ML18227B281

Submit Appendix a, Revised ECCS Analysis

ML18227B281
Person / Time
Site:	Turkey Point
Issue date:	09/27/1978
From:	Florida Power & Light Co
To:	Office of Nuclear Reactor Regulation
References
Download: ML18227B281 (74)

Text

2'%00 PSIA 630 2250 PSIA 620 CP I

+elo CV eoo I

590 5

580 2IOO PSIA 1900 PSIA I700 PSIA 570 560 Note:

These curves are applicable with steam generator tube plugging

< 15 percent.

550 0

60 80 loo RATED POWER

{PERCEHT) l20 Figure 2.1 1.

Reactor Con Thermal and Hydraulic Safety Llml<

Three Loop Operat~

9/27/78

i5 II lg

"0 650 640 2400 PSIA 630 2250 PSIA 620 OO 610 2100 PSIA, 600 i

590 580 1900 PS1A 1700 PSIA.

570 Note:

These curves are applicable with steam generator tube plugging 15 percent and ( 19 percent.

560 0

20 40 60 80 100 120 140 Rated Power (Percent)

Figure 2.1-la Reactor Core 'Thermal and Hydraulic Safety Limits, Three Loop Operation 9/'27/78

II

/

Reactor Coolant Tem erature Overtempera-ture AT ATo Kl 0.0107 (T 574) + 0.000453 (P-2235) f(Aq) hT Indicated BT at rated

power, F

0 T

= Average temperature, F

P Pressurize".

pressure, psig f(Aq) ~ a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during startup tests such that:

For (q

q ) wi'thin +10 percent and -14 percent' b

where q

and qb are the percent power in the'op t

and bottom halves of the core respectively, and q

+ q is total core

.power in percent of rated.

power, f(Aq) = 0-

/

For each percent that the magnitude of (q q )

exceeds

+10 percent, the Delta-T trip set point shall be automatically reduced by 3.5 percent of its value at interim power.

For each percent that the magnitude of (q q )

exceeds

-14 percent, the Delta-T trip set point

'i shall be automatically reduced by 2 percent of.

its valu at interim power.

K (Three Loop Operation)

~ 1.095*

1 (Two Loop Operation)

~ 0.88

• Kl = 1.095 for steam generator tube plugging

< 15 percent Kl = 1.08 for ste'am generator tube plugging ) 15 percent and

< 19 percent 2.3-2 9/27/78

0

Over-power AT' AT dT 1 ~ 11 K K2 (T T

f'Aq) 1 dt 5T 'ndicated.'hT at rated power, F

0 T

~

Average temperature, F

T' Indicated. average temperature at nominal conditions and rated power, F

K1 K2 0 for decreasing average temperature, 0.2 sec./F for increasing average temperature 0.00068 for T equal to or more than T';

0 for T less than T'ate of change of temperature,,

F/sec dT dt f(hq)

As defined above Pressurizer Low Pressurizer. pressure equal to or greater than 1835 psig.

High Pressurizer pressure equal to or less than 2385 psig.

High Pressurizer water level equal to or less than 92% of full scale.

Reactor Coolant Flow Low reactor coolant flow equal to or greater than 90% of normal indicated flow Low reactor coolant pump motor frequency equal to or greater than 56.1 Hz Under voltage on.reactor. coolant pump motor bus - equal to or greater than 60% of normal voltage Steam Generators, Low-low steam generator water level equal to or greater than 5% of narrow range instrument scale

• This factor is 1.11 for steam generator tube plugging

< 15 percent.

This factor is 1.10 for steam generator, tube plugging > 15 percent and

< 19 percents

203-3 9/27/78

t'

6.

DNB PARAMETERS The following DNB related parameters limits shall be maintained during power.operation:

a.

Reactor Coolant System Tavg

< 578.2'F b.

Pressurizer Pressure

> 2220 psia*

c.

Reactor Coolant Plow

> 268,500 gpmt With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal power to less than 5% of rated thermal power using normal shutdown procedures.

Compliance with a.

and b. is demonstrated by verifying that each of the parameters is within its limits at least once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Compliance with c. is demonstrated by verifying that the parameter is within its limits after each refueling cycle.,

/

• Limit not applicable during either a THERMAL POWER ramp increase in excess of (5%)

RATED THERMAL POWER per minute or a THER?QJ.

POWER step increase in excess of (10%)

RATED THERMAL POWER.

t Reactor Coolant Flow > 268,500 gpm for steam generator tube plugging

< 15 Reactor Coolant Flow > 263,130 gpm for steam generator tube plugging

> 15% and

< 19%.

3.1-7 9/27/78

i5 k~

reactivity insertion upon ejection greater than 0.3%

b, k/k at rated power.

Inoperable rod worth shall be determined within 4 weeks.

b.

A control rod shall be considered inoperable if (a) the rod cannot be moved by the CRDM, or (b) the rod is misali'gned from its bank by more than 15 inches, or (c) the rod drop time is not met.

c. If a control rod cannot be moved by the drive mechanism, shutdown margin shall be increased by boron addition to compensate for the with-drawn worth of the inoperable rod.

5.

CONTROL ROD POSITION INDICATION If either the power range channel deviation alarm or,the,rod deviation mon-itor alarm are not operable rod positions shall be logged once per shift and after a load change greater than 10% of rated power.

If both alarms are inoperable for two hours or more, the nuclear overpower trip shall be reset to 93% of rated power.

6.

POWER DISTRIBUTION LIHITS a.

Hot channel factors:

1.

With steam generator tube plugging

< 15%, the hot channel factors (defined in the basis) must meet the following limits at all times except during low power physics tests:

Fq (Z)

< (2.22/P) x K(Z), for P

>.5 Fq (Z)

< (4.44) x K(Z), for P

<.5 FgH

< 1.55

[1 + 0.2 (1-P)]

Where P is the fraction of rated power at.which the core is operating; K(Z) is the function given in Figure 3.2-3; Z is the core height location of Fq.

2.

With steam generator tube plugging

> 15% and

< 19%, the hot channel factors must meet the following limits at all times except during low power physics tests:

Fq (Z)

< (2.05/P) x K(Z) I for P

>.5 Fq (Z)

< (4.10) x K(Z) I for P

<.5 FNH 1 5 ~1+0 2 (1 P Where P, K(Z), and Z are defined in l. above.

If predicted F

exceeds 2.05 with tube plugging 15% and

< 19%, then power will be limited to the rated power multiplied by the ratio of 2.05 divided by the predicted Fq, or augmented surveillance of hot channel factors shall be implemented.

b.

Following initial loading before the reactor is operated above 75% of rated power and at regular effective full rated power monthly intervals thereafter, power distribution maps, using the movable detector system shall be made, to conform that the hot channel factor limits of the speci-fication are satisfied.

For the purpose of this comparison, 3'-3 9/27/78

Ih I

SAFETY EVAL'UATION Re:

Turkey Point Units 3 and 4

Docket Nos.

50-250 and 50-251 Revised ECCS Analysis I.

Introduction This evaluation supports several changes to the Technical Specifications which have been brought about by the potential plugging of additional steam generator tubes at Turkey Point Units 3 and 4.

Technical Specification 3.1.6.c (Reactor Coolant Plow), and 3.2.6.,a (Hot Channel Factors),

the 'Over-temperature hT and Overpower hT equations, and Figure 2.1-1 (Reactor Core Thermal and'ydraulic Safety Limits, Three Loop Operate;on) will be affected.

II..

Revised ECCS Anal sis The attached ECCS Analysis (see Appendix A) constitutes a

reanalysis of a hypothetical loss of coolant accident (LOCA) for Turkey Point Units 3

& 4.

The previously demonstrated limiting break (DECLG,. CD = 0.4) was reanalyzed changing only the following parameters:

1)

RCS flow =

98% of the thermal design value 263,130 GPM')

Steam generator tube plugging

=

19%

(uniform)

3) Total peaking factor (Fq~')

= 2. 05 4)

No change in nominal Tavg,.

however, instrument uncer-tainty was subtracted from TIN rather than added.

III.

RCS Flow An evaluation has been performed to address the operation of Turkey Point Units 3 and 4 at 98$ rated Thermal Design Flow.

The evaluation was performed consistent with the following assumptions:

98% Thermal Design Flow, gpm S.G.

Tube Plugging, Maximum Power,.

Mwt Tavg at 100% Power,

'F 87,710 19 2, 200 574. 8

ik gl

Safety Evaluation Pa e Two 0

hT at 100% Power,

'F 1 t at 100% Power,

'F inlet Fg maximum

57. 1 546. 2 2.05 A reduction in the steady state primary 'flow will affect all of the FSAR Chapter 14 transients.

However, by using excess margin available and technical specification reductions in allowed core peaking factors, a

2% change in flow will not change the.safety conclusions in the FSAR.

The FSAR transients can be divided into two categories:

DNB'imited, and Fuel or Reactor Vessel Integrity Limited.

These are discussed below with the method used to offset penalties associated with flow reductions.

A.

DNB Limited Transients The primary means of DNB protection for these transients is the Over-Temperature Delta-T Protection System.

Although credit might not have been taken in the FSAR, this system assures DNB protection limits are not exceeded for the following transients:

Rod Withdrawal at Power, Boron Dilution at Power, Excessive Heat Removal due to Feedwater Malfunction, Startup of an Inactive Loop, Excessive Load Increase and Loss of External Electrical Load..

Revised Technical Specification core limits have been developed which incorporate a

2%, reduction in thermal design flow.

A red'uction in the Kl term of the Over-.Temperature Delta-T setpoint equation from 1.09 to 1.08 will assure adequate protection.

In addition, to the above there is considerable margin to DNB limits (DNBR = 1.24) in nearly all of the above transients.

Since a

2$ reduction in flow results in approximately a

2% reduction in DNBR, there is still adequate margin available.

The DNB transients not protected by the Over-Temperature Delta-T setpoints are:

Rod Misalignment, Loss of Flow and Steamline Break.

For all of these cases the flow reduction corresponds to less than a

2% reduction in minimum DNBRs, which can be accommodated with margin in the current design.

b I'

Safety Evaluation Pa e Three 0

B.

Fuel or Vessel Integrit Limited Transients Rod Withdrawal from Subcritical The current safety analysis shows large margins to safety limits with the peak heat flux being considerably less than 100% of rated power.

Thus a

2% reduction in flow would have a negligible effect on peak fuel or clad temperatures.

Boron Dilution The relatively long duration of the transient means that flow does not affect the operator action times during refueling or startup operation.

In

addition, the effect of 25% steam generator tube plugging on boron dilution has been analyzed.

This analysis conserva-tively bounds the 19% plugging analysis.

(Appendix B)

Locked Rotor A reduction in flow will slightly increase peak system pressure

(

< 50 psia) from the value shown in the Cycle 3 RSE.

However, the results are still consider-ably below the vessel faulted stress limits.

The peak fuel and clad.temperatures would also be affected.

How-

ever, the hot spot peaking factor has been reduced, due to LOCA considerations, from 2.32 to 2.05.

This 11%

reduction in hot spot energy would more than compensate for the 2% reduction in flow.

Thus vessel and fuel limits would not be exceeded due to a flow reduction.

Loss of Normal Feedwater/Station Blackout The results of this accident are highly sensitive to the residual (decay) heat generation due to the long duration of the transient after trip.

Residual heat generation is directly proportional to the initial power level preceding the trip.

The analysis in the FSAR assumed the power to be 102% of the maximum turbine rating (2300 Mwt).

Thus the total energy input to the system would be +

5% less than originally assumed.

Therefore this affect alone would more than compensate for a 2% flow reduction.

lh

~,

4

0 Safety Evaluation Pa e Pour Rupture of a Control Rod Drive i4Iechanism Sensitivity studies have shown that a

2-o reduction in flow will result in less than a 40'F increase in fuel and clad peak tempera-tures.

The current analysis shows that for a 40 F

increase, all fuel and clad integrity limits can be met with margin.

Loss of Coolant Accident An Appendix K LOCA analysis is attached (Appendix A) for l9% tube plugging and 98%

thermal design flow.

Thus it has been shown, that a 2% reduction in thermal design flow will not result in any safety limit violation.

i5 0

r

APPENDIX A REVISED ECCS ANALYSIS TURKEY POINT UNITS 3

6 4

Ih

~ '

TABLE 1 TIME SEQUENCE OP EVENTS CD = 0,4 DECLG (Sec)

SThRT Rx Trip Signal S.I'ignal Acc. Infection End of. Blowdown Bottom of Core Recovery Acc. Empty Pump Injection End of Bypass 0.0

.556

.7000 28.061 46.789 60.979 25.7000 27.815

TABLE 2 LARGE BRI?AK CD =0.4 DECLG Results Peak Clad Temp. 'F Peak Clad Location Ft.

Lo 1 Z /H20 Rx (

x)%

Local Zr/H20 Location Ft.

Total Zr/H 0 Rxn Hot Rod Burst Time sec Hot'Rod Burst Location Ft, 2195. 37 6.0 12.3951 6.0

<0.3 22.80 6.0 Calculation Core Power Mwt 102% of Peak Linear Power kw/ft 102% of Pt.aking Factor (At License Rating)

Accumulator Water Volume (per tank) 2200 11.650 2.05 875 ft3 Fuel region + cycle analyzed UNITS 3 and 4

Cycle Region 3

. TABLE 3 LARGE BREAK CONTAINMENT DATA (DRY CONTAINMENT)

NET FREE VOLUME INITIALCONDITIONS Pressure Temperature RWST Temperature Service Water Temperature Outside Temperature

1. 55x10 Ft 14.7 psia e%9O oF 39 'F 63 'F 39 4F SPRAY SYSTEM Number of Pumps Operating Runout Flov Rate Actuation Time 1'450 gpm 26 secs SAFEGUARDS. FAN COOLERS Number of Fan Coolers Operating Fastests Post Accident Initiation of Fan Coolers 26 secs

0 il I

LARGE BREAK TABLE 3 (Continued)

CONTAINMENT.DATA (DRY CONTAINMENT)

STRUCTURAL HEAT SINKS Thickness (In)

Area (Ft )

2 Steel 0;03 Steel 0.063 Steel O.ll Steel 012 Steel

0. 24 Steel: 0.2898 Concrete 24.0 Steel 0.4896 Steel 0.6396 Steel 0.8904 Steel 1.256 Steel 1.56..

Steel 2.0 Steel 2.75 Steel 5'

Steel 9'

Stainless 0.14 Concrete 24.0 Stainless 0'4 Stainless 2.126 Stainless 0.007 Concrete 24.0

.31,400 107$ 158 56,371 57$ 185 9$ 931 136,000 23,677 6,537 4,915 27$ 802 5,307 668 1,268.7 1,277 '

260,4 14,392 768 3704 102,400 59,132

il JI

0.95 0.85 0.80 IO 4

'6 8 IO 2

4 6 8IO 2

4 6

8 IO 2

4 '

8 IO T IHE (SECOHDS)

Figure 1

Fluid Quality-DECLG (CD = 04)

il il

30 20 CD LJJ cv' 0

I 0

I CD

-IO F.

-20

-30

.IO 2

0 6

8 IO 2

0 6

8 IO 2

tI 6

8 IO 2

0 6

8 IO T IME (SECONDS)

Figure 2

Mass Velocity - DECLG (CD = 0.4)

ih P

I

2 l0 I

c4 I

4.

2 I 02 8

6 O

2 Io' 6

I00 0.

25 50 75 I00 TIME (SECONDS)

I 25 l50 l75 200 Figure 3 Heat Transfer Coefficient - DECLG (C> = 0.4)

0 ih

2500 2000 l500 LJJ I 000 500 0

0 IO l5 TIME (SECOIIDS) 20 30 Figure 4

Core Pressure

- DECLG (C> = 0.4)

0

llx exl04 7xl0" 5xl 0 hC 3xlo" I xi 04

-lxl0" 0

IO l5 TIME (SECONDS) 20 25 30 Figure 5

i Break Ftow Rate-OECLG (Cp =0.4)

15

75 50 a-25 Ch Ul 0

-25 5

CD

-50 0

5 l0 l5 T iVE (SEC0H~S) 20 25 30 Figuie 6

Core Pressure Drop DECLG (CD = OA)

0 II

'4j

2500 2000 I 500 IOOO LLI W

O 500 0

0 25 50 75 IOO 125 TIME (SECONDS) l50 l75 200 Figure 7

Peak Clad Temperature

- DECLG (Cp = 0.4)

0 4$

2000 l 750 l500 I250 I 000

'50 500 250 0

0 25 50 75 IOO TIIIE (SECOHDS) l25 150 I 75 200 Figure 8

Fluid Temperature-DECLG (Cp =OA)

0 lp

l0000 7500 5000 2500 0

lt g

-2500 IV BOTTOM

-5000

-7500

-l0000 0

l0 T ICE (SECONDS) 20 25 30

-Figure 9

Core Flow - Top and Bottom - DECLG (CD = OA)

20.0 17.5 ZO 15.0 u

12.5 10.0 7.5 LC 5.0 2.5 0 ~ 0 0

25 50 75 100 T IWE (SECONDS) 125 150 175 200 Figure 10a Reflood Transien'.

- DECI.G (Cp = 0.4)

ih Ig C

I ~

2.00

.75 L

I.50 0.50

'0. 25

0. 00 0

25 50 I 00 I25 TIME (SECONDS)

I50 I 75 200 225 Figure 10b Reflood Transient - DECLG (CD = 0.4)

0 4$

I ll

6000 5000 LJJ tI000 CD 3000 2000 I 000 0

0 Io I5 TIHE (SECONDS) 20 25

-30 F igure ll Accumulator F low (Blowdown) - OECLG (CO = 0.4) 4 4

UJ 3

CO 2

0 0

00 l

I I

t l

t 60 80 I 00 I 20 I I40 I 60 I 80 200 220 TIME (SECONDS)

Figure 12 Pumped ECCS Flow (Reflood). DECLG (Cp = 0.4)

4>

( ~

'lO

30 CO 20 15

!0 0

0 40 80 I20 I60 200 240 280 320 360 400 TIME (SECONDS)

Figure l3 Containment Pressure

- DECLG (Cp = 0.4)

Cl II

I.O 0.8 2

0.6 0.2 0.0 0

IO l5 TIME (SECONDS) 20 25 30 Figure 14 Core Power Transient

- DECI G (CD ='.4)

0 Ck

5.5xl06

%.5xl06 w

3.5xI06 UJ LLI 2.5xl 0 l.5xl 06 5.0xl06

-5.0xl06 0

IO l5 TIME (SECONDS) 20 25 30 Figure 15 Break Energy Released to Containment

- DECLG (Cp = 0.4)

II C

\\'

iy

l 200 I 000 800 600 400 200 0

0 loo 200 300 400 500 600 700 800 900 looo TlhfE (SECONDS)

Figure g6 Containment Wall Hea. Transfer Coefficient - DECLG (Cp = 0.4)

0 r

r ~

t

APPENDIX B REACTIVITY INSERTION RATE vs.

BORON CONCENTRATION TURKEY POINT UNITS 3 4

I+I 1

t

Boron Dilution Anal sis Section 14.1.5 of the Turkey Point Units 3

6 4 FSAR shows that for a boron dilution event the operator has sufficient time to identify the problem and terminate the dilution before the reactor returns critical or loses shutdown capabili.'ty.

The standard acceptance criteria and FSAR calculated values for operator action are summarized below:

NODE FSAR (minutes)

ACCEPTANCE CRETE RXA (minutes)

Refueling Startup Power a.

Manual Control b.

Automatic Control

> 120 240 15 30 15 15 15 Steam generator tube plugging has no affect on the analysis at refueling conditions since only the reactor vessel volume is assumed active.

The coolant loop volume is conservatively assumed stagnant.

For dilution during startup and at. power, there is an effect due to the reduction in primary coolant volume.

The effective volume of primary coolant in the steam generator tubes is conservatively assumed to be reduced by 25%

(> 510 cubic feet).

Thus the total volume assumed in the analysis has been reduced from 7800 cubic feet to 7290 cubic feet.

This translates into approximately a

7% reduction in the originally calculated dilution time from startup conditions (240 minutes)

This is still significantly greater than the required operator action time, therefore no safety concern exists.

l 4~

~ g A

1 h

J

For dilutions during power operation a highly conservative reactivity insertion rate of 1. 1 x 10 5

8 k/sec was assumed in the FSAR consistent with an initial boron concentration of 1200 ppm.

FSAR figure 14.1.5-1 (Reactivity Insertion Rate vs Boron Concentration) has been reca'lculated consistent with the lower primary value and is attached.

The results show that the reactivity insertion rate assumed in the FSAR is still conservative.

Therefore no additional analysis is required.

Et should be noted, however, that the FSAR analysis is still highly conservative with respect to the current cycles since the analysis assumed that only 1$ shutdown margin is available.

The Turkey Point Units have been designed such that

>.2.5% shutdown margin is always available for BOL conditions.

The result is that operator action times would be

> 70 minutes with the more realistic value.

Oi t

II

2.0 Variation in Reactivity Insertion Rate vith Initial Boron Concentration for a Boron Dilution Hate of 230 GPh!

1.0 0

0 1000 2000 2500

'nitial Boron Concentration, PPM

4i i ~F

• ~

E

+K FLORIDA POWER 5 LIGHT COMPANY INTER-OFFICE CORRESPONDENCE TO FROM SUBJECTS R. E. Uhrig A. D. Schmidt TURKEY POINT UNITS 3

6 4

PROPOSED TECH SPEC CHANGE ECCS REANALYSIS (19% S/G PLUGGING)

LOCATION DATE COPIES TO Miami,, Florida January 26, 1978 J.

R. Sensen/C.

O. Woody G. E. Liebler/932.1 TP H

E., Yaeger/J.

K. Hays R. J. Acosta N. F. Ajluni G. D. Whittier PRN-LI-78-20 The subject proposal is attached for your review and forwarding to the NRC. It has been reviewed and approved by the PNSC and CNRB.

A. D. Schmidt MAS/lah Attachment COURTESY WINS FRIENDS...FOR FLORIDA...FOR YOUR COMPANY...FOR YOU!

FORM ZOOS REV. Zna

4l if, C~