F ,if 5h b- .a. m

F ,if 5h" b-. .

m .s.

, .a".20 (. 'I007 r'W10j *

.

* j# g :,e6&G'(

FloridaP.o. e . . e. ._rw .

g a,m

N $# OCS d. O

May 21, 1979 3 ,;ec E

;..'cs ,-

to ,-

Mr. James P. O'Reilly, DirectorU.S. Nucicar Regulatory Commission -

'

Office of Inspection and Enforcement101 Marietta Street, Suite 3100Atlanta, Ca 30303

Subject: Docket No. 50-302Operating License No. DPR-72I.E. Bulletin 79-05B

Dear Mr. O'Reilly:

Enclosed is our response to Items 5 and 7 of 1.E. Bulletin 79-05B datedApril 21, 1979.

Our response describes a generic design for implementation of safety gradereactor trips into the Reactor Protection System at Crystal River Unit 3for loss of main feedwater and turbine trip. The detailed design and

procurement of this modification will require 12 months following NRCapproval of the generic design. This modification would be installed atthe first ref ueling outage or outage of suf ficient duration following this12 month period.

Item 7 of I.E. Bulletin 79-05B required the submittal of those technicalspecifications which must be modified as a result of our response to thisbulletin. However, as per our discussions with Mr. Hugh Dance of yourstaf f, Florida Power Corporation will be submitting changes to the CR #3technical specifications f ollowing issuance of the SER f or CR #3.

2170 1687906280093 ,

G n r i Office 320i inirt tounn street sovin. P O Bon 14042 St Petersbu g Fiorida 33733 , 813 - 666 5 t 51r

a

. .

.

$

Mr. James P. O'Reilly May 21, 1979

Should you require further discussion concerning this submittal, pleasecontact this office.

Very truly yours,

FLORIDAPOWERCORPOP.AI40N1

W, f. BrauTat_

W. P. StewartManager, Nuclear Operations

WPSemhM08D6

cc: U.S. Nuclear Regulatory CommissionOffice of Inspection and EnforcementDivision of Reactor Operations InspectionsWashington, D.C. 20555

File: 3-0-3-a-3

0 169

-

. .

. ..

.

,

.

NEW SAFETY-GRADE REACTOR

TRIPS FOR RPS-1

2!70 170

,** , , .

,* ...

.

This report describes the implementation of safety-grade reactor trips into*

the RPS-I for loss of main feedwater and turbine trip.

Loss of Main Feedwater Trip - Control oil pressure switches on both mainfeedwater pumps will input an open indication to the RPS on feedwater

pump trip. Contact buffers in the RPS will sense the contact inputs andinitiate an RPS trip when both pumps have tripped. This trip will be bypassed

below a predetermined flux level, typically 20% FP. Reference Figure 1.

Turbine Trip - Contact outputs from the main turbine electro-hydraulic controlunit will input an open indication to the RPS on turbine trip. Contact

buffers in the RPS will sense the contact inputs and initiate an RPS tripwhen a turbine trip is indica,ed. This trip will be bypassed below at

predetermined flux level, typically 20% FP. Reference Figure 2.

B&WPressure switches for both trips will be supplied by the customer.will supply all RPS cabinet mounted equipment. Attachment 1 lists the cabinet

mounted equipment and gives the trip response time. Attachment 1 also gives

the contact buffer isolation voltage and the customer requirements for the

contact inputs.

Figure 1 is a simplifiad drawing of the main feedwater pump trip.

Figure 2 is a simplified drawing of the turbine trip.

Drawing 51079DGB-1 shows the generic logic for the new trips.

Drawing 51079MLG-1 is a legend for the generic logic drawing.

2170 1/1

.- Attachment 1

, ,-,

. . .

CABINET MOUNTED EQUIPMENT FOR ADDITION OF RPSTRIPS ON LOSS OF MAIN FEEDb'ATER AND TURBINE TRIP

,

3 Contact Buffers .

2 Bistables Per Channel

2 Auxiliary Relays j

Modules will be installed in a pre-wired mounting case and tested asa unit prior to shipment. The mounting case is to be installed in an

"

empty row of each RPS channel and connections made to the RPS wiring.

Trip response time of the RPS cabinet mounted equipment will be 1 150 ms.

Isolation of the contact buffer module is 600 volts with the contactinput lines not grounded.

Customer contact input requirements:

Continuous 90 ma, P-P

Surge 250 ma, P-P

Voltage 118 VAC

Closed contact indicates pump running

Open contact indicates pump tripped

2170 172

.

TYPICAL RPS CHANNEL - >*

EXISTING

TRIP STRING. c-

__ q_ _ _ _ _ _ . _ _ _ _ _ _

ADDED EQUIPMENT g' .

*FOR LOSS OF MFW TRIP | N

MFW PUMP A |-

1 | oTRIPPED :

|~T .

. * ~

CONTACTS CONTACT l m_-

BUFFER g,

I !|

'

MFW PUMP B _L |'

0'. TRIPPE0 I''

CONTACTCONTACTS :

'

.

BUFFER

l'<.

_q- _ _ _ .

-|- 1 .

'

I - 1 |! |FLUX

,'

2 -

i j .

'

| BISTABLE-

.L_ _ _ _ _ ._ _ _ _ __

' .

.T -

.

2/4

TRIP

LOGIC'

RPS TRIP DN LOSS OF MAIN FEE 0 WATER

(SIMPLIFIED) Figure 1~ if

TRIP

TO CROCS' ~

.

'

.

.

TYPICAL RPS CHANNEL ExlSTING,

TRIP STRINGc' .

~ei~.

ADDED EQUIPMENT FOR .

! kTURBINE TRIP.

1 1 -

NTURBINE : T |

TRIPPED 'CONTACT | n

: '

BUFFER |CONTACTS, .

.

I-

.1_ j_ ,

,,

l- I _t_'

'

|I*' -

Flux.- ;

II BISTABLE - --

.| .|.

L_..______ _ _I,

~, .;

T2/4TRIP

-

RPS TRIP DN TURBINE TRIP ' __

LOGIC

_ _ _ _ . _

(SIMPLIFIED) Figiire 2

.

~ TRIP

TO CROCS

. ,

@-.

. . . ..

. . . . .. .

. . ..

%

..

.

.

ANTICIPATORY TRIP

FUNCTIONS

FOR

177 FA PLANTS

2i70 175

.

.' -'*

'

'..

4

.

INDEX

.

1.0 INTRODUCTION

2.0 ASSESSMENT OF POSSIBLE ANTICIPATORY TRIPS

3.0 FUNCTIONAL ANALYSIS

4.0 CONCLUSIONS AND SUMMARY-

APPENDIX A: LOSS OF ONE FEEDPUMP ANALYSIS

2170 1,67

.

9 ,--

. ..

.

1.0 INTRODUCTION

For the purposes of this report, an anticipatory trip is defined as a

trip function that would sense the start of a loss of OTSG heat sink

and actuate much earlier than presently installed reactor trip

signals. Possible anticipatory trip signals indicative of changes in

OTSG heat removal are: turbine trip, loss of main feedwater, low

steam generator 1cvel.

This report evaluates the ef fectiveness of anticipatory trips compared

to the existing high RC pressure trip for a LOFW. Qualitative and

elimination of thequantitative arguments are presented which support

level trips in the steam generators from final design considerations

of anticipatory trips.

Functional response is presented in terms of a parametric study of

time to trip. Thus, irrespective of the plant specific trip signals

and actuation time, the hardware design can proceed with greater

flexibility. That is, by presenting system parameters, such as

pressurizer fill time, as a tunction of time to trip, then if one

plant's turbine trip signal occurs 2.1 secs after initiation of the

event and another plant's trip signal occurs at 2.5 secs, this study

will still be applicable to both.

ECemhM08D6

2170 177

* . . , -*

. . .

9

Page 2

Some of the results presented in this report have already been

submitted to the NRC in Reference 1*, the balance of the information

will be submitted by May 21, 1979. The analyses are performed with

the revised setpoints, i.e. , high RC pressure trip at 2300 psig and

PORV setpoint at 2450 psig. It is shown that anticipatory trips

provide additional margin between the peak RC pressure af ter the

reactor trip and the PORV setpoint, but provide littic additionalwithmargin in the longer term repressurization to the PORV setpoint

continued delay of auxiliary feedwater initiation.

2.0 ASSESSMENT OF POSSIBLE ANTICIPATORY TRIPS

In accordance with Bulletin 79-058, item 5, an evaluation for design

basis for anticipatory trips on turbine trip, loss of main feedwater,

and low steam generator level has been completed.

Ref. 1: B&W teport " Evaluation of Transient Behavior and Small Reactor

Coolant System Breakers in the 177 Fuel Assembly Plant", dated

May 7, 1979.

2170 178''S**"" *

* =<. ..

.

Page 3

.

The evaluation showed low steam generator level not to be anticipatory

and, therefore, it has not been recommended as an anticipatory trip

Figure 2-1 shows the OTSG startup level f rom site data andfunction.

the CADDS ca.Aulated OTSG mass inventory as functions of time

following the TM1-2' event. The time of reactor trip on high RCthat a steampressure is noted on the figure and clearly demonstrates

generator level trip would not have been anticipatory for a level

setpoint that would not interfere with normal operations and

The initial rapid f all in OTSG level occurs as the turbinemaneuvers.

of thestop valves close, momentarily stopping steam flow out

The mass inventory increases during this period due togenerators.

the loss of flow f riction a P. I,y the time the reactor trip occurs, at

8 secondr., steam flow is re-established through the bypass system,

flow fr!.ction aP re-establishes the level and both mass and measured

level .ta rt to decrease uniformly. An OTSG level trip set to trip on

the initial drop shown in Figure 2-1 would need to be set

restrictively high f or normal plant maneuvers and/or lower power

levels.

Further level information (in terms of mass inventory) is given in the

figures for the analysis in Section 3.0. The results for those cases

also indicate that the steam generator low level trip fu,nction would

not be sufficiently fast to be considered anticipatory.

ECSemhM08D6

2170 179

. r ,.-

_

. ..*

.

.

.

F igur e. 2 -1

LOFW (TMI-2 EVENT)-

',2.0 , , ,

160-

.

TRIP DN HIGH RC.

1.6 PRESSURE~

p

120 3-

g-

.=,

a --

8 1.2 fg ;_

\ m"

: / 80~.

g ( \ a

3 \ $. .8 - \

M S40 E-

g

\\.4 - s~~~~_--.--_-

0-

.

0 i i i i.

0 20 40 60 80

Time, seconds

--- -- TOT AL S/0 MASS, CADOSSTART-UP LEVEL, TMI-2 SITE DATA.-

.

2170 180

.<, ..

.

Page 4

Anticipatory trips for loss of feedwater and turbine trip can be

designed to trip the reactor in a more expedient manner than the high

RC pressure trip f or some overheating transients. An anticipatory

trip will provide more margin to PORV setpoint during the initial

overpressurization resulting f rom loss of feedwater and/or turbine

trip. These trips will provide slightly more time to PORV setpoint

and pressurizer fill for delayed auxiliary feedwater initiation

conditions.

3.0 FUNCTIONAL ANALYSIS

A series of LOFW evaluations was performed at 100% lull power

(2772 MWt) with a reactor trip assumed on an anticipatory signal.

With the new high RC pressure setpoint of 2300 psig, a reactor trip

would be expected at about 8 seconds af ter the LOFW. The anticipatory

trip study considered reactor trips with 0.4 sec, 2.5 sec, and 5 see

delays f rom time zero. These studies also included sensitivities to

AFW failure and reactor coolant pump coastdown.

The anticipatory trip study modeled a generic 177 FA plant, and is

considered applicable to raised or lowered loop designs. A feedwater

coastdown similar to that estimated to have occurred at the March 28th

THI-2 event was used to generate sepsrate heat demands for each CADDS

ECSemhM08D6

2i70 181

* .=,, .,

.

Page 5

analysis. The heat demands will change as the reactor trip time is

delayed, because the additional heat input will boil of f the fixed

steam generator inventory at dif f erent rates.

For the cases where" AFW flow was modeled, 1000 gpm was assumed,

starting at 40 seconds. With proper steam generator level and

pressure control, the system parameters will begin to stabilize at

195-290 seconds, depending on trip delay time and RCP operation; see

Tabic 3-1 and Figure 3-12. The PORV will not be actuated, nor would

the pressurizer fill or empty.

With the assumption of no AFW, the PORV will be actuated about three

minutes into the event, as a result of system swell; the pressurizer

fills as 10-12 minutes (see Tabic 3-1). A delay of reactor trip of

2-3 seconds is seen to reduce PORV time to actuate by about one

minute, and pressurizer fill by about 2 minutes. For PORV setpoints

other than 2450 psig, the times will vary and can be determined f rom

Figures 3-3, 3-4, 3-8 and 3-10.

In each of these cases, the mass addition and cooling effect of

expected make-up system operation is not rudeled. One make-up pump

running will add about 10 inches per minute to pressurizer level, and

~1/2% heat de ma nd. It should be noted that the May 7 report used a

ECSemhM08D6

2170 182

.

**r, ,

.

.

.

Page 6

heat demand which reproduced the TMI-2 LOFW event; it has been

reported by the operator that two make-up pumps were running f rom 13

see into the event, creating a higher heat demand than the

assume. This difference isanticipatory trip studies of the report

shown in Figure 3-l'2.

'

The steam generator heat demands, reactor power, RC system pressure,

pressurizer level, and RC inlet / outlet temperatures are given in

Figures 3-7 through 3-11 for the trip on high RC pressure (t=8 see)

The ef f ects of delayed auxiliary feedwater initiation are alsocase.

shwon on the high RC pressure trip curves.

2170 183

ECHemhM08D6

: ...

. ._

.

TABLE 3-1

LOFW EVENT

(LOFW at T=0 sec)

.

TIME OF REACTOR REACTOR AUXILIARY PORV PRESSURIZER S/G Li. VEL CD J)J0L

TRIP (" DELAY") C00LAST PUMPS FEEDWATER OPERATES FULL (400")(P =1025 psig)stm

195 sec-

0.4 Run at 40 sec -

225 sec.

-

2.5 Run at 40 see -

275 sec5.0 Run at 40 see --

0.4 Run None 235 sec 790 see -



2550.4 Coastdown at 40 see --

0.4 Coastdown None 190 sec 700 sec -

LOFW EVENT - TIME =0 sec

REACTOR TRIP AT 2300 PSIG

TIME OF TRIP RCP IJV PORV PRESS. FULL S/C LEVEL CONT.

260 sec8.0 Run at 40 sec --


2170 184

.

.

.

.

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Steam Generator Heat Demand Vs Time Following Loss gFigure 3-6

of Main Feedwater From Rated Power 7-

o40 S AFW DELAY rs

a 1.0 ~_

E "e .8

%- .6 -

E AFW AFW LEVEL~4 ,

n .2 - [ CONTROL -

M INITIATION

0

'.

120 S AFW DELAY

1.0g ,

,

2 .8 --

8 .6 AFW AFW LEVEL-* g

E 4 INITIATION CONTROL~

) /~8 .2 -

' '

0.0.

E 1.0 INFINITE AFW DELAY-

g.8g.6 ,

E .4 - .

M .2 -

0.0-

' ' ' ' '' -

100 200 300 400 500 600

.

Time,~s

.

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Pressurizer Pressure Vs Time Following Loss ofFigure 3-9Main Feedwater From Rated Power

i | I I I'

-

2600 -

mcs~

hON

2400 - "~

3- KEY ca,

,

40 S. AFW INITIATION DELAY,

,

_..g- 120 S. AFW INITI ATION DELAY-

5 >-INFINITE AFW INITIATION DELAY

-

{ 2200 ,

AU \

#

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\f,

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- 100 200 300 400 500 600'

.Time, s.

'F m ye W q ,,

4 diisd W Einl R$ JtlyG A21$ Aid Aand M M M M Jm.J M M W W

Figure 3-10 Pressurizer Level Vs Time Following Loss of,

}!ain Feedwater From Rated Power<w

I I 4 I I

KEY a40 S. AFN INITIATION DELAY s

~.

~

120 S. AFW INITIATION DELAY~~--

00 ~

!< INFINITE AFW INITIATION DELAY~

.

..

} 300 --

i' -

t.=$m

0 200 - -

% -

'~~ben

N100 -s-

x :___-.___ ____

'

.

I t I I'

100 200 300 400 500 600,

Time, s,

.-

Figure 3-11 Core Inlet and Outlet Temperature Vs Time FollowingLoss of Main Feedwater From Rated Power

tn-

Cwi I I I I -

KEY g40 S. AFW INITIATION DELAY s

120 S. AFW INITTAll0N DELAY h_ _ . _ . , -

INFINITE AFW INITIATION DELAY' ~ ''

660-

-

.

W.'' -

. -- 640 -

2 -

5 .

m-

- -

2- 620- .g

1~

-

600-OUT1.ET

-

_

R 580 -

*-

'

,

5 E h,'

560' ' ' ~~'

~,

g - > %_, - ,-=. % wo

INLET -

540 - -

,

-

520 -

_

' ' ' ' '500

100 200 300 400 50ti 600'

- Time, s,

.

c. I j;a,aoaao :xi ca 2a c= = - .. , , , . .

. ,

.

', .

- - 'figure 3-12

ie

,

LOSS OF FEEDIATER AT T = 0 SEC NO AUXILIARY TEEDIATER.

.

t

i b'| 1 i i i

I.*

.

3,i -

.

800 PRES $URl2ER flLLS -

j - *. ,

! ! (RCPRUNNING).

1 v ,

700 -

4

-

,

!

RCP C0AST00;.;,

O!-

600 -

oa i

!6-

2'

INCLUDES C00LINC EFFECT OF WAKEUP/HPl%= 500

-

| 4 5 f L0ir AS REFORTED IN REFEP.ENCE I.

-

.,ie

f E !"

I,'

.vi ;

400 - .-

; ...

t

!

| FORY DPER11ES' -'

300 -

| (RCPRURNING),

- -

100 ,

E!w

;.

|RCP C045100lN;

I'100 - ,

.

il i i 1 i i9

0.4 1.0 7.0 3.0 4.0 5.0 6.0

Tip: to reactor trip ("telay") . seconos

2170 196-.

..

.

.

. .

.

4.0 CONCLUSIMS AND SUMMARY.

I spectrum of delay times, representing anticipatory trips, has been

anal; ad for the loss of feedwater transient. The spectrum included trips

at time zero, with a 0.4 second instrument delay, up to high RC pressure

trip at time 8.0 secoads. Since a high RC pressure trip occurs . cry soon

after a leis of heat sink (overpressurization) transient from 100% FP, only

turbine trip and direct loss of feedwater detection trips would be considered

Anticipator;>.

For all *. rips considered, including high RC pressure, the PORV is not

actuated when normal system operations occur. The pressure rise in the primary

side is less for the anticipatory trips providing additional margin to PORV

lift. If auxiliary feedwater is significantly delayed, then an anticipatory

trip will, at best, provide about 1 minute additional time to PORV lift and

about 3 minutes additional time to filling of the precsurizer. These results

can be seen in Table 4-1 which shows the segocnce of events for a LOFW

transient with trip on high RC pressure (2300 psig) and trip at time zero.

,

2170 197

-

.

.

. .

.

.

.

TABLE 4-1. LOFU-SEQUENCE OF EVENTS COMPARISON

40-s 120-s TRIP AT ZERO,

EVENT AFW DELAY NO AF'a' NO ADJ

Loss of feedwat.er iaitiated 0 0 0 0 (trip occurs)(0.4 delay)

High-pressure trip (2300 psig) 8 8 8 a

PORV opens (2450 psig) a a 175 235

Peak RCS pressure 10 10 175 235

Pressurizer full a a 620 790

* Does not occur for these cases

.

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APPENDIX A*

LOSS OF ONE FEEDWATER PUMP

A special analysis was performed at the B&W Owner's Group request.

This analysis considered the loss of one main feedwater pump with the

plant operating at 100% FP, RC pumps running, no power runback or

auxiliary feedwater initiation and a RC high pressure trip setpoint at

2300 psig. The base parameters for this study are the same as those used

in the " realistic" analysis presented in Section 4.2 of the May 7, 1979,

B&W Repott for 177 FA plants.

The objectives of this study were two-fold:

1) Determine if the PORV will lift under a loss of one feedwater

pump oituation, and,

2) Determine if the OTSG level would be a viabic anticipatory trip,

i.e., how rapidly does the steam generator inventory decrease

in relation to the time a high RC pressure trip would occur.

RC system pressure and pressurizer level as functions of time are

shown in Figu'es A-2 and A-3, respectively. Reactor trip occurs on high

pressure (2300 psig) in 15.8 seconds af ter the loss of one main feedwater

pump. Figure A 4 shows the steam generator mass as a function of time and

only 30% of the mass is boiled off by the time the reactor trip occurs. This

is insuf ficient inventory decrease to cause a level trip in an anticipatory

mode. Figure A-2 shows that no PORV actuation results from this transient.

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Figure A-1'

FEEDWATER COAST 00WN TO 50'J

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100

.

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90 -

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80 -

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70 -

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2170 200'

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0 20 40 60 80 100

Time, seconds-

-

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Figure A-2

.FEEDWATER C0AST00WN TG 505

,

2400,

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Time, seconds

2170 201-

.

. .

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Figure A-3

FEEONATER C0ASTDOWN TO 50',

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Figure A-4

LOSS OF ONE FEEDPUMP-RAMP TO 50:; IN 10 SEC

TOTAL S/G MASS, CADDS

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| 2170 203

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2170 207

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F ,if 5h b- .a. m

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