ML17326B134

Nonproprietary Version of Rev 2 to DC Cook,Unit 2 Cycle 5,5%.Steam Generator Tube Plugging Limiting Break Loca/Eccs Analysis.

ML17326B134
Person / Time
Site:	Cook
Issue date:	08/07/1984
From:	Chandler J, Kayser W, Stout R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17326B133	List: ML17326B131 ML17326B134
References
XN-NF-84-21-(NP, XN-NF-84-21-(NP)-R02, XN-NF-84-21-(NP)-R2, NUDOCS 8408090209
Download: ML17326B134 (80)

Text

XN-NF-84-21(NP )

Revision 2 Issue Date: 8/7/84 DONALD C ~ COOK UNIT 2 CYCLE 5 - 5/o STEAM GENERATOR TUBE PLUGGING LIMITING BREAK LOCA/ECCS ANALYSIS Prepared -by:

W. V. ayser, Manager PWR Safety Analysis Concur:

. C. C an er, Lea ng neer Reload Fuel Licensing Approve: CSr out, Manager Licensing & Safety Engineering ggg g Concur:

.'. organ, Proposals anag r 5 Customer Services Engineering Approve: 87~%/

. A. o er, Manag Fu 1 Engineering 8 Technical Services gf

,E3(CGM NU.CLEAR .COMPANY, INC.

KNIMIolY-MIlIl'I'.II.II.f. IlI)I'Y

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XN-Nf-84-21(NP)

Revision 2 TABLE OF CONTENTS Section ~Pa e

1.0 INTRODUCTION

....:.................................. 1 2.0 S UMMARY o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3 3.0 LIMITING BREAK LOCA ANALYSIS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5

3. 1 LOCA ANALYSIS MODEL ........................... 5 3 .2 RESULTS 7

4.0 CONCLUSION

S ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 64

5.0 REFERENCES

...............,.......................... 65

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XN-NF-84-21(NP)

Revision 2 LIST OF TABLES Parcae Table 2.1 D.C. Cook Unit 2 LOCA/ECCS Analysis Summary ........ 4 3.1 Donald C. Cook Unit 2 System Input Parameters ~ ~ ~ ~ ~ ~ 9 3.2 1.0 OECLG Break Analysis Parameters ................ 10 3.3 D.C. Cook Unit 2 1.0 OECLG Break Event Times ... ...

~ 11 3.4 1.0 OECLG Break Fuel Response Results for C ycle 5 ............................................ 12 3.5 1.0 OECLG Break Fuel Response Results with an All ENC Core .................................... 13

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. XN-NF-84-21(NP)

Revision 2 LIST OF FIGURES

~Fi ure Pacae 3.1 RELAP4/EM Blowdown System Nodalization for D.C. Cook Unit 2 .... 14 3.2 Oowncomer Flow Rate During Blowdown Period, 1.0 OECLG Break ................................... 15 3.3 Upper Plenum Pressure during Blowdown Period, 1.0 DE(LG Break ....... 16 3.4 Average Core Inlet Flow during Blowdown Period, 1.0 OECLG Break ........ 17 3.5 Average Core Outlet Flow during Bl owdown Period, 1.0 DECLG Break ..................... 18 3.6 Total Break Flow during Blowdown Period, 1.0 OECLG Break ............................. 19 3.7 Break Flow Enthalpy during Blowdown, 1.0 DECLG Break .................................... 20 3.8 Flow from Intact Loop Accumulator during Blowdown Period, 1.0 OECLG Break .................. 21 3.9 Flow from Broken Loop Accumulator during Blowdown Period, 1.0 OECLG Break .................. 22 3.10 Pressurizer Surge'ine Flow during Blowdown Period, 1.0 OECLG Break ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 23 3.11 Heat Transfer Coefficient during Blowdown Period at PCT Node, 1.0 OECLG Break, 2.0 MWO/kg Case . 24 3.12 Clad Surface Temperature during Blowdown Period at PCT Node, 1.0 OECLG Break, 2.0 MWD/kg Case ..........,. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 25 3.13 Depth of Metal-Water Reaction during Blowdown Period at PCT Node, 1.0 DECLG Break, 2.0 MWD/kg Case . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 26

iv XN-NF-84-21(NP)

Revision 2 LIST OF FIGURES (Cont. )

~Fi ure Pa<ac 3.14 Average Fuel Temperature during Bl owdown Period at PCT Location, 1.0 OECLG Break, 2.0 MWD/kg Case . 27 3.15 Hot Assembly Inlet Flow during Blowdown

,

Period, 1.0 OECLG Break, 2.0 MWD/kg Case ........... 28 3.16 Hot Assembly Outlet Flow during Blowdown Period, 1.0 DECLG Break, 2.0 MWD/kg Case ........... 29 3.17 Heat Transfer Coefficient during Blowdown Period at PCT Node, 1.0 OECLG Break, 10.0 MWO/kg Case ................................... 30 3.18 Clad Surface Temperature during Blowdown Period at PCT Node, 1.0 OECLG Break, 10.0 MWD/kg Case ................................... 31 3.19 Depth of Metal-Water Reaction during Blowdown Period at PCT Node, 1.0 DECLG Break, 10.0 MWD/kg Case .................. 32 3.20 Average Fuel Temperature during Blowdown Period at PCT Location, 1.0 DECLG Break, 10.0 MWD/kg Case ................................... 33 3.21 Hot Assembly Inlet Flow during Blowdown Period, 1.0 OECLG Break, 10.0 MWO/kg Case .................. 34 3.22 Hot Assembly Outlet Flow during Blowdown Period, 1.0 OECLG Break, 10.0 MWD/kg Case .......;.......... 35 3.23 Heat Transfer Coefficient during Blowdown Period at PCT Node, 1.0 OECLG Break, 47.0 MWD/kg Case .................. . . . . ..... 36 3.24 Clad Surface Temperature during Blowdown Period at PCT Node, 1.0 OECLG Break, 47.0 MWO/kg Case ................................... 37 3.25 Depth of Metal-Water Reaction during Slowdown Period at PCT Node, 1.0 OECLG Break, 47.0 MWD/kg Case ....... .......................... 38

XN-NF 21( NP )

Revision 2 LIST OF FIGURES (Cont. )

~Fi ere Pa<ac 3.26 Average Fuel Temperature during Blowdown Period at PCT Location, 1.0 OECLG Break, 47.0 MWO/kg Case e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 39 3.27 Hot Assembly Inlet Flow during Blowdown Period, 1.0 OECLG Break, 47.0 NWD/kg Case'....... 40 3.28 Hot Assembly Outlet Flow during Blowdown Period, 1.0 OECLG Break, 47.0 HWD/kg Case .......... 41 3.29 Accumulator Flow during Refill and Ref lood Periods, Broken Loop, 1.0 OECLG Break ............. 42 3.30 Accumulator Flow during Refill and Ref lood Periods, Intact Loop, 1.0 DECLG Break ............. 43 3.31 HPSI 8 LPSI Flow during Refill and Ref lood Periodg, Broken Loop, 1.0 OECLG Break ............. 44 3.32 HPSI 8 LPSI Flow during Refill and Ref lood Periods, Intact Loop, 1.0 DECLG Break ............. 45 3.33 Containment Back Pressure, 1.0 DECLG Break ........ 46 3.34 Normalized Power, 1.0 OECLG Break, 2.0 MWO/kg Case ..... 47 3.35 Normalized Power, 1.0 DECLG Break, 10.0 NWD/kg Case .............................. 48 3.36 Normalized Power, 1.0 OECLG Break, 47.0 NWD/kg Case ............. 49 3.37 Reflood Core Mixture Level, 1.0 OECLG Break, C ycle 5 Core ........................ 50 3.38 Reflood Downcomer Mixture Level, 1.0 DECLG Break, Cycle 5 Core ..................... 51 3.39 Reflood Upper Plenum Pressure, 1.0 OECLG Break, Cycle 5 Core ............................... 52

vi XN-NF-84-21(NP)

Revision 2 LIST OF FIGURES (Cont.)

~Fi ure ~Pa e 3.40 Core Flooding Rate, 1.0 OECLG Break, Cycle 5 Core ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ 53 3.41 Ref lood Core Mixture Level, 1.0 OECLG Break, All ENC Core ...................................... 54 3.42 Reflood Oowncomer Mixture Level, 1.0 OECLG All ENC Core ............ 'reak, 55 3.43 Ref lood Upper Plenum Pressure, 1.0 DECLG Break, All ENC Core ............................... 56 3.44 Core Ref looding Rate, 1.0 OECLG Break, A 11 ENC Core ....................-......-.... .. 57 3.45 TOODEE2 Cladding Temperature versus Time, 1.0 OECLG Break, 2. MWO/kg Case, Cycle 5'ore ..... 58 3.46 TOOOEE2 Cladding Temperature versus Time, 1.0 OECLG Break, 10. MWO/kg Case, C ycle 5 Core ................... -..... ~ 59 3.47 TOOOEE2 Cladding Temperature versus Time, 1.0 OECLG Break, 47. MWO/kg Case, Cycle 5 Core ... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 60 3.48 TOOOEE2 Cladding Temperature versus Time, 1.0 OECLG Break, 2. MWO/kg Case, All ENC Core ..... 61'2 3.49 TOODEE2 Cladding Temperature versus Time, 1.0 DECLG Break, 10. MWO/kg Case, All ENC Core 3.50 TOOOEE2 Cladding Temperature versus Time, 1:0 OECLG Break, 47. MWO/kg Case, All ENC Core 63

XN-NF-84-21(NP )

Revision 2

1. 0 I NTRODUCT ION Large break LOCA/ECCS analyses were performed in 1982( ~2) to support operation of the D.C. Cook Unit 2 reactor at 3425 MWt with ENC fuel. Reference 1 presented analytical results for a spectrum of postulated large break LOCAs.

The limiting break was identified as the 1.0 double ended cold line guillotine (DECLG) break. Reference 2 presented results for the previously identified limiting break using the EXEM/PWR(3) ECCS models, except GAPEX was used as the fuel performance model in place of RODEX2. The RODEX2 code was not approved

(

by the NRC for use in ECCS analyses in 1982. Therefore the NRC-approved GAPEX(4) code was used to calculate fuel properties at the initialization of the LOCA calculation. The Reference 2 report documented the results of calculations with one and two LPSI pumps operating. At equivalent core peaking limits, higher peak cladding temperatures (PCTs) were calculated in the LOCA analysis when two LPSI pumps were assumed operating. The Reference 2 analysis with two LPSI pumps operating was performed for Cycle 4 operation of D.C. Cook Unit 2.

This report documents the results of a LOCA/ECCS analysis to support operation of the D.C. Cook Unit 2 reactor for Cycle 5 at a thermal power rating of 3425 MWt, with up to 5/. of the steam generator tubes plugged, with two LPSI pumps operating, and for ENC fuel exposed up to a peak rod average burnup of 47 MWD/kg. Results are also reported for the case in which the entire core is ENC fuel. The calculations were performed using the EXEM/PWR LOCA/ECCS

XN-NF-84-21(NP )

Revision 2 models, including fuel properties calculated at the start of the LOCA transient with ENC's generically approved RODEX2 code.(5)

XN-NF-84-21(NP)

Revision 2 2.0

SUMMARY

LOCA/ECCS calculations were performed to determine core peaking limits which permit operation of the O.C. Cook Unit 2 reactor within guidelines specified by 10 CFR 50.46 and Appendix K.(6) The calculations assumed operation:

1) At a thermal power of 3425 MWt;

2) With 5X average steam generator tube plugging;

3) With the Cycle 5 core configuration (85/ ENC fuel); and

4) With the entire core ENC fuel.

The calculations were performed for the previously identified limiting break, the 1.0 OECLG break, with full ECCS flow.

The results of the analysis are summarized in Table 2.1. The analysis supports operation of the O.C. Cook Unit 2 reactor for Cycle 5, and future cycles with ENC fuel, at a total peak limit (FqT) of 2.04 and a corresponding T

F<H limit of 1.55.

XN-NF-84-21(NP)

Revision 2 Table 2. 1 O.C. Cook Unit 2 LOCA/ECCS Analysis Summary Peak Rod Average Burnup (MWO/kg) 2.0 10.0 47.0 FT Q

2.04 2.04 2.04 T

1.55 1.55 1.55 Results for the Cycle 5 Core Confi uration (85K ENC Fuel)

Peak Cladding Temperature (oF) 2007 2014 1993 Maximum Local 2r-H20 Reaction (X) 4.6 4.7 4.5 Total Zr-H20 Reaction <1.0 < 1.0 < 1.0 Results with Entire Core of ENC Fuel Peak Cladding Temperature (oF) 2022 2030 2008 Maximum Local Zr-H20 Reaction (/) 4.9 5.0 4.7 Total Er-H20 Reaction <1.0 < 1.0 < 1.0

XN-NF-84-21(NP)

Revision 2 3.0 LIMITING BREAK LOCA ANALYSIS This report supplements previous LOCA/ECCS analyses performed and documented for D.C. Cook Unit 2. A spectrum of LOCA breaks was performed and reported in XN-NF-82-35.(1) The limiting LOCA break was determined to be the large double-ended guillotine break of the cold leg or reactor vessel inlet pipe with a discharge coefficient of 1.0 (1.0 DECLG). Reference 2 established that for D.C. Cook Unit 2 it is more limiting in the LOCA analysis to assume no failure of a LPSI pump. The analysis performed and reported herein considers:

1) That 5X of the steam generator tubes are plugged;

2) That 85/ of the Cycle 5 core is composed of ENC fuel;

3) That both LPSI pumps are operational; and

4) That ENC fuel may be exposed to a peak average burnup of 47 MWD/kg.

3.1 LOCA ANALYSIS MODEL The Exxon Nuclear Company EXEM/PWR-ECCS evaluation model was used to perform the analyses required. This model(3) consists of the following computer codes: RODEX2(5) code for initial stored energy; RELAP4-EM(") for the system blowdown and hot channel blowdown calculations; ICECON(8) for the computation of the ice condenser containment backpressure; REFLEX(3 g) for computation of system ref lood; and TOODEE2(3~ID~11) for the calculation of final fuel rod heatup.

XN-NF-84-21(NP )

Revision 2 The Oonal d C. Cook Uni t 2 nuclear power pl ant is a 4-1 oop Westinghouse 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".

The system nodalization is depicted in Figure 3. 1. The unbroken loops were assumed symmetrical and modeled as one intact loop with appropriately scaled input. Pump performance 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 rates, and heat transfer were determined from appropriate correlations.

The Cycle 4 LOCA analysis(2) assumed that lX of the steam generator tubes were plugged. In the current analysis, the plant was modeled assuming asymmetric steam generator tube plugging: 3.33/ of the tubes plugged in the intact loops, and 10.0/ of the tubes plugged in the broken loop. The larger plugging in the broken loop results in higher PCTs. The primary coolant flow at full power was reduced by 1.1/ from the current measured flow at the plant to account for the assumed average 5X steam generator plugging. Additionally, the core model assumed that the core is 85/ ENC fuel, whereas the previous analysis assumed the Cycle 4 core configuration. Calculations were also performed for the case in which the core is all ENC fuel, representative of

XN-NF-84-21(NP)

Revision 2 Cycle 6 and beyond. ENC fuel has a smaller rod diameter than the Westinghouse fuel it replaces. To offset the impact of increased flow area on the LOCA analysis results, the core power was reduced from 3425 MWt to 3411 NWt. System input parameters are given in Table 3.1.

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. Chopped cosine axial power profiles are assumed with the maximum axial peaking factor used in the analysis given in Table 3. 2. 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 3. 1.

3.2 RESULTS Table 3.3 presents the timing and sequence of events as determined for the large 'break guillotine configuration with a discharge coefficient of 1.0 for full ECCS operation. Table 3.4 presents the results of the exposure analysis for Cycle 5 composed of 85K ENC fuel. Table 3.5 presents the results of the exposure analysis for a core composed of all ENC fuel.

Results of the analyses are given in Figures 3.2 to 3.43. Figures 3.2 to 3. 10 provide plots of key system blowdown parameters versus times.

Figures 3. 11 to 3.28 provide plots of key core responses during the blowdown period. Figures 3.29 to 3.32 provide the ECCS flows in the broken and intact loop during the refill period. Figure 3.33 presents the containment pressure during the LOCA. Figures 3.34 to 3.36 present the normalized power during the LOCA for the three exposure cases analyzed. Figures 3.37 to 3.40 provide results from the reflood portion of the transient for the case in which 85K of

XN-NF-84-21(NP)

Revision 2 the core is ENC fuel. Figures 3.41 to 3.44 provide the reflood results for the case in which the core is composed entirely of ENC fuel. Finally, Figures 3.45 to 3.50 provide the response of the fuel during the refill and reflood periods of the LOCA transient for the fuel burnup cases 'investigated.

XN-NF-84-21(NP)

Revision 2 Table 3.1 'Donald C. Cook Unit 2 System Input Parameters Thermal Power, MWt* 3425 Core, MWt 3411 Pump, MWt 14 Primary Coolant Flow, Mlbm/hr 143.1 Primary Coolant Volume, ft3 11,768 Operating Pressure, psia 2250 In,let Coolant Temperature, oF 542 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 (each of four) 950 Accumulator Pressure, psia 636 Steam Generator Heat Transfer Area, ft2-SG1, SG2, SG3, SG4 11,588, 3(12,446)

Steam Generator Secondary Flow, ibm/hr- 3.505 x 106 S'G1, SG2, SG3, SG4 3(3.764 x 106)

Steam Generator Secondary Pressure, psia 799 Reactor Coolant Pump Head, ft 277 Reactor Coolant Pump Speed, rpm 1189 Moment of Inerti a, bm-f t2 1 82,000 Cold Leg Pipe, I.D. in. 27.5 Hot Leg Pipe, I.D. in. 29.0 Pump Suction Pipe, I.D. in. 31.0 Fuel Assembly Rod Diameter, in. 0.360 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, Bare Rod, ft2** 53.703 Refueling Water Storage Tank Temperature, oF 80 Accumulator Water Temperature, oF 120

• Primary Heat Output used in RELAP4-fM Model = 1.02 x 3425 = 3493.5 Mwt

• ."ENC Fuel Parameters.

10 XN-NF-84-21(NP)

Revision 2 Table 3.2 1.0 OECLG Break Analysis Parameters Peak Rod Average Burnup (MWO/kg) 2.0 10.0 47.0 Total Core Power (MWt)* 3411 3411 3411 T

Total Peaking (F~) 2.04 2.04 2.04 Fraction Energy Oeposited in Fuel

~ Fully Moderated Core 0.974 0.974 0.974 Voided Core 0.954 0.954 0.954 Cycle 5 (85/ ENC Fuel)

Peaking

'

Axial x Engineering 1.316 1.316 1.316 T

~

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Enthalpy Rise (F~H) 1.55 1.55 1.55 All ENC Core Peaking

. Axial x Engineering 1.316 1.316 1.316 T

. Enthalpy Rise (F~H) 1.55 1.55 1.55

• 2% power uncertainty is added to this value in the LOCA analysis.

XN-NF-84-21(NP)

Revision 2 Table 3.3 0. C. Cook Unit 2 1.0 OECLG Break Event Times Event Time (sec.)

Start 0.00 Break Initiation 0.05 Safety Injection Signal 0.65 Accumulator Injection Broken Loop 3.2 Intact Loop 15.5 End of Bypass 24.31 Safety Pump Injection 25.65 Start of Ref lood 40.48 Accumulator Empty Broken Loop 44.2 Intact Loop 52.9

12 XN-NF-84-21(NP )

Revision 2 Table 3.4 1.0 DECLG Break Fuel Response Results for Cycle 5

. Peak Rod Average Burnup (MWD/kg) 2.0 10. 0 47.0 Initial Peak Fuel Average Temperature (oF) 2151 2060 1629 Hot Rod Burst Time (sec) 69.5 70.5 78.5 Elevation (ft) 7.0 7.0 7.75 Channel Blockage Fraction .25 .28 .47 Peak Clad Temperature

. Time (sec) 287 288 269 Elevation (ft) 9.63 9.63 9.38

'. Temperature (oF) 2007 2014 1993 Zr-Steam Reaction

~ Local Maximum Elevation (ft) 9.63 9.63 9.38

~ Local Maximum (X)* 4.6 4.7 4.5 Core Maximum < 1.0 <1.0 < 1.0

• Values 400 sec into LOCA transient.

13 XN-NF-84-21(NP)

Revision 2 Table 3.5 1.0 OECLG Break Fuel Response Results with an All ENC Core Peak Rod Average. Burnup (MWO/kg) 2.0 10.0 47.0 Ini ti al Peak Fuel Average Temperature (oF) 2151 2060 1629 Hot Rod Burst

. Time (sec) 69.1 70.1 78.3 Elevation (ft) 7.0 7.0 7.75

. Channel Blockage Fraction .25 .28 .47 Peak Clad Temperature

. Time (sec) 292 292 274

. Elevation (ft) 9.63 9.63 9.38

. Temperature (oF) 2022 2030 2008 Zr-Steam Reaction

. Local Maximum Elevation (ft) 9.63 9.62 9.38

. Local Maximum (X)* 4.9 5.0 4.7

. Core Maximum <1.0 <1.0 <1.0

• Values 400 sec into LOCA transient.

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erne 8 K6 g

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1 ~ 0 DC COOK 2 17X17 ~ DECL BONs 57 AVE PLUG,10X 343 ~ 3% PLUGS CI QO u)S0 K

~3 k- Q K

C) y 4-9 5K O

1R, 16 zn 2.4 TAHE AFTER BREAK ( SEC)

F iqnre 3.P Downcomer Flow Rate Dnring Blowdown Period, 1.0 DECLG Break

1.0 DC COOK 2 17X17 DECL BDN 5X AVE'. PLUG 10/ 3+3-3X PLUG.

12 2D 3a TIHE AFTER BREAK f SEC )

Figure 3.3 Upper Plenum Pressure During Olowdown Period, 1.0 OECLG Break

al 1.0 DC COOK 2 17X1'j DECL 8DNi

~ 5l AVE. PLUG,10K 3+3.31 PLUG, D

tv Ul X'.

y 4J CKo Kg O

4 4J K

0 Qo CXg I

1P 1C 20 RA TAHE AFTER BREAK ( SEC)

Figure 3.cl Average Core Inlet Flow during Blowdown Period, 1.0 DECLG Break

1 0 OC COOK 2 17X17 e OECL BONo 5X AVE PLUGS 10 <'+3 ~ 3 X PLUG>

~R LU V)

K gg) CI

>o C) po mS Q O I-D O

4Jy

<a D

C9 Cl 12 1C 20 2$ 2I TXME AFTER BRFAK l SEC) f iqure 3.5 Average Core Outlet Flow during Blowdown Period, 1.0 OCCLG Break

D bio 1 0 OC COOK 2 17X17aOECL BONt 5~ AVE ~ PLUG ~ 10~ 3+3 ~ 3> PLUGS KJ

~8 wR K

O Dg

<o R

up X

D Z

12 16 2.0 RA TXHE AFTER ORF"'K ( SEC )

Figure 3.6 Total Break Flow during Blowdown Period, 1.0 DECLG Break

1 0 OC COOK R 1.0 OECLG BON

<<a8

%5 p Vessel Side IjJ wl Pump Side

~

TIME (SEC)

Figure 3.7 Break Flow Enthalpy Ouring Blowdown, 1.0 DECLG Break

tJ 1.0 DC COOK 2 17X17 ~ DECL BOH. 5X AVE. Pl UG.10K 3'+3-3X PLUG 4J CA g g

CK Oa QO L

o P-8 cr. Q K

Og 12 1C 20 t4 TIHE AFTER 8REAK ( SEC )

Figure 3.0 Flow from Intact Loop Accumulator during Blowdown Period, 1.0 DFCLG Break

tJ 1 0 DC COOK 2 17X17 r DECL BDNr 5> AVE PLUGS 101 3+3 3X PLUGS 4J3 CA w

~

K e4 ld

$

CK fL Og K

4J.

og K

O 1- o QO K

K Og CK 12 1C 20 2l a2 TXHE AFTER BREAK ( SEC )

Figure 3.9 F)ow from Broken l.oop Accumulator during Blnwdown Period, 1.0 DECLG Break

1.0 DC COOK 2 17X17 ~ DECL BOND 57 AVE. PLUGi10l 3+3-3X PLUG>

lZ 16 20 24 2a 32 TAHE AFTER BREAK ( SEC)

Figure 3.10 Pressurizer Surge Line Flow during Blowdown Period, 1.0 DL'CLG Break

l- 0 OC COOK Z 1 0 OECLG X

8 Vl ZI I

O CO I

t0 ti TZ~E ~ SE.C)

Figure 3.11 lleat Transfer Coeff icient during Blowdown Period at PCT Node, 1.0 DECLG Break, 2.0 NHD/kg Case

1.0 DC COOK 2. 1.0 OE'CLG tD ZI TIVE t SEC)

Figure 3.12 Clad Surface Temperature during Blowdown Period at PCT Node, 1.0 DECLG Break, 2.0 NWD/kg Case

r X.O DC COOK a a.O DEcLc Jo

~I M

bjm 4

g gO g 4 nl i 1t N RO tl TXVE t SEC)

Figure 3.13 Oepth of Metal-Water Reaction during Blowdown Period at PCT Node, 1.0 OECLG Break, 2.0 MWO/kg Case

CD c4 U)3 1 0 OC COOK R 1.0 DECLG OI 5II 5

oJ 8

B R >c 8 2:I

(

EA ll g

0 CO I

lS RD a

TXME l SEC)

Figure 3.10 Average Fuel Temperature during Blowdown Period at PCT Location, 1.0 OECLG Break, P.U NWO/kg Case

i.O OC COOK Z. X-0 OECLG Q. Ã N U 3f.

TIVE (SEC)

Figure 3.l5 llot Assembly Inlet Flow during Blowdown Period, 1.0 DECLG Break, 2.0 HWD/kg Case

1.0 Dc COOK 2 1-0 DE'CLG R

8

>c 2.'

Vl p

O CQ I

1$ t0 ?I IO o TXHE ( SEC )

Figure 3.16 I<ot Assembly Outlet I low during Blowdown Period, 1.0 DECLG Break, 2.0 I'1HD/kq Case

i.o DC COOK 2. x.o DECL&

8 X Xl ZI EA 0 CO I

R9 24 32 40 U

TIME (SEC)

F igure 3.17 lleat Transfer Coeff ir.ient during Blowdown Period at PCT Node, 1.0 l)ECLG Break, 10.0 MWD/kg Case

1.0 DC COOK Z 1.0 DECLG R7 OC I

Vl O CO I

ZO ZL l4 o

TINE' SEC )

Figure 3.18 Clad Surface Temperature during Blowdown I'eriod at PCT Node, 1.0 DFCLG Break, l0.0 NWD/kg Case

1.0 OC COOK 2 1.0 OECLG H

X z4 Oo H

I-CJ CK bl I

4J CC i8

<o lY H

N og h

Qg 8QX

( HI lA I

0 CO I

Z.O U (0 o TIME ( SEC )

Figure 3.19 Depth of Metal-Water Reaction during Blowdown Period at PCT Node, 1.0 DECLG Break, 10.0 HWD/kg Case

1.0 DC COOK 2 1.0 DECLG QX 8 I Vl 0 CO I

ZO tl 3t (0 o TINE ( SE'C )

Figure 3.20 Average Fuel Teotperature during Blowdown Period at PCT Location, 1.0 DECLG Break, 10.0 HWD/kg Case

1 0 OC COOK R 1.0 OECLG Kl M 8 RI Vl O CO I

Lt 1C ZO R.l 10 TIVE (SE'C)

Figure 3.21 )lot Rssembly tnlet Flaw during Olowdown Period, 1.0 OLCI.G llreak, 10.0 llWO/kg Case

1.0 OC COOK 2. 1 0 DfCLG R >C 8 2:I CA O CO I

M iM za Z( 3C 10 D

TINE (SfC)

Figure 3.22 tlot Assembly Outlet Flow during Blowdown Period, 1.0 OECLG Break, 10.0 NHD/kg Case

l-0 OC COOK 2 =

1 0 OECLt Z

lD X

Vl ZI O CO I

Mm 1k lS t0 t4 2$

U TIvE t SEC)

Figure 3.23 lleat Tansfer Coefficient during Blowdown Period at PCT Node, 1.0 DLCLG Break, 47.0 MHD/kg Case

1-0 OC COOK P. 1.0 QECLG LS za Kl TETE (SE'C)

Figure 3.24 Clad Surface Temperature during Blowdown Period at PCT Node, 1.0 DECLG Break, 47.0 Hll0/kg Case

~ N 1-0 OC COOk 2 1.0 DECL&

~S Qo

~8 Oo 0

~s kg 4J lZ gO CI$

gp O AJ R

8 >c KI Vl

)

0 00 I

0 lf 1S EO Rl tl "3E 1 TINE (SEC)

Figure 3.25 Depth of Metal-Water Reaction During Blowdown Period at PCT Node, 1.0 DECLG Break, 47.0 MWD/kg Case

1 0 OC COOK 2 1.0 OECLG R >c lD M I

Vl 0 CO I

It LC t0 U tl 3R. '3C l0 o TIt>E ( SEC 1 Figure 3.26 Average Fuel Temperature dur ing Blowdown Period at PCT Location, 1.0 DECLG Break, 47.0 MWO/kg Case

1 0 OC COOK 2 1 0 DE'CLG I 1t lS 5l RA ee TAHE (SEC)

Figure 3.27 (lot Assembly Inlet Flow during Blowdown Period, 1.0 OECLG Break, 47.0 MHD/kg Case

1.0 OC COOK 2, 1 ~ 0 OCCLG lf fO zl tl TIvE (SEC)

Figure 3.28 Hot Assembly Outlet Flow during Blowdown Period, 1.0 OECLG Break, 47.0 MWD/kg Case

~

8 Oy 1t 15 tD ZI 3a TIME AFTER EOB Y ( SEC )

I. igure 3.29 Accumulator Flow during Refill and Ref lood Periods, Broken Loop, 1.0 DECLG Break

1$ tD -

H El TXHE AFTER EOBY ( SEC )-.

Figure 3.30 Rccumulator Flow during Ref ill and Ref lood Periods, Intact Loop, 1.0 OECLG Break

300 250 200 150-100 50-R >c ED M I

0 Vl ll 0 50 100 150 200 250 300 350 0 I

CO Time (sec) After. Start I Figure 3.31 IIPSI 0 LPSI Flow during Refill and Ref lood Periods, Broken Loop, 1.0 OfCLG Break

C 1000 V)

CQ 800-O 600 400 P

O O 200 0

0 50 100 150 200 300 350 Time (sec) After Start I ignre 3.3? Hi'Sl 5 I.PSI Flow during Refill and Reflood Periods, Intact I oop, 1.0 I)ECLG llreak

22 21 20 19 P

18 R

0 16 O

O 15 g) X lD M

'C Ul pI 0 50 100 150 200 250 300 350 0 00 Time (sec)-After Start I Figure 3.33 Containment Back Pressure, 1.0 OECLG Break a

CK 4J O

0 a

4J t'4

~l K

LL oX SRI Vl Z

0 00 I

~4 Q (0 CO l$ 0 200 Z40 o TINE ( SECONDS )

Figiire 3.34 Horn" li"ed Pokier, '.C DECLu Break, Z.O Hl'ID/kg Case

Zl X 8

( I lh O OO I

H.O 160 200 210 210 320 3CO 100 TINE ( SE'CONDS )

Eigure 3.35 Normalized Power, 1.0 BECLG Break, 10.0 MWO/kg Case

i0 Q.O NO ZOO ZEO iso 320 TINED ( SECONDS )

I'igure 3.36 Normal ized Power, 1.0 DECLG Break, 47.0 NWD/kg Case

DCC2 REFl 000.1 0 OECLG FULL ECCS FLOV FQ=2.04 1 55 3525 HMT.HIX CORE LESS SPCR R

CD

>c I

Vl 0 CO I

PO tV 40 180 Z00 ZA0 Z00 3ZO 400 TIHE FROM BOCREC ( SEC ) a I=i@ore 3.3/ lief loocl Core Mixlisre I.evel, 1.0 OECI.G Break, Cycle 5 Core

DCC2, REFLOODo 1.0 DECLG FULL ECCS FLOW FQ=2.0$ 1.55 3425 HWTiHIX CORE LESS SPCR R >C

(

ID I

Vl O 00 I

40 80 160 200 2AO 280 <00 TIHE FROH BOCREC ( SEC ) U l inure 3.3A Reflood fjowocoioer Mixture Level, 1.0 DECLG Break, Cycle 5 Core

OCC2 REFLOOD, 1.0 DECLG FULL ECCS fLOW F9=2.0i 1.55 3425 HMT,HIX CORE LESS SPCR R >cR

(

CD I

I/l 0 CO I

FO fO 40 ao 160 200 240 2SO 360 400 TIHE FROH BOCREC ( SEC ) o I icjur>> 3.39 Hei'lood tipper Plenum Pressure, 1.0 OECLG Break, Cycle 5 Core

OCC2 REFLOOD.1.0 OECLG FULL ECCS FLOW FQ=2.04 1.55 3525 HWTiHIX CORE LESS SPCR 8R

>C

( RI Vl 0 CO I

~ PO 40 SO 120 160 200 RAO 280 380 400 TIME FROH BOCREC ( SEC) a I ignre 3.40 Core I-looding Rate, 1.0 DECI.G Break, Cycle 5 Core

OCC2. REFLOODo Al L ENC CORE FOH = 1-55 40 40 aso zaa zoo zea 3zo eao TXHE FROM BOCREC (SEC)

I igure 3.41 lhefloori Core HixIore l.evel, 1.0 DECI.G Break, All ENC Core

OCC2. REFLOOO. ALL EHC CORE FOH = 1.55 LLJ PA X

H K

oO O

a 40 IO 160 200 240 ZIO 360 400 TIME FROM 80CREC ( SEC )

I igure 3.02 Ref loorl Onwncomer Mixture Level, 1.0 OECLG Break, All ENC Core

OCC2 REFLOOOo ALL ENC CORE FOH <<1 55 X

8R 2I

(~'

I/I 0 CO I

44 $0 X60 Zao Z.40 nl 0 MO '360 400 TINE FROH BOCREC (SEC) a Eigure 3.03 lief lood Upper Vier>iun Pressiire, 1.0 OECl.G Break, All ENC Core

OCC2 REFLOOD, ALL EHC CORE FOH = 1.55 0 126 L60 5)0 240 320 360 ioo TIME FROM BOCREC (SEC)

I ignore 3A4 Core Ref loodirlg Rat.e, 1.0 DECI.G Break, All ENC Core

FQ=2.05 FOH ~1.55 - 2 HM

1. PCT HOOE (HOOE 22 AT S-62 FT- )

tL g 2. RUPTUREO HOOE (NOOE 11 AT 7.00 FT )

5~

al I

bl CL Da K"

K td CL

'K id/

C9 z,

H Cl Qo O

CI CI

%.0 40.0 S0.0 120.0 160.0 200.0 Z(0.0 200.0 320.0 360.0 TIME SECONDS I'igure 3.45 TOOOEE2 Cladding Temperature versus Time, 1.0 DECLG Break,

2. HllD/Kg Case, Cycle 5 Core

FQ=2.04 FOH ~1.55 10 H 1- PCT NODE (NODE 22 AT $ .62 FT )

2.. RUPTURED NODE (NODE 11 AT 1-00 FT- )

40-0 80.0 160.0 200.0 240.0 ZS0.0 32,0.0 360.0 40 .0 TIME SECONDS I:ignite 3.06 TOODCE2 Cladding Temperature versus Time, 1.0 DECI.G Break,

10. IIWD/Kg Case, Cycle 5 Core

N 1- PCT

)JODO'NODf Rl AT $ .31 FT )

RUPTURED HODR (0

hJ l'NODE 1<.AT 1.15 FT- )

hJ K

C9~

UJ ~

Q c5 I

)LJ f)'

o CJ:

LL lLJ 0

K QJ CI

)-5 C9 M

Cl QoD g

0 RX 8 RI Vl 0 0)

I Cl Cl Q).0 40.0 80.0 120.0 160.0 ?00 ' 2<0.0 210.0 320.0 360.0 40 .0 TINE SECONDS I igrrre 3.47 ~

l000l.:L2 Claddirrg Temperature versus Time, 1.0 OECLG Break, 4/. Hll0/Kg Case, Cycle 5 Core

- FQ~R.O{ FOB =1.55 " 2, HM

i. PCT NODE (NODE 22 AT $ .62 FT. )

2- RUPTURED NODE (NODE ).L AT 1.00 FT. )

40.0 a0.0 160.0 200.0 2(0-0 260.0 320.0 360.0 lORO TIHE SECONDS I'igure 3.0A TOOOLC2 Cl i~lrliug TemperaLure versus Time, 1.0 OLCl.G Break,

2. NWO/Kg Case, All l:NC Core

FQ~L.0$ FOH =1 55 10 H 1- PCT NODE t NODE 22, AT 5.62 FT. )

RUPTURED NODE t NODE 11 AT 1.OO FT. )

40.0 ao.o 120.0 160.0 200.0 2io.o 280-0 320.0 TXHE SECONDS I i gurt 3.49 f00f)EE2 Clarlding funperature versus 1 ime, 1.0 DECLG Break,

10. NWf)/Kg Case, All ENC Core

FQ<2.04 FOH 3.=1-55 - h7 M

1. PCT HQOE (HOOE 21 AT 0 31 FT )

Z. RUPTURED HOOf (HOOE 14 AT 7.15 FT 1

40. 0 80.0 120.0 160.0 200.0 240.0 280.0 320.0 TIME SECONDS Figure 3.50 100DEE2 Cladding Temperature versus Time, 1.0 DECI G Break, c17. HWD/Kg Case, Rll ENC Core

l 64 XN-NF-84-21 (NP )

Revision 2

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 for operation with ENC 17xl/ fuel operating in accordance with the LHGR limits noted in Table 2.1. 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 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 radio-activity remaining in the core.

XN-NF-84-21(NP)

Revision 2

5.0 REFERENCES

XN-NF-82-35, "Donald C. Cook Unit 2 LOCA ECCS Analysis Using EXEM/PWR Large Break Results," Exxon Nuclear Company, Inc., Rich-land, WA 99352, April 1982.

(2) XN-NF-82-35, Supplement 1, "Donald C. Cook Unit 2 Cycle 4 Limiting Break LOCA-ECCS Analysis Using EXEM/PWR," Exxon Nuclear Company, Inc., Richland, WA 99352, November 1982.

(3) XN-NF-82-20(P), Rev. 1, August 1982; and Supplement 4, July 1984, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Up-dates," Exxon Nuclear Company, Inc., Richland, WA 99352.

(4) XN-73-25, "GAPEXX: A Computer Program for Predicting Pellet-to-Cladding Heat Transfer Coefficients," Exxon Nuclear Company, Inc.,

Richland, WA, August 13, 1973.

(5) XN-NF-81-58(A), Rev. 2, "RODEX2: Fuel Rod Thermal-Mechanical Re-sponse Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA

. 99352, February 1983.

(6) "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.

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

(8) XN-CC-39, Rev. 1, "ICECON: A Computer Program Used to Calculate Containment Backpressure for LOCA Analysis ( Including Ice Condenser Plants)," Exxon Nuclear Company, Inc., Richland, WA 99352, November 1977.

(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 99352. May 1979.

(10) XN-NF-82-07(A), Rev. 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Inc., Richland, WA 99352, March 1982.

G.N. Lauben, NRC Report NUREG-75/057, "TOODEE2: A Two-Dimensional 1>>

(12) D.C. Cook Unit 2 Technical Specification, Appendix "A" to License No.

DPR-74, Amendment No. 48.

66 XN-NF-84-21(NP)

Revision 2 (13) XN-NF-82-32(P), Supplement 2, "Plant Transient Analysis for the Donald C. Cook Unit 2 Reactor at 3425 MWt: Operation with 5% Steam Generator Tube Plugging," Exxon Nuclear Company, Inc., Richland, WA 99352, February 1984.,

(14) XN-NF-84-21(P), "Donald C. Cook Unit 2, Cycle 5, 5X Steam Generator Tube Plugging, Limiting Break LOCA/ECCS Analysis," Exxon Nuclear Company, Inc., Richland, WA 99352, February 1984.

( 15) Letter, H. R. Denton (NRC) from J. C. Chandler (ENC), Re: Support-ing Documentation for Unit 2 Technical Specification Changes for Cycle 5 Reload, dated May 7, 1984 (JCC:076:84).

(16) XN-NF-84-21(P), Revision 1, "Donald C. Cook Unit 2 Cycle 5 - 5X Steam Generator Tube Plugging, Limiting Break LOCA/ECCS Analysis,"

Exxon Nuclear Company, Inc., Richland, WA 99352, May 1984.

XN-NF-8'4-21(NP )

Revision 2 Issue Date: 8/7/84 DONALD C COOK UNIT 2 CYCLE 5 Sio STEAN, GENERATOR TUBE PLUGGING LIMITING BREAK LOCA/ECCS ANALYSIS Distribution J. C. Chandler W. V. Kayser G. F. Owsley H. G. Shaw T. Tahvili AEP/H.G. Shaw (10)

USNRC/J.C. Chandler (15)

Document Control (3)