ML20205M890

Forwards Third Request for Addl Info Re Util 861230,870115 & 0217 Submittals Concerning Analysis of Firewater Cooldown from 82% of Full Power.Major Concerns Re Effects of Transient Loading Due to Seismic Motion or Flow

ML20205M890
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/30/1987
From:	Heitner K Office of Nuclear Reactor Regulation
To:	Robert Williams PUBLIC SERVICE CO. OF COLORADO
References
TAC-63576, NUDOCS 8704020521
Download: ML20205M890 (5)

Text

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March 30, 1987 Docket No. 50-267 Mr. R. O. Williams, Jr.

Vice President, Nuclear Operations Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201-0840

SUBJECT:

ANALYSIS OF FIREWATER C00LDOWN FROM 82% POWER OPERATION -

THIRD REQUEST FOR ADDITIONAL INFORMATION We are reviewing your submittals dated December 30, 1986, January 15 and February 17,1987 (P-86682, P-87002 and P-87055) concerning your analysis of a firewater cooldown from 82% of full power. In order to complete this review, we are making a third request that you provide certain additional information.

Our request for this infonnation is enclosed.

We have performed a preliminary review of your submittals regarding the structural and mechanical effects on the steam generator during the firewater cooldown. Our major concerns pertain to the effects of transient loading due to seismic motion or flow induced vibration which appear to have been excluded without proper justification, and to the material characteristics of Sanicro

31. Our position is that loading due to the safe shutdown earthquake (SSE) nust be considered if the initiating event is any event other than an SSE.

The information requested in this letter affect fewer than 10 respondents; therefore, OMB clearance is not required under P.L.96-511.

Sincerely, original signed by Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B -

Office of Nuclear Reactor Regulation

Enclosure:

DISTRIBUTION:

As stated Docket J11e/ JPartlow NRC PDR HThompson cc w/ enclosure: Local PDR ACRS (10)

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March 30, 1987 Docket No. 50-267 Mr. R. O. Williams, Jr.

Vice President Nuclear Operations ,

Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201-0840

SUBJECT:

ANALYSIS OF FIREWATER C00LDOWN FROM 82% POWER OPERATION -

THIRD REQUEST FOR ADDITIONAL INFORMATION We are reviewing your submittals dated December 30, 1986, January 15 and February 17, 1987 (P-86682, P-87002 and P-87055) concerning your analysis of a firewater cooldown from 82% of full power. In order to complete this review, we are making a third request that you provide certain additional information.

Our request for this infomation is enclosed.

We have performed a preliminary review of your submittals regarding the structural and mechanical effects on the steam generator during the firewater cooldown. Our major concerns pertain to the effects of transient loading due to seismic motion or flow induced vibration which appear to have been excluded without proper justification, and to the material characteristics of Sanicro

31. Our position is that loading due to the safe shutdown earthquake (SSE) must be considered if the initiating event is any event other than an SSE.

The information requested in this letter affect fewer than 10 respondents; therefore, OMB clearance is not required under P.L.96-511.

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Sincerely,

l. C.bu Kenneth L. Heitner, Project Manager I Standardization and Special Projects Directorate Division of PWR Licensing-B Office of Nuclear Reactor Regulation

Enclosure:

As stated l cc w/ enclosure:

See next page

g Mr. R. O. Williams Public Service Company of Colorado Fort St. Vrain cc:

Mr. D. W. Warembourg, Manager Albert J. Hazle, Director Nuclear Engineering Division Radiation Control Division Public Service Company Department of Health of Colorado 4210 East Ilth Avenue P. O. Bcx 840 Denver, Colorado 80220 Denver, Colorado 80201 Mr. David Alberstein, 14/159A Mr. R. 0. Williams, Acting Manager GA Technologies, Inc.

Nuclear Production Division Post Office Box 85608 Public Service Company of Colorado San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L. Brey, Manager

• Nuclear Licensing and Fuel Division Mr. P. F. Tomlinson, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Public Service Company of Colorado Denver, Colorado 80201 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Senior Pesident Inspector U.S. huclear Regulatory Commission Mr. R. F. Walker P. 0. Box 840 Public Service Company of Colorado Platteville, Colorado 80651 Post Office Box 840 Denver, Colorado 80201-0840 Kelley, Stansfield & 0'Donnell Public Service Company Building Comitment Control Program Room 900 Coordinator 550 15th Street Public Service Company of Colorado Denver, Colorado 80202 2420 W. 26th Ave. Suite 100-D Denver, Colorado 80211 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive Suite 1000 Arlington, Texas 76'J11 Chairman, Board of County Comissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1 Denver Place 99918th Street, Suite 1300 Denver, Colorado 802J2-2413

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-i Enclosure REQUEST FOR ADDITIONAL INFORMATION

1. The stress evaluation regions.in'the steam generator during the single.

1 cycle cooldown from power using firewater in the EES tube bundles did 'rct consider the effects due to flow induced vibration or dynamic loading due to an SSE.

a. Provide justification for not including'the effects due to flow-induced vibration.

b. Provide the basis, method and results of the stress evaluations which include the effects due to SSE loading for the Reheater tubes and their supports, the EES. tubes and the EES tube support

, structure, the superheater helical tube bundle and its support structure, and the superheater downcomer.

a 2. In Attachr:ent 7 of P-86683, it is stated on pg. 3 that during the 90 minute period the tube metal temperature is less than 1300*F, while calculation No. 68-02 of Attachment 9 indicates that the maximum of. all tube temperatures in this transient is 780*F. Provide a reconciliation and basis of these two values, indicating which is the correct value.

, 3. Provide the detailed methodology used for perfonning the creep collapse

analysis of the reheater tube, and the User's Manual for the computer

!. program " BUCKLE". Provide assurance that the time to reach yield stress 4 in the maximum stressed point is shorter than the time at which the ovality becomes unbounded, i.e., show that collapse can not occur at a

, maximum stress which is lower than the yield stress.

I 4. Provide a basis for the statement on p. 16 of Attachment 7 that "the structural integrity of the steam generator is not likely to be i 4

compromised due to excessive tube / plate interaction loads during the EEe l firewater cooldown event," in particular during simultaneous SSE )

loading. I

5. It is stated that Sanicro 31 reheater tube material is a European Alloy 800 - type material, which met the required material specifications of i Alloy 800 Grade 2 at the time when the steam cenerators were built..

Provide the chemical composition of this material and .its strength levels i (i.e., yield and ultimate strengths at appropriate temperatures). In .

addition address the metallurgical treatment of Sanicro 31 in comparison with the treatment of the ASME-Code Alloy 800 material. Further, explain why the Sanicro 31 will not degrade to below the minimum levels allowed for the Code material because of the temperature cycles which the steam generator tubes experience.

6. It is implied that Sanicro 31 is equivalent to an Alloy 800 H material, e.g., SB-163 or SB-407. If this is the case, justify the value of 23,300 psi psi forat SG-163 380*F for andS*20,000 psi for SB-407 at 380*F. Current ASME Code values in I

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7. Figure 4.4 is incomplete as submitted. The temperature profile for the time in hours from 0 to 2 is missing. Please supply this information.

8. The possibility of accumulated creep-fatigue damage contributing to tube failure is not explicitly addressed in this submittal. Please provide an evaluation of this potential failure mechanism in the context of the creep-collapse failure mode that was assumed.

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9. With regard to the Fort St. Vrain (FSV) Technical Specifications, 5.3.11 -

Steam Generator Bimetallic Welds Surveillance and SR 5.3.12 - Steam Generator Tube Leaks Surveillance, describe the results of the surveillance requirements including the number of times and the dates that they have been implemented. Provide any other information that is available concerning degradation of the stesm generator tubes.

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