ML20141N598

Forwards Request for Addl Info Re Snupps 860108 & 0211 Submittals of Steam Generator Tube Rupture Analysis.Info Requested Re Max Overfill Case,Atmospheric Relief Valves & Fission Products Released to Intact Steam Generators

ML20141N598
Person / Time
Site:	Wolf Creek
Issue date:	03/13/1986
From:	Youngblood B Office of Nuclear Reactor Regulation
To:	Koester G KANSAS GAS & ELECTRIC CO.
References
NUDOCS 8603170376
Download: ML20141N598 (4)

Text

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Docket No.: 50-482

;' . MAR 131986

?-

'Mr. Glen L. Koester

's Vice President - Nuclear Kansas. Gas & Electric Corpany 200 North Market Street-Post Office Box 208

. Wichita, Kansas 67201

Subject:

Review of Steam Generator Tube Rupture Analysis The NRC staff is continuing its review of your steam generator tube rupture analysis !ubmitted by SNUPPS by letter dated January 8,1986 and February 11, 1986. The information requested in the enclosure is necessary to permit the staff to complete its reveiw.

Please provide the requested information within 15 days of your receipt of i this letter. If all of the requested infonnation request cannot be provided within the requested time provide a schedule for the timely submittal of all remaining items.

Sincerely,

\$\

B. J. Youngblood, Director PWR Project Directorate f4 Division of PWR Licensing-A

Enclosure:

As stated 1 ,'I]

NRC PDR Local PDR PRC System NSIC PWR#4 Rdg N)uncan BJYoungblood OELD ACRS (10)

JPartlow BGrimes, EJordan 9p(. 3 s 8 los PWR#4/DPWR-A PWR#4/DPNR-A P0'Connor/mac BJYoungblood 03/gs /86 03/p /86 0603170376 960313 PDR ADOCK 05000402 p PDR

_ _ _ _ _ - _ _ _ - - _ _ _ - m.- -_--+1 -

p Mr. Glenn L. Koester Wolf Creek Generating Station s Kansas Gas and Electric Company Unit No. I cc:

Mr. Nicholas A. Petrick C. Edward Peterson, Esq.

Execative Director, SNUPPS Legal Division 5 Choke Cherry Road Kansas Corporation Conenission Rockville, Maryland 20850 State Office Building, Fourth Floor Topeka, Kansas 66612 Jay Silberg, Esq.

Shaw, Pittman, Potts & Trowbridge Regional Administrator, Region IV 1800 M Street, NW U.S. Nuclear Regulatory Comission Washington, D.C. 20036 Office of Executive Director for Operations Mr. Donald T. McPhee 611 Ryan Plaza Drive, Suite 1000 Vice President - Production Arlington, Texas 76011 Kansas City Power & Light Company 1330 Baltinore Avenue Mr. Allan Mee Kansas City, Missouri 64141 Project Coordinator Kansas Electric Power Cooperative,Inc.

Chris R. Rogers, P.E. P. O. Box 4877 Manager Electric Department Gage Center Station Public Service Comission Topeka, Kansas 66604

' P. O. Box 360 Jefferson City, Missouri 65102 Resident Inspector / Wolf Creek NPS c/o U.S. Nuclear Regulatory Comission Regional Administrator, Region Ill P. O. Box 311 U.S. Nuclear Regulatory Comission Burlington, Kansas 66893 799 Roosevelt Road Glen Ellyn, Illinois 60137 Mr. Robert M. Fillmore State Coporation Comission Brian P. Cassidy, Regional Counsel State of Kansas Federal Emergency Management Agency Fourth Floor, State Office Building Region I Topeka, Kansas 66612 J. W. McCormack POCH Boston, Massachusetts 02109 Senior Resident Inspector / Wolf Creek NPS c/o U.S. Nuclear Regulatory Comission Terri Sculley, Director P. O. Box 311 Special Projects Division Burlington, Kansas 66839 Kansas Corporation Comission State Office Building, Fourth Floor Topeka, Kansas 66612 Mr. Gerald Allen Public Health Physicist Bureau of Air Quality & Radietion Control Division of Environment ,

Kansas Department of Health

' and Environment ~

Forbes Field Building 321 Topeka, Kansas 66620

l ENCLOSURE I t

REQUEST FOR ADDITIONAL INFORMATION STEAM GENERATOR TUBE RUPTURE (SGTR) ANALYSIS CALLAWAY & WOLF CREEK PLANTS

1. The SNUPPS analysis for the SGTR maximum overfill case states that at the time of break flow termination, the steam volume below the outlet nozzle is very small. Thus, the margin to overfill for this case is minimal, and a slight change in assumptions or calculational results could result in overfill. As an example, the SNUPPS analysis apparently assumes reactor trip at 100% power. This assumption may not be the most conservative from a standpoint of margin to overfill and is also probably not realistic when compared to the Ginna SGTR event. A more realistic scenario may involve turbine runback to some lower power followed by overtemperature delta T trip. At lower power levels the steam generator should have a larger liquid inventory because of reduced void fraction, assuming the SG level remains constant. Thus, starting maximum auxiliary feedwater flow at a lower power level may result in more rapid overfill. Discuss whether this scenario (i.e., lower void fraction) was considered in your analysis and what effect it would have on the margin to overfill.

2. Explain the basis for ':he large difference for reactor trip time between the " failed open AFW control valve" case and the " stuck open ARV" case

! and the effect of these assumptions en the analysis results.

3. The " stuck open ARV" case assumes that the atmospheric relief valve (ARV) is isolated in 20 minutes by manually closing the ARV block valve. State how this time period was estah'ished and whether it is realistic considering that this operation wou s be performed in a ,

location subject to adverse conditions including high temperature, radiation and noise.

. o C

l 2

4. Appendix E " Bases for ARV Technical Specification" states: "An ARV is considered operable if the block valve is closed solely because of leakage". The SGTR analysis assumes that the operator initiates RCS cooldown in less than 30 minutes by opening the intact SG ARVs. Since the operator may have to open the ARY block valves manually if the above Technical Specification is implemented, demonstra'.e that this can be accomplished within the stated time frame considering the concerns regarding this operation expressed in Question 3.

5. In your analysis, you assumed that the fission products released to the intact steam generators were not released to the environment. Provide an analysis demonstrating that the fission products released to the intact steam generators will be retained in the steam generators during the cool

- down phase.

5 O

i

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