ML20236F998

Forwards Rept Which Represents Best Estimate Analytical Study of Potential for Increased Pressurizer PORV Opening Resulting from Turbine Trip W/O Reactor Trip Below 30% Power Transient (P-8 Setpoint Study),Per NUREG-0737

ML20236F998
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	10/30/1987
From:	Ainger K COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.10, TASK-TM 3789K, NUDOCS 8711020369
Download: ML20236F998 (12)

Text

,.

q w[-1 f/

[.T Common sealth Edison 1 1

A-"

.l 2 f: One First Ntional Plaza, Chica00, Illinois

-]

7 l Address Reply to: Post Office Box 767.

)

!A v

% d-Chicago, Illinois 60690 0767 -

.]

p,

)'

October-30, 1987 i

f,U.S. Nuclear Regulatory Commission-ATTN:, Document Control Desk

~ Washington, DC 20555 l

l

Subject:

' Byron Station Units 1 and 2 Braidwcod Station Units 1 and 2-Anticipatory Reactor Trip Upon Turbine Trip NRC' Docket Nos. 50-454/455 and 50-456/457 i

Reference (a): September.30,f l987,fletter from K. A. Ainger

-j to U.S. NRC-1 1

Gentlemen:

Reference'(a) transmitted an application for a license amendment

-involving a plant' modification and associated' Technical Specification changes.

The modification increases the setpoint for enabling the anticipatory reactor trip upon turbine trip from greater than 10% reactor power to greater than 30%

~

-reactor power. The NRC Staff has re'juested Commonwealth Edison to address a concern regarding the potential; increase.in probability of-a stuck-open-pressurizer power' operated relief valve (PORV) as_a result of implementing the modification described above. The NRC position pertaining to this issue is

' addressed in NUREG-0737, Item II.K.3.10.

I

-Enclosed is a report which presents a best' estimate analytical study

-to show that no additional pressurizer.pORV challenges are expected due to j

implementation-of an interlock system that would eliminate' direct reactor t

trips on' turbine trips below 30% power.

please' direct any further questions regarding:this matter to this

-office.

-Very truly yours, 87 I

K. A.'Ainger j

Nuclear Licensing Administrator

-]

1 Im l

5

't j

)

-Enclosure l

l cc: ; Byron Resident Inspe'ctor I

Braidwood Resident Inspector._

j NRC Region III office i

i-8711020369 871030 h

1 L

PDR ADOCK 011000454 -

I 1

l-3789K-P PDR 1

l J

i i,'4. f,;,,,9

[

(

e f

'I

.i ATTACHIENT 4 j'

. A STUDY OF THE POTENTIAL FDR INCREASED PRESSURIZER FORV l

l-OPENING RESULTDG FROM TURBINE TRIP WITHOUT REACTOR 1 RIP L

BELW 305. POWER.1RANSIENT E

(P-8 SETPOINT STUDY) e t

1 l

i i

)

(

8

__m

.__.m.._____

T q.',

l I

~ ]. 9, '.

8,.;

INTRODUCTION The Commonwealth Edison' Byron and Braidwood Units are designed with 50% load rejection' capability.

As~a result of this capability, the Westinghouse design-I criterion is1that11oad_ rejections up to 50%'should not require a reactor trip l

u

. if all other ' control systems function properly.

To take advantage of this-capability, Westinghouse has proposed to implement an interlock system that would eliminate. direct reactor trips on turbine trips below 50% power, thereby i

decreasing unnecessary challenges to the reactor protection system and increasing plant availability.

For Byron and Braidwood Units, the existing

)

permissive setpoint P-8 (P-8 is currently used to allow one loop loss of flow 1

below 30% power) has been chosen to provide the proposed interlock function

~

(i.e., delete reactor trips on turbine trips below 30% power). The NRC has a

- expressed concerns regarding the potential increase in probability of a

.)

stuck open pressurizer power operated relief valve (PORV) following the i

- implementation of deletion of reactor trip on turbine trip from high power

' (power. level greater than 10%).

The NRC position is addressed in NUREG-0737, Item II-K.3.10. The information contained.herein presents a best estimate analytical study to show-that no additional pressurizer PORV challenges are.

l expected due to implementation of an interlock system that would eliminate 1

direct reactor trips on turbine trips below 30% power for Commonwealth Edison Byron and Braidwood Units.

.II.

SYSTEM TRANSIENT ANALYSIS-

II.1 -Description-of the Analysis:

A best estimate analytical study was performed to determine the transient plant response to a turbine trip without a reactor trip from 30% power.

The analysis was performed using the LOFTRAN computer code (I) model of the Byron and Braidwood Units. -This computer model simulates the overall thermal /

hydraulic / nuclear response of the NSSS a's well as the various control and 1.

(1) LOFTRAN Code Description, WCAP-7878, Rev. 0 - Rev. 3.

l 6484e:1d/061087 A-2 i

q

^

f

.i,

.)

-protection. systems. Since the object of this study was primarily to determine the peak in pressurizer pressure following the initiation of the. transient, assumptions were made that would contribute to~a conservatively high prediction of pressurizer pressure.

These assumptions were the following.

1..Beginning-of-Life (BOL) reactivity parameters were used since this gives the minimum moderator feedback, and consequently, the minimum l

-decrease in nuclear power as a result of the initial increase in primary coolant temperature during the transient.

L i

2.

Transients were initiated from 32% power (2% calorimetric error in

adverse direction):since 30% power is the maximum proposed value for the P-8 permissive setpoint that would permit'a turbine trip without

]

actuating a direct reactor trip.

Transients initiated from a lower power level.would be less severe with respect to predicting the peak' in pressurizer pressure.

This is true since the peak in pressurizer

.)

pressure is directly related to the amount of energy that'must be stored in the primary system during the mismatch between core power production and, secondary system load.

A turbine trip from a lower

' initial power level simply results in a smaller power mismatch and

'I this results in'a smaller peak in pressurizer pressure.

In the limit, the initial power level'of the transient would be reduced to 10% power which-is currently the power level at which a turbine trip without a reactor trip is permitted.

3.

The pressurizer model in LOFTRAN is conservative with respect to over predicting peaks in pressurizer pressure. This is because the pressurizer pressure calculation model in LOFTRAN is isentropic.

Comparison studies (2) have shown that such models (isentropic model) i overpredict pressurizer pressure during pressure-increase transients.

(2) Baron, R. C., " Digital Model Simulation of a Nuclear Pressurizer," Nuclear

. Science and Engineering 52, 283-291 (1973).

' 6484s:1d/061087 A-3

=

4 71.

.'4.

Byron and Braidwood units are implementing a program to reduce the l

l potential:for initiation of primary water stress. corrosion cracking'in g

the-steam: generator tubes by lowering the primary side temperatures with respect.to'the or_igina1l design temperature. Since the reduced-i T-hot parameters (i.e., primary temperatures,. control systems setpoints) are more severe with respect to predicting the peak in the

^

pressurizer pressure, these reduced T-hot parameters are used in the best: estimate analysis.

11.2 Analysis Results:,

-The' expected-system response to a turbine trip without'a reactor. trip from 30%

power is shown in figures 1A through 6A.

For. normal plant operation.with all i

normal control. systems' assumed operational, the pressurizer pressure does not j

reach-thepointofpressurizerPORVactivation(PORVsetpointforthe-Byron

'i and Braidwood Units is 2350 psia).

Results also indicated that steam i

generator PORV does not open during this transient. Note that in figures 1A-

'through 6A,-transient! initiates after 10 seconds of steady state.-

III.

FAILURE' MODES ANALYSIS.

.j For normal plant operation with-all normal control systems assumed

operational, this transient-does not result in opening the pressurizer PORVs.

However,<the NRC has: expressed a concern that the implementation of a turbine trip without a' reactor trip below 30% power permissive should not result in l

increased challenges to the pressurizer PORVs even in the event of " degraded" q

control system performance.

A sensitivity study was therefore performed in

{

which certain failures were assumed to occur in the control systems that

)

influence the course of this transient in' order to determine their effect on the potential for pressurizer PORY challenges.

There are three main control systems that act during this transient:

the

. steam dump system, the pressurizer spray system, and the rod control system.

The steam dump system consists of twelve valves which are arranged into four l

banks, three valves per bank. A single credible failure was assumed to be'the 6484e:1d/061087 A-4

^

(

- failure of two banks of steam dump valves to open on demand following the turbine trip.

In the pressurizer spray system, the type of failure assumed was a reduction in spray flow capacity (due, for example, to a sticking spray I

- controlvalve).

The failure that was assumed in the rod control system was I

the failure of the power mismatch channel.

The purpose of the power mismatch channel signal is to provide a fast feed-forward signal to the rod control system during a rapid change in turbine load.

If this signal is not present, than the rods are controlled only by the Tavg error signal which has a much slower response and thus it takes longer time to begin driving the rods into the core following the turbine trip.

III.1 Failure Mode Analysis Results:

The results of the failure mode sensitivity study showed that for normal system operation and for any single failure that was considered, the pressurizer PORVs did not open.

In fact, it takes a combination of multiple control system failures to result in pressurizer PORV opening during the transient. '

IV. CONCLUSIONS Based on the best estimate analysis results the following conclusions are made:

1.

For normal plant operation with all normal control systems assumed operational, the implementation of a system (P-8 permissive) that i

permits a turbine trip without actuating a direct r'eactor trip below 30% power will not result in opening the pressurizer power-operated relief va'Ives.

2.

For any single failure in the control system that was considered in the analysis, the implementation of a system (P-8 permissive) that p'ermits a turbine trip without a direct reactor trip below 30% power will not result in opening the pressurizer power-operated relief valves.

It was found that, it takes a combination of multiple control system failures to result in pressurizer power-operated relief valves opening during the transient.

6484e:1d/061087 A-5 i

___-m____.___m.

i y

{#

l-CAE/CCE P-8 BEST EST.ANA.FROM 307. POWER CAE/CCE P-8 TT WITHOUT RT FROM 307. POWER

.55 T.

.5 xoz.45 u.

.4.

U c.55 m

L.

.5 mu

. 8 25 o.

m

.2 cw tj.15 a

~

~

.1

.05 0.

O.

20.

.40.

60.

80.

100, 120. 140. 160. 180, 200.

TIME CSEC1 Figure 1A A-6

CAE/CCE P-8 BEST EST.ANA.FROM 307. POWER CAE/CCE P-8.TT WITHOUT RT FROM 307. POWER 2400.

E y 2550.

E g

a 2500, d-

.d a.

m'2250.

w

+

Cf

.g.2200.

d a.

'2150.

l l

l 2100.

O.

20.

40.

60.

80.

100. 120. 140. 160. 180. 200.

l TIME (SEC) ll Figure 2A A-7

3 g

[. hj.I

)(?

/

-i L,

, m. s 1 c/ >y e

n m :.

e r, e., ; }:

g

+s,.

1 y

o q

t

'y

..s.j.{

7 w

-7 a ;}t 4

i g

m m; - "

o

>, d i

)

i s

.l i f.l r

. h, -f..

, 4,,.

i t-

~l.;1

..s-

.-q :

+

j

. t,

e.

., p )( r./

, L. '

[,"A

'gf' f

j,r l

-/

,j im

^'t/

,f.

9 a

Jj >,

~

,l.

~,..

)

L

/.

(;.

c

g..

i p,-

.,, q, k il'~ A-

,.e

3. <,

e.,'

l

-ec

'i a

,y

' s.

d s,,

/,

r. r.

.<c7n W

-N...

um g'., y O, m p ST.ANA.,FRON;307.* POWER - -

...j -

t-cj -

5

)n 9'

1E

" L.,f

.CAE/CCEJP-9)GGS.h H

1C

.. D / '1

!. 'k ;

!/f/7'

)

~

wa.

e Tf f

,l y y CAE/CCE1P-8 TT M1THOUT:P)T FROM-307.:P0yER T

+

is s

' ' 9 ' '. ff,.' []: '

~

j. / !-

'f}

~~d, r

W 7

4 }.,.

or-l b

i'[ J D tO (j!' -

n),

s

~-

Ef 7590.

p'O -

, 's g...,t

- g,s h s

n q.

2 n-585.

1 x

y y

. d590.1

,. &a

d!/

at d

.m w.

79 v,p

. c 9

a.;

e:

.y a

-i 9

575

-r y

., p.

t n ;.

"., se j

r ;w.

-2, a.,,.

'j-,

,f j

' ' E570.:

,, n o. J1.

f o.

-e-

.j n

s -

>1 y4 g

.. ' ?; u 11 9t rg..

7 F, c,'-

wa.565.

.i o

4 m

.,r. n

4. 4, 7a, a

E".

,,~

+

- 7 w

V:560.

T

&T-'

4

.w 1,555.

g

+1

. w.

i. p g.

s'

'C 550'.

't ~

4

/-

I 545.-

1 y-

/

g,

-540.

O..

20.

.40.-

60.

80.

100. 120. 140. 160~. 180.'200.

. TIME (SEC)

' I i

V st-ll s'

ll '

-y

~

'T l.

S.

1 i

,ei a

y l

c

,o a.

Figure 3A

/ l

~~A-8"

, '/

1

.(

./

}O

.M.

• g:..

i>

,.c, ;. q,.-. *

>4

,s

.y. ; s,

'2 $_

.. +!.i.

, ( f t;:l w.

CAE/CCE[P-8BESTEST.ANA.FROM307.. POWER-

q, CAE/CCE P-8.TT.WITHOUT RT'FROM 30% POWER' y^.

4 t

1200i Hx

,f !150.

.u:

.w

.r..-

To?!l00.

._1.

.o..

g.

$!! 050.

a u :a,

t

-u, u

w-

~

N !j @@@,.

m.

..e o

O en.

en

wul 950' 4

.a;

+.

900,.-

O.

-20..

40.

-60. 180.

100. 120.-140. 160. 180. 200.

p' T :.

=

(SEC)'

m -

. TIME.

w., 4 9

.w j',

Figure.4A A-9 e

.3 -

y

,,cl$f.

m,<4

(;,

p;

+

n a

. w..

L l

(

lc i

4 4

.j'f.

'CAE/CCE P_-8.BEST EST.ANA.FROM 307. POWER' m:

CAE/CCE'iP-8.TT WITHOUT RT FROM:307. POWER

.'!'160..

.e w

1140; 4

.- 11120..

r

_c :c j

((1100'.

p c

~

u 1080.

l

... m -

u a'1060.

,m

.w.

u

Nn.'1040.

r 1020.

c-

.w

.n.1000..

1 980.-

~1

.-.960.

940.-

.0.

. 20.. 40.

60.

80. '100..120. 140. 160. 180. 200.

TIME

'(SEC)

Figure 5A A-10 t%

l-.

..o.

9g e

.f T

e l

l CAE/CCE P-8 BEST EST.ANA.FROM 30% POWER CAE/CCE P-8.TT WITHOUT RT FROM 307. POWER 1.

~

.8 1

5.8 z

E3 7

Qo.6 a-m u.

.5 234

. u.

E.5-u)

W.2

.1 0.

O.

20.

40.

60.

80.

100. 120. 140. 160. 180. 200.

TIME (SEC)

Figure 6A A-11 l

_ _ _ _ _ _ _ _ _ _. _ _ _ _