ML20213H105

Forwards Request for Addl Info Re Util Steam Generator Tube Analysis,Per 860108,0211,0401 & 0904 Submittals.Response Requested within 45 Days of Ltr Receipt to Continue Review on Current Schedule

ML20213H105
Person / Time
Site:	Wolf Creek
Issue date:	11/12/1986
From:	Oconnor P Office of Nuclear Reactor Regulation
To:	Koester G KANSAS GAS & ELECTRIC CO.
References
NUDOCS 8611190096
Download: ML20213H105 (6)

Text

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.g-Docket No.: 50-482 NOV 121986 Mr. Glenn L. Koester Vice President - Nuclear Kansas Gas & Electric Corppany 201 florth Market Street Wichita, Kansas 63166

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE SNUPPS STEAM GENERATOR TUBE RUPTURE ANALYSIS The staff is continuing its review of your submittals dated January 8,1986, February 11,1986, April 1,1986, and Septctrber 4,1986. To pennit us to continue our review on our current schedule, we require the infonration re-quested in Enclosure 1 a 2 to this letter be provided.

Please provide the requested information within 45 days of your receipt of this letter.

Sincerely,

\9 Paul W. O'Connor, Project Manager ,

PWR Project Directorate #4 Division of PWR Licensing-A

Enclosures:

As stated cc: See next page DISTRIBUTION:

-Docket File local PDR NRC PDR FWRi4 Reading BJYoungblood Reading P0'Cerror MDuncan ACRS(10 EJordan JParticw BGrimes NThompson PURt4/DPhR-A RS F08 \'O PWR MWR-A P0'Cerror/ rad CBerlinger VBenaroya BJY u blood 11/;2/86 11 /86 11/L/86 11 86 8611190096 861112 PDR ADOCK 05000482 P PDR

f' o,, UNITED STATES 3 o NUCLEAR REGULATORY COMMISSION

7. p WA SHINGTON, D. C. 20555

' ... .. ** NOV 121986 Pocket No.: 50-482 Fr. Glenn L. Koester Vice President - fluclear Kansas Gas & Electric Cortrany 201 f; orth l'arket Street Pichita, Kansas 63166

SUBJECT:

RE00EST FOR ADDITIONAL INFORMATION RELATED TD THE SNUPPS STEAM GENERATOR TUBE RUPTURE ANALYSIS The staff is continuing its review of your submittals dated January 8,1986, February 11,1986, April 1,1986, and September 4,1986. Tc rernit us to continue our review on our current schedule, we require the infonnation re-quested in Enclosure 1 a 2 to this letter be provided.

Please provide the requested information within 45 days of your receipt of this letter.

Sincerely, I

[{J, (,, PSv5LW Paul W. O'Connor, Project fianager PilR Project Directorate #4 Division of PWR Licensing-A

Enclosures:

As stated cc: See next page

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Mr. Glenn L. Koester Wolf Creek Generating Station

Kansas Gas and Electric Company Unit No. I cc

Mr. Nicholas A. Petrick Mr. Gary L. Haden, Director Executive Director, SNUPPS Research & Energy Analysis 5 Choke Cherry Road Kansas Corporation Commission Rockville, Maryland 20850 ith Floor - State Office Building

~lopeka, Kansas 66612-1571 i

Jay Silberg, Esq.

Shaw, Pittman, Potts & Trowbridge Regional Administrator, Region IV 2300 N Street, NW U.S. Nuclear Regulatory Commission Washington, D.C. 20037 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 Conpany 1330 Paltimore Avenue Mr. Allan Mee kansas City, Missouri 64141 Project Coordinator Vansas Electric Power Cooperative,Inc.

Chris R. Rogers, P.E. P. O. Box 4877 Manager, Electric Departn.ent Gage Center Station Public Service Commission Topeka, Vansas 66004 P. O. Box 360 l Jefferson City, Missourf 65102 Resident Inspector / Wolf Creek NPS c/o U.S. Nuclear Regulatory Commission Regional Administrator, Region III P. O. Box 311 i U.S. Nuclear Regulatory Commission Burlington, Kansas 66893

, 799 Poosevelt Road Glen Ellyn, Illinois 60137 Mr. Brian Moline i Chief Legal Counsel Senior Resident Inspector / Wolf Creek Kansas Corporation Commission

, c/o U. S. Nuclear Regulatory Ccenission 4th Floor - State Office Duilding i

P. O. Box 311 Topeka , Kansas 66612-1571 Burlington, Kansas 66839 Mr. Robert Elliot, Chief Engineer Utilities Division Kansas Corporation Conunission 4th Floor - State Office Building Tcpeka, Kansas 66612-1571 Pr. Gerald Allen Public Health Physicist Bureau of Air Quality & Radiatico

Control Division of Environment e

! Kansas Department of Health

and Environnent forbes Field Building 321 Topeka, Kansas 66620 t

ENCLOSURE 1 REGUEST FOR ADDITIONAL INFORFATICU St.1'PPS STEAM GENERATOR TUBE RUPTURE (SGTR) ANALYSIS CALLAUAY I, LOLF CREEK Peferences: (1) Letter frera Nicholas A. Petrick, SNUPPS, to Harold P.

Denton, NRC "Steara Generator Tube Rupture Analysis -

SNUPPS" Septer.ber 4, 1986 (2) SNUPPS Report " Steam Generator Single-Tube Rupture Analysis for SNUPPS Plants" Deceraber 1985

1. In Reference (1), item 1, you indicated an average break flow rate of 35 lbm/sec. for the design basis SGTR, based on the Burnell correlation, which would result in a margin to solid secondary water relief of 271 cubic feet. You also indicated that an increase in average break flow rate greater than 127. of the calculated flow could result in liquid secondary relief. The staff believes that, based on this response, there is sufficient uncertainty ir the break flow calculations, as well as in operator action times (see Enclosure 2) and interval between safety injection (SI) terminatier erd break flow termination (see question 2 below), to warrant assuming that safety valve (SV) liquid relief can occur. Since the SVs ere not specifically designed for liquid or two phase flow, this coulc result in valve malfunction. During the Ginna SGTR event, liquid relief thrcugh a SV prevented proper valve reseating, with consecuent continued secondary leakage.

Therefore please perfonn an analysis which assumes a design basis SGTP with loss of offsite power (LOOP), closure of the ruptured SG tcain steari isolation valve (MSIV), and SG overfill resulting in liquid relief through one SV. The analysis should further assume that the SV fails partially open with an eff ective flow area of approximately fi cf total SV flow area, with consequent continued secondary blowdown until RHR cut-in conditions are reached. Please provide the radiological consequences for this scenario for both Callaway and Wolf Creek Plants, utilizing design basis accident methodology as outlined in f.UREG 0800 Section 15.C.3, end ICRP-2 Dose Conversion Factors (" TID Factors"), in order to derronstrate that the guidelines of 10 CFR Part 100 are not exceeded.

2. The information in Reference (1), itera 6, as well as the figures in Reference (2) indicate that the analysis was only carried cut tc the tir..e of SI ternination. The estimate of additfor.61 leakage between SI termination and primary to secondary pressure equilibratice do not appear to have sufficient bases. Therefore, please exteno your analyses to the pressure equilibtation tirre, or provide additional bercs for the estitutes provided.

ENCLOSURE 2 REQUEST FOR ADDITIONAL lhF0khATI0!!

SP:UFPS TOPICAL REPORT:

Steam Generator Tube kupture Anelysis for Sf;UPPS PLANTS 1.

The discussions in Sections 2.1 and 2.2 differentiate between identifying SGTR occurrence and identifying which steam generaters (SGs) have ruptured tube (s), and suggest indications for these identifications.

However, it is not clear that this distinction between thc tre different diacrostic activities is considered in the analyses reported. The symptoms you have identified, their alarms, and operator resronses may not be appropriate for the event scenario assurrptions. Include explicit consideration of the symptoms, indicating instrurxntation. alarms or procedural directives, and operator responses, including times for each in the timetables and analyses for each diagnostic activity. Do this separately for the identification of SGTR occurrence and for the identification of which SG(s) have ruptured.

2. The report references the k'estinghouse Emergency Resperse Guidelines (ERGS) to idertify which operator actions are performed in response to SGTR scenarios. There does not seem to be adequate plert specific information for the EPGs to provide adequate guidance, to adequately identify ano quantify SGTR diagnostics. For instance, since the ERGS do not identify radieretivity control as a critical safety function, it is not clear thot the operator would be properly directcd to consult radiation monitoring ec,uipment or that such consultation would be timely. Icentify (1) instrumentation and centrols which the analyses assume the operator will use for diagnostic purposes (2) the procedures that will be signalled by each; (3) alanrs er precedural directives which will alert the operater to use them.

Discuss the sensitivity, responsiveness, availability, and qualification of these instrurcents and controls.

3. The discussion of Section 2.0 refers to operator actiers based on ERG Section E-3, SGTP; however, this section does not ciscuss event milestones and operator actions prior to entry irte the EECs and actions based on ERG Section E-0 Address this portion of the SCTR scenario.

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4. The operator acticns assumed in the analyses of the report are predicated on a predetern.ir.ed course of operator action which is ret detailed or justified.

Observetions reported in WCAP-10599 indicate that during the ERG velidation program operator uncertainty and incorrect interpretations hasc occurred. Specific incidents cited in this report are particularly applicable to SGTR events. Justify for the SGTR Analysis scerarfct the asSUFed course of operator actions, given NUREG-C800 assumptions end itcir conscquent indications and ERG-instructed responses to those potential indications.

5. Scenarios postulated in the SNUPPS SGTR analyses presere the identification by the operator that a SGTk event is in progress and that he has transitioned to the E-3 procedure. This presumption is not adequately justified for a NUREG 0800 scenario.

For a NUREG G8CC scenario itemize step-by-step, from time of tube rupture to time of event termination, i.e., cold shutdown, all events accompanying syrptors, alarms operator actions, and times associated with each. This description should include details prior to entry into the ERGS and all transitiers in the ERGS. All operator behavior should be justified, including assumptions that the cperator would not make erroneous transitions. Assumptions required by NUREG 0800, e.g. , loss of offsite pcher, stuck rod, and their impact on operator actions should be considered. Also, otter activities appropriate to operation during SGTR scenarios (e.g., interaction with Emergency Plan Emercer.cy Action Levels) should be accounted for.

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