ML20237H131

Forwards Response to Request for Addl Technical Info on Reactor Risk Ref Document (NUREG-1150),per NRC

ML20237H131
Person / Time
Site:	Sequoyah
Issue date:	08/11/1987
From:	Gridley R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML20237H133	List: ML20237H131 ML20237H155 ML20237H173 ML20237H196 ML20237H257 ML20237H274 ML20237H281 ML20237H298 ML20237H348 ML20237H361 ML20237H365 ML20237H400 ML20237H412 ML20237H421 ML20237H430 ML20237H438 ML20237H452 ML20237H464 ML20237H480 ML20237H492 ML20237H501 ML20237H508 ML20237H517 ML20237H532 ML20237H547 ML20237H559 ML20237H593 ML20237H602 ML20237H616 ML20237H629 ML20237H641 ML20237H649 ML20237H654 ML20237H659 ML20237H671 ML20237H682 ML20237H689 ML20237H696 ML20237H699 ML20237H703 ML20237H714 ML20237H719 ML20237H726 ML20237H735 ML20237H739 ML20237H770 ML20237H796 ML20237H852 ML20237H899 ML20237H967 ... further results
References
RTR-NUREG-1150 NUDOCS 8708170074
Download: ML20237H131 (12)

Text

--_-- - _ - - - - _ _ _ - _ - _ _

i TENN ESSEE- VALLEY ' AUTHORITY CHATTANOOGA TENNESSEE 37401 SN 157B Lookout Place ,

{

~FD - N AUG 111987 33' U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Attention: Document Control Desk Washington, DC 20555 Gentlemen:

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 - NUREG-1150 (REACTOR RISK REFERENCE DOCUMENT)

Please refer to my letter dated July 31, 1987.

Enclosed is TVA's response to the request for additional technical information on Scquoyah submitted by James G. Keppler's letter to S. A. White dated July 23, 1987. If you have any questions, please telephone D. L. Williams at (615) 632-7170.

Very truly yours, TENNESSEE V EY AUTHORITY R. L. Gridley, Director Nuclear Safety and Licensing Enclosures.

cc (Enclosures):

Mr. G. G. Zech, Assistant Director Regional Inspections Division of TVA Projects Office of Special Projects U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. J. A. Zwolinski, Assistant Director ..

for Projects Division of TVA Projects  ;

Office of Special Projects U.S. Nuclear Regulatory Commission 4350 East West Highway 'Og6 EWW 322 ig Bethesda, Maryland 20814 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee, 37379 6700170074 870811 An Equal Opportunity Employer PDR ADOCK 05000327 1 _ _ _ _ _ _ _ _

PDR

ENCLOSURE Response to Request for Additional Information Letter from James'G. Keppler to S. A. White Dated July 23, 1987 l Comments

1. Please provide the probability data and the calculational basis to  !

justify a frequency for -loss of offsite power of less than 7E-2 per year.

8. We are currently using Cluster 7 from NUREG-1032 for probabilities for failure to restore offsite ac power. Please provide any justification or

-data you have to support a lower (i.e., better) curve for recovery of ac power at Sequoyah.

R_esponse

-Questions 1 and 8 related to the frequency and duration of loss of  !

offsite power. NUREG-1032, " Evaluation of Station Blackout Accidents at I Nuclear Power Plants (Draft)," is cited as the reference for this information in the NUREG-il50 analysis. However, TVA's analysis based on .

the criteria outlined in NUREG-1032 indicates that Sequoyah'should be  ;

considered a cluster group 2 plant for loss of offsite power initiating  !

event frequency instead of cluster group 7'as assumed in the NUREG-1150 analysis. Therefore, we believe using the cluster group 2 data is appropriate for SQN. Attachment A provides information previously.

supplied to EI Services (NRC subcontractor on this program).

Comment

2. Please provide your estimate of battery depletion time during station blackout sequences and the basis.

Response

The Sequoyah design basis for battery depletion time under station 3 blackout conditions is two hours (reference FSAR section 8.3.2). The  !

125V dc batteries at Sequoyah are tested at least once every 18 months in accordance with SI-105 (Attachment B) to ensure a minimum discharge time of two hours. Also, the manufacturer specification on the batteries specifies a minimum of two hours to discharge.

i A separate TVA analysis supporting the four-hour battery depletion time used by NRC in the NUREG-1150 analysis does not exist. However, additional information is available to support this assumption. A successful test simulating loss of ac power was performed by TVA as a part of the special low power testing program at Sequoyah. Also, as part of a station blackout risk study for TVA's Browns Ferry Nuclear Plant performed by Oak Ridge National Laboratory (NUREG/CR-2182), TVA determined that the four hour battery depletion time initially assumed in the study could be extended to seven hours. We expect that a similar analysis for the Sequoyah batteries would provide comparable results.

TVA plars to pursue this matter as part of the current industry initiatives to address NRC's unresolved safety issue A-44. Given this data and our experience with risk-based evaluations, it is our judgment that the battery depletion time used in NUREG-1150 is reasonable.

n t l

. Comt ;_n <

3. A primary sequence of interest is the reactor coolant pump (RCP) seal loss of coolant accident (LOCA) after loss of all seal cooling due-to i total loss of plant service water or component cooling water. Please provide any test results or engineering analyses to establish:

a. the time to failure of RCP seals after loss of seal cooling (and leak rate); and

b. the ability of the centrifugal charging pumps to operate after a failure of the component cooling system.

Response

a. The time of failure of RCP seals after loss of seal cooling (and the leak rate) is available in Westinghouse NCAP-10541, " Reactor Coolant Pump Seal Performance Following a Loss of All AC Power," Revision 2.

A copy of this HCAP has previously been submitted to NRC by the Westinghouse Owners' Group (reference Westinghouse letter OG-206 dated December 10, 1986, from L. D. Butterfield to J. Lyons). This information applies to SQN.

b. Testing has been conducted to show the centrifugal charging pump (CCP) operation is not degraded by the loss of component cooling (CCS) for a period of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This testing is discussed in an April 7, 1987 letter from H. J. Hitchler (Westinghouse) to H. Mims (TVA), a copy of which is provided as Attachment C.

Comment

4. Please provide thermal / hydraulic analysis results showing containment pressure time histories for small (2" and 1/2") LOCAs after actuation of the containment sprays.

Response

Containment pressure time histories for small break LOCAs (2-inch) are covered by IDCOR Task 23.1S; "Sequoyah Nuclear Plant: Integrated Containment Analysis," a copy of which is provided as Attachment D.

Similar information on 1/2-inch LOCAs is not available; however, it is our understanding that the submittal of this IDCOR report is responsive to the staff's request.

Comment

5. Please provide the results of analyses which show rates of temperature rise in the switchgear rooms upon loss of heating, ventilation or air conditioning (HVAC) to the rooms.

Response

Engineering calculation TI-ECS-61, Revision 1 (Attachment E) provides

" Electric Board Room Temperature Profiles Versus Time Upon Loss Of Cooling." This calculation assumes loss of all HVAC and considers operator actions. In addition, engineering calculation TI-ECS-69 (Attachment F) provides a similar analysis assuming availability of an air handling unit to circulate air within the affected spaces.

Comment

6. Please provide plant-specific data for diesel generator failure to start and diesel generator test / maintenance outages.

Response

Attachment G provides information concerning diesel generator availability at Sequoyah. TVA agrees that the diesel generator failure rates used in NUREG-ll50 are reasonable.

Comment

7. Please send latest revision of all emergency procedures, functional restoration procedures, emergency contingency actions, and operating procedures for loss of component cooling water, essential raw cooling water, instrument air, and a vital bus.

Response

Attachment H identifies the procedures requested by NRC. Copies of these procedures are enclosed in this transmittal.

l l

1 I

.___--______-_____-_____.D

9 2 Q

&77ACHMEUT* A i

1-

i

.t.

T

'l E E 's 8 4 5 3

• \ s

.$ .z -

M 7 55'O I '

N Ikq.. Offsite

_----.. Power E k h8 Cluster

.01 = s e 5,9 0

s a% s ..s 2

y

~

' s ,, 's % N

-.%~~~~*

' s, w -g43

~% ?

< I Q

C 001 N - .~

N

u. 'S.'

0  :

3 N' -a. ' g' D' .%\

1 s

<  : N

%o ' -o. 6 E ~

'~ c 3 Q

m .0001 -

0,~O2 ,

W 5 1

.00001 ' ' i i e i i ,

0 2 4 6 8 10 12 14 16 DURATION (Hours) i i

l I

Figure A.14 Estimated frequency of occurrence of losses of offsite power exceeding specified durations for nine offsite power clusters NUREG-1032 A-42

_ _ _ _ _ _ _ _ __ _ _ 1

A)

Table A.11 Identification of grid (GR), offsite power system design (I),

severe weather (SR), and extremely severe weather (SS) factors included in nine cluster groups Cluster Group 1 Cluster Group 2 (Cont'd) Cluster Group 3 (Cont'd)

GR I SR SS GR I SR SS GR I SR SS 1 1 1 1 2 1 3 1 1 3 5 1 1 1 2 7 2 1 4 1 1 3 5 2 1 2 1 1 2 1 4 2 2 1 3 2 1 2 2 1 2 2 1 2 2 1 5 1 2 1 1 1 2 2 2 2 2 1 5 2 2 1 2 1 2 2 3 1 2 2 3 2 2 2 1 1 2 2 4 1 2 2 5 1 2 2 2 1 2 2 4 2 ,, 2 2 5 2 2 3 1 1 2 3 1 2 Cluster Group 2 2 3 2 1 2 3 2 2 2 3 3 1 GR I SR SS Cluster Group 3 2 3 3 2 2 3 4 1 1 1 1 2 GR I SR SS 2 3 4 2 1 1 2 2 2 3 5 1 1 1 3 2

~

1 1 3 1 2 3 5 2 1 1 4 1 1 1 5 1 1 1 4 2 1 1 5 2 Cluster Group 4 1 2 1 2 1 2 3 2  !

1 2 2 2 1 2 5 1 GR I SR SS 1 2 3 1 1 2 5 2 1 2 4 1 1 3 1 2 1 1 1 4 1 2 4 2 1 3 2 2 1 1 2 4

$h3 1 3 1

2 1 1 3 3 1 1 1 3 4 1 3 3 2 1 1 4 4 2 1 1 2 1 3 4 1 1 1 5 4 2 1 2 2 1 3 4 2 1 1 6 3 NUREG-1032 A-44

yp

~.

o N

M l Table A.6 Grid reliability / recovery jl Grid reliability (G) h Grid reliability group (G) Frequency of grid loss l

l

( ) Less than 1 per 60 site-years (0.01/ site year) h G2

' > 1 per 60 site years and

< 1 per 30 site years h; (0.03/ site year)

{ G3

> 1 per 20 site years and j < 1 per 6 site years

{0.1/siteyear)

G4 Greater than or equal to 1 per 6 site years

,, (0.3/ site year)

Recovery (R)

Recovery from grid blackout group (R) Recovery capability OR1 Plant has capability and procedures to recover offsite t

l (non-emergency) AC power to the site within 1/2 hour following a grid blackout.

R2 All other plants not in R1.

Grid reliability / recovery (GR')

Grid reliability / Grid reliability Recovery from grid recovery group (GR) group (G) blackout group (R) bGR1 G1 R1 M G2 R1 GR3 G3 R1 GR4 G4 R1 GR5 G1 R2 GR6 G2 R2  !

GR7 G3 R2 i,

NUREG-1032 A-20

Y ~

h' ll .  !

I l 01 Table A.3 Mean time to restore offsite, pov r and statistical test values '

for plant design groupings [I '

ll.

Group Design Mean time to restore ;i designation features

• offsite power (hrs) p II A3 and (B3, B4 or B2b) O.13 h

12 A3 and (81 or B2a) 0.21 yi I3 oB4, h or andB2b) 0.50 ll p

It (Al or A2) and 4 (81 or B2a) 0.97  ;;

7 Statistical Test Values f!l Design

i Test factor F value ,, Pr F ,

Test for A 7.01 0.0132 "

interaction B 1.67 0.2062 d A*B 0.85 0.3637 c l Test for b I A 6.92 0.0135 ,d main effects f ,. B 2.68 0.1125 p,,i

• A1, A2, A3, B1, 82, 83, and 84 are defined in Table A.2. l Note: Frequency of plant-centered loss-of-offsite power events was 0.056 [

per site year. p ,

!j

,I il I

j .

, t i

! i

'! i

.,.4 i

i i

-I i i

i NUREG-1032 A-9 i i

i j 4 -

)

!-l's L!

able A.2 Definitions of offsite power system design factors lL i Majur design factor Design features d U l j h Independence of offsite power 1. All offsite power sources are- .)

o b, sources to the nuclear power connected to the plant through d f plant one switchyard. f h

, All'offsite power sources are connected to the 71 ant through two or more switc1 yards, and the switchyards are electrically connected.

l

3. All offsite power sources are i connected to the plant through  !

j < two or more switchyards'or

( j separate incoming transmission lines, but at least one of the AC sources is electrically

,,. independent of the others.

B. Automatic and manual transfer 1. If the normal source of AC power .

schemes for the Class 1E buses fails, there are no automatic  !

when the. normal source of AC transfers and one or more manual power fails and when the backup transfers to preferred or alter- i sources of offsite power fail nate offsite power sources.

2. If the normal source of AC power fails, there is one automatic transfer but no manual transfers to prefer-red or alternate off- -

site power sources. [

a. All of the Class IE buses in a unit are connected to i the same preferred power j source after the automatic I transfer of power sources.

b. The Class 1E buses in a l unit are connected to L separate offsite power h sources after the auto-

'l matic transfer of power 1 sources.

After loss of the normal AC power source, there-is one auto-  !

, matic transfer. If this source I ji fails, there may be one or more manual transfers of power q

sources to preferred or alter-

i!.; nate offsite power sources.

.t ,

li l NUREG-1032 A-6 h

'l I

l 1l

'I f!!;

Table A.2 (continued) l

!f igi Major design factor Design features.  !

i.

~

a. All of the Class 1E buses in a unit are connected to f!i one preferred power source '

, after the first automatic -

[$ '

I transfer, l i

b. The Class 1E buses in a Gq unit are connected to sepa- lqti rate offsite power sources i lj

_after the first automatic kl, transfer. '

f p

4. If the normal source of AC power q fails, there is an automatic hl  :

transfer to a preferred source i.

of power. If this preferred i

[f' l

- source of power fails, there is an automatic transfer to another n ,,

ej .

ll  :

source of offsite power. .

)

a. All of the Class 1E buses $h in a unit are connected to I

(

the same preferred power l! .

source after the first l

'l automatic transfer. '

b. The Class 1E buses in a unit are connected to sepa- g q

rate offsite power sources g j after the first automatic L  !

transfer of power sources. !l l

-4 ,

1 lj':

l t

I Y

s .$ i p  !

l

(

[ I i

t

1

'f I,

NUREG-1032 A-7  ! I y

( h ..

1 Table A.8 Severe-weather-induced loss-of-of tYite power frequency / recovery N

Severe-weather-induced loss-of-offsite power frequency (S) f,i Frequency of severe weather-induced [

Frequency group (S) loss of offsite power  !!

Less than 1 per 350 site-years -

(0.002/ site-year)

S2 > 1 per 350 site years and

< 1 per 120 site years (0.005/ site year) '

l-S3 Greater than or equal to 1 per 120 site years (0.015/ site years)

Recovery (R)

Recovery from severe-weather-induced '

loss-of-offsite power groups (R) Recovery capability  :

l

\ l Plant has capabi_lity and procedures to j  ; '

recover offsite (non-emergency) AC h I power to the site within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> l j following a severe-weather-induced '

loss of offsite power. '

1e R2 All other plants not in R1 Severe-weather-induced loss-of-offsite power frequency / recovery (SR)  ;

i Severe-weather-induced loss-of- 1 offsite power frequency / recovery i group (SR) Frequency group (S) Recovery group (R) '

SR1 S1 R1 , ,

2 S2 R1 .

l SR3 S3 R1 SR4

}

51 R2 i SR5 S2 R2  ! l SR6 S3 R2 l

l I

I i

NUREG-1032 A-31 i l

l _

-y . ,

{'

~

,t i

f I

Table A.9 Extremelysevere-weaJher induced loss-of-offsite power j frequency

+- ,

Extremely severe-weather-induced i' Frequency groups (SS) loss-of-offsite power frequency 551 _Less than 1 per 3500 site years ,

(0.0002/ site year) 552 > 1 per 3500 site years and

< 1 per 1200 site years

-(0.0005/ site year) 553 > 1 per 1200 site years and

< 1 per 350 site years (0.0015/ site year) 554 > 1 per 350 site years and

< 1 per 120 site years  ;

(0.0,05/ site year) '

555 Greater than or equal to 1 per 120 site years (0.015/ site year)

~

l J

i l

l I

l l

l I

1 l

l NUREG-1032 A-33 l

(