ML17320A946

Nonproprietary Rev 2 to Plant Transient Analysis for DC Cook Unit 2 Reactor at 3,425 Mwt Operation W/5% Steam Generator Tube Plugging

ML17320A946
Person / Time
Site:	Cook
Issue date:	03/03/1984
From:	Adams F, Chandler J, Kayser W SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17320A942	List: ML17320A941 ML17320A944 ML17320A945 ML17320A946 ML17334A516
References
XN-NF-82-32(NP), XN-NF-82-32-(NP)-R02, XN-NF-82-32-(NP)-R2, NUDOCS 8403080223
Download: ML17320A946 (118)

Text

XN-NF-82-32 (NP)

Revision 2

Issue Oate:

3/3/84 PLANT TRANSIENT ANALYSIS FOR THE DONALD C.

COOK UNIT 2 REACTOR AT 3425 Nwt OPERATION WITH 5X STE GENERATOR TUBE PLUGGING Prepared by:

~

am PWR Safe,.iy Analysis Reviewed by: ~

QA9/Ff K yser, anager PWR Safety Analysis Concur:

an er, ea ngineer Reload Fuel Licensing Concur:

organ, Nanager Pro osals Im Customer Services Engineering Approve:

. B. Stout, Nanager Licensing E Safety Engineering Approve g

G. A. Sofer, Nanager Fuel Enaineerino E Technical Services E@4oiM MUCLEAR COMPANY,In+.

Ba03080223 840302 PDR ADOCK 050003lh

'DR

NUCLEAR REGULATORYCOMMISSION DISCLAIMER IMPORTANTNOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT

'LEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear

Company, Inc.

It is being sub-mined 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 Nudear-fabricated reload fuel or other technical services provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief.

The informadon 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.

Vflthout 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 dan'ages resulting from the use of, any information, ap.

paratus, method, or process disclosed in this document.

XN-NF-FOO, 766

XN-NF-82-32 (NP)

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ACKNOWLEDGEMENT Exxon Nuclear Company's T.R. Lindquist has significantly contributed to this work.

,l I

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TABLE OF CONTENTS Section

~Pa e

1.0 INTRODUCTION

1 2.0

SUMMARY

o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~

~

~ ~ ~

~

~

~ ~ ~ ~

~

~ ~ ~ ~ ~ ~ ~ ~ ~

~

~

2 3.0 CALCULATIONALMETHODS AND INPUT PARAMETERS.........

7 4.0 TRANSIENT ANALYSIS.................................

18

4. 1 UNCONTROLLED ROD WITHDRAWAL...................

18 4.2 LOCKED ROTOR..................................

19 5.0 6.0 4.3 LOSS Of EXTERNAL ELECTRICAL LOAD..............

4.4 DECREASED FEEDWATER HEATING............,......

21 4.5 EXCESSIVE LOAD INCREASE INCIDENT..............

DISCUSSION 97 REFERENCES.........................................

103

L

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LIST OF TABLES Table

~Pa e

2.1 2.2 3.1 3.2 3.3 Applicable Fuel and Vessel Design Limits...........

5 II Sugary of Results.................................

6 Operating Parameters Used in PTSPWR2 Analysis of Donald C.

Cook Unit 2................'..

ll Donald C.

Cook Unit 2 Trip Setpoints...............

12 Donald C.

Cook Unit 2 Fuel Design Parameters Exxon Nuclear Fuel.................................

13 3.4 Donald C.

Cook Unit 2 ENC Kinetics Parameters......

14 4.1 Event Sequence for Fast Rod Withdrawal.............

23 4.2 Event Sequence for Slow Rod Withdrawal.............

24 4.3 Event Sequence for Locked Rotor....................

25 4,4 Event Sequence for Loss of. External Load...........

26 k

4.5 Event Sequence for Decreased Feedwater Heating.....

27 4.6 Event Sequence for Excessive Load Increase.........

28

t

XN-NF-82-32 (NP)

~ Revision 2

LIST OF FIGURES

~Fi ure

~Pa e

3.1 3.2 3.3 PTSPWR2 Block Diagram..............................

15 Axial Power Profile Used in the Transient Analysis for D.C.

Cook Unit 2......................

Scram Curve Used in the Transient Analysis for D.C.

Cook Unit 2...............................

17 4.1 4.2 MDNBR vs. Reactivity Insertion Rate for Rod Withdrawal Transients..............................

29 Thermal Power for Fast Rod Withdrawal..............

30 4.3 Core Heat Flux for Fast Rod Withdrawal.............

31 4.4 4.5 RCS Temperatures for Fast Rod Withdrawal-Hot Leg, Core Average, Cold Leg and Core Inlet Temperatures.................................

32 Pressurizer Pressure for Fast Rod Withdrawal.......

33 4.6 Pressurizer Liquid Volume for Fast Rod Withdrawal.....................................

34 4.7 4.8 4.9 e

Steam Generator Pressure for Fast Rod Withdrawal...

35 Thermal Power for Slow Rod Withdrawal..............

36 Core Heat Flux for Slow Rod Withdrawal.............

37 4.10 4.11 RCS Temperatures for Slow Rod Withdrawal-Hot Leg, Core Average, Cold Leg, and Core Inlet Temperatures.................................

38 Calculated Overtemperature hT Setpoint for Slow Rod Withdrawal................................

39 4.12 4.13 Pressurizer Pressure for Slow Rod Withdrawal.......,

40 Pressurizer Liquid.Volume for Slow Rod Withdrawal 41 4.14 Steam Generator Pressure for Slow Rod Withdrawal...

42

iv XN-NF-82-32 (NP,'evision 2

.a LIST OF FIGURES (Cont. )

~Fi use P acae 4.15 Thermal Power for Locked Rotor with 0ffsite Power................................

43 4.16 4.17 Core Heat Flux for Locked Rotor with Offsite P ower........................................

Core Flow for Locked Rotor with Offsite Power 44

~

~ ~ ~ ~ ~

45 4.18 4.19 RCS Temperatures for Locked Rotor with Offsite Power - Hot Leg, Core Average, Cold Leg, and Core Inlet Temperatures............................

46 Pressurizer Pressure for Locked Rotor with 47 Offsite Power......................................

4.20 Scram Reactivity for Locked Rotor with 48 Offsite Power......................................

4.21 Thermal Power for Locked Rotor with Loss of Offsite Power...................................

49 4.22 Core Heat Flux for Locked Rotor. with Loss of Offsite Power...................................

50 4.23 4.24 4.25 Core Flow for Locked Rotor with Loss of 0ffsite Power................................

RCS Temperatures for Locked Rotor with Loss of Offsite Power - Ho" Leg, Core Average, Cold

Leg, and Core Inlet Temperatures.............

Pressurizer Pressure for Locked Rotor with Loss of Offsite Power........................

~

~

~

~

~

~

51

~

~ ~ ~

~

~

52

~

~

~

~

~

~

53 4.26 Pressurizer Liquid Volume for Locked Rotor with Loss of Offsite Power 54 4.27

Scram Reactivity for Locked Rotor with Loss of Offsite Power.............

~

~

~

~ ~ ~

~

~ ~

~ ~

~

55 4.28 Thermal Power for Loss of Load.....................

XN-NF-82-32(NP)

Revision 2

L'IST OF FIGURES (Cont.)

~Fi ere

~Pa e 4.29 Core Heat Flux for Loss of Load....................

57 4.30 4.31 RCS Temperatures for Loss of Load - Hot Leg, Core Average, Cold Leg, and Core Inlet emperatures......................................

T Pressurizer Pressure for Loss of Load..............

58 4.32 4.33 4.34 4.35 4.36 4'7 4.38 4.39 4.40 Pressuri'zer Liquid Volume for Loss of Load.........

60 Steam Generator Pressure for Loss of Load..........

61 Steam Flow for Loss of Load........................

62 Thermal Power for Decreased Feedwater Heating......

Core Heat Flux for Decreased Feedwater Heating.....

64 RCS Temperatures for Decreased Feedwater Heating - Hot Leg, Core Average, Cold Leg, and Core Inlet Temperatures.......,..................

65 Pressurizer Pressure for Decreased Feedwater Heating e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~

~

~ ~ ~

~

~

~ e

~

~

~

~

~

~

~

~

~

~

66 Pressurizer Liquid Volume for Decreased Feedwater Heating..................................

67 Core Reactivity for Decreased Feedwater Heating....

6B 4.41 4.42 4.43 4,44 Steam Generator Enthalpies for Decreased Feedwater Heating:

Bottom of Downcomer, Top of Downcomer, and Feedwater........................

69 Steam Generator Pressure for Decreased Feedwater Heating..................................

70 Steam Generator Power for Decreased Feedwater Heating..................................

71 Thermal Power for Decreased Feedwater Heating with Automatic Rod Control.........................

72 f

vi XN-NF-82-32 (NP',

Revision 2

LIST OF FIGURES (Cont.)

~Fi ure

~Pa e

4.45 4.46 4.47 4.48 Core Heat Flux for Decreased Feedwater Heating with Automatic Rod Control.............

RCS Temperature's for Decreased Feedwater Heating with Automatic Rod Control - Hot Leg, Core Average, Cold Leg, and Core Inlet Temperatures................-.."".."..""..

Pressurizer Pressure for Decreased Feedwater Heating with Automatic Rod Control Pressurizer Liquid Level for Decreased Feedwater Heating with Automatic Rod Control

~

~ ~ ~

73 74 75 76 4.49 Core Reactivity for Decreased Feedwater Heating with Automatic Rod Control...............

77 4.50 4.51 4.52 Control Rod Speed for Decreased Feedwater Heating with Automatic Rod Control RCCA Reactivity for Decreased Feedwater Heating with Automatic Rod Control Thermal Power for Load Increase................

78 79 80 4.53 Core Heat Flux for Load Increase..'.................

81 4.54 RCS Temperatures for Load Increase

- Hot Leg, Core Average,,Cold

Leg, and Core Inlet Temperatures......................................

82 4.55 4.56 4.57 4.58 Pressurizer Pressure for Load Increase.............

83 I

Pressurizer Liquid Volume for Load Increase........

84 Core React ivity for Load Increase..................

85 Steam Generator Pressure for Load Increase.........

4.59 Steam Generator Power for Load Increase............

4.60 Steam Flow for Load Increase.......................

vii XN-NF-82-32 (NP',

Revision 2

LIST OF FIGURfS (Cont.)

F inure 4.61

~Pa e

Thermal Power for Load Increase with Automatic Rod Control 89 4.62 Core Heat Flux for Load Increase with Automatic Rod Control..............................

90 4.63 4.64 RCS Temperatures for Load Increase with Automatic Rod Control - Hot Leg, Core Average, Cold Leg, and Core Inlet Temperatures..............

91 Pressurizer Pressure for Load 'Increase with Automatic Rod Control..............................

92 4.65 Pressurizer Liquid Volume for Load Increase with Automatic Rod Control.........................

93 4.66 Core Reactivity for Load Increase with Automatic Rod Control..............................

94 4.67 Control Rod Speed for Load Increase with Automatic Rod Control..............................

95 4.68 RCCA Reactivity for Load Increase Qith Automatic Rod Control..............................

96

XN-NF-82-32(NP)

Revision 2

1.0 INTRODUCTION

Plant transient analyses-are presented to support future cycle operation of the D.C. Cook Unit 2 nuclear power plant at 3425 MWt with SX average steam generator tube plugging.

The major impacts of steam generator tube plugging are a slight reduction in primary coolant flow rate and a slight degradation of primary to secondary system heat transfer.

These effects motivated a re-analysis of those events previously demonstrated to be limiting with respect to thermal margin and reactor vessel pressurization.

An asymmetric tube plugging distribution can adversely affect thermal margin in the locked rotor accident and is therefore assumed as the basis for analysis in the simulation of that event.

Three of the steam generators were assumed 6.7X plugged, and the steam generator with the locked rotor was assumed to be unplugged.

Plant transient analyses to support the 3425 MWt core power level have been reported previously(>-).

Section 2.0 of this report provides a

summary of the results for this analysis.

Section 3.0 describes the calculational methods and input parameters employed.

A more detailed description of individual event simulations is given in Section 4.0.

Those events which are considered in the plant FSAR'(2) but not reanalyzed here're discussed in Section 5.0.

'D

XN-NF-82-32{NP)

Revision 2

2.0

SUMMARY

The plant transient analysis reported here has been performed to evaluate the response of the Donald C.

Cook Unit 2 reactor core and reactor protection system (RPS) during anticipated operational occurrences (AOOs) and postulated accidents (PAs).

The analysis supports operation of ENC reload fuel in D.C.

Cook Unit 2 at a core power of 3425 MWt with an average steam generator tube plugging level of 5X.

The fuel and vessel design criteria to be satisfied in the analysis are listed in Table 2.1.

The key results of the analysis are given in Table 2.2.

The results confirm that applicable fuel and vessel design criteria are met for the previously identified limiting FSAR transients.

The least MDNBR calculated for any AOO event occurred in a slow rod withdrawal event and is well above the XNB correlation safety limit of 1.17.

Peak pressure calculated for any event was 2575 psia, and occurred in the loss of load event.

This is well below the criteria at maximum system pressure of 2750 psia.

The MDNBRs for DNB-limiting transients reported in Table 2.2 have been calculated using ENC's automated-crossflow core thermal hydraulics simulation methodology(3);

A confirmatory analysis of the locked rotor with concurrent loss of offsite power has also been performed.

Radiological release for this postulated accident is within 10 CFR 100 limit.'(")

The transient events reanalyzed and reported here comprise an adequate scope of analysis to assure the safe operation of D.C.

Cook Unit 2 with 5X steam generator tube plugging.

Those anticipated operational occurrences and postulated accidents whose results most closely approach specified acceptable

XN-NF-82-32(NP)

Revision 2

fuel design limits or the vessel pressurization limit have been reanalyzed and have been shown to satisfy applicable criteria.

Section 5.0 demonstrates that the results of the FSAR events are bounded by the results of events which have been analyzed here or that the significant conclusions of the FSAR analyses of these events remain valid under the conditions of this analysis.

Thermal margin for the Cycle 5 (XN-2) core is significantly greater than that of the Cycle 4 core due to a reduction in the F<H which was analyzed and N

to the larger number of ENC fuel assemblies present in Cycle 5.

An F~NH of 1.55 was analyzed which reduces the peak assembly power by 3.15, relative to the previous analysis(1) value of 1.60.

Further, the larger number of ENC assemblies present in Cycle 5 significantly reduces the coolant mass flux penalty suffered by ENC fuel in Cycle 4.

The combined effect of these factors on available thermal margin more than outweighs the effects of the calculated 1.1X RCS flow reduction due to tube plugging.

Core safety limits and the overtemperature AT (OT-AT) reactor trip setpoint reported in the previous analysis(>)

are conservatively applicable to Cycle 5 and future cycles with an F~H of 1.55.

Adequate functioning of the, N

OT-bT trip function has been demonstrated in the rod withdrawal analyses presented in Section 4.1.

Of the FSAR events analyzed, the slow rod withdrawal is the most limiting AOO.

The calculated MDNBR for this event included allowances for uncer-tainties in core operating conditions as discussed in Section 3.0.

The event tripped on OT-bT, which includes all applicable measurement and calibration uncertainties.

Thus, the reported MDNBR values in Table 2.2 for the events

XN-NF-82-32(NP)

Revision 2

which tripped on OT-hT include a

double accounting for 'ore parameter uncertainties.

Elimination of doubly accounting for these uncertainties in the MDNBR evaluation results in the MDNBR increasing to 1.44 for the slow rod withdrawal transient.

The slow rod withdrawal transient remains the limiting AOO analyzed.

XN-NF-82-32(NP)

Revision 2

Table 2.1 Applicable Fuel and Vessel Oesign Limits Event Class Anticipated Operational Occurrence A

1 icable Criteria Peak System Pressure

< 2750 psia MDNBR Calculated with XNB Critical Heat Flux Correlation

> 1.17 Postul ated Acc ident Radiological Release below 10 CFR 100 Limits Peak System Pressure

< 2750 psia

Table 2.2 Summary of Results Transient Max imum Power Level (MWt Maximum Core Average Heat Flux (Btu/hr-ft2)

Maximum Pressurizer Pressure sia)

MDNBR (XNB)

Initial Conditions for Transients Uncontrolled Rod Withdrawal 9

8.4 x 10 3 hp/sec Uncontrolled Rod Withdrawal 8

7.4 x 10-6 gp/sec Loss of Flow-Locked Rotor Locked Rotor*

Loss of Load Decreased Feedwater Heating 3425.

4616.

4140.

3824.

3996.

3610.

3743.

197580.

220260.

234360.

199700.

198690.

200760.

215910.

2250.

2300.

2320.

2308.

2380.

2526.

2252.

1.878 1.625 1.265 1.276 0.698 1.878 1.679 Excessive Load Increase 4048.

Decreased Feedwater Heating with Automatic Rod Control 3767.

217130.

232630.

2260.

2252.

1.627 1.532 Excessive Load Increase with Automatic Rod Control 4042.

229180.

2260.

1.569 With concurrent loss of offsite power.

Radiological release is within 10 CFR 100 limits.

XN-NF-82-32(NP)

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3.0 CALCULATIONAL METHODS AND INPUT PARAMETERS The D.C.

Cook Unit 2 plant transient analysis was performed using the Exxon Nuclear Plant Transient Simulation Model for. Pressurized Water Reactors (PTSPWR2)(5).

The PTSPWR2 code is an Exxon Nuclear digital computer program which models the behavior of pressurized water reactors under normal and abnormal operating conditions.

The computer code is based on the solution of the basic transient conservation equations for the primary and secondary coolant systems.

The transient conduction equation is solved for the fuel

rods, and a point kinetics model is used to evaluate the core neutronics.

The program calculates fluid conditions such as flow, pressure, mass inventory and steam quality, heat flux in the core, reactor power, and reactivity during the transient.

Various control and safety system components are included as necessary to analyze postulated events.

A hot channel model is included to trace the departure from nucleate boiling (DNB) during transients.

The DNB evaluation is based on the hot rod heat flux in the high enthalpy rise subchannel and uses the XNB correlation(<) to calculate the DNB heat flux.

Model features of the PTSPWR2 code are described in detail in Reference 5.

Calculational methodology employed in this analysis is in accordance with ENC standard plant transient analysis methodology for PWRs(7).

A diagram of the system model used by PTSPWR2 is shown in Figure 3.1.

As illustrated, the PTSPWR2 code models the reactor, two independent primary coolant loops including 'all major components:

pressurizer,

pumps, steam generators, and the steam lines, including all major valves (turbine stop valves, isolation valves, pressure relief valves, etc.).

PTSPWR2 loop 2 is a lumped loop model of D.C.

Cook Unit 2 primary loops 2, 3 and 4.

XN-NF-82-32(NP)

Revision 2

The present calculations were performed using the NOV76A version of ENC's PTSPWR2 code with appropriate updates.

Updates are included to describe the D.C.

Cook Unit 2 plant control systems.

Steady state measurement and instrumentation errors are taken into account to ensure conservatively calculated values of MONBR.

The corres-ponding plant initial conditions in the MONBR calculations are as follows:

Reactor Power 3425 MWt + 2X (68.5 MWt) for calor ometric error.

Inlet Coolant Temperature 542.2

+ 4oF for deadband and measurement error.

Primary Coolant System Pressure 2250 - 30 psia for steady state fluctuation and measurement

'rrors.

XN-NF-82-32(NP)

Revision 2

Primary Coolant Flow*

143.1 Mlbm/hr - 3.5X for measurement uncertainty.

The simultaneous application of the above parameter uncertainties minimizes the initial minimum ONB ratio in a bounding fashion.

It is noted that the above steady state errors are not generally included in the plant system modeling, but rather are used to conservatively bound the calculated MDNBR.

Table 3.1 shows a list of operating parameters used in the analysis.

Unless otherwise noted, the transient simulations reported herein have assumed that pressurizer spray and power-operated relief valves are fully operable in order to maintain system pressure at a minimum value.

This results in the most conservative estimation of the MDNBR.

These pressure control functions are assumed inoperable in those events simulated for comparison to the system pressurization criteria.

The trip setpoints incorporated into the PTSPWR2 model for D.C. Cook Un.it 2 are based on the Technical Specification limits.

These limiting trip setpoints are modeled in the plant transient analysis to demonstrate the adequacy of the reactor protection system for operation at a 3425 MWt rating with 55 steam generator tube plugging.

Reactor trip setpoints and scram delay times associated with them are listed in Table 3.2.

Adequate allowance has been made for trip instrument channel measurement uncertainties and cali-bration errors.

The ENC fuel design parameters for O.C.

Cook Unit 2 are summarized in Table 3.3.

Table 3.4 lists the neutronics parameter values which are

• Value includes a 1. 15 reduction from the current measured flow of 144.7 Mlbm/hr to account for increased loop resistance due to 5%%d steam generator tube plugging.

10 XN-NF-82-32(NP)

Revision 2

calculated to conservatively bound the D.C.

Cook Unit 2 core for both the beginning and end of cycle.

A design axial power profile with a peaking factor FZ

= 1.55 was used in the analysis.

This profile is shown in Figure 3.2.

The scram reactivity curve used in the analysis is shown in Figure 3.3.

This curve is taken from the O.C.

Cook Unit 2 FSAR.(2)

Scram delay times employed in the plant transient simulations are sufficiently conservative with respect to Technical Specif ication 1 imits on reactor trip system performance to assure conservative simulation of reactor scram.

In Figure 3.3, the scram reactivity is normalized to the total rod worth.

XN-NF-82-32(NP)

Revision 2

Table 3.1 Operating Parameters Used in PTSPWR2 Analysis of Donald C.

Cook Unit 2 Core Total Core Heat Output, MWt Total Core Heat Output, MBtu/hr Heat Generated in Fuel, X

System Pressure, psia 3425.

11,688 97.4 2250 Hot Channel Factors Total Peaking Factor, Fq T

Enthalpy Rise Factor, F~

H N

Coolant Flow Rate, Mlbm/hr Effective Core Flow Rate, Mlbm/hr Coolant Average Temperature, oF Heat Transfer Average Heat Flux, Btu/hr-ft2 Steam Generators Total Steam Flow Mlbm/hr, per lead Steam Temperature, oF Steam Pressure, psia Feedwater Temperature, oF Tube Plugging, X

2.47 1.55 138.1 131.9 574.1 197,580 3.70 518.

799.

431.

5.0

Table 3.2 Oonald C.

Cook Unit 2 Trip Setpoints

~Set oint Used-in Analysis

~oe1 a T ime High Neutron Flux Low Reactor Coolant Flow High Pressurizer Pressure Low Pressurizer Pressure Low-Low Steam Generator Water Level Overtemperature hT*

109K 90K 2400 psia 1965 psia 21K of span TAyE

= 574.1oF Po

= 2250 psia K]

= 1.267 K3

=.000926 e

llSX 87K 2425 psia 1940 psia OX of span TAyE

= 574.1oF Po

= 2250 psia Kl

= 1.452 K3

=.000744 0.5 sec 1.0 sec 2.0 sec 2.0 sec 2.0 sec 6.0 sec The overtemperature bT trip is a function of pressurizer

pressure, coolant average temperature, and axial offset.

The TAyE and Po setpoints, and the setpoint bias Kl are contained within the functional relationship.-

The bias constant Kl employed in the analysis includes allowance for applicable trip channel uncertainties.

Other constants in the overtemperature hT setpoint as it appears in the Technical Specification (gains, lead and lag constants) were incorporated without change in the analysis.

13 XN-NF-82-32(NP)

Revision 2

Table 3.3 Donald C.

Cook Unit 2 Fuel Design Parameters-Exxon Nuclear Fuel Fuel Radius Inner Clad Radius Outer Clad Radius Active Length Number of Fuel Rods in Core

.1515 inches

.1550 inches

.1800 inches 144.0 inches 50,952

14 XN-NF-82-32(NP)

Revision 2

Table 3.4 Donald C.

Cook Unit 2 Kinetics Parameters Supported by the Plant Transient Analysis Parameter Beginning-

~of-C cle Value End-of-

~Cc 1 e

-Moderator Coefficient

(~p/oF x 104)

+0.5

-3e9 Doppler Coefficient (lp/oF x 105)

-1.0

-1.7 Pressure Coefficient (hp/psia x 106)

-.60

+4.3 Delayed Neutron Fraction (X) 0.61 0.510 Total Rod Worth (X gp) 4.00 4.00

15 XN-NF-82-38(NP)

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TUROINE THROTTLE VALVE RELIEF A!ID SAFETY VALVES STEAN HEADER ATMOSPHERIC OUHP Bypass Valve ATHOSPHERIC RELIEF NID SAFETY DUMP VALVES STEAN LlliE STEAW LINE tiOOE 2 SL 22 ISOLATIOCT VALVE STEAt1 LlttE STEAtl LINE 1

N 0 1

SL 11 ISOLATlel VALVE STEA'1 DONE STEAN 00ttE FEED'A'lER SEPARATORS AttD ORTO5 SP 2 REI.IEF VALVES SAFETY VALVES SEPARATORS IVID DRYERS SP I FEED TER SPRAT PR PRESSURIZER HEATER UP UPPER PLEtilll OUTLET INLET SG 2 INLET OUTLET P CIUH PLEtettt REACTOR COOLANT PUMPS P2, 3

A 4 LP LNIER PLEIR1AI REACTOR COOLANT PUMP $ 1 80RATED COLO LEG HATER INJECTION

~gP EHTHALPT JlttO 80ROtl COICEttTRATICtt TlliE DELAY EERIE REREEERCC COLO LEG INJ ECT Iet 80RATED HATER Figure ".1 PTSPMR2 System Model

1.0 0.5 0

.8 0

.2

.3

.4

.5

.6

.7 (l3ottom)

Axial Height (Fraction of active fuel length)

Figure 3.2 Axial power profile used in transient analysis of 0.

C.

Conk Ijnit 2

.9

.0 (Top)

X) OC 8 K C

I Ql 0

CO I

M GJ o

0.8 0.6 G:

C)z 0.<

0.2 3.5 0

0.5 1

1.5 2

2.5 3

TIVE FIN"TER SCRBt1 (SECONOS)

Figure 3.3 Scram Curve Used in the Transient Analysis for D.C.

Cook Unit 2 R OC 8 K

(

I EA O

CO M PO I

fO GJ a

18 XN-NF-82-32(NP)

Revision 2

4.0 TRANSIENT ANALYSIS 4.1 UNCONTROLLED ROD WITHDRAWAL The withdrawal of control rods adds reactivity to the reactor core causing both the power level and the core heat flux to increase.

Since the heat extraction from the steam generator remains relatively constant, there is an increase in primary coolant temperature.

Unless terminated by manual or automatic action, this power mismatch and the resultant coolant temperature rise could eventually result in a

DNB ratio of less than

1. 17.

While the inadvertent withdrawal of control rods is unlikely, the reactor protection system is designed to terminate such a transient while maintaining an adequate margin to DNB.

Two potential causes for such an incident are:

1) operator error; and 2) a malfunction in the reactor regulating system or rod drive control system resulting in continuous withdrawal of a control rod group.

In this incident, the reactor may be tripped by an overtemperature LT setpoint, the high nuclear overpower

setpoint, or the overpower g T setpoint.

Additionally, the primary coolant temperature increase is limited in magnitude by the steam generator safety valve setpoint.

A series of rod withdrawal simulations was performed at various reactivity insertion rates to demonstrate the adequacy of the reactor protection system for this event.

Figure 4.1 summarizes the results of this study.

Ample margin to DNB is demonstrated for the range of possible rod withdrawal events.

Figures '.2 through 4.7 show plant responses for a fast rod withdrawal from full power.

The reactivity insertion rate is 8.42 x 10 3

AA'sec.

A nuclear overpower trip (118K) occurs at 0.13 seconds.

The DNB ratio

19 XN-NF-82-32(NP)

Revision 2

drops from an initial value of 1.878 to 1.625.

Pressure increases to a maximum of 2300 psia, with core'verage temperature increasing 4.4oF.

The event sequence for the fast rod withdrawal is given in Table 4.1.

The system responses to a slow rod withdrawal of 7.42 x 10 6~sec are depicted in Figures 4.8 through 4.14.

The overtemperature LT trip setpoint is reached at 67.3 seconds.

The minimum DNB

- ratio during the transient is 1.265.

The event sequence for the slow rod withdrawal is given in Table 4.2.

Part power rod withdrawal analyses are discussed in Section 5.0.

4.2 LOCKED ROTOR In the unlikely event of a seizure of a primary coolant pump, flow through the core is abruptly reduced.

The reactor is tripped by the resulting low flow signal.

The coolant enthalpy rises, decreasing the margin to DNB.

The locked rotor transient was analyzed assuming four loop operation with instantaneous seizure of one pump from 3425 NWt.

This case was shown in the reference cycle analysis to be more severe with respect to DNB penetration than a locked rotor with three loop operation.

A second case was analyzed which assumed a concurrent loss of offsite power.

The transient responses for the Locked Rotor event are shown in Figures 4. 15 to 4.20.

The reactor is scrammed at 0.03 seconds by a low flow signal.

A 1.0 second scram delay time is conservatively assumed.

Core

'verage temperature increases 11.5oF with system pressure reaching 2308 psia at 3.8 seconds.

The MDNBR for the locked rotor is 1.276 at 1.9 seconds.

20 XN-NF-82-32(NP)

Revision 2

A second locked rotor event which additionally assumes a loss of offsite power is also simulated.

The intact primary coolant pumps are assumed to initiate a coastdown concurrently with the locked rotor.

Calculated MDNBR for the event is 0.698.

Radiological release for this event is bounded by the LOCA accident and is within 10'FR 100 limits. (

)

This event has not been analyzed in the FSAR and is not considered to be part of the plant licensing basis.

Plant response to this event with maximum pressurization assumptions is depicted in Figures 4.21 to 4.27.

Pressurizer pressure control systems have been assumed inoperable in order to maximize pressure response.

The MDNBR has been evaluated with minimum pressure response assumptions.

The event sequences for the locked rotor cases are given in Table 4.3.

4.3 LOSS OF EXTERNAL ELECTRICAL LOAD This simulation considers plant behavior upon a trip of the turbine-generator" without a direct reactor trip.

The event is simulated to assess the adequacy of the pressurizer safety valve capacity to maintain reactor coolant system pressure below the ASME code limit of 1105 of design pressure (2750 psia).

Transient responses are evaluated from 3425 MWt for the most severe pressurization accident:

loss of load at beginning-of-cycle (BOC) with a positive moderator coefficient and no automatic reactor control.

Figures 4.28 to 4.34 depict the plant responses following a loss of load from full power.

A high pressure trip occurs at 6.98 seconds, with peak pressurizer pressure reaching 2526. 1 psia.

The first set of steam line safety valves opens at 11.9 seconds, relieving 45K of the steam flow.

The setpoint of the second set of safety valves is reached at 18 seconds.

The average

21 XN"NF-82-32(NP)

Revision 2

primary coolant temperature increases 24.5oF above the nominal value.

The MDNBR does not decrease below its initial value.

The event sequence for the loss of load is given in Table 4.4.

4.4 DECREASED FEEDWATER HEATING Failure of bleed steam to any of the six pairs of feedwater heaters could result in a 75 Btu/lb decrease in feedwater enthalpy.

The event is simulated by imposing a

15 second feedwater enthalpy ramp at a rate of -5 Btu/lb-second.

Results of the decreased feedwater heating event are given in Figures 4.35 through 4.43.

A new steady state is established relatively early in the transient.

The MDNBR of 1.679 characterizes this steady state.

The overtemperature hT reactor trip precludes penetration of the XNB critical heat flux correlation safety limit during transients such as the decreased feedwater heating event which are characterized by slow excursions in core power, coolant temperature, and pressure.

A second case was simulated assuming automatic rod control.

A bounding EOC D-bank worth consistent with power dependent insertion limits was employed (-1.2X hp).

Primary side response is depicted in Figures 4.44 through 4.49.

Figures 4.50 and 4.51 demonstrate RCCA control action.

The MDNBR for the event is 1.627 and occurs at 258 seconds.

Secondary system response is similar to that depicted in Figures 4.41 through 4.43.

Event sequences for these events are given in Table 4.5.

4.5 EXCESSIVE LOAD INCREASE INCIDENT Excessive load incidents may be initiated by sudden opening of the turbine control

valves, steam dump

valves, and/or the steam bypass to

22 XN-NF-82-32(NP)

Revision 2

condenser valve.

This results in rapid increase in steam flow which causes cooldown of the primary system.

Automatic Rod Control action or a large negative (EOC) moderator coefficient can result in a

power increase.

Protection against damage to the reactor core as a consequence of an excessive load increase is provided by the high nuclear flux, low steam generator

pressure, and overtemperature 6 T setpoints.

A rapid 20K load increase is simulated at end-of-cycle conditions.

To minimize the calculated

MONBR, the pressurizer heaters are assumed inoperab le.

System responses to the 20K load increase are shown in Figures 4.52 through 4.60.

The power increase dr iven by moderator cooldown continues until the high nuclear overpower trip setpoint (118K of rated power) is reached at 56.5 seconds.

At the time of trip, significant primary system depres-surization has further reduced available thermal margin.

The decreasing primary coolant temperature mitigates the thermal margin decay, resulting in an MONBR for the event of 1.532 shortly after reactor trip.

The case with automatic rod control is shown in Figures 4.61 through 4.66.

Automatic Rod Control action is demonstrated in Figures 4.67 and 4.68.

Control action resulting largely from the temperature deviation channel results in a more rapid power increase than observed in the uncontrolled case.

To maximize power response, the temperature control program employed a linear gain between 0 and 120K of rated turbine demand.

The reactor trips on high nuclear flux at 21.2 seconds.

The MONBR of.1.569 occurs shortly thereafter.

Event.sequences for the load increase transients are given in Table 4.6.

23 XN-NF"B2-32INP)

Revision 2

Table 4. 1 Event Sequence for Fast Rod Withdrawal Event Time (seconds)

Uncontrolled RCCA Bank Withdrawal begins High Neutron Flux Setpoint reached Scram Results in Rod Motion 0.0 0.13 0.63 Minimum DNBR occurs 1.55

24 XN-NF-82-32(NP)

Revision 2

Table 4.2 Event Sequence for Slow Rod Withdrawal Event Uncontrolled RCCA Bank Withdrawal begins Overtemperature hT Setpoint reached Scram Results in Rod Motion Minimum DNBR occurs Time (seconds) 0.0 67.3 73.3 74.0

25 XN-NF-82-32(NP)

Revision 2

Table 4.3 Event Sequence for Locked Rotor Event Time (seconds)

CASE 1:

Locked Rotor with Offsite Power Available Single Primary Coolant Pump seizes Loop Low Flow Trip 'Setpoint reached Scram Results in Rod Motion Minimum DNBR occurs Peak RCS Pressure reached 0.

0.03 1.03 1.9 3.2 CASE 2:

Locked Rotor with Concurrent Loss of Offsite Power Single Primary Coolant Pump seizes Loop Low Flow Trip Setpoint reached Scram Results in Rod Motion Minimum DNBR occurs Peak RCS Pressure reached 0.

0.03 1.03 2.4 4.0

26 XN-NF-82-32,'NP)

Revision 2

Table 4.4 Event Sequence for Loss of External Load Event Time (seconds)

Loss of Load High Pressurized Pressure Setpoint reached 0.

6.98 Scram Results in Rod Motion 8.98 Peak Pressure reached 18.1

27 XN-NF-82-32(NP)

Revision 2

Table 4.5 Event Sequence for Decreased Feedwater Heating Event Time (seconds)

CASE 1:

Uncontrolled "Feedwater Enthalpy Begins to Decrease from the Steady State Value at 5. Btu/lb/second 0.

Feedwater Enthalpy reaches Minimum Value 15.0 Minimum DNBR reached (NOTE: Reactor scram does not occur.)

260.

CASE 2:

Automatic Rod Control Feedwater Enthalpy Begins to Decrease from the Steady State Value at 5. Btu/lb/second 0.

Feedwater Enthalpy reached Minimum Value Minimum DNBR reached 15.0 258.

28 XN-NF-82-32(NP)

Revision 2

Table 4.6 Event Sequence for Excessive Load Increase Event Time (seconds)

CASE 1: Uncontrolled BOX Increase in Load Demand reached High Neutron Flux Trip Setpoint reached Scram Results in Rod Motion 10.

56.0 56.5 Minimum DNBR occurs 57.0 CASE 2: Automatic Rod Control 205 Increase in Load Demand reached High Neutron Flux Trip Setpoint reached Scram Results in Rod Motion 1.0 21.2 21.7 Minimum DNBR occurs 21.5

t1DNBR VS INSERTION RATE High Nuclear Flux Trip Overtemperature hT Trip lxlO-5 lxl0 4 lxl0-3 lxlO-2 INSERTION RATE (DELYA-RHO/SEC)

Figure 4.1 NDNBR vs. React'ivity 'Insertion Rate, for Rod Withdrawal Transients X7 OC tD C/l O

CO hD I

PO ~

hJ a

FRST ROD NITHDRRl<RL B.E-3 DKiSEC 3500 3000 2500 2000 1500 LEGEND o-PL 1000 QX S K C

I Eh I

0 CO W fO I

0 0

2 3

5 6

7 IINE (SEC) 9 10 ll figure 4.2 Thermal Power for Fast Rod Withdrawal

F'RST ROD HITHDRRNRL 8. E-3 DK/SEC CU 3000000 I

2500000 LEGEND o

Q' 2000000

~~

1S00000 1000000 500000 0

5 6

TISE tsE..")

Figure 4.3 Core Heat Flux for Fast" Rod N'ithdrawal 9

10 11 RX 8 R C

I De g 0 Q)

~ PO I

M ~

PO O

. 610 FRST ROO HITHORRNRL 8. E-3 OK/SEC 580 4

C9 439 sso LEGEND o -

TCIO 0-TCFI TCL1 THL1 P

s70 560 o-sso X7 >C 8

I Eh I

O CO I

5 6

TJHE (SEC) 10 11 Figure 4.4 RCS Temperatures for Fast Rod Withdrawal - Hot Leg, Core Average, Cold Leg, and Core Inlet Temperature

F'RST RDD NITHDRFIWRL 8. E-3 DK/SEC 2250 LEGEND o-PPR 2200 lK 07 21SO 43 lA 2100 20SO 2

3 5

6 TIt1E t SEC)

S 10 11 Xl OC 8 R

(

I Ul ~

I O CO I

o Figure 4.5 Pressurizer Pressure for'Fast Bod Withdrawal

))00 FRST ROD NITHDRRj~RL 8.E-3 DK/SEC 4

SSD C3 LEGENO CFRPR GJ 800 (3

850 C3 800 700 0

s e

TIt.~E tSEC) 9

)0 11 Figure 4.6 Pressurizer Liquid Yolume for Fast Rod Withdrawal

F'RST RDD NITHDRRHRL 8.E-3 DK/SEC SSO Q

800 43 CY M

Cf) 850 0

LEGEND P001 800 750 0

5 6

7 TIVEt SEC) 9 10 11 XlX S R I

Vl I

O 00 I

PO (A a

SLQN ROD NITHDRRNRL B.E-6 DK/SEC 4000 LEGEND D-PL 2500 ta3 C)~

20eO 1500 l0 20 30 40 SO 60 70 80 80 100 110 rrriC ~Scci Figure 4.8 Thermal Power for Slow Rod Withdrawal R OC 8 2:

(

IZ Ul-I O C)

I o

%000000 SLOW ROO NITHDRRNRL B.E-B DK/SEC 3500000 3000000 2500000 2000000 1S00000 1000000 C7 C3 Q)0000 10 20 30 00 SO 60 70 80 80 100 110 TINE (SE")

Figure 4.9 Core Heat Flux for Slow Rod Withdrawal

SLOW RDD WITHDRRWRL B. E-6 DK/SEC 620 6)0

~

600 CS Cl SSO V) 580

~~

570 4l 560 540 0

)0 20 30 40 50 60 70 80 80

)00

))0 TINEtSEC)

Figure 4.10 RCS Temperatures for Slow Rod Withdrawal - Hot Leg, Core Averaae.

Cold Leo and Core Inlet Temeratures LEGEND o -

TGIO 0-TCR TCL1 o

THLl

,R OC 8 R

(

I lh I

O CO fQ I

PO a

as SLON ROD NITHDRRNRL 8.E-6 DK/SEC 4

8S C3 LEGEND

< - DELTSP Cl 80 lY C3 7S C3 70 65 10 20 30 40 50 60 70 80

. 90 100 110 TIVE(SEC)

Figure 4.11 Calculated Overtemperat'ure al" Setpoint for Slow R Oc e M I

Mo~

Ql O CO D M I

~ Col M

2400 SLOW ROD WITHDRRWRL B.E-B DK/SEC 43 2100 2000 1800

~

1800 1700 10 20 30 00 50 60 70 80 80 100 110 TII1E t SEC)

Figure 4.12 Pressurizer Pressure for Slow Rod Withdrawal

SLON ROD NITHDRRNRL 8.E-6 DK/SEC 1300 1200 1100 C3 1000 800 600 10 20 SO

~O SO 60 Z0 80 80 1OO>>O IIf1E (5EC)

Figure 4.13 Pressurizer Liquid Volume for,Slow Rod Withdrawal

1100 SLON ROD NITHDRRNRL B.E-6 DK/SEC tA 4l 800 CO

'43 SS0 7SO 0

10 20 30 CD 50 60 70 80 90 100 110 TI (ATE (SEC j Figure 4.14 Steam Generator Pressure for Slow Rod Withdrawal A7 X 0)

I C/l ll O

CO I

U

4000 LOCKED ROTOR WITH OF'F'SITE POWER 3500 3000 LEGENO PL 2500 2000 1500 1000 SOD 0

0.5 4.5 5

S.S 1.5 2

2.S 3

3.S TINE (SEC)

Figure 4.15 Thermal Power for Locked Rotor with Offsite Power R >C lD(

I 4o g Vl I

0 CO I

PO GJ

3400000 LOCKED ROTOR WITH OF'FSITE POWER 3200000 3000000 2800000 CQ 2600000 OC 2400000 4J 2200000 2000000 1800000 0.5 1.5 2

2.S 3

3.S 4.5 5

T I l1E (5CC)

Figure 4.16 Core Heat Flux for Locked Rotor with Offsite Power S.S XI >C S R

(

I lA me I O

CO FO I

U

<0000 LOCKED ROTOR WITH OF'F'SITE POWER 38000 36000 l/)

Kl 34000 C) 4 32000 C)

C3 LEGEND NLPCR 30000 28000 0

O.S 1.5 2

2.5 3

TINE (SEC) 3.S i.S 5

5.5 RK ED ~

I lh Tl O Q)

N M I

M 4JM U

Figure 4.17 Core Flow for Locked Rotor with Offsite Power

6]0 LOCKEO ROTOR NiTH OFF'SITE PONER 600 590 4

C9 4J9 560 LEGEND 0--

TC/0 0-TGB TCL1 THLl 570 560 C3~

SS0 S<0 0.5 1

1.5 2

2.5 3

TI t1E (SEC) 3.5

<.5 5

5.5 X7 'K 8 R

(

I Me R Vl ll I~

O CO I

~ 4J D

Figure 4.18 RCS Temperature for Locked Rotor with Offsite Power-Hot Leg, Core Average, Cold Leg, and Core Inlet Temperatures

2310 LOCKED ROTOR WITH OF'F'SITE POWER 2300 22SO CC lA~

2280 43 2270 2260 2250 22<0 2230 2220 0

0.5 1.5 2

2.5 3

TIl1EtSEC) 3.5 4.S 5

S.S LEGENO 0-PPR XlX e M I

ao 2 Vl ll I

O CO I

~ Ca) a Figure 4.19 Pressurizer Pressure for Locked Rotor with Offsite Power

LOCKED ROTOR PITH OF'F'SITE PONER LEQENO o "

DKGTL Q

-3 C3 0

0.5 1.5 2

2.5 3

T I l1E (SEC) 3.5 4.5 5

5.5 XI>C lD ~

C I

Mo K lh 0 0)

I PO lA a

Figure 4.?0 Scram Reactivity for Locked Rotor with Offsite Power

LOCKED ROTOR W/0 OF'F'SITE POWER LEGEND 0~

PL 43 2000 0

]SOD M

1000 0

1 2

3 4

5 6

7 8

9 10 ll 12 TIt1E (SEC)

Figure 4.21 Thermal Power for Locked Rotor with Loss of Offsite Power 77 A 8 R C

I2 Eh ll wo0 Co I

M 4O o

LOCKED ROTOR W/0 OF'FSITE POWER 2500000 Kl 2000000 4

1500000 43x:

1000000 3

0 5

6 1

8 S

10 11 12 TleE tscc)

Figure 4.22 Core Heat Flux for Locked Rotor with Loss of Offsite Power

LOCKED ROTOR W/0 OF'FSITE POWER 35000

~CD 3OOOO tf)

X:

Q3 2S000 C) 4

~

20000 C)

CD LEGEND a -

HLPCR 15000 10000 2

3 0

S 6

7 8

9 10 1 1 12 TIt1E (SEC)

Figure 4.23 Core Flow for Locked Rqtor with Loss of offsite Power Xl>C 0 R

(

I Ql O CO I

M o

610 LOCKED ROTOR N/0 OFFSITE PONER SSO 4

C9 43 SBO

!.EQEND 0-TCIO o -

TCR TCL1 THLl 570 K

Cd 560 C3~

55O Sio 0

1 Figure 4.24 2

3 4

5 6

7 8

9

}0 11 12 TIt1E (SEC)

RCS Temperatures for Locked Rotor with Loss of Offsite Power-Ilot Leg, Core Average, Cold Leg, and Core Inlet Temperatures K OC

0) M

(

I Ul O CO I

PD ~

FO

LOCKED ROTOR N/0 OFTSITE POWER 2350 LEGEND PPR 4l 2250 Q

~

2200 2150 2100 0

1 2

3 4

5 6

7 8

S 10 11 12 Xl >C lD M I

lh ll O CO I

PO a

TI l1E (SEC)

Fi ure 4.25 Pressurizer Pressure for Locked Rotor with Loss of Offsite Power

1100 LOCKED ROTOR W/0 OF'F'SITE POWER 1050 4

)OOO Ll CD SSO C3 CD SOD 850 0

1 2

5 6

7 8

9 10 1) 12 IIt1E (SEC)

Figure 4.26 Pressurizer Liquid Volume for Locked Rotor with Loss of Offs ite Power

LOCKED ROTOR W/0 OF'F'SITE POWER V)

"2 o

-3 C3 LEGENO DKCTL "5

C3 td

<<6 CC 7

C3 V)

"8

<<9 0

8 9

10 11 12 2

3 4

5, 6

7 TIVE(SEC)

Figure 4.27 Scram Reactivity for Locked Rotor with Loss of Offsite Power Zl )C 8 R

(

I Vl Tl me O CO I

PO Ca>

FO a

0000 LOSS OF LOBO DC COOK 2 3500 3000 2500 2000 1500 1000 500 0

0 8

10 12 TIt1E (SEC) 16 18 20

ÃIK 8 2.'

I lh I

O CO I

m GJ a

Figure 4.28 Thermal Power for Loss of Load

3500000 LOSS OF'ORD DC GOOK 2 3000000 2500000 CQ 2000000 OC 1500000 1000000 500000 0

2 0

6 8

10 12 TI I"IE (SEC)

Figure 4.29 Core Heat Flux for Lossof Load 16 18 20

630 LOSS.OF'ORD DC COOK 2 620 610

~

6OO C9 43 C3 590 580

~~

570 43 560 SSO 580 8

10 12 TIf1E (SEC) 16 18 20 LEQENO o -

TCIO 0-TCR TCL1 o -

THL1 R OC S R

(

I Ul O CO W PO I

bJ lA Figure 4.30 RCS Temperatures for Loss of Load - Hot Leg, Core Average, Cold Leg, and Core Inlet Temperatures

2550 LOSS OE LORD DC COOK 2

2500 (C

V) 2450 LEGENO o-PPR 2<00 2350 43 2300 2250 8

10 12 TI HE (SEt')

Figure 4.31 Pressurizer Pressure for Loss of Load 16 18 20 R OC 8 R

(

I Wo I O

CO I

U

1450 LOSS OI'ORD DC COOK 2 1 F00 1350 1300 C3 1250 1200

)n 1150 1100 Q

1050 1000 0

8 10 12 TIt1E (SEC)

Figure 4.32 Pressurizer Liquid Volume for Loss of Load 16 18 20 LEQENO a -

CF'HPR R 0K 8 K Ip Ul O CO I

~ GJ o

1150 LOSS OF'ORD DC COOK 2 1100 1050 LEGDID 0 "

P001 1000 V)

CL 950 43 900 850 lA 800 750 8

10 12 TI t1E (SEC)

F igure 4.33 Steam Generator Pressure'or L'oss of Load 16 18 20 P Oc tD M

(

I

> Z Ul

~o I 0 CO D FO I

~ CA U

4500 LOSS OF'ORD DC COOK 2 4000

~

3500 C3 43 V) 300D LEGEND o - NDPSLT 250D 2000 lA 150D a=

1000 500 0

8 10 12 II t1E (SEC)

Figure 4.34 Steam Flow for Loss of Load 16 18 20

ÃI>C 8 2:

I Vl Tl I

0 CO W PO I

PO o

DECRERSED F'EEDNRTER ENTHRLPY-DC COOK 2 3750 3700 3650 3600 Ca3 C3 0

3550 (C

43 3500 3<50 3400 50 100 150 T I t1E (SEC) 200 250 300 F igure 4. 35 Thermal Power for Oecreased

'Feedwater Heat ing

DECRERSED EEEDWRTER ENTHRLPY-DC CQDK 2 3S50000 3SOOO00 3450000 3500000 OC 3350000 3300000 C) 3280000 3200000 0

50 100 1SO 200 TI (ATE (sec) 250 Figure 4.36 Core Heat Flux for Oecreased Feedwater Heating 300

DECRERSED F'EEDWRTER ENTHRLPY-DC COOK 2'10 SSO 4

~

SeO C3 LEGEND o -

TCIO 0-TCR TCL1 THLl 570 560 550 540 530 0

50 100

'150 T I /ATE (SEC) 200 250 300 Xl OC lD(

I Vl ll O

CO PO I

M CO D

figure 4.37 RCS Temperatures for Oeereased feedwater Heating-

DECRERSED F'EEDWRTER ENTHRLPY-DC COOK 2 22SS 22SO CC lA~

22<S 43 22<0 la3~

22SS 2230 2225 50 100 1SO 200 TIHE (SEC) 250 300 F igure 4.30 Pressurizer Pressure for Decreased Feedwater Heating

DECRERSED F'EEDNRTER ENTHRLPY-DC COOK 2 1030 1025 1020 101S C3 1010 1005 O

C3 1000 LEGEND GF'Nf'R 995 990 885 SO 100 150 T1f1E (SEC) 300 RX 8 R I

0 CO 1

U Figure 4.39 Pressurizer Liquid Volume

DECRERSED F'EEDWRTER ENTHRLPY-DC CDQK 2

0. 20

0. 15

0. 10 CC

0. 05 C)

C3

0. 00 "0. 05 C3 4l "0. 10

-0. 15 "0. 20 0

250 100 50 150 T t f1E (SEC)

Figure 0.40 Core Reactivity for Decreased Feedwater Heating 300 LEGEND o-OK o-OKHQO

~ - OKPRES OKOOF' Oc 8 K

(

I lh Il I

O CO I

ru ~

U

DECRERSED F'EEDNFITER ENTHRLPY-DC COOK 2 ISO 460 440 Gl 420 lA

<00 0

380 4J

~

360 4

lA 3iO 320 SO 100 1SO 200 TIl1E (SEC) 250 300 LEGEND HB01 o -

HF'Nl HT01 R >C 8 R I2 Ch~. I O 00 W PO I

PO GO D

Figure 4.41 Steam Generator Enthalpies" for Oecreased Feedwater

'r

DECRERSED F'EEDWRTER ENTHRLPY-DC COOK 2 795 790 785 tA 780 lA 775 V) 770 cs 765 760 755 50 100 150 TIl1E (SEC) 200 250 300 R OC S

I Ul

~0 I

O CO I

PJ 4J Figure 4.42 Steam Generator Pressure for Decreased Feedwater Heating

DECRERSED F'EEDWRTER ENTHRLPY-OC COOK 2 220000 215000 t.ECCNO QOB 210000 Kl 205000 C9 200000 V) 195000 50 100 150 200 II f1E (SEC)

Figure 4.43 Steam Generator Power for Decreased Feedwater Heating

DEC.

F'EEDNRTER ENTHRLPY-DC,COOK 2-RRC 37SO 3700 3500 150 200 TIt1E (SEC)

Figure 4.44 Thermal Power for Decreased Feedwater fleating with Automatic Rod Control X7 0<

8 2:

(

I Ul ll O 0)

W POI a

DEC.

FEEDHRTER ENTHRLPY-DC COOK 2-RRC 3500000 3450000 34M000 35SD 000 330Il000 C3 C3 3250000 XlK 8 R

(

I Vl ll I

0 CO I

50 100

]50 200 KO 300

, TIf1E(SEC)

Figure 4.45 Core Heat Flux for DecreasedFeedwater Heating with Automatic Rod Control

DEC.

F'EEDWRTER ENTHRLPY-DC COOK 2-RRC 610

~

S80 CS Cl S80 cA 43 S70 tC 43 550 540 0

)SO 200 TIt1E (SEC)

Figure 4.46 RCS Temperatures for Decreased Feedwater Heating with Automatic Rod Control - Hot Leg, Core Average, Cold Legana t;ore Inde.

temp"r-cur~'EGENO TCIO o-TGA TCL1 o -

THL XlOC 8 K

(

I EA I

O CQ I

PO lA FO o

2280 OEC.

FEEOWRTER ENTHRLPY-OC. COOK 2-RRC LEGEND PPR 22C CD M I

Mo~

Vl I

0 CO I

M4J U

28.5 J

OEC.

F'EEOHRTER ENTHRLPY-DC COOK 2-RRC LEGEND LEVN CO 26.5 0

150 KO TINE (SEC)

Figure 4.40 Pressurizer Liquid Level for Oecreased Feedwater Heating with Automatic Rod Control

0.05 DEC.

F'EEDWRTER ENTHRLPY-DC COOK 2-RRC 0.00

.05

~~ -0.]0 LEGEHD o-OK o -

OKgQO 4 - OKPRES OKOOP "0. 15

-0.X

-0. 25 lSO 200 TIt1E (SEC)

Figure 4.49 Core Reactivity for Decreased Feedwater Heating with Automatic Rod Control Xl>C (D M I

lh Tl I

O CO K fO I

M~

o

0.0 DEC.

FEEDWRTER ENTHRLPY-DC COOK 2-RRC C3td Q

CO J

CD

-0.6 CD C3

~Q 7 "0. 8 SO 100

]SO 200 250 71OE (SEC)

Figure 4.50 Control Rod Speed for Decreased Feedwater Heating with Automatic Rod Control

0. 45 DEC.

F'EEDNRTER ENTHRLPY-DC COOK 2-RRC

0. 40 0.35 0.30

0. 20 LEGEND a

OKRCCfl O. 20

0. 1S 43 (C

0. l0 C3 CD 0.05 0.00 lSO 200 TIVEtSEC)

Figure 4.51 ACCA Reactivity for Decreased Ft:edwater Heating with r

R OC 8 K

(

I Vl I

O CO I

PO ~

a

EXCESSIVE LORD INCRERSE-DC COOK 2 3500 2SOO

~

20OO 1SOO 1000 10 20 30 10 TIt1E (SEC)

Figure 4.52 Thermal Power for Load Increase

EXCESSIVE LORD INCRERSE-DC COOK 2 3500000 (Yi 4-

'M$0000 I

2500000 LEGEND n-Pf OC.

2000000

~~

.>sooooo C)

>OO0000 500000 0

10 Fi ure 4.5 30 40 TIVE tSCC~

~ t 60 70 XlK 8 R

(

I Vl 0 00 I

M VP O

610 EXCESSIVE LORD INCREASE-DC CQQK 2 600 590 4

~

560 LEGEND o -

TCIO o-TCR TCL}

o -

THL1 570 560 550 SCQ 5'30 Figure 4.54 20 60 30

~0 50 TIVE(SEC)

RCS Temperature'es for Load Increase

- Hot Leg, Core

Average, Cold Leg, and Core Inlet Temperatures XIK 8 2.'

Vl ll I

O CO PD I

rv ~

U

EXCESSIVE LORD INCRERSE-DC COOK 2 22SO 2200 LEGEND PPR 2150

~

2100 20SO 2000 1850 0

10 20 30 40 TIt1E (SEC) 70 X >c 0)(

I Wo M

1l I

0 CO I

M ~

PQ Fi ure 4.55 Pr

>r'

1100 EXCESSIVE LORD INCRERSE-DC COOK 2 1000 C3 700 C3 a00 Q

500 10 20 30 i0 TIAC (SEC j 70 Figure 4.56 Pressurizer Liquid Volume for Load Increase

EXCESSIVE LORD INCRERSE-DC COOK 2

tA CC p

a C3 2

Q 4JlK LEGEND OK o-OKHOO

~ - OKPRES OKOGP

-8 10 30 40 T1HE(SEC) 60 70

% OC 0 R I

J EA I

O Po

'D FO I

PO 4J a

kJ Fiqure 4.S7 Core Reactivit

SSO EXCESSIVE LORD INCRERSE-DC COOK 2 850 V)

Q 800 4J V) af)

IS0 Q

LEGEND PD01 CS 700 10 20 30

<0 TI I1E (SEC) 60 70 RK 8 R

(

I 4o2 Ul~. I 0 00 K PO I

PO o

f figure 4.58 Steam Generator Pressure for Load Increase

EXCESSIVE LOBO INCRERSE-DC COOK 2

LEGEND QOA

>soooo 43 (Q

100000 C9 SDOOO U)

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~.

I 0 00 I

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10 F i ure 4.59 Steam 30 en T ICE iSEC) 60 70

5000 EXCESSIVE LORD INCRERSE-OC COOK 2

<S00 4000 c3 43 3500 Kl 3000 a

2500 CC 2000 V) 1500

>000 SOO 0

la 20 30

<0 II f1E (SEC) 60 LEGEND o - NDOSLT RK 0 R C

I lh I

O CO I

~ GJ PD U

Figure 4.60 Steam Flow for Load Increase

EXC..

LORD INC.-DC COOK 2-RRC 4000 3500 3000 LEGEND PL 2500

~

20M 1500 1000 500 0

0 10 15 20 TIt1E (SEC) 25 30 35 X7 0<

8 R C

I We +

Vl M I

0 CO N M I

PO a

Figure 4.61 Thermal Power for Load Increase with Automatic Rod Control

~ I

4000000 EXC.

LORD INC.-DC COOK 2-RRC 3500000

~~

3OOOOOO I

2500000 Kl LEGEND QT 2000000

~~>sooooo C3 ioooooo 500000 0

Figure 4.62 25 30 ls 20 TIt1E (SEC)

Core Heat Flux for Load Increase with Automatic Rod Control 35 XlX 8 RI EA ll O CO I

PO ~

PO lg

620 EXC.

LORD INC.-DC COOK 2-RRC 610

~

Sso C9 43 Cl 580 570

~~

560 550 540 530 10 15 20 TIVE(SEC) 25 30 35 LEGEND TCIO 0-TCR TCL1 o -

THL1

à >C CD M I

O CO Figure 4.63 RCS Temperatures for Load Increase with Automatic Rod Control - Hot Leg, Core Average, Cold Leg, and Core

2300 EXC.

LORO INC.-OC COOK 2-RRC 2250 CC 2200 0

LEGEND n-PPR 43 2150 2100

~~

2050 2000 1950 10 15 20 TIt1E (SEC) 30 R >C tb'(

I ao Z Ul ll I

O CO W PO I

~ GJ o

Figure 4.64 Pressurizer Pressure for Load Increase with Automatic Rod Control

1100 EXC.

LORO INC.-DC COOK 2-RRC 1000 800 C3 LEGEND CFNPR la3 soo C3 Cl 100 600 500 10 15 20 TIVE (SEC) 30 3S Xl>C 8 K I

Vl I

O CO I

M~M Figure 4.65 Pressurizer Liquid Volume for Load Increase with 1

EXC.

LORD INC.-DC COOK 2-RRC Vl lK 0

(C.

C3 C) 2 D

bl LEGENO a-OK o-OKt100

> - OKPRES OKOOP

-6 10 5

]5 20 25 30 TIt1E (SEC)

Figure 4.66 Core Reactivity for I oad Increase with Automatic Rod Control 35 XI X 8 2:

I g

0 COM I

M CA U

0.0 EXC.

LORD INC.-DC COOK 2-RRC 0,2 C3 43 lA "0.3 LEGEND

~ - RSPEED C3~ -0V 0U)

-0.5 "0.6 C)

C3 C3

-0.7

-0.8 10 15 20 TIME(SEC) 25 35 R OC fD M C

IZ lA ll I

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Figure 4.67 Control Rod Speed for Load Incre'ase with Automatic Rod

0. 30 EXC.

LORD INC.-DG COOK 2-RRC 0.25 lY 0.20 C3 C3

0. 15

0. 10 43 lY.

LEGENO a - OKRCCR C3

0. 05

0. 00 10 15 20 TItlat (SEC) 25 30 35 R OC lb M I

EA O 00 I

a Figure 4.68 RCCA Reactivity for Load Increase with Automatic Rod Control

97 XN-NF-82-32(NP)

Revision 2

5. 0 DISCUSSION The transient analysis as performed by ENC for Donald C.

Cook Unit 2 nuclear power plant demonstrates adequate margin to applicable fuel and vessel des ign 1 imits for a

mixed ENC/West inghouse core dur ing normal operation, anticipated operational occurrences, and postulated accidents.

The following transients were analyzed using the ENC PTSPWR2 plant transient simulation model at a core power of 3425 MWt.

1)

Rod Withdrawals between 8.42 x 10-3 and 7.42 x 10-6~p/s 2)

Locked Primary Coolant Pump Rotor 3)

Locked Primary Coolant Pump Rotor with Concurrent Loss of Offsite Power 4)

Decreased Feedwater Heating 5)

Excessive Load Increase 6)

Loss, of Load These transients were considered because they were shown in the D.C. Cook Unit 2 FSAR(2) (reference analysis) to have the least margin to thermal margin limits.

The applicable fuel and vessel design limits for the transients are a

minimum DNB ratio of

1. 17 calculated with the XNB critical heat flux correlation and a peak system pressure of 2750 psia.

For the locked rotor

accident, the fuel design criterion is that a small fraction of the core may experience boiling transition.

For the locked rotor with concurrent loss of offsite power, radiological release may not exceed 10 CFR 100 limits.

Other transient events considered in the reference analysis are not reanalyzed here, either because the reference analysis results remain valid

,98 XN-NF-82-32(NP)

Revision 2

for those events under the conditions of this analysis, or because other events which have been reanalyzed. here have been shown in the reference analysis to be more limiting.

The reference analysis considered RCCA withdrawal transients initiated from a variety of core power levels equal to or less than 3391 Ml<t.

The full i

power cases are shown to be the most limiting of the cases considered with respect to MONBR.

The analyzed 5X steam generator tube plugging level coupled with the, loading of ENC fuel will not affect this fact, and the full power cases will continue to be the most limiting of the RCCA withdrawal events under the conditions of this analysis.

Part power RCCA withdrawal cases are therefore not reanalyzed.

The results of the full power RCCA withdrawal event also bound the possible results of the Chemical and Volume Control System Malfunction transient.

During this transient, reactivity is added to the core by the addition of unborated primary coolant makeup water.

The system response is similar to that for the slow rod withdrawal transient analyzed in Section 4.1, r

with a reactivity insertion rate of about 1.0 x 10 5 dk/sec.

This insertion rate is bounded by the range of insertion rates included in this analysis.

The reference analysis of the RCCA drop transient demonstrated

that, neglecting radial power distribution effects associated with the event, the MONBR monotonical ly increases in time from the initial value.

The MONBR which occurs during the event may therefore be conservatively evaluated by a steady

99 XN-NF-82-32(NP)

Revision 2

state MDNBR calculation performed at rated initial conditions of core power, temperature,

pressure, and flow, and which employs a radial peaking factor augmentation to account for the adverse core radial power distribution which characterizes the event.

The radial peaking augmentation factor at 3425 MWt core power for the mixed core loading considered here is 1.2.

A steady state MDNBR calculation employing this peaking augmentation factor and performed as described will result in an MDNBR well above the XNB correlation safety limit of 1.17.

Since the MDNBR calculated during the transient will not exceed the steady state MDNBR thus obtained, it is concluded that the result of the RCCA drop transient at 3425 MWt meets the fuel design limit on MDNBR.

The loss of normal feedwater simulation reported in the reference analysis was performed at the Engineered Safety Features design thermal power of 105K, of rated.

This power level exceeds the 3425 MWt rated power assumed in this analysis.

Steam generator tube plugging of 5X will result in less than 3oF higher primary coolant temperatures than shown in the reference analysis.

Calculations indicate that this small increase in primary coolant temperature will not result in the expulsion of primary liquid from the pressurizer safety relief valves.

Tube plugging will decrease the tendency for steam generator dryout due to reduced heat transfer effectiveness.

The result of the loss of normal feedwater event presented in the reference analysis will not therefore be significantly impacted by 5'A steam generator tube plugging.

The startup of an inactive loop was shown in the reference analysis to be significantly less limiting than the uncontrolled RCCA withdrawal event.

Since neither the loading of ENC fuel nor the analyzed 5X steam generator tube

100 XN-NF-82-32(NP)

Revision 2

plugging level will alter the relative severity of these two events, the results of the uncontrolled RCCA withdrawal will continue to bound the results of the inactive loop startup.

The inactive loop startup event is therefore not reanalyzed here.

The results of the loss of AC power event were shown in the reference analysis to be enveloped by the results of the four pump coastdown and loss of normal feedwater events.

The flow degradation aspect of the loss of AC power event has been reanalyzed here in Subsection 4.2 as a loss of flow transient (locked rotor with loss of offsite power).

Adequate long term decay heat removal is demonstrated by the loss of normal feedwater simulation reported in the reference analysis.

Results of the loss of AC power event have therefore been adequately bounded by the combination of the 4

pump coastdown event reported in Subsection 4.2.

The loss of normal feedwater event is discussed above..

Results of the small steam line break reported in the reference analysis are judged to remain valid for the conditions of this analysis.

The event is independent of rated

power, since it is initiated from hot zero power conditions.

Core kinetics parameter tables employed in the reference analysis bound the core configurations considered in this analysis.

The impact of steam generator tube plugging is to reduce primary to secondary heat

transfer, increasing primary to secondary system temperature differences.

The system temperature datum is established by the magnitude of the break flow, which is conservatively considered to be independent of tube plugging level.

Primary system temperatures in this event will then be increased by tube

101 XN-NF-82-32(NP)

Revision 2

plugging, with consequently lesser requirements for shutdown margin.

The small steam line break is therefore not reanalyzed since the reference analysis is bounding.

The main feed line break (MFLB) event was shown in the reference analysis to be independent of fuel type, since applicable fuel design limits are never approached during the event (MDNBR increases monotonically).

The loss of normal feedwater (LONF) event results in greater volumetric expansion of the primary liquid than occurs during the MFLB because primary coolant expansion during the MFLB is mitigated by extraction of primary heat due to steam generator blowdown.

Adequate auxiliary feedwater system capacity to prevent uncovery of the core was demonstrated in the reference analysis of the LONF event.

Since the conclusion of that analysis is judged to remain valid for the 3425 MWt rating with 5X tube

plugging, and bounds the primary coolant expansion and thus the potential for core uncovery in the MFLB, it is concluded that the, core will remain covered throughout an MFLB initiated from the 3425 MWt level.

Analysis of the MLFB event is therefore not considered.

The-rod ejection transient is addressed in Reference 8.

The results of certain operational incidents are not significantly dependent on fuel type or small changes in rated power level.

These include:

RCCA Misalignment

~

Turbine Generator Overspeed

~

Fuel Handling Incident

~.

Accidental Waste Gas Release

102 XN-NF-82-32(NP)

Revision 2

Radioactive Liquid Release

~

Steam Generator Tube Rupture These incidents as discussed in the reference cycle analysis were shown to be protected by administrative controls, redundancy of alarms, and/or integrity of system components.

The conclusions drawn for these incidents as given in the reference analysis remain valid and these events are not reanalyzed here.

103 XN-NF-82-32(NP)

Rev>s>on 2

6.0 REFERENCES

(1)

XN-NF-82-32, Rev.

1, "Plant Transient Analysis for the Donald C.

Cook Unit 2 Reactor at 3425 MWt," Exxon Nuclear Company, Inc., Richland, WA, April 1982.

(2)

Donald C.

Cook Unit 2 Nuclear Plant Final Safety Analysis Report, as completed in 1982.

(3)

XN-NF-75-21, Rev.

2, "XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Opera-tion," Exxon Nuclear Company, Inc., Richland, WA, September 1982.

(4)

XN-NF-82-90, Supplement 1,

"D.C.

Cook Unit 2 Potential Radiological Consequences'f Incidents Involving High Exposure Fuel," Exxon Nuclear Company, Inc., Richland, WA, February 1984.

(5)

XN-74-5, Rev.

2, "Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTSPWR)," Exxon Nuclear Company, Inc., Richland, WA, May 1975.

(6)

XN-NF-621(A)-,

Rev.

1, "Exxon Nuclear DNB Correlation for PWR Fuel Design,"

Exxon Nuclear Company, Inc., Richland, WA, April 1982.

(7)

XN-74-5, Rev.

2, Supp.

2, "Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR),

Supplement 2:

Methodology and Applicat.ions,"

Exxon Nuclear

Company, Inc., Richland, WA, January 1984.

(8)

XN-NF-83-85, "D.C. Cook Unit 2 Cycle 5 Safety Analysis Report,"

Exxon Nuclear Company, Inc., Richland, WA, October 1983.

XN-NF-82-32(NP)

Revision 2

Issue Date: 3/3/B4 PLANT TRANSIENT ANALYSIS fOR THE DONALD C.

COOK UNIT 2 REACTOR AT 3425 NWt OPERATION WITH 5X STEAN GENERATOR TUBE PLUGGING Distribution F. T. Adams J.

C. Chandler R. A. Copeland N. F. Fausz W. V. Kayser T.

R. Lindquist G. F. Owsley H. G.

Shaw G. A. Sofer R. B.

Stout'.

Tahvili AEP/H.G.

Shaw (10)

Document Control (5)