ML20086M889

Forwards Response to IE Bulletin 79-21, Temp Effects on Level Measurements. Analysis of Adequacy of Steam Generator Low Level Setpoints Encl

ML20086M889
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	02/09/1984
From:	Barnes P COMMONWEALTH EDISON CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
8126N, IEB-79-21, NUDOCS 8402170134
Download: ML20086M889 (8)

Text

.

N Commonwealth Edison A

) One First N:tiornt Plaza Chictgo, lihnois Q

F>

T O J Iddress Reply to: Post Offica Box 767 y

Chicago, Illinois 60690 February 9, 1984 i

Mr. James G. Keppler Regional Administrator U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137

Subject:

Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Response to IE Bulletin 79-21

" Temperature Effects on Level Measurements" NRC Docket Nos. 50-454/455 and 50-456/457 4

References (a):

August 13, 1979 letter from J.

G.

Keppler to B. Lee.

(b):

January 26, 1982 letter from T. R. Tramm to H. R. Denton i

Dear Mr. Keppler:

Reference (a) addressed water level measurement errors for operating reactors.

Reference (b) committed us to address water level measurement error at Byron Station Units 1 and 2.

The attachment to thin letter 3s the analysis performed which provides information concerning the adequacy of the present steam generator low level setpoints.

The analysis utilized the containment pressure safeguards-actuation setpoint as a backup to the steam generator low level trip for auxiliary feedwater initiation.

Westinghouse incorporated this I

analysis into the statistical setpoint study.

FSAR Section 14.2.8 addresses the containment high pressure safety injection in the accident analysis for.feedwater system pipe break.

We believe that.this submittal adequately addresses Byron Station SER Item 7.2.2.3.

8402170134 840909 PDR ADOCK 05000454

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FEB 13196(-

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. To the best of my knowledge and belief the statements contained in the Attachment are true and correct.

In some respects these statements are not based on my personal knowledge but upon information furnished by other Commonwealth Ediaan employees, consultants or contractors.

Such information has been reviewed in accordance with Company practice and I believe it to be reliable.

Please sddress questions regarding this matter to this office.

Respectfully,

f. d. W" P. L. Barnes Nuclear Licensing Administrator cc:

Document Control Desk RIII Resident Inspector - BY/BW SUBSCRIBED and SWORN to before me this day of % m )f M, 1984 G.,A aMada Notary Public 8126N I

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~ _,.,_

e Byron /Braidwood STEAM GENERATOR REFERENCE LEG Th'e potential for steam generator level bias was originally addressed by Westinghouse.

Westinghouse postulated that a high-energy line break, specifically a feedwater line break, could result in a heatup of the steam generator level measure-ment reference leg.

7c increase reference leg water column temperature would result in a decrease of the water column density with a consequent increase in the indicated steam generator water level.

This bias could delay or prevent reactor trip and auxiliary feedwater initiation which are based on low-low steam generator water level.

The loss of fe'edwater accident is protected by a reactor trip on low-low steam generator level.

Steamline breaks and LOCA's arc protected by other reactor trips and safety injection signals.

The low-low steam generator level trip setpoint was normally set at 37.5% (Unit 1) and 17% (Unit 2) of the narrow range level instrument span.

As a result of the Westinghouse concern, a

review of the potential inaccuracies was performed at Zion Station and concluded that a maximum of 4.5% inaccuracy could exist exclusive of hcat induced bias.

Westinghouse generically estimated the worst bias that might exist for the 1cw-low steam gen.erator level due to reference leg hectup would be 10% for temperatures up to 230'F.

Above 280*F, reactor trip and auv.iliary feedwater actuation would occur on a high containment pressure signal.

By utilizing the present containment ' trip setpoint of 5 psig, the maximum temperature reached prior to unit trip is 160'F.

This corresponds to a 2.55% ' temperature induced bias, rather 4

than the 10% generically determined by Nectinghouse.

Based on the above values, the steam genera tor low-low trip setpoint can be 37.5% without affecting plant safety.

9 e

e 4

0 4

P D

6 w

-g e

-r

+,

T'*

l

~ 7 ATTACHMENT A.

TERMS 4

PT - Total Pressure in Containment

• P* s - Partial Pressure of Steam

- Partia'l Prossure of ' Air P

a T, - Tertperature of Containment V' - Volume P

Pressure R' - Number of Air Molecules in Containment 3

4 1

11.

Assumptions (and Justification) 1.

Ierfect Gas Iaw Applies'. PV=R ' T '

t 2.

Law of Partial Pressure Applies'P =P =P T

3 a

3.

Initial Conditions:

Trange=65-120*F-Assume T;=90*F i

Prange=.1 to 4.3 psig Assume =0 psig=14,7 psia

' P :-0. 69 psia S

i P =14.01 psia:

4 a

4.

Saturation Conditions Simultaneoucly).

(Liquid and Vapor phases present

~

5.

Localized heating effects are minimized due to the--

routing of' the four referrence legs.

taps are located within the same 90* quadrant.)-(However, the four i

k s

,F l

g

~1-4 A

t

F

^ ~

'..~

1

.~

\\

t' f

L-6.

Equilibrium. conditions with respect to pressure (pressure can be considered 'tx> be in equilibrium

~

almost instantaneously. because pressure fronts move with sonic velocity).

i.

Temperature equilibrium is reached slower than pressure- (trip will be initiated prior to contain-ment temperature reaching equilibrium conditions and. calculation is conservative).

8.

Containment'high pressure trip setpoint = 5 psig

_= 19. 7 psia 9.

a.

No air being added PV PV

=-

a aa Ta i

Ta f

b.

Vol.ume. is Cons tan t

~,. _,,, -

Pa

=

Pa 10.

Bias Formula E =: HL=pt, CAL - p 7 "H ~

Pf, CAL - Pg, CAL I Where:

E = Level error due to reference leg heat up as a fraction of level span.

H = Level span = Vertical distance between-narrow range taps on S.

G.

(233").

H = Height of reference leg = maximum vertical L

distance from lower tap to water level'in condensing POT cn upper tap.

(233")

pt, CAL = Water density at containment tempera-ture and S.

G. pressure for which the level indication'was calibrated - (90*F, 62.12i/FT3) ph = Water density in reference leg. at time of in-te re s t.

pg, CAL pg, CAL = Difference between saturated water density and dry saturate,d s team density a t -the S. G. pressure for which the level indication-sys tem was-calibra ted.

An upper v-bound prassure must' be assured.

11.

Calibration' temperat'ure = - 90 F 9

Cal..bration pressure (@ 50% power) = 105 5' psia-

• (S.. G. Secondary pressure) e w

+-

p

-g

,~-m

,e ywa r

-w-

.r

' q C.

Calculatio ns --

la.

Maximum' temperature (Containment at 90*F)

P

=

P a

a T

1.

T a

a f

a t

P T,f = Pa,f a.i x

a Ta,i Pa,f (x + 460) x 14.01 psia

=

550 lb.

T = 180*F Pa,f = (180 + 460) x'14.01 = 16.30 psia 550 P3 = 7.51 psia PT = 23.81 psia

~

t ic.

T = 140 Pa,f = (140 +-460) x 14.01 psia 550

= 15.28 psia P3 = 2.89 PT = 18.17 psia

~ ~ '

t ld.

T = 150 -'

P,f = (150 + 460) x 14.01 psia a

550

= 15. 5 3, psia P3_= 3.71 PT

• 19*24

~

- 4 h

i 3_

.,0..'

s e

le.

T = 160 P

,f_= (160 + 460) x 14.01 psia a

550

= 15.53 psia t

P3 = 4.749 psia PT = 20.53 160*F gives a conservation valub of 20.53 psia, which is over the containment high pressure trip setpoin t.

2.

Reference Leg E=HL=DL CAL - O n

H p f, CAL P g, CAL P'L, CAL @ 90*F = 62.12 I/FT3 P L 0 160*F = 61.01 t/FT3 P!, CAL Ag, CAL

• 1 - l_

i V

V 0 1050 psia f

g, CAL 1

1

= 45.87156 - 2.3590

=

.0218

.4239

= 43.51 t/FT3 Hn = 233 = 1.00 F

737 E = 1 (62.12 - 61.01) 43.51

~'

= 2.55%

3.

Setpoint conservatism Instrument Error =

4.5%

(Zion Information),

Temperatur'c induced error = 2. 55%

7.05%'

Based on the Accident Analysis on page 15.2-19 of the

'S.TR, which assumed a reactor trip at -10% of low low level setpoint, it.can be seen that:

v 10%

l

7. 0E S.

2.95% is the conservatism in the setpoints pres'ently being ii.ed.

~ 1

• ~.'.

• I.

SAR References

~

15.. Accident Analysis page 15.0 - 33 Limi' ting trip point assumed in analysis 32. 3t-low SG level 87.4%-Hi--SG level.

page 15.2 - 19.Feedwatar System line break sectioq 15.2.8.21 Reactor trip is assumed to be initiated when the low-low level trip setpoint minus 10% of the L

narrow range span in-the' faulted SG is reached.

16.

Technical Specifications Page 16.2-6 U1 SG low-low level 3,37% Trip

> 36% Allowable

~

V.

U2 SG low-low level 3,16% Trip-3,15% Allowable

-II.

Precautions, Limitations & Setpoints Document Revised by CAW-4194 - (CBW-3420) of March 17, 1982.-

Section lAlc - SG Control system level +5% (pagc 1)

• Section 1B2G - SG Low-low level 40.8% Unit 1 17.0% Unit 2

• Section 1B2H - 'SG Hi-Hi leve'l 82.4 % Unit 1 78.1% Unit 2 Section c-SG level Control.66% Unit 1 (Page 32)~

50% Unit,2 (Page 33)

• Change with March 17, 1982 letter' III.

Setpoint Study WCA 9 640 N'ovember, '19 79

~

Section 3. 3.12 - Low-lou SG level Trips 3-71-Unit 1 17%-Unit 2 Section 3. 3.13 - Hi-hi-SG level Trips

.82%-Unit 1 78%-Unit 2, i

N 1.

9

-