ML20207P890

Nonproprietary Rev 2 to DC Cook Unit 1 Limiting Break K(Z) Loca/Eccs Analysis

ML20207P890
Person / Time
Site:	Cook
Issue date:	01/15/1987
From:	Holm J, Stitt B, Tahvili T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17324B208	List: ML20207P890
References
XN-NF-85-115(NP, XN-NF-85-115(NP)-R02, XN-NF-85-115(NP)-R2, NUDOCS 8701200375
Download: ML20207P890 (58)

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XN-NF-85-ll5 L NPi I REVISION 2 I

l D.C. COOK UNIT 1 I LIMITING BREAK KLZ?

I LOCA/ECCS ANALYSIS

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ANUARY 1987

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I RICHLAND, WA 99352 I

.;g EXXON \UC_ EAR COV 3A\Y NC.

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3 XN-NF-85-115(NP) g Revision 2 Issue Date: 1/15/87 D. C. COOK UNIT 1 LIMITING BREAK K(Z) LOCA/ECCS ANALYSIS Prepared by: / y/ kf,[n' / //3/f7 T.~Tahvi11 PWR Safety Analysis i

Prepared by: Y B. 0.-Stitt

/-/3 '<37 PWR Safety Analysis Concur: $J. 5. 1plm, Manager t ///g / r 7 PWR Safety Analysis

, Concur: MG. t. ff Ward, Manager

///7/M i ReToad Licensing I

Concur: v. i %2rt l1 L!} Gbu ilo,hrj J. N. Morgan, Manager '

Customer 1ervices Engineering l w l Approved by:

k 46,Wk?

W H. E. Williamson, Manager Licensing & Safety Engineering Approved by: 'dr - a. ,_ 1 64 4% 4 '/

G.'L. Ritter, Manager /

Fuel Engineerina & Technical Services I

g ERON NUCLEAR COMPANY NC.

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XN-NF-85-115(NP)

I Revision 2 I

NUCLEAR REGULAToaY COMMtssioM ItEPoRT DtsCLAIMER IMPonTANTNOTtcI MEG ARDING CONTENT 5 ANO (15E OF

• MIS DOCUMENT PLEASE READ CAREFULLY This 'tenanasal resort was derived tarossa rensarta and deveicomsat procreamssommeares by Esses NessaarCassessy.Iac.!:as means asos ttaa by Esans Nessame to tas U1 Nestent Requassory (*'=====aas as part ( 2 toenement emeersbesses to fassiisase safety amedysse by licassess of the U1 Nesteer Regatasury '*- wassa sullas Ezaos Nusisar.(sericated retend fast er osaar tomansaat serviese provides by Ezzoa Nuctear for lisat weenr power rensesre and it is tres and correst to the best of Izzon Nuclears knowlessa.is(ermannen,and beslet. The intermassoa costaaned heresa may

he uend by its UA Nasisar Reemansory Casamasensa ta its review of

ass l resort, and ender the cursus of tas renessuve screamsets, by licensees or i sostlemas batere tas U1 Nessant Requistory Co==inasos unica are l

censomere of Zaman Nestaar la casar demeestranea of constiaase with the UA Nestant Regataser? ^- 's regasatsoes.

Esses Nessaar's werreause and .- "me :cacernans the tuorect messer of raia desaamans are tasse est forta la tae seroement between Ezzca T. Nestser and the cessesser to wanea taae desament is issued. Accortingly, I

esesos as osaarwise eserunney wides ia smaa screement, sestner Ezzon Numiser ser any perses senas os age beasan A. Maass say warras 7. or repressatanos azorses or imoaned, wita ressant to the assuracy, comoisteness, or ese(miassa of the tatorsmanes searmi==4 in this e d - ===e or taas tae ass of any informanos.

seestarea. ==amad or proomes diaeva==a in taas desammet will see intriage gravassey owned risats or E. Assamme may llaantities wita ressest to tae use of. or for i damaeas reseitiat froaa the use of, any inforumanos.

seestasus. ==shed or prosess dissaceed in taas ,

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TABLE OF CONTENTS Section Page

1.0 INTRODUCTION

................................................ 1 2.0 K(Z) LOCA ANALYSIS ................................. .... .. 5 2.1 LOCA Analysis Model......................................... 5 2.2 Results..................................................... 5

3.0 REFERENCES

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,l W LIST OF TABLES I8 Section Pace 1.1 0.C. Cook Uni t i LOCA-ECCS Analysi s Resul ts. . . . . . . . . . . . . . . . . 3 2.1 0.C . Cook Uni t i System 0ata. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 2.2 Fu e l De si gn P ar amet ers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 2.3 0.C. Cook Unit 1 LOCA-ECCS Analysis Results, Event Times.... 11

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iii LIST OF FIGURES Figure Pace 1.1 Comparison of Power Distributions Analyzed to Limits........ 4 2.1 Blowdown System Nodalization for 0.C. Cook Unit 1........... 12 6 2.2 Axial Peaking Factor versus Relative Heignt, 1.0 DECLS Break, 80C......................................................... 13 2.3 Axial Peaking Factor versus Relative Heignt, 1.0 DECLS Break, 0 E0C......................................................... 14 2.4 Downcomer Flow Rate, 1.0 OECLS 8reak........................ 15 2.5 Upper P lenum Pressure, 1.0 DECLS 8reak. . . . . . . . . . . . . . . . . . . . . . 16 2.6 Average Core In let Flow, 1.0 DECLS Break. . . . . . . . . . . . . . . . . . . . 17 2.7 Average Core Ou tlet Flow, 1.0 DECLS Bre ak . . . . . . . . . . . . . . . . . . . 18 8 2.3 Total Break Flow, 1.0 DECLS Break........................... 19 2.9 Bre ak Junction Enthal py, 1.0 0ECLS. . . . . . . . . . . . . . . . . . . . . . . . . . 20 9 2.10 Flow from Intact Loop Accumulator, 1.0 DECLS Break..........

2.11 Containment Back Pressure, 1.0 DECLS Break..................

21 22 2.12 Hot Channel Heat Transfer Coefficient, 1.0 CECLS Sreax, 8 80C......................................................... 23

2. '.3 Clad Surf ace Temperature, 1.0 OECLS Break, 80C. . . . . . . . . . . . . . 24 2.14 Depth of Metal-Water Reaction, 1.0 DECLS Brea'k, 30C......... 25

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2.15 Hot Channel Average Fuel Temperature, 1.0 DECLS Break, 30C.. 26 2.16 Hot Assembly Inlet Flow, 1.0 DECLS Break, 30C. . . . . . . . . . . . . . . 27 8

- 2.17 Hot Assembly Outlet Flow, 1.0 DECLS Break, 80C. . . . . . . . . . . . . . 28 2.18 Hot Channel Heat Transfer Coefficient,1.0 DECLS., Break, E0C......................................................... 29 2.19 Clad Surf ace Temoerature, 1.0 DECLS Break, ECC. . . . . . . . . . . . . . 20

, 2.20 Depth of Metal-Water Reaction, 1.0 DECLS 3riax, 50C......... Il 2.21 Hot Channel Average Fuel Temperature, 1.0 DECLS 3reak, E0C......................................................... 32 g 2.22 Hot Assemoly Inlet Flow, 1.0 DECLS 3reak, E0C............... 33 y 2.23 Hot Assemoly Outlet Flow, 1.0 DECLS 3 re ak , E0C. . . . . . . . . . . . . . 24 2.24 Normali zed Power, 1.0 DECLS Break, 30C. . . . . . . . . . . . . . . . . . . . . . 35 i

8 XN-NF-85-115(NP)

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Floure Page

2. 25 Noma li zed Power, 1.0 DECLS Bre ak , E0C. . . . . . . . . . . . . . . . . . . . . . 26 2.26 Reflood Core Mixture Level, 1.0 DECLS Break, 80C............ 37 2.27 Reflood Downcomer Mixture Level, 1.0 DECLS Break, 80C....... 38 2.28 Reflood Upper Plenur Pressure, 1.0 OECLS Break, 80C......... 39 g

2.29 Core Flooding Rate. 1.0 DECLS Break, 80C. . . . . . . . . . . . . . . . . . . . 40 y 2.30 Reflood Core Mixture Level, 1.0 DECLS Break, E0C............ 41 2.31 Reflood Downcomer Mixture Level, 1.0 DECLS Break, EOC....... 42 2.32 Reflood Upper Plenum Pressure,1.0 OECLS Break, E0C.. .. ... .. 43 2.33 Core Flooding Rate, 1.0 DECLS Break, E0C. . . . . . . . . . . . . . . . . . . . 44 g 2.34 T000EE2 Cladding Temperature vs Time,1.0 DECLS Break. W B0C......................................................... 45 2.3!i T000EE2 Cladding Temperature vs Time,1.0 OECLS Break, E0C......................................................... 46 8g 2.3ti Hot Channel Factor Nomalized Envelope for Fg = 2.04, K ( Z ) Fu ncti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 i I 1

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1.0 INTRODUCTION

AND

SUMMARY

This document presents analytical results for a postulated large break loss-of-coolant accident (LOCA) for the O.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 0.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 previous analyses as the most limiting break.(1,2)

The LOCA analyses were performed for a full core of ENC fuel using the _

models outlined in Section 2.1. The maximum allowable linear heat generation rate (including the 1.02 factor for power uncertainty) is 14.3 kW/ft, corresponding to a maximum total power peaking i' actor of 2.04 I I (F Q), and nuclear enthalpy rise of 1.51 (F g),

The present LOCA ECCS analyses were performed for Beginning-of-Cycle (BOC) fuel (2,000 mwd /MTM) and exposed fuel at End-of-Cycle (EOC) with a conservatively low peak red average burnup of 9,000 mwd /MTM to maximize ,

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 anticipated power shapes at the EOC exposure was used. These power shapes

( are shown in Figure 1.1 and compared to theq F (2) limit. All of the ENC fuel currently in the O.C. Cook Unit i reactor is at exposures greater than 20,000 mwd /MTM..

An earlier sensitivity study (0I using the ENC evaluation models had shown that the peak clad taperature (PCT) increased when maximum LPSI ' low was assumed. A similar sensitivity study was performed with the current power -

distributions and evaluation models to predict the PCT for both the maximum LPSI flow case and the case when one LPS! pumo fails. The sensitivity study showed that the maximum PCT was reached wnen full LPSI 9 -

XN-NF-85-115(NP)

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flow was assumed. The results reported in this document are for the case with no failure of LPSI pumps.

The calculational basis and results of the present analyses are summarized in Table 1.1. For the 80C case, the maximum calculated PCT is 1853*F, and w occurs at 72 seconds from the start of the transient With the EOC exposure, the maximum calculated PCT is 1918'F, and occurs at 202 seconds from the start of the transient In bor.h cases, the total metal water reaction is less than one percent of the zircalo) in the core. The results of the analyses show g that within the limits established, the O.C. Cook Unit I nuclear reacter E 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 p 2200 F limit. -

(2) The amount of fuel element cladding that reacts chemically with water g or steam does not exceed if. of the total amount of zircaloy in the 15 reactor.

(3) The cladding temperature transient is terminated at a time when the I

core geometry is still amenable to cooling. The hot fuel red cladding oxidation limits of 17% are not exceeded during or after quenching.

(4) The care temperature is reduced and decay heat is removea for an extended period of time, as required by the long-lived radioactivity remaining in the core.

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mM M gM M M) 8Qg M : M ~ -M > H Mr M, M&M lable 1.1 0.C. Cook thalt 1 LOCA-ECCS Analysis Results - K(Z)

BOC 20001610/MIM Peak EOC 9000 MWD /MIM Peak AnalysisResuM Rod Average Exposure Rod Average Exposure Peak Clad lemperature (PCI), of 1853 1918 Iime of PCI, sec. 71.9 202 Iocal tr/Il2 0 Reaction (max.), 5* 1.21 2.92 Iotal11 2 aesieration, C 1 of Intal Zr Reacted < l .0 <l.0 liot Rod !!urst Ilme, sec. 50.2 58.0 w

Calcislational Basis Iicense Core Power,IWt 3250 3250 Power Used for Assalysis, HWt** 3315 3115 Peak i inear Power f or Asialysis, kW/f t** 14.3 13.1 lotal Peak isig Factor, t ag I 2.04 1.95 E nti: 41py Rise, Nuclear, t{gg 1.51 1.51 gg Casaputer value .38 400 seconds Oh Int luding 1.02 8.at tor Ior power inicer14init les

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Revision 2 E., g 2.0 KfD LOCA ANALYSIS I This report provides the results of a LOCA/ECCS analysis performed for the I 0.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 0.C. Cook Unit I, with results reported in XN-NF-76-51.(15) The limiting LOCA break was determined to be an equivalent double-ended split break of the cold leg (1.0DECLS).

The analysis in this report is expected te be conservative with respect to ENC fuel at peak rod average exposures of up to 47,000 mwd /MTM. The peak clad temperatures are dependent upon the initial stored energy, which for the EXEM/PWR models increases from 0 to about 2000 mwd /Mlli and then decreases with exposure. An analysis for a plant similar to 0.C. Cook Unit 1 (D.C. Cook Unit 2 Reference 18) but with a 17x17 fuel geometry rather than a 15x15 fuel geometry demonstratas that over the exposure range of 0 to 47,000 mwd /NTM, the peak clad temperature occurs at the exposure point corresponding to the peak stored energy.

2.1 LOCA Analysis Model The Exxon Nuclear Company EXEM/PWR ECCS evaluation model I3) was used to I perform the analyses. This model consists of the following computer codes: R00EX2 I4I code for initial rod stored energy and internal fuel rod g gas inventory; RELAP4-EMIll) for the system blowdown and hot channel B blowdown calculations; ICECON(12) for the computation of ice conconser containment backpressure: REFLEX (3,5,13) for comautation of system reflood; and T000EE2(3,5,14) for the calculation of final fuel rod heatuo.

The quench and heat transfer coefficient models used in the reflood I

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l portion of the transient are based on the Fuel Cooling Test Facility g (FCTF) test data and are reported in References 10 and 17. W The O.C. Cook Unit i nuclear reactor is a four-loop Westinghouse pressurized water reactor with an ice condenser containment. The reactor coolant system is nodalized into control volumes representing reasonably g homogeneous regions, interconnected by flow-paths or ' junctions.* The 5 system nodalization is as depicted in Figure 2.1. The pump performance characteristic curves are supplied by the NSSS vendor. The transient behavior was detemined from the governing conservation equations for mass, energy, and somentum. Energy transport, flow rates, and heat transfer are datamined from appropriate correlations. System input '

parameters are given in Table 2.1.

The reactor core is modeled with heat generation rates datamined from I

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 perfomed at 102". of rated power. The fuel design parameters are shown in Table 2.2.

2.2 Ensul.ta 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(2) curve is defined such that the Linear Heat Generation Rates (LNGRs) determined by the NSSS vendor analysis are unchanged. The ((Z) function i is shown in Figure 2.36.

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l Table 2.3 presents the timing and sequence of events as detemined for the split break with a discharge coefficient of 1.0. Comparison of the system 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 I systsuo blowdown analysis. Unless othemise 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 i 2.24 through 2.25 show the nonsalized power calculation results. The reflood calculation results are shown in Figures 2.25 through 2.33.

The maximum peak cladding temperature (PCT) calculated for the 1.0 DECLS break at 90C is 1853*F (Figure 2.34). The maximum local metal water reaction in this case is 1.217. after 400 seconds, and the total core I metal-water reaction is less than 1"..

is 1918'F (Figure 2.35),

The local metal-water reaction is 2.92".. with a total metal-water reaction of less than 1"..

An earlier sensitivity study (0) using the DiC evaluation models had snown that the peak clad temperature (PCT) increased when maximum LPS! flow was assumed. A similar sensitivity study was performed with the current power I_ _

i XN-NF-85-ll5(NP)

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distributions and evaluation models to predict the PCT for both the I

maximum LPSI flow case and the case when one LPSI pump fails. The sensitivity study showed that the maximum PCT was reached when full LPSI flow was assumed. Only the more conservative results for the full LPSI flow case are reported. The results of the analyses show that within the limits established, the O.C. Cook Unit I nuclear reactor satisfies the criteria specified by 10 CFR 50.46(16) for operation at the rated system power level.

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Table 2.1 Donald C. Cock Unit 1 System Data Primary Heat Output, NWt 3250.*

Primary Coolant Flow, Ibm /hr 135.6 x 10 6 Primary Coolant Volume, ft 3 I Operating Pressure, psia 11,890.

2250.

Inlet Coolant Temperature F 536.3 Reactor Vessel Volume, ft 3 4602.

3 Pressurizer Volume, Total, ft 1800.

Pressurizer Volume, Liquid, ft 3 1080.

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

Accumulator Volume. Liquid, ft 3 929.

Accumulator Pressure, pisa 636.

Steam Generator Heat Transfer Area, ft 2 ,

50,985.***

Steam Generator Secondary Flow, lbm/hr 3.53 x 10 6 Steam Generator Secondary Pressure, psia 758.

Reactor Coolant Pump Head, ft 227.

Reactor Coolant Pump Speed, rpm 1190.

2 I Moment of Inertia, lbm-ft / rad Cold Leg Pipe, I.D., in 82,000.

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**

I Active Core Height, in Fuel Heat Transfer Area, ft 2 144.0**

52,200 Fuel Total Flow Area, ft 2 50.91

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

• • ENC fuel parameters.

l Represents value corresponding to 1% steam generator tube plugging used in the analysis.

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Table 2.2 Fuel Odsign Parameters i I

l Cladding, 0.0., in 0.424 Cladding, 1.0., in 0.364 g Cladding Thickness, in 0.030 5 Pellet 0.0., in 0.3565 Diametral Gao, in 0.0075 Pellet Density, %TD 94.0 Active Fuel Length, in 14.0 Rod Piten, in 0.563

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Revision 2 11 I Table 2.3 0.C. Cook Unit 1 LOCA/ECCS Analysis Results.

Event Times I

I Event Time (sec)

Start 0.00 Break Initiation .05 Safety Injection Signal .65 Accumulator Injection, Broken Loop 2.0 Accumulator Injection, Intact Loop 15.5 End-of-Bypass 22.9 Safety injection Flow 25.7 i Accumulator E.w ties, Broken Loop 26.2 Start of Reflood 39.6 Accumulator Emoties, Intact Loop 50.7 Peak Clad Temperature Reached BOC (Chopped cosine axial power peaking) 71.9 I EOC (Upskewed axial power peaking) 202.0 I.

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3.0 REFERENCES

I XN-NF-81-07, "LOCA/ECCS Reanalysis for 0.C. Cook Unit 1 Using the ENC 1.

WREM-IIA PWR ECCS Evaluation Model," Exxon Nuclear Company, Inc.,

Ricniand, WA, February 1981.

I 2. XN-NF-83-61, "D.C. Cook Unit 1 LOCA/ECCS Analysis for Extended Exco-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, Maren 1982;

.I Supplement 2, March 1982; Supplement 3, June 1985; and Supplement 4, July l 1984 Exxon Nuclear Company, Inc., Richland, WA.

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

Richland, WA, March 1984.

5. XN-NF-82-07(P)(A), Revision 1, Exxon Nuclear Company ECCS Claading Swelling and Rupture Model," Exxon Nuclear Company, Inc., Ricnland, WA, August 1992.

6. Letter, H.G. Shaw (ENC) to H.L. Sobel (AEP), HGS:062:32, dated Feo-ruary 23, 1982.

7. XN-75-41, " Exxon Nuclear Company WREM-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-aased Generic PWR ECCS Evaluation Model Update ENC WREM-II," July 1976; Supplement 1, Septemaer 1976; and Supplement 2, November 1976, Exxon Nuclear Company, Inc., Ricnland, 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), Rev.1, Volume II, "PWR 17x17 Fuel Cooling 7est Progrsn:

Reflood, Quench, Carryover, and Heat Transfer Cor 21ations,* Exxon Nuclear Company, Inc., Richlano, WA, January 1986.

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

12. XN-CC-39, Revision 1, *ICECON: A Comouter Program used to Calculate Containment Backpressure for LOCA Analysis (Including Ic2 Concenser Plants),' Exxon Nuclear Company, Inc., Richland, WA, Novemoer 1977.

I )

I I _

XN-NF-85-115(NP) g Revision 2 g l

l

13. XN-NF-CC-52(P), Volume II, " REFLEX and REFLEX /UPI PWR Reflood Computer I

Program - User's Manual," Exxon Nuclear Company, Inc., Richland, WA, 3 February 1985.

g

14. G.M. Lauben, "T00DEE2: A Two-Oimensional Time Dependent Fuel Element Thermal Analysis Program," 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., Richlanc, WA, Octobe'r 1976.

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

I 17. XN-NF-85-105(P), " Scaling of FCTF-Based Reflood Heat Transfer Cor-

! relation for Other Bundle Designs " Exxon Nuclear Canpany, Inc.,

l Richland, WA, October 1985.

18. XN-NF-86-68(P), Rev.1, " Donald C. Cook Unit 2 Limiting Break LOCA/ECCS Analysis,101 Steam Generator Tube Plugging, and K(Z)," Exxon Nuclear Company, Inc., Richland, WA, April 1986.

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I XN-NF-85-115(NP) l Revision 2 Issue Date 1/15/07 I

I

0. C. COOK UNIT 1 LIMITING BREAK K(Z) LOCA/ECCS ANALYSIS 1

'I Distribution J. S. Holm H. G. Shaw I T. Tahvili H. E. Williamson

8. D. Stitt R. C. Gottula G. L. Ritter I -

keerican Electric Power /HG Shaw (12)

Document Control (5)

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