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 I

I

1 XN-NF-77-18 PLANT TRANSIENT ANALYSIS OF THE PALISADES REACTOR FOR OPERATION AT 2530 Mwt JULY 1977 j

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

I I

I I,,

I I

-~

I I

I.

Issue Date:

07/18/77 PLANT TRANSIENT ANALYSIS OF THE PALISADES REACTOR FOR OPERATION AT 2530 Mwt JULY 1977 Prepared By G. E. Koester J. D. Kahn G. R. Correll D. J. VandeWalle (CPCo)

~1---=-"-~~L-J_, *--,_;_.-. --;'-,/,/,1,1_,!.1/l Approved:) e,/<-£*1.:>*,.-..._ /-r-r

//./

I /

Approved:

K. P.

G~1braith, Manger Nuclear Safety Engineering

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

~k~

I

, H/71 Approved: ~.--uh. ci~i-.____...-------. / /

W. S. Nechodom, Manager Licensing and Compliance XN-NF-77-18 EJ${0N NUCLEAR COMPANY, Inc.

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

Without derogating from the foregoing, neither Exxo.n Nuclear nor any person acting on its behalf:

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 will not infringe privately owned right~.

B.

Assumes any liabilities* with respect to the use of, or for damages resulting fro~ the use of, any information, apparatus, method, or process disclosed in this document.

I II I

I I

- I a

I I

I

.I

I I

I*

I I

.I I

I I

I,,

I I

I 1*

I I.

TABLE OF CONTENTS

1.0 INTRODUCTION

AND

SUMMARY

2.0 CALCULATIONAL METHODS AND INPUT PARAMETERS.

3.0 TRANSIENT ANALYSES........

. 3. l UNCONTROLLED ROD

~JITHDRAWAL.

XN-NF-77-18 Page l

5 17 17 3.1.l Control Rod Withdrawals From 102% of Rated Power.

18 3.1.2 Control Rod Withdrawals From 52% of Rated Power 19 3.2 CONTROL ROD DROP INCIDENT...

43 3.3 LOSS OF COOLANT FLOW INCIDENTS 51 3.3.l Four-Pump Coastdown 3.3.2 Locked Rotor....

3.4 EXCESSIVE FEEDWATER INCIDENTS.

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

3. 5 EXCESSIVE LOAD INCREASE INCIDENT....

3.5. l Excessive Load From 102% of Stretch Power 3.5.2 Excessive Load From Hot Standby

3. 6 LOSS OF LOAD INCIDENT.

3.7 LOSS OF FEEDWATER FLOW INCIDENT.

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

3.8.l Steam Line Break From 102% 6f Rated Power (2530 Mwt)............

3.8.2 Steam Line Break From Hot Standby 52 53 66 66 67 78 79 79 92 111 118 120 120

TABLE OF CONTENTS (Conti n~.ed)

3. 9 SINGLE ROD WITHDRAWAL.

3.10 ROD EJECTION INCIDENT.

4.0 REFERENCES

ii XN-NF-77-18 Page 137 144 149 I

I

I I

I I

I I,,

,I I

I I

.I

I.'

I I

I I

I I

I I.

I I

I I*

I I

Table

1. 1

2. 1 2.2 2.3 2.4

3. 1 3.2 3.3 3.4 LIST OF TABLES

SUMMARY

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

NOMINAL OPERATING PARAMETERS USED IN PTSPWR2 ANALYSIS OF PALISADES AT 2530 Mwt PALISADES FUEL DESIGN PARAMETERS EXXON NUCLEAR FUEL KINETICS PARAMETERS...............

TRIP SETPOINTS FOR OPERATION OF PALISADES REACTOR AT 2530 Mwt..........

LOSS OF LOAD TRANSIENT RESULTS.........

MAXIMUM EJECTED ROD WORTHS AND PEAKING FACTORS.

ROD EJECTION INCIDENT ANALYSES PARAMETERS ROD EJECTION INCIDENT RESULTS.*....

iii XN-NF-77-18 Pag~

3 8

9 10 11 95 146 147' 148

I I

I II, I

I I

1 I

I.

I I

~'

I*

I,,

I

'I Figure

2. 1 2.2.

2.3 2.4 2.5

3. 1 3.2 3.3 3.4 3.5 3.6

3. 7 '

3.8 3.9 XN-NF-77-18 LIST OF FIGURES PTSPWR2 SCHEMATIC FOR PALISADES.

AXIAL POWER PROFILE FOR 102% OF STRETCH POWER OPERATION.. *..

AXIAL POWER PROFILE FOR 52% OF STRETCH POWER OPERATION....

PALISADES THERMAL MARGIN LIMITING OPERATING CONDITIONS {BUNDLE POWER, INLET TEMPERATURE, AND PRIMARY PRESSURES) 100% POWER = 2530 Mwt PALISADES SCRAM CURVE............

CONTROL ROD l~ITHDRAWAL INCIDENT REACTIVITY ADDITION RATE VS MDNB'R INITIAL POWER LEVEL = 2580. 6 Mwt.

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

@ 1.4 X lo-4 Ap/SEC..........

CORE TEMPERATURE RESPONSES,4CONTROL ROD WITHDRAWAL FROM 105% POWER@ 1.4 X 10-Ap/SEC........

PRIMARY LOOP TEMPERATURE CH~NGES, CONTROL ROD WITHDRAWAL FROM 105% POWER @ 1. 4 X 10-Ap/SEC..'......

~..

PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, CONTROL ROD WITHDRAWAL FROM 102% POWER@ 1.4 X 10-4 Ap/SEC LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, CONTROL ROD WITHDRAWAL FROM 105% POWER@ 1.4 X lo-4 Ap/SEC POWER, HEAT fLUX, AND SYSTEM FLOWS, CONTROL ROD WITHDRAWAL FROM 102% POWER@ 1.0 X lo-5 Ap/SEC CORE TEMPERATURE RESPONSES, 5CONTROL ROD WITHDRAWAL FROM 102% POWER@ 1.0 X 10-. Ap/SEC........

iv 12 13 14 15 16 21 22 23 24 25 26 27 28 29

'1,

'*1

Figure 3.10 3.11

3. 12 3.13

3. 14

3. 15

3. 16

3. 17 3.18

3. 19 3.20 3.21 3.22 3.23 XN-NF-77-18 LIST OF FIGURES (Continued)

PRIMARY LOOP TEMPERATURE CHANGES, CONTROL ROD WITHDRAWAL FROM l 02% POWER @ l. 0 X 10-5 tip/SEC.

PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATO~S, CONTROL ROD WITHDRAWAL FROM l 02% POWER @ l. 0 X lo-

tip/SEC.

LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, CONTROL ROD WITHDRAWAL FROM 102% POWER@ 1.0 X lo~5 ~p/SEC.

30 31 32 POWER, HEAT FLUX, AND SYSTEM FLOWS, CONTROL ROD WITHDRAWAL FROM 52% PO~JER @ 6. 0 X 10-4 tip/SEC.. *......

33 CORE TEMPERATURE RESPONSES4 CONTROL ROD WITHDRAWAL FROM 52% POWER @ 6.0 X lo-tip/SEC.........

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

35 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERAT~RS, CONTROL ROD WITHDRAWAL FROM 52% 'POWER @ 6.0 X 10-tip/SEC 36 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS CONTROL ROD WITHDRAWAL FROM 52% POWER @ 6.0 X lo-4 tip/SEC 37 POWER, HEAT FLUX, AND SYSTEM FLOWS, CONTROL ROD WITHDRAWAL FROM 52% POWER @ 6. 0 X l o-5 tip/SEC.

38 CORE TE~PERATURE RESPONSE~S CONTROL ROD WITHDRAWAL FROM 52% POWER @ 6.0 X 10 tip/SEC..............

39 PRIMARY LOOP TEMPERATURE CHANGES, CON~ROL ROD

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

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

45 v

'I I

I I

I 1*

'./

,*' I

.I I

I I

'I I

I I.

I I

1 I

I.

I I

I Figure 3.24 3.25 3.26 3.27 3.28 3.29 3.30 3.31 3.32 3.33 3.34 3.35 3.36 3.37 3.38 3.39 3.40 XN-NF-77-18 LIST OF FIGURES (Continued)

CORE TEMPERATURE RESPONSES, CONTROL ROD DROP..

PRIMARY LOOP TEMPERATURE CHANGES, CONTROL ROD DROP.

PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, CONTROL ROD DROP.

LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATOR, CONTROL ROD DROP..

REACTIVITY FEEDBACK, CONTROL ROD DROP...

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

POWER, HEAT FLUX, AND SYSTEM FLOWS, 4 PUMP TRIP CORE TEMPERATURE RESPONSE, 4 PUMP TRIP...

PRIMARY LOOP TEMPERATURE CHANGES, 4 PUMP TRIP PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, 4 PUMP TRIP..

LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, 4 PUMP TRIP.

POWER, HEAT FLUX, AND SYSTEM FLOWS, LOCKED ROTOR.

CORE TEMPERATURE RESPONSES, LOCKED ROTOR..

PRIMARY LOOP TEMPERATURE CHANGES, LOCKED ROTOR.

PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, 46 47 48 49 50 55 56 57 58 59 60 61 62 63 LOCKED ROTOR.

64 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, LOCKET ROTOR.

65 POWER, HEAT FLUX, AND SYSTEM FLOWS, REDUCTION OF FEEDWATER ENTHALPY.

68 vi

),

Figure 3.41 3.42 3.43 3.44 3.45

.3.46 3.47 3.48 3.49 3.50 3.51 3~52 3.53 3.54 XN-NF-77-18 LIST OF FIGURES (Continued)

CORE TEMPERATURE RESPONSES, REDUCTION OF FEEDWATER ENTHALPY.

~...

69 PRIMARY LOOP TEMPERATURE CHANGES, REDUCTION OF FEEDWATER ENTHALPY...............

70 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, REDUCTION OF FEEDWATER ENTHALPY..........

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

72 POWER, HEAT FLUX, AND SYSTEM FLOWS, INCREASE IN FEEDWATER FLOW AT 52% POWER..........

CORE TEMPERATURE RESPONSES, INCREASE IN FEEDWATER FLOW AT 52% POWER......................

PRIMARY LOOP TEMPERATURE CHANGES, INCREASE IN FEEDWATER FLOW AT 52% POWER...................

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

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

POWER, HEAT FLUX, AND SYSTEM FLOWS, EXCESSIVE LOArr INCREASE......

CORE TEMPERATURE RESPONSES, EXCESSIVE LOAD INCREASE PRIMARY LOOP TEMPERATURE CHANGES, EXCESSIVE LOAD INCREASE..............

PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, EXCESSIVE LOAD INCREASE..............

73 74 75

... 76 77 81 82

... 83

.. 84 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, EXCESSIVE LOAD INCREASE...

..... 85 vii

'I I

I

~-*

I

.I

.1*

I I

I 1

'I I,

.-. I I

Figure I

3.55 I

3.56 3.57 I

I :I 3.58 3.59 I

3.60 I

3.61 I.

3.62 I

3.63

,-*-*. I 3.64 l

3.65 I

3.66 I

3.67 3.68 I

I I

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

Page POWER, HEAT FLUX, AND SYSTEM FLOWS, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY...

86 CORE TEMPERATURE RESPONSES, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY.

87 PRIMARY LOOP TEMPERATURE CHANGES, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY..

88 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATOR, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY.....

.... 89 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY..

..... 90 REACTIVITY FEEDBACK, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY..............

91 POWER, HEAT FLUX, AND SYSTEM FLOWS, LOSS OF LOAD - CASE l........

96 CORE TEMPERATURE RESPONSES, LOSS OF LOAD - CASE l....

97 PRIMARY LOOP TEMPERATURE CHANGES, LOSS OF LOAD - CASE 1.......

98 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, LOSS OF LOAD - CASE l...............

99 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, LOSS OF LOAD - CASE 1...............

POWER, HEAT FLUX, AND SYSTEM FLOWS, 100 LOSS OF LOAD - CASE 2.................... 101 CORE TEMPERATURE RESPONSES, LOSS OF LOAD - CASE 2....

PRIMARY LOOP TEMPERATURE CHANGES, LOSS OF LOAD - CASE 2...

viii

............. 102

............. 103

I 1,

I.

Figure I

3.83 3.84 I

3.85 I

3.86 3.87 I

3.88 I

3.89

- ~'\\ I 3.90 I

3.91 3.92 I

J*~

3.93 I

3.94

. I 3.95 1*

3.96 I

3.97 I

I I

/

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

HIGH PRESSURE SAFETY INJECTION FLOW VS PRESSURE POWER, HEAT FLUX, AND SYSTEM FLOWS, STEAM LINE BREAK..............

CORE TEMPERATURE RESPONSE, STEAM LINE BREAK PRIMARY LOOP TEMPERATURE CHANGES, STEAM LINE BREAK.

PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, STEAM LINE BREAK..................

. LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS,

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

124 125 126 127 128 129 130 REACTIVITY FEEDBACK, STEAM LINE BREAK POWER, HEAT FLUX, AND SYSTEM FLOWS, STEAML INE BREAK FROM HOT STAND by..

...... 131 CORE TEMPERATURE RESPONSES, STEAMLINE BREAK FROM HOT STANDBY...................

PRIMARY LOOP TEMPERATURE CHANGES, STEAMLINE BREAK FROM HOT STAND by..........

PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, STEAMLINE BREAK FROM HOT STAND BY.........

LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, STEAMLINE BREAK FROM HOT STAND BY.........

REACTIVITY FEEDBACK, STEAMLINE BREAK FROM HOT STAND BY POWER, HEAT FLUX, AND SYSTEM FLOWS, SINGLE CONTROL ROD WITHDRAWAL CORE TEMPERATURE RESPONSES, SINGLE CONTROL ROD WITHDRAWAL..............

x 132 133 134 135 136 139 140

Figure 3.98 3.99

3. 100 LIST OF FIGURES (Continued)

PRIMARY LOOP TEMPERATURE CHANGES, SINGLE CONTROL ROD WITHDRAWAL............

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

LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, SINGLE CONTROL ROD WITHDRAWAL...........

xi XN-NF-77-18

.... 141 142 143 I

I

,1 I

I.

I I

I I,,

I I

I,,

I I

I 1-1--

XN-NF-77-18

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:

l.

2.

3.

4.

5.

6.

7.

8.

Control Rod Withdrawal Incidents Control Rod Drop Incident Loss of Coolant Flow Incidents

- Four Pump Coast Down

- Locked Rotor Excessive Feedwater Incidents Excessive Load Increase Incidents Loss of Load Incidents Loss of Feedwater Flow Incident Steam Line Break Incidents Incident*

Classification I I II III IV II II I I II IV

9.

Uncontrolled Withdrawal of an Individual Control Rod II I

10.

Control Rod Ejection Incident IV The plant transients 1 through 9 were analyzed using the Exxon Nuclear Company plant transient simulation code PTSPWR2. (l) The control rod ejection incident was analyzed using XTRAN. (2) The criteria to be satisfied for the Class II and III incidents are:

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

,- i) peak primary pressure~ 2750 psia ii)

MDNBR > 1.30 xrl-NF-77-18 iii) primary to secondary differential pressure, ~P ~ 1530 psid.

The criterion of concern for the steam line break is that the end-of-cycle (EOC) shutdown margin shall be adequate to avoid excessive (>1%)

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 (to ensure no.fuel meltdown) and 2) that the peak system pressure be less than the vess~l design pressure(~ 2750 psia).

The results of the analyses are summarized in Table 1.1.

The lowest 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 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.

In all other cases, there is at least 95% probability with 95% confidence that no fuel rod in the core will experience DNB.

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

The above analysis is valid for a maximum power peaking, F6, of 2.55 and an axial power peaking factor of 1.40, with the axial peak located at X/L <

60.

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

I I

I

~-. I I

I I

TABLE l. l

SUMMARY

OF RESULTS Maximum Maximum Maximum Primary-Maximum Core Average Pressurizer Secondary MDNBRfi Transient - Class.

Po\\'1er Level Heat Flux Pressure LIP (Mwt)

(Btu/hr:n2)

(psi a)

(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

-4 w

Rod Withdrawal @ 6.0 x 10 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 :z:

Entha 1 PY - II 2590 169,910 2019 1272 1. 75 I

-....J Increased Feedwater Flow

-....J I

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

Transient - Class Excessive Load From 102% Power - II From Hot Standby - II Loss of Load - II Loss of Feedwater - II Steam Line Break From 102% Power - IV From Hot Standby - IV Uncontrolled Withdrawal of an Individual Control Rod-III Control Rod Ejection Incident-IV Maximum Power Level (Mwt) 2870 258 2838 2673 464 694 2841 399,740 TABLE l. l (Continued)

Maximum Core Average Heat Flux (Btu/hr*ft2) 178,780 17,075 176,415 172,905 30,960tt 45,530 182,515 1111 Maximum Pressurizer Pressure (psia) 2394 2162 2125 2260 Maximum Primary-Seconda ry 6P (psid) 1287 1363 1388 1238 1297 t +

I' 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.

MDNBR~

1.74 3.60t

1. 39

1. 65

1. 30t

1. 41 t

1. 44

+tt I

+:>

I

z I ::z

"'Tl I

-.....J

. I __,

co

- - - **- - - -*-* ~:- - - ~ - ~ - - -

I I

I XN-NF-77-18 2.0 CALCULATIONAL METHODS AND INPUT PARAMETERS

~ -

The transient analyses for operation of the Palisades reactor.at 2530 Mwt were performed using the Exxon Nuclear Plant Transient Simulation model for Pr~ssurized Water Reactors (PTSPWR2)_(l)

The PTSPWR2 code is a digital computer program developed to describe the behavior of pressurized water reactors subjected to abnormal operating conditions.

The model is based on the solution of the basic transient conservation equations for the primary and secondary coolant system, on the transient conduction equation for the fuel rods, and on the point kinetics equation for the core neutronics.

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

Various control and safety system components are included as necessary to analyze desired transients. A hot channel model is used to evaluate the departure from nucleate boiling ration (DNBR) during transients.

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 Barnett CHF correlation(4) are used to predict DNB or CHF depending on the system conditions.

The models contained within PTSPWR2 code are described in detail in Reference 1.

A block diagram of the PTSPWR2 model is depicted in Figure 2.1.

. For these analyses, the core parameters calculated using the PTSPWR2

/

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 used to compute DNB heat fluxes at primary system pressures above 1000 psia

r-XN-NF-77-18 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-mined by averaging the critical heat flux determined by both correlations.

The initial conditions for the transient analyses are based on steady-state operations at 2530 Mwt (excluding pumping power) with uncertainties applied to ensure a conservative*analysis; i.e. minimize DNBR, ma~imize system pressure, and maximize pressure differential between primary and secondary:

Reactor Power Average Core Inlet Coolant Temperature Primary Coolant System Pressure 2530 + 2% Mwt 537.5 + 5°F 2060 ~ 50 psia 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 are given in Table 2.2.

The kineti~s parameters for beginning-of-cycle (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 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 conservative effect during a particular transient was evaluated.

The set of kinetics parameters used for each transient case is described in the sec-tion dealing with the representative transient.

A skewed axial power profile (Fz = 1.4 at Z/L = 0.6) was used for all 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 I

1 I

I I

I*

I I

I I XN-NF-77-18 analyses from 52% of rated power.

The axial power profiles used are depicted in Figures 2.2 and 2.3.

Unless otherwise indicated, the total peaking applied in the analyses T

T was FQ = 2.55 for the analyses from 102% of rated power and FQ = 3.74 for 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 (2530 Mwt), respectively.

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 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-nical specification requirement of a minimum of 90% of full insertion at 2.5 seconds.

Parameters dependent on transient type are discussed in each transient description ?ection.

L Core TABLE 2.1 NOMINAL OPERATING PARAMETERS USED IN PTSPWR2 ANALYSIS OF PALISADES AT 2530 Mwt Total Core Heat Output, MWt, Total Core Heat Output, Btu/hr Heat Generated in Fuel,%

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

Core Inlet Coolant Temperature, °F Average Core Coolant Temperature, °F Hot Channel Factors:

Total Peaking Factor, F~

Radial Peaking Factor, Fr Heat Transfer 2

Core Average Heat Flux, Btu/hr-ft Steam Generators Total Steam Flow, Mlbs/hr Secondary Steam Pressure, psia Feedwater Temperature, °F Number of Active Steam Generator Tubes, S.G.#1 S.G.#2

Includes leakage and flow measurement uncertainty XN-NF-77 2530 8633.4 x 106 97.5 2060 121. 7 114. 5 537.5 565.

2.55 l.45 168 '180 l 0. 97 730 435.

6590 6775 I

I I

I 11 I

I I

I

I I

I TABLE 2.2 PALISADES FUEL DESIGN PARAMETERS EXXON NUCLEAR FUEL Fuel Radius Inner Clad Radius Outer Clad Radius Active Length Active Fuel Rods Per Bundle

0. 175 inches 0.179 inches 0.207 inches 131.8 inches 208 XN-NF-77-18

I XN-N F-77-18 I

TABLE 2.3 I

KINETICS PARAMETERS' I

_?ymbol Parameter Value I

Beginning of End of Cycle Cycle I

aM Moderator Coefficient I

(L\\p/°F) x 1 o4

+ 0.50

- 3.50 aD Doppler Coefficient I

( L\\p/ o F) x 105

- 1.09

- 1.38 ap Pressure Coefficient I

(L\\p/psia) x 106

- 1.00

+ 7.00 I

av Moderator Density I

Coefficient (L\\p/lb/ft3) x 103

-.962 6.413 I

aB Boron Worth Coefficient 4

(L\\p/ppm) x 10 )

- 0.80

- 1. 00 I

13eff 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 I

- 2.0% at Hot standby I

I I

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

/

TABLE 2.4 TRIP SETPOINTS FOR OPERATION OF PALISADES REACTOR AT 2530 Mwt Setpoint Uncertainty Used in Analysis High Neutron Flux 106.5%

+ 5. 5%

112.o~;

Low Reactor Coolant Flow 955;

+ 2.0%

93.0%

High Pressurizer Pressure 2255 psi a

+ 22 psi 2277 psi a Low Pressurizer Pressure 1750 psi a

+ 22 psi 1728 psi a Low Steam Generator Pressure 500 psia

+ 22 psi 478 psi a Low Steam Generator Level*

6 feet

+10 in 6 feet 10 in.

Therma 1 Margin**

p = f{TH,TC}

+165 psi p - 165 psi a

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 (T8) and the average cold leg temperature (TC).

The functional relationship of the -variables is derived from Figure 2.4.

Delay Time 0.4 sec 0.6 sec 0.6 sec 0.6 sec 0.6 sec 0.6 sec 0.6 sec I __,

I

z I :z.,

I

-...J

-...J I

co

FEED WATER l\\J\\/'v

'0000' PUMP 2B FLOW RES I STANCE ENTHALPY ANO BORON CONCENTRATION TIME OELAY AIMOSl'HElllC Ul!MP SAf[IY VAi VES VALVLS NOil! I SI 11 NOil!

\\Sill.Al \\ill!

VAi.VF 00 2 STEAM OOME

~ll l\\M llf 1'11! !{

TLJIH!llH. SfOI' VALVL l!VIWi'.1 VAi VL I 'i01 Al ION VAi VI A1MU'51'11l t~ J l.

AJ LT Y IJI: MP VALV[';

VAi VLS DO I SEPARATOR'*

ANO OHYrns Si' I II XN-NF-77-18 RrL !Fr VAi.Vi S SAITTY V.~L VI;

__ 0

_11[11 J

WMll' SG 2 OUTLET PLENUM SG 2 INLET PLENUM l'tlt11' 21\\

SPRAY PH l'IIrSSUHI II II UP UPPER PLENUM CP CORE II' 1.ilWI II l'Ll,NllM SG l INLET PLENUM

',I, I SG l OUTLET PLENUM PUMP If\\

PllMI' 111 FIGURE 2.1 PTSP~JR2 SCHEMATIC FOR Pi\\LISADES I

I I

I

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

s...

0

.µ u

'° LL.

01 c:

r-

~

IB l.O 0..

s...

Q) 3 0

0..

r-x c::(

. 5 o,,_~.......... ~..-.ii....-~_._~__.---~.....1-~~~_.::.,.J..::,.;..:;;...;,;;.::.;;.i_:.:....;....;.;,.;.;:i;.;;:.;;~.;;;1 0

. l

.2

. 3

.4

. 5

.6

.1

.8 Bottom Z/L (Non-Dimensional Axial Distance)

(Including Unheated Length)

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

.9 1.0 Top I __,

w I

2 I

2 "Tl I

-....J

-..J I

co

2.0 N

LL s....

/

0 +'

u

'° LL O')

c:

r-

~

'° aJ 0..

s....

aJ 3:

0 0..

'°

r-x c::(

.5

.2

. 3

.4

.5

.6

.7

.8 Z/L (Non-Dimensional Axial Distance)

(Including. Unheated Length)

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

. 9 1.0 Top I.......

+:>

I z

I z,.,

I

-....J

-....J I

co

- - - - - - - *- I- - - - - -.. -I - - -* -

I I

I u..

0

~

QJ

i

..µ

'° QJ CL E

QJ I--

..µ QJ c::

QJ 0

u XN-NF-77-13 550

% Maximum Bundle Power FIGURE 2.4 PALISADES THERMAL MARGIN LIMITING OPERATING CONDITIONS (BUNDLE POWER, INLET TEMPERATURE, AND PRIMARY PRESSURES) 100% POWER = 2530 Mwt

100

...c

+-'

s...

0

~

80

µ 60 u

!tl QJ er:

+-'

c:

QJ u

s...

QJ 0...

20 20 40 60 Percent Control Rod Insertion FIGURE 2.5 PALISADES SCRAM CURVE I __,

O"I I

z I ::z "Tl I

I co

I I

I I,,

. 1 I

I I

I I XN-NF-77-18 3.0 TRANSIENT ANALYSES

3. l UNCONTROLLED ROD WITHDRAWAL

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

The withdrawal of more than one control rod due to an operator error or a reactor regulating system or rod drive control system malfunction causes an increase in both core power level and core heat flux.

Since the heat extraction from the steam generators remains relatively constant until actuation of the steamline safety valves, an increase in primary coolant tern-perature results.

Unless terminated by either manual or automatic action, the control rod withdrawal would eventually result in a minimum DNB 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 against DNB in this event is provided by the high neutron flux, thermal margin/low pressure, and high pressurizer pressure reactor trips.

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

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

(1)

Control rod withdrawal as a function of withdrawal rate from full power (102%), and, (2)

Control rod withdrawal as a function of withdrawal rate from 52% rated power conditions.

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

XN-NF-77-18 3.1.l Control Rod Withdrawals From 102% of Rated Power The uncontrolled rod withdrawal transients from 102% of rated power were analyzed with rod withdrawal rates up to an equivalent of

-5 30 x 10 Ap/sec.

To envelope the operation during a cycle, the analyses was made assuming both:

1)

Maximum reactivity feedback, assuming the largest negative values for Doppler and moderator coefficients -

corresponding to EOC condition,_ i.e., -1.66 x 10-5 Ap/°F and -3.5 x 10-4 Ap/°F respectively, and

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

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

The result of the analysis from 102% of rated power is shown 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 and minimum feedback.

The protection system function which terminated the transient varies with reactivity insertion rate as indicated in Figure 3.1.

For all reactivity insertion rates, MDNBR is greater than l.30.

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

I I

I I.

I I

I

I I

I

I I

I I

I

'I 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:

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

where P = Power level in fraction of rated.

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

A maximum inserted rod worth of 1.5% ~P was assumed which is conservatively high considering the control rod insertion limits.

The analysis was made for both maximum feedback and minimum feedback.

As in the full load case, the pressurizer sprays were assumed operable to conservatively reduce the margin to DNB.

Results for this case are shown in Figure 3.2 which plots minimum transient DNB ratio versus reactivity insertion rate for both maximum and minimum feedback.

The protection system functions which terminate the transient are also indicated in Figure 3.2.

For the maximum feedback (EOC) 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-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

1-* XN-NF-77-18 for a reactivity insertion rate of 60 x 10-5 ~p/sec with minimum feedback, and *in Figures 3.18 through 3.22 for a reactivHy insertion rate of 6 x 10-5 ~p/sec with maximum feedback.

I I

I I.*

I I:

l1 I

I I

I

2.0 1.8 0

.µ ttl

~

ca 1.6 z

Cl E

I E

c:

=-----

E 1.4

1. 2

1. 0.

5 1 x rn-

Minimum Feedback

--- Maximum Feedback Thermal Margin Trip

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

High Neutron Flux Trip 1 x 10-4 Reactivity Addition Rate, ~p/sec FIGURE 3. 1 CONTROL ROD WITHDRAWAL INCIDENT REACTIVITY ADDITION RATE VS MDNBR INITIAL POWER LEVEL = 2580.6 Mwt I

N _.

I x

2 I

2 "Tl I

-....J

-....J I

co

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

0 tO er.

2.0 No Reactor Trip l.8

Minimum Feedback

--- Maximum Feedback

/

c::::

z: 1.6 '-..

/Thermal Margin Trip

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

~ 1.4 l. 2

l. 0 5

l x 10-

/

T.

/

rip

...................... /,,,.\\- ~""-

~

High Pressurizer Pressure Trip l x lO -4 Reactivity Addition Rate, 60/sec FIGURE 3.~ CONTROL ROD

\\flTHDRA\\~AL INCIDENT REACTIVITY ADDITION RATE VS MDNBR INITIAL POWER LEVEL = 1315.6 Mwt I

N N

I

z:

I :z:

"Tl I

-....J

-....J I

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 1

120r--~~-t~~~-r~~~+-~~_.:.5i--_..:...TO~T~A~L~~~T~EA~H.:__.=.L~IN~E~FL~O~W~~\\--~~--i~~~-L-~~--l I

1 1

1 -

100 U:

1--2-'""~*~.;~~0J:

1

'[=;=,4;:;;:5+=11'=~.4=js~K~:~~4~sf=J~C~~q:=-J5f-~£-~----4~5j-~~~~-4-:5r--~~-_~4-s;--::......-;~:\\~~~~

~

0 w f-a:

o::g 0

\\ \\J LL 0

f-z t380 0::

w a..

40 20 O 0'--~---;2¢--~~~5L-~--;?-;f--~~~1~0~~-1~2:--~__Jl~5~~-1~7~~__j2L0~~-2~2~~_J25 TIME. SEC FIGURE 3.3 POWER, HEAT FLUX, AND SYSTEM FLOWS, CONTROL -ROD \\.IITHDRAWAL FROM l 02%

POl~ER

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

SEQ.

1228 23.JUN 77 05 *.13 *03 I

N w I

z I :z "Tl I

-....I

-....I I

00

LL (9

w 0

,PAi_ I SP.DES ++ CONTROL ROD WI THDRANAL FROM 2 5 8 0. 6 MWT ++ 0. 0 0 0 H Lr;/sec 180 180 140 120 100 800 E; 0 0

~ rj I

1.i 11VE.

l="iUEL TEMPER! ~TURE i

CORE I!ULET TEMPEi ~ATURE

~-

i

3.

11\\'E.

C 1DRE COOi...AN TEMP.

I

~-

Ci...RO Tl -MPERATURE I

~

u

i.

1 1

1 l

l l

'-I I

I I I

I~

~

I I

~

u I ! I l

I i I I i

'i

j

'.j

j

j J

J

.j

~

t.

('.*

t.

L I

5

?

10

~2 15 1?

20 TIME. SEC F rs~~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.

I\\ \\

j -

J 22 25 I

N

~

I

z:

I z

"'Tl I

--..J

--..J I

LL (9

w D

16 PAL I SADES ** CONTROL ROD hi I THDRAl-JAL FROM 2 5 8 0. 6. Mh!T ** O. O O O 14 :_:/sec

~j

'-" 1nr,.....

lN AVE.

PRlMAR'f t.:U DLANl 11:.Mt-'

LlJOP 1

12 8

0

-8

-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

/

~

v

~

u...;-

~

l 2---

3 lj 3 q 3 q l 2 3 ti 3 ti 1 2

~..

1 2 3 q 3 q

['\\.

~,,

\\

2 5

?

10 12 15 17 20 22 25 TIME. SEC FIGURE 3. 5 PRIMARY LOOP TEMPERATURE CHANGES, SEQ.

1229 23 JUN 77 O? 133 105 CONTROL ROD4 WITHDRAl>JAL FROM l 02% POWER

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

I N

CJ1 I

z I :z

"'Tl I

-...J

-...J I

a:

Cf) ct.

280 240 200 1 so 120 80 40 0 0 PALISADES +* CONTROL ROD WITHDRAWAL FROM 2580. Ei MWT ++ 0. 00014 tip/sec

i.

STEAM DOME PRESS JRE CHANGE LOOP t 2

STEAM I DOME PRESS JRE CHANGE LOOP 2 3*.

PRESSU! PIZER PRES,URE CHANG!

' I I i-3 __..,,.,

~

v--~ ~;

l.----'

~

c----

~

_Ll-2---

I L-3--

1 2

~

i. '2 i

1 2 l

3 i 2 I

., 2 5

10 12 15 1?

20 22 25 TIME.

SEC 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.

I N

CJ'\\

I z*

I 2..,

I

-...J

-...J I

co

~o 30

20

~

I u z i-;

10 0

-10

-20

-3~

PAL I SP.DES ++ CONTROL ROD WI THORRWAL FROM 2 5 8 0. 6 MWT ++ 0. 0 0 0 14 :.,J/sec 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 3

~

3 -

L--3---

i----

~

1 2 1 2 1 2 3 -

1 2 1 2 1 2 1

~

3 t

\\

\\ \\

2 5

10 12 15 11 20 22 25 TIME, SEC 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.

I N

-....J I

z I :z "Tl I

-.....i

-....J I

PAL I SADES ** CONTROL ROD WITHDRAWAL FROM 2 5 8 0. 8 MWT ** 0. 0 0 0 0 1 :..,c/sec 14 0 I I

i I

120 100 D w l-a:

cr:so LL 0

1--z wso u a:

w Cl..

40 20 0 0 I

I I

I I I I

I l

I I I I I

i I I

1 POWER EVEL 2

HEAT F ux 3

TOTAL PR£MARY CO DLANT FLOW

~

TOTAL EEDWATER LOl-J 5

TOTAL t>TEAM UNE FLOl-J 1

n

~

t I) t,.,

~

l 2 l ')

'.)

lj 5

~ lj 5

~ lj 5 lj 5

,u

~

\\

,,~

"\\

\\\\ \\ '

l 10 20 30 40 50 80 70 80 TIME. SEC FIGURE 3.8 PO\\~ER, HEAT FLUX, AND SYSTEM FLO\\~S, SEQ.

426 23 JUN 7709 1 46*1J2 CONTROL ROD_~ITHDRAWAL FROM l 02~, POWER

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

90 100 i

N co I

z I :z

'Tl I

I co

180 6ALISAOES ++ CONTROL ROD WITHDRAWAL FROM 2580 6 MWT ** o 00001 *0/sec 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

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

1 400,'----~~'~~~~~~~~~~=1::--~---:-1::--~-L~~J__~__L~__J o

10 20 30 40 50 80

?O 80 90 100 TIME. SEC FIGURE 3.9 CORE TEMPERATURE RESPONSES, CONTROL ROD S~ITHDRAWAL FROM 102% POV/ER

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

SEQ.

426 23 JUN i? 09 *'18 *'13 2

I 2,,

I N

'° I

I

'-J

'-J I

CD.

LL

(!J w

0 30 20 10 0

-10

-20

-3 0 PALISADES ** CONTROL ROD WITHDRAWAL FROM 2580. 6 MWT ** 0. 00001llo/sec 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

<I CHANG IN HOT L G - COLO EG TEMP.

bIFFERENCE

,_J ~

LJ_J---~

LD----

1_1---

3 1 2 - ~

3 <I 3 <I 3

11 3 <I

~

\\

10 20 30 40 50 TIME. SEC 80 70 80 90 FIGURE 3.10 PRIMARY LOOP TEMPERATURE CHANGES, CONTROL ROD WITHDRAWAL FROM 102~ POWER

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

SEQ.

~26 23 JUN 71' Oil 11j81t1<1 LOOP 1 LOOP 2 100 I w 0

I z

I z.,.,

I

-....J

-....J I

350 300 250 200 a:

(f)

Q_.

150 100 50 0 0 PALISADES+.-

CONTROL_ ROD WITHDRAWAL FROM 2580. S MWT ++ 0. 00001*:,..:./sec 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

/

I I

I 1

~

/

~

/v v

?/

/

~

11---

l 10 20 30 40 50 60 7' 0 80 90 TIME. SEC 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. 100 I w I

z I z

""Tl I

-...J

-...J I

co

100 80 80 40 en w

I u z

20 PAi_ I SADES +-t CONTROL ROD WITHDRAWAL FROM 2 5 8 0. 8 MWT ** 0. 0 0 0 0 1 !..:;.,/sec 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

)

3/ v

~

J 2 '2 1 2 1

n 0

~

£.

1 2 I I I I

-20 I I I I

' \\

\\ ~

10 20 30 4 0 -

.50 b 0

?O 80 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.

90 100 I w N

I z

I z "Tl I

"'-J

"'-J I

14 0 120 100 0 w f-a:

o::s 0 l.l..

0 f-z

~60 0::

w a....

40 20 0 0 PALISADES ++ CONTROL ROD WITHDRAWAL FROM 1315. 6 MWT ++ 0. 00060

~c/s~c 1

POWt::~ 1.EVEL 2

HEAT Fl ux 3

TOTAL I ~RI MARY CO DLANT -FLOW A

TOTAL EEDWATER LOW TOTAL

~TEAM LlNE FLOW y \\

3 3 -

3 I 3

3 3

I/

D

/ v-I'\\

v v

~

y

~

/

/~

~

q 5

" 5 q 5

\\

4

\\ \\

~

............. r-1-----

5

--2 2

4 6

8

-10 TIME, SEC 12 14 16 FIGURE 3.13 POWER, HEAT FLUX, AND SYSTEM FLOWS, CONTROL ROD WITHDRAWAL FROM 52% POWER

@ 6. o X 10-4 6,:, / SEC.

SEQ.

10116 24 JUN 7? 14: 56: 26 18 I

l I i

i i ' I

~

i i 2 C:

I w w

I

z:

I :z:

"'Tl I

-.....J

-.....J I

LL

(!)

w 0

180 160 14 0 120 100 800 600 PALISADES ++ CONTROL ROD WITHDRAWAL FROM 1315. 6 MWT ++ 0. 00060 /Jo/sec

'-.l 1

AVE.

F llEL TEHPERI HURE 2

CORE I bLET TEMPE! ~ATURE 3

AVE.

cnRE COOLAN TEMP.

lj CLAD T MPERATURE IJ u

/ ~

u v

"'-~

~ ~

U1

<l lj lj

'1

j

j

" 3

~ 3 3

" 3

c.

2 8

10 12 14 16 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.

18

~

I w

+:>

I

z:

I :z:.,

I

-....J

-....J I

LL (9

w 0

12 10 8

6 4

2 0

-2 0 PALISADES ++ CONTROL ROD WITHDRAWAL FROM 1315. 6 MWT ++ 0. 00060 :...:./sec 1 2 2

6 l

2 8

CHANG CHANG CHANG CHANG 10 TIME.

IN AVE.

. LOOP l IN AVE.

RlMARY CO LANT TEMP

LOOP 2 IN HOT L G -

COLD EG TEMP.

lFFERENCE LOOP 1 IN HOT L G -

COLD EG TEMP.

IFFERENCE LOOP 2 12 14 16 18 SEC FIGURE 3.15 PRIMARY LOOP TEMPERATURE CHANGES, CONTROL ROD J"JITHDRAWAL FROM 52% POYJER

@ 6. 0 X 10-tip/SEC.

SEQ.

10 4 6 2 4 JUN ?7 14 : 5 6 : 2 S 20 I w c..n I

z:

I :z:

I

-....J

-....J I

co

er:

CJ)

[L 14 0 120 100 80 60 40 20 0 0 PALISADES ++ CONTROL ROD WITHDRAWAL FROM 13-15. 6 MWT ++ 0. 00060 !:;p/sec 1

STEAM 'OME P:r

~RE CHANGE LOOP I 2

STEAM* I JOME PRES URE CHANGE LOOP 2

3.

PRES SUI LZER PR S' PURE CHANGI

/,~

~/, r II I vi v I

/

~

v lJ 1 2 1 2 3 2

4 6

8 10 12 14 16 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.

18 20 I w O'\\

I x :z I :2:

"'Tl I

......a

......a I

50 40 30 (f)2 0 w

I u z 10 0

-2~

PALISADES *+ CONTROL ROD WITHDRAWAL FROM 1315. 6 MWT ** 0. 00060 t::;/sec l 2 1 2 -

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

2 1 2.,

6

1.

2 3

8 CHANGE [N STEAM 'EN.

WATER LEVEL.

LOl~P i CHANGE IN STEAM

~EN. WATER LEVEL.

LO~P 2 CHANGE lN PRESSU [ZER WATEi LEVEL 10 TIME. SEC 12 14 16 SEQ.

1046 24 JUN 77' 14 :56131 FIGURE 3.17 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, CONTROL ROD WITHDRAWAL FROM 52% POWER @ 6. 0 X 1 o-4 lip/SEC 18 20 I w I

x

z:

I ::z:,,

I

-..J I

OJ

110 100 90 D w l-a:

tl'.:8 0 LL 0

1-z

~70 Q'.: w a...

GO 50 I

PALISADES ++ CONTROL ROD WITHDRAWAL FROM 1315. G MWT ++ 0. 00006 :C.;.,/sec I

~j PQl-JER I EVEL HEAT F~UX 3

TOTAL

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

it TOTAL 1

EEOWATER fL.,OW 3

3 3

5 TOTAL

~TEAHJ_INE FLOW~ -

I I

NOTE:

Dscillati bns due tp I

,afety va lve hyste res is I

51 I

n

~

r ~,

\\

I I

~

5 I

~.

~

w I

I

~~

'I : I

~ r:I I

' ~~ ~

I I I c

I i 11 I I 1 2 1 2 1 2 I i I I

I

. q I

I 4

~

4

\\( ~

iq*

a '5 5.

I I

.+ 0 80 120 160 200 2i+O 280 320 TIME. SEC -... **

FIGURE 3.18 POvJER, HEAT FLUX, AND SYSTEM FLOHS, SEQ.

114823 JUN 7i'04*l?*ti8 CONTROL ROD5 \\.IITHDRA\\*/AL FROM 52'i POWER

@ 6.0 X 10-

~p/SEC. 360 I

t I

[

I i

I 400 I w 00 I

x :z I :z.,

I

'-l

'-l I

/ - - - - - - - - -* - - - - - - - - - -

LL CD w 0

\\

PALISADES ++ CONTROL* ROD WITHDRAWAL FROM 1315 8 MWT ++ 0 00008 ~c/sec 180 I~

~j AVE.

Fl~EL TEMPER HURE CORE I ~LET TEMPE )ATURE i80 1~0 120 100 800 800 3~. AVE.

CVRE COOLAN TEMP.

~

CLAD T MPERATURE

~

I NOTE:

Oscilla ~ions due to safety ~alve hys lteresis 0

' i I

1 1

v

~

I 1

l 1

L... B--

I i

I

'I

'I lj i

ll tj

"' 3

"' *3

" s n

~

j

[

lf 0 80 120 180 200 240 280 320 TIME. SEC FIGURE 3.19 CORE TEMPERATURE RESPONSES, CONTROL ROD )1/ITHDR.L\\\\.JAL FROM 52% POlffR

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

SEQ.

1148 23 JUN 77 04: 11' *~9 380

! I I

I I w

\\.0 I

x :z I

z:

"'Tl I

-...J

"-J I

Lt.

(!)

w 0

60 50 40 30 20 10 0

-1 ~

I I

PALISADES ++ CONTROL ROD WITHDRAWAL FROM 1315. 8 MWT +* 0. 00008 !.-;/sec

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 I

I I

NOTE:

Osei 11 at ions du e to I

safet_~ valve hj steresis I

I 2

2 I

)r-r"-

v ~

v

,~

v

~*

/ v I

':! 'I v

~

'rl'-'vvvvv wvv

,~V v£ I

3

~

3 'I I

I I

40 80 120 160 200 240

. 2 80 320 380 TIME. SEC FIGURE 3. 20 PR I MARY LOOP TEMPERATURE CHANGES, SEQ.

t H s 23 JUN 7? o :+, t 7, 50 CONTROL ROD5WITHDRAWAL FROM 525~ POWER

@ 6.0 X 10-6~/SEC.

I I

I 400 I

~

0 I

z I :z.,

I

'-l

'-l I

i

a:

1-1 Cf)

D-.

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

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

I

+::>

I

z I :z,,

I

-.....J

-...J I

-~--------

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~~~~~-----.

80 cnSO w

I u z 40 20 0

NOTE:

Oscillations due o

safety v lve hyst resis

2.

3.

CHANGE IN STEAM EN.

~ATER LEVEL. LO P 2 CHANGE fN PRESSU !ZER WATE LEVEL I

- 2 ~'---~---:4~0:--~~870~~-1~2~0=--~~1~8~0~-.~2~0~0:--~~2L4~0~----,2~8-0~~-3L2_0~~3~G-0~___14QQ TIME. SEC 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.p.

N I

z I :z I

-...,J

-...,J I

co

I I

.1 I

I I

'I XN-NF-77-18 1.?

CONTIWL ROD DIWP I NC J DENT

~ *-

-~ *--*--* **-

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

The control rod drop incident is defined as the inadvertant re-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 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-ed by either of the following two independent means:

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

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

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 was assumed to be in the manual mode, therefore conservatively inhibiting automatic rod insertion during the transient.

When the reactor regulating system is in the automatic mode, rod withdrawal is also prohibited by the rod drop protection system.

The pressurizer spray was assumed operable in order to minimize any pressure rise and therefore minimize DNBR.

The analy-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 radial peaking factors are given below.

Rod Group Rod Worth A

minimum, -0.04 ~P 4

maximum, -0.12 ~P BOC FR

l. 60

l. 66 EOC F R

l. 60

l. 64

Case 2

3 4

The-results of the Dropped Rod Worth

( Llp)

-:0.04

-0. 12

-0.04

-0. 12 analysis are Reactivity Coefficient BOC BOC EOC EOC XN-NF-.77-18 summarized below.

Applied F

---'---R-MDNBR l.60 l.42 l.66

l. 35

1. 60 1.42 1.64 1.40 The lowest MDNBR occured for Case 2 (maximum dropped rod worth at BOC).

Although the minimum dropped rod worth for cases 1 and 3 showed a higher return to power after rod drop than did the maximum dropped rod for cases 2 and 4, the lower radial peaking associated with the lower dropped rod worth resulted in a higher MDNBR.

The system responses for the case yielding the lowest MDNBR are shown in Figures 3.23:through 3.28.

I I

I

14 0 12 0 100 D w l-a:

~80 LL 0

1-z wso u

()'.'.: w CL 4 0 20 0 0

., '! 5 3 'I

~

\\'---rr-2

2 10 20 PALISADES ++ CONTROL ROD DROP 1

2 3

'I 5

3 "

t; 3..

2 -

2 30 40 POWER EVEL HEAT F ux TOTAL bRIMARY CO DLANT FLOl-J TOTAL EEOl-IATER LOW TOTAL STEAM UNE FLOW 3

c:

3 3

~

2 2

=:: ~

50 TIME. SEC 60 70 3

3 3

~

\\

~

I\\

~

1~

i-=-___

11-t-__

f'--.-.___.

5 5

80 90 100 FIGURE 3.23 POWER, HEAT FLUX, AND SYSTEM FLOWS CONTROL ROD DROP SEQ.

3 1 0 0 1 JUL 7'i' 1 0

5 6

0 5.

I.p.

U1 I

z I z

""Tl I

-....J I

LL

. CD w

0 PALISADES ** CONTROL ROD DROP 180 J 1

AVE.

r:-* JEL TEMPER HURE 2

CORE r ~LET TEMPE 1ATURE 3

AVE.

C)RE CODLAN TEMP.

.j CLAD T MPERATURE 160 u I~

u '

1 1

14 0 1 r--

1 r-IJ

-I\\

u

\\

12 0 100 I\\

I

~

.j Lj Lj 4

r--l-___

800 4

LI 3

3 3

3 Lj 3

.., 4

~

L

£.

3 600 I

I 10 20

.3 0 40

.tJO 60 70 80 FIGURE 3.24 CORE TEMPERATURE RESPONSES, CONTROL ROD DROP TIME, SEC SEQ.

310 01.JUL ?'? 10 :56 :06

')

< 4 90 100 I

-+::>

O"l I

z I z

-n I

-....J

-....J I

-~---- -* - - - - ----- -- - --

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

LL CD w 0

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 1 2 ~

~

3 :i 3 q 3 "

3 4 3 ".

3 4 0

~ ~

~

\\

'~

~

\\

~

~-)

l 2 1 2

\\

'-3" ---

-10

-20

-30

-40

-50

-6~

10 20 30 40 50 60 70 80 90 100

TIME, SEC FIGURE 3.25 PRIMARY LOOP TEMPERATURE CHANGES, CONTROL ROD DROP SEQ.

3 l 0 0 1 JUL 77 10 : 5 6 : 0 7 I

.+:::>

-.....i I

z I :z

"'Tl I

-.....i

-.....i I

a:

(f) 0...

400 300 200 0

-3~'G 10 2-0 PALISADES ** CONTROL ROD DROP

.3 0

1.

STEAM JOME PRESS JRE CHANGE LOOP 1

2.

STEAM JOME PRESSl~RE CHANGE LOOP 2

3.

'+ 0 PRESSU' >I ZER PRES PURE CHANG1

.t5 0 TIME, SEC 60 70 3

80 FIGURE 3.26 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, CONTROL ROD DROP SEQ.

110 0 1 JUL 77 10 : 5 6 : 0 8 90 100 I

~

co I

z I :z

"'Tl I

'-J

'-J I

co Im'* -- --

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

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

t8 I u z 20 0

-20

-40

-60

-80

-10 PALISADES

~ CONTROL ROD DROP

1.

CHANGE [N STEAM I bEN.

i.JATER LEVEL. LO OP l

2.

CHANGE IN STEAM

~EN. i-JAT"" 0 CC:-f LO

~p 2

3.

CHANGE LN po---

~iTER 1-JRTE g L VEL i 2 1 2 l 2 1 2

~

\\

~

I I

~

""3

\\

~

\\

~

'~-

~

u

~

3

~

1~

u 10 20 30 4_ 0 50 TIME. SEC 60 70 80 FIGURE 3.27 LEVEL CHANGES IN PRESSURIZER AND STEAM SEQ.

310 01 JUL 77 10'~6:10 GENERATOR, CONTROL ROD DROP i

I I I I i

I I

I 1

I I i I

I j

~

)

_j 1

I I

I l

I I I 3 ~

\\

90 1 CJ 0 I

~

l.O I

z I :z I

-....J

-....J I

3 2

0 (f) 0::

er:

_J

_J o-1 d

-2

-3

-~ *o PALISADES ** CONTROL ROD DROP 2

2 2

1 3-,_;

lJ.

.:J q 1

3 q 1

3 q 10 20 30 FIGURE 3.28 REACTIVITY FEEDBACK, CONTROL ROD DROP 1

2 3

. q 2

40 MO DERR OR REACTI 1 ITY DOPP LE R REACTIVI y BORON

~EACTtVITY TOTAL

~EACT IV I TY 2

2 3 q 3 4 1

50 TIME, SEC so

~

2/

3 "

3 3

1 '

1 4

lY ~

70 80 SEQ.

310 01 JUL ?7 10*56:11 2

3 1

90 100 I

c..n 0

I z

I z,,

I

-...J

-...J I

co

!1 I

I I

I I.

I I

1:

I

I I

II I

! XN-NF-77-18 3.3 LOSS OF COOLANT FLOW INCIDENTS A loss of primary coolant flow may result from a loss of electrical 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 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 coolant pumps from operation at 102% of rated power (2530 Mwt) without turbine generator assist.

(b)

Seized rotor of one primary pump, under full load operating conditions.

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

The combination of the positive BOC moderator and the minimized Dapper coefficient maximizes the core power and hot rod heat flux during these transients.

The following control systems were assumed active or inactive to decrease the transient pressure and thus minimize the predicted margin to DNB:

Pressurizer spray - active Pressurizer heaters inactive Steam dump - inactive XN-NF-77-18 3.3.l Four-Pump Coastdown The individual primary loop flow rates for the four-pump

/

coastdown inci9ent were determined from RELAP4-EM(?) calculations of an electrical failure at 102% of stretch power.

Figure 3.29 shows the results of the calculations as fractional core flow vs time.

The RELAP4-EM pump model describes the interaction of the centrifugal pumps and the coolant fluid using the four-quadrant homologous curves which are constructed empiri£ally by the NSSS supplier.(8) From these curves, along with the pump rotor moment of inertia and the primary loop hydraulic characteristics, values of head and torque are uniquely defined by the volumetric flow and pump speed.

Also shown in Figure 3.29 are the Palisades reactor test results of an unassisted primary coolant flow coastdown from hot zero 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 the two power levels. Since the primary pressure and temperature remain essentially constant during the four-pump coastdown incident, no void formation will occur that would alter the applicability of these results.

The transient responses for the four~pump coastdown incident are shown in Figures 3.30 through 3.34.

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

The reactor trip occurs at 1.58 seconds.

System pressure peaks at 2073 psi a.

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

I I

.1

I I

1.*

I I

1 I

I I XN-NF-77-18 3.3.2 Locked Rotor For the case of the locked rotor instantaneous seizure 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 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 signal.

Core average temperature increased to 579.2°F with a peak primary 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.

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

The flows of the four primary coolant cold legs are no longer equal under three pump operation since reverse flow occurs in the cold leg of the locked pump during the locked rotor incident.

The PTSPWR2 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 operation when the temperatures,of the cold legs are nearly equal.

The core average temperature,on which the point kinetic calculations are based, is then a g6od representati~~ of the uniform core cross-sPctinnal tem-perature distribution.

A study conducted at Battelle Columbus on the Palisades core show that there is only a small change in the core flow 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, XN-NF-77-18 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-perature distribution in the core during the locked rotor incident has no effect on the applicability of the point kinetics model since at the time of MDNBR (2.4 seconds) the maximum calculated difference between the temperatures of the three operating cold legs is only 3°F.

The resulting non-uniform temperature distribution in the core is negligible.

Twenty (20) seconds after the incident, the temperatures of all cold legs are nearly equal as is shown in Figure 3.37.

I I

I

'I I

I I

r.*-1 I!

I

1 I

I I

I

- --- -. - -... -I* --.. - - -: --.

~ - - -) -) -

100 90 80

~70 3:

~ 60 LL..

'; 50 1.1..

4-0 40

-+-'

QJ

~ 30 QJ c..

20 l ()

0 5

Palisades Test Results at Hot Zero Power-RELAP4-EM 102% of Rated Power (2530 Mwt) l 0 15 20 Time From Pump Trip (Sec) 25 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 I

CJ1 CJ1 I

z I z "Tl I

-...J

-...J I

co

120 100 0 w I-a:

0:::8 0 LL 0

I-z wso U*

er: w (l....

40 20 I PALISADES ** 4 PUMP TRIP FROM 2580. 6 MW **

2 3

'I 5

DOWER EVEL

~

~EAT F UX TOTAL RIMARY CO LANT FLOW TOTAL EEDWATER LOW TOTAL TERM LINE. FLOW i

i 1 3

5

~-+-~-+-~~~~+--+-~~~~~~l 0 0 2

4 s

8 10 12 14 (s*

18 20 TIME, SEC FIGURE 3.30

POWER, H~AT FLUX, AND SYSTEM FLOWS, 4 PUMP TRIP SEQ.

i 8 8 8 15.JUN ?? 19 ' 2 'I ' 0 :J I

U1 O"l I

z

z "Tl I

-,,.I

-..J I

-~ -!-

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

~~--;-*--*-*--:- 7 -,---.

~-[ -;--) - ---

LL

(.'J w 0

PALISADES ++ 4 PUMP TRIP

~ROM 2580 G MW +*

18 OIJ I

1 AVE. F C.EL TEMPERrURE I

2 CORE I ~LET TEMPE ATURE 3

AVE. C ~RE COOLAN TEMP.

I

<t CLAD T MPERATUREI Ou I

i 1G l

1 100

~.

u

'~

-v

~

I v

~ ~

I r---_ 1 14 0 120

r--_ i-L---

800 800

.:l 3

-z 2

2 2

40~

2 8

FIGURE 3.31 CORE TEMPERATURE RESPONSE; 4 PUMP TRIP G

q tl 3

3 3

j

~.,

~

10 12 14 16 18 20 TIME. SEC SEQ.

18 8 8 15 JUN 7r 19 : 2 'I : 0 5 I

(.Tl

"'-J I

z I :z "Tl I

"'-J

. "'-J I

LL

(!)

w 0

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

8 4

0

-4

-8

-12 2

2 3

~

8 CHANG IN AVE.

RI MARY CO LANT TEMP. LOOP 2 CHANG IN HOT L G -

COLO EG TEMP.

IFFERENCE LOOP 1 CHANG. IN HOT L G -

COLD EG TEMP.

IFFERENCE L_OQP 2 2

10 12 16 18 TIME. SEC FIGURE 3. 32 PRIMARY LOOP' rEMPERATtJ:RE CHAN.GES, 4 PUMP' TR:IP*

SEQ.

1888 15 JLJN 77 19 '2~ :Q6 20 I v.,

<X>

I

z:

I"

z:,,

I

-....J

-....J I

co I

I

-- -)

-*~ - - -*'..'. -' -!' - - -;---i---+--! *-.... -.' -- -1

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

a:

Cf)

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

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

LOOP l LOOP 120 80 3

40 0

18 20

-4 0'--------::--'---_i_---=-1-----_i____-__J __

__L __

_l__ __

_J__ __

I o

2 4

G 8

10 12 14 113 FIGURE 3.33 TIME, SEC PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, 4 PUMP TRIP SEQ.

1888 15 JUN 7? l.'3 :24 :Q8 I

(.]1

\\.0 I

z I *z

""Tl I

-....J

-....J I

co J

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

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 20 I

~

r--z___

/

r--z___

1 2 10 0

cn-10 w

~

- \\

~

I I

I I u I

z

-20 t

I I

I

-30 I

i

' j

-4Q i I

I I

I I I

-5~

2 FIGURE 3.34

\\j I\\

I

~...

2 t 2

~** -*

I

.2~

i

-... 1-..

~

I I

I i

I I

G 8

10 12 14 lG T lME. SEC LEVEL CHANGES IN PRESSURIZER AND STEAM GE.NERATO.RS; 4 PU~1P fRIP SEQ.

18 8 8 15 J JN ?? 1.9 '2 4

0 9 r

.1 2....

I I

I 18 20 I

O'l 0

I.

z I z "Tl I

--.J

".J I

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 4 0 20 0 0

~

I 2

8 10 TIME, SEC FIGURE 3.35 POWER, HEAT FLUX, AND SYSTEM FLOWS, LOC KEO ROTOR 5

5 12 14 18 18 20 SEQ.

1882 15 JUN Ti' 19

18

39 I

O'l z

I z

"'Tl I

-....J

-....J I

co

LL

(!)

w D

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

18 0 -I l

AVE. F :1EL TEMPER ATURE 2

CORE I ~LET TEMPE ~ATURE I

3.

AVE.

C lRE COOLAN TEMP.

'I CLAD T MPERATURE 180 u 1

1

~

u

~

u

~ ~

,1 r

~

I

~ -

14 0 120 100 800 I

~

I

~

'I

'I r

II lj r

,j

~

3 3'

I 5

600 2

2

c.

2 8

FIGURE 3.36 CORE TEMPERATURE RESPONSES, LOCKED ROTOR 10 TIME. SEC II

~

I I

I 12 14 1 ()

18 SEQ.

lee 2 15.J JN ?i' 19 : 18 '4 1

'f i

I 20 I

7l N

I

z I :z I

-..._J

-..._J I

co

LL (0

w 0

30 20 10 0

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

2 3

~

2 2

1 1

LOOP 1 LOOP 2

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

-4 ~

2 8

8 10 12 14 18 18 20 TIME, SEC FIGURE 3.37 PRIMARY LOOP TEMPERATURE CHANGES, SEQ.

1862 15 JUN 7? 19 1 t8 *IJ2 LOCKED ROTOR I

CJ) w I

x :z I z

"'T1 I

I 00

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

er:

(f)

(L_

300 200 150 10 0 50

.o

-5~

I i

! I I

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

1.

STEAM nOME PRESS JRE CHANGE LOOP t 2

STEAM nOME PRESS JRE CHANGE LOOP 2 3

PRESSU HZER PRES >URE CHANGI 2

2 2

2 1

2 bP'

=---

ci--L__

~

~ v

/

L

/

3 3

~ ----r----z_

~

lA

~ ~

~

2 8

10 12 16 18 20 TIME. SEC FIGURE 3.38 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, LOCKED ROTOR SEQ.

18 8 2 15 J LJN 7? 19 : 18 : -! 3 I m

-!=:>

I z

I :z.,

I

-...J

-...J I

80 60 t+ 0

(.()2 0 w

I u z 0

-20

-t+O I

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

I q

CHANGE IN STEAM ;EN.

~ATER LEVEL. LOr 1 2

CHANGE IN STEAM EN.

wATER LEVEL. LO P 2

31.

CHANGE IN PRESSU~IZER WATE ~ LEVEL I

I 3

3

~

--1-_ ~

1 2 g --

~

'J

~

~ ~

~

1--1-_

.........._ 2

~ -------t_____

2 2

~

1------

J---

~ ~

~

_LJ__---

2 8

10 12 1 t+

16 18 20 TIME, SEC FIGURE 3.39 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, LOCKED ROTOR SEQ.

18 8 2 15 JUN ??' 19 : i 8 ' 4 5 I

O)'

t1l I

z I

2 "Tl I

'.J

'-I I

co XN-NF-77-18 3.4 EXCESSIVE FEEDWATER INCIDENTS Exce.ssive heat removal by a feedwater system malfunction can cause the primary system to depressurize and the reactor power to increase due to negative moderator reactivity coefficients.

Protection against these undesirable conditions and high steam generator water levels is provided by high steam generator water-level alarms, and by reactor trips due to high neutron flux, low pressurizer pressure, thermal margin/low pressure, or low steam generator pressure.

The incidents which result in the most severe excessive cooling action by the feedwater are analyzed:

Reduction in feedwater enthalpy from 102% of rated power Increased feedwater flow from 52% of rated power 3.4.1 Reduction in Feedwater Enthalpy An instantaneous enthalpy drop of 58 Btu/lb is assumed for the decrease in enthalpy incident from 102% of rated power (2530 MWt).

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

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

The plant responses to the transient resulting from the decrease in feedwater enthalpy are shown in Figures 3.40 through 3.44.

I I

I I -,

I

' I I

I I XN-NF-77-18 The decrease in the feedwater enthalpy leads to a decrease in the primary coolant temperature and pressure, along with a reduction in secondary pressure in the steam generators. A low pressurizer pressure trip signal occurs at 124.7 seconds.

The core average temperature decreases to 523°F and the pressurizer pressure decreases to 1708 psi a.

An MDNBR of l.75 for this case occurs at 8.0 seconds.

3.4.2 fncreased Feedwater Flow From 52% Power The feedwater flow is increased from 50% to 100% in 8 seconds to initiate the increase in feedwater flow incident.

The reactor 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 which automatically closes the feedwater regulating valves.

End-of-cycle (EOC) reactivity coefficients are assumed, since the negative moderator co-efficient acts to increase core power during the transient.

The nominal EOC Doppler coefficient is multiplied by 1.2.

The initial reactor pressure is 2010 psia, 50 psi below the nominal value of 2060 psia.

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

The reactor power 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.

PALISADES ++ REDUCTION OF FEEdWAlER ENTHALPY FROM 2580. S MW 14 0 1

POWER I EVEL 0 w f-a:

100 Cl'.:8 0 LL 0

f-z W~)O u

er:

w Q_

rt 0 20 0 0 2

HEAT F, ux 3

TOTAL I l>RIMARY CO DLANT FLOW lj TOTAL EEO WATER LOW 5

TOTAL

~TEAM LINE FLOW 1 2 1 2 ~

c:

.1 2 3

'i n 3 3

3 3

c p

1

'I

~

-f-4-_~

~\\

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

\\

I..

l I\\

\\ \\

\\ I\\.

~

\\: ~

\\

1 2 5

20 40 60 80 100 120 14 0 160 TIME, SEC FIGURE 3.40 POWER, HEAT FLUX, AND SYSTE~ FLOWS, REDUCTION OF FEEDWATER ENTHALPY SEQ.

1?42 22 JUN 77 19 *H *04 f\\ 2 5

180 3

ti

-* - --* ----"II---*-----.. - - ----*--1 1-5 200 I

O"\\

co I

z:

I :z:

I

-....i

-...J I

co

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

LL C!J w 0

180 16 0 14 0 120 100 800 600 40~

PALISADES ** REDUCTION OF FEEDWATER ENTHALPY FROM 2580. 6 MW

~

~~

RVE.

F UEL TEMPER! ATURE I

CORE r: 11LET TEMPE bfffURE

3.

'1VE.

c:ORE COOLAN TEMP.

.J.

CLAD T MPERRTURE

~

l 1

1 1

IJ

--r--_ r----_

~

-IJ

\\

'l q

\\

~

q*

q

'I 3

3 3

3 7

'Z 3

3

~ 9 3 q 2 3 'l

'Z c:

~

20 40 GO 80 100 120 14 0 180 TIME. SEC FIGURE 3.41 CORE TEMPERATURE RESPONSES, REDUCTION OF FEEDWATER ENTHALPY SEQ.

1?4222 JUN ?7 t.9:H*05 T

180 2 3 '!-

200 I

0)

"° I

z I :z I

-..J

-..J I

co

10

-- 0 LL

~

w 0

-iO

-20

-30

-40 I I I

-50 I

PAL I SADES *

REDUCT 1 ON OF FEEDWATER ENTHALPY FROM 2 5 8 0. 5 MW 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 3 4 3 4 3 4 4 CHANG IN HOT L. G -

COLO EG TEMP.

)IffERENCE

~OOP 2

. 2 3 '!

---.... ~

~

~ ' ~ ~

~

i 2

1_3----

\\

~.

3 '!

3 4 20

+ 0 60 80 100 120 i 4 0 150 180 200 TIME. SEC FIGURE 3.42 PRIMARY LOOP TEMPERATURE CHANGES, REDUCTION OF FEEDWATER ENTHALPY SEQ.

t ?' i 2 2 2 JUN ? : 13 !I : iJ 6 I

-....J 0

I

z I :z

"'Tl I

-....J

-....J I

- - *- -- - --*- {-* - - \\-... - - - - -

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

cc u::

0.

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

I I, 3-_2 -+'

=-=,

2=---1 100!--~r-----t-----+----t----+----+----+~c..__..=-t----+------i

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

20 u fJ GO go "l'J:j 120 14 0 160 FIGURE 3.43 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, REDUCTION OF FEEDWATER ENTHALPY SEu.

i?'42 22.JlJN ?7 'i.3 "l7 *08 3

180 200

--- -- -----*--=-*-~~~

I I

z I :z

""Tl I

I co

PALISADES ** REDUCTION OF FEEDWATER ENTHALPY FROM 2580. 6 MW 20 -- - ---------:----- - _____ : ______

~

1

-- ~~:~~-r~~-~~~:~ ~~~:--~~~~=: ~~~~~: t~~- ~ ---~-------*

1

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

i I

I I

0

+-

--+-

j

-201 i

I !

i I

3 I

-4 Oi c.n w

I u z I-<

-so:

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

-10*0

-1 20--- ---- --~----*- -- -

1J 2 0 40 60 80 100 TIME. SEC FIGURE 3.44 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, REDUCTION OF FEEDWATER ENTHALPY 120 14 0 160 180 SEQ.

17'4 2 '2 2 JUN 1'7' 19 I 47 : 10 200 I

--.,J N

I

z I ::z,.,

I

--.,J

--.,J I

co

- - - *-.. - - ---- j- - - --

11 0 100 90 0 w l-a:

tl'.:g 0 LL 0

1-z t3'i'O Cl'.:

w CL so 50 PALISADES ++ INCREASE IN FEEDWATER FLOW RT 52% POWER 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.~.. -. I '-""

1 ')

1 2 l 2 L Z l

L ;!

1 ;!

~ ~

1 2 -

r--._

5 20 FIGURE 3.45 5

5 5

5

5.

5 5

40 60 80 100 120 14 0 160 TIME, SEC POWER, HEAT FLUX, AND SYSTEM FLOWS, INCREASE IN FEEDWATER FLOW AT 52% POWER

SEO, 19 12 2 1 JUN 7 7 2 0 ' 2 2 I 5 2 l

5 5

180

' 200 I

w I

z I :z "Tl I

-...,J I

co

. _c*.. -

LL (9

w 0

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

~

l

./

100 900 800 700 600 4

4 4

4 tj 4

4 lj 4

3 3

3 50~

20 Lf 0 80 80 100 120 140 i60 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 1

180 4

3 200 I

-..J

..j:::>

i

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

LL

(!)

w 0

3 2

1 0

-1

-2

-3 PALISADES ** INCREASE IN FEED~ATER FLOW AT 52% POWER I

q.

1 CHANG IN AVE.

0RIMARY CO PLANT TEMP. LOOP 1 3 q "

~

n

.A.lllll1o.

"""~.... -111'11 l-lll LUO!

. v -

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 II I

'r! v\\

\\

1 2 1 2 1 2 l

L l 2 1 2

..J,__2 l 2 0

20 40 60 80 100 120 14 0 160 180 TIME. SEC FIGURE 3.47 PRIMARY LOOP TEMPERATURE CHANGES, SEQ.

1912 21 JUN 77 20 122 *54 INCREASE IN FEEDWATER FLOW AT 52% POl~ER 200 I '-l 01 I

x :z I :z

"'Tl I

'-l

'-l I

80 40

-20

-40 PALISADES ** INtREASE IN FEEDWATER FLOW ~T 52% PO~ER l

STEAM OOKE PRESS ~RE CHANGE LOOP l

2.

STEAM DOME PRESS 1lRE CHANGE LOOP 2

3.

PRESSU HZER PRES ~URE CHANG 3~

~

3.__........

\\\\

~

3.

_]:..---

\\

~.*

............ ~

2.

1 2 1

1 1 2 t 2 1 2

-20

~o 60 80 100 120 14 0

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

2 3

1 2 180 200 I

-...J I

z I z "Tl I

-...J

-...J I

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 u z 1--;

3 3

3 40

~

20 r

J.

0 3

3 3

-2~

20 40 60 80 100 120 140 16 0 TIME. SEC FIGURE 3.49 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, INCREASE IN FEEDWATER FLOW AT 52% POWER SEQ.

19 12 2 1 J UN 77 2 0

2 2

5 7 3

180 200 I

-...J

-...J I

z I z.,

I

-...J

-...J I

XN-NF-77-18 3.5 EXCESSIVE LOAD INCREASE INCIDENT Excessive load incidents may be initiated by opening of the turbine control valves, atmospheric steam dump valves and/or the steam bypass to condenser valve.

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

Along with an increase in nuclear power, there is a decrease in main steam pressure and in primary coolant temperature and pressure.

Protection against damage to the reactor 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.

Although excessive load incidents are much less severe than steam line breaks, 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-ing these more credible incidents.

The excessive load increase incidents were analyzed from 102% of rated power and from hot zero power (rods at power dependent insertion limits).

Both cases were analyzed with two loop (four pump) operation as required by the technical specifications for these operating conditions.

End-of-cycle (EOC) kinetics parameters were used to maximize core power during the reactor cooldown.

For the excessive load incident in which primary system pressure is decreased sufficiently to activate the safety injection system (initiated from hot standby), minimum capabilities of the boron injection system were assumed.

The use of the kinetics para-meters and assumptions concerning the availability of the boron injection system are detailed in Section 3.8 (Steam Line Break).

I I

I

I*

I I

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

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

Increased steam 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 as well as the primary coolant system pressure decrease.

After trip the 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 average temperature reaches 532°F, the steam dump and bypass valves close, terminating blowdown of the steam generators'.

There is no return to power following the trip and the minimum DNB ratio during the incident is l.74.

3.5.2 Excessive Load From Hot Standby The system responses to the suddert opening of all steam 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 is 2110 psia, 50 psi above the nominal value of 2060 psia. This is conserva-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 at 17 seconds following the opening of the valves.

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

The maximum core average heat flux is 17,075 Btu/hr ft2.

Borated water from the high pressure safety injection system reaches the core at XN-NF-77-18 80 seconds.

Boric acid from the charging pumpings reaches the core region after 128 seconds, terminating the core power increases.

The. critical heat flux was determined using the modified Barnett CHF correlation. (4) 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.

The local peaking factors occurring in the area of the stuck control rod after return to power were calculated using XTG.( 9) Local effects of Doppler and void feedback were included in the calculations of the local peaking factors.

The negative feedback effects due to voiding were not used, however, in the determination of overall core 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 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 T

FQ _= 16. 0.

I I

1*

I I

' 14 0 0 w l-a:

120 100 er: 8 0 LL 0

1-z tj6 0 er:

w a_

40 20 0 0 PAL I SADES +* EXCESS I VE LOAD INCREASE FROM 2 5 8 0. 6 MW 51 I

I

~~

POWER EVEL I

HEAT F ux I

31 TOTAL 'RI MARY COi DLANT FLOl-l I

[

I TOTAL EEOWATER I 11

LOW i

s]

i TOTAL

>TEAM LINE FLOW 1

I T

I I

L~

i I

1 2

\\

I I

\\

3 3

~

\\

~

~'

r\\

~

\\~

~

~ ~ ~

I~

rl_

~

I 5 1 2 1 2

'I i: 1 2 lj Ii 1 2 v

10 20 30 40 50 60 70 80 90 TIME. SEC FIGURE 3.50 POWER, HEAT FLUX, AND SYSTEM FLOWS, EXCESSIVE LOAD INCREASE SEQ.

445 27 JUN 7714156*04 I

i I

I 3

I I

I i I

I i

I I

\\

I 4 si 100 I co I

co

LL

(.!J w 0

180 16 0 14 0 12 0 10 0 800 600 PALISADES ** EXCESSIVE LORD INCREASE FROM 2580. 6 MW u

1 AVE.

F ~EL TEMPER! HURE 2

CORE 1 ~LET TEMPEi 'ATURE 3

AVE.

C ~RE COOLAN TEMP.

tj CLAD T HPERATURE u

l 1

__..,... n u

\\

J

\\

J

\\

~

4

-..l_____

l 3

1 I

l 3---

., 4

(.

I) 3 4

') 3 4 2 3 4 2 3 4 2 3 lj 2 3 lj 10 20 30 40 50 TIME. SEC 60 70 80 FIGURE 3.51 CORE TEMPERATURE RESPONSES, EXCESSIVE LOAD INCREASE SEQ.

445 27 JUN 71' 14 :56 :1)5 l

2 3 lj I

90 100 I co N

I z

I z..,.,

I

'...J

'...J I

co

LL

(!)

w 0

10 0

-10

-20

-30

-4 0

-50

-6~

PALISADES ++ EXCESSIVE LOAD INCREASE FROM 2580. 6 MW l

CHANG I IN AVE.

PRIMARY CO JLANT TEMP. LOOP l

,/ n 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

~

\\ ~

\\

'\\

~

"----. ~

1 2 1 2 1 2 1 2

'~

~

~II 3 tj 3 tj 3 tj 3 tj 10 20 30 40 50 60

'1' 0 80 90 TIME. SEC FIGURE 3. 52 PRIMARY LOOP TEMPERATURE CHANGES, EXCESSIVE LOAD INCREASE 100 I

OJ w

I

z I :z "Tl I

--...J

--...J I

OJ

er:

U?

Q_

400 300 200 100 0

-10

-20 PAL I SADES ++ EXCESS I VE LOAD INCREASE FROM 2 5 8 0. 5 Mhl l

STEAM OOME PRESS! IJRE CHANGE LOOP l 2

STEAM OOME PRESS IJRE CHANGE LOOP 2 3

PRESSU bIZER PRES URE CHANG l 2 l 2 l 2 l 2 l 2 l 2 1 2 I

~

I I

IJ I

j I L u

10

~

r-----1-__

20 30 40 3

50 TIME. SEC 3

3 60

?O 80 3

-::GURE 3.53 PRESSURE CHANGES IN PRESSURIZER AND SEQ. 445 27 JUN ??14*56*08 STEAM GENERATORS, EXCESSIVE LOAD INCREASE l 2 3

90 100 I

CD I

z I ::z I

-....J

-....J I

CD

~------

20 0

-20 (f)-4 0 w

I u z --60

-80

-10

-1~

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

~

-~

1 CHANGE [N STEA!i 'EN.

l-IATER LEVEL. LO p t

~ l 2

CHANGE [N STEAM 'EN.

wATER LEVEL. LO OP 2 3

CHANGE [N PRESSU ~rZER l-IRTEI ~ LEVEL

\\ \\

~

\\

.J

~

u 10 20

" ~

~

\\

~

3 30 40 50 TIME. SEC 1.

1 -

~

?

-:i 80

?O 80 3

FIGURE 3.54 LEVEL CHANGES IN PRESSURIZER AND SEQ.

445 27 JUN 7714=56110 STEAM GENERATORS, EXCESSIVE LOAD INCREASE 1 -

3 90 100 I co

(.J1 I

z I

z:

I

"'-J

"'-J I

co

_J

1If0 12 0 100 0 w f-a:

tY80 LL 0

f-z tj60 CY w

Q_

40 20 0 0 PALISADES ** EXCESSIVE LOAD INCREASE FROM HOT STANO-BY 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 3

3 3

~

5 5

5 1 2 l

l

_M f) ij

'I 4

- " ~ 'I 4

4 l~

l ~

1 '

1

~

20

~o 60 80 100 120 14 0 160 TIME. SEC FIGURE 3.55 SEQ ti' 2Ei 2 3 JUN ?"!' 13 I 3? : l 6 PO~ER, HEAT FLUX, AND SYSTEM FLOWS, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY 3

5 l

'I 180 5

200 I co en I

z I z I

-...J

-...J I

co

LL (9

w 0

PALISADES ** EXCESSIVE LORD INCREASE FROM HOT STANO-BY 5?5i-~~--,-~~~--,~~-:--,--~~,1'.

~i'\\\\TF~=-rn=;--.==:-;:;~~=-~-r-~~~-,----=~~~~~~

2 475r--~~-r-~~--1-~~:--t~~----jf--~~+-~-/--+-~~-+~~-4~~---=::~~~

4 o~~~--;-:-;:;--~~4~0~~~6~0~~~8~0~--:---:-'-:10~0=--~~1~20~~-1L4_0~~1~6-0-,~--118-0~~~200 TIME. SEC FIGURE 3. 56 CORE TDlPERATURE RESPONSES, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY SEQ.

1726 23 JUN i? 19

3?

19 I co

-....J I

z:

I :z:

"'Tl I

-....J

-....J I

co

LL

(!)

w 0

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.

I t>RIMARY CO PLANT TEMP. LOOP 2

--3 '!

3 CHANG!

IN HOT L G -

COLD EG TEMP.

HFFERENCE LOOP l 3 "

3 "

3 lj ti CHANG1 HJ unT r 1 b:o

"'" n TC:-.. 0

'r i:-r:-i:-i:;ii:-1Jri:-

nnP 9 0

-25

-50

-?5

-10

-12

-1~

~

1

-u C.)

-u 20 FIGURE 3.57

~

~9

~ '

~~

--L1____ -LL__ -LL__ -LL 40 60 80 100 120 140 160 TIME, SEC PRIMARY LOOP TEMPERATURE CHANGES, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY SEQ.

1726 23 JUN 71' 19

37

20 180 200 I co co I

z I :z:.,

I

-.....J

-.....J I

co

a:

Cf)

~

0

-20

-40

-60

-80

-10

-12

-1~

PALISADES ** EXCESSIVE LOAD INCREASE FROM HOT STAND-BY

~

1.

STERH JUME PRESSI JRE CHHNGI::

LOOP t

2.

STEAM l DOME PRESSI JRE CHANGE LOOP 2

3.

PRESSiJf ~ IZER PRES ;URE CHANG!

\\~

~'J u~ ~

~

\\

J

\\' ~

~

1 2 u

\\

1 2 1 2 I

u

\\

uu

~

Ju

~

r---L 3

3 3

~

JU 20 40 80 80 100 TIME; SEC 120 140 160 FIGURE 3.58 PRESSURE CHANGES IN PRESSURIZER AND SEQ.

112623JUNn19 1 37*22 STEAM GENERATOR, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY 1 2 3

180 200 I

OJ

\\.0 I

z I :z

-n I

-...J

-...J I

20 0

(f)-4 0 w

I u z 1--;

PALISADES ++ EXCESSIVE LOAD INCREASE FROM HOT STAND-BY 3

20 40 Ei 0 3

2 3

80 CHANG N ST AM CHANGE IN STEAM EN.

WATER LEVEL. LO P 2 CHANGE IN PRESSU IZER WATE LEVEL 3

3 3

3 100 120 14 0 180 TIME. SEC 3

FIGURE 3.59 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY SEQ.

l 7 2 8 2 3 JUN 'i'7 18 i 3 7 '2 3 3

180 200 I

<.O 0

I z

""Tl I

-....J

-....J I

OJ

(f')

0::

a:

_J 10 8

6 62 0

0

-2

-4 0 PALISADES ** EXCESSIVE LORD INCREASE FROM HOT STANO-BY 1

!10DERA OR REACT! 1 ITY 2

DOPPLE R REACTIVI y 3

BORON REACTIVITY ll TOTAL REACTIVITY 1

1

~

,~

v l.--"'""

/

y 1~

4 4

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

4 4

4

~

~ ----

j 3

r-t__

v 2

2

" 3 4/

v 20 '

4 0 60 80 100 120 14 0 160 TIME. SEC FIGURE 3.60 REACTIVITY FEEDBACK, EXCESSIVE LOAD INCREASE FROM HOT STAND-BY SEQ.

l 7 2 6 2 3 JUN 1'l' 1 g I 3 7 I 2 4 1

4 2 -

v 180 200 I

\\,D i

z I z,,

I

-.....J

-.....J I

'*. XN-NF-77-18 3.6 LOSS OF LOAD INCIDENT 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 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-denser are available to remove energy from the primary coolant system.

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-surizer and steam generator safety valves would assure that both primary coolant system and steam generator pressure do not exceed design limits.

The acceptance criteria for this event are:

The pressurizer safety valves shall limit reactor coolant system pressure to a value below the ASME code limit of 110% of design pressure (2750 psia).

The pressurizer safety valves shall limit the pressure differential between the primary and secondary systems to less than 1530 psid.

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

The most probable cause of a rapid loss of load is a turbine trip.

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

I I

I I:

I..

1:

I I

I

1 I

I I

II I

I I

!I I

I I

I XN-NF-77-18 primary coolant and main steam systems are adequately protected during a complete loss of load transient. Several cases were analyzed to determine the closest approach to each of the above criteria.

Tran~ient responses are calculated from 102%' of stretch power.

Beginning-of-cycle (BOC) kinetic coefficients were conservatively assumed, with an 0.8 multiplier applied to the Doppler coefficient.

The worst case with regard to peak primary side pressure (Case l) is the transient initiated from 2110 psia with no pressurizer relief, pres-surizer spray, steam dump or steam bypass allowed.

Figures 3.61 through 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 highest transient primary-secondary ~p (Case 2) is the Loss-of-Load initiated from 2110 psia and with no primary side pressure reducing effects from either pressurizer spray or relief valves, but allowing depressurization 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 secondary pressure, giving a maximum ~P between the primary and secondary sid~s of 1388 psid during the transient.

Figures 3.66 through 3.70 show the results for this case.

Both of these cases were tripped on high pressurizer pressure.

The case yielding the lowest MDNBR (Case 3) is the transient initiated from 2010 psia with the pressurizer spray and relief valves operable, 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 XN-NF-77-18 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 results of this case.

The Loss-of-Load transient results are summarized in Table 3.1 In none of the cases were the Palisades Technical Specifi~ation Limits of l) peak primary system pressure (-5_ 2750 psia), 2) primary-secondary AP

(~ 1530 psid), or 3) MDNBR

(~l.30) violated.

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 2110 772 No No 2394 1371 2

2110 772 No Yes 2382 1388 3

2010 772 Yes No 2274 1247 Technical specification limit on primary pressure is 2750 psia Technical specification limit on primary-secondary t.iP is 1530 psid during transients.

MDNBR

l. 55

l. 57 l.39 I

<.O (J1 I

z I :z

"'Tl I

-....J

-...J I

r-,

HO 120 100 0 w

' I-a:

0:'.:80 LL 0

I-z wso u

O:'.:

w 0....

40 20 0 0 PALISADES++ LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 6

8 12 2

3

~

5 P ~E HEAT F UX TOTAL RIHARV CO LANT FLOW TOTAL

~EDWATER LOW TOTAL TEAM LINE FLOW 16 20 24 28 TIME. SEC 3

1 32 FIGURE 3.61 POWER, HEAT FLUX, AND SYSTEM FLOWS, LOSS OF LOAD - CASE l srn.

4 o o t r J UN rr 1 o, o 5, o s

~

I

<.O O"l I

1 z

I z..,.,

I 36 40

-...J

~ -...J I

co

lL.

(!)

w Cl 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

~

.J 160 HO 0------

__ 1 1

1

\\

~

\\

u u

120 100

~

, ~

'/

800 r1----_

1 4

'l

~

1

'2

'i.

~ 4

'J

'S

~

'2 't

()

'2 "

{,

600 8

12 16 20 2q 28 32 FIGURE 3.62 CORE TEMPERATURE RESPONSES, LOSS OF LOAD - CASE l TIME. SEC SEQ.

'i 0 0 lT JUN ir l 0

0 5 r 0 'i 36 I) '2 4

~o I

\\D

--..J I

x z I z "Tl I

--..J

--..J I

LL

(!J w

0 PALISADES *+ LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 30 20 10 0

-10

-20

-t+ ~*.

8 12 16 20 TIME. SEC FIGURE 3.63 PRIMARY LOOP TEMPERATURE CHANGES, LOSS OF LOAD - CASE l 28

. LOOP t

. LOOP 2 IFFERENCE LOOP t IFFERENCE LOOP 2 32 36 SEQ.

q 0 0 l 7 J UN 77 10 I 0 5 I 0 5 Lf 0 I

<.D co I

z:

I :z:

I

-....,J

-....,J I

co

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

cr::

(j) 0...

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

2 3

LOOP 2 1oor--~.__r-~,L--t-~~+-~~-t-~~-1-~~-t-~~-+--~~-+-~~-i-~~-l 0

8 12 16 20 24 28 32 TfME. SEC FIGURE 3;64 PRESSURE CHANGES IN PRESSURIZER AND SEQ. qoo 17 JUN 7710105107 STEAM GENERATORS, LOSS OF LOAD -

CASE l 36 40 I

\\.0

\\.0 I

Z.

I z

"'Tl I

'-J

'-J I

co

120 80 If 0 (J}o w

I u z

-If 0

-80

-12

-1~

l 3 _..,...-

~

u

~

v PALISADES ** LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 3

':i./

t..--

v v 1 2

_V

~

8 12 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 r--t.._

~

l 2

~

'~

~

...l 2 ---

"' ~

~

16 20 TIME. SEC 24 28 32 36 40 FIGURE 3.65 LEVEL CHANGES IN PRESSURIZER AND SEQ.

4 0 0 17 JUN 77 10 : 0 5 I 0 8 STEAM GENERATORS, LOSS OF LOAD - CASE I

0 0

I

z I :z:.,

I

-...J

-...J I

co

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

14 12 10 0 w l-a:

0::8 0 LL 0

1-z t360 a::

w Q_

20 0

0 1 2 01~

5 PALISADES ** LOSS OF LOAD FROM 2580 6 MW & 2110 PSIA 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 1

1 1~-

2 2\\

\\"'\\

~

3

~

31

\\

~

1 \\

~

s\\

5

/

\\

5 r--- l._ N-

~

r-----.& r-----.&

~

_J

\\,

~

5

~

""' ~

\\

r-...!

N...__

r-L_

--r--t--_

1 1

1 3

~

r--.._

~

2 1

0 0 8

16 20 24 28 32 36 40 TIME. SEC FIGURE 3. 66 POWER," HEAT FLUX, AND SYSTEM FLOWS, LOSS OF LOAD - CASE 2 SEQ.

2 3 5 17 JUN 7 7 10 I 0 0 I 2 2 I _..

0 _..

I x :z I :z

"'Tl I

I

LL

(!)

w D

PALISADES +-t LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 180 'U 1

AVE.

F ~EL TEMPER HURE 160 14 0 120 100 800 600 2

CORE ! ~LET TEMPEi ~ATURE 3

AVE.

CDRE COOLAN TEMP.

CLAD T MPERATURE 0

1 l

1 U----

\\

lJ

~

u u

4 4

~

-:i 3

v

~

£

£ 4

8 12 16 FIGJRE 3.67 CORE TEMPERATURE RESPONSES, LOSS OF LOAD - CASE 2

\\ I\\

~

4 v

I 20 TIME. SEC 4

24

-............ r---__ J1---_ ' 1 LI

~

n 4

28 32 36 SEQ.

2 3 5 1 7 JUN 7 7 10 : 0 0 : 2 4 I) 40 I -'

0 N

11 z

I z

"'Tl I

-....J

-....J I

--liill: ---*;----- ~.. -'

.. --)-~--.,

1 u..

(,!)

w 0

30 20 10 0

-10

-20

-30 1 2./

\\

PALISADES ++ LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 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

~

CHANG lN HOT L G -

COLO,...EG TEMP.

DIFFERENCE v

,/

N v

~

~ '!-.....

~tj

.:I

'+

~ '\\

~

LOOP t LOOP 2

~

I 0 w I

8 12 16 20 24 28 32.

36 40 z

I z "Tl I

'..J

'..J I

TIME, SEC FIGURE 3.68 PRIMARY LOOP TEMPERATURE CHANGES, LOSS OF LOAD - CASE 2

. SEQ.

235 17 JUN 77 10 1QO 125

er:

(j)

CL 500 400 300 200 100 0

-10

-2~

PALISADES** LOSS OF LORD FROM 2580.B MW & 2110 PSIR l

STEAM JOME. PRESS JRC. CHANGE LOOP 1 2

STEAM lOME PRESS JRE CHANGE LOOP 2 3

PRESSUI >[ZER PRES URE CHANG

~

3 J

1~

1 2 2/

/

~~

rt..g_

v

~

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

v

~ ~

./

1

£ v

~

~~

u

~

" u 8

12 16 20 TIME. SEC 24 28 32 36 40 FIGURE 3.69 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, LOSS OF LOAD - CASE 2 SEQ.

235 11' JUN?? 10:00:26 I

0 _p.

I z

I z "Tl I

-....J

-....J I

co

__ _. ____ --"- -) *- _

.. --->- - -).. -..... -

60 4 0 20

(/)0 w

I u z

-40

-60

-8~

PALISADES **LOSS OF LOAD FROM 2580. 6 MW & 2110 PSIA 2

3 8

12 16 CHANGE !N STEAM EN.

WATER LEVEL. LO P 1 CHANGE !N STEAM EN.

WATER LEVEL. LO P 2 NGE !N PRESSU !ZER WRTE LEVEL 3

20 24 28 32 TIME, SEC FIGURE 3.70 LEVEL CHANGES IN PRESSURIZER AND SEQ.

235 17 JUN 7710:00:2a STEAM GENERATORS, LOSS OF LOAD - CASE 2 36 40 I.....

0 U1 I

z I z.,

I

-...J

-...J I

co

14 0 120 100 D w I--cc Cl:'.:80 LL 0

1--z t36 0 Ct'.: w 0....

40 20 0 0 PALISADES ++ LOSS OF LOAD FROM 2580. 6 MW & 2010 PSIA l

POWER EVEL 2

HEAT F ux 3

TOTAL 'RI MARY CO ~LANT FLOW

'I TOTAL EEDWATER 1 LOW 5

TOTAL

>TEAM LINE FLOW 1

1~

L]:..---

~

l i 2

Y\\

')

3 3

I.:> \\

(\\ "

i~

\\

I v v

-~

~

---.....:. ~

\\

---.::. ~

v -

~

\\

~

~--

\\\\

/

~

\\i

'\\

"-z __

'\\

\\

~',

\\_

~

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

---~

1 2

1 1

1 8

12 16 20 24 28 32 36 40 TIME, SEC FiS~RE 3.71 POWER, HEAT FLUX, AND SYSTEM FLOWS, LOSS OF LOAD - CASE 3 SEQ.

378 17 JUN 7? OS 11~3 rJJ2 I

C)

O"l I

z I :z,,

I

-....J I

co

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-~~~

I l

AVE.

F1lEL TEHPERIATURE 2

CORE [ LET TEHPE"ATURE 3

AVE. C DRE COOLAN TEMP.

~

CLAD T MPERATURE

\\

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

I\\

i120~u~~~r--~~--t~~~-t-~~~t-~~._-~+-~~~l--~~-f.-~~~.+-~~___,~~~_J 1oop~~~~r--~~--t~~~-+-~~~-t-~~~~~,:--~~1--~~-+-~~~-!.-~~~~~~_J 8oor-~--t~~-+~~-t-~~-+-~~+-~~~~~~,,..--.+-~~...!-~~L-~_j

~r----_ l r----r1--

lj 600

~

II t:

4 0~'--~--:4:"--~~8-::--~~~1~2~~~1~6~~-2~0::--~_J2_4~~-2L8~~-1..~~--1~~__J 32 36 40 TIME. SEC FIGURE 3.72 CORE TEMPERATURE RESPONSES, SEQ.

3? 8 1? JUN 7? 0 9 : 4 3

4 4 LOSS OF LOAD - CASE 3 I

C)

-.....J I

z:

I :z:

"'Tl I

-.....J

-.....J I

OJ

LL (9

w 0

30 20 10 0

1 2 PALISADES ** LOSS OF LOAD FROM 2580. 6 MW & 2010 PSIA 1

2 3

IN AVE.

RIMARY CO LANT TEMP

LOOP 1 IN AVE.

RIHARY CO LANT TEMP

LOOP 2 IN HOT L G -

COLD EG TEMP.

IFFERENCE LOOP 1 IN HOT L G -

COLD EG TEMP.

IFFERENCE LOOP 2

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

-4~

'+

8 12 16 20 TIME.

FIGURE 3.73 PRIMARY LOOP TEMPERATURE CHANGES, LOSS OF LOAD - CASE 3 24 28 32 36 40 SEC SEQ.

3 'i' 8 1 7 JUN 'i'7 0 9 1 'I 3 1 4 5 I -'

0 00 I

z I :z "Tl I

-.....J

-.....J I

():)

(-/ -

(-:*

,.-ii\\ --

1..

--~ --

~

a:

1-1 en CL PALISADES++ LOSS OF LORD FROM 2580. 6 MW & 2010 PSIA 40or--~~-.~~~~~~~~~___,.~-==---..-~~-,--.--~~~:._:::_:___:_:____;,__::....:...:..:__~~~~

LOOP 1 0

-3~

'+

8 12 2

3 16 20 TIME.

LOOP 2 24 28 32 SEC 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 3

36 40 I

0

\\..D I

z I z..,,

I

"-..J

"-..J I

co

?5 50 25 cnO w

I u z

-25

-50

-?5

-1~

PALISADES ++ LOSS OF LOAD FROM 2580. 6 MW & 2010 PSIA 1

CHANGE IN STEAH EN.

WATER LEVEL. LO P 1 2

CHANGE I~ STEAM EN.

WATER LEVEL. LO P 2 3

CHANGE IN PRESSU IZER WATE LEVEL 3 ~-,.._-

3 3

4 8

12 16 20 24 2 8*

32 TIME.

SEC 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

0 I

z

I ;z:,,

I 36 40

-....J

-....J I

co

I I

-1 I

I I

1, I

I I.,,

I I

-I.

1 I

-111-XN-NF-77-18

3. 7 LOSS OF FEEDWATER FLOW -INCIDENT A loss of feedwater flow incident may arise due to the rupture of 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-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 is unchecked, the normal primary coolant system heat sink would be reduced and eventually eliminated.

The result would be an increase in core inlet temperature with only the presssurizer relief valves and auxiliary feedwater system available for the removal of decay heat, until a controlled system cooldown is initiated.

The reactor protection system provides reactor protection through a reactor trip activated by low water level in each 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 condition requires the most rapid response from the reactor control and 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 below the nominal value of 2060 psia.

The feedwater flow is reduced to zero in two seconds.

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

-112-XN-NF-77-18 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 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 accommodate decay heat and pump heat for an additional 15 minutes, at which 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 incident to restore partial feedwater flow by activation of the auxiliary feedwater system.

I I

I I,,

I I

,I I

.I,,

I I

I

.I I

I I

140 120 100 0 w l-a:

crso LL 0

1-z t3SO er w

a...

40 20 0 0 PAL I SADES ++ LOSS OF FEED-WATER FLOW FROM 2 5 8 0. 6 MW 1

POWER EVEL 2

HEAT F ux

3.

TOTAL lRIMARY CO ~LANT FLOW lj TOTAL EEDWATER LOW 5

TOTAL :TEAM LINE FLOW 1 z.......

5 1 t) 1 "

~ \\\\

3 3

3

~

3 3

\\

r----....._s

\\ ~

\\

E \\

\\

~II

\\

~5 5

~

\\ ' ~

1 ~

-r-l--_

~

2

, 2 1 2 lj

'I lj lj lj 4

lj lj 10 20 30 40 50 60 70 80 TIME. SEC FIGURE 3. 76 PQ1_4ER, HEAT FLUX, AND SYSTEr1 FL0\\4S, LOSS OF FEEDWATER FLOW SEQ.

784 20.JLJN 77 13 I 22 I 17 3

5 1 2 lj 90 5

100 I.......

w I

z

1. :z

"'Tl I

'-J

'-J I

co

_J

LL l'.D w 0

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 1

l l

-\\

u

\\

14 0

\\

J

\\

u

\\

il LI ti

~

-L______

ll 1

I

.j

,J v

,J - -

~?tr 1

~ ti I) 3 4 c:

') 3 ti

') 3 4 120 100 800 600 10 20 30 40 50 60 70 80 TIME. SEC FIGURE 3. 77 CORE TEMPERATURE RESPONSES, LOSS OF FEEDWATER FLOW SEQ.

'i' 8LI 20 JUN '('I' 13 122 I 18 l

') 3 ti 90 100 I __.

+:>

I z

I z I

-...J

-...J I

LL

(.9 w

0 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

~

1 2 0

3 q 3 lj

~

\\

t'* ~

- ~

~

r---_3 LL 1 I)

-............ ---L.L ~

--l._2 3 11 3 lj

-20

-40

-........ ~

-6~

10 20 30 40 50 60

?O 80 90 100 TIME, SEC FIGURE 3.78 PRIMARY LOOP TEMPERATURE CHANGES, LOSS OF FEEDWATER FLOW SEQ.

78'1 20 JUN 77 13 122 119 I

U"1 I

z I ;z,,

I

-.....J

-.....J I

co

a:

en 0...

800 soo 400 200 0

-20

-'+ 0

-s~

PALISADES ** LOSS OF FEED-WATER FLOW FROM 2580. S MW

l.

STEAM I DOME PRESSi ~RE CHANGE i....OOP t

2.

STEAM DOME PRESS' ~RE CHANGE LOOP 2

3.

PRESSU RIZER PRES >URE CHANG!

1 2

/..-

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

~

l 2

~

~ L?

-rW---_

12~

1 2

~

L..?___

""- ~

u

~

1---1.--_ --.1.__

--.1.__~

3

~

u

\\

u 20 30

'f 0 50 80 70 80 90 100 TIME. SEC FIGURE 3. 79 PRESSURE CHANGES IN PRESSURIZER AND SEQ.

7 8i 20 JU~ 77 13 :22 *2i.

STEAM GENERATORS, LOSS OF FEEDWATER FLOW I __,

O"\\

I

z I :z "Tl I

-....J

-....J I

--11--~ -.*, -

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

50 0

10 PALISADES ** LOSS o~ FEED-WATER FLOW FROM 2580. 6 MW 20 30 3

CH~~GE IN STEAM EN.

WATER LEVEL. LO P t CHANGE IN STEAM EN.

WATER LEVEL. LO P 2 CHANGE IN PRESSU iZER WATE LEVEL

~o so so

?O 80 TIME. SEC FIGURE 3.80 LEVEL CHANGES IN PRESSURIZER AND srn.

784 20.Ji.JN??:.: :22:22 STEAM GENERATORS, LOSS OF FEEDUATER FLOH 90 1 0 iJ I __,

'-l I

x :z:

I :z:,,

I

'-l

-....J I

-118.:.

XN-NF-77-18 3.8 STEAM LINE BREAK The break of a steam line results in a sharp reduction in steam generator inventory.

The resultant pressure decrease causes an increase in energy demand from the primary coolant system which reduces primary coolant. temperature and pressure.

With a negative moderator reactivity coefficient the core would return to a power level, following reactor trip, sufficient to cause core damage if unchecked.

The steam line break was analyzed from two initial reactor conditions:

l) 102% of rated power (2530 MWt) and 2) hot zero power (rods at power dependent insertion limits).

Both cases were analyzed with two loop (four pump) op~ration as required by the technical specifications for these operating conditions. fhe steam line brea~ was assumed to occur 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 coolant. Critical flow was assumed at the break and was calculated using

( l 0) the Moody curve.

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.

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

This relationship Ii I

I I'

I I

I,,

I I

I

I

I I

1 11 I

I

1.

I I

I

.:-119-:

XN-NF-77-18 is for end-of-cycle conditions when the moderator coefficient is most negative.

For conservatism the slope of the moderator reactivity curve 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 decreased by 20% to minimize the negative reactivity feedback during the transient.

Minimum capability of the boron injection system was assumed which 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.

The HPSI flow versus pressure curve used is shown in Figure 3.83. A low pressurizer pressure signal (1571 psia) initiates the safety injection 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 pressure safety injection pumps has been accounted for in the analysis.

The concentrated boric acid (10,940 ppm) introduced by the charging pump~ was conservatively assumed to reach the primary loop 80 seconds after the safety injection signal.

No credit was taken for the effects of the resident low concentration borated water being swept into the primary loop from the lines of either system.

The critical heat flux was determined using the modified Barnett CllF correlation(J) for these transients.

The local peaking factors occurring

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 of the local peaking factors.

The negative feedback effects due to voiding were not used, however~ in the determination of overall core power.

-120-XN-NF-77-18 3.8.l Steam Line Break From 102% of Rated Power (2530 Mwt)

The system responses to a steam line break initiated from 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 steam isolation valves close at 7.6 seconds~ after which steam loss is confined to one steam generator.

Borated water from the high pressure, safety injection system reaches the core after 35 seconds.

Boric acid introduced by the charging pumps reaches the core after 96 seconds, terminating power increase.

The steam generator associated with the ruptured line empties at 126 seconds, after wh"ich cooldown of the primary system is essentially terminated.

The peak power reached during the transient after control rod insertion was 464 Mwt at 15 seconds, Core average heat flux peaks at 30,960 Btu/hrMft2.

At the time of maximum core average heat flux, the margin 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 temperature) and applying the corresponding local hot rod peaking (F~ = 18.2),

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.

3.8.2 Steam Line Break from Hot Standby The system response to a steam line break initiated from hot standby with a shutdown margin of 2% Ap are shown in Figures 3.90 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 seconds.

The maximum core average heat flux is 45,530 Btu/hr n 2.

Borated I

I I

I

-121-XN-NF-77-18 water from the high pressure safety injection system reaches the core at 35 seconds.

Boric acid from the charging pumpings reaches the core region after 96 seconds, terminating the core power increases.

As with the break from 102% of rated power. the lowest margin 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, and 376°F inlet temperature), the minimum CHF ratio was calculated to be 1.41 with a local hot rod peaking factor.of F~ = 16.0.

r 1.06

1. 05 -

l.04 l.03 4-4-

Q)

~

>, l.02

.µ

1. 01 u ro Q.J c::::

l.00

.99

.98

.97

1. 0 0.9 0.8 Core Average Density (gm/cc)

FIGURE 3.81 VARIATION OF REACTIVITY WITH CORE AVERAGE DENSITY AT END-OF-CYCLE 0.7 I __,

N N

I x

:z:

I :z:.,

I

'-I

-.:..J I

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

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

<J I

.µ

r- '

r-

.µ u

co Q) 0::::

s::

01

' *r-Q)

Ol s::

co

..i:::

u 0

20 40 60 80 100 Power (% of 2530 MWt)

FIGURE 3.82

VARIATION OF REACTIVITY WITH POWER AT CONSTANT CORE AVERAGE TEMPERATURE 120 140 I --'

N w z

I z

"'Tl I

-...J

-...J I

c..

en 3

0 r-l.J....

l 000 500 0 l--~.1....-~.J......~-L-.~....L....~......l-~-'-~~~----'~~"'--~......... ~_,_~...._..__......_~~~--

o 500 l 000 l 500 Primary System Pressure (psia)

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

N

..j:::>

I

z:

I :z:.,,

I

'-I

'-I I

co

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

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.

TOTAL... EED;.JRTER f t...ow 1201--T 1 ~~+-~~~+-~~-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 100 D w I-a:

cr:g 0 LL 0

I-z w60 u

er:

w 0...

40 20 0 0

, __ ~

I I

I I i I

l I

\\N

!. 2 \\

J.3----µ1 _.=.£-~1"----IT"""

1 :

I

\\)------'lj-----1--___:__lj __._ __

_:4:..__iL........-

~

4 1

I

\\

20 40 60 80 100 120 14 0 T H1E, SEC 1 2 4*

160 FIGURt 3.SJ POWER, HEAT FLUX, AND SYSTEM FLOWS, STE.4.~i LINE BREAK SEQ.

11'03 23 JUN 71' 16 *29 :QI)

I i I

I

~ 2

~

l 180 200 I ___..

N

()1 I

z I.

z,,

I

-....J

-....J I

co

LL C9 w 0

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

I AVE.

i: ~EL TEMPER~TURE I

I

2.

CORE [ ~LET TEMPEf RTURE I

3, AVE. C DRE COO~RN TEMP.

~.

CLAD T MPERATURE

]

I I

I

'-Ji I

1()0 14 0 800

~ \\

IJ \\

.\\

I I

\\

I I

I\\

"'i j

~~- ~1 1

l l

l I

r-1--_

~":

t

'.l I

'.l 4

'.l 4 3 "

I 1

4

~ 1 4

') 3 4 r

I I

I 12 0 100 600 Lt 0 0 20~

40 60 80 100 120 20 14.0 160 TIME, SEC FIGURE 3.85 CORE TEMPERATURE RESPONSE, STEAM LINE BREAK SEQ.

L 703 23.JUN 71' 16 129 *OF3 t

I) 3 4 180 I i I

I I J

I I I I

I 200 l -*

z I :z I

-....J

-....J I

co

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

LL CD i...LJ 0

40

!J

-40

-80

-12

-16 i I

(\\\\_____

~

'\\

~

u J

PR~ISRCES ** STEAM LINE BREAK FROM 2580. 6 MW I

I

!.J CHANG. [ N AVE.

~R rMAR'f CO Di....ANT TEMPj.

LOOP 2 i

I 2

CHANG IN AVE.

PRIMARY CO b:..RNT TEMP*. LOOP 1 I

I I

3 CHANG IN HOT L G -

COL.. O EG TEMP.

)!FFERENCE I

I L.OOP 2 i

i

~

CHANG IN HOT L '.:G -

COLD EG TEMP.

)IFFERENCE LOOP i I

I i

I I

I i

I

~

l 3

I

?

r---.::__

3 I

3

\\

3

?

~ 4 I I

4 4

I 4

4

~--

j 4

~~ ~

~

~ ~

?

~

,____2 2

2 2

I 1

1 l

r----_ ~

. -2 ~ u 20 40

[) 0 80 100 TIME. SEC 120 140 160 180 FIGURE 3.86 PRIMARY LOOP TEMPERATURE CHANGES, STEAM LINE BREAK SEO.

i ?' 0 3 '23.JUN I ?' 1 {) : 2 9 : 0 9 I

i I

I I

I 200 I

N

-...J I

x ::z I ::z "Tl I

-...J

-...J I

(f.l CL PALISADES ** STEAM LINE BREAK FROM 2580. 6 MW 800

~:I STEAM

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

3.

PRESSi.JI IZER PRES: SURE CHRNGI I

l l

I 400 0~

I

\\

I

~

\\ \\

~

/I~

~

i 2

~

--~

2 r-1--_

2 2

1 I

1 1

0

-40 l

I

'\\

1 1

\\

1-l \\

I I

I L -

r-.J 'J

~

i

-80

-12

~

3 3

~ 3 3

_t---

'-! \\..i i

-16 I

I l

I i

I L... -

. lJ lJ r

-2~

40 50 80

.oo 120 14 0 160 TIME. SEC FIGURE 3.87 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, STEAM LINE BREAK SEQ.

l !'03 23 JUN ?7 16: 29 I 10 2

1 3

180 200 I

N co I

z I z..,.,

I

-....J

-....J I

co

c.nO w

r:

u z -

3 3

1 1

14 0 160 TP"1E, SEC FIGURE 3.88 LEVEL CHANGES IN PRESSURIZER AND STEAM SEO.

iro3 23 JUN 'i"i' 18*29'12 GENERATORS, STEAM LINE BREAK I

i I '

2 I

I I

I I -'

i N

\\.0 i

I

' I l

3 I I I

I i

x

z I

I :z "Tl I

180 200

-....J

-....J I

co

13 10

?

4 (J) a:::.

a:

_J 61 0

-2

-5

-8 0 PAL I SADES *

STEAM LI NE BREAK FROM 2 5 8 0. 5 MW t

l"lULJt:.t'(A UI"< Rt.Ht.: I I Ill 2

DOPP LE R REACTIV I y 3

BORON I REACTIVITY l\\

TOTAL I REACTIVITY l

1 1

.J----~

~

y 2 /

2

')

~

J 3 3

3..

., tj tj tj tj I

3 v---

I 3

3

~

i v

I

, I /

I I I i

I !

I 1v I

I I

I !

I I

20 40 GO 80 100 TIME. SEC 120 14 0 I

I I

I 1 1 I l I

I 2

I I

I tj 3

I l

I I I I i

I i

I I

180 2

4 3

FIGURE 3.89 REACTIVITY FEEDBACK, STEAM LINE BREAK srn. tr o 3 2 3 JUN n u;

2 9 1 13 1

2 tj 180 200 I __,

w 0

, I

z I :z.,

I

-....J

-...J I

14 0 120 100 0 w I-CC

~80 LL 0

1-z t}SO a:: w CL 40 20 0 0 PALISADES ** STEAMLINE BREAK FROM HOT STAND BY l

POWER

.EVEL 2

HEAT F ux 3

TOTAL

~R I11ARY CO bLANT FLOW TOTAL EEDWATER LOI-I 5

TOTAL HEAM LINE FLOW 3

3 3

'3 3

3 3

r

~

""-~

r-----_p_L 5 l 2

~ t 2 c 1 2 1 ')

5 1 2 5 l 2

~

~J p

2

'l lj If q

q lj q

q y

20 40 60 80 100 120 14 0 160 TIME. SEC FIGURE 3.90

-POWER, HEAT FLUX, AND SYSTEM FLOWS, STEAMLINE BREAK FROM HOT STAND BY

.. SE~. 396 214 JUN 7r 10 1qQ *36 3

I 5 1 2 5

lj 180 200 I

w I

z I :z

"'Tl I

-...J I

co

LL

(.0

~

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

l l=IVE.

I=" JEL TEMPER TURE l

I I

2 CORE II lLET TEMPE ~AT URE I I 3

AVE.

C DRE COOLAN TEMP.

I I 4

CLAD T' HPERATl.JRE GOO J-1 i

/

r-----__

r-1.._

l

/

5 i; 0 r-----

I I

~

I i i--1

'\\\\

I/

J I

~ u I

I l

~

520

~8:

11

~

I t I

~

i I lj r--.

'I

'i

~

~3

. 4

'I

~

3 3

'I I'\\

3 3

L

.* 2 I

2 3

2 2

20 40 80 80 100 120 14 0 16 0 TIME. SEC FIGURE 3.91 CORE TE~PERATURE RESPONSES, STEAMLINE BREAK FROM HOT STAND BY SEQ.

3 9 S 2 4 JUN H 10 ' 'l 0

3 6

-~

180 4 -

3

~

200 I __,

w N

I z

I 2 '1 I

-..J

-..J I

co

LL

(!)

w D

80 40 0

-80

-12

-16

-2~

PALISADES ++ STEAMLINE BREAK FROM HOT STAND BY 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 4

CHANG c-IN HOT L G -

COLD EG TEMP.

)IFFERENCE LOOP 1---1_

\\

v

<I

~

3 3

ti 4

4 4

I

z:
I :z:
'Tl I
--.J
--.J I

~---------

0
-50
-10
-15 (JI w I u z
-20
-25
-30
-3~
PALISADES
~+ STEAMLINE BREAK FROM HOT STAND BY
\\\\
l.
CHANGE IN STEAH 1 ;EN.
i-JATER LEVEL.
ri~o 2 2
2 2
2 CHANGE IN STEAH 1 :EN.
lolATER LEVEL. :...ono *,
3 CHANGE IN PRESSUI ~IZER 1-JATE! ~ LEVEL IJ
~ u u
lJ
,)
~
IJ
\\
"' ~
\\
3 N..
3 3
3 3
3 3
3
~
~
~ ~
... J
~
~
~
~
20 40 80 80 100 120 14 0 16 0 TIME, SEC FIGURE 3.94 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, STEAMLINE BREAK FROM HOT STAND BY SEQ.
3 9 6 2 4 JUN ?'i' 10 '~ 0 : 4 2 3
~
180 200 I.
w Ul I
z:
I ::z
"'T1 I
--...J
--...J I
10 (f)'
O'.:
er:
_J
_J 02 D
0
-2
-4 (1 PALISADES ** STEAMLINE BREAK FROM HOT STANO BY 1
HO DERR OR REACT! llITY
2.
DOPPLE R REACTIVI y 3
BORON REACTlVITY TOTAL REACTIVlTY 1
1 1
1 1
1 1-------
v I
(
I I~ "
2 3 f--_
lj lj 3
ti
4.
lj ti I ~
v 3
_}
3
~
3 2
_2 2
2 2
2 I
20.
40 60 80 100 120 140 160 TIME. SEC FIGURE 3.95 REACTIVITY FEEDBACK, SEQ.
3 9 6 2 4 JUN 11 10 ' 4 0 1 ti 3 STEAMLINE BREAK FROM HOT STAND BY 1
lj 3
2 180 200 I _.
w CJ)
I
z:
I :z:,,
I
--.J
--.J I
I I
I I
I I,,
I I
I I
I I
I I
I I
I I
-137-XN-NF-77-18 3.9 SINGLE ROD WITHDRAWAL In the unlikely event of the withdrawal of a single control rod, 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 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-ing system would insert control rods to compensate for the withdrawn rod and power level and all other system parameters would remain relatively 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-drawal of one or more control rod banks, the high neutron flux, thermal margin, and high pressurizer pressure reactor trip signals provide protection for the event.
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 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; and one assumed the least negative values for Doppler and moderator co-5
-4 efficients i.e., -.87 x 10-Ap/°F and 0.5 x 10 Ap/°F respectively.
For both cases, the design axial peaking factor was assumed while a radial 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 group 4 rod and with the remainder of group 4 at its insertion limit.
r---**
-138-XN-NF-77-18 As discussed above, the withdrawal of a single control rod from 102% of rated power with the reactor regulating system in the manual mode was evaluated assuming both minimum and maximum feedback.
As the miriimum feedback case was found to* be most limiting from a DNB standpoint, the 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/
minute from its full power insertion limit.
The rod was assumed to be 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, 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 included in the analysis in order to minimize the pressure rise and therefore 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 with an MDNBR of l.44.
I I
I I
I I
I I
I I
I I
I I
I I
I I
I
0 i...L1 l-a::
14 0 120 100 Ct'.:8 0 LL 0
1-z
~so O'.:
w
()__
40 20 0 0 PALISADES ** SINGLE CON~ROL ROD WITHDRAWAL +* MINIMUM FEEDBACK I
l POJ..JER ' EVEL 2
~EAT Fl ux I I 3
TOTAL I RIMARY CO nLANT FLOhl I
\\
TOTAL I EEOhlATER LOhl 5
TOTAL TEAM LINE FLOW 1
1 1
~
1 -
£
\\
t "
1.-
'\\
1 "
l 2 J
q
J
., q 5
~ 'I 5
., q 5
., q 5 q 5
'I
\\ \\
I
~
I \\
I\\.
I cl~
\\,,.,.... v "' "
\\.~A
,vv VfVt---~
I 8
12 lb 20 24 28 32 3b 40 TIME. SEC FIGURE 3.96 POWER, HEAT FLUX, AND SYSTEM FLOWS,
. SINGLE CONTROL ROD WITHDRAWAL SEQ.
ll'F23.JJN 7?'1J4*22,IJ3 I __,
w l.O I
x z I z "TJ I
-....J
-....J I
Li..
(.9 w
0 180 180 HO 120 100 800 800 PALISADES ** SINGLf CONTROL ROD WITHDRAWAL ** MINIMUM FEEDBACK u
I I
l.
AVE.
FUEL TEMPER ~'.i.JRE 2
CORE 11µLET TEMPE ~ATURE
3.
AVE.
C!DRE COOLAN TEMP.
CLAD Tl MPERATURE PJ 1
1 l
~
1 1
1 1
l
\\
PJ
\\
I u
\\
I
' *I\\
PJ I
I I
'l tj tj tj tj 3
3 3
.j
~
.j
£ L
z
£ z
z I
I J.
12 lG 20 TIME. SEC FIGURE 3.97 CORE TEMPERATURE RESPONSES, SINGLE CONTROL ROD WITHORA~iAL
"'~
tj
.j
..j 3
-41 z
. l 28 32
- 'f 0 SEQ.
1 18? 2.3,JUN 77 0 'I ' 2 2 : IJ 4 I __,
~
0 I
x z I z "Tl I
I CXJ
~------------------
PALISADES ++ SINGLE CONTROL ROD NITHDRAwAL ** MINIMUM FEEDBACK 30 I
1 CHANG IN AVE.
'RIMARY C01 DLANT TEMP.
LOOP 1
2.
CHANG IN AVE.
I 'RIHARY COi OLANT TEMP.
LOOP 2 3
CHANG IN HOT L G -
COLD I EG TEMP.
I HFFERENCE LOOP 1 ii.
CHANG IN HOT L G -
COLD I EG TEMP.
HFFERENCE LOOP 2 20 1 2 10 3 4 1 2 1 2 1 2 1 2 -
1 2 1 2 3 4 3 4 3 4 3 4 0
i-101-~~-t-~~---+-~~---+-~~___,~~~-t-~~---+-~~-+-~~----;-~~t---+-~~--i
-2 0 ----------+--------~---*- -- -----+---**-- **-------*---+--------+---
-3 oi-*---t---------- ------*-*-------- ----- ---**--------- *----t----~
- 4 ~ -------4~--- *-*** ***- _8 ___________ i 2-* ---*-18--------20------24-------- --28------*-3 2*** -----3-6.
TIME. SEC FIGURE 3.98 PRIMARY LOOP TEMPERATURE CHANGES, SINGLE CONTROL ROD WITHDRAWAL SEQ.
1187 23 JUN 77 04: 22: 06 40 1*
z I z,..,
I
-.....J I
a:
(.!)
CL 350 300 250 200 150 100 50 0 c PALISADES ** SINGLE CONTROL ROD WITHDRAWAL ** MINIMUM FEEDBACK 1
STEAM I DOME PRESSI URE CHANGE LOOP l 2
STEAM I DOME PRESS IJRE CHANGE LOOP 2 3
PRESSUI RIZER PRES ~URE CHANGI I
/ ~
I I I I
3 ii_,.,.---
~
~
~
~
lJ
__-2--
U3.----
I
---3---
i. 2 l
2 ---
I 2 3 LH--1-- 1 2 1 2 1
~
8 12 16 20 24 28 32 TIME, SEC F1GURE 3.99 PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS, SINGLE CONTROL ROD WITHDRAl~AL SEQ.
118 'i' 2 3.JUN ? ? 0 4 '2 2 : 0 7 3 ---....
3S 40 I
+:>
N I
z I z.,
I
-...J
-...J I
PALISADES ++ SINGLE CONTROL ROD WITHDRAWAL +* MINIMUM FEEDBACK 30
1.
CHANGE IN STEAM iEN.
~ATER LEVEL. LO JP 1
2.
CHANGE IN STEAM EN.
~ATER LEVEL. LOI lP 2
3.
CHANGE !N PRESSUr!ZER WATE ~ LEVEL
.j 20 I
i....-----
~
]--"
~
___3---
~
1 2 1 2
~
1~
1 2 1 2 1 9
~
10 3
1 'l
\\
\\
-10
\\
I
\\
I
\\
-20
-30 8
12 16 20*
24 28 32 TIME. SEC FIGURE 3. 100 LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS, SINGLE CONTROL ROD
\\~ITHDRAWAL SEQ.
1187' 23 JUN 77 Qq 122 rQ9 l
36 I
I i I I
2 40 I __,
.p.
w I
x z I z I
-....J
-....J I
-144-XN-NF-77-18 3.10 ROD EJECTION INCIDENT Hypothetical control rod ejection incidents for the Palisades core were analyzed with the XTRAN digital computer code( 2), a two-dimensional (r-z) program which solves the space and time dependent neutron
. diffusion equation with fuel temperature and moderator density reactivity feedbacks.
Hot full power (HFP) and hot zero power (HZP) core conditions in combination with beginning-of-cycle (BOC) and end-of-cycle (EOC) condi-tions were considered in determining and analyzing the most limiting hypothetical rod ejection incidents.
The criteria to be satisfied for 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 design pressure (~ 2750 psia).
To assure the conservatism of the ejected rod worths and power peaking factors utilized in these analyses these parameters were calculated 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 intended to be inserted at the power level under consideration.
In all cases, the dropped rod was assumed to be in the core quadrant diagonally opposed to the ejected rod, thus amplifying the reactivity insertion and peaking factors associated with the postulated rod ejection.
The ejected rod worths and power peaking factors thus calculated by the XTG core simulator code(g) are shown in Table 3.2.
The ejected rod worths and I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I 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 at BOC-HZP and EOC-HFP conditions.
Values of the significant independent variables utilized in th~
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 fuel to the moderator, and b) assuming no heat transfer from the fuel (adiabatic conditions).
The results of the analyses are summarized in TaQle 3.4.
With 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*
acceptable consequences in that no fuel pellets realize an average enthalpy 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 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 be less than 200 psi. Since nominal system pressure is 2060 psia, the primary coolant system would not be overstressed and the pressurizer relief*
valve (set at 2400 psia) would not open during a rod ejection incident.
Conclusion The above results are bounded by the results of the rod ejection accident analyzed in Cycle I, and are therefore acceptable.
line melting of the fuel is expected following rod ejection.
enthalpy of 250 cal/gm is not exceeded.
No gross center A fuel pellet
-146-XN-NF-77-18 TABLE 3.2 MAXIMUM EJECTED ROD WORTHS AND PEAKING FACTORS*
HZP**
HFP***
BOC EOC BOC EOC Ejected Rod Worth, % Ap 1.02 0.95
0. 17
0. 24 Maximum Power Peaking Factor, F~
12.86 l 0. 56
3. 16 All calculations considered the presence of a dropped rod across the core from the ejected rod.
3.47 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.
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.
I I
I I
I I
I I
I I
I I
I I
I I
I I
I
1 I
I I
I I
I I
I I
I I
I I
I I
I I
I
-147-TABLE 3.3 ROD EJECTION INCIDENT ANALYSES PARAMETERS BOC-HZP Initial Power Level, Mwt l
Max. Ejected Rod Worth, % ~P 1.24 Delayed Neutron Fraction 0.0060 Fuel-to-Clad Gap Heat Tr~nsfer Coefficient, BTU/hr-ft -°F 500/0*
Film Heat T2ansfer Coefficient, BTU/hr-ft -°F,
10,000 Clad Thermal Conductivity, BTU/hr-ft -°F 9.39 Clad Heat Capacity, BTU/lb - °F 0.0773 Doppler Coefficient Multiplier 0.80
Cases run with and without heat transfer.
XN-NF-77-18 EOC-HFP 2580.6 0.60 0.0045 500/0*
10,000 9.39 0.0773 0.'80
-148-TABLE 3.4 ROD EJECTION INCIDENT RESULTS BOC-HZP Non,...
Adiabatic Adiapatic Max. F~ (after ejection) 13.48 13.48 Peak Reactor Power Normalized to 2530 MW 158 158 Average Enthalpy of Hottest Fuel Pell e.t, cal /gm 215 247 XN-NF-77-18 EOC-HFP Non-Adiabatic Adiabatic
6. 77
6. 77 43 39 170 200 I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I
-149-XN-NF-77-18
4.0 REFERENCES
1.
Kahn, J. D., Descri tion of the Exxon Nuclear Plant Transient Simulation Model for Pressurixed Water Reactors PTSPWR,
Exxon.Nuclear Company, XN-74-5, Revision l, May 1975.
2.
Morgan, J. N., XTRAN-PWR:
A Computer Code for the Calculation of Rapid Transients in PWR 1s with Moderator and Fuel Temperature Feedback, XN-CC-32, September, 1975.
3.
Tong, L. S., Boilin Crisis and Critical Heat Flux, USAEC Office of Information Services TIB-8 97
4.
Hughes, E. D., A Correlation of Rod Bundle Critical Heat Flux 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 To Determine the Distribution of Coolant During Steady-State and
}ransient Core Operation, Exxon Nuclear Company, XN-75-21, April 1975.
6.
Galbraith, K. P., et al., Definition and Justification of Exxon Nuclear Company DNB Correlation for PWR's, XN-57-48 (October 1975).
7.
WREM:
Water Reactor Evaluation Model (Revision l),NUREG-75/056, May 1975..
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.
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.
10.
Moody, E. T., Maximum Flow Rate of a Single Component, Two Phase Mixture, Transactions of the ASME, Journal of Heat Transfer, February 1965.
I I
I XN-NF-77-18 I
I I
DISTRIBUTION G. R. Correll I
R. H. Ehlers K. P. Galbraith J. D. Kahn I
R. H. Kelley 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)
I I
I I
I I
I I

ML18347B304

Contents

Text

1.0 INTRODUCTION

SUMMARY

4.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

SUMMARY

SUMMARY

4.0 REFERENCES

Navigation menu

ML18347B304

Text

1.0 INTRODUCTION

SUMMARY

4.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

SUMMARY

SUMMARY

4.0 REFERENCES

Navigation menu

Search