Text

Proposed Tech Specs Allowing Plugging of First Row Tubes on All Steam Generators to Aid in ECCS Analysis
	ML20032E035
Person / Time
Site:	Farley
Issue date:	11/16/1981
From:	; ALABAMA POWER CO.
To:	;
Shared Package
ML20032E030	List: ML20032E029; ML20032E035;
References
	NUDOCS 8111190535
	Download: ML20032E035 (41)
	v • d • e

',

POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-Fn(Z)

LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationship:

n F (Z) 3,[2.311 [K(Z)] for P > 0.5 g

P F (Z) 5. [4.62] [K(Z)] fpr P 3, 0. 5 g

where P = THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1 ACTION:

With F (Z) exceeding its limit:

9 a.

Comply with the following ACTION:

1.

Reduce THERMAL POWEk at least 1% for each 1% F (Z) exceeds g

the limit within 15 minutes-and similiarly redece-the -

i Power Range Neutron Flux-High Trip Setpoints within the next 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s: POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit. The g

Overpower AT Trip Setpoint reduction shall be performed with the reactor in at least HOT STANDBY.

b.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the~ reduced limit required by(a, above; THERMAL POWER may then be incre provided Fg within its limit.

f

..[

~

i t

FARLEY - UNIT 1 3/4 2-5 8111190535 811116 PDR ADOCK 05000348 P

PDR

~.

i 3/4.2 POWER DISTRIBUTIO!i LIMITS

~

BASES The. specifications of this section provide assurance of fuel integ-iL rity during Condition I (!iorral Operation) and II (Incidents of Moderate Frequency) events by:

(a) maintaining the minimum DNBR in the core > 1.30 1

during normal operation ar.d in short term transients, and (b) limitiiig the fission gas release, fuel pellet temperature _ & cladding mechanical properties to within ass ed design criteria.

In addition, limiting the peak linear p]wer density during Condition I events provides assurance that the initial conditicns assumed for the LOCA analyses are met arid -

the ECCS acceptance criteria linit of 2200*F is not exceeded.

4 The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F(Z)

Heat Flux Hot Channel Factor, is defined as the maximum local 0

heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing the man-ufacturing tolerances on fuel pellets and rods.

p Nuclear Enthalpy Rise Hot Channel Factor, is defined as the.

N aH ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

F*Y(2)

Radial Peaking Factor, is defined as the ratio of peak power.-

density to average power density in the horizontal plane at core elevation Z.

f 3/4.2.1 AXIAL FLUX DIFFERE?;CE (AFD)

The limits on AXIAL FLUX DIFFEF.EriCE assure that the F (Z) upper n

bound envelope of 2.31 times the normalized axial peaking factor is not l

exceeded during either norr.al operation or in the event of xenon redis-tribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under'these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

r FARLEY - UNIT 1 B 3/4 2-1 Amendment No.10

POWER DISTRIBUTION LIMITS 3

3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fp LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

q F (Z) 1 [2.31] [K(Z)] for P > 0.5 A

P F (Z) $-[ 4.62 ) [K(Z)] for P 1 0.5 q

where P _ THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure (3.2-2) for a given core height location.

APPLICABILITY: MDDE 1 ACTION:

With F (Z) exceeding its limit:

q a.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit n

within 15 minutes and similiarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower delta T Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit.

The o

Overpower delta T Trip Setpoint reduction shall be performed with i

the reactor in at least HOT STANDBY.

b.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, above; THERMAL POWER may then be increased provided F (Z) is n

demonstrated through incore mapping to be within its Timit.

FARLEY-UNIT 2 3/4 2-4

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a) maintaining the minimus DNBR in the core greater than or equal to 1.30 during nomal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of.2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local 0

heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods and measurement uncertainty.

Fh Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

Fxy(Z)

Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the norizontal plane at core elevation Z.

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper q

bound envelope of 2.31 times the normalized axial peaking factor is not l

exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high p u er levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target flux difference value is necessary to reflect core burnup i

I considerations.

FARLEY-UNIT 2 B 3/4 2-1

SAFETY EVALUATION FOR CHANGES TO UNIT 1 AND 2 TECHNICAL SPECIFICATION 3.2.2

Background:

Alabama Power Company proposes to plug the first row tubes of all steam generators at Farley Nuclear Plant - Unit 1.

The purpose of this action will be to increase.the power generation reliability of the unit.

Plugging the first row of steam generator tubes will correspond to a tube plugging level-higher than that assumed in the Farley large break LOCA/ECCS analysis currently on file with the NRC. Thus, a revised analysis which assumes a higher tube plugging level must be submitted to the NRC for review and approval.

In addition, as a result of penalties assessed by the NRC against the February 1978 version of the Westinghouse ECCS Evaluation Model, a re-vision to the Fg limit in Technical Specification 3.2.2 will be required.

Farley Unit 2 is docketed as part of this request in order to have an approved analysis on file with the NRC to allow Alabama Power Company an option of-plugging if it becomes necessary due to first row cracking developing on this unit.

References:

(1) FSAR Section 15.4.1.

(2) Unit 1 and 2 Technical Specification 3.2.2.

(3) Farley Plant ECCS Analysis with 5% Tube Plugging of Steam Generators, (Attachment I)

(4) Fuel Rod Burst / Blockage Evaluation Calculation Sheet, (Attachment II).

Bases:

Plugging the first row of steam generator tubes will impact the current Farley Plant safety analysis assumptions. The safety analysis that would be affected as a result of plugging the first row of tubes would be the large break LOCA/ECCS analysis which is described in Section 15.4.1 of the FSAR. The Farley Plant LOCA/ECCS analysis utilized the February 1978 version of the Westinghouse ECCS Evaluation Model. The results of this analysis confirm that the Farley Plant

.g ATTACHMENT I

~

sm 6

JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 ECCS ANALYSIS WITH 5% TUBE PLUGGING OF STEAM GENERATORS m.

i:

i l

~.

1 i

I.

Introduction / Method of Analysis The Loss-of-Coolant Accident (LOCA) has been reanalyzed for the Joseph j

M. Farley Nuclear Plant Units 1 and 2.

This analysis was performed for the 0.4 DECLG break which is the limiting break for the Farley Plant. The following information concerning the analysis supercedes that provided

. in FSAR Section 15.4.1 on Major Rea: tor Coolant System Pipe Ruptures.

The results of this analysis are consistent with the acceptance criteria provided in Reference (1).

The description of the various aspects of the LOCA analysis is given in WCAP-8339(2). The individual computer codes which comprise the Westinghouse Emergency Core Cooling System (ECCS) evaluation model are described in de-tail in separate reports (3-6) along with code modifications specified in references 8, 10, and 11. The analysis presented herein was perfomed with the February 1978 version of the evaluation model which includes modi-fications delineated in references 12, 13, 14, and 15.

NRC has also approved the removal of 650F of the fuel temperature conservatism related to the Westinghouse PAD code as is noted in Reference 16. This change has been included in this analysis.

II.

Results The analysis of the loss-of-coolant accident is perfomed at 102 percent of the licensed core power rating. The peak linear power and total core power used in the analysis are given in Table 1.

Since there is margin between the peak linear power density used in this analysis and the peak linear power density expected during plant operation, the peak clad temperature calculated in this analysis is greater than the maximum clad temperature expected to exist.

Table 2 presents the occurence time for various events throughout the accident transient, for the 0.4 DECLG, the limiting break.

Table 1 presents selected input values and results from the hot fuel rod thermal transient calculation.

For these results, the hot rod is defined as the location of maximum peak clad temperatures. That location is speci-fled in Table 1 for the break analyzed. The location is indicated in feet I

s.

Safcty Evaluatisn Page 2 meets the acceptance criteria. of 10CFR50.46 (i.e., Limiting Break of CD = 0.4; PCT = 21580F with an FQ of 2.32 and steam generator tube plugging level o'f 1.50%). Since plugging the first row of steam generator tubes would correspond to "a tube plugging level of 2.78% and invalidate the current Farley Plant

~

ECCS analysis, Alabama Power Company has performed a revised ECCS analysis based on a tube plugging level of 5.0%. A copy of this revised analysis is included as Attachment I.

The results of this analysis confinn that the Farley Plant meets the acceptance criteria of 10CFR50.46 (i.e., Limiting Break of Cp = 0.4; PCT = 21820F with an FQ of 2.32 and steam generator tube plugging level of 5.0%).

Subsequent to the NRC approval of the February 1978 version of the Westinghouse ECCS Evaluation Model, the NRC assessed Westinghouse plant's penalties associated with the NRC position that the fuel rod burst and blockage model used in the Westinghouse ECCS Evaluation Model was not conservative. The fuel rod burst and blockage model penalties were assessed against the current Farley Plant ECCS analysis as described in Alabama Power Company letters dated January 10, 1980 for Unit I and August 6,1980 for Unit 2 to the NRC. The Farley Plant was shown to meet the acceptance criteria of 10CFR50.46 (i.e., Limiting Break of CD = 0.4; PCT = 22000F with an FQ of 2.32 and steam generator tube plugging level of 1.50%)

by utilizing credit from improved analytical techniques (i.e., UHI Technology) to offset the penalties associated with the NRC fuel rod burst and blockage model.

I The fuel rod burst and blockage model penalties were assessed against the revised Farley Plant LOCA/ECCS analysis as described in Attachment II. The Farley Plant was shown to meet the t.cceptance criteria of 10CFR50.46 (i.e., Limiting Break of CD = 0.4; PCT = 22000F with an FQ of 2.31 and steam generator tube plugging level of 5'0%) by utilizing credit from improved analytical techniques (i.e., UHI Technology) to offset the penalties associated with the NRC fuel rod burst and block-

~

Safety Evaluation Page 3 age model. Thus, a change to Technical Specification 3.2.2 will be required to revise the FQ Limit from 2.32 to 2.31.

Farley Unit 1 is currently under a forced outage due to a turbine generator repair and will be ready for Cycle 4 criticality by mid-February 1982. Farley Unit 2 is approaching one fourth of the first cycle fuel burnup level. The slight reduction in the FQ limit from 2.32 to 2.31 will have no operational impact on Unit 1 Cycle 4 or Unit 2 Cycle 1 which were initially analyzed using a heat flux hot channel factor, FQ, of 2.32. Both Farley Units 1 and 2 are operated well below the FQ limit. The expected maximum total peaking factor F

during normal operations of the reactor including load following maneuvers for beginning, middle, and end of cycle conditions are below the 2.31 limit.

The surveillance requirements necessary to assure that this limit is not exceeded are provided in plant Technical Specification 4.2.2.

Plugging of the steam generator tubes to the 5% level is calculated to result in an increase in the core inlet temperature of 1.6 'F with an accompanying increase in the core average temperature of approximately half that amount.

The design basis for non-LOCA transients is a core average temperature of 577.2*F. The effect on core flow due to the steam generator tube plugging is insignificant. Therefore, plugging the steam generators tubes to a level of 5%

does not invalidate the current non-LOCA transient analyses.

==

Conclusion:==

The proposed change to Technical Specification 3.2.2 does not involve an unre-viewed safety question as defined by 10CFR50.59.

l which presents elevation above the bottom of the active fuel stack.

Table 3 presents a summary of the various containment systems parameters and structural parameters which were used as input to the COCO computer code (6) used in this analysis.

s

_ Tables 4 and 5 present reflood mass and energy releases to the contain-ment, and the broken loop accumulator mass and energy release to the contairinent, respectively.

The results of several sensitivity studies are reported (9).

These results are for conditions which are not limiting in nature and hence are reported on a generic basis.

Figures 1 through 17 present the transients for the principle parameters for the break sizes analyzed. The following items are ne d :

3: Quality, mass velocity and clad heat transfer coeffi-Figures 1 cient for the hot rod and burst locations.

6: Core pressure, break flow, and core pressure drop.

Figures 4 The break flow is the sum of the flowrates from both ends of the guillotine break. The core pressure drop is taken as the pressure just before the core inlet to the pressure just beyond the core outlet.

Figures 7 9: Clad temperature, fluid temperature and core flow.

The clad and fluid temperatures are for the hot rod and burst locations.

11 : Downcomer and core water level during reflood, and Figures 10 flooding rate.

13 : Emergency core cooling system flowrates, for both Figures 12 accumulator and pumped safety injection.

15 : Containment pressure and core power transients.

Figures 14 Figures 16 17 : Break energy release during blowdown and the contain-ment wall condensing heat transfer coefficient for the worst break.

I

, III. Conclusions For breaks up to and including the double ended severance of a reactor coolant pipe, the Emergency Core Cooling System will meet the deceptance Criteria as presented in 10CFR50.46.(1) That is:

0 1.

The calculated peak clad temperature does not exceed 2200 F based on a total core peaking factor of 2.32 2.

The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircalloy in the reactor.

3.

The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The cladding oxidation limits of 17% are not exceeded during or after quenching.

4.

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

IV.

References 1.

" Acceptance Criteria for Emergency Core Cooling Systems for Light llater Cooled Nuclear Power Reactors",10CFR50.46 and Appendix K of 10CFR50.46. Federal Register, Volume 39, Number 3. January 4,1974.

2.

Bordelon, F. M., Massie, H. W., and Zordan, T. A.,

" Westinghouse l

ECCS Evaluation Model-Sumary," WCAP-8339, July 1974.

l l

3.

Bordelon, F. M., et al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-8302 (Proprietary Version), WCAP-8306 (Non-Proprietary Version), June 1974.

4.

Bordelon, F. M., et al., "LOCTA-IV Program: Loss-of Coolant Transient l

Analysis,"WCAP-8301 -(Proprietary Version), WCAP-8309 (Non-Proprietary' l

Version), June 1974.

l s.

^

., References (Cont'd) 5.

Kelly, R. D., et al., " Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code)." WCAP-8170 (Proprietary Version), WCAP-8171 (Non-Proprietary Version), June 1974.

-6.

Bordelon, F. M., and Murphy, E.T., "Contairnent Pressure Analysis Code (C0CO),"kf,AP-8327 (Proprietary Version), WCAP-8326 (Non-Proprietary Version), June 1974 7.

Federal Register, "Tupplement to the Status Report by the Directorate of Licensing in the matter of Westinghouse Electric Corporation ECCS Evaluation Model conformance to 10CFR50.46 Appendix K " November 1974.

8.

Bordelon, F.

M., et al., "The Westinghouse ECCS Evaluation Model:

Supplementary Information." WCAP-8471 (Proprietary Version) WCAP-8472, (Non-Proprietary Version), January 1975.

9.

Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies,"

WCAP-8340 (Proprietary Version), WCAP-8356 (Non-Proprietary Version),

July 1974.

10.

" Westinghouse ECCS Evaluation Model, October,1975 Versions," WCAP-8622 (Proprietary Version), WCAP-8623 (Non-Proprietary Version) November,1975.

11. Letter from C Eicheldinger of Wes og'iouse Electric Corporation to D. B. Vassalo of the Nuclear Regula.sy Comission, letter number NS-CE-924, January 23, 1976.

12.

Kelly, R. D., Thompson, C. M., et al., " Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing large LOCA's During Operation With One Loop Out of Service for Plants Without Loop Iso-lation Valves " WCAP-9166 February,1978.

13.

Eicheldinger, C o " Westinghouse ECCS Evaluation Model, February 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-A (Non-Pro.

prietary Version). February 1978.

4

5-References (Cont'd)

14. Letter from T. M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Comission, letter number NS-TMA-1981, November 1, 1978.

15. Letter from T. M. Anderson of Westinghouse Electric Corporation to Tedesco of the Nuclear Regulatory Comission, letter number NS-TMA-2014 December 11, 1978.

16. Letter from J. F. Stolz of the Nuclear Regulatory Comission,

" Review of WCAP-8720. " Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations," March 27, 1980.

O

%e

..n

TABLE 1 LARGE BREAK - ANALYSIS INPUT AND RESULTS Quantities in the Calculations NSSS Power Mwt 102% of 2652 s

Peak Linear Power kw/f t 102% of 12.37 Peaking Factor (At License Rating) 2.32 3

Accumulator Water Volume (ft )

1025/ tank Steam Generator Tube Plugging Level 5.0 percent (uniform)

Resul ts DECLG Cn = 0.4 Peak Clad Temp. OF 2181.86 Peak Clad Location Ft.

7.5 Local Zr/H O Rxn (Max)%

7.68 2

Local Zr/H O Location Ft.

7.5 2

Total Zr/H O Rxn %

< 0.3 2

Hot Rod Burst Time Sec.

28.4 Hot Rod Burst Location Ft.

6.0 Fuel Region and Cycle Analyzed UNIT REGIONS CYCLE 1

All 4

2 All 1

g

TABLE 2 LARGE BREAK TIME SEQUENCE OF EVENTS DECLG C 0.4

=

D (Sec)

START 0.0 Rx Trip Signal 0.495 S. I. Signal 0.95 Acc. Injection 16.4 End of Blowdown 29.469 Bottom of Core Recovery 39.497 Acc. Empty 50.497 Pump Injection 25.95 26.574 End of Bypass he m

TABLE 3 Dry Containment Data NET FREE V0.

2.3 X l'06 3

fg INITIAL CONDls.s...

' Pressure 14.7 psia

. Temperature 900F RWST Temperature 350F Service Water Temperature 350F Outside Temperature 400F SPRAY SYSTEM Number of Pumps Operating 2

Runout Flow Rate 5550.0 gpm Actuation Time 48.0 sec FAN COOLERS Number of Fan Coolers Operating 4

Fastest Post-Accident Initiation of Fan Coolers 27.4 see STRUCTURAL HEAT SINKS 2

Wall Area (ft )

Composition Thickness (Ft) 1 75000 Steel / Concrete 0.02083/3.75 2

4700

, Steel / Concrete 0.05/3.75 3

69800 Concrete 0.75 4

77000 Steel 0.0058 5

80500 Steel 0.0113 6

47600 Steel 0.03 7

23600 Steel 0.0583 8

11800 Steel 0.1917 9

13275 Concrete 9.0 10 7900 Steel / Concrete 0.0208/1.5 11 12500 Steel 0.0087

~..

_ _ _. _ _. _ - - _ - ~

~'

TABLE 4 REFLOOD MASS AND ENERGY RELEASES (0.4 DECLG)

Time Mass Flow Energy Flow (Sec)

(Lb/Sec)

(Btu /Sec) 30.497 0

0 40.347 0.7941 1033.13 40.947 0.77189 1004.25 45.109 31.61 41,130.5 54.384 69.31 87,585.3 68.684 276.64 137,828.7 85.984 334.91 146,170.4 105.484 343.29 141,398.

126.784 348.94 135,756.8 174.684 358.91 123,586.6 230.784 367.46 113,317.3 297.484 376.09 103,633.3 381.684 386.78 94,609.1 l

l t,

__,, ~ _ _

TABLE 5 BROKEN LOOP ACCtMJLATOR PRSS ANO ENERGY RELEASE (0.4 DECLG Break)

Time (s ec.)

Mass (1b/sec)

Energy (Btu /sec)

' 0.0 4196.253 243382.65

.1.01 3773.851 218883.36 5.01 2857.074 165710.286 10.01 2300.609 133435.344 15.01 1961.240 113751.904 20.01 1730.312 100358.082 25.01 1568.717 90985.61 26.710 1527.974 88622.494 26.712 1527.832 88614.242 26.719 1527.773 88610.851 26.779 1526.339 88527.654 26.979 1521.390 88240.635 27.181 1516.435 87953.252 27.381 1511.550 87669.896 27.580 1506.724 87390.00 27.780 1501.966 87114.012 27 980 1497.251 86840.582 28.181 1492.536 86567.090 28.380 1487.903 86298.356 28.581 1483.272 86029.789 28.781 1478.717 85765.601 28.981 1474.195 85503.318 e

%e

f h

.O O

O O

.O en si 1.4000 J0$[PH FARLE Y UtilT I2 5 PCT TUDE PLUG FE0 70 HODEL ALA 0.4 0[CLC RE AllALY$l$ VITH UPP[R H[ A0 AT Tie 0T GUALITY OF FLUl0 GURST. G.03 til i PEAR. 7.50 FTiel I,2500

.w t"

e en

---^

^

1.0000 c

V l

3 0,7500 es.o b.

l.

8 0,5000

,{

1 0,2500 t

0,0

,~

"g83 2

3 n'ii 2 Rg3 '

O 8 8 888883 k u M "*023hs -

a e

8

2 3

8 3

3 Las

n en..

e M

e a i, d d dL.h s

s e a a s asa.:-

4

..:.; a a.: age e

a e e sen.g

.s a

a ilHE ISECl FIGURE' I

FLUID QUALITY T

7-DECLG(CD = 0.4)

D g

m-

. mi oo m w ece

- wui was j

was

' ww.

j l,

[

war dr n'

_, I t

.$E.t1 w

= 0 o

I t-

8C0'0,1

.=o

= ::

e w es i

o00*0*

"ii ;

1

. *0 m 4

h 3

w owca v

t G

>+

ro w

w on e

w mwo

!;,/~~

91,

g si d8 E a."~s,i ono-

>e

=

uw,

[*o 8 @ *8

$d 8 )

1 l

l uws aw I

.CC w t C'

I 2: O

~

,g n*n

>w OC00*E N

O" l

e 0

0000*2 W

cr Eau l

3 N,

a.

.C.

E o.

th i&..

q DCC4*0

/

i a

00C2*0 f3

/

1 C;CB'O

/

l 1 CC05'O

'iDC)*O e(

ux0 m,0 000;*0 l

l I

8 8

i i

i 7

e

(

(335-ilus1:

4:33:13A ssy:i

=

~

l

.s.

l'

.m I..

l l

r M*CD q

g I,

1 f

' 7 EE 1

p

(

x z

ll W

e

v H

" E.*

I O

==

s w

wE E

tu (w

g, e.

g gg.ge; y

I C

& w **

l Q s'*%

s4 T

z 8

2 5, *.,

w i

a wo

=w e

s g

ao w

en H 4'

g

S,*

1 u

2 J #

p Q

o "..

L tr O pv z-D 8V e g

)

HJ k

CO

-=

f

""e

~5

.lj i

1 o

00 0c2 Esg i

. 3

=;w

)

>w U C

" o C.

.c. a o I

l w

e Ema 3

z g

w*4 a

en SOE E

44

[

42 cr:m l'

a l _,l lj l_l_l_l l

l

= uc

s_

2-n.g e

e m

~~ a s gm+ ass g 3. - psg 3 wsa s s a 8

g33

a e ea a a e ass a e

_m s ;,,

g.

J-sw-21sint!! 1:3313:ss3os s:: vat sy3n

%=, S.

e w

,---,m,

8M0 Rn0 L

E D

0 H

8 7T O

CHT E

T T

A G

RM0 UD LA P E le i

Hf E

C DR E

)

S 4

U(.

't E.

g.

T P PP E

T C

M R0 UO T

U I

P H

S=

T S

T 5I l

V(

ED 2$

RC P(

l 1

G 4

Y:

f O

.OMM EL L

D tUl O OE l

T RC A

a T

l AD CD YE N

E L

E R

R A CO 4

LC F

C i E l

D P

E E

4. E R

S R

OJ 0 U U

G S

AE I

AP F

LR

. Sn?.

e Q.

i g

.4

.O Q

0 0

0 Q

5 0

0 2

2 1

1 5

f.s wg;t*

(

t

.i i,

i 4

4

p r

n R

n n

O' s

e

.a 1.00C+05 JOSEPH FARLEY Utili ! 2 5 PCT TUDE PLUC FEB 70 MODEL ALA 0.4 DECLC RE ANALY5is WITH UPPER HE AD AT THOT OI:E AK FLOW 7 209 I

'l O

E

[

E 5.00E*04 4

f l

2.5eE 04 5,

-i N

=

i

-2.50E 04 *

-5.oe[+04 1

4

-7.50t+04

' 1.ceE*05 i

E 4

li q

.a e.

a e

a Tit 4E ISEC) j FIGURE 5

BREAK FLOW RATE DECLG(CD = 0.4)

O 1

4 j

D D

.O

=

i e

70.000 JOSEPH FARLEY U:3IT I 2 5 PCT TUCE PLUG FEB 78 N00EL ALA 0.4 DECLC RE Af4 ALYSIS WITH UPPER HE AD* AT THOT CORE PR.OROP l

50.G~,0

(

O U.

t

%g 25.000

=

a

. u 9.0 VM-T

-25.000

-50.000 I

-70.000 fl o

o e

o a

3 8

3 o

TillE ISECl

~

FIGURE 6

CORE PRESSURE DROP DECLG(CD = 0 4)

e e

o e **

00'005 C

[

o 4

L

(

CO*CCD e

m 5$

C o2 =an h> o

= = tf w>

tem 4

Q MW 4 4.

c0*C*s o==

D a<-

r O (

M of w

a

$"o O W "T w

A

E o.

wI

_a E wO D '*

u's CW H

e"*

C ze CO

- o JO g

UV e

D E g o --

00*CM MJ CU 3 <:r o ww a

>w Q, C2 w" o Z

d g

-d;

<o b

=

[ o U, w,o w

. a>

E o

7o4 o

6-4 44

.e.e k

C3'CCI e

C e

l l

~~ ~

g*0 e

e a

=

f g

(

I 8

2 g

g t

4 s

fd 5333330) 003 ACH* dM31**f.Y GY13 e

o#

a 9no O'

Ae

$8,

.I e

l L

e E

l D

i y~

F A

0 H

0 5

8

{

7T 7

0 l

Di T

E R

F TA AE L

G'D EnV.

E P

\\

h C

R ULA}

U)

PE T4 1

Hl A

E i

~

DR F E

S R0 UE 1

I E

T P 0

[

P=

0 P

T x

U M

M C

6 I

T ED P

H TC T

5I.

C Wf$

DG 2SR IL 0

I 1S8 UC A.

i p.

1_

Y LE f

58.o FD L

l t

Ul s

A YCR E

E 8

L R

R U

AC T

F A

E L

C HE R

R DE U

P P

G

(

4. [H I

O F

0T J

0 1

AF l

1 A

U L

L i

(

$ 3-4H e

p i.

i l-.

^

j p

p t

l 9o 0

0 0

0, 0

0 O

0 0

O 0

o t.

0 G.

0 7'.

C 5

c 0

o 0

0 0

5 2

o 5

2 1

1 1

l 7

s 2

0 C 0=oM.

s.-

5-(*

2 i

,a4 1

111 4

1'

800'0G I,

GOG*Ct J

E wo i

C O

o E

l-e O

~g m

ex w-a ca r

o t00*CE h

52 av ew-1 O

+

0 g30 w

E5 C

N n

3

-ta

(

yon Eh C$U e

s tu J

~

ne s1 EU l

(

~C=

}

ow Ua M2 5 s-g Coo *0?

i

,3o e

"" w c e w

O3$

E u

3 xw c

B"

- u.

Ro$

S 5t

/

000*Ch 0

1 l

4n I

co e

=

9 9

~

035/311 31 RWA 01J-2

%e e.

._...-..___.~,...,,,r.

,,,n-.-,

,~

l-t 8dL D

D 8dU f

e L

E 0

0 H

0 7

7i ES 0

t RL 0i T OE E

~

F CV T

A E

G 8

E.

L UD LA PE TR I

l NE E I C

Dl E ET)

UE r

! IA4 T P SW.

P T

E f 0

U i

C H nR P

l H

f RG=

T 5I TM V

OD 25 DCC ONC 1

15 OWG Y

=

LDL T

I t

L 3&%

FUC A

NU H E

E A

RED YER E

L R

A C)

L 0

T F

F l C(

1 i EL P DE E

V E

4. E R

S L

O U

J0 R

G E

I AT LA F

AU 8 g-

/

D y

I D

l.l od 0

0 0

0 D

0 0

0 5

0 0

0 7

5 2

0 5

0 3

0 2

1 1

T 5

2 O

'Cta 5 eW,3 s

~

8de O

O 8$.

L E

D O

O M

8 71 0 BH T

E "Y

F T

,A T

G 3$

TI

~

UD NC LA PE EO)

H l

E C

TL4 CR E

SE UE S

NV0 TP I

v A

T E

TE=

RT U

C H

P l

H f

T LD 3I DNC W

OIC 2$

O G

i$

LEL t

V FRC O

tU N S$~

i L

EOE liA RCD A

YE RI E

C L

1 E

RA C!

1 L/

F CN HE E

I P D [t R

E T

U

4. R S

A G

OJ0 F

I O

O A O L

L AF 0d-D

)h p

9o 0

0 0

0 p

0 1

0 5

5 0

7 5

7 0

7 2

0 2

t 1

t t

0 0

0 Cm 5~ q. o3 F

uS O

R.

a3 s

L E

)

0 O

0 NW t

t O

O D

FT O

W BH O

ET L

F B

T A

(

C g.g UD W

LA O)

P E L4 I

H C F.

E C

DR 0

U R

E

(

T P

0=

P E

T U

M l C

P AD I

H T

T LC 5I UC V

MG 2S UL I

CC n

l S

Y CE T

L gd AD I

A NU H A

YE ER 2

L 1

RA C F L E

C HEV R

P DO U

E L

G

4. F I

S O

F J0.M U

AC LC AA nod T

p p

oo p

0 0

E O

0 0

0 3

C.

0 0

0 G

0 0

0 t

9 5

4 2

0

$vs3

>3w f"<

f.

8

,t

.I s!

e l'

i

\\

C

_a m,

w

-+

- _ +

4 e---

M

-Os -

==--

---+

'- T

-M g

N 80

- ;=

M g

e m-2=~

-'N T

=

a s--

d; -

W A

z aA.

-a m_

% ag

~4-

..r*

.a W

. m

-x gg

<ipC _ g

_ _ + _ _. - - _

_4 a-w W -2

&~.

4

_ __. _,* ;._-;,4 3_

e

>c=w o,x

+

.__. a s

-Q LJ4.5 3

-Z 2, 4

~

,.,==_.,

a 9

ft d

-~~-

w.~

a

~

_ ~ - -

WE g =_

gi

- =

a

_,-e=

= m-se e

a

_4 g-no g M_+._-

Iy

-. +. -

LP w --

? -'=-*

a-rg

- =

2-

-.g

-__-+__

- y---

og w

---+

=

-kZ M

'.T.Z ~__L _ Q-

=: _.

]' :}

~~

=== j

-- - =

r..- - - + - ~

$=. - - - * - - -

T_-= =-d =

3=

~ ~ * -

- - =

_ =-_ - ' * = - =

-Q 1 =s --

... 3 =_ 251E =.I E _~ ---_: =~iE

'--- ~ _ '-- - - - =.Si. - :

~~~

=-

EHE---

y

_;-=4

=

.m 3

-g

=*.: ~-

.. _. ~~~--OCT=-

=

~-

.=T~--

222 -' -

- - - -..i_..

==

t--. =t-

.tu b.-

^t --

--:~ -. _ _ - -. :_-=

=.r.

.-a_

y_

.-._z--

.=-

l~

~C

-~' 23 =

C -~.Z'- ':! =

~ ~ -. = ~ ~ ~ ~ - ' Y. -

==

~~--*

.L]- -'

=~**.!.~~*--

.- !=

---' = =

L+-

~.~ 3 =

--2: -

-~"--*

P-~

22

~= ~ ~ = '

' - ~ "

._*21- ;==

~C"'

=.=

_g=

=

.-. -b b--

--b' Zb= - --

(,

bb b b...bb k ** --- p.=._ -'C-".-

.-l~~ =

* *. "=

C

=

~= =

=

= ~ ~ "

~;-

C"~-

6

--- Q. :- ~~

r. un

--~'

- - =. -.=-

'"*E

(

=

."*:.!*J r_]

*.C~:

C* -'.C*' M *."Z -.~4

~

0-"2=

55 T~

5 '-t r iEir-E5 5 23-s i

55 NN NN W

'~

~#-'~*~~~'==

~~

d~

=C

.j-@ne.O C-1,

26]CI== -~=4

. 22C J

=

U "n 2:1x=

=

.g:c.CJ q=5=b *=OI:

=Q=g =un

==

n :

a.-'

a

~-2:43*C" :*

4o

=

==

. :Q==

-~*g;E: - :

n.:==

=

= :=+---* -- 4: :-

= =*

10==*

r a

- == =r-.... + _ _

- N"=N

== :~~~':

=*.D"=*C= =r} C : U3 = = CO _ :: 25 10

-.-!U T

47 h

"'It =- 22-*

=**."-

===:=-.J

x.: -.n =

.1 = =1:2==.::1=

~I'E: i~.

.=: CL.--

-ef.C==

=J

w.._ =

.y

.=_

(.

. _..=... E rr 2:

-.. =...

._=J = _._J===== =:- ;! =

. _ ~ --- _ -

O

=

--: =12 n

2:n

=

::n.= 2:4= =

=

__=_

__]".}~f'=]y[h~_Cib ~EE !=}= =%

3

-m = _3 "O

...g_

==

i=

$ %f

__ [MhhMh

{M

]~2

=

--.._._a=

-==t=

=

=_._-

==

=

==

- - - - =.

==.

- =.__._r

n. _..:n

=1 n

=*- =-'- ~-~ 'g;;

==..

_u;;j== _g-.._.:

-a--

a:= =

=

= =

====-.;

=

==

~~;1 ~~'-

=

=_r -.=.

..=

. =. _

.l.-.

'- ':"%...-. = -.===

.... =

._..:.=.

_= =.. =. - =-- =._. - -l.

.=.....=n. ~.: :

... =...s.r..d =.-..:]=

=. -. = - : d=-- =-- =--- _- j :

a

%e

6

[

-M

4.

_m.Teie

.ee.i

.e

=

sum..ud

-mm.-..-

94 P ummum' p

_gy=

-m-

-+*++.v-

-++ g;7;-

~

1 e.

e.u.m.

.-y_

2 2

e a-E

-r-

-o y

~*

. L'1 -'

H

-4 m

[

-o.

--+-

- ]-

M

~~*

p-

~._. h, P

_.a_

qu

=J. _ _~ !

--+

C*

__m,,

eo

_.C -

p

_a-y

._=__.

-._=

w p

w _-

h n

" ;.; C *,.C.

_4 C_[. _

z e

--^

- 1: -

-J 4=;- --j w

7- -

4

=._

.A,. __

3,._

4_

A_

r

_._4

_y I ---

~?~~ 55f-m sq m

?C^-

k,,iq

.M-

'i a

d.__

m

. y=

-*-'--*7~~e--

g

,1 t-4 -33.5

-~

-r-a._

-==

m

-+

w=

N n

--+~,+y.,,.,..__Q_V__=._

~ i.

^.-

-- ; f &

.._t_*'-

. n:

'] = r-!rh'C g

_4 -

_u_

n:f.-

-_ M-"s---- _o..

iM _a E L.J =

-+

=

--+

i

.. --g==c _-

4

us --* m

_9..

_ 4. =--- " ' M-- - - - '

~~ ~ N9

~

.._e-.-.. _ _

.+-g,,_

-.4__

m, c_

. _ _,._ g _ ggQ --

s

-p

+-;,

"C' m _..d 1

f w

j., r. - u

--.4 --

-o

=r----"--

" - - ~ ' ' - - -

,"f.-'C* 21

-- i

__ g -

e C'.- - 1La

--s=4 =

J

_a__

3

'. m

- - - + -

f

,f

^^O g.-

___i

--+-

1

---+

_4

-;=

-4

---_i-y o:

_-j __

p-p._

iN f

4-

.M _ _

=3--

-~

g m

' -- ~.-.'

'*"1~~

' _ '., ~

_4

~

g

_.- If

=

=._.g=._

--"I -- - - + -

(,0

_f. _-

g _..

_4

- _.. g_

- _.7

-a

.w f

=

--*L..

'~

.d

-t=

Z

._g-+. ---j_

3 33

-.!EC CZ

.---+ ; r:=_=n__

=__ = :--

==

23_- --

n =-=: ~~5== =._.r--. -

=

_.- g-._=_-__-= __ 6,. _ _=_-_._

3

=

3 4_

i h

n* =...- - -I _- _.-.u__

_w

.=

e

=

=.. _

==

e

- - - * - - - -._:;n --$=. _ _2;-__

' - - - -. CC =. - - -----= La

.C*" - *. -.--

~I T.

.L

- ' ' - ~

-- -~

_n

=

2_**= _"_7 7._-'. - ~ " -

-O7.*-Z. - !_~~-.T.*.-_.-'.'2=_._ -] =_.. 7,.=_'

_'~_-*_:_-'. _ ' '.'.. ' '.

- ".__C*-' =...= =C**'I-T =u *-.:.-*".=

(

=E =E =;=-

Ir.. __i.s. _ =}.E_1..';=======

1 s E._s=

== a

.=1E =g=

==.. _

._ a_

(

_~~1= m..,=

=2-

-.4.

4, ~1 =. _l. =.........p..._.:_ _ -. :r._., _..

...=....

==.=. = --t=_ =._i=___ _. _

= _..=. _ =.

=3.=_...=_2.__-

.a__

=.:=l:~2

'. J3 C."~.

rrk*!*_ l=,= :s _.g g

= o';

J*~

=.

~~,.l***

~Z*.

." $ I *.!*.

C.."1_..

sic

-~ -~

'~ ~ I'c-"1*:-t = -.7 7

_ _.1"-"]

J ' * *~~.l.

~' _l o'* -"

'~

".C'$7Z

- 3=

==

O

!l E}=, =ggg gg_Q =s= = Lc

;

o

_1_ =4 :

_Q==,=

1g _.Ts-

= Tm^*

=_=.

=* = ^ ; =-Q====;

1 --=

=1,====1=-

=

T'====

==

=.

_._ _. =

==

=

c==

.

2=_

,' :==

w-

=

==

=.F=.'_- +.==_ ;;======.O-['.="

223 =

. ' ~ - -

.:: f.? 'l=.=2 ;

g =.a =

.- _4...CC+CC 21==

0L ;r =

n :h":- =- r

\\

=

= =

=

-1

a_

= - -

- =- -:=

=

2.._==_.. Fr__ _n_.r._.=..;.=._.=...a_=.sa.__(9.r. s..a)_3.8_1..s.s..3_M_

LNMe..il.V...EN..O..W_

E...._E

.a.. _=..Ea.. s._..=. _a,a_..

I.s_i._a=

~~

~

7 ='

2 C' IIC

- l

=

'~.t=

'*-._=._

C.

- M.=

' ' ' * ~

' ~ '

[=1.

$.[C UN*. ~ ".

".*2,,Z

.=_**d.

~

~C

~ ~ *

=l

-*=

7;*;.;-' -* ;=

-*I;*: :. ;--;=

_-- - n Nrj =

-- ; =

;;r;I.. u _. -*===

. 3

- d

-E---

&~

~--40

=

0;

=:.; __===-e

=

-- =

2S =.2

._. 2.5 n;n:;n

-'=

==

._.=* 3 =_

-7=.-

~~

=

n----- =4xn

=

_g = =12-- :-- -

_-_.3 =.-1==

- - - =

=

C93 D

9 g.3 D

o si L

E D

O M

O D

FT O

tH iT F

s T

T A

G 39.2, N

E UD LA*

I S

P E C N)

.H l

E E

A4 DR S

U E

(

R T P T0 P

C T

U H

C T

R=

l P

H E

f 5i WD V

OC 2$

. P(G l

EL t

5 Y

RC T

L

.R9C IM A OE D

A

's CD U N YC R

E L

R 5

A C L

1 F

C HE P D ER E

4 U

S OJ 0 G

R I

AW F,

E LO AP Q.2 D

ad D

0 0

8 C.

4 2

0 D

l 0

0 0

9 0

a sS u w > a e.

f.

e k

i 3

,4 i1

, i, 1

j

1' 1i3 4

,j1 i: s

.,ji i

l'1

aRP.

I l O

O e

=M$

e L

E D

0 O

H 8

7T 0

BH T

E F

D T

E A

C' n%M S

UD A

LA E

PE L

H

)

E C

E DR E

RT UE S

f

(

P F

T P

YE T

i U

N GH C

RN P

H T

5l E1 T

~

Ne v

ET 2S N

I KO I

S T

AC O.

tiA

= 8 E

Y L

l tO U N BT 8

A YE ER L

RA C 6

L F

1 CY

_N F l Pi EC l

E DH E

E R

4. [N S

U OJ0 G

K I

A AE LR AB g.S m

O f

ad l'

i 7

F F

F 0

0 0

0 O

0

+

t E

2 E

D

+

f 8

8 8

8 0

C 0

0 0

8 0,

0 5

7 1

1 3

S

> ?~ s*l u4 l*u G f,u 1

t.

e '

,i l

4 i

ji 4

ji!4 i1 l;

i!'

j i ;

a'

!j

l 1

1

_-a w- -

U4 Msts

-_4 wwu.-

-m, W S

--mm--.~

a ^

=-

M

~

w-

,.sm

--g.p= -JB00- _ _ _

_ u_ -

33 I

.m z.

q

- + - -

1r S.

, 00- =;.- 3T h

1

- :- a*

_r,-

2-___

M L.

7m

. ='.. _-

_.._i__. - _ -

3 Q

k x

+-

~

3kEMM.

.5t==~=

=p_

e

]

_i_____._.

= C~

Q

.= =

=- 3 :

==

=th

C

_T

= =c_._

l~

h =-'- ' _ = " '

=

E

_L w'" -'

_ i._.

1 3.

-_-v_-_

- i

[

-- A

v=

=b a

] -----

ri t-~

~

-+-

o

=e

=

-- 2:=_E400.. =_ -

=w

.h Q

M

~

~~

U

- = =

N*r MN

~

- - - --~

p

=, -

==

=

4,

=x =_=e=.

ht u

7 __

~~

'~"_ = _ = __ ;

. ---h.

E3h

__3._

de

=; a _.

_y_.____-_

g__ g._

g

=<

N.

I

___.__ =t -.

=-p

=2____

=>_

st$

= El e-

- ~== =. -

---= =M=

=

_ _ _ ;=_=_ =

~

--= =':

LAJ _._

.. ~=

-._=_

=

g m::

L*-

~----

rE.__=

~ ~. = _ _. _

.=_=_

==

__ g_.==__=_

= _.w --_

= &" =-=200-

=:....

=___-__ _ __ _

M.._.C.=-

= _ _.. -._

-_=_= _..__

m_

=. _. = _.=_n

=.

= s =_..=_.

= _=..

=

_..;=

m

= _. < _ _

_.1 -==

=

~ _.

_~

~

1]._

.. _ d_ 5~~_. _ _.::

2

_.. =_.. - - _. _ ~ ~== ::

~

4,._..

i

---C-Z

==

.:-i---==

._t----=nr

. :: = :.-- :-- =

=

.C-' =

' - - - -M**=

C-2--

O* ;I

-i----

ZL

. L.__

=O-.

-*=

.= _

u-W

...g.'

.._=_

- 2:__n = _. _.=. -.

. :.:=.. = _M_ r-

=

.= :::.= _, _-.f _ r..:.... _.

..=.

[..

.=.

_ -. =. - -.-.f..__.~_.. _ = = =__=32.._,2:.:n t..

y

=

.:._=_..=.=_

._=

x.,==

s =

==

=

u 3.

=

.--- n*1 =- =in:

2 *. :_

-. - : =

=:*_: =.:

d=

2-2

=

un a==. =..

._--= _..

-- 100.' ;-+t=_==

==

_..r=

=

==

= =

(

.i.._.___.=. _ _= _..

.....=. =. =...=.. = _

.h t in:

=

.--- =

=

==

=

un ze:

=

:.=*

p*::==.

.:. :::$=-

.==

= :_.

=

2:- ~: ': 2

'.. = _

M2:

T:

=

=

=.

22

.**: = -21 =

~~

C:

=

J.,

.._=__

=...=.. =...

=.=..: ;..

Z..

..=_

^*~

~7

_.= _

3.

..v

=.=,.._.-._.: =.. _

=

nr

= =--== un u -}= =:

un

=

N=_.

l=_--===

=._

==

=== =-

=

m: un un =

=

_--5 E==:=.x..=

==

=.=:==

I =.=

n u.

n. n. :.n. f n...=.

=.~1._.

...?..

. :? =_ _=._

=

=_

=

= =

-,-n

- - =

=

=

==

=.. n. r. -

_.. _ =

..=._=. _ _.

=

= :

g

--):.:= =

=

==

s=

.3:=.J--()f)g :: :-m)=:

= :: =: =::::=.==.=2002s =E a -g= = a a 2300==-

=

==

-r.= -:=

=

nt:

- - =- :.

x-

=

- -- - E

(

==

2:

=

--= ===

~

--- =

=====q.

=

g. = _=a=_

= _..

. _. =

_=_.. =..

.._..=.=.

.=.:.~

==

=--

==

_:

n= =

_=. = _ _ =_==

=._=.._=.. _=.

. = _.,.

_ =..

.:_=.. =.. =.

_ :.:,.__sSIE{SEf.=)-... =.

= _...=.. _

. =. _=.=

__m.

= _.. =. =. =.=... =...

.J__-

_--. =. _ _.= _ _

=_.=..==....m--

. =..=..:.:._u :.

._. - -. =. =...=.=_

=:_ =. _

L

--=1EE =

=

==

=

a..

t=

=

=:- - -

EE.= :.: =

22: =

===4

=ap ~_l = :+:1

?'j:l:M::

E

=

=nr

=

=j. =

=.!=

=

ca =

=

u:

2:: ru =

ur n FIGURE 17 CONTAINf1ENT WALL

" ~-

HEAT TRANSFER C0 EFFICIENT

I t

ATTACHMENT II i

l 4

JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 i

t 1

i I

FUEL ROD BURST / BLOCKAGE EVALUATION CALCULATION SHEET 2

I I

l i

1 l

l

%e

-,-nw_,

-,a.,

,..,,-an.,

,.. ~,..

n,

FUEL ROD BURST / BLOCKAGE EVALUATION CALCULATION SHEET A.

Evaluation of the potential impact of using fuel rod models preshnted in draft NUREG-0630 on the Loss-of-Coolant Accident (LOCA) analysis for Joseph M. Farley Units 1 and 2.

This evaluation is based on the limiting break LOCA analysis identified as follows:

BREAK TYPE - DOUBLE ENDED COLD LEG GUILLOTINE BREAK DISCHARGE COEFFICIENT 0.4 WESTINGHOUSE ECCS EVALUATION MODEL VERSION February, 1978 CORE PEAKING FACTOR 2.32 HOT ROD MAXIMUM TEt1PERATURE CALCULATED FOR THE BURST REGION Of THE CLAD -

1998.70F = PCTB-ELEVATION -

6.0 Feet.

HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR A NON-RUPTURED REGION OF THE CLAD -

2181.90F = PCTg ELEVATION -

7.5 Feet.

CLAD STRAIN DURING BLOWDOWN AT THIS ELEVATION 0.865 MAXIMUM CLAD STRAIN AT THIS ELEVATION 6.416 --Percent Percent Maximum temperature for this node occurs when the core reflood rate is less than 1.0 inch par second and reflood heat transfer is based on the Steam Cooling calculation.

AVERAGE HOT ASSEPELY ROD BURST ELEVATION -

6.0 Feet HOT ASSEMBLY BLOCKAGE CALCULATED -

43.0 Percent 1.0 BURST N0DE The maximum potential impact on the ruptured clad node is expressed in letter NS-Tf'A-2174 in terms of the change in the peaking factor limit (F ) required to maintain a peak clad temperature (PCT) of 22000F Q

and in terms cf a change in PCT at a constant FQ. Since the clad water reaction rate increases significantly at temperatures above 22000F, individual effects (such as APCT due to changes in several fuel rod models) indicated here may not accurately apply over large ranges, but a simultaneous 0

change in FQ which causes the PCT to remain in the neighborhood of 2200 F justifies use of this evaluation procedure.

s s.

. l 4

From NS-TMA-2174:

For the Burst Node of the clad:

- 0.01 AFQ a., 1500F BURST NODE APCT Use of the NRC burst model could require an FQ reduction of 0,015

~

The maximum estimated impact of. using the NRC strain model is a required FQ reduction of 0.03.

Therefore, the maximum penalty for the Hot Rod burst node is:

(.015 +.03) (1500 /.01) = 6750F F

APCT

=

1 Margin to the 22000F limit is:

[

aPCT 22000F - PCTB = 01.3

=

2 The FQ reduction required to maintain the 22000F clad temperature limit is :

.01 AFQ APCT )

( 1500F )

(APCT AFQ

=

B 1

2 (675 - 201.3)

(

)

=

0.03 (but not less than zero)

=

2.0 NON-BURST NODE The maximum temperature calculated for a non-burst section of clad typically occurs at an elevation above the core mid-plane during the core reflood phase of the LOCA transient. The potential impact on that maximum clad temperature of using the NRC fuel rod models can be estimated by examining two aspects of the analyses. The first aspect is the change in pellet-clad gap conductance resulting from a difference in clad strain at the non-burst maximum clad temperature node elevation. Note that clad strain all along the fuel rod stops after clad burst occurs and use of a different clad burst model can change the time at which burst is calculated. Three sets of LOCA analysis results were studied to establish an acceptable sensitivity to apply generically in this evaluation. The possible PCT increase resulting from a change in strain (in the Hot Rod) is +20 F per percent decrease in strain at the maximum clad temperature 0

t locations. Since the clad strain calculated during the reactor coolant system blowdown phase of the accident is not changed by the use of NRC fuel rod models, the maximum decrease in clad strain that must be considered here is the difference between the " maximum clad strain" and the " clad strain during blowdown" indicated above.

~..

..,_-r_m,

.r.-_,

m.,,

y

.m..-..,._,_,,.m__,.

s-

~

Therefore:

3 I.01 strain) (MAX STRAIN - BLOWOOWN STRAIN)

APCT

( 01) (0.06416 0.00865) s

=

111.00F

=

The second aspect of the analysis that can increase PCT is the flow blockage calculated. Since the greatest value of blockage indicated by the NRC blockage model is 75 percent, the maximum PCT increase can be estimated by assuming that the current level of blockage in the analysis (indicated above) is raised to 75 percent and then applying an appropriate sensitivity formula shown in NS-TMA-2174.

Therefore:

0 1.25 F (50 - PERCENT CURRENT BLOCKAGE)

APCT

=

4 0

+ 2.36 F (75-50) 1.25 (50 - 43) + 2.36 (75-50)

=

67.80F

=

If PCT occurs when the core reflood node is greater than 1.0 inch per g

second APCT4 = 0.

The total potential PCT increase for the non-burst node is then APCT3 + APCT4 APCT 111.0 + 67.8 = 178.80F

=

S Margin to the 22000F limit is 22000F - APCT 18.10F 2200 - 2181.9 APCT

=

6 N

The FQ reduction required to maintain this 22000F clad temperature limit is (from NS-TMA-2174).

(APCT5 - APCT ) Ib I

aFQ

=

I 6

N 0.16 (but not less than zero)

AFQ

=

N 0

The peaking factor reduction required to maintain the 2200 F clad temperature limit is therefore the greater of AFQ and AFQ '

g N

0.16 or: A FQ

=

PENALTY s

s.

?

f. ~ [ B.

The effect on LOCA analysis results of using improved analytical and modeling techniques (which are currently approved for use in the Upper Head Injection plant LOCA analyses) in the~ reactor coolant system blowdown calculation (SATAN computer code) has been quantified via an analysis which has recently been submitted to the NRC for review. Recognizing that review of this analysis is not yet complete and that the benefits associated with these model improve-ments can change for other plant designs, the NRC has established a credit that is acceptable for this interim period to help offset penalties resulting from application of the NRC fuel rod models. That credit for two, three, and-fdur loop plants is an increase in the LOCA peaking factor limit of 0.12, 0.15, and 0.20, respectively.

C.

The peaking factor limit adjustment required to justify plant operation for this interim period is determined as the appropriate AFQ credit identified in Sectinn (B) above, minus the a FQ calculated in Section (A) above PENALTY (but not greater than zero).

0.16 0.15 FQ ADJUSTMENT

=

-0.01 1

i

-._,_m_.

__..__.,__m___.

-

ML20032E035

Text

Background:

References:

26]CI== -~=4

==

= =*

===:=-.J

~I'E: i~.

=J

=

==

====-.;

3,._

us --* m

==

1 s E._s=

== a

;

_1_ =4 :

1 --=

=

.

2=_

,' :==

=

= =

-1

=

=

._=

d=

=

= :_.

..=_

_.= _

l=_--===

I =.=

=

. _. =

==

_:

===4

=ap ~_l = :+:1

=.!=

=

Navigation menu

Search