ML20138A925

Submits Response to Request for Addl Info Re Generator Tube Rupture Evaluation

ML20138A925
Person / Time
Site:	Catawba
Issue date:	04/22/1997
From:	Mccollum W DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M98107, TAC-M98108, NUDOCS 9704280285
Download: ML20138A925 (4)

Text

. . _ .. __ . . _ _ . - . . _ . _ - . _ . . _ __ m_.=__ _ . , _.

.ll .

l l

• DukeIbwerCompany Wtustst R McCow v. Jx A

• Catau ba Nudear C+neration Department Vicehesident 4800 ConcordRoad (803)K11-3310 Offue York, SCTJT45 (8031831J126 fax

) DUKEPOWER April 22, 1997 g

( i U..S. Nuclear' Regulatory Commission Attention: Document Control Desk

[

I Washington, D.C. 20555

Subject:

Catawba Nuclear Station, Units 1 and 2 Dockets Nos. 50-413 and 50-414 Request for Additional Information Regarding the l Operating License Amendment for the Steam l

! Generator Tube Rupture Evaluation i (TAC Nos. M98107, M98108) )

Per phone conference on April 21, 1997, the NRC requested -

clarification of details regarding the additional l information submitted by Duke Power to the NRC in letters i

dated April 2, 1997, April 10, 1997 and April 16,1997 for the Steam Generator PORV Technical Specification Amendment.

The proposed amendment revises Section 3/4.7.1.6 of the Technical Specifications to require four instead of three steam generator power operated relief valves (PORVs) and Section 15.6.3 of the Updated Final Safety Analysis Report '

(UFSAR) to require four instead of three PORVs and allow credit for local manual operation of the PORVs. The additional information requested is provided in the  !

enclosure and should supply the clarification necessary to complete the amendment request.

We request that you review the additional information on a schedule consistent with the urgency of the original request. If you need additional clarification of the response to the questions or have additional questions please contact Martha Purser at (803)-831-4015.

f

' Sincerely,

) William R. McCollum, J .

9704290285 970422 PDR ADOCK 05000413 O l P PDR _ l, l{ . illul1l, u., .

' 11"'l,u n ,

l 1

naen ,mw w l

i

I' l

Document Control Desk Page 2 .

April 22,1997 l

i xc (with attachments):

L.A. Reyes, Regional Administrator, Region II P.S. Tam, Senior Project Manager, ONRR I

l R.J. Freudenberger, Senior Resident Inspector, CNS i

t i

s 0

~.

l Clarification of Request for Additional Information for l Catawba Uniti and Unit 2 Regarding Proposed Amendments for the Steam Generator PORVs l

Reference:

Phone Conference with the state NRR branch on April 21, 1997 and April 14, 1997 regarding letters from Duke Power to NRC dated April 2, 1997, April 10, 1997 and April 16, 1997.

Request: Please provide time dependent data and assumptions for Steam Generator Tube Rupture Dose Analysis.

Response

The radiation doses of a steam generator (S/G) tube rupture (SGTR) . Ath a failure which leaves a power operated relief valve on only one intact S/G available for operation from within the control room has been analyzed. The time dependent thermal hydraulic input used to calculate radiation doses following this SGTR are presented in Table 1.

. JR ..

s

.pTable.1 , .

3

• / m, '

e n ,

, ?SGTR Thermal-Hydraulic Input.Dataj Time Span Break Flow Flash Steaming Rate (Min) (1bm/ min) Fraction (lbm/ min) 0 - 20.15 3600 0.18023 64110 20.15 - 3558 0.26361 20625 22.5 22.5 - 25.5 3312 0.26361 16287 25.5 - 28.5 3222 0.24683 10397 28.5 - 31.5 3162 0.11714 9108 31.5 - 34.5 3162 0.11552 8450 34.5 - 37.5 3138 0.11333 7987 37.5 - 3138 0.11111 7788 38.138 l 38.138 - 3138 0.11111 7324 j 39.5 39.5 - 41.5 3090 0.11091 7324 41.5 - 3090 0.10864 7151 43.5 43.5 - 45.5 3060 0.10884 6993 45.5 - 47.5 3102 0.10663 4042 47.5 - 49.5 3090 -0 10584 4042 h

, . _. _ . . . _ ~ ~ . . _ . . _ . _ . . _ . . _ _ . _ _ . . . _ _ _ _ _ _ . . _ . . . _ _ . _ . . _ . . _ _ _ _ . _ . . . - . . _ _ _ _ . _

I i . .

/ 44 .. Table,1, Continuedt

< ~

s. .

. .y: ,

Time Span Break Flow Flash Steaming Rate (Min) (lbm/ min) Fraction (1bm/ min) 49.5 - 51.5 3054 0.10544 3915 51.5 - 53.5 3054 0.10425 3915 53.5 - 55.5 3024 0.10325 3805  ;

55.5 - 57.5 3024 0.10226 2949 57.5 - 59.5 2916 0.09729 2885 59.5 - 61.5 2892 0.06672 2885 61.5 - 63.5 2790 0.05581 2830 63.5 - 65.5 2772 0.04753 2830 65.5 - 66.5 2730 0.04329 2781 66.5 - 67.5 2682 0.03913 2781 67.5 - 2628 0.03477 2781 68.833 68.833 - 70 2562 0.03477 2737 70 - 72 2562 0.03021 2737 72 - 74 2178 0.01959 2737 74 - 76 2280 0.01629 2737 76 - 76.785 2250 i 0.02016 2675 76.785 - 80 2232- '

1 0.02016 2675 80 - 82 2286 0.01067 2675 82 - 84 2058 0.00892 2626 84 - 86 792 0.00137 2626 86 - 88 1746 0 2626 88 - 90 1830 0 2584 90 - 92 1914 0 2584 92 - 94 1806 0 2584 94 - 96 1542 0 2546 1 96 - 98 1398 0 2546 98 - 100 1104 0 2546 100 - 102 966 0 2546 102 - 104 684 0 2502 104 - 106 528 0 2502 l 106 - 355 0 0 392.8 j l

Additional information is s.ven as follows: Letdown flow of 76 GPM is assumed for the first 20 minutes of the transient. I The coolant inventory is put at 481,637 lbm. Integrated ,

break flow is 285,000 lbm. The releases of steam from the l ruptured S/G was found to be 1,281,000 lbm before trip and J L 491,500 lbm after trip. The figures presented here for j integrated break flow and steam releases from the ruptured S/G supersede the corresponding information enclosed with the i letter from W.R. McCollum to the USNRC, April 2, 1997.

!~

a I

d l

{.

i - .- .. -.

.- _ ._ . _ _ , _ , __