ML20151P433

Requests Addl Info Re Steam Generator Tube Rupture Event Methodology within 120 Days of Ltr Receipt

ML20151P433
Person / Time
Site:	Fort Calhoun
Issue date:	08/03/1988
From:	Milano P Office of Nuclear Reactor Regulation
To:	Morris K OMAHA PUBLIC POWER DISTRICT
References
NUDOCS 8808100031
Download: ML20151P433 (6)

Text

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August 3, 1988 Docket No. 50-285 Mr. Kenneth J. Morris Division Manager - Nuclear Operations Omaha Public Power District 1623 Harney Street Omaha, Nebraska 6910P.-2247

SUBJECT:

REQUEST FOR SUPPLia. ENTAL INF0PPATION ON STEAM GENERATOR TUBE RUPTURE EVENT METHODOLOGY

References:

(1) OPPD letter, LIC-87-598, R. Andrews to NRC, dated September 30, 1987 (2) OPPD Nuclear Analysis, Reload Core Analysis Methodology, Transient and Accident Methods and Verification, OPPD-NA-8303-P, Revision 1, dated November 1986 Omaha Public Power District (0 PPD) requested, in reference (1), the review and approval of the proposed Steam Generator Tube Rupture Event Methodology for incorporation into reference (2). The NRC technical staff has reviewed this submittal and found that further information is necessary. This was addressed in a telephone conversation with Mr. K. Holthaus and other members of your staff on July 28, 1988. In this conver::ation it was stated that information addressing the steam generator overfill scenario and the worst-case offsite dose scenario would be required. The infonnation needed is detailed in the enclosure to this letter. Additionally, it was determined that the inclusion of this event methodology into reference (2) was not necessary for the Cycle 12 reload analysis submittal.

Therefore, OPPD is required to submit, within 120 days of the receipt of this letter, a response which addresses the items in the enclosure.

The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under Public Law 96-511.

Sincerely,

/s/

Patrick D. Milano, Project Manager n Project Directorate IV co u Division of Reactor Projects - III,

$$@ IV, V and Special Projects

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Enclosure:

om As stated C0 0 cc w/ enclosure:

g'd See next page oo DISTRIBUTION 8@  : Docket File; ' '

NRC PDR Local PDR D. Crutchfield g ~L.'Rubenstein J. Calvo A. Thadani W. Hodges occ J. Callan, RIV P. Milano A. Gilbert T. Westerman, RIV

@@a P. Noonan 0GC-Rockville E. Jordan B. Grimes ACRS(10) Plant File D4 Reading \

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. . August 3,1980 '

1 Docket No. 50-285 Mr. Kenneth J. Morris Division Manager - Nuclear Operations Omaha Public Power District 1623 Harney Street Omaha, Nebraska 69102-2247

SUBJECT:

REQUEST FOR SUPPLEMENTAL INFORMATION ON STEAM GENERATOR TUBE RUPTURE EVENT METHODOLOGY

References:

(1) OPPD letter, LIC-87-598, R. Andrews to NRC, dated September 30, 1987 (2) OPPD Nuclear Analysis, Reload Core Analysis Methodology, Transient and Accident Methods and Verification, OPPD-NA-8303-P, Revision 1, dated November 1986 Omaha Public Power District (0PPL' ' quested, in reference (1), the review and approval of the proposed Steam Ger - tor Tube Rupture Event Methodology for incorporation into reference (2). The NRC technical staff has reviewed this submittal and found that further infometion is necessary. This was addressed in a telephone conversation with Mr. K. Holthaus and other members of your staff on July 28, 1988. In this conversation it was stated that infomation addressing the steam generator overfill scenario and the worst-case offsite dose scenario would be required. The infomation needed is detailed in the enclosure to this letter. Additionally, it was detemined that the inclusion of this event methodology into reference (2) was not necessary for the Cycle 12 reload analysis submittal.

Therefore, OPPD is required to submit, within 120 days of the receipt of this letter, a response which addresses the items in the enclosure.

The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under Public Law 96-511.

Sincerely,

/s/

Patrick D. Milano, Project Manager Project Directorate IV Division of Reactor Projects - III, IV, V and Special Projects  ;

Enclosure:

i As stated cc w/ enclosure:

See next page DISTRIBUTION Docket File NRC PDR Local PDR D. Crutchfield L. Rubenstein J. Calvo A. Thadani W. Hodges l J. Callan, RIV P. Milano A. Gilbert T. Westerman, RIV l P. Noonan 0GC-Rockville E. Jordan B. Grimes ACRS(10) Plant File D4 Reading LA:PDQ PM:PD4 S D:PD4 4 l PNoonad PMilano ores JCalvo  !

08/l/88 08/( /88 08//g88 08/ 3 /88 l

.Tj# * *%,),"' UNITED STATES NUCLEAR REGULATORY COMMISSION 4 C E WASHING TON, D. C. 20655

'I August 3, 1988

%,.....j Docket No. 50-285 Mr. Kenneth J. Morris Division Manager - Nuclear Operations Omaha Public Power District 1623 Harney Street Omaha, Nebraska 69102-2247

SUBJECT:

REQUEST FOR SUPPLEMENTAL INFORMATION ON STEAM GENERATOR TUBE RUPTURE EVENT METHODOLOGY

References:

(1) OPPD letter, LIC-87-598, R. Andrews to NRC, dated September 30, 1987 (2) OPPD Nuclear Analysis, Reload Core Analysis Methodology, Transient and Accident Methods and Verification, OPPD-NA-8303-P, Revision 1, dated November 1986 OmahaPublicPowerDistrict(0 PPD) requested,inreference(1),thereviewand approval of the proposed Steam Generator Tube Rupture Event Methodology for incorporation into reference (2). The NRC technical staff has reviewed this submittal and found that further infonration is necessary. This was addressed in a telephone conversation with Mr. K. Holthaus and other members of your staff on July 28, 1988. In this conversation it was stated that infonnation addressing the steam generator overfill scenario and the worst-case offsite dose scenario would be required. The information needed is detailed in the enclosure to this letter. Additionally, it was determined that the inclusion of this event methodology into reference (2) was not necessary for the Cycle 12 reload analysis submittal.

Therefore, OPPD is required to submit, within 120 days of the receipt of this letter, a response which addresses the items in the enclosure.

The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under Public Law 96-511.

Sincerely, N%fL -

Patrick D. Milano, Project Manager Project Directorate IV Division of Reactor Projects - III, IV, V and Special Projects

Enclosure:

As stated cc w/ enclosure:

See next page

. Mr. Kenneth J. Morris Fort Calhoun Station Omaha Public Power District Unit No. 1 cc:

Harry H. Voigt, Esq.

LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, NW Washington, D.C. 20036 Mr. Jack Jansen, Chairman Washington County Board of Supervisors Blair, Nebraska 68008 Mr. Phillip Harrell, Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 309 Fort Calhoun, Nebraska 68023 Mr. Charles B. Brinkman, Manager Washington Nuclear Operations C-E Power Systems 7910 Woodmont Avenue Bethesda, Maryland 20814 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Harold Borchert, Director Division of Radiological Health Department of Health 301 Centennial Mall, South P.O. Box 95007 Lincoln, Nebraska 68509 W. G. Gates, Manager Fort Calhoun Station P. O. Box 399 Fort Calhoun, Nebraska 68023

l l ISSUES REQUIRING RESOLUTION STEAM GENERATOR TUBE RUPTURE EVENT METHODOLOGY l

FORT CALHOUN STATION (CE)

The at:ident analysis methodology for the Steam Generator Tube Rupture (SGTR)

Event proposed for incorporMion into the Updated Safety Analysis Report I (USAR), Chapter 14, Section 14, has been determined to be inadequate. Two

! scenarios exist that could result in worst-case consequences: the steam l generator overfill scenario and the worst-case offsite dose scenario. These scenarios should be addressed in a revised submittal. The following criteria should specifically be considered when evaluating both aspects for a design basis SGTR event.

A. Assumptions

1. Assume a most limiting single active failure for the steam generator overfill scenario (i.e. possibly the auxiliary feedwater flow control valve failere full open or the steam generator atmos heric dump valve (ADV) failure closed, associated with the faulted SG .

2. Assume a most limiting single active failure for the worst-caso radiological offsite dose (possibly a stuck open ADV associated with the faulted SG).

a. Determine whether the worst case for dose is overfill with maximized break flow, or an uncovered tube break with minimized break flow as in the North Anna event.

b. Determine whether the loss of offsite power would be more conservative at the onset of the SGTR event or at the time of reactor trip.

3. Assume a maximum Technical Specification leakage in the unaffected steam generator (1 gpm).

4. Assume fuel failure for any rod with MDNBR below 1.19 as a result of the SGTR event.

B. Infomation Needed to Present the Results of the SGTR Analysis

1. Sequence of events on a time scale, from the onset of the tube rupture to the pressure equalization between the primary and secondary coolant systems.

2. Operator acticr times for identifying the faulted SG, isolation of the faulted SG, initiation of cooldown, depressurization, etc. This should clso be presented on a time scale.

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3. Discu'ision of the issue of a potential SG overfill, including the integrity of the steam lines under a water-filled condition and, if applicable,the effects of water flow through the safety valves, since these were not designed for this service.

4. Discussion of the basis for the operator action times utilized (i.e.

simulatorruns,etc).

5. Major transient curves including RCS pressure, secondary system prassure, DNBR, SG water level, leak rates for the faulted and intact SGs, etc.

6. Amount of fuel failure based on DNBR.

7. Calculated radiological consequences as corrpared to the limits set forth in 10 CFR Part 100 (2 hours and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />), including pre-accident and coincident fodine spikin based on tech spec limits (usually 100/E)g and noble gas inventory

8. Radiological parameters and curves including mass flow loss with respect to time, flashed fractions, and partition and decontamination factors in accordance with SRP 15.6.3.

9. Discussion of the consequences of a postulated break at the top of the tube bundle, as it had occurred at North Anna.

Contact:

A. Gilbert l

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