ML19209B442

Requests Addl Info Re Possibility of Main Steam Line Break Inside Containment as Result of Overpressurization of Containment at Sites

ML19209B442
Person / Time
Site:	McGuire, Mcguire
Issue date:	09/21/1979
From:	Baer R Office of Nuclear Reactor Regulation
To:	Parker W DUKE POWER CO.
References
NUDOCS 7910090740
Download: ML19209B442 (6)

Text

y* ) yn m

m o

• UNITED STATES f

g

, y r>

g NUCLEAR REGULATORY COMMISSION

.g p

(

.: y WASHINGTON, D. C. 20555 SEP 211979 Docket Nos. 50-369 and 50-370 Mr. William 0. Parker, Jr.

Vice President, Power Production Duke Power Company P. O. Box 2178 422 South Church Street Charlotte, North Carolina 28242

SUBJECT:

P0TENTIAL OVERPRESSURIZATION OF CONTAINMENT IN THE EVENT OF MAIN STEAM LINE BREAK (McGuire Nuclear Station, Units 1 and 2)

In a letter dated September 10, 1979, we were informed by the Virginia Electric and Power Company that overpressurization of the containment at North Anna 3 and 4 could occur as a result of a main steam line break inside containment.

This overpressurization resulted when auxiliary feedwater flow was included in the analysis. We are currently assessing the generic implications of this matter and request that you provide us with certain information (see Enclosure) regarding the McGuire design.

We request that this information be provided by October 22, 1979.

Please contact us if you have any questions regarding this matter.

Sincerely, l fW Robert L. Baer, Chief Light Water Reactors Branch No. 2 Division of Project Management

Enclosure:

Request for Additional Information ces w/ enclosure:

See next pages 1117 089 3910090 790 g

Mr. William 0. Parker, Jr.

Vice President, Steam Production Duke Power Company P. O. Box 2178 422 South Church Street Charlotte, North Carolina 28242 cc: Mr. W. L. Porter Duke Power Company P. O. Box 2178 422 South Church Street Charlotte, North Carolina 28242 Mr. R. S. Howard Power Systems Division Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. E. J. Keith EDS Nuclear Incorporated 220 Montgomery Street San Francisco, California 94104 Mr. J. E. Houghtaling NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 Mr. Jesse L. Riley, President The Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28207 J. Michael McGarry, III, Esq.

Debevoise & Liberman 1200 Seventeenth Street, N. W.

Washington, D. C.

20036 Robert M. Lazo, Esq., Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Connission Washington, D. C.

20555 Dr. Cadet H. Hand, Jr., Director Bodega Marine Lab of California P. O. Box 247 Bodega Bay, California 94923

.11i7 090

Mr. William 0. Parker, Jr.

cc: Anthony Z. Roisman, Esq.

Natural Resources Defense Council 917 - 15th Street, N. W.

Washington, D. C.

20555 Richard P. Wilson, Esq.

Assistant Attorney General State of South Carolina 2600 Bull Street Columbia, South Carolina 29201 1117 091

SEP 211979 Enclosure Request for Additional Informt. tion McGaire Nuclear Station, Units 1 and 2 In a letter dated September 10, 1979, the NRC was informed by Virginia Electric and Power Company that overpressurization of the containment at North Anna 3 and 4 could occur as a result of a main steam line break inside containment. This overpressurization resulted when auxiliary feedwater flow was included in the analysis.

NRC is currently assessing the generic implications of this letter.

To assist us in determining if a similar circumstance could occur at your facility, you should take the following actions.

1) Rev ew your original analysis of this event, and provide NRC with the i

assumptions used during this analysis.

Particular emphasis should be placed on describing how auxiliary feedwater flow (AFF) was accounted for in your original analysis.

(Reference to previously submitted information is acceptable if identified as to page number and date.) Any changes in your design which would impact the conclusions of your original analysis should be discussed. We are particularly concerned with design changes that could lead to an underestimation of the containment pressure following a MSLB inside containment.

2) opecifically, provide ti..

3110 wing information for the analyses performed to determine the maximum containment pressure for a spectrum of postulated main steam line breaks for various reactor power levels:

a.

Specify the auxiliary feedwater flow rate that was used in your original containment pressurization analyses.

Provide the basis for this assumed flow rate.

1117 092

v.

_ stP 211373 b.

Provide the auxiliary feedwater rated flow rate, the run out flow rate, and the pump nead capacity curve of your current design.

c.

Provide schematic drawings to show the auxiliary feedwater system arrangement in your current design.

.d.

Provide the time span over which it was assumed in your original analysis that AFF was added to the affected steam generator folicwing a MSLB inside containment.

e.

Discuss the design provisions in the auxiliary feedwater system used to terminate the auxiliary feedwater flow to tha affected steam generator.

If operator accion is required to perform this function, discuss the information that will be available to the operator to

-alert him of the need to isolate the auxiliary feedwater to the affected steam generator, the time when this information would beceme available, and the time it would take the operator to complete this action.

If termination of auxiliary feedwater flow is dependent on automatic action, describe the basic operation of the auto-isolation system. Describe the failure modes of the system.

Describe any annunciation devices associated with the system.

f.

Provide the single active failure analyses which specifically identifies those safety grade systems and components relied upon to limit the mass and energy release and the containment pressure response.

1117 093

., ~

SEP 21(979 The single failure analysis should include, but not necessarily be limited to: partial loss of containment cooling systems and failure of the auxiliary feedwater isolation valve to close, g.

For the single active failure case which results in the maximum containment atmosphere pressure, provide a chronology of events.

Graphically, show the containment atmosphere pressure as a function of time for at least 30 minutes following the accident.

For this case, assume the auxiliary feedwater flow to the broken loop steam generator to be at the pump run out flow (if a run out control system is not part of the current design) for the entire transient if no automatic isolation to auxiliary feedwater is part of the current design.

h.

For the case identified in (g) above, provide the mass and energy release data in tabular form. Discuss and justify the assumptions made regarding the time at which active containment heat removal systems become effective.

1117 094