.83K after 400 seconds, and the total core metal-water reaction reached is less than 1.0%, all well below the limits set by the criteria of 10 CFR 50.46.

The results of the limiting break calculation are essentially the same as those reported for the EXEM/PWR example problem(2). Only small corrections to system input, use of the mixing vane model, and a modifica-tion of the power distribution to correspond to an F~H of 1.55 exist between the two calculations. A PCT decrease of 36oF was calculated.

I Table 2.1 Donald C. Cook Unit 2 Large Break Events Time (seconds)

DECLG DECLG DECLG 1.0 DECLS 0.6 DECLS (CD=1.0) (CD=0.8) (CD=0.6) (8.25 ft2) (4.95 ftZ Events Start 0.0 0.0 0.0 0.0 0.0 Initiation of=-Break 0.05 0.05 0.05 0.05 0.05 Safety Injection Signal 0.65 0.65 0.65 0.65 0.65 Accumulator Injection, Intact Loop 15.6 15.6 17.0 15.6 16.1 Accumulator Injection, Broken Loop 3.3 3.3 5.0 2.2 7.6 End of Bypass (EOBY) 24.2 24.64 26.76 24.2 24.76 Bottom of Core Recovery (BOCREC) 40.71 41.14 43.24 40.71 41.21 Accumulator Empty, Intact Loop 51.37 51.34 52.86 51.42 51.61 Safety Pump Injection 25.65 25.65 25.65 25.65 25.65 Peak Cladding Temperature Reached 287 221 165 248 232

I XN-NF-82-35 Tahle 2.2 D. C. Cook tJnit 2 System Input Parameters Primary Heat Output, MWt* 3425 Primary Coolant Flow, ibm/hr 142.7 x 106 Primary Coolant Volume, ft3 11,892 Operating Pressure, psia 2250.

Inlet Coolant Temperature, oF 542 Reactor Vessel Volume, ft3 4,945 Pressurizer Volume, Total, ft3 1800 Pressurizer Volume, Liquid, ft3 1080.

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

Accumulator Volume, Liquid, ft3 (each of four) 950.

Accumulator Pressure, psia 636 Steam Generator Heat Transfer Area, ft2 51,500 Steam Generator Secondary Flow, ibm/hr 4(3.685 x 106)

Steam Generator Secondary Pressure, psi a 820 Reactor Coolant Pump Head, ft 277.

Reactor Coolant Pump Speed, rpm 1189.

Moment of Inertia, ibm-ft2 82,000.

Cold Leg Pipe, I.D. in. 27.5 Hot Leg Pipe, I.O. in. 29.0 Pump Suction Pipe, I.O. in. 31.0 Fuel Assembly Rod Oiameter, in.** 0.360 Fuel Assembly Rod Pitch, in. 0.496 Fuel Assembly Pitch, in. 8.466 Fueled (Core) Height, in. 144.0 Fuel Heat Transfer Area, ft2** 57,327 Fuel Total Flow Area, ft2** 53.703

• Primary Heat Output used in RELAP4-EM Model = 1.02 x 3425 = 3493.5 MWt

• • ENC fuel parameters

l PRESSURIIER STEAK STEAM GENERATOR GENERATOR Q HOOEL VOLUME 19 60 Q PLOH OUNCTION 56 Q40 18 19 INTACT LOOP BROKEN LOOP 20 REACTOR VESSEL Q4s Q46 46

'17 20 Qi 39 59 BREAK LOCATION 55 Q4s

)7 21 3 HOT LEG +2 15 15 HOT LEG Q9 Q Q)5 22 Q21 13 13 14 14 27 48 52 26 26 Q)2 Gs Qs Q47 Q53 Q49 Q24 PUMP 41 Q39 Q3S PUMP 57 Q7 24 Q41 28 Q37 44 Q3 Q43 Q48 PSO 38 34 10 CONTAINMENT 28 u36 43 u33 37 33 10 42 g3 23 QS Q42 Q44 40 31 31 QO ACCURILATOR ACCUMULATOR Figure 2.1 System Blowdown Nodalization for the Donald C. Cook Unit 2 PWR

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1.8 1:6 FZ = 1.355 1.4 1.2 1.

0.6 0.4 0.2 0.0 00 0 ~ 1 0.2 08 0.4 0.5 06 0.7 0.8 09 10 X/L Figure 2.2 Axial Peaking Factor versus Rod Length for Donald C. Cook Unit 2 ECCS Analysis

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Figure 2.ll Flow From Broken Loop Accumulator D. C. Cook Unit 2; DECLG (CD=1.0)

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Figure 2.13 Clad Surface Temperature PCT Node D. C. Cook Unit 2; DECLG (CD=1.0)

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Figure 2.15 Hot Channel Heat Transfer Coefficient PCT Node-P. C. Cook Unit 2; DECLG (Cp=l.0)

I 1.0 QECLC Q. C ~ COOK 2, . HOT CHANNEL 17 X17 FUEL 1C ZIj "4 ZS TXYE >. TER "RE".K ( SEC l Figure 2.16 Hot Assembly Inlet Flow - P. C. Cook Unit 2; PECLG (Cp=l.p)

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Figure 2.17 Hot Assembly Outlet Flow - D. C. Cook Unit 2; DfCLG-(CD=1.0)

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c f

DC COOK UNIT 2 . c NORMALIZED POWER 17X17 FUEL 1 0 OECLG

~ UDUF= 84 3425HWT

~

s c.

o 1

auJ cJ

~Z0 c.CC 200 Q40 c.80 TIME AFTER BREAK (SEC)

Figure 2.19 . Normalized Power D.. C. Cook Unit 2, DECLG (CD=1 o)

l

0. C- COO< P REFLOCO 17X17 FUEL 1~ 0 DECLG BOL, 3425 MWT Cr C'J Q)

CI C3 Q ed

~i UJ CL OC I

Tl n I c6 ;20 i.6C Z(iC "-4G "8v- ".ZC CO TXHE F'ROid BGCREC t SEC ) I GO cjl Figure 2.20 Core Flooding Rate - D. C. Cook Unit 2; DFCLG (CD=1.0)

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D. C ~ CC'OK Z REFLUGQ 17X17 FUEL 1 ~ G DECLG 80L: 34~5 AMT 2G iGG 2GC "4G 2S~ 566 4GG TIRE FR H L'CCREC ( SEC )

Figure 2.21 Ref lood Downcomer Mixture Level-D. C. Cook Unit 2; DECLG (CD=1.0)

l l

l

D C- COOK 2, REFLOCD 17X17 FUEL 1 0 DECLC BOL '4?5 NIT 2G I.Ci0 2C" 24G Sv ".ZG 3GC 4GC

')

TAHE FROH 90CREC i SEC Figure 2.22 Ref lood Coie Mixture Level - D. C. Cook Unit 2; DECLG (Cp=l.0)

0. C- COOK 2 REFl COD 17X17 FUEL 1. G DECLC BPL 3425 MMT Sv'ZG c.CC RGC ?4C  ?.SG r)')G Geo <OC T j:HE FROYi 9 OCREC ( SEC )

Figure 2.23 Ref lood Upper Plenum Pressure - D. C. Cook Unit 2; DECLG (CD=1.p)

II S

I

D. C. CCOK 2 REFLOCD 17 X17 FUEL '

ZQ 1 ~ 0 DECLG BOL 4 P 5 HM'.

c'l I

ILI fL I

~

ILJ c'l Q

K ILI I-Z cv I-(T Ko Q 8 I

~

(

(A OoP MG i6G ZGC c4G ZSG 3c!G 360 TAHE FROM BOCREC I SEC )

Figure 2.24 Ref lood Core Saturation Temperature-D. C. Cook Unit 2; DECLG (CD=1.0)

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D. C. Cook Unit 2, ENC 17x17 Fuel, 1.0 DECLG Fq=2.10 BOL PCT HQQE (NQDE Z~ AT 9.3/ fT. )

Z. RUPTUREG !'OOE (NOOE 1C AT C.'i5 fi. i 16-C %6 C

~

pn ZGC." "46.0 Figure 2.25 TOODEE2 Cladding Temperature vs Time D. C. Cook Unit 2 ; DECLG (CD 1 0)

I 37 XN-NF-82-35 3.0 EXPOSURE SENSITIVITY A sensitivity study was performed to evaluate the effect of fuel exposure on the LOCA ECCS results. The analysis was performed. using the EXEN/PWR ECCS model documented in XN-NF-82-20(P) and XN-NF-82-20(P),

Supplement 1(2). The results indicate that D. C. Cook Unit 2 can operate T

with ENC 17x17 fuel in compliance with NRC 10 CFR 50.46 criteria with a Fq limit of 2.10 throughout exposure.

The calculation performed was an analysis which bounds the entire fuel history. The most limiting fuel conditions calculated during the fuel history were input to the LOCA ECCS models. For the maximum power fuel rod, the limiting fuel conditions correspond to end-of-life (EOL) exposure conditions (47,000 NWD/NTH peak rod burnup). The fuel average temperature and internal rod pressure are highest at EOL for the hot rod. Decay power is highest for EOL conditions and was therefore used in the analysis.

Stored energy was calculated to be highest at beginning-of-life (BOL) for the average rod in the core. Therefore, BOL fuel parameters are used for average core conditions.

The analysis included RELAP4/hot channel, REFLEX/reflood and TOODEE ~heatup calculations. The RELAP4/hot channel calculation uses the RELAP4/blowdown calculated plenum conditions as boundary conditions and is used to calculate fuel temperatures during the blowdown portion of the LOCA transient. Hot channel plots are given for the fuel average temperature at the PCT node (Figure 3.1), clad surface temperature at the PCT node (Figure 3.2), depth of metal-water reaction at the PCT node (Figure 3.3), heat t'ransfer coef-ficient at the PCT node (Figure 3.4), hot assembly inlet flow (Figure 3.5),

I 38 XN-NF-82-35 and hot assembly outlet flow (Figure 3.6). Ref lood plots are given for the core inlet flooding rate (Figure 3.7), downcomer mixture level (Figure 3.8),

core liquid level (Figure 3.9), upper plenum pressure (Figure 3.10), and core saturation temperature (Figure 3. 11). The TOODEE/heatup plot of peak cladding temperature for the PCT node and the rupture node is given in Figure 3.12.

Table 1.2 summarizes the results of the TOODEE/heatup analysis.

T The results show that the D. C. Cook Unit 2 can operate with a constant Fq of 2.1 throughout the life of the fuel.

-1 0 QECLC t L 0 C CQQ<2., HO 'HANNcL 17 X17 FUL 3 4 Zb HMT E'OL OC I

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CO 26 c4 3c'. 3G 4G I

TTNE' SEC ) CA CJl Figure 3.1 RELAP4/Hot Channel Fuel Average Temperature at the PCT Node for 1.0 OECLG Break from Bounding Fuel Rod History Analysis

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~

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1 ~ 0 DECLC BL Q. C. COO<2. HO CH NNEL 17X17 FUEL 342.5 MitlT EO'G 2$ 29 TINE ( SEC )

Figure 3.2 RELAP4/Hot Channel Clad Surface Temperature at the PCT Node for 1.0 DECLG Break from Bounding Fuel Rod History Analysis

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1 0

~ OECLC 8" 0 C- COOK2. HOT CHALONE. 17 X17 FULL 3 4 25 ".PWT EOL 2 IG 26 24 TJ.Nc. ( SEC )

Figure 3.3 RELAP4/Hot Channel Depth of Metal-Water Reaction at the PCT Node for 1.0 DECLG Break from Bounding Power History Analysis

lt ll'

1. 0 DECLG BL D. C COO><2. HOT CHI:INNEL 17X17 FUEL 3425 HMT EOL OC I

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2 ZG c" 4 3Z 36 CO TjHE ( SEC ) I CA Vl Figure 3.4 RELAP4/Hot Channel Heat Transfer Coefficient at the PCT Node for 1.0 DECLG Break from Bounding Fuel Rod History Analysis

1~ 0 DECLC BL D- C ~ COO/P. HOT CHANNEL 17 X 17 FUEL 3 4 R5 MMT E'OL Lz 16 za 24 zs 32 TIME ( SEC )

Figure 3.5 RELAP4/Hot Channel Hot Assembly Inlet Flow for 1.0 OECLG Break from Limiting Power History Analysis

~

E

1 0 DE'CLG BL

~ 0. C ~ COOK2. MOT CMANNCL 17 X17 FUEL 3 4 2.5 NWT E'OL 16 P.G ~C 36 Tj:flE ( SEC )

Figure 3.6 RELAP4/Hot Channel Hot Assembly Exit Flow for 1.0 DECLG Break from Bounding Power History Analysis

D. C COOK 2 REFLUOD 17X17 FUEL

~ 1 0 DECLC

~ EOL 3425 HUT i'GG ZGG TTHE FROH BOCREC t SEC 2.40

)

"Si Figure 3.7 REFLEX Core Inlet Flooding Rate for 1.0 DECLG Break from Bounding Fuel Rod History Analysis

5 D. C. COOK 2 REFLOOD 17X17 FUEL 1 0 DECLC EOL 3425 HMT 20 iC)0 200 240 2SC 320 3GC 400 TINE F ROM 90CREC t SEC )

Figure 3.8 REFLEX Downcomer Mixture Level for 1.0 DECLG Break from Bounding Fuel Rod History Analysis

I

~

~

~

~

~

II

~

~

D. C. COOK 2, REFLOOD 17 X17 FUEL 1 0 DECLG EOL 342.5 HWT 40."

4G 2G I.60 ZGG I".40 ZSG SZG SGC TINE F RQH 8 OCREC ( SEC )

Figure 3.9 REFLEX Core Mixture Level for 1.0 DECLG Break from Bounding Fuel Rod History Analysis

I 5

~

S

D.C.COOK 2. REFLOOD >7Xa7 FUEL 1.0 DECLC EO'425 MMT 40 30 I.60 200 c40 c.SG 3ZG 360 400 TIME .-ROM BOCREC < SEC )

Figure 3.10 REFLEX Upper Plenum Pressure for 1.0 OECLG Break from Bounding Fuel Rod History Analysis

I 0- C- COOK 2. REF'LOOD 17X17 FUEL 1. 0 DECLG EOL 3425 MW

'2G

-Ge 2GC c.<C 2SG "26 SGO 4GC TTHE FRGH BOCREC ( SEC )

Figure 3.11 REFLEX Core Saturation Temperature for 1.0 DECLG Break from Bounding Fuel Rod History Analysis

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D. C. Cook Unit 2, ENC 17xl7 Fuel, 1.0 DECLG, = 2.10 EOL Fq O

CD cJ

1. PCT i(COE (NQQE Zl AT o 37 FT )

CI Z,----RUPTUREQ NODE V) 4J (HCDE 11 AT 7.CC FT. I QJ ci

~C~

~ CD Cf.

LQ LL TO I- cq (3

Z.

Cl Qo CC. Z4

~

CD nJ

~ 'C CD (iC. C B)0-0 ".$ 6 C TAHE SECGNDS Figure 3.12 TOODEE/Heatup Peak Cladding Temperature Plot for Rupture Node and PCT Node for 1.0 DECLG Break from Bounding Fuel Rod History Analysis

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51 XN-NF-82-35

4.0 CONCLUSION

S For breaks up to and including the double-ended severance of a reactor coolant pipe, the Donald C. Cook Unit 2 Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10 CFR 50.46 with ENC 17x17 fuel operating in accordance with the LHGR limits noted in Table 1.1. That is:

1. The calculated peak fuel element clad temperature does not exceed the 2200oF 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 transient is terminated at a time when the core geometry is still amenable to cooling. The hot fuel rod cladding oxidation limits of 17K are not exceeded during or after

'I quenching.

4. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radio-activity remain'ing in the core.

5 XN-NF-82-35 5oO REFERENCES

1. Exxon Nuclear Com an WREN-Based Generic PWR ECCS Evaluation Mo e , N- - , u y , and supp ements an rev>sions thereto, Exxon Nuclear Company, Inc. Richland, WA.

2. Exxon Nuclear Com an Evaluation Model EXEM/PWR ECCS Model U dates, e ruary , an supp ements upp. , Mare 1982; Supp. 2, March 1982), Exxon Nuclear Company, Inc.,

Richland, WA.

3. "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. Federal Register, Volume 39, Number 3, January 4, 1974.

4. Fuel Rod Thermal-Mechanical Res onse Evaluation Model, XN-NF-81-58(P),

ugust , xxon uc ear ompany, nc., >c an , WA.

5. U. S. Nuclear Regulatory Commission Letter, T. A. Ippolito (NRC) to W. S." Nechodom (ENC), SER for ENC RELAP4-EM Update, March 1979.

6. ICECON: A Com uter Pro ram Used to Calculate Containment Back-ressure for L A Ana sis Inc udin ce on enser P ants ,

ev. , ovem er , xxon uc ear ompany, nc.,

Richland, WA.

7. Exxon Nuclear Com an WREM-Based Generic PWR ECCS Evaluation o e ate ay , xxon uc ear ompany, nc., schland; WA.

8. Exxon Nuclear Com an ECCS Claddin Swellin and Ru ture Model, arc xxon uc ear ompany, nc.,

Richland, WA.

9. G. N. Lauben, TOODEE2: A Two-Dimensional Time De endent Fuel Element Therma na s>s ro ram, eport ay

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XN-NF-82-35 Issue Date: 04/30/82 Distribution M. J. Ades D. J. Braun J. C. Chandler R. E. Collingham G. C. Cooke S. E. Jensen'.

V. Kayser J. E. Krajicek J. N. Morgan G. F. Owsley G. A. Sofer T. Tahvili D. M. Turner H. G. Shaw J. D. Kahn P. J. Valentine H. G. Shaw/AEP (5)

Document Control (10)

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