ML18047A570

PWR Main Steam Line Break W/Continued Feedwater Addition (B-69), Request for Addl Info

ML18047A570
Person / Time
Site:	Palisades
Issue date:	06/29/1982
From:	Herrick R, Vosbury F FRANKLIN INSTITUTE
To:	Peter Hearn NRC
Shared Package
ML18047A569	List: ML18047A570
References
CON-NRC-03-81-130, CON-NRC-3-81-130 IEB-80-04, IEB-80-4, RAI-C5506-135, NUDOCS 8209130162
Download: ML18047A570 (4)

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REQUEST FOR ADDITIONAL INFORMATION PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION <B-69)

CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT NRC DOCKET NO. 50-255 FRC PROJECT C5506

• NRC TAC NO. 46850 .

FRC ASSIG.NM~NT.5 NRC CONTRACT NO. NRC-03-81-130 FRCTASK 135 Prepared by I

Franklin Research Center Author: F. W. Vos bury

• 20th and Race Street ..

Philadelphia, PA 19103 FRC Group Leader: R. c. Herrick Prepared for

.Nuclear Regulatory Commission Washington, 0. C. 20555

• Lead NRC Engineer: P. Hearn June 29, 1982

( 8209130162. 820907 f t n k l i n Resea.rch Center *.

PDR ADOCK" 05000255 ' l 0 PDR '

i..._l. A Division of The Franklin Institute The Benjamin Franklin Pertcway. Phila. Pa. 19103 (215) 448-1000

i* e. ...

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. RAI-C5506-135 BACKGROOND An evaluation of the information contained in the May 9, 1980 [l] and April 26, 1982 [2] letters from Consumers Power Company (CPC) to the U.S.

Nuclear Regulatory Commission (NRC) relating to IE Bulletin 80-04 (3],

• Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition* - .*

revealed an item of concern. Additional info~mation relating to this concern is needed before a final evaluation can be made regarding the potential for exceeding the containment "design pressure or worsening the reactor return-to-power response.

CONCERN

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• As a result.of conducting an accident analysis review pursuant to IE I

Notice 79-22 (4), CPC discovered that during a main steam line break (MSLB) accident, the condensate pumps would continue to feed the ruptured steam generator at an excessive rate which would result in a more severe restart and a higher containment pre~sure than was previously analyzed. CPC

, reported this by" Licensee Event Report (LER)79-041 (5]. This and several other similar events at different plants led to the issuance of IE Bulletin 80-04.

To correct the condition, the Licensee modified the control circuitry of the feedwater regulating valves (CV-0701 and CV-0703) and the f eedwater bypass valves (CV-0734 and cv....0735), so as to close the valves on steam generator low pressure.

In response to e request for additional information [6] asking the Licensee to evaluate the potential for a single active failure in the main feedwater (MFW) system that could cause the greatest feedwater flow to the affected steam generator during a MSLB accident, the Licensee stated:

• since the regulating valves and the bypass valves are downstream of both the condensate pumps and MEW pumps, this modification assures that flow to the ruptured steam generator will be less than that assumed in the safety analysis (2].*

.. ~nklin Re~earch Center A OMoion OI The Fninklin 1nMitute

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RAI-CSS06-l35.

Review of the MFW and condensate systems revealed that a single active failure to one of the feedwater regulating valves or feedwater bypass valves could cause it to remain open after receiving the low steam generator pressure trip signal.

Failure of these valves would still allow the condensate pumps to continue pumping water through the MFW pumps into the ruptured steam generator

[2, 5). The MSLB analysis Ill assumed that:

• Main feedwater flow rampdown from full at time of trip to zero flow at 60 seconds post-reactor trip.*

Since a single active failure ot.~he.feedw~ter regu;atfng valve or feedwater bypass valve* feeding the r~ptured'.. steam generator woulcf'cause a higher MFW flow rate than that assumed in the .MSLB analysis, further 'analysis or corrective action is required to demonstrate ~at Ute containment design

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pressure is not exceeded and that a worsening of the reactor return-to-power does not occur.

REQUEST Please provide the following information:

l. A determination of the MFW flow to the affected steam generator during a MSLB accident, assuming a single activ~*failure to either a feedwater regulating valve or feedwater bypass valve which would cause it. to remain open.

2. An analysis of the containment pressure response for a postulated MSLB to determine if the containment design margin is exceeded using the main feedwater flow rate identified in Item l above.

3. An analysis of the core reactivity response to a MSLB taking *into I account the effects of the MFW flow identified in Item l.

~nklin Research Center

• A DMsion ol The Frenldin lnslitute

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RAI-C5506-135

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REFERENCES \

1. s. R. Frost (CPC) \i . '

Letter to J. G. Keppler (NRC)

Subject:

Response to IE Bulletin 80-04 I

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I May 9, 1980 I.; .

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2. B. D. Johnson (CPC)

Letter to D. M. Crutchfield (NRR, ORB) *< ...

Subject:

Response to IE Bulletin 80-04, Additional Information April 26, 1982

3. *Analysis of a PWR .Ma~n Steam Line Break with Continued F!edwater Addition* .

NRC Office of inspection and Enforcement, February e, 1980 IE Bulletin 80-04

4. *0verpressurization of the Containment of a PWR Plant After a Main Steam Line Break*

.NRC Office of Inspection and En~orcement, Octob~r 1, 1979 IE Information Notice 79-24

s. Licensee Event Report (LER), NRC Docket No. 50-255, LER 79-041 dated Ocotober 30, 1979 Summarized in USNRC LER.Monthly Report of January 3, 1980

6. D~ M. Crutchfield (NRR, ORB) ,

Letter to Consumers Power Company

Subject:

Request for Add~tional Information, IE Bulletin 80-04 March 8, 1982

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~nklin Res~erch Center A DMsion ol The F r...i.Jin lnalitute .I

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