Text

Plant Transient Analysis of the Operation at 2530 Mwt
	ML18347B304
Person / Time
Site:	Palisades
Issue date:	07/18/1977
From:	Correll G, Kahn J, Koester G, Vandewalle D; Exxon Nuclear Co
To:	; Office of Nuclear Reactor Regulation
References
	XN-NF-77-18
	Download: ML18347B304 (164)
	v • d • e

I XN-NF-77-18 I

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PLANT TRANSIENT ANALYSIS OF THE PALISADES I REACTOR FOR OPERATION AT 2530 Mwt I

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I JULY 1977 j I

I I EJ${.ON NUCLEAR COMF'ANV, Inc.

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I I Issue Date: 07/18/77 XN-NF-77-18 I

I PLANT TRANSIENT ANALYSIS OF THE PALISADES I REACTOR FOR OPERATION AT 2530 Mwt JULY 1977 I

,, Prepared By G. E. Koester J. D. Kahn G. R. Correll D. J. VandeWalle (CPCo)

I I -~1---=-"-~~L-J_,*--,_;_.-.

Approved:) e,/<-£*1.:>*,.-..._ /-r-r--;'-,/,/,1,1_,!.1/l I /

K. P. G~1braith, Manger

//./

Nuclear Safety Engineering I Approved: ~~7-ff-77 G. A. So~anag I Fuel Design and Engineering I L. H. Steves, Manage I Contract Performance

~k~ I ;, H/71 /

Approved:~.--uh. ci~i-.____...------- . /

-~ W. S. Nechodom, Manager Licensing and Compliance I

I I EJ${0N NUCLEAR COMPANY, Inc.

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I IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY I This technical report was d.erived through research and develop-ment programs sponsored by Exxon Nuclear Company, Inc. *It is being I

submitted by Exxon Nuclear to the USNRC as part of the tec.hnical con-tribution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear-fabricated reload fuel o~ other techni-cal services provided by Exxon Nuclear for light water power reactors I

and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this repprt, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of compliance with the USNRC's regulations.

I Without derogating from the foregoing, neither Exxo.n Nuclear nor any person acting on its behalf: I A. Makes any warranty, expressed or implied, with respect to the accuracy, completeness, or useful-ness of the information contained.in this docu-ment, or that the use of any information, apparatus, method, or process disclosed in this document

- I B.

will not infringe privately owned right~.

Assumes any liabilities* with respect to the use of, or for damages resulting fro~ the use of, any information, apparatus, method, or process a

disclosed in this document.

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I I XN-NF-77-18 I*

TABLE OF CONTENTS I Page I

1.0 INTRODUCTION

AND

SUMMARY

2.0 CALCULATIONAL METHODS AND INPUT PARAMETERS.

l 5

I 3.0 TRANSIENT ANALYSES . . . . . . . . 17

.3. l UNCONTROLLED ROD ~JITHDRAWAL. 17 I 3.1.l Control Rod Withdrawals From 102% of Rated Power. 18

.I 3.1.2 Control Rod Withdrawals From 52% of Rated Power 3.2 CONTROL ROD DROP INCIDENT. . .

19 43 I 3.3 LOSS OF COOLANT FLOW INCIDENTS 51 52 3.3.l Four-Pump Coastdown I 3.3.2 Locked Rotor . . . . 53 I 3.4 EXCESSIVE FEEDWATER INCIDENTS.

3.4.l Reduction in Feedwater Enthalpy 66 66 I 3.4.2 Increased Feedwater Flow From 52% Power 67

,, 3. 5 EXCESSIVE LOAD INCREASE INCIDENT . . . .

3.5. l Excessive Load From 102% of Stretch Power 78 79 3.5.2 Excessive Load From Hot Standby 79 I 3. 6 LOSS OF LOAD INCIDENT. 92 I '

3.7 LOSS OF FEEDWATER FLOW INCIDENT.

3.8 STEAM LINE BREAK . . . . . . . .

111 118 I 3.8.l Steam Line Break From 102% 6f Rated Power (2530 Mwt) . . . . . . . . . . . . 120 1* 3.8.2 Steam Line Break From Hot Standby 120 I

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I TABLE OF CONTENTS (Conti n~.ed)

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3. 9 SINGLE ROD WITHDRAWAL.

3.10 ROD EJECTION INCIDENT.

137 144 I

4.0 REFERENCES

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LIST OF TABLES

I Table

-*-- Pag~

I 1. 1

2. 1

SUMMARY

OF RESULTS . . . . . . . . . . . .

NOMINAL OPERATING PARAMETERS USED IN 3

PTSPWR2 ANALYSIS OF PALISADES AT 2530 Mwt 8 I 2.2 PALISADES FUEL DESIGN PARAMETERS EXXON NUCLEAR FUEL 9 I 2.3 2.4 KINETICS PARAMETERS . . . . . . . . . . . . . . .

TRIP SETPOINTS FOR OPERATION OF PALISADES REACTOR 10 AT 2530 Mwt . . . . . . . . . . 11 I 3. 1 LOSS OF LOAD TRANSIENT RESULTS . . . . . . . . . 95 I. 3.2 3.3 MAXIMUM EJECTED ROD WORTHS AND PEAKING FACTORS.

ROD EJECTION INCIDENT ANALYSES PARAMETERS 146 147' I 3.4 ROD EJECTION INCIDENT RESULTS . * . . . . .* 148 I

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LIST OF FIGURES II, Figure

'1, PTSPWR2 SCHEMATIC FOR PALISADES. 12 I 2. 1 2.2. AXIAL POWER PROFILE FOR 102% OF STRETCH POWER OPERATION . . *. . 13 I 2.3 AXIAL POWER PROFILE FOR 52% OF STRETCH POWER OPERATION . . . . 14 I 2.4 PALISADES THERMAL MARGIN LIMITING OPERATING CONDITIONS {BUNDLE POWER, INLET TEMPERATURE, AND PRIMARY PRESSURES) 100% POWER = 2530 Mwt 15 I 2.5 PALISADES SCRAM CURVE. . . . . . . . . . . . 16 I 3. 1 CONTROL ROD l~ITHDRAWAL INCIDENT REACTIVITY ADDITION RATE VS MDNB'R INITIAL POWER LEVEL = 2580. 6 Mwt . 21

1 3.2 CONTROL ROD WITHDRAWAL INCIDENT REACTIVITY ADDITION RATE VS MDNBR INITIAL POWER LEVEL = ~315.6 Mwt 22 I 3.3 POWER HEAT_ FLUX, AND SYSTEM FLOWS, CONTROL ROD WITHDRAWAL FROM 105% POWER

@1.4 X lo-4 Ap/SEC . . . . . . . . . . 23 '*1 I. 3.4 CORE TEMPERATURE RESPONSES, CONTROL ROD WITHDRAWAL FROM 105% POWER@ 1.4 X 10- 4 Ap/SEC . . . . . . . . 24 I 3.5 PRIMARY LOOP TEMPERATURE CH~NGES, CONTROL ROD WITHDRAWAL FROM 105% POWER @1 . 4 X 10- Ap/SEC. .' . . . . . . ~ . . 25 3.6 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, I ~'

CONTROL ROD WITHDRAWAL FROM 102% POWER@ 1.4 X 10-4 Ap/SEC 26

3. 7 ' LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, I* 3.8 CONTROL ROD WITHDRAWAL FROM 105% POWER@ 1.4 X lo-4 Ap/SEC 27 POWER, HEAT fLUX, AND SYSTEM FLOWS, CONTROL ROD I,, 3.9 WITHDRAWAL FROM 102% POWER@ 1.0 X lo-5 Ap/SEC CORE TEMPERATURE RESPONSES, CONTROL ROD WITHDRAWAL 28 FROM 102% POWER@ 1.0 X 10-.5 Ap/SEC . . . . . . . . 29 I

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I LIST OF FIGURES (Continued)

Figure I 3.10 PRIMARY LOOP TEMPERATURE CHANGES, CONTROL ROD WITHDRAWAL FROM l 02% POWER @l . 0 X 10-5 tip/SEC. 30 I 3.11 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATO~S, CONTROL ROD WITHDRAWAL FROM l 02% POWER @l. 0 X lo-

tip/SEC. 31 I

3. 12 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, 3.13 CONTROL ROD WITHDRAWAL FROM 102% POWER@ 1.0 X lo~5 POWER, HEAT FLUX, AND SYSTEM FLOWS, CONTROL ROD

~p/SEC. 32 I

WITHDRAWAL FROM 52% PO~JER @6. 0 X 10-4 tip/SEC . . * . . . . . . 33

3. 14 CORE TEMPERATURE RESPONSES CONTROL ROD WITHDRAWAL I

FROM 52% POWER @6.0 X lo- 4 tip/SEC. . . . . . . . . 34

3. 15 PRIMARY LOOP TEMPERATURE CHANGES, CONTROL ROD WITHDRAWAL I FROM 52% POWER @6.0 X lo-4 tip/SEC. . . . . . . . . . . . 35

3. 16 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERAT~RS, 1*

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CONTROL ROD WITHDRAWAL FROM 52% 'POWER @6.0 X 10- tip/SEC 36 ,*'

3. 17 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS CONTROL ROD WITHDRAWAL FROM 52% POWER @6.0 X lo-4 tip/SEC 37 I

3.18 POWER, HEAT FLUX, AND SYSTEM FLOWS, CONTROL ROD WITHDRAWAL FROM 52% POWER @6. 0 X l o-5 tip/SEC . . . . . . 38

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3. 19 CORE TE~PERATURE RESPONSE~S CONTROL ROD WITHDRAWAL FROM 52% POWER @6.0 X 10 tip/SEC. . . . . . . . . . . . . . 39 I 3.20 PRIMARY LOOP TEMPERATURE CHANGES, CON~ROL ROD

,WITHDRAWAL FROM 52% POWER.@ 6.0 X 10- tip/SEC . . . I--

3.21 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, CONTROL ROD WITHDRAWAL FROM 52% POWER @6.. 0 X l o-5 tip/SEC 40 41 3.22 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS CONTROL ROD WITHDRAWAL FROM 52% POWER @6.0 X 10-~ tip/SEC 42 3.23 POWER, HEAT FLUX, AND SYSTEM FLOWS CONTROL ROP DROP; . . . . . . . . . . . . . . . . . . . . . . 45

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XN-NF-77-18 I LIST OF FIGURES (Continued)

I Figure 3.24 CORE TEMPERATURE RESPONSES, CONTROL ROD DROP . . 46 I 3.25 PRIMARY LOOP TEMPERATURE CHANGES, CONTROL ROD DROP. 47 I 3.26 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, CONTROL ROD DROP. . . . . . . 48

1 3.27 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATOR, CONTROL ROD DROP. . . . 49 3.28 REACTIVITY FEEDBACK, CONTROL ROD DROP . . . . 50 I 3.29 PALISADES PRIMARY COOLANT FLOW COASTDOWN COMPARISON OF PLANT TEST RESULTS AT HOT ZERO POWER WITH RELAP4-EM I. PREDICTION AT 102% OF RATED POWER . . . . 55 3.30 POWER, HEAT FLUX, AND SYSTEM FLOWS, 4 PUMP TRIP 56 I 3.31 CORE TEMPERATURE RESPONSE, 4 PUMP TRIP . . . 57 3.32 PRIMARY LOOP TEMPERATURE CHANGES, 4 PUMP TRIP 58 I 3.33 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS,

),

4 PUMP TRIP . . . . . . . . 59 I 3.34 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, 60 4 PUMP TRIP . . . . . . .

I 3.35 POWER, HEAT FLUX, AND SYSTEM FLOWS, LOCKED ROTOR. 61 3.36 CORE TEMPERATURE RESPONSES, LOCKED ROTOR . . 62

'I 3.37 3.38 3.39 PRIMARY LOOP TEMPERATURE CHANGES, LOCKED ROTOR.

PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, LOCKED ROTOR. . . . . . . . . . . . . .

LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, 63

. . 64 I LOCKET ROTOR. . . . . . . . . . 65 3.40 POWER, HEAT FLUX, AND SYSTEM FLOWS, REDUCTION OF I FEEDWATER ENTHALPY. . . . . 68 I vi I

'I XN-NF-77-18 I I ~-*

LIST OF FIGURES (Continued)

Figure 3.41 3.42 CORE TEMPERATURE RESPONSES, REDUCTION OF FEEDWATER ENTHALPY. ~ . . . . ...

PRIMARY LOOP TEMPERATURE CHANGES, REDUCTION OF 69

'I FEEDWATER ENTHALPY. . . . . . . . . . . . . . . . . . . 70 .I 3.43 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, REDUCTION OF FEEDWATER ENTHALPY . . . . . . . . . . . . . . 71 .1*

3.44 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, 3.45 REDUCT.ION OF FEEDWATER ENTHAL!PY . . . . . . . .

POWER, HEAT FLUX, AND SYSTEM FLOWS, INCREASE IN

. . . . . 72 I

FEEDWATER FLOW AT 52% POWER . . . . . . . . . . 73 ,,,.

.3.46 CORE TEMPERATURE RESPONSES, INCREASE IN FEEDWATER FLOW AT 52% POWER. . . . . . . . . . . . . . . . . . . . . . 74 3.47 PRIMARY LOOP TEMPERATURE CHANGES, INCREASE IN FEEDWATER FLOW AT 52% POWER . . . . . . . . . . . . . . . . . . .

I 75 3.48 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, INCREASE IN FEEDWATER FLOW AT 52% POWER . . . . . . ... . 76 I

3.49 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, INCREASE IN FEEDWATER FLOW AT 52%POWER. 77 I

3.50 POWER, HEAT FLUX, AND SYSTEM FLOWS, EXCESSIVE LOArr INCREASE . . . . . . 81 I

3.51 3~52 CORE TEMPERATURE RESPONSES, EXCESSIVE LOAD INCREASE PRIMARY LOOP TEMPERATURE CHANGES, 82 1 .,,

EXCESSIVE LOAD INCREASE . . . . . . . . . . . . . . ... . 83 3.53 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, EXCESSIVE LOAD INCREASE . . . . . . . . . . . . . . . . . 84 3.54 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, EXCESSIVE LOAD INCREASE . . . . . . . . . . . . . . . . 85 vii

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LIST OF FIGURES (Continued)

I Figure Page I 3.55 POWER, HEAT FLUX, AND SYSTEM FLOWS, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY. . . . 86 3.56 I CORE TEMPERATURE RESPONSES, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY. . . . . 87 3.57 PRIMARY LOOP TEMPERATURE CHANGES, I

I EXCESSIVE LOAD INCREASE FROM HOT STAND-BY. . . . 88

I 3.58 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATOR, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY. . . . . . . . . 89 3.59 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, I 3.60 EXCESSIVE LOAD INCREASE FROM HOT STAND-BY. .

REACTIVITY FEEDBACK, EXCESSIVE LOAD INCREASE

. .... 90 I 3.61 FROM HOT STAND-BY . . . . . . . . . . . . . .

POWER, HEAT FLUX, AND SYSTEM FLOWS, 91 LOSS OF LOAD - CASE l . . . . . . . . 96 I. 3.62 CORE TEMPERATURE RESPONSES, LOSS OF LOAD - CASE l . . . . . . . . . . . . . . . . . . 97 I 3.63 PRIMARY LOOP TEMPERATURE CHANGES, LOSS OF LOAD - CASE 1 . . . . . . . 98 I 3.64 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, LOSS OF LOAD - CASE l . . . . . . . . . . . . . . . 99 l'-

3.65 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, LOSS OF LOAD - CASE 1. . . . . . . . . . . . . . . 100 3.66 POWER, HEAT FLUX, AND SYSTEM FLOWS, I LOSS OF LOAD - CASE 2. . . . . . . . . . . . . . . . . . . . 101 3.67 CORE TEMPERATURE RESPONSES, I 3.68 LOSS OF LOAD - CASE 2. . . .

PRIMARY LOOP TEMPERATURE CHANGES,

. ........... . 102 I LOSS OF LOAD - CASE 2. . . . ........... . 103 viii I

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I XN-NF-77-18 1,

LIST OF FIGURES (Continued)

I. Figure I 3.83 3.84 HIGH PRESSURE SAFETY INJECTION FLOW VS PRESSURE POWER, HEAT FLUX, AND SYSTEM FLOWS, 124 I 3.85 STEAM LINE BREAK . . . . . . . . . . . . . .

CORE TEMPERATURE RESPONSE, STEAM LINE BREAK 125 126 I 3.86 3.87 PRIMARY LOOP TEMPERATURE CHANGES, STEAM LINE BREAK. 127 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, I 3.88 STEAM LINE BREAK . . . . . . . . . . . . . . . . . .

. LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, 128

. STEAM LINE BREAK . . . . . . . . . . .

I 3.89 REACTIVITY FEEDBACK, STEAM LINE BREAK 129 130

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I 3.90 POWER, HEAT FLUX, AND SYSTEM FLOWS, STEAML INE BREAK FROM HOT STAND by . . . . . . . . 131 3.91 I CORE TEMPERATURE RESPONSES, STEAMLINE BREAK FROM HOT STANDBY . . . . . . . . . . . . . . . . . . . 132 3.92 PRIMARY LOOP TEMPERATURE CHANGES, I STEAMLINE BREAK FROM HOT STAND by . . . . . . . . . . 133

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3.93 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, I 3.94 STEAMLINE BREAK FROM HOT STAND BY . . . . . . . . . 134 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS,

.I-* 3.95 STEAMLINE BREAK FROM HOT STAND BY . . . . . . . . .

REACTIVITY FEEDBACK, 135 STEAMLINE BREAK FROM HOT STAND BY 1* 3.96 POWER, HEAT FLUX, AND SYSTEM FLOWS,

. . . . . . . . . . . . . 136 SINGLE CONTROL ROD WITHDRAWAL ........... 139 I 3.97 CORE TEMPERATURE RESPONSES, SINGLE CONTROL ROD WITHDRAWAL . . . . . . . . . . . . . . 140 I

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,1 LIST OF FIGURES (Continued)

Figure I PRIMARY LOOP TEMPERATURE CHANGES, 3.98 SINGLE CONTROL ROD WITHDRAWAL . . . . . . . . . . . . . . . . 141 I.

3.99 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, SINGLE CONTROL ROD WITHDRAWAL . . . . . . . . . . . 142 I

3. 100 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, SINGLE CONTROL ROD WITHDRAWAL . . . . . . . . . . . 143 I

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1.0 INTRODUCTION

AND

SUMMARY

This report presents the results of the plant transient analysis per-formed for operation of the Palisades Plant at a core thermal power level of 2530 MWt. The analysis was performed for the following abnormal occurrences based on 2530 Mwt operation:

Incident*

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Classification Control Rod Withdrawal Incidents II I 2. Control Rod Drop Incident II

3. Loss of Coolant Flow Incidents I - Four Pump Coast Down III I 4.

- Locked Rotor IV Excessive Feedwater Incidents II I

,, 5.

6.

7.

Excessive Load Increase Incidents Loss of Load Incidents Loss of Feedwater Flow Incident II II II I 8. Steam Line Break Incidents IV

9. Uncontrolled Withdrawal of an Individual Control Rod II I I 10. Control Rod Ejection Incident IV I The plant transients 1 through 9 were analyzed using the Exxon Nuclear

,-- Company plant transient simulation code PTSPWR2. (l) The control rod ejection I incident was analyzed using XTRAN. ( 2 ) The criteria to be satisfied for the Class II and III incidents are:

I

Classification consistent with current FSAR Incident Classifications 1- for PWR's.

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I i) peak primary pressure~ 2750 psia ii) MDNBR > 1.30 I iii) primary to secondary differential pressure, ~P ~ 1530 psid.

The criterion of concern for the steam line break is that the end-of-I cycle (EOC) shutdown margin shall be adequate to avoid excessive (>1%) I clad damage. *The criteria for the rod ejection incident isl) that the energy deposition in the fuel as a result of the incident be ~ 280 cal/gm I (to ensure no .fuel meltdown) and 2) that the peak system pressure be less than the vess~l design pressure(~ 2750 psia).

I The results of the analyses are summarized in Table 1.1. The lowest I MDNBR for class II and III events was l .35, which is above the acceptable ~-.

minimum of l.30. The locked rotor incident, a class IV event, was analyzed I and the MDNBR was found to be 1 .27. This result is acceptable for this low probability incident, comparable to a rupture of the primary coolant system.

I In all other cases, there is at least 95% probability with 95% confidence I

that no fuel rod in the core will experience DNB.

The results of the rod ejection incident showed the energy deposition I in the fuel was below the criteria of 280 cal/gm, and the peak transient system pressure was below the criteria of 2750 psia.

I The above analysis is valid for a maximum power peaking, F6, of 2.55 I

and an axial power peaking factor of 1 .40, with the axial peak located at X/L <

60. I Operation of the Palisades reactor at 2530 MWt is therefore justifiable on the basis of the above plant transient analysis and results.

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TABLE l. l

SUMMARY

OF RESULTS Maximum Maximum Maximum Primary-Maximum Core Average Pressurizer Secondary Transient - Class. Po\'1er Level Heat Flux Pressure LIP MDNBRfi (Mwt) (psi a) --

(Btu/hr:n2) (psid)

Initial Conditions*

For Transients 2580.6 169,600 2010 1238 l. 75 Uncontrolled Rod Withdrawal - II Rod Withdrawal @ l .4 x lo-4 lip/sec from 102% Power 2838 183,050 2103 1290 l. 52 Rod Withdrawal @ 1.0 x lo- 5 lip/sec from 102% Power 2833 182,970 2161 1319 l. 45 I Rod Withdrawal @ 6.0 x 10 -4 w I

lip/sec-from 52% Power 3188 143,910 2113 1169 2.00 Rod Withdrawal @ 6.0 x lo- 5 Ap/sec from 52% Power 1942 124, 110 2133 1221 l.89 Control Rod Drop - II 2196 165 ,240 tt ** 1238 1.35 Loss of Coolant Flow Four Pump Coastdown - III 2629 169,600 2073 1240 l. 39 Locked Rotor - IV 2650 169,600 2080 1250 l. 27 Excessive Feedwater Flow Incidents - II ><

z:

Reduction in Feedwater I Entha 1PY - II 2590 169,910 2019 1 . 75

z:

1272 I

-....J Increased Feedwater Flow -....J I

from 52% Power - II 1484 97,500 2036 1240 3.00 co

TABLE l. l (Continued)

Maximum Maximum Maximum Primary-Maximum Core Average Pressurizer Seconda ry Transient - Class Power Level Heat Flux Pressure 6P MDNBR~

(Mwt) (Btu/hr*ft 2 ) (psia) (psid)

Excessive Load From 102% Power - II 2870 178,780 ** 1287 1.74 From Hot Standby - II 258 17,075 ** 1363 3.60t Loss of Load - II 2838 176,415 2394 1388 1. 39 Loss of Feedwater - II 2673 172 ,905 2162 1238 1. 65 Steam Line Break From 102% Power - IV 464 30,960tt *** *** 1. 30t From Hot Standby - IV 694 45,530 ** *** 1. 41 t I

+:>

Uncontrolled Withdrawal I of an Individual Control Rod-III 2841 182,515 2125 1297 1. 44 Control Rod Ejection Incident-IV 399,740 1111 2260 t +

I' +tt

Initial conditions are for 102% of rated power (including measured error and control board allo0ances)._

- Pressure decreases from initial value.(2060 + 50 psia).

- - The criteria on primary secondary 6P is not applicable for steam line breaks.

Calcul~ted using the modified Barnett CHF correlation.

Maximum heat flux after return to power.

Not applicable for control rod ejection incident.

Does not include rod bow penalty.

Average enthalpy of hottest fuel pellet < 247 cal/gm. ><

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2.0 CALCULATIONAL

-- METHODS AND INPUT PARAMETERS I ~ -

The transient analyses for operation of the Palisades reactor .at 2530 I Mwt were performed using the Exxon Nuclear Plant Transient Simulation model for Pr~ssurized Water Reactors (PTSPWR2)_(l) The PTSPWR2 code is a digital I computer program developed to describe the behavior of pressurized water reactors subjected to abnormal operating conditions. The model is based on I the solution of the basic transient conservation equations for the primary I and secondary coolant system, on the transient conduction equation for the fuel rods, and on the point kinetics equation for the core neutronics. The I

/

program calculates fluid conditions such as flow, pressure, mass inventory and quality, heat flux in the core, reactor power, and reactivity during I the transient. Various control and safety system components are included I as necessary to analyze desired transients. A hot channel model is used to evaluate the departure from nucleate boiling ration (DNBR) during transients.

I The DNBR evaluation is based on the hot rod heat flux for the subchannel with the highest enthalpy rise. The W-3 DNB correlation( 3 ) or the modified I 4 Barnett CHF correlation( ) are used to predict DNB or CHF depending on the system conditions. The models contained within PTSPWR2 code are described I in detail in Reference 1. A block diagram of the PTSPWR2 model is depicted I in Figure 2.1 .

. For these analyses, the core parameters calculated using the PTSPWR2 I code were used as boundary conditions to a transient thermal-hydraulic code~ * )

5 6 for evaluation of the minimum DNBR or minimum CHFR. The W-3 correlation was I used to compute DNB heat fluxes at primary system pressures above 1000 psia I

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I and the Modified Barnett CHF correlation was used for system pressure below 725 psia. Between 1000 psia and 725 psia the critical heat flux was deter- I mined by averaging the critical heat flux determined by both correlations.

The initial conditions for the transient analyses are based on steady-I state operations at 2530 Mwt (excluding pumping power) with uncertainties I

applied to ensure a conservative*analysis; i.e. minimize DNBR, ma~imize system pressure, and maximize pressure differential between primary and secondary: I Reactor Power 2530 + 2% Mwt Average Core Inlet Coolant I

Temperature 537.5 + 5°F Primary Coolant System Pressure 2060 ~ 50 psia I The steady-state operating conditions for the core and the hot assembly are summarized .in Table 2.1. The fuel design parameters for the ENC fuel I

are given in Table 2.2. The kineti~s parameters for beginning-of-cycle I (BOC) and end-of-cyc*le (EOC) conditions are listed in Table 2.3. The BOC and EOC moderator coefficients represent bounding values to ensure conservative I

calculations for Cycle 3 as well as future reload cycles .. The BOC and EOC Doppler.coefficients were increased or decreased by 20%, such that the most I

conservative effect during a particular transient was evaluated. The set I of kinetics parameters used for each transient case is described in the sec-tion dealing with the representative transient. I A skewed axial power profile (Fz = 1.4 at Z/L = 0.6) was used for all *1 the transients analyzed, from 102% of rated power (2530 Mwt) while a skewed chopped cosine axial power profile (Fz = 1.75 at Z/L = 0.7) was used for the I I

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analyses from 52% of rated power. The axial power profiles used are depicted I in Figures 2.2 and 2.3.

I was FQ Unless otherwise indicated, the total peaking applied in the analyses T T

= 2.55 for the analyses from 102% of rated power and FQ = 3.74 for I the analyses from 52% of rated power. These peaking factors result in the steady-state MDNBR values of 1.75 and 3.37 at 102% and 52% of rated power I (2530 Mwt), respectively.

I The trip setpoints and their associated delay times to scram are given in Table 2.4. The rod scram curve used in the PTSPWR2 analysis is shown in I Figure 2.5. The time for fuel rod insertion was conservatively taken to be 3.0 seconds from the time of rod release. This is adequate to meet the tech-I nical specification requirement of a minimum of 90% of full insertion at 2.5 I seconds. Parameters dependent on transient type are discussed in each transient description ?ection.

I I

I XN-NF-77 I TABLE 2.1 I

NOMINAL OPERATING PARAMETERS USED IN PTSPWR2 ANALYSIS OF PALISADES AT 2530 Mwt I Core Total Core Heat Output, MWt, I

2530 Total Core Heat Output, Btu/hr 8633.4 x 10 6 I Heat Generated in Fuel,%

System Pressure, psia 97.5 2060 11 Total Coolant Flow Rate, Mlbs/hr Effective Core Flow Rate, Mlbs/hr*

121. 7 114. 5 I

Core Inlet Coolant Temperature, °F 537.5 I Average Core Coolant Temperature, °F 565.

Hot Channel Factors: I Total Peaking Factor, F~

Radial Peaking Factor, Fr 2.55 l.45 I

Heat Transfer I Core Average Heat Flux, Btu/hr-ft 2 168 '180 Steam Generators I Total Steam Flow, Mlbs/hr Secondary Steam Pressure, psia l 0. 97 730 I

Feedwater Temperature, °F 435. I Number of Active Steam Generator Tubes, S.G.#1 6590 I S.G.#2 6775

Includes leakage and flow measurement uncertainty I I

I L

I I XN-NF-77-18 I TABLE 2.2 I PALISADES FUEL DESIGN PARAMETERS EXXON NUCLEAR FUEL I

I Fuel Radius 0. 175 inches I Inner Clad Radius 0.179 inches I Outer Clad Radius 0.207 inches I Active Length 131.8 inches I

Active Fuel Rods I Per Bundle 208 I

I I

I

I XN-N F-77-18 I

TABLE 2.3 I

KINETICS PARAMETERS'

_?ymbol I

Parameter Value Beginning of End of I

aM Cycle Cycle I

Moderator Coefficient (L\p/°F) x 1o4 + 0.50 - 3.50 I aD Doppler Coefficient

( L\p/ o F) x 10 5 - 1 .09 - 1 .38 I

ap Pressure Coefficient I (L\p/psia) x 10 6 - 1 .00 + 7.00 I

av Moderator Density Coefficient I (L\p/lb/ft 3) x 10 3 - .962 6.413 aB Boron Worth Coefficient I

(L\p/ppm) x 104 ) - 0.80 - 1. 00 I

13 eff Delayed Neutron Fraction, % .75 .45 I

a.CRC Net* Rod Worth (% L\p)** - 2.90 - 2.90 I

Total rod worth minus stuck rod worth 2.0% at Hot standby I

I I

I

- ..... - *- .. - - - - - - - -**, -\ *- - - -

TABLE 2.4 TRIP SETPOINTS FOR OPERATION OF PALISADES REACTOR AT 2530 Mwt Setpoint Uncertainty Used in Analysis Delay Time High Neutron Flux 106.5% + 5. 5% 112 .o~; 0.4 sec Low Reactor Coolant Flow 955; + 2.0% 93.0% 0.6 sec High Pressurizer Pressure 2255 psi a + 22 psi 2277 psi a 0.6 sec Low Pressurizer Pressure 1750 psi a + 22 psi 1728 psi a 0.6 sec Low Steam Generator Pressure 500 psia + 22 psi 478 psi a 0.6 sec Low Steam Generator Level* 6 feet +10 in 6 feet 10 in. 0.6 sec I I

Therma 1 Margin** p = f{TH,TC} +165 psi p - 165 psi a 0.6 sec

/

Below ~perating level.

- The thermal margin trip setpoint is a functional pressurizer pressure (P) ><

setpoint, varying as a function of the average hot leg temperature (T ) :z 8 I and the average cold leg temperature (TC). The functional relationship of the -variables is derived from Figure 2.4.

z I

-...J

-...J I

co

I XN-NF-77-18 I

l\J\/'v FLOW RES I STANCE

'0000' ENTHALPY ANO BORON CONCENTRATION TIME OELAY I

I TLJIH!llH. SfOI' VALVL

~ll l\M l!VIWi'.1 I

llf 1'11! !{

VAi VL AIMOSl'HElllC A1MU'51'11l t~ J l.

Ul!MP SAf[IY ;AJ LT Y IJI:MP VAi VES VALVLS VALV['; VAi VLS NOil! I SI 11 NOil!

\Sill.Al \ill! I 'i01 Al ION II I VAi.VF VAi VI 00 2 STEAM OOME DO I I I

FEED SEPARATOR'*

ANO OHYrns

__0J _11[11 WATER RrL !Fr VAi.Vi S SPRAY SAITTY V.~L VI; Si' I WMll' I

l'IIrSSUHI II II PH

',I, I I

I I

I SG 2 SG 2 UP SG l SG l OUTLET INLET INLET OUTLET PLENUM PLENUM UPPER PLENUM PLENUM PLENUM CP CORE I l'tlt11' 21\

II' PUMP If\ I 1.ilWI II l'Ll,NllM PUMP 2B PllMI' 111 I

FIGURE 2.1 PTSP~JR2 SCHEMATIC FOR Pi\LISADES I

I

- - .. - '- .. - - !- - - - -* -', -I - - - -

s...

0

.µ u

'° LL.

01 c:

r-

~

IB l.O 0..

s... I Q) 3 w 0 I 0..

r-x c::(

.5 o,,_~ ..... ~..-.ii....-~_._~__.---~.....1-~~~_.::.,.J..::,.;..:;;...;,;;.::.;;.i_:.:....;....;.;,.;.;:i;.;;:.;;~.;;;1 2

0 .l .2 .3 .4 .5 .6 .1 .8 .9 1.0 I 2

Bottom Z/L (Non-Dimensional Axial Distance) Top "Tl I

-....J (Including Unheated Length) -..J I

co FIGURE 2.2 AXIAL POWER PROFILE FOR 102% OF STRETCH POWER OPERATION

2.0 N

LL s....

/ 0

+'

u LL'° O')

c:

r-

~

0..

'° aJ I

s.... .......

aJ +:>

3: I 0

0..

'°x

r-c::(

.5

.2 .3 .4 .5 .6 .7 .8 .9 1.0 Top z Z/L (Non-Dimensional Axial Distance) I (Including. Unheated Length) ,.,

z I

-....J

-....J I

FIGURE 2. 3 AXIAL POWER PROFILE FOR 52% OF co STRETCH POWER OPERATION

XN-NF-77-13 I 0 u..

~

I QJ

i

..µ

'°!....

QJ I CL E

QJ I--

..µ I

QJ c::

....... 550 QJ I

0 u

I I

I I % Maximum Bundle Power I FIGURE 2.4 PALISADES THERMAL MARGIN LIMITING OPERATING CONDITIONS (BUNDLE POWER, INLET TEMPERATURE, "

I AND PRIMARY PRESSURES) 100% POWER = 2530 Mwt I

100 80

...c

+-'

s...

0

~

µu 60

!tl QJ er: __,I O"I I

+-'

c:

QJ u

s...

QJ 0...

20

z I

z "Tl I

20 40 60 .........

I Percent Control Rod Insertion co FIGURE 2.5 PALISADES SCRAM CURVE

I XN-NF-77-18 I I 3.0 TRANSIENT ANALYSES I 3. l UNCONTROLLED ROD WITHDRAWAL

.. ~-----------------

The withdrawal of more than one control rod due to an operator I error or a reactor regulating system or rod drive control system malfunction I causes an increase in both core power level and core heat flux. Since the heat extraction from the steam generators remains relatively constant until I actuation of the steamline safety valves, an increase in primary coolant tern-perature results. Unless terminated by either manual or automatic action, I the control rod withdrawal would eventually result in a minimum DNB I ratio of less than 1.30. The reactor protection system is designed to terminate such transients before an MDNBR of l. 30 is reached. Protection I against DNB in this event is provided by the high neutron flux, thermal margin/low pressure, and high pressurizer pressure reactor trips.

drawal of a single control rod is discussed in Section 3.9 .

The with-

.1 In order to examine the adequacy of the protection* system, the following incidents were analyzed:

I (1) Control rod withdrawal as a function of withdrawal rate from full power (102%), and, I (2) Control rod withdrawal as a function of withdrawal rate from I 52% rated power conditions.

Control rod withdrawals from 75% of rated power were not evaluated in this I study because this case is bounded by the 102% and 52% cases.

I I

I

I XN-NF-77-18 I

3.1.l Control Rod Withdrawals From 102% of Rated Power I

The uncontrolled rod withdrawal transients from 102% of I rated power were analyzed with rod withdrawal rates up to an equivalent of 30 x 10 -5 Ap/sec. To envelope the operation during a cycle, the analyses I was made assuming both:

1) Maximum reactivity feedback, assuming the largest I

negative values for Doppler and moderator coefficients -

5 I

corresponding to EOC condition,_ i.e., -1.66 x 10- Ap/°F and -3.5 x 10- 4 Ap/°F respectively, and I

2) minimum reactivity feedback, assuming the least negative values of Doppler and moderator coefficients correspond-I ing to BOC conditions, i.e., -.87 x 10- 5 Ap/°F and 0.5 I x 10- 4 Ap/°F respectively.

The effect of the pressurizer spray was included in the I analysis in order to minimize the pressure rise and therefore to minimize the MDNBR.

I.

The result of the analysis from 102% of rated power is shown I graphically in Figure 3.1 (located at the end of Section j.1), which plots minimum transient DNB ratio versus reactivity insertion rate for both maxinium I and minimum feedback. The protection system function which terminated the transient varies with reactivity insertion rate as indicated in Figure 3.1.

I For all reactivity insertion rates, MDNBR is greater than l .30. Sample system I transient results are shown in Figures 3.3 through 3.7 for a reactivity in-5 sertion rate of 14 x 10- Ap/sec with maximum feedback, and in Figures 3.8 I through .3.12 for a reactivity insertion rate of l x l0- 5 Ap/sec with minimum feedback.

I I

XN-NF-77-18 3.1.2 Control Rod Withdrawals From 52% of Rated Power

. -5 Control rod withdrawal rates up to 60 x 10 ~p/sec were analyzed assuming the following axial and radial peaking factors:

I FA= 1.75 and I FR= 1.45 (1.0 + 0.5 [1-P]),

where P = Power level in fraction of rated.

I The above relationship between radial peaking factor and power level conservatively represents the increase in radial peaking that I occurs at reduced power levels, and corresponds to the greater control rod I insertion which is allowed at reduced power.

A maximum inserted rod worth of 1.5% ~P was assumed which I is conservatively high considering the control rod insertion limits. The analysis was made for both maximum feedback and minimum feedback. As in the I full load case, the pressurizer sprays were assumed operable to conservatively

I reduce the margin to DNB.

Results for this case are shown in Figure 3.2 which plots I minimum transient DNB ratio versus reactivity insertion rate for both maximum and minimum feedback. The protection system functions which terminate the I transient are also indicated in Figure 3.2. For the maximum feedback (EOC)

I there is insufficient inserted rod worth to cause reactor trip, and the~efore the transient is effectively terminated when all control rods are fully with-I drawn. For all reactivity insertion rates, MDNBR is greater than 1.30.

Sample system transient results are shown in Figures 3.13 through 3.17

'I I

I

1-*

I XN-NF-77-18 I

for a reactivity insertion rate of 60 x 10- 5 ~p/sec with minimum feedback, I

and *in Figures 3.18 through 3.22 for a reactivHy insertion rate of 5

I 6 x 10- ~p/sec with maximum feedback.

I I

I I.*

I I I:

l1 I

I I

I

Minimum Feedback

- - - Maximum Feedback 2.0 1.8 0

.,.... Thermal Margin Trip

.µ ttl

------- - -"\\- - - - --

~

ca z 1.6 .

Cl .

E

I E

.,.... --- -- N I

I c:

E 1.4

=-----

High Neutron Flux Trip

1. 2

1. 0 . 5 x 2

1 x rn- 1 x 10- 4 2 I

"Tl Reactivity Addition Rate, ~p/sec I

-....J

-....J I

co FIGURE 3. 1 CONTROL ROD WITHDRAWAL INCIDENT REACTIVITY ADDITION RATE VS MDNBR INITIAL POWER LEVEL = 2580.6 Mwt

2.2 r-----------..-------.-----,----,---:r---,.--,--,--..------------,.------~---.----.----

Minimum Feedback

- - - Maximum Feedback 2.0 No Reactor Trip l.8 /

0 /

.,..... /

.;....> /

tO ;'

er. /

c::::

z: 1.6 '-.. /Thermal Margin Trip "

I N

N I

.,.... '-.. ',..._ ,../"/'\High Neutron Flux c:

.,..... ' ........... / / Trip

~ 1.4

~

/,,,.\- ~""-

High Pressurizer l .2 Pressure Trip

l. 0 5 -4

z:

l x 10- l x lO I

z:

"Tl Reactivity Addition Rate, 60/sec I

-....J

-....J I .

FIGURE 3.~ CONTROL ROD \flTHDRA\~AL INCIDENT REACTIVITY ADDITION RATE VS MDNBR INITIAL POWER LEVEL = 1315.6 Mwt

PALISADES ++ CONTROL ROD ~ITHORAWRL FROM 2580 6 MWT *+ O 00014 le/sec 14 0 l POWER l EVEL - I 2 HEAT Fl ux I 3 TOTAL bRfHARY COOLANT FLOW 4 TOTAL . EEOWATER LOW ,

120r--~~-t~~~-r~~~+-~~_.:.5i--_..:...TO~T~A~L~~~T~EA~H.:__.=.L~IN~E~FL~O~W~~\--~~--i~~~-L-~~--l 1 1 1 1 1 1 - I U:1--2-'""~*~.;~~0J::'[=;=,4;:;;:5+=11'=~.4=js~K~:~~4~sf=J~C~~q:=-J5f-~£-~----4~5j-~~~~-4-:5r--~~-_~4-s;--::......-;~:\~~~~ ~

100 1

0 w

f-a:

o::g 0

\ \J LL 0 I N

f- w z I t380 0::

w a..

40 20

z I

z O 0'--~---;2¢--~~~5L-~--;?-;f--~~~1~0~~-1~2:--~__Jl~5~~-1~7~~__j2L0~~-2~2~~_J25 "Tl I

-....I

-....I TIME. SEC I 00 FIGURE 3.3 POWER, HEAT FLUX, AND SYSTEM FLOWS, SEQ. 1228 23 .JUN 77 05 *.13 *03 CONTROL -ROD \.IITHDRAWAL FROM l 02% POl~ER

@ l. 4 X l o-4 LIP/SEC.

,PAi_ I SP.DES ++ CONTROL ROD WI THDRANAL FROM 2 5 8 0. 6 MWT ++ 0. 0 0 0 H Lr;/sec 180 rj

~

1.i 11VE. l="iUEL TEMPER! ~TURE i CORE I!ULET TEMPEi ~ATURE i I "

~-

3. 11\'E. C DRE COOi...AN TEMP.

1 I

I ~- Ci...RO Tl -MPERATURE 180 u

~

1 i.

1 1 I\

l 1 l 1 l

140 '-

I II II

\

I~

120 ~

LL !

I I (9

w I N

~

I 0

~

100 u I

I 800 l I

iI I

i 'i :j :j :j :j :j '.j :j :j J :j E; 0 0! ' ;J .j .< ~

,. ;J

.:. t. ('.* ::'. t. L <'. - '- :<

z:

I I '

- z

"'Tl 5 ? 10 ~2 15 1? 20 22 25 I

--..J

--..J TIME. SEC I Frs~~r<E 3. 4 CORE TEi-1PER'HURE RESPONSES, SEQ. 1229 23 JUN F 1)5 :33 *O'l COl'lTROL ROD \*IITHDRA~JAL FROM l 02', PO~*!ER

@ 1.4 x io-4 :C.c.' SEC.

PAL I SADES ** CONTROL ROD hi I THDRAl-JAL FROM 2 5 8 0. 6. Mh!T ** O. OOO14 :_:/sec 16 lN AVE. PRlMAR'f t.:U DLANl 11:.Mt-' LlJOP

~j

'-" 1nr, ..... 1 CHANG !N AVE. l~RIMARY CO DLANT TEMP LOOP 2 3 CHANG !N HOT L G - COLD EG TEMP. HFFERENCE LOOP 1 4 CHANG! fN HOT L G -- COLD EG TEMP. 'D [ FFERENCE LOOP 2 12

/

8

~

v

~ ~

u...;-

LL ~

l 2 - - -~

I (9

3 lj 3 q 3 q N w l 2 3 ti 3 ti CJ1 D 1 2 ~ ..

1 2

~ ,, 3 q 3 q ['\. I 0 -

- \

-8

\

z I

z

"'Tl

-1~ 2 5 ? 10 12 15 17 20 22 25 I

-...J

-...J TIME. SEC I FIGURE 3. 5 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. 1229 23 JUN 77 O? 133 105 CONTROL ROD WITHDRAl>JAL FROM l 02% POWER

@ 1.4 X 10- 4 l:!.P/ SEC.

PALISADES +* CONTROL ROD WITHDRAWAL FROM 2580. Ei MWT ++ 0. 00014 tip/sec 280 i. STEAM DOME PRESS JRE CHANGE LOOP t 2 STEAM IDOME PRESS JRE CHANGE LOOP 2 3*. PRESSU! PIZER PRES ,URE CHANG!

240 200 '

1 so I

a:

Cf) ct.

120 I N I

CJ'\

I 80 3 __..,,.,

i-v--~ ~;

l.----'

~

40

_Ll-2--- ~

c----

~

~ 1 2 I ><

L-3-- ""-

i. '2 z*

~ 1 2 l I

. - 3 i 2 I i ..,

2 0 0 2 5 10 12 15 1? 20 22 25 I

-...J

-...J TIME. SEC I co FIGURE 3.6 PRESSURE CHANGES IN PRESSURIZER AND SEO. i22923 JUN 77 0~:33:06 STEAM GENERATORS, CONTROL ROD WITHDRAWAL FROM 102% POWER@ l .4 X io-4 6P/SEC.

PAL I SP.DES ++ CONTROL ROD WI THORRWAL FROM 2 5 8 0. 6 MWT ++ 0. 0 0 0 14 :.,J/sec

~o "- CHANGE IN STEAM. *EN. WATER LEVEL. LOI pP. *l l 2 CHANGE !N STEAM ~EN. WATER LEVEL. LOI DP 2 3 CHANGE IN PRESSU OI2ER lo!ATEI ) LEVEL 30

20 3 -

~

10 L--3---

i----

3

-~

I u ~ I

~

N z -....J

-~

i-; I 1 2 1 2 1 2 3 1 2 1 2 1 2 , !:>

0 3

.,,_ 1 ~

t

-10

\

-20

\

-3~ 2 5 10 12 TIME, SEC 15 11 20 22

\ 25 z

z I

"Tl I

-.....i

-....J I

FIGURE 3.7 LEVEL CHANGES IN PRESSURIZER AND SEG. t22923 JUN ?r 05*33*08

. STEAM GENERATORS, CONTROL ~OD WITHDRAioJAL FROM 102% POWER@ 1.4 X 10- tip.fSEC.

PAL I SADES ** CONTROL ROD WITHDRAWAL FROM 2 5 8 0. 8 MWT ** 0. 0 0 0 0 1 :..,c/sec 14 0 1 POWER EVEL I 2 HEAT F ux TOTAL PR£MARY CO DLANT FLOW I 3

~ TOTAL EEDWATER LOl-J 120 i 5 TOTAL t>TEAM UNE FLOl-J I

I t ,., 1 n ~

~

l ') t I) l 2 ., . '.) .,, lj 5 ~ lj 5 ~ lj 5 lj 5 100 ,u I

~

D w

l-a:

cr:so I LL 0

1--

I I

II \ N i

co z \

I wso

,,~

u a:

w Cl..

40 I "\

lI I

I I

I \\

20 i I \ '

l

z I

z I 'Tl I

0 0 10 20 30 40 50 80 70 80 90 100 '-..!

TIME. SEC "' I co FIGURE 3.8 PO\~ER, HEAT FLUX, AND SYSTEM FLO\~S, SEQ. 426 1

23 JUN 7709 46*1J2 CONTROL ROD_~ITHDRAWAL FROM l 02~, POWER

@ l .0 X l0 llc./S EC .

6ALISAOES ++ CONTROL ROD WITHDRAWAL FROM 2580 6 MWT ** o 00001 *0/sec 180 11 l=lVE. l="i JEL. TEMPER!~TURE

~ CORE I ~L.ET TEMPEIRATURE i=lVE. CilRE COOL.AN TEMP.

i CL.AD T MPERRTuRE 1

1 1

12oow~~~t-~~-t~~~-1-~~~+-~~---+~~~~,1--~~-l-~~~-i-~~~L-~~_J I

N

'° I

2 I

2 400,'----~~'~~~~~~~~~~=1::--~---:-1::--~-L~~J__~__L~__J I o 10 20 30 40 50 80 ?O 80 90 100 '-J

'-J I

TIME. SEC CD.

FIGURE 3.9 CORE TEMPERATURE RESPONSES, SEQ. 426 23 JUN i? 09 *'18 *'13 CONTROL ROD S~ITHDRAWAL FROM 102% POV/ER

@ l.O X 10- t:.p/ SEC~.

PALISADES ** CONTROL ROD WITHDRAWAL FROM 2580. 6 MWT ** 0. 00001llo/sec 30 1 CHANG IN AVE. DRIMARY CO DLANT lEHP . LOOP 1 2 CHANGt:. IN AVE. t>RIMARY CO bLANT TEMP . LOOP 2 3 CHANG IN HOT L G - COLO EG TEMP. DIFFERENCE LOOP 1

<I CHANG IN HOT L G - COLO EG TEMP. bIFFERENCE LOOP 2 20 10

,_J~

~

LJ_J---

1_1--- LD---- 3 <I 3 1 2 - ~ 3 <I 3 <I 3 11 0

~

LL I

(!J w w 0 I

0

-10

-20

\

-3 0

\

z I

z.,.,

I 10 20 30 40 50 80 70 80 90 100 -....J

-....J I

TIME. SEC FIGURE 3.10 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. ~26 23 JUN 71' Oil 11j81t1<1 CONTROL ROD WITHDRAWAL FROM 102~ POWER

@ l .0 X 10- 5 ~=/SEC.

PALISADES+.- CONTROL_ ROD WITHDRAWAL FROM 2580. S MWT ++ 0. 00001*:,..:./sec 350 I 1 STE'1M DOME PRESS ~RE CHANGE L..OOP l i :2 STEAM DOME PRESS ~RE CHANGE L..OOP 2 i 3 PRES SU! bfZER PRES >URE CHANGI 300 I

/

I 250 1 I

200 a: I w

(f) --'

Q_ . I

~

150

~

""' ~

100

/v /

50

?/ _,..., v

/ ~

~ ~ ><

z

~ z I

l ~ 11--- ""Tl I

0 0 10 20 30 40 50 60 7' 0 80 90 100 -...J

-...J I

TIME. SEC co FIGURE 3.11 PRESSURE CHANGES IN PRESSURIZER AND SEQ. ~26 23 JUN n 09 :-ts '15 STEAM GENERATORS, CONTROL RgD WITHDRAWAL FROM 102% POWER@ 1.0 X 10- 60/SEC.

PAi_ I SADES +-t CONTROL ROD WITHDRAWAL FROM 2 5 8 0. 8 MWT ** 0. 0 0 0 0 1 !..:;.,/sec 100 1 CHANGE IN STEAM 'EN. WATER LEVEL. LO )p 1 2 CHANGE IN STEAM ,EN. 1-JATER LEVEL. LO )p 2 3 CHANGE IN PRESSUI ~rzER WATEI ~ LEVEL 80 80 en 40 w

v

)

I I u w z

N I

3/

20

~

0 J 2 '2 ~ 1 2 ..

1 n

~ £.

1 2

\

I II '

I

-20 I I

I I

\ ~

z z

I "Tl I

10 20 30 40 - .50 b0 ?O 80 90 100 "'-J

"'-J I

TiME. SEC FIG'Jc;E 3.12 LEVEL CHANGES IN PRESSURIZER AND srn. 426 23 .JUN?? os *48 *H STEAM GENERATORS, CONTROL R~ l~ITHDRA\o/AL FROM l 02:; POWER @ l . 0 X lo- 6 ::/SEC.

PALISADES ++ CONTROL ROD WITHDRAWAL FROM 1315. 6 MWT ++ 0. 00060 ~c/s~c 14 0 1 POWt::~ 1 .EVEL 2 HEAT Fl ux 3 TOTAL I ~RI MARY CO DLANT -FLOW TOTAL EEDWATER LOW A TOTAL ~TEAM LlNE FLOW 120 y \

3 3 - 3 I 3 3 3 100 0

w f-a: I/ /

v-D I'\

o::s 0 l.l..

0 v y v ~ I

~

w f- w z /

I

~60 0::

w

/~ ~ ~

a.... q 5 q 5

\" "

" 5 4 40 '

20

\ ~ .

\ .............

5 r-1-----

--2 Il Ii i

i I' ><

z:

~ I i :z:

"'Tl i I 0 0 2 4 6 8 -10 12 14 16 18 2 C:

-.....J

-.....J I

TIME, SEC FIGURE 3.13 POWER, HEAT FLUX, AND SYSTEM FLOWS, SEQ. 10116 24 JUN 7? 14: 56: 26 CONTROL ROD WITHDRAWAL FROM 52% POWER

@ 6 . o X10-4 6,:, / SEC.

PALISADES ++ CONTROL ROD WITHDRAWAL FROM 1315. 6 MWT ++ 0. 00060 /Jo/sec 180 '-.l 1 AVE. F llEL TEHPERI HURE 2 CORE I bLET TEMPE! ~ATURE 3 AVE. cnRE COOLAN TEMP.

lj CLAD T MPERATURE 160 IJ *-

14 0 u

~

v

/

120 u

~

LL I

~

(!)

w w +:>

0 ~

I 100 U1 -

800

.. <l lj lj '1 600 3 - 3

.., ;j

" ;j " 3

~

c. " 3

z:

I

z:

I 2 8 10 12 14 16 18

~

-....J

-....J I

TIME. SEC FIGURE 3.~4 CORE TEMPERATURE RESPONSES, SEQ. 1046 24 JUN 'ii' 14 '56 *26 CONTROL ROD ~JITHDRA\*JAL FROM 52': POWER

@ 6.0 X 10- 6f:'/SEC.

PALISADES ++ CONTROL ROD WITHDRAWAL FROM 1315. 6 MWT ++ 0. 00060 :...:./sec 12 l CHANG IN AVE. . LOOP l 2 CHANG IN AVE. RlMARY CO LANT TEMP

LOOP 2 CHANG IN HOT L G - COLD EG TEMP. lFFERENCE LOOP 1 CHANG IN HOT L G - COLD EG TEMP. IFFERENCE LOOP 2 10 8

6 LL I

(9 w w c..n I

0 4

2 1 2 0

z:

I

z:

I

-2 2 6 10 12 14 16 18 20

-....J 0 8 -....J I

TIME. SEC co FIGURE 3.15 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. 10 4 6 2 4 JUN ?7 14 : 5 6 : 2 S CONTROL ROD J"JITHDRAWAL FROM 52% POYJER

@ 6. 0 X 10- tip/SEC .

PALISADES ++ CONTROL ROD WITHDRAWAL FROM 13-15. 6 MWT ++ 0. 00060 !:;p/sec 14 0 LOOP I 1

2

3.

STEAM

'OME P:r ~RE CHANGE STEAM* I JOME PRES URE CHANGE PRES SUI LZER PR S' PURE CHANGI LOOP 2 120

/,~

100 80

~/, r er:

CJ)

[L 60 II vi w

I O'\

I I

40

/

vI 20 1 2 3 ~

v lJ 1 2 -

x

z I

2:

"'Tl I

0 0 2 4 6 8 10 12 14 16 18 20 ......a

......a I

TIME. SEC FIGURE 3.16 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 1046 '24 JUN ?7 14 :56 :30 STEAM GE.NERATORS, CONTROL _R~D; WITHDRAWAL FROM 52--c POWER@ 6.0 X 10 '-'p./SEC.

PALISADES *+ CONTROL ROD WITHDRAWAL FROM 1315. 6 MWT ** 0. 00060 t::;/sec 50 1. CHANGE [N STEAM 'EN. WATER LEVEL. LOl~P i 2 CHANGE IN STEAM ~EN. WATER LEVEL. LO~P 2 3 CHANGE lN PRESSU [ZER WATEi LEVEL 40 30 (f)2 0 w

I I u w z........, ........

I 10 1 2 .,

1 2 0

l 2

~__;;:3~--1----~

x

z:

I

z:

-2~

I 2 6 8 10 12 14 16 18 20 -..J I

TIME. SEC OJ FIGURE 3.17 LEVEL CHANGES IN PRESSURIZER AND SEQ. 1046 24 JUN 77' 14 :56131 STEAM GENERATORS, CONTROL ROD WITHDRAWAL FROM 52% POWER @ 6. 0 X1o- 4 lip/SEC

PALISADES ++ CONTROL ROD WITHDRAWAL FROM 1315. G MWT ++ 0. 00006 :C.;.,/sec 110 PQl-JER I EVEL I ~j HEAT F~UX I

3 TOTAL ~RIMAR'\' CO )LANT FLOl-J

it TOTAL 1*EEOWATER fL.,OW 3 3 3 3 5 TOTAL ~TEAHJ_INE FLOW~ :

100 - I t

I I

NOTE: Dscillati bns due tp

,afety va lve hyste res is I!

90 I [

D w

l-a: Ii tl'.:8 0 --

51 I r ~, \

~

LL n 0

5 I

~ ~ ~

I I w I

w 00 1-z I ' I

~70 I ~~

Q'.:

w a...

I

'I : I ~ r:I I I

I ~~'! ~ '

I GO 1 2 cI I i 11 I

I 1 2 1 2

.q I

I i

I I 50 4 iq*

'.* .. I 4 ~

\( ~

a '5 - 5 .

x I I :z I

I I I '

z I

.+ 0 80 120 160 200 2i+O 280 320 360 400 '-l

'-l I

TIME. SEC- ... **

FIGURE 3.18 POvJER, HEAT FLUX, AND SYSTEM FLOHS, SEQ. 114823 JUN 7i'04*l?*ti8 CONTROL ROD \.IITHDRA\*/AL FROM 52'i POWER

@ 6.0 X 10- 5 ~p/SEC.

- - - - - - - - -* - - - - - - - - - - \

/

\

PALISADES

++ CONTROL* ROD WITHDRAWAL FROM 1315 8 MWT ++ 0 00008 ~c/sec 180I~ AVE. Fl~EL TEMPER HURE

~j CORE I ~LET TEMPE )ATURE !

I 3~ . AVE. CVRE COOLAN TEMP.

~ CLAD T MPERATURE I i80 ~

I NOTE: Oscilla ~ions due to ,

safety ~alve hys lteresis 1~0 0 i

I 120 -v 1 1

~

LL

~ I

~

CD w w I ,,..

\.0 I

0 1 l 1 L...

100 B--

800 I

i I I

- lj

'I 800 i

~ j

" ll

"' *3 tj

" s

'I

"' ., n -

~

"' 3

[ x

z

- I

z:

"'Tl I

lf 0 80 120 180 200 240 280 320 380 -...J

"-J TIME. SEC ------ I FIGURE 3.19 CORE TEMPERATURE RESPONSES, SEQ. 1148 23 JUN 77 04: 11' *~9 CONTROL ROD ) /ITHDR.L\\.JAL FROM 52% POlffR 1

@ 6.0X io-::> l'lof-iEC.

PALISADES ++ CONTROL ROD WITHDRAWAL FROM 1315. 8 MWT +* 0. 00008 !.-;/sec 60 *, t. CHAN GI !N AVE. I~R IMHRY CUI 'LANT TEMPt. COOP :

2 CHANGI IN AVE. !~RI MARY COi bLANT TEMP . LOOP 2 3 CHANGI IN HOT L G -- COLD -EG TEMP. IFFERENCE LOOP t

'I CHANGI IN HOT L G - COLD EG TEMP. IFFERENCE LOOP 2 50 I I I

I I NOTE: Osei 11 at ions du e to safet_~ valve hj steresis II 40 I

2

)r-r"- v ~

-2v I

,~

30

~*

v Lt.

I

(!) ~

w 0 v

0 I 20

/ ':! 'I I

10 v ~

'rl'-'vvvvv wvv

,~V I

0 v£ 3 ~

I 3 'I I

- I I

I ! I I ><

- I :z I

I I I I .,

z

-1 ~ 40 80 120 160 200 240 . 2 80 320 380 400 I

'-l

'-l I

TIME. SEC FIGURE 3. 20 PR I MARY LOOP TEMPERATURE CHANGES, SEQ. tH s 23 JUN 7? o:+ , t 7, 50 CONTROL ROD WITHDRAWAL FROM 525~ POWER

@ 6.0 X 10- 5 6~/SEC.

PALISADES ++ CONTROL ROD WITHDRAWAL FROM 13 15. 8 MWT ** 0. 0 0 0 0 8 L.,/sec 280.-~~-.~~~~~~--,,--~~~1~.1~=sr=E=AH,.,-;o:=oM=E=-=pR=E=ss"""""R=E~C=H=AN~G=E~L=o=op~1-,-~~~-r-~~--,

3 2~ STEAM OHE PRESS RE CHANGE LOOP 2

~ PRESSU IZER PRES VRE CHANG I

I 2~0~~~-t-~~~+--~~-+-~~~+-~~-+~~~+-~~--+~~~-+-~~~~~~~

I NOTE: Oscilla ons due ~o safety v lve hys~resis a:

1-1 I

+::>

Cf)

I D-.

1201--~~-+-~-t-~-+--r-~~1--~~-+-~~~--~~~1--~~-+-~~~--~~--1f--~~----;

80 40

z I

0 0 40 80 120 180 200 240 280 320 380 400 -.....J

-...J I

TIME. SEC FIGURE 3.21 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 1 t48 23 JUN n 04: ti' *51

. STEAM GENERATORS, CONTROL ~OD WITHDRAWAL FROM 52% POWER @ 6. 0 X 10- ~p/SEC.

PAL I SADES ** CONTROL ROD WITHDRAWAL FROM 13 15. G MWT * + 0. 0 0 0 0 G ;:., 0 /sec 120 1~~~--,-~~~-,~~~,-~~--,--r--r;;cH1rAwN~GE=-TOiuN'S~TMEA~M;--:;:;;EMN~.-w~A~T=E~R~L~E=vE~L-.~L~O==-P~l~~~~~-----.

2. CHANGE IN STEAM EN. ~ATER LEVEL. LO P 2

3. CHANGE fN PRESSU !ZER WATE LEVEL NOTE: Oscillations due o safety v lve hyst resis 80 cnSO w

I I u .p.

z N I

40 20 0

z I

I .,,

z

- 2 ~'---~---:4~0:--~~87 0~~-1~2~0=--~~1~8~0~-.~2~0~0:--~~2L4~0~----,2~8-0~~-3L2_0~~3~G-0~___14QQ I

-...,J

-...,J I

TIME. SEC co FIGURE 3.22 LEVEL CHANGES IN PRESSURIZER AND srn. 11'!823 JUN 770~'1?'*53 STEAM GENERATORS, CONTROL ~D WITHDRAWAL FROM 52% POWER @ 6. 0 X 10- tp/SEC.

I I XN-NF-77-18 I

1.? CONTIWL ROD DIWP I NC.....J DENT I

~ -~ *--*--* **- -***--~ ~.-~ ... --*

The control rod drop incident is defined as the inadvertant re-I lease of a single control rod causing it to drop into the reactor core.

A control rod drop results in a rapid decrease in reactor power, which I could be followed by a return to power with a distorted power distribution as a result of the non-symmetric rod pattern. Dropped rods may be detect-I ed by either of the following two independent means:

I (1) a limit switch on each individual rod which indicates a fully inserted rod.

I (2) a high negative rate of change of neutron flux signal which is generated by the out-of-core power range neutron I flux channels. "-

I It was assumed that the turbine steam demand did not vary in this analysis in order to improve the operating margin. The reactor regulating system I was assumed to be in the manual mode, therefore conservatively inhibiting

.1 automatic rod insertion during the transient.

system is in the automatic mode, rod withdrawal is also prohibited by the When the reactor regulating I rod drop protection system. The pressurizer spray was assumed operable in order to minimize any pressure rise and therefore minimize DNBR. The analy-I sis was performed for the maximum and minimum expected dropped rod worths at both beginning and end of cycle conditions. These worths. and the resulting I radial peaking factors are given below.

I Rod Group A

Rod Worth minimum, -0.04 ~P BOC FR EOC F R

l. 60 l. 60 I 4 maximum, -0.12 ~P l. 66 l. 64

'I

I XN-NF-.77-18 I

I The-results of the analysis are summarized below.

Dropped Rod I Worth Reactivity Applied Case ( Llp)

-:0.04 Coefficient BOC F

---'---R-l.60 MDNBR I

l.42 2 -0. 12 BOC l.66 l. 35 I 3 -0.04 EOC 1. 60 1.42 4 -0. 12 EOC 1.64 1.40 I

at BOC).

The lowest MDNBR occured for Case 2 (maximum dropped rod worth Although the minimum dropped rod worth for cases 1 and 3 showed I

a higher return to power after rod drop than did the maximum dropped rod I for cases 2 and 4, the lower radial peaking associated with the lower dropped rod worth resulted in a higher MDNBR. I The system responses for the case yielding the lowest MDNBR are shown in Figures 3.23:through 3.28.

I I

I

PALISADES ++ CONTROL ROD DROP 14 0 1 POWER EVEL 2 HEAT F ux 3 TOTAL bRIMARY CO DLANT FLOl-J

'I TOTAL EEOl-IATER LOW 5 TOTAL STEAM UNE FLOW 12 0 3

100

"- ., '! 5 3 'I ~ 3 " t; 3 .. <:; 3 c: 3 - 3 3 3

\'---

~

D rr-2

2 w

2 l-a:

- -

2 2

~80 2

=:: ~ ""-

~

LL 0 I

.p.

~.

U1 1- I z

wso u ~

\

()'.'.:

w CL 40 20 I\ 1~ 11-t-__

~

z I

i-=-___ z

""Tl I

f'--.-.___. 5 5 "

-....J I

0 0 10 20 30 40 50 60 70 80 90 100 TIME. SEC FIGURE 3.23 POWER, HEAT FLUX, AND SYSTEM FLOWS SEQ. 3 10 0 1 JUL 7'i' 10

5 6

0 5 .

CONTROL ROD DROP

PALISADES ** CONTROL ROD DROP 180 J 1 AVE. r:-* JEL TEMPER HURE 2 CORE r ~LET TEMPE 1 ATURE 3 AVE. C)RE CODLAN TEMP.

.j CLAD T MPERATURE 160 u I~

14 0 u 12 0 IJ

' ,..!__ 1 1

1 r--r 1

-- -1 I\

LL I

. CD -+::>

w O"l I

0 ,_

100 u 800

\ I\

600 3

.j 3

Lj 3

4 3

4 I

3 LI Lj

~ r--l-___

~ L £. £. 3 3

- .., 4 ') :< 4 z

" "' ' ' z I

-n I

-....J I I -....J I

10 20 .3 0 40 .tJO 60 70 80 90 100 TIME, SEC FIGURE 3.24 CORE TEMPERATURE RESPONSES, SEQ. :310 01 .JUL ?'? 10 :56 :06 CONTROL ROD DROP

~-------~------

PALISADES *~ CONT-ROL ROD DROP 10 1 CHANG IN AVE. ~R IHARY CO PLANT TEMP . LOOP 1 2 CHANG IN AVE. bR !MARY CO PLANT TEMP . LOOP 2 3 CHANG IN HOT L G - CQLD EG TEMP. DIFFERENCE LOOP 1 CHANG IN HOT L G - COLD EG TEMP. DIFFERENCE LOOP 2 0

1 2 "

~~ 3 :i 3 q 3 "

3 4 3 " . 3 4

~

-10

~ ~

~ \

-20 ~

LL CD

'~ I

.+:::>

w -.....i 0

-30

~ \ I

~

-40

~-) \

l 2 1 2

-50

' '-3"

z

z I

"'Tl I

-.....i

-.....i

-6~ 90 100 I

10 20 30 40 50 60 70 80

TIME, SEC FIGURE 3.25 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. 3l0 0 1 JUL 77 10 : 5 6 : 0 7 CONTROL ROD DROP

PALISADES ** CONTROL ROD DROP 400 1. STEAM JOME PRESS JRE CHANGE LOOP 1

2. STEAM JOME PRESSl~RE CHANGE LOOP 2

3. PRESSU' >I ZER PRES PURE CHANG1 300 200 a:

....... ~

I co (f) I 0...

0

z I

z

"'Tl I

3 '-J

'-J

-3~'G 70 I

10 2-0 .3 0 '+ 0 .t5 0 60 80 90 100 co TIME, SEC FIGURE 3.26 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 110 0 1 JUL 77 10 : 5 6 : 0 8 STEAM GENERATORS, CONTROL ROD DROP

---~-~~--~-~-------

PALISADES *~ CONTROL ROD DROP 20 1. CHANGE [N STEAM IbEN. i.JATER LEVEL. LO OP l 2.

3.

CHANGE IN STEAM ~EN. i-JAT"" 0 CC:-f LO CHANGE LN po--- ~iTER 1-JRTE g L VEL

~p 2 i;

I I i 2 I 1 2 l 2 1 2 I 0

~

i

\

I I

~ I I

I

-20 1

~

t8I

-40 ""3 ~

\ \

I i

II I

~ ~

I u ~

z j l.O I

~ *) _j

-60 1

'~-

II I

-80 I

~

Il I

I

~

I

-10 u

~ - 3 3 ~\ ><

z I

~

z I

~ -....J

-....J 1~

--- I u

10 20 30 4_ 0 50 60 70 80 90 1 CJ 0 TIME. SEC FIGURE 3.27 LEVEL CHANGES IN PRESSURIZER AND STEAM SEQ. 310 01 JUL 77 10'~6:10 GENERATOR, CONTROL ROD DROP

PALISADES ** CONTROL ROD DROP 3 1 MO DERR OR REACTI 1 ITY 2 DOPP LE R REACTIVI y 3 BORON ~EACTtVITY

. q TOTAL ~EACT IV I TY 2

2

~

2 2 2/

2 2 2 2

0 1

3-,_; lJ. ! .:J q 1 3 q 1 3 q 3 q

. 3 4 1

3 "

3 1

3 (f) 1 '

I 0:: c..n er: 0 I

_J

_J o-1 d

-2

-3 ><

~

4

" z z

I lY I

-...J

-...J I

-~

o 10 20 30 40 50 so 70 80 90 100 co TIME, SEC FIGURE 3.28 REACTIVITY FEEDBACK, SEQ. 310 01 JUL ?7 10*56:11 CONTROL ROD DROP

!1 I XN-NF-77-18 I

I 3.3 LOSS OF COOLANT FLOW INCIDENTS A loss of primary coolant flow may result from a loss of electrical I power to the primary coolant pumps or from a mechanical failure, such as primary pump shaft seizure. The loss of primary coolant results in a rapid I. increase in coolant temperature, which combined with reduced flow, reduces the margin to DNB.

The loss of coolant flow has been analyzed for the two most severe cases:

(a) Four-Pump Coastdown; loss of power to all four primary I coolant pumps from operation at 102% of rated power (2530 Mwt) without turbine generator assist.

I (b) Seized rotor of one primary pump, under full load 1: operating conditions.

The analysis of the loss of coolant flow incidents assumed I beginning-of-cycle (BOC) kinetic coefficients with a 0.8 multiplier applied to the BOC Doppler coefficient for conservatism. The combination of the

I positive BOC moderator and the minimized Dapper coefficient 'maximizes the I core power and hot rod heat flux during these transients. The following control systems were assumed active or inactive to decrease the transient II I

pressure and thus minimize the predicted margin to DNB:

Pressurizer spray - active

Pressurizer heaters - inactive

Steam dump - inactive

I XN-NF-77-18 I

I 3.3.l Four-Pump Coastdown The individual primary loop flow rates for the four-pump

/

I coastdown inci9ent were determined from RELAP4-EM(?) calculations of an electrical failure at 102% of stretch power. Figure 3.29 shows the results I

of the calculations as fractional core flow vs time. The RELAP4-EM pump I

model describes the interaction of the centrifugal pumps and the coolant fluid using the four-quadrant homologous curves which are constructed I empiri£ally by the NSSS supplier.( 8 ) From these curves, along with the pump rotor moment of inertia and the primary loop hydraulic characteristics, I

values of head and torque are uniquely defined by the volumetric flow and I

pump speed. Also shown in Figure 3.29 are the Palisades reactor test results of an unassisted primary coolant flow coastdown from hot zero I power. The small difference between this curve and the RELAP4-EM curve at 2580.6 Mwt reflects the small change in primary loop pressure drop at I

the two power levels. Since the primary pressure and temperature remain essentially constant during the four-pump coastdown incident, no void I

formation will occur that would alter the applicability of these results. I The transient responses for the coastdown incident are shown in Figures 3.30 through 3.34.

four~pump The loss of power to I all pumps results in a reactor trip on a low flow signal.

trip occurs at 1 .58 seconds.

The reactor System pressure peaks at 2073 psi a. An I

MDNBR of 1.39 is reached at 3.1 seconds after initiation of the incident. I I

I

.1

I I XN-NF-77-18 I

3.3.2 Locked Rotor I For the case of the locked rotor instantaneous seizure I of the rotor is assumed. Credit is not taken for any additional flow re-sistance through the pump as a result of seizure and flow reversal. The I transient responses for the locked rotor incident are shown in Figures 3.35 through 3.39. The reactor is tripped at 0.9 seconds by a low flow I signal. Core average temperature increased to 579.2°F with a peak primary I system pressure of 2080 psia, well below the relief valve setting of 2400 psia. The MDNBR reached* during the transient is l .27 at 2.4 seconds.

I MDNBR's of less than 1.30 are acceptable for this class of transient (IV) because of the transient's low probability of occurrence.

I The flows of the four primary coolant cold legs are no longer equal under three pump operation since reverse flow occurs in the 1.*

cold leg of the locked pump during the locked rotor incident. The PTSPWR2 I code assumes that the flows of the four cold legs are uniformly mixed in the lower plenum. This is a reasonable approximation under four pump I operation when the temperatures,of the cold legs are nearly equal. The core average temperature,on which the point kinetic calculations are based, I is then a g6od representati~~ of the uniform core cross-sPctinnal tem-I perature distribution. A study conducted at Battelle Columbus on the Palisades core show that there is only a small change in the core flow

1 distribution when operation goes from four to three pumps. In addition, the study concludes that the core flow is well mixed in the upper plenum, I

I I

I XN-NF-77-18 I

I but that the cold leg flows are not uniformly mixed in the lower plenum under three pump operation. However, the potential for non-uniform tem- 'I perature distribution in the core during the locked rotor incident has no effect on the applicability of the point kinetics model since at the I

time of MDNBR (2.4 seconds) the maximum calculated difference between the I temperatures of the three operating cold legs is only 3°F. The resulting non-uniform temperature distribution in the core is negligible. Twenty I (20) seconds after the incident, the temperatures of all cold legs are r.*-

nearly equal as is shown in Figure 3.37.

1 I!

I

1 I

I I

I

- --- -'. - - ... -I* '-- .. - - -: --. - - -) -) - ~

100 90 80

~70 Palisades Test Results at Hot Zero Power-3:

~

LL..

60 I

1.1..

50 CJ1 CJ1 I

4-0

-+-'

40 QJ

~

QJ 30 RELAP4-EM 102% of Rated c..

Power (2530 Mwt) 20 l ()

0 5 l0 15 20 25 z I

Time From Pump Trip (Sec) z "Tl I

-...J

-...J I

co FIGURE 3.29 PALISADES PRIMARY COOLANT FLOW COASTDOWN COMPARISON OF PLANT TEST RESULTS AT HOT ZERO POWER WITH RELAP4-EM PREDICTION AT 102% OF RATED POWER

PALISADES ** 4 PUMP TRIP FROM 2580. 6 MW **

DOWER EVEL ~

2 ~EAT F UX 3 TOTAL RIMARY CO LANT FLOW

'I TOTAL EEDWATER LOW 5 TOTAL TERM LINE. FLOW 120 100 0

w I-a:

0:::8 0 LL I 0 U1 O"l I

I-z wso U*

er:

w (l....

40 3 5 20 I i i 1

.,:z><

~-+-~-+-~~~~+--+-~~~~~~l

z "Tl I

-,,.I 0

0 2 4 s 8 10 12 14 (s* 18 20 :-..J I

TIME, SEC FIGURE 3.30 POWER, H~AT FLUX, AND SYSTEM FLOWS, SEQ. i 8 8 8 15 .JUN ?? 19 ' 2 'I ' 0 :J 4 PUMP TRIP

-~

.----------~~-- ~~--;-*--*-*--:- ~-[

PALISADES ++ 4 PUMP TRIP ~ROM

' . 2580 G MW +*

18 OIJ I 1 AVE. F C.EL TEMPERrURE I

2 3

CORE AVE.

I ~LET TEMPE ATURE C~RE COOLAN TEMP.

! <t CLAD T MPERATUREI I

I 1G Ou i l 1

~.

14 0 u-

'~ -

120 v-

~~

LL I

(.Tl

(.'J "'-J w I I

0 ,_

100 v

~

I ~r---_

1 800

r--_i-L---

" " " " q tl 800 *' *' .:l 3 3 3 - 3 ., -,;; :;

~

-z 2 2 2 <;

- ;j G

~

z I

z "Tl I

40~ 2 8 10 12 14 16 18 20

"'-J

. "'-J I

TIME. SEC FIGURE 3.31 CORE TEMPERATURE RESPONSE; SEQ. 18 8 8 15 JUN 7r 19 : 2 'I : 0 5 4 PUMP TRIP

PALISADES ** 4 PUMP TRIP FROM 2580. 6 MW **

12 .~~~-.--~~~-.,----~~~-r--~~--,-,,.--~C~H~AN~G;r-~IN..--r;;AV~E~.----.;-;R~luMANR~Y~CO'i"l)!LnoAN~T~TE~~~lP~,~L~OONP""'""*.-~~---,

2 CHANG IN AVE. RI MARY CO LANT TEMP . LOOP 2 3 CHANG IN HOT L G - COLO EG TEMP. IFFERENCE LOOP 1

~ CHANG. IN HOT L G - COLD EG TEMP. IFFERENCE L_OQP 2 8

2 4

0 LL I v.,

(!) <X>

w I 0

-4

-8

-12

z:

I"

z:

I 2 8 10 12 16 18 20 -....J

-....J I

TIME. SEC co FIGURE 3. 32 PRIMARY LOOP' rEMPERATtJ:RE CHAN.GES, SEQ. 1888 15 JLJN 77 19 '2~ :Q6 4 PUMP' TR:IP*

I I

- -} -: -/ - - ... -, .. - (-\ - -1* ,-,' -: - -.; - -

PALISADES ** 4 PUMP TRIP FROM 2580. 6 MW **

240-~~~,-~~-,-~~~.--~~-,-~~-=--~------=-=~~~~~~~~~~~~~~

LOOP l LOOP 120 I

a:

(.]1

\.0 I

Cf)

CL 80 3

40 0

z I

z

""Tl

-4 0'--------::--'---_i_---=-1-----_i____-__J_ __ _ L_ __l___ __J___ _. : _ __ _ I I

-....J o 2 4 G 8 10 12 14 113 18 20 -....J I

TIME, SEC co FIGURE 3.33 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 1888 15 JUN 7? l.'3 :24 :Q8 STEAM GENERATORS, 4 PUMP TRIP J

PAL I SADES ** 4 PUMP TR IP FROM 2 5 8 0. 6 MW **

20 1. CHANGE IN STEAM ~EN. i.JATER LEVEL. LOTP 1

2. CHANGE IN STEAM I~EN. WATER LEVEL. LOr 2

3. CHANGE IN PRESSU ,IZER 1-JATEI R LEVEL I ;., *n

.. 10 I

- ~

~

r--z___

1 2 .. / r--z___ .......... ;,

0 ~ ~

~

cn-10

\ I w

\j r I O'l I

u II 0

I.

z

!--; I I\

-20 t

I I

-30 I I ~ ... . -- -** - .. - ..

i I

, 2 t 2 .1 2 .... ..

""' ~**

I j'

i .. .2~.

- ...1-.. I

-4Q i

~

. ~ II I

!I ! I ! >::

i I I

z I

II z I

-5~

I I ! I I .. .. I "Tl I

--.J 2 G 8 10 12 14 lG 18 20 ".J I

T lME. SEC FIGURE 3.34 LEVEL CHANGES IN PRESSURIZER AND STEAM SEQ. 18 8 8 15 J JN ?? 1.9 '2 4

0 9 GE.NERATO.RS; 4 PU~1P fRIP

PAi_ISADES ++ LOCKED ROTOR FROM 2580. S MW *+

140,--~~---r-~~~,-~~~,-~~-;-,-~~PQ~w~'E~R.-;:::E,~VE~L~~,.-----~~---r-~~~-,--~~~~~~-

I ~j ~EAT F ux I 3 I

TOTAL lR I MARY COOLANT FLOW I ~ TOTAL EED'-'RTER I LOW I 120i--;~~--t~~~-t-~~~t--~~~5+-_:_TO~!~A~L-f-'..T~EA~M~L~l~N~E~F~L~OW::_~-l--~~--l-~~~_j__~~--l

~ I

. I O'l 40 5 5 20 z

I z

"'Tl I

-....J 0 0 2 8 10 12 14 18 18 20 -....J I

TIME, SEC co FIGURE 3.35 POWER, HEAT FLUX, AND SYSTEM FLOWS, SEQ. 1882 15 JUN Ti' 19

18

39 LOC KEO ROTOR

PALISADES ++ LOCKED ROTOR FROM 2580. 6 MW **

18 0 - l AVE. F :1EL TEMPER ATURE I 2 3 .

CORE AVE.

I ~LET TEMPE ~ATURE C lRE COOLAN TEMP.

'I CLAD T MPERATURE I

180 u 1 1

~

14 0 u

~

120 u

~ ~

~

LL I 7l N

(!) I w

D :_ ,1 100

~

rI ~~

800 I

- ~~ -.!__

Ir ~ 'I

'I II lj 600 r ,j ., ~ 3 II

., ~

~

I

2 2 <

-c. 3' " 5 - - .. 'f

! i ><

I :z I

z I I I I 2 8 10 12 14 1 () 18 20

-..._J

-..._J I

TIME. SEC co SEQ. lee 2 15 .J JN ?i' 19 : 18 '4 1 FIGURE 3.36 CORE TEMPERATURE RESPONSES, LOCKED ROTOR

PALISADES ** LOCKED ROTOR FROM 2580. 6 MW ++

30 2

3 LOOP 1

~ LOOP 2 20 2 2 10 1

1 0

LL I CJ)

(0 w w I

-1oi--~---jr-~--t~~-f~~-+~~-+-~_2:::.~~~--l--~~-+-~~.J_~__J 0

x

z lj I z

"'T1 I

-4 ~ 2 8 8 10 12 14 18 18 20 -.....!

I TIME, SEC 00 FIGURE 3.37 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. 1862 15 JUN 7? 19 1 t8 *IJ2 LOCKED ROTOR .

., .... **~~---------'------------------------

PALISADES ** LOCKED ROTOR FROM 2580. 6 MW **

300 I 1. STEAM nOME PRESS JRE CHANGE LOOP t i

2 STEAM nOME PRESS JRE CHANGE LOOP 2 I 3 PRESSU HZER PRES >URE CHANGI I

l 2 2 2

.2

,,-:-- 1 2

bP' =--- ci--L__

~ -

v 200

~

150

/

I er: m

-!=:>

I (f)

(L_

10 0 /

50

~

L ~

3 3

r----z_

~

lA ~ '-.......... "'!

.o

~

~ ><

z I

~ .,:z I

-5~ 2 8 10 12 16 18 20 -...J

-...J I

TIME. SEC FIGURE 3.38 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 18 8 2 15 J LJN 7? 19 : 18 : -! 3 STEAM GENERATORS, LOCKED ROTOR

PALISADES ** LOCKED ROTOR FROM 2580. 6 MW +*

80 ! q CHANGE IN STEAM ;EN. ~ATER LEVEL. LOr 1 CHANGE IN STEAM EN. wATER LEVEL. LO P 2 I II 2

31. CHANGE IN PRESSU~IZER WATE ~ LEVEL .

60 I

t+ 0

(.()2 0 w 3 I

I 3 O)'

u --1-_ t1l z........

~

I

~--

1 2 g ~ ~ 'J 0 ~

~

-20

~ ....._ 2

~

1--1-_

~ ----- --t_____ 2

- 2 2 2 2 2

~

J---

-t+O

~ ~

1------

~

-...!._ _LJ__--- ~ z 2

I "Tl I

'.J 2 8 10 12 1 t+ 16 18 20 '-I I

TIME, SEC co FIGURE 3.39 LEVEL CHANGES IN PRESSURIZER AND SEQ. 18 8 2 15 JUN ??' 19 : i 8 ' 4 5 STEAM GENERATORS, LOCKED ROTOR

I XN-NF-77-18 I

3.4 EXCESSIVE FEEDWATER INCIDENTS I

Exce.ssive heat removal by a feedwater system malfunction can I cause the primary system to depressurize and the reactor power to increase due to negative moderator reactivity coefficients. Protection against I these undesirable conditions and high steam generator water levels is provided by high steam generator water-level alarms, and by reactor trips I

due to high neutron flux, low pressurizer pressure, thermal margin/low I pressure, or low steam generator pressure. The incidents which result in the most severe excessive cooling action by the feedwater are analyzed: I Reduction in feedwater enthalpy from 102% of rated power Increased feedwater flow from 52% of rated power I

3.4.1 Reduction in Feedwater Enthalpy I An instantaneous enthalpy drop of 58 Btu/lb is assumed for the decrease in enthalpy incident from 102% of rated power (2530 MWt). I No credit is taken for the reduced rate of cool-down that would be provided by the reduction of extraction steam . . In addition, the pressurizer heaters I

. are assumed inoperable. This allows a more rapid depressurization of the I

primary system, ensuring a conservative prediction of the margin to DNB.

Beginning-of-cycle (BOC) reactivity coefficients were assumed, with the nominal Doppler coefficient being multiplied by 0.8. The initial reactor pressure of 2010 psia, 50 psi below the nominal value of 2060 psia.

I The plant responses to the transient resulting from the I

decrease in feedwater enthalpy are shown in Figures 3.40 through 3.44.

I I

I

I I XN-NF-77-18 I

I The decrease in the feedwater enthalpy leads to a decrease in the primary coolant temperature and pressure, along with a reduction in secondary I pressure in the steam generators. A low pressurizer pressure trip signal occurs at 124.7 seconds. The core average temperature decreases to 523°F I and the pressurizer pressure decreases to 1708 psi a. An MDNBR of l .75 for this case occurs at 8.0 seconds.

I 3.4.2 fncreased Feedwater Flow From 52% Power The feedwater flow is increased from 50% to 100% in 8 I seconds to initiate the increase in feedwater flow incident. The reactor I is assumed to be operating in the manual mode at 52% of stretch power.

No credit was taken for the high water level alarm in the steam generators I which automatically closes the feedwater regulating valves. End-of-cycle (EOC) reactivity coefficients are assumed, since the negative moderator co-I efficient acts to increase core power during the transient. The nominal I EOC Doppler coefficient is multiplied by 1.2. The initial reactor pressure is 2010 psia, 50 psi below the nominal value of 2060 psia.

I The system responses during the increased feedwater flow incident are shown in Figures 3.45 through 3.49. The reactor power I

I and the primary temperatures approach asymptotic values within one minute as shown in Figures 3.45 and 3.47. The MDNBR decreases from an initial steady-state value of 3.37 to a minimum value of 3.00 during the incident.

I I

PALISADES ++ REDUCTION OF FEEdWAlER ENTHALPY FROM 2580. S MW 14 0 1 POWER I EVEL 2 HEAT F, ux ..

3 TOTAL Il>RIMARY CO DLANT FLOW lj TOTAL EEO WATER LOW 5 TOTAL ~TEAM LINE FLOW 1 2 1 2 ~ 3 3 3 3 3 3 c: .1 2 3 'i n 3 <; c 100 *' 1 ;.::

p

'I ~ <!

-f-4-_~

r--.:::::::::::~

~\

0 w

f-a:

Cl'.:8 0 LL 0

\ *O"\

co I

f-z W~)O u

I..

l I\ I

\ \

er:

w Q_

\ I\.\

rt 0 20

~~

0 0 20 40 60 80 100 120

\: ....._

14 0

.. 5 1 2 160 5

f\

180 2 ti 5

200

z:

I

z:

I

-....i

-...J TIME, SEC I co FIGURE 3.40 POWER, HEAT FLUX, AND SYSTE~ FLOWS, SEQ. 1?42 22 JUN 77 19 *H *04 REDUCTION OF FEEDWATER ENTHALPY

... - - - *- --'.- - -~* .. - I- - -*-- - -

PALISADES ** REDUCTION OF FEEDWATER ENTHALPY FROM 2580. 6 MW 180 ~ RVE. F UEL TEMPER! ATURE

~~

3.

CORE r: 11 LET TEMPE bfffURE

'1VE. c:ORE COOLAN TEMP.

I

.J. CLAD T MPERRTURE 16 0 ~

l 1 1

1 14 0 :IJ r--_

r----_

120 IJ- ~

LL I 0)

C!J w "° I

0 100 '-!

\

800

\

600 3

'l 3

q 3

q* ..

q 'I ,;

\ ~ ' T 3

7 <:. 'Z 'Z c:

3

~

3 ~ 9 3 q 2 3 'l 2 3 '!-

z I

z 40~ 20 40 GO 80 100 120 14 0 180 180 200 I

-..J

-..J TIME. SEC I co FIGURE 3.41 CORE TEMPERATURE RESPONSES, SEQ. 1?4222 JUN ?7 t.9:H*05 REDUCTION OF FEEDWATER ENTHALPY

PAL I SADES *

REDUCT 1ON OF FEEDWATER ENTHALPY FROM 2 5 8 0. 5 MW 10 i CHANG IN AVE. ~RIMHr<T CO ~LAN1 Tt.nt-' . Luu~ l 2 CHANG IN AVE. :ORIMARY CO DLANT TEMP . LOOP 2 3 CHANG IN HOT L G - COLD EG TEMP. HF FERENCE LOOP 1

., 2 3 '! 3 4 3 4 3 4 4 CHANG IN HOT L. G - COLO EG TEMP. )IffERENCE ~OOP 2

- -- 0 --....._,, ---....

~

-iO

~

~~

-20 LL

~

w

~ I

-....J 0

' ~~

I 0

-30 i 2 1_3---- ~

-40

\

-50 I I

~. ...........__ 3 '! 3 4 II ><

z

z I

"'Tl 20 :+ 0 60 80 100 120 i 40 150 180 200 I

-....J

-....J TIME. SEC I FIGURE 3.42 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. t ?' i 2 2 2 JUN ? : 13 !I : iJ 6 REDUCTION OF FEEDWATER ENTHALPY

---~-----~---------

3 0 01:-----,----------------,s=-:-=:::.=-:.v..,.--.-=:=-:,:--=*::~?R~'E=s=sl-=:JR--=oE-c=-:-"-=.9~NG~::~_--,'J~'J~P*-._-----~-----,

2 s~E~~ IDJ~:: cR::~s~RE c~RNGE ~oo? 2 I 3 CRESS~~=~~~ FREStURE CHANG 200r.---+----r-----+------'----.....__---+-----1------+----+---~ 1

-+'

1 i I 2

I I, =---1 100!--~r-----t-----+----t----+----+----+~c..__..=-t----+------i 3-_2 *=-=,

cc I u:: I 0.

-lQ!Fr----+----+----~-?---+---~----+--~--l-----1-----i~--~

3

z I

z

""Tl 20 u fJ GO go "l'J:j 120 14 0 160 180 200 I I

co FIGURE 3.43 PRESSURE CHANGES IN PRESSURIZER AND SEu. i?'42 22 .JlJN ?7 'i.3 "l7 *08 STEAM GENERATORS, REDUCTION OF FEEDWATER ENTHALPY

PALISADES ** REDUCTION OF FEEDWATER ENTHALPY FROM 2580. 6 MW 1 1 20 - - - ---------:----- - _____:______ ~ -- ~~:~~-r~~-~~~:~ ~~~:--~~~~=: ~~~~~: t~~- ~ ---~-------*

. ! 1 2 3j CH I ER WATER LEVEL ' '

1 2 1 2 1 i ; I

-+- --+-

1 :

' I I 0 ----- j 3 I

-201 i

I iI c.n *-4 Oi' w I

--.,J I

u N I

z I-<

-so:

-8 0 ----------------~----

-10*0

z I

-1 20--- ---- --~----*- -- - ---- - -----------. ---*- - -- -------- -*------------ --- ---*--- ---------------------- - - - - - - - ,.,

z 1J 20 40 60 80 100 120 14 0 160 180 200 I

--.,J

--.,J TIME. SEC I co FIGURE 3.44 LEVEL CHANGES IN PRESSURIZER AND SEQ. 17'4 2 '2 2 JUN 1'7' 19 I 47 : 10 STEAM GENERATORS, REDUCTION OF FEEDWATER ENTHALPY

- - - *- _.. - - ---- j- - - -- __ _ .___ .._. - - -

PALISADES ++ INCREASE IN FEEDWATER FLOW RT 52% POWER 11 0 l POWER EVEL 2 HEAT F ux 3 TOTAL )RI MARY CO )LANT FLOW lj

- lj - lj - lj lj rnTAI ~i:-i:-nwi:iri:-o :-r nw 1.5 u v

-1 " IU l n .. ,_,I\ '-1.~ .. -.

100 " I '-""

90 0

w l-a:

tl'.:g 0 LL I 0

1-w I

z t3'i'O Cl'.:

w CL . '

so 1 ') 1 2 l 2 L Z l <! L ;! 1 ;! l ;!

~

1 2 ~

50 r--._ 5 5 5 5 5 5. 5 5 5 5

z I

z "Tl 20 40 60 80 100 120 14 0 160 180 ' 200 I

-...,J TIME, SEC "' I co SEO, 19 12 2 1 JUN 7 7 2 0 ' 2 2 I 5 2 FIGURE 3.45 POWER, HEAT FLUX, AND SYSTEM FLOWS, INCREASE IN FEEDWATER FLOW AT 52% POWER

. _c* .. - - _

PALISADES ** INCREASE IN FEEDWRTER FLOW RT 52% POWER 12 0 u I

1. l'lVE. l='I JEL TEMP ERi RTURE 2 CORE Ill/LET TEMPE RATURE 3 l'\VE. ClbRE COOLAN TEMP.

.i CLAD Tl MPERATURE f.J 1 1 1 1 1 1 1 1

100

""l ~

./

900 LL I

-..J (9 ..j:::>

w i 0

800 700 600 4 4 4 4 tj 4 4 lj 4 4 3 3 3 3 3 3 3 3 3 3 50~ 20 Lf 0 80 80 100 120 140 i60 180 200 TIME, SEC FIGURE 3.46 CORE TEMPERATURE RESPONSES, SEQ. 19 12 2 l .J i..iN ? ? 2 0 '2 2 1 93 INCREASE IN FEEDWATER FLOW AT 52% POWER

~-------~~-----

PALISADES ** INCREASE IN FEED~ATER FLOW AT 52% POWER 1 CHANG IN AVE. 0 RIMARY CO PLANT TEMP . LOOP 1

.v I

- 3 q " ~ -

q . n -*

.A.lllll1o. """~ .... -111'11 l-lll

LUO! "

3 CHANG IN HOT L G - COLD EG TEMP. DIFFERENCE LOOP 1 11j CHANG IN HOT L G - COLD EG TEMP. DIFFERENCE LOOP 2 q

3 2

II 1

I LL

(!)

w 0 'r! 01 I

'-l I

0

-1 v\

\

-2

\ l L 1 2 1 2 1 2 1 2

..J,__2 l 2 l 2 x

z I

z

"'Tl

-3 20 160 180 200 I 0 40 60 80 100 120 14 0 '-l

'-l TIME. SEC I FIGURE 3.47 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. 1912 21 JUN 77 20 122 *54 INCREASE IN FEEDWATER FLOW AT 52% POl~ER

PALISADES ** INtREASE IN FEEDWATER FLOW ~T 52% PO~ER 80 l STEAM OOKE PRESS ~RE CHANGE LOOP l

2. STEAM DOME PRESS 1lRE CHANGE LOOP 2

3. PRESSU HZER PRES ~URE CHANG 40 3 3

~

3~

~ I

-...J

~

I

~

\\

3 .__........

3. -*-- *- . _]:..--- ~ --- '

-20

\ :

-40

~.*

~

1 2. 1 2 t 2 1 2 1 2 1 2 1 2 z

I z

"Tl

-20 ~o 60 80 100 120 14 0 . 160 180 200 I

-...J

-...J TIME. SEC I FIGURE 3. 48 PRESSURE CHANGES IN PRES SUR I'ZER AND STEAM GENERATORS, INCREASE IN FEEDWATER FLOW AT 52% POWER

PALISADES ** INCREASE IN FEEOWRTER ~LOW AT 52% POWER 12or-~~r-~~-,-~~-,~~-;-;~r;-;-;;~TTU-==..-.-=-::o---.:::-===r.-~;;:.._~~~~...-----~~

1 CHANGE IN STEAM ~EN. ~ATER LEVEL. LOIDP 1 I

2. CHANGE IN STEAM !SEN. ~ATER LEVEL. LOlbP 2 ! I

3. CHANGE IN PRESSUt I ZER i-JATE ~ LEVEL 1 1oor--~~r-~--ir---~---1f--~~f--~--1~~--1~~--i~~--i~~--l~~_JI I

i 80 cn60 w I I -...J u -...J z

1--;

I 40

~

20 r 0

J. 3 3 3

3 3 3 3 z

I z.,

-2~ 20 40 60 80 100 120 140 16 0 180 200 I

-...J

-...J TIME. SEC I SEQ. 19 12 2 1 J UN 77 2 0

2 2

5 7 FIGURE 3.49 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, INCREASE IN FEEDWATER FLOW AT 52% POWER

I XN-NF-77-18 I

I 3.5 EXCESSIVE LOAD INCREASE INCIDENT Excessive load incidents may be initiated by opening of the I turbine control valves, atmospheric steam dump valves and/or the steam bypass to condenser valve. This results in rapid increase in steam flow I

which causes cooldown of the primary system. Along with an increase in I nuclear power, there is a decrease in main steam pressure and in primary coolant temperature and pressure. Protection against damage to the reactor I core as a consequence of an excessive load increase is provided by the high neturon flux, low steam generator pressure, and thermal margin trip settings.

I Although excessive load incidents are much less severe than steam line breaks, I which result in a much more rapid primary system cooldown, the analysis was performed to assure that the integrity of the reactor core is maintained dur- I ing these more credible incidents.

The excessive load increase incidents were analyzed from 102% of I

rated power and from hot zero power (rods at power dependent insertion I

limits). Both cases were analyzed with two loop (four pump) operation as required by the technical specifications for these operating conditions. I End-of-cycle (EOC) kinetics parameters were used to maximize core power during the reactor cooldown. For the excessive load incident in which I

primary system pressure is decreased sufficiently to activate the safety I

injection system (initiated from hot standby), minimum capabilities of the boron injection system were assumed. The use of the kinetics para- I meters and assumptions concerning the availability of the boron injection system are detailed in Section 3.8 (Steam Line Break).

I I

I

XN-NF-77-18 3.5.1 Excessive Load From 102% of Stretch Power I At the full power condition, the turbine control valves, I the steam dump valves and the bypass valves are suddenly opened.

system responses are shown in Figures 3.50 through 3.'54.

The Increased steam I flow causes the reactor power to increase to 2870 Mwt at 11.0 seconds. High neutron flux trip occurs at 10.6 seconds. Correspondingly, the steam pressure I as well as the primary coolant system pressure decrease. After trip the I turbine control valves are closed, but the steam generators continue to blow-down at a lower rate through the steam dump and bypass valves. When the core I average temperature reaches 532°F, the steam dump and bypass valves close, terminating blowdown of the steam generators'. There is no return to power I following the trip and the minimum DNB ratio during the incident is l .74.

I 3.5.2 Excessive Load From Hot Standby The system responses to the suddert opening of all steam I durnp and bypass valves from hot standby with a shutdown margin of 2% /\p are shown in Figures 3.55 through 3.60. The initial reactor core pressure I is 2110 psia, 50 psi above the nominal value of 2060 psia. This is conserva-I tive, since it delays the boron injection during depressurization of the pri-mary system. The reactor is tripped on an overpower signal of 10% of 2530 Mwt I at 17 seconds following the opening of the valves.

After rod insertion following the reactor trip, the core I returns to power at 75 seconds, peaking at 10.2% of stretch power at 140 2

I seconds. The maximum core average heat flux is 17,075 Btu/hr ft . Borated water from the high pressure safety injection system reaches the core at I

I

I XN-NF-77-18 I

I 80 seconds. Boric acid from the charging pumpings reaches the core region after 128 seconds, terminating the core power increases. I The. critical heat flux was determined using the modified 4

Barnett CHF correlation. ( ) For conservatism, the most reactive control rod I

was assumed to be stuck out of the core when evaluating the shutdown I capabilities of the control rod. The local peaking factors occurring in the area of the stuck control rod after return to power were calculated using I 9

XTG.( ) Local effects of Doppler and void feedback were included in the calculations of the local peaking factors. The negative feedback effects I

due to voiding were not used, however, in the determination of overall core 1*

power. The lowest margin to critical heat flux occurs at the time of maxi-mum core average heat flux. For the core conditions at this time (10.2 I percent core heat flux, 743 psia, and 414°F inlet temperature) the minimum CHF r~tin was calculated to be 3.60 with a hot rod local peaking factor of I

T FQ _= 16. 0. I I

I I

PAL I SADES +* EXCESS I VE LOAD INCREASE FROM 2 5 8 0. 6 MW

' 14 0

~~

POWER EVEL ux I I HEAT F I

I

- 51

[ Ii I i

31 11

s]

TOTAL TOTAL TOTAL

'RI MARY COi DLANT FLOl-l EEOWATER I LOW

>TEAM LINE FLOW I I i

120 T 1

I I I I i I 1 2 L~ \ I 3 I 3 3 3 3 3 3

\ 3 3 100

~~

0 II w

l-a: I i

er: 8 0

~

LL 0

1-z tj6 0

\ ~'

I Ii I

co I

r\

er:

~

w a_

~

40

\~ ~ ~" I I

I 20 I~-

~

~ rl_ ~ ~ ~

I\

I 5 1 2

- - 1 2 'I i: 1 2 lj Ii 1 2 4 si I

v 0 0 10 20 30 40 50 60 70 80 90 100 TIME. SEC co FIGURE 3.50 POWER, HEAT FLUX, AND SYSTEM FLOWS, SEQ. 445 27 JUN 7714156*04 EXCESSIVE LOAD INCREASE

PALISADES ** EXCESSIVE LORD INCREASE FROM 2580. 6 MW 180 u 1 AVE. F ~EL TEMPER! HURE 2 CORE 1 ~LET TEMPEi 'ATURE 3 AVE. C ~RE COOLAN TEMP.

tj CLAD T HPERATURE 16 0 u n

l 1 __..,...

14 0 u 12 0 J

\

LL I co

(.!J N w I 0

10 0 J

\ \

800 600 3 4

~ -..l_____ l ..

1 I l l

(.

3 -- - ., 4 I) 3 4 ') 3 4 2 3 4 2 3 4 2 3 lj 2 3 lj 2 3 lj z

I z..,.,

I I 10 20 30 40 50 60 70 80 90 100 '...J

'...J I

TIME. SEC co FIGURE 3.51 CORE TEMPERATURE RESPONSES, SEQ. 445 27 JUN 71' 14 :56 :1)5 EXCESSIVE LOAD INCREASE

PALISADES ++ EXCESSIVE LOAD INCREASE FROM 2580. 6 MW 10 l CHANG I IN AVE. PRIMARY CO

,/n JLANT TEMP . LOOP l 2 CHAN GI IN AVE. PRIMARY CO )LANT TEHP . LOOP 2 3 CHANG1 rN HOT L G - COLD EG TEMP. ) IFFERENCE LOOP 1

~ CHANG IN HOT L G - COLD EG TEMP. ) IFFERENCE LOOP 2 0

~ ~

-10 ~

-20 \~ '\

\ ~ ,"----.

LL I

(!) OJ w

w I 0

-30

~

~ 1 2 1 2 1 2 1 2

-4 0

'~

-50

~

~II 3 tj 3 tj 3 tj 3 tj

z I

z "Tl

-6~ 10 20 30 40 50 60 '1' 0 80 90 100 I

--...J

--...J I

TIME. SEC OJ FIGURE 3. 52 PRIMARY LOOP TEMPERATURE CHANGES, EXCESSIVE LOAD INCREASE

PAL I SADES ++ EXCESS I VE LOAD INCREASE FROM 2 5 8 0. 5 Mhl 400 l STEAM OOME PRESS! IJRE CHANGE LOOP l 2 STEAM OOME PRESS IJRE CHANGE LOOP 2 3 PRESSU bIZER PRES URE CHANG 300 200 l 2 l 2 l 2 l 2 l 2 l 2

,,. l 2 1 2 100 er:

U?

Q_

0 I CD I

I

-10 '-!

~~ ~

~

II

-20 IJ

~ ,

I j

~ r-----1-__ 3 3 3 3 3

z I

I ::z L I u

10 20 30 40 50 60 ?O 80 90 100 -....J

-....J I

TIME. SEC CD

-::GURE 3.53 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 445 27 JUN ??14*56*08 STEAM GENERATORS, EXCESSIVE LOAD INCREASE

~

PAL I SADES + + EXCESS I VE LOAD I NCR ERSE FROM 2 5 8 0. 6 MW 20 -~

1 CHANGE [N STEA!i 'EN. l-IATER LEVEL. LO p t

~

l 2 3

CHANGE [N STEAM 'EN. wATER LEVEL. LO OP 2 CHANGE [N PRESSU ~rZER l-IRTEI ~ LEVEL 0

-20

\ \

(f)-4 0

~~ ,

\ \"

w I I co

~~

u z

-60 1 .

~

1 -

(.J1 I

-80

~

-10 .J

~ ~

3 -

? -:i 3 3 z

I

z:

I

-1~ u

~

10 20 30 40 50 80 ?O 80 90 100 "'-J

"'-J I

TIME. SEC ---'

co FIGURE 3.54 LEVEL CHANGES IN PRESSURIZER AND SEQ. 445 27 JUN 7714=56110 STEAM GENERATORS, EXCESSIVE LOAD INCREASE

_J

PALISADES ** EXCESSIVE LOAD INCREASE FROM HOT STANO-BY 1If0 1 POWER EVEL 2 HEAT F ux 3 TOTAL l>R1MARY CO )LANT FLOW ti TOTAL EEDl-JATER LOl-J 5 TOTAL pTEAM UNE FLOW 12 0 3 3 3 3 3 3 3 3 3 3

100 0

w f-a:

tY80 LL I 0 co en f- I z

tj60 CY w

Q_

40

~

5 -

20 -----..:. "'""-- 5 5 5 5 5 5 1 2 l <: l <: ><

l~

_M '

l f)

~

ij

'I 1 '

4 1

~ ~

" ~ 'I 4 4 l '

'I z

I z...,.,

I 0 0 20 ~o 60 80 100 120 14 0 160 180 200 -...J

-...J I

TIME. SEC co SEQ ti' 2Ei 2 3 JUN ?"!' 13 I 3? : l 6 FIGURE 3.55 PO~ER, HEAT FLUX, AND SYSTEM FLOWS,

EXCESSIVE LOAD INCREASE FROM HOT STAND-BY

PALISADES ** EXCESSIVE LORD INCREASE FROM HOT STANO-BY 5?5i-~~--,-~~~--,~~-:--,--~~,1'.~i'\\TF~=-rn=;--.==:-;:;~~=-~-r-~~~-,----=~~~~~~

2 LL I

(9 co

-....J w I 0

475r--~~-r-~~--1-~~:--t~~----jf--~~+-~-/--+-~~-+~~-4~~---=::~~~

z:

I

z:

"'Tl 4 o~~~--;-:-;:;--~~4~0~~~6~0~~~8~0~--:---:-'-:10~0=--~~1~20~~-1L4_0~~1~6-0-,~--118-0~~~200 I

-....J

-....J I

TIME. SEC co SEQ. 1726 23 JUN i? 19

3?

19 FIGURE 3. 56 CORE TDlPERATURE RESPONSES, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY

PALISADES ** EXCESSIVE LOAD INCREASE FROM HOT STANO-BY 25 1 CHANG! IN AVE. t:iRIMARY CO JLANT TEMP

LOOP 1 2 CHAN GI IN AVE. It>RIMARY CO PLANT TEMP . LOOP 2

--3 '! 3 CHANG! IN HOT L G - COLD EG TEMP. HFFERENCE LOOP l 3 lj ti HJ r 1b:o ,..,_ TC:- .. 0 'r ; nnP 9

.! 3 " CHANG1 unT i:-r:-i:-i:;ii:-1Jri:-

3 " "'" n 0

-25

~

-50

~ 1

~

LL I

(!) co co w I 0

~9

-?5

~

-10 u-C.)

' ~~

--L1____ -LL__

-LL__ -LL

z I

z:

-1~ u 20 40 60 80 100 120 140 160 180 200 I

-.....J

-.....J TIME, SEC co I

SEQ. 1726 23 JUN 71' 19

37

20 FIGURE 3.57 PRIMARY LOOP TEMPERATURE CHANGES, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY

PALISADES ** EXCESSIVE LOAD INCREASE FROM HOT STAND-BY 0 1. STERH JUME PRESSI JRE CHHNGI:: LOOP t

~

2. STEAM l DOME PRESSI JRE CHANGE LOOP 2 PRESSiJf ~ IZER PRES ;URE CHANG!

\~

3.

~'J

-20 u~ ~

-40 J

~ \

-60 u

\' ~~ 1 2

. -" 1 2 1 2

\\

1 2 I

a:

....... OJ

\.0 Cf) I I

~

-80 u uu

\

-12 Ju

~

~r---L 3 3 3 3

z I

z

-n

~

I

-1~ JU 20 40 80 80 100 120 140 160 180 200 -...J

-...J I

TIME; SEC FIGURE 3.58 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 112623JUNn19 1 37*22 STEAM GENERATOR, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY

PALISADES ++ EXCESSIVE LOAD INCREASE FROM HOT STAND-BY 20 CHANG N ST AM 2 CHANGE IN STEAM EN. WATER LEVEL. LO P 2 3 CHANGE IN PRESSU IZER WATE LEVEL 0

(f)-4 0 w I I <.O u 0 z

1--;

I 3 3 3 3 3 3 3 3 z I

z

""Tl I

20 40 Ei 0 80 100 120 14 0 180 180 200 -....J

-....J I

TIME. SEC OJ SEQ. l 7 2 8 2 3 JUN 'i'7 18 i 3 7 '2 3 FIGURE 3.59 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY

PALISADES ** EXCESSIVE LORD INCREASE FROM HOT STANO-BY 10 1 !10DERA OR REACT! ITY 1

2 DOPPLE R REACTIVI y 3 BORON REACTIVITY ,

ll TOTAL REACTIVITY 8

1 6 1 1

~

,~

v l.--"'""

(f') I 0:: \,D a:

_J 62 y

/ i 0

1~ 2 3 2 3 2 3 ll ~ ~'.-!

4 4 4 4 4 4

r-t__ -

0 j

~

v ~ 3 2 2 " 3 2 -

v v

4/

-2 z

I z,,

-4 20 ' 180 200 I

0 40 60 80 100 120 14 0 160 -.....J

-.....J I

TIME. SEC FIGURE 3.60 REACTIVITY FEEDBACK, EXCESSIVE SEQ. l 7 2 6 2 3 JUN 1'l' 1 g I 3 7 I 2 4 LOAD INCREASE FROM HOT STAND-BY

I I

XN-NF-77-18 I

3.6 LOSS OF LOAD INCIDENT I

In the event of a complete loss of load while the reactor is operating at full power,there would be a significant reduction in the I rate of heat removal from the primary coolant system. Under these dr-cumstances the steam dump to the atmosphere and steam bypass to the con-I denser are available to remove energy from the primary coolant system. I If credit is not taken for steam dump to the atmosphere and steam bypass to the condenser, the actions of the pressurizer relief valves and pres- I surizer and steam generator safety valves would assure that both primary coolant system and steam generator pressure do not exceed design limits.

I:

The acceptance criteria for this event are: I ..:

The pressurizer safety valves shall limit reactor coolant system pressure to a value below the ASME 1:

code limit of 110% of design pressure (2750 psia) .

The pressurizer safety valves shall limit the pressure 1:

differential between the primary and secondary systems I to less than 1530 psid.

There shall be no core damage (MDNBR ~ 1.30) during I the transient.

The most probable cause of a rapid loss of load is a turbine I

trip. This analysis considers plant behavior upon a trip of the turbine- I generator without a direct reactor trip in or~er to demonstrate that the I

I I

1 I XN-NF-77-18 I

I primary coolant and main steam systems are adequately protected during a complete loss of load transient. Several cases were analyzed to determine II the closest approach to each of the above criteria. Tran~ient responses are calculated from 102%' of stretch power. Beginning-of-cycle (BOC)

I kinetic coefficients were conservatively assumed, with an 0.8 multiplier applied to the Doppler coefficient.

I The worst case with regard to peak primary side pressure (Case l)

I is the transient initiated from 2110 psia with no pressurizer relief, pres-surizer spray, steam dump or steam bypass allowed. Figures 3.61 through

!I 3.65 show the transient responses for this case. The peak pressurizer pressure reached during the transient is 2394 psi a. The case yielding the I highest transient primary-secondary ~p (Case 2) is the Loss-of-Load I initiated from 2110 psia and with no primary side pressure reducing effects from either pressurizer spray or relief valves, but allowing depressurization I of the secondary side via atmosphere steam dump and steam bypass to the condenser. This results in a combination of high primary pressure and low I secondary pressure, giving a maximum ~P between the primary and secondary I sid~s of 1388 psid during the transient.

the results for this case.

Figures 3.66 through 3.70 show Both of these cases were tripped on high I pressurizer pressure.

The case yielding the lowest MDNBR (Case 3) is the transient I initiated from 2010 psia with the pressurizer spray and relief valves operable, I but the steam dump and steam bypass to the condenser was inoperable giving a peak core average temperature of 593.?F. This results in a combination of low I

I

I XN-NF-77-18 I

I primary pressure and low inlet subcooling giving an MDNBR of l .39. Reactor trip occurs on high power level. Figures 3.71 through 3.75 show the I results of this case.

The Loss-of-Load transient results are summarized in Table 3.1 I

In none of the cases were the Palisades Technical Specifi~ation Limits of I l) peak primary system pressure (-5_ 2750 psia), 2) primary-secondary AP

(~ 1530 psid), or 3) MDNBR (~l.30) violated. I I

I I

1*

I I

TABLE 3.1 LOSS OF LOAD TRANSIENT RESULTS Max. t.iP**

Initial Initial Allow Pressurizer Allow Atmospheric Peak* During Primary Secondary Relief Valve Steam Pump and Pressure Transient Case Pressure Pressure Opening and Spray Condenser Bypass osia psid MDNBR 2110 772 No No 2394 1371 l. 55 2 2110 772 No Yes 2382 1388 l. 57 3 2010 772 Yes No 2274 1247 l.39 I

<.O (J1 I

Technical specification limit on primary pressure is 2750 psia

- Technical specification limit on primary-secondary t.iP is 1530 psid during transients.

z I

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PALISADES++ LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA HO P ~E 2 HEAT F UX 3 TOTAL RIHARV CO LANT FLOW

~ TOTAL ~EDWATER LOW 5 TOTAL TEAM LINE FLOW 120 3 ~

100 0

w

' I-a:

0:'.:80 LL I

<.O 0 O"l I

I-z wso u

O:'.:

w 0....

6 40 20 1 1 ><

z I

0 0 8 12 16 20 24 28 32 36 40 ~

-...J

-...J I

TIME. SEC co FIGURE 3.61 POWER, HEAT FLUX, AND SYSTEM FLOWS, srn. 4 oo t r J UN rr 1o , o5 , os LOSS OF LOAD - CASE l

PALISADES ** LOSS OF LOAD FROM 2580. 6 M~ & 2110 PSIR 180 J l AVE. F lJEL TEMPER HURE 2 CORE I ~LET TEMPE ~ATURE 3 AVE.* C1 bRE COOLAN TEMP.

~ CLAD Tl MPERATURE

~

160 .J

__ 1 HO ~

1 1

\

lL.

(!)

120 u

\ I

\D

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I Cl 100 u

800

~,

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'/

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'l

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., -4

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{, ;.: : "' () '2 " I) '2 4 x

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8 12 16 20 2q 28 32 36 ~o --..J

--..J I

TIME. SEC FIGURE 3.62 CORE TEMPERATURE RESPONSES, SEQ. 'i 0 0 lT JUN ir l 0

0 5 r 0 'i LOSS OF LOAD - CASE l

PALISADES *+ LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 30 . LOOP t

. LOOP 2 IFFERENCE LOOP t IFFERENCE LOOP 2 20 10 0

LL I

<.D

(!J co w I 0

-10

-20

z:

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z:

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I 8 12 16 20 28 32 36 Lf 0 -....,J

-....,J I

TIME. SEC co FIGURE 3.63 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. q00 l 7 J UN 77 10 I 0 5 I 0 5 LOSS OF LOAD - CASE l

---~---------------

PALISADES++ LOSS OF LOAD FROM 2580. S MW & 2110 PSIA 500r-~~-,~~~~~~~-,~~----,,---...:rFr;:;:;-;~;::-;:;~~~..,,....,,==~~L~OO=P,..:...;_1___:,.--=-::....:....:_~~~~------

2 LOOP 2 3

I cr::

\.0

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1oor--~.__r-~,L--t-~~+-~~-t-~~-1-~~-t-~~-+--~~-+-~~-i-~~-l 0

Z.

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8 12 16 20 24 28 32 36 40 '-J

'-J I

TfME. SEC co FIGURE 3;64 PRESSURE CHANGES IN PRESSURIZER AND SEQ. qoo 17 JUN 7710105107 STEAM GENERATORS, LOSS OF LOAD - CASE l

PALISADES ** LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 120 l CHn111GI:. lN S!t.nn pt.N. WM I t.t< LEv'EL. LUI ~p l 2 CHANGE IN STERt1 ~EN. WATER LEVEL. LO bP 2

3. CHANGE IN PRESSUI ~riER WATEI b* LEVEL 80 3

':i . /

t..-- r--t.._

If 0 v ~

(J}o l 3 _..,...- v .' -

l 2

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'~ ~...l

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TIME. SEC co FIGURE 3.65 LEVEL CHANGES IN PRESSURIZER AND SEQ. 400 17 JUN 77 10 : 0 5 I 0 8 STEAM GENERATORS, LOSS OF LOAD - CASE

PALISADES ** LOSS OF LOAD FROM 2580 6 MW & 2110 PSIA 14 0 1 !-'OWt::R L.t. Vt.L 2 HEAT F ux 3 TOTAL hR[MARY CO DLANT FLOW tj TOTAL EEOWATER LOl-l 5 TOTAL ~TEAM LINE FLOl-l 12 0 1

1 1~-

2\

1 2 2

\"'\ ,, ~ ':! 3 3 01~

10 ~

31 \ ~

1 \

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20

r-L_ 1 N...__

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1 1 2

- x

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"'Tl I

0 0 8 16 20 24 28 32 36 40 ........

I TIME. SEC FIGURE 3. 66 POWER," HEAT FLUX, AND SYSTEM FLOWS, SEQ. 235 17 JUN 7 7 10 I 0 0 I 2 2 LOSS OF LOAD - CASE 2

PALISADES +-t LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 180'U 1 AVE. F ~EL TEMPER HURE 2 CORE ! ~LET TEMPEi ~ATURE 3 AVE. CDRE COOLAN TEMP.

CLAD T MPERATURE 160 0 1

U----

14 0 lJ l 1

\

120 u

~ -

LL

(!)

w D

\ I\ -'

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11 I

100 u

~

800

~

r---__ J1---_

4 ' 1 600 4

~

v

" -:i

., 4

., v 4 LI

~ ~

£ 3

£ " "- "- ~ n "> 4 I) ':! "

z I

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-....J 4 8 12 16 20 24 28 32 36 40 -....J I

TIME. SEC FIGJRE 3.67 CORE TEMPERATURE RESPONSES, SEQ. 2 35 17 JUN 7 7 10 : 0 0 : 2 4 LOSS OF LOAD - CASE 2

. .. 1 - . ....- -

PALISADES ++ LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 30 l CHANG IN AVE. "R U1ARY CO JLANT TEMP . LOOP 1 2 CHANG 1:. lN AVE. ~RlHARY CO )LANT TEMP . LOOP 2 3 CHANG 1:. lN 1-jOT L G - COLD EG TEMP. brFFERENCE LOOP t

~ CHANG lN HOT L G - COLO ,...EG TEMP. DIFFERENCE LOOP 2 20 v --......

,/

10 1 2 ./

v ' N

~

0

~

u..

(,!)

w 0

-10

\ ~ ~ '!- ..... ~ w I

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-30

~

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z I

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8 12 16 20 24 28 32 . 36 40 '..J

'..J I

TIME, SEC FIGURE 3.68 PRIMARY LOOP TEMPERATURE CHANGES, . SEQ. 235 17 JUN 77 10 1QO 125 LOSS OF LOAD - CASE 2

PALISADES** LOSS OF LORD FROM 2580.B MW & 2110 PSIR 500 l STEAM JOME. PRESS JRC. CHANGE LOOP 1 2 STEAM lOME PRESS JRE CHANGE LOOP 2 3 PRESSUI >[ZER PRES URE CHANG 400 *-

300

~ 3 J 1~ 1 2 2// ~~ rt..g_

200 vv ~

er:

~ I

~

0 (j) _p.

I CL

£v

./ 1 100

"' ~

0

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-2~ u" 8 12 16 20 24 28 32 36 40 -....J

-....J I

TIME. SEC co SEQ. 235 11' JUN?? 10:00:26 FIGURE 3.69 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, LOSS OF LOAD - CASE 2

. .__ .._ -! - --: __,___.____ - "- -) *- _.. --->- - -) .. -. .... -

PALISADES **LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 60 CHANGE !N STEAM EN. WATER LEVEL. LO P 1 2 CHANGE !N STEAM EN. WATER LEVEL. LO P 2 3 NGE !N PRESSU !ZER WRTE LEVEL 40 3

20

(/)0 w

I .....I u 0 z U1 I

-40

-60 z

I z

I

-8~

-...J 8 12 16 20 24 28 32 36 40 -...J I

TIME, SEC co FIGURE 3.70 LEVEL CHANGES IN PRESSURIZER AND SEQ. 235 17 JUN 7710:00:2a STEAM GENERATORS, LOSS OF LOAD - CASE 2

PALISADES ++ LOSS OF LOAD FROM 2580. 6 MW & 2010 PSIA 14 0 l POWER EVEL !

2 HEAT F ux 3 TOTAL 'RI MARY CO ~LANT FLOW

'I TOTAL EEDWATER 1 LOW 5 TOTAL >TEAM LINE FLOW 120 1

L]:..---

~

l i 1~ 2 Y\ 3 3

- .., .,- ') -.!

100 I v" I .:> \

i~

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w ~ ~

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20 '-i!. .......__ ...!__

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TIME, SEC co FiS~RE 3.71 POWER, HEAT FLUX, AND SYSTEM FLOWS, SEQ. 378 17 JUN 7? OS 11~3 rJJ2 LOSS OF LOAD - CASE 3

180~~~~--r~P_A_L_I_SIAD_E_s~*-*--rL_o_s_s---..,o~F~LruOA~Oc--;:F~R~O~M~2~5~8~0~6~M~W_;__;&=-=2~0~1~0_:,_P~S~I~A~-r-~~~

l AVE. F1lEL TEHPERIATURE 2 CORE [ LET TEHPE"ATURE I 3 AVE. CDRE COOLAN TEMP.

~ CLAD T MPERATURE

\

140~u~~-t-~~j--~-f~---lH-\~~-!-~~+-~--+~~-4-~~--I-~~

I\

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C)

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- lj II r----r1--

600 ~

t:

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-.....J

-.....J I

TIME. SEC OJ FIGURE 3.72 CORE TEMPERATURE RESPONSES, SEQ. 3? 8 1? JUN 7? 0 9 : 4 3

4 4 LOSS OF LOAD - CASE 3

PALISADES ** LOSS OF LOAD FROM 2580. 6 MW & 2010 PSIA 30 1 IN AVE. RIMARY CO LANT TEMP

LOOP 1 2 IN AVE. RIHARY CO LANT TEMP

LOOP 2 3 IN HOT L G - COLD EG TEMP. IFFERENCE LOOP 1 IN HOT L G - COLD EG TEMP. IFFERENCE LOOP 2 20 10 1 2 0

I LL -'

0 (9 00 w I 0

-101--~~-+~~~-1--~~~-hol'~~~+--~~--+~~~-+-~~~-+-~~~+-~~~1--~~-1

z I

z "Tl I

-4~ 8 12 16 20 24 28 32 36 40 -.....J

'+ -.....J I

TIME. SEC ():)

FIGURE 3.73 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. 3 'i' 8 17 JUN 'i'7 0 9 1 'I 3 1 4 5 LOSS OF LOAD - CASE 3

--~

- (-/

~

- (-:* -

,.-ii\ ----*1 . ., ..

PALISADES++ LOSS OF LORD FROM 2580. 6 MW & 2010 PSIA 40or--~~-.~~~~~~~~~___,.~-==---..-~~-,--.--~~~:._:::_:___:_:____;,__::....:...:..:__~~~~

LOOP 1 2 LOOP 2 3

a:

1-1 I

0 en \..D I

CL 0

3 z I

z I

-3~ 8 12 16 20 24 28 32 36 40

"-..J

'+ "-..J I

TIME. SEC co FIGURE 3. 74 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 31'8 17 JUN 1'1' 09 1q3 :LJ6 STEAM GENERATORS, LOSS OF LOAD - CASE 3

PALISADES ++ LOSS OF LOAD FROM 2580. 6 MW & 2010 PSIA

?5 1 CHANGE IN STEAH EN. WATER LEVEL. LO P 1 2 CHANGE I~ STEAM EN. WATER LEVEL. LO P 2 3

3 ~-,.._-

CHANGE IN PRESSU IZER WATE LEVEL 50 3

25 3

cnO w I I

u z

0 I

-25

-50

-?5

z

I

z

I

-1~ 4 8 12 16 20 24 2 8* 32 36 40

-....J

-....J I

TIME. SEC co FTGURE 3. 73 LEVEL CHANGES IN PRESSURIZER AND srn. 37'8 17 .JUN 7? 09 "13 *'18 STEAM GENERATORS, LOSS OF LOAD - CASE 3

I

-111- XN-NF-77-18 I

-1 3. 7 LOSS OF FEEDWATER FLOW -INCIDENT A loss of feedwater flow incident may arise due to the rupture of I the feedwater cross-over line downstream of the main feedwater pumps or a condensate pump fault which would cause low suction pressure on both feed-I water pumps. When operating at full power, there would be no corresponding decrease in steam flow from the steam generators. If loss of main feedwater I--- is unchecked, the normal primary coolant system heat sink would be reduced 1, and eventually eliminated. The result would be an increase in core inlet temperature with only the presssurizer relief valves and auxiliary feedwater I system available for the removal of decay heat, until a controlled system cooldown is initiated. The reactor protection system provides reactor I protection through a reactor trip activated by low water level in each I. steam generator with additional protection for the reactor provided by the

,, high pressurizer pressure and thermal margin trips.

The transient was initiated from 102% of rated power (2530 Mwt).

Only complete loss of feedwater is assumed in this analysis since this I condition requires the most rapid response from the reactor control and I portection system. Beginning-of-cycle (BOC) kinetic coefficients were conservatively assumed, with an 0.8 multiplier applied to the Doppler

_, coefficient. The initial reactor pressure is 2010 psia which is 50 psi

-I. below the nominal value of 2060 psia.

zero in two seconds .

The feedwater flow is reduced to Figures 3.76 through 3.80 show the system responses during the loss-of-feedwater flow incident. The reactor trips at 26.7 seconds on low

1 steam generator water level. This precipitat~s a turbine trip and activates I

I

-112- XN-NF-77-18 I

I the atmospheric and condenser steam dump systems. An MDNBR of 1.65 occurs 0.3 seconds after trip. The atmospheric steam dump valves are controlled by the average primary coolant temperature following a turbine trip. The atmospheric dump valves are completely closed 38.3 seconds following the I

reactor trip, after which decay heat is removed via the steam bypass to the condenser. The water inventory in the steam generators is adequate to I

accommodate decay heat and pump heat for an additional 15 minutes, at which I-,,

time the mass inventory of each steam generator is no less than 1700 lbs.

Hence, the operator will have 16 minutes after the initiation of the

,I incident to restore partial feedwater flow by activation of the auxiliary I

feedwater system.

.I I

I I

.I I

I I

PAL I SADES ++ LOSS OF FEED-WATER FLOW FROM 2 5 8 0. 6 MW 140 1 POWER EVEL 2 HEAT F ux

3. TOTAL lRIMARY CO ~LANT FLOW lj TOTAL EEDWATER LOW 5 TOTAL :TEAM LINE FLOW 120 5 1 1 " 3 1 z....... t) 3 3 3 3

< ., ~ \\ ~

3 100 0

w l-a:

\r----....._s crso

~

\

LL I 0 .......

w 1-z t3SO E

\ I

\ \

er ~II w

a...

40

\~ ~5 5 5 5

\ '

20

~ 1

~

~ -r-l--_

2 , 2 ><

1 2 1 2 :z 1.

z lj 'I lj lj 4 lj lj lj

" 40 lj "'Tl I

0 0 10 20 30 50 60 70 80 90 100 '-J

'-J I

TIME. SEC co FIGURE 3. 76 PQ1_4ER, HEAT FLUX, AND SYSTEr1 FL0\4S, SEQ. 784 20 .JLJN 77 13 I 22 I 17 LOSS OF FEEDWATER FLOW

_J

PALISADES ** LOSS OF FEED-WATER FLOW FROM 2580. 6 MW 180 u l. AVE. F EL TEMPER HURE

2. CORE II ~LET TEHPE )AT URE

3. AVE. ClRE COOLAN TEMP.

il. CLAD T MPERATURE 160 J l

1 l

-\

14 0 u

\

120 J

\

LL I l'.D +:>

w I 0

100 u 800

\

~ -L______

600

.j il c:

,J LI

-v ti

" ,J ll

-- ~?tr

" ~ ti 1

I) 3 4 I

') 3 ti 1

') 3 4 l

') 3 ti z

I

' z I

10 20 30 40 50 60 70 80 90 100 -...J

-...J TIME. SEC I FIGURE 3. 77 CORE TEMPERATURE RESPONSES, SEQ. 'i' 8LI 20 JUN '('I' 13 122 I 18 LOSS OF FEEDWATER FLOW

PALISADES ** LOSS OF FEED-WATER FLOW FROM 2580 6 MW 80 1 CHANG! IN AVE. )Rit1ARY CO JLANT TEMP

LOOP 1 2 CHANG IN AVE. )RIMARY CO )LANT TEMP .* LOOP 2 3 CHANGI IN HOT L G - COLD EG TEMP. I~ I FFERENCE LOOP t q CHANG I IN HOT L G - COLD EG TEMP. !~IFFERENCE LOOP 2 60 40 20

~

LL I

(.9 w ~ U"1

~

I 0 1 2 '.':! ,,

0

~

-- 3 q 3 lj

\

-20

\ t'* ~

~ ~

-40

~ r---_3 LL

~ 1 I)

---L.L ~ --l._2 3 11 3 lj

~

z I

z

-6~ ?O 80 90 100 I

10 20 30 40 50 60 -.....J

-.....J I

TIME, SEC co FIGURE 3.78 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. 78'1 20 JUN 77 13 122 119 LOSS OF FEEDWATER FLOW

PALISADES ** LOSS OF FEED-WATER FLOW FROM 2580. S MW 800 l. STEAM IDOME PRESSi ~RE CHANGE i....OOP t

2. STEAM DOME PRESS' ~RE CHANGE LOOP 2

3. PRESSU RIZER PRES >URE CHANG!

soo 400 1 2

--...... ~I) v

/..-

200

~~

~

I a: l 2 __,

~

L?

en 0... 12~ ....

-rW---_ 1 2 O"\

I 0

~ ""- ~ ---- -L..?___

-20 u

~

1---1.--_

--.1.__

--.1.__~ 3

~

-'+ 0 u

\ ><

z I

z

-s~ u" ..

20 30 'f 0 50 80 70 80 90 100 "Tl I

-....J

-....J TIME. SEC I FIGURE 3. 79 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 7 8i 20 JU~ 77 13 :22 *2i.

STEAM GENERATORS, LOSS OF FEEDWATER FLOW

--11--~ - .*,-

- - :-* - - .. - -* *- *- *-*) - *- - /-- - - - -

PALISADES ** LOSS o~ FEED-WATER FLOW FROM 2580. 6 MW 50 CH~~GE IN STEAM EN. WATER LEVEL. LO P t CHANGE IN STEAM EN. WATER LEVEL. LO P 2 3 CHANGE IN PRESSU iZER WATE LEVEL 0

I

'-l I

x

z:

I

z:

10 20 30 ~o so so ?O 80 90 10 iJ I

'-l

-....J TIME. SEC I FIGURE 3.80 LEVEL CHANGES IN PRESSURIZER AND srn. 784 20 .Ji.JN??:.: :22:22 STEAM GENERATORS, LOSS OF FEEDUATER FLOH

Ii

-118.:. XN-NF-77-18 I

3.8 STEAM LINE BREAK I

The break of a steam line results in a sharp reduction in steam I' generator inventory. The resultant pressure decrease causes an increase in energy demand from the primary coolant system which reduces primary I coolant. temperature and pressure. With a negative moderator reactivity coefficient the core would return to a power level, following reactor trip, I

sufficient to cause core damage if unchecked. I The steam line break was analyzed from two initial reactor conditions: l) 102% of rated power (2530 MWt) and 2) hot zero power I (rods at power dependent insertion limits).

two loop (four pump) op~ration Both cases were analyzed with as required by the technical specifications I

for these operating conditions. fhe steam line brea~ was assumed to occur I at the exit nozzle of steam generator No. 2. This results in the fastest' steam generator blowdown, and the most rapid cooldown of the primary I

coolant. Critical flow was assumed at the break and was calculated using the Moody curve.

( l 0) I I

The reactor tripped and the isolation valve in the unbroken steam line closed on a low steam generator pressure signal following the steam line break. For conservatism, the most reactive control rod was assumed to be stuck out of the core when evaluating the shutdown capabilities of the control rod. Both analyses assumed a shutdown margin of -2.0% ~p.

I The variation of reactivity as a function of moderator density I used in the analysis is depicted in Figure 3.81. This relationship I

I I

I

.:-119-: XN-NF-77-18 is for end-of-cycle conditions when the moderator coefficient is most 1 negative. For conservatism the slope of the moderator reactivity curve 1 was increased by 20%. The reactivity variation as a function of core power is shown in Figure 3.82. The slope of this curve was conservatively I decreased by 20% to minimize the negative reactivity feedback during the transient.

I Minimum capability of the boron injection system was assumed which

1. implies the operation of only two of the three available high-pressure safety injection pumps (HPSI) and only two of the available three charging pumps.

I The HPSI flow versus pressure curve used is shown in Figure 3.83. A low pressurizer pressure signal (1571 psia) initiates the safety injection I system. The time required to sweep the lines of low concentration borated water prior to the introduction of 1720 ppm borated water from the high I pressure safety injection pumps has been accounted for in the analysis. The I concentrated boric acid (10,940 ppm) introduced by the charging pump~ was conservatively assumed to reach the primary loop 80 seconds after the I safety injection signal. No credit was taken for the effects of the resident low concentration borated water being swept into the primary loop from I the lines of either system.

I CllF The critical heat flux was determined using the modified Barnett correlation(J) for these transients. The local peaking factors occurring I *in tile vacinity of the stuck control rod were calculated using XTG. (g)

Local effects of Doppler and void feedback were included in the calculations I of the local peaking factors. The negative feedback effects due to voiding I were not used, however~ in the determination of overall core power.

I

-120- XN-NF-77-18 I

I 3.8.l Steam Line Break From 102% of Rated Power (2530 Mwt)

The system responses to a steam line break initiated from I 2580.6 Mwt are depicted in Figures 3.84 through 3.89. The reactor is tripped at 1.0 second by a low steam generator pressure signal. The main I steam isolation valves close at 7.6 confined to one steam generator.

seconds~ after which steam loss is Borated water from the high pressure, I

safety injection system reaches the core after 35 seconds. Boric acid I introduced by the charging pumps reaches the core after 96 seconds, terminating power increase. The steam generator associated with the I ruptured line empties at 126 seconds, after wh"ich cooldown of the primary system is essentially terminated. The peak power reached during the I

transient after control rod insertion was 464 Mwt at 15 seconds, Core I average heat flux peaks at 30,960 Btu/hrMft 2 .

At the time of maximum core average heat flux, the margin I

to the critica*1 heat flux *is minimized. Using the core conditions for this time (19 percent core average heat flux, 386 psia, and 387°F inlet I

temperature) and applying the corresponding local hot rod peaking (F~ = 18.2), I

the minimum CHF ratio was ca)culated to be 1.30. An MCHFR of less than 1.30 would be an acceptable result for this lo\'J probability incident. I 3.8.2 Steam Line Break from Hot Standby The system response to a steam line break initiated from I

hot standby with a shutdown margin of 2% Ap are shown in Figures 3.90 I

through 3.95. *After rod insertion following th~ reactor trip, the core returns to power at 20 seconds, peaking at 27.4% of rated power at 95 I seconds. The maximum core average heat flux is 45,530 Btu/hr n 2 . Borated I

I

I I -121- XN-NF-77-18 I

I water from the high pressure safety injection system reaches the core at 35 seconds. Boric acid from the charging pumpings reaches the core region I after 96 seconds, terminating the core power increases.

As with the break from 102% of rated power. the lowest margin I to critical heat flux occurs at the time of maximum core average heat flux.

For the core conditions at this time (27 percent core heat flux, 582 psia, I and 376°F inlet temperature), the minimum CHF ratio was calculated to be I 1.41 with a local hot rod peaking factor.of F~ = 16.0.

I I

I

r 1.06

1. 05 -

l.04 l.03 4-4-

Q)

~

>, l.02

.µ

...... I N

.µ u 1. 01 N I

ro Q.J c::::

l.00

.99

.98

.97 x

1. 0 0.9 0.8 0.7 * :z:

I Core Average Density (gm/cc) .,

z:

I

'-I

-.:..J I

FIGURE 3.81 VARIATION OF REACTIVITY WITH CORE AVERAGE DENSITY AT END-OF-CYCLE

~

<J I

.µ

r- '

r-

.µ u I co --'

N Q) 0:::: w

' *r-s::

01 Q)

Ol s::

co

..i:::

u z

0 20 40 60 80 100 120 140 I z

Power (% of 2530 MWt) "'Tl I

-...J

-...J I

FIGURE 3.82

VARIATION OF REACTIVITY WITH POWER AT CONSTANT CORE AVERAGE TEMPERATURE

l 000 c..

en I 3 500 N

..j:::>

0 I r-l.J....

0 l--~.1....-~.J......~-L-.~....L....~......l-~-'-~~~----'~~"'--~.........~_,_~...._..__......_~~~--

o 500 l 000 l 500 :z:

I Primary System Pressure (psia) .,,

z:

I

'-I

'-I I

co FIGURE 3.83 HIGH PRESSURE SAFETY INJECTION FLO\.I VS PRESSURE

P9LISROES ++ STEAM LINE 9RERK FROM 2580. 6 MW 140r--i-----.---'-----:--!---.,.~~-~1-r------=oo~w=E~R~'~E~~E=~~--,--~~~--~-~~~~-~

2. f-iEAT F1 :JX I

3. TOTAL.. :iR :~ARY cokRNT FL..O;.J \

i. f TOTAL.. . EED;.JRTER t...ow

1201--T1 ~~+-~~~+-~~-4~~~~5+*~~TO~T~A~L..~S~T~EA~~~*~L..~rN~E~JF~L~O=;.J~~~~~--L~~~__L~~~!

\ _2..3__J---~?~-t'~_.::....2~-t-~3:___-t~_::...3~-r-~-2_.-r---3~-r~_:3::,__-t_ __::3_-t~_:~~*~i

,__~ I I 100 "

l D

w I-a: I cr:g 0 LL 0 I I ..

I- I N

()1 z i I I

w60 u

er:

w 0...

40 I

i 20 I

\N !.

2

\ J.3----µ1 _.=.£-~1"----IT""" ~

~ I 1 2 ><

2 I 4*

~

z 1:

1 I

\)------'lj-----1--___:__lj__.__ __:4:..__iL........-

\

4

" "" ~ :l

z I.

I 0 0 20 40 60 80 100 120 14 0 160 180 200 -....J

-....J I

T H1E, SEC co FIGURt 3.SJ POWER, HEAT FLUX, AND SYSTEM FLOWS, SEQ. 11'03 23 JUN 71' 16 *29 :QI)

STE.4.~i LINE BREAK

PALISADES ** STEAM LINE BREAK FROM 2580. 6 MW 1()0 u I I

I 2.

AVE. i: ~EL TEMPER~TURE CORE [ ~LET TEMPEf RTURE 3, AVE. C DRE COO~RN TEMP.

II

~. CLAD T MPERATURE .

] I I

14 0 '-Ji I

i I I I

I J

12 0

~\

100 IJ

\

.\

LL I C9 w

II I

I I -*

l 0

800 II I 600 I\

\ "'i I

I j

~~- ~": ~1 I

,., '.l "

1 l l l r-1--_ t t

I I!

,., '.l 4 I I

,., '.l 4

" 1 4 ~ 1 4 ,., 3 " ') 3 4 I) 3 4 Lt 00 I r I I :

z I

I I

I I .,.,

z 20~ 20 40 60 80 100 120 14.0 160 180 200 I

-....J

-....J TIME, SEC co I

FIGURE 3.85 CORE TEMPERATURE RESPONSE, SEQ. L703 23 .JUN 71' 16 129 *OF3 STEAM LINE BREAK

- ... - - - .. *- l- - - - - - - -- -* - - -

PR~ISRCES ** STEAM LINE BREAK FROM 2580. 6 MW I !.J CHANG. [N AVE. ~R rMAR'f CO Di....ANT TEMPj. LOOP 2 i

I I

I 2 CHANG IN AVE. PRIMARY CO b:..RNT TEMP* . LOOP 1 I

I :

I I 3 CHANG IN HOT L G - COL.. O EG TEMP. )!FFERENCE L.OOP 2 I

i i ~ CHANG IN HOT L '.:G - COLD EG TEMP. )IFFERENCE LOOP i !

i I 40 I !

I i I i I

(\\_____

i I I

-- I

--....._ ~ l

!J 3

~

' ?

I r---.::__

3 3 I

-40

\ 3 ? ?

LL CD i...LJ 0

-80

'\ ~

~--

4 I :j II 4 4

4 4 4 4 4 I

N I

-...J I

~~

-12 u~ ~ I

~"--~~

I

- ?

-16 J

~

~r----_ -----~ ,____2 1

2 1

2 l

2 x

z

z I

.,... I "Tl I

. -2 ~ u 20 40 [) 0 80 100 120 140 160 180 200 -...J

-...J I

TIME. SEC FIGURE 3.86 PRIMARY LOOP TEMPERATURE CHANGES, SEO. i ?' 0 3 '23 .JUN I ?' 1 {) : 2 9 : 0 9 STEAM LINE BREAK

PALISADES ** STEAM LINE BREAK FROM 2580. 6 MW 800

~:I STEAM )QME PRESS! VRE CHRNGE LOOP 2 I STEAM DOME PRESS! JRE CHRNGE LOOP l I

3. PRESSi.JI IZER PRES: SURE CHRNGI I 400 l I l I

0~ I 0 I

~

\

/I~

~

-40 i

I

~ r-1--_ 1 1 1 2

-- ~

1 1 2 2 2 2 I

N l

(f.l 1 1 1 co I '\

CL

-80 1-l \ ! I I

-12 L -

i r-.J 'J

\~ I I '

3

~

3 3

3 _t---

-16 '-!*- \..i-

~

! -:: -:: 3 i

I l ><

z I

i - I I I z..,.,

L... -

-2~. lJ lJ 40 50 80

,.oo 120 14r 0 160 180 200 I

-....J

-....J TIME. SEC I co FIGURE 3.87 PRESSURE CHANGES IN PRESSURIZER AND SEQ. l !'03 23 JUN ?7 16: 29 I 10 STEAM GENERATORS, STEAM LINE BREAK

I i

I 2 I I

I I

c.nO w I

r: -'

u z

3 3 3 3 3 3 3 i

i I'

l N

\.0 I

I I

II x

1 1 . i ::z I

z I "Tl I

-....J 14 0 160 180 200 -....J I

TP"1E, SEC co FIGURE 3.88 LEVEL CHANGES IN PRESSURIZER AND STEAM SEO. iro3 23 JUN 'i"i' 18*29'12 GENERATORS, STEAM LINE BREAK

PAL I SADES *

STEAM LI NE BREAK FROM 2 5 8 0. 5 MW 13 t l"lULJt:.t'(A UI"< Rt.Ht.: I I Ill 2

3 DOPP LE R REACTIV I y BORON IREACTIVITY I

l\ TOTAL IREACTIVITY I

I 10 l

~

1 1 I1 1

.J---- 1l

? y

.--- ~ I I

4 2 / 2 ')

.:. :.! ~

2 I

2 2 (J) I a::: . w a:

_J I

I 0

, I I

61 0 J -,

3 3 ., tj tj tj tj tj 3 .. 4 tj v--- 3 I

3

~

I 3 3 I 3

-2

, I i v I l I

/

I I I I I! I i

I Ii

-5 I I I 1v I I i I

z I !

z I

-8 I I I 0 20 40 GO 80 100 120 14 0 180 180 200 -....J

-...J I

TIME. SEC FIGURE 3.89 REACTIVITY FEEDBACK, srn. tr o3 2 3 JUN n u;

2 9 1 13 STEAM LINE BREAK

PALISADES ** STEAMLINE BREAK FROM HOT STAND BY 14 0 l POWER .EVEL 2 HEAT F ux 3 TOTAL ~R I11ARY CO bLANT FLOW TOTAL EEDWATER LOI-I 5 TOTAL HEAM LINE FLOW 120 3 3 3 3 3 '3 3 3 3 3

r 100 0

w I-CC

~80 LL I

0 --'

w 1- I --'

z I t}SO

~ ""-

a::

w CL 40

~

r-----_p_L 1 ')

5 l 2 ~ t 2 c 1 2 5 1 2 5 l 2

~~

5 1 2 5 20 p

~J

z 2

- 'l lj If q q lj q q lj :z I

0 0 y

20

- - 40 60 80 100 120 14 0 160 180 200

"'Tl I

-...J I

TIME. SEC co FIGURE 3.90 -POWER, HEAT FLUX, AND SYSTEM FLOWS, .. SE~. 396 214 JUN 7r 10 1qQ *36 STEAMLINE BREAK FROM HOT STAND BY

PALISADES ** STEAM.LINE BREAK FROM HOT STAND BY I

I l l=IVE. I=" JEL TEMPER TURE I

I I

2 3

4 CORE AVE.

CLAD II lLET TEMPE ~AT URE C DRE COOLAN TEMP.

T' HPERATl.JRE l

GOO J- 1 r-----__ r-1.._

i /

r----- ~

5 i; 0 l /

520 Ii I

"\\ I i--1 LL

(.0

~

c J I/ -------- w N

I I

u

~8:I I

I l

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~~

11 I

t I

iI

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'I

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. 4

~3

'i ~

3 3 3

'I

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I'\

-~

L

.* 2 3 z 2 ..

2 3 I 2 3 I ~

2

'1 I

20 40 80 80 100 120 14 0 16 0 180 200 -..J

-..J I

TIME. SEC co FIGURE 3.91 CORE TE~PERATURE RESPONSES, SEQ. 39S 2 4 JUN H 10 ' 'l 0

3 6 STEAMLINE BREAK FROM HOT STAND BY

PALISADES ++ STEAMLINE BREAK FROM HOT STAND BY 80 1 CHANG E IN AVE. i)R IHARY CO JLANT TEMP

LOOP 2 2 CHANG .._ IN AVE. ~RIMARY CO )LANT TEMP . LOOP r 3 CHANG .._ IN HOT L G - COLD ,..E'G TE'HP. DIFFERENCE L.00P 2

~1---1_

4 CHANG c- IN HOT L G - COLD EG TEMP. )IFFERENCE LOOP !.

40 -

v <I ~ 3 3 3 3 0

\

\\,

.. ti 4 4 4 4 4

0.... I t .

-?5 u L 1 1 1 1 1 1 1

-10 uu-
\
-12
~
~
Cl
\ ~
~ 3 3 3 3 3 3 3 ><
z:
I
z:
~U
- 20 40 60 80 lOO 120 140 160 180 200
'Tl I
--.J
--.J I
TIME. SEC FIGJ~E 3.93 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 396 24 J.UN n1o:qQ:qo STEAM GENERATORS, STEAMLINE BREAK FROM HOT STAND BY

~---------

PALISADES ~+ STEAMLINE BREAK FROM HOT STAND BY 0 l. CHANGE IN STEAH ;EN. i-JATER LEVEL. ri~o 2
\\"'
1 2 2 2 2 CHANGE IN STEAH 1:EN. lolATER LEVEL. :...ono *,
3 CHANGE IN PRESSUI ~IZER 1-JATE! ~ LEVEL
-50
-10 IJ
\~ \ 3 N.. 3 3 3 3 3 3 3 3
~
-15 u~ ~
~
(JI w
~
I.
I u -'
w z Ul I
. .J
-20 u
~
-25 lJ
~
~
-30 - ~
~
,)
z:
I
z
"'T1 I
-3~ IJ
~
--...J 20 40 80 80 100 120 14 0 16 0 180 200 --...J I
TIME, SEC FIGURE 3.94 LEVEL CHANGES IN PRESSURIZER AND SEQ. 396 2 4 JUN ?'i' 10 '~ 0 : 42 STEAM GENERATORS, STEAMLINE BREAK FROM HOT STAND BY
PALISADES ** STEAMLINE BREAK FROM HOT STANO BY 10 1 HO DERR OR REACT! llITY
2. DOPPLE R REACTIVI y 3 BORON REACTlVITY TOTAL REACTIVlTY 1 1 1 1 1 1 1-------
1 I
v (f)'
O'.:
er:
_J
_J 02 D
( w I
CJ)
I I 2 3 I~ f--_
3 " lj lj ti 4. lj ti lj 0 v
~
I 3
_}
3 3
3
-2
~ _2 ..
2 2 2 2 2 2 -
I ><
z:
I
z:
-4 80 100 120 140 I
(1 20. 40 60 160 180 200 --.J
--.J TIME. SEC I FIGURE 3.95 REACTIVITY FEEDBACK, SEQ. 3 96 2 4 JUN 11 10 ' 4 0 1 ti 3 STEAMLINE BREAK FROM HOT STAND BY
I I -137- XN-NF-77-18 I
3.9 SINGLE ROD WITHDRAWAL I In the unlikely event of the withdrawal of a single control rod, I the system response would be similar to that of a withdrawal of a control rod group. Localized radial peaking would be more severe, however, due to I
the asymmetric rod pattern which would result. With the reactor regulating system in the manual mode, power would increase until the reactor either tripped or the rod was fully withdrawn. In the automatic mode the regulat-I ing system would insert control rods to compensate for the withdrawn rod and power level and all other system parameters would remain relatively I constant. Only the most limiting case., i.e., the case with the system in the manual mode, is analyzed in this section. As in the case of the with-I drawal of one or more control rod banks, the high neutron flux, thermal I margin, and high pressurizer pressure reactor trip signals provide protection for the event.
I Two cases of withdrawal of a single control rod from 102% of rated power (2530 MWt) with the reactor regulating system in the manual mode were I analyzed; one assumed the largest negative values for Doppler and moderator coefficients i.e., -1.66 x 10- 5 Ap/°F and -3.5 x 10- 4 Ap/°F respectively; I and one assumed the least negative values for Doppler and moderator co-I efficients i.e., -.87 x 10- 5 Ap/°F and 0.5 x 10 -4 Ap/°F respectively. For both cases, the design axial peaking factor was assumed while a radial I peaking factor of 1 .6 was assumed. This is the highest radial peaking factor which would occur at any time in cycle life with a single withdrawn I group 4 rod and with the remainder of group 4 at its insertion limit.
I I
r---**
I
-138- XN-NF-77-18 I
I As discussed above, the withdrawal of a single control rod from 102% of rated power with the reactor regulating system in the manual mode I was evaluated assuming both minimum and maximum feedback. As the miriimum feedback case was found to* be most limiting from a DNB standpoint, the I
results of this case are presented here. It was assumed in the analysis that a single group 4 control rod was withdrawn at a rate of 46 inches/
I minute from its full power insertion limit. The rod was assumed to be I worth 0.1% Ap when inserted tG its full power insertion limit. The total worth of the rod is approximately 0.3% Ap assuming full insertion of group 4, I
thus 0.1% Ap represents a conservative upper limit of the worth of a single group 4 rod inserted only 25%. The effect of the pressurizer spray was I
included in the analysis in order to minimize the pressure rise and therefore I minimize MDNBR. Figures 3.96 through 3.100 show the transient response of the system for this case. The reactor trips on high power at 30 seconds I
with an MDNBR of l .44.
I I
I I
I I
I I
PALISADES ** SINGLE CON~ROL ROD WITHDRAWAL +* MINIMUM FEEDBACK 14 0 EVEL II I
2 l POJ..JER '
~EAT Fl ux I 3 TOTAL I RIMARY CO nLANT FLOhl
\ TOTAL I EEOhlATER LOhl 5 TOTAL TEAM LINE FLOW 120 1 1 1 - \
1 ~ £ '
l 2 1 " -
t "
1 .- - <.
'\
., J "< q :J ., q 5 ~ 'I 5 ., q 5 ., q 5 - q 5 'I ., .,
100
\
I\
0 i...L1 l-a::
~
I Ct'.:8 0 LL 0
1-z
~so
\ I\.
I w
l.O I
cl~
I O'.: !
w
()__
40 \,,.,.... v "' "
20
\ ,vv.~AVfVt--- ~
x z
I z
0 0 I 32 40 "TJ I
-....J 8 12 lb 20 24 28 3b -....J I
TIME. SEC FIGURE 3.96 POWER, HEAT FLUX, AND SYSTEM FLOWS, SEQ. ll'F23 .JJN 7?'1J4*22,IJ3
. SINGLE CONTROL ROD WITHDRAWAL
PALISADES ** SINGLf CONTROL ROD WITHDRAWAL ** MINIMUM FEEDBACK 180 u l. AVE. FUEL TEMPER ~'.i.JRE 2 CORE 11µLET TEMPE ~ATURE
I 3. AVE. C!DRE COOLAN TEMP.
I CLAD Tl MPERATURE 180 PJ '
1 1
1 l ~
l 1 HO PJ 1 1
\
Li..
120 u ,
\ \
I
__,I
(.9 w
0 100 PJ I '
I\ ~
0 I
"'~
I 800 I
I 800 3
'l 3
tj 3
tj
.j tj
~
tj
.j tj
.j
" . .j
" - 3 z z ;: -41 z
£ L z £ x
z I
z l
I I "Tl I
J. 12 lG 20 28 32 - 'f 0 ........
I TIME. SEC CXJ FIGURE 3.97 CORE TEMPERATURE RESPONSES, SEQ. 1 18? 2.3 ,JUN 77 0 'I ' 2 2 : IJ 4 SINGLE CONTROL ROD WITHORA~iAL
PALISADES ++ SINGLE CONTROL ROD NITHDRAwAL ** MINIMUM FEEDBACK 30 1 CHANG IN AVE. 'RIMARY C01 DLANT TEMP. LOOP 1 I 2.
3 CHANG CHANG IN IN AVE. I 'RIHARY COi OLANT TEMP. LOOP 2 HOT L G - COLD I EG TEMP. IHFFERENCE LOOP 1 ii. CHANG IN HOT L G - COLD I EG TEMP. HFFERENCE LOOP 2 20 1 2 10 1 2 1 2 1 2 1 2 1 2 - .. 1 2 3 4 3 4 3 4 3 4 3 4 0
i-101-~~-t-~~---+-~~---+-~~___,~~~-t-~~---+-~~-+-~~----;-~~t---+-~~--i 1*
-2 0 ----------+--------~---*- - - - - - - - + - - - * * - - **-------*---+--------+---
-3 o i - * - - - t - - - - - - - - - - - - - - - - * - * - - - - - - - - - - - - - ---**--------- * - - - - t - - - - ~
z I
z
- 4 ~ -------4~--- *-*** ***- _8 ___________ i 2-* ---*-18--------20------24-------- --28------*-3 2*** -----3-6. 40 I
-.....J I
TIME. SEC FIGURE 3.98 PRIMARY LOOP TEMPERATURE CHANGES, SEQ. 1187 23 JUN 77 04: 22: 06 SINGLE CONTROL ROD WITHDRAWAL
PALISADES ** SINGLE CONTROL ROD WITHDRAWAL ** MINIMUM FEEDBACK 350 1 STEAM IDOME PRESSI URE CHANGE LOOP l 2 STEAM IDOME PRESS IJRE CHANGE LOOP 2 3 PRESSUI RIZER PRES ~URE CHANGI 300 II ~
250
/
200 I
a:
(.!)
CL 150 I
I N I
+:>
I 3 3 ii_,.,.--- ---....
100 50
~
- ~
~
~
---3---
__-2--
l 2 ---
U3.---- lJ I ><
z I
1 2 3 ~ LH--1-- 1 2 1 2 i. 2 I .,
z I
0 c 12 16 20 24 28 32 -...J 8 3S 40 -...J I
TIME, SEC F1GURE 3.99 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 118 'i' 2 3 .JUN ? ? 0 4 '2 2 : 0 7 STEAM GENERATORS, SINGLE CONTROL ROD WITHDRAl~AL
30 PALISADES ++ SINGLE CONTROL ROD WITHDRAWAL +* MINIMUM FEEDBACK
1. CHANGE IN STEAM iEN. ~ATER LEVEL. LO JP
2. CHANGE IN STEAM EN. ~ATER LEVEL. LOI lP 2 1
II i
3. CHANGE !N PRESSUr!ZER WATE ~ LEVEL I I
20 i....-----
.j I
]--"
~
10 ......
~
___3---
~
1 2 1 2 ~ 1~ 1 2 1 2 1 9 ' ~
3 "' 1 'l
\\
- I
.p.
w I
-10
-20
\
-30 I
I
\\ l 2 -
x z
I z...,.,
I 8 12 16 20* 24 28 32 36 40 -....J
-....J I
TIME. SEC FIGURE 3. 100 LEVEL CHANGES IN PRESSURIZER AND SEQ. 1187' 23 JUN 77 Qq 122 rQ9 STEAM GENERATORS, SINGLE CONTROL ROD
\~ITHDRAWAL
I
-144- XN-NF-77-18 I
3.10 ROD EJECTION INCIDENT I
Hypothetical control rod ejection incidents for the Palisades I core were analyzed with the XTRAN digital computer code( 2 ), a two-dimensional (r-z) program which solves the space and time dependent neutron I
. diffusion equation with fuel temperature and moderator density reactivity feedbacks. Hot full power (HFP) and hot zero power (HZP) core conditions I
in combination with beginning-of-cycle (BOC) and end-of-cycle (EOC) condi- I tions were considered in determining and analyzing the most limiting hypothetical rod ejection incidents. The criteria to be satisfied for I the rod ejection incident are l) that the energy deposition in the fuel be
~ 280 cal/gm and 2) that the peak system pressure be less than the vessel I
design pressure (~ 2750 psia). I To assure the conservatism of the ejected rod worths and power peaking factors utilized in these analyses these parameters were calculated I by considering the ejection of a control rod from a control rod configuration containing a fully-inserted (dropped) rod in addition to the control rods I
intended to be inserted at the power level under consideration. In all I cases, the dropped rod was assumed to be in the core quadrant diagonally opposed to the ejected rod, thus amplifying the reactivity insertion and I peaking factors associated with the postulated rod ejection. The ejected rod worths and power peaking factors thus calculated by the XTG core I
simulator code(g) are shown in Table 3.2. The ejected rod worths and I I
I I
-145- XN-NF-77-18 power peaking factors at HZP are maximum at BOC, and HFP worth and peaking are maximum at EOC. Accordingly, the rod ejection incident was analyzed I at BOC-HZP and EOC-HFP conditions.
Values of the significant independent variables utilized in th~
I I .
analyses are given in Table 3.3 .. Each of the two major cases, BOC~HZP and EOC-HFP, were analyzed twice: a) assuming heat transfer from the I fuel to the moderator, and b) assuming no heat transfer from the fuel I (adiabatic conditions).
The results of the analyses are summarized in TaQle 3.4. With I ,_.
very conservative values assumed for roq worths an.d peaking factors, both the full- and zero-power rod ejection incidents have been shown to have*
I acceptable consequences in that no fuel pellets realize an average enthalpy I greater than 280 cal/gm. The maximum pellet enthalpy. is less than 250 cal/gm.
The total energy ~roduced during the first 4.92 seconds (at which I time the transient has been terminated) of the HZP transient is 10,950 MW-sec.
The core pressure surge resulting from this energy release is calculated to I be less than 200 psi. Since nominal system pressure is 2060 psia, the I primary coolant system would not be overstressed and the pressurizer relief*
valve (set at 2400 psia) would not open during a rod ejection incident.
I Conclusion The above results are bounded by the results of the rod ejection I accident analyzed in Cycle I, and are therefore acceptable. No gross center A fuel pellet I line melting of the fuel is expected following rod ejection.
enthalpy of 250 cal/gm is not exceeded.
I I
I XN-NF-77-18
-146-I I
TABLE 3.2 MAXIMUM EJECTED ROD WORTHS AND PEAKING FACTORS* I BOC HZP**
EOC BOC HFP***
EOC I
Ejected Rod Worth, % Ap 1.02 0.95 0. 17 0. 24 I Maximum Power Peaking Factor, F~ 12 .86 l 0. 56 3. 16 3.47 I I
I I
All calculations considered the presence of a dropped rod across the core from the ejected rod. I
Calculated with regulating rod Groups 4, 3 and 2 fully-inserted; Group 2 is actually limited to ~40% insertion at HZP by the Tech.
Spec. power dependent insertion limits. I
"*** Calculated with regulating rod Group 4 fully-inserted; Group 4 is actually limited to 25% insertion at HFP by the Tech. Spec.
power dependent insertion limits.
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1 I -147- XN-NF-77-18 I
TABLE 3.3 I ROD EJECTION INCIDENT ANALYSES PARAMETERS I
BOC-HZP EOC-HFP I Initial Power Level, Mwt l 2580.6 Max. Ejected Rod Worth, % ~P I Delayed Neutron Fraction 1.24 0.0060 0.60 0.0045 I Fuel-to-Clad Gap Heat Tr~nsfer Coefficient, BTU/hr-ft -°F 500/0* 500/0*
I Film Heat T2ansfer Coefficient, BTU/hr-ft -°F , . 10,000 10,000 Clad Thermal Conductivity, I BTU/hr-ft -°F 9.39 9.39 Clad Heat Capacity, BTU/lb - °F 0.0773 0.0773 I Doppler Coefficient Multiplier 0.80 0.'80 I
I
Cases run with and without heat transfer.
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-148- XN-NF-77-18 I
I TABLE 3.4 ROD EJECTION INCIDENT RESULTS I BOC-HZP EOC-HFP I Non,... Non-Max. F~ (after ejection)
Adiabatic Adiapatic 13.48 13.48 Adiabatic Adiabatic
6. 77 6. 77 I
Peak Reactor Power Normalized to 2530 MW 158 158 43 39 I
Average Enthalpy of Hottest Fuel Pell e.t, cal /gm 215 247 170 200 I I I I
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4.0 REFERENCES
I 1. Kahn, J. D., Descri tion of the Exxon Nuclear Plant Transient Simulation Model for Pressurixed Water Reactors PTSPWR ,
I 2.
Exxon .Nuclear Company, XN-74-5, Revision l, May 1975.
Morgan, J. N., XTRAN-PWR: A Computer Code for the Calculation of Rapid Transients in PWR 1 s with Moderator and Fuel Temperature I Feedback, XN-CC-32, September, 1975.
3. Tong, L. S., Boilin Crisis and Critical Heat Flux, USAEC Office I of Information Services TIB- 8 97
4. Hughes, E. D., A Correlation of Rod Bundle Critical Heat Flux I For Water in the Pressure Rang~ 150 to 725 psia, Idaho Nuclear Corporation, IN-1412, July 1970.
5. Galbraith, K. P. and Patten, T. W., XCOBRA-IIIC: A Computer Code I To Determine the Distribution of Coolant During Steady-State and
}ransient Core Operation, Exxon Nuclear Company, XN-75-21, April 1975.
I 6. Galbraith, K. P., et al., Definition and Justification of Exxon Nuclear Company DNB Correlation for PWR's, XN-57-48 (October 1975).
I 7. WREM: Water Reactor Evaluation Model (Revision l),NUREG-75/056, May 1975 ..
I 8. C-E/EPRI, Two-Phase Pump Performance Program, Quarterly Technical Progress Report No. 1, January 1 to April l, 1975, C-E Power System, Combustion Engineering, Inc., Windsor, Connecticut.
I 9. Stout, R. B., XTG: A Two-Grau Three Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing and Users Manual PWR Version ,
XN-CC-28, Rev/ 4, July, 1976.
I 10. Moody, E. T., Maximum Flow Rate of a Single Component, Two Phase Mixture, Transactions of the ASME, Journal of Heat Transfer, I February 1965.
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I I DISTRIBUTION G. R. Correll I R. H. Ehlers K. P. Galbraith J. D. Kahn R. H. Kelley I G. E. Koester
T. L. Krysinski
c. E. Leach I G. F. Owsley G. A. Sofer
w. s. Nechodom I CPCo/R. H. Ehlers ( l 00)
I NRC/G. F. Owsley (50)
I Document Control(lO)
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ML18347B304

Contents

Text

1.0 INTRODUCTION

SUMMARY

4.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

SUMMARY

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4.0 REFERENCES

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ML18347B304

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1.0 INTRODUCTION

SUMMARY

4.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

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SUMMARY

4.0 REFERENCES

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