ML20214A224

Forwards Request for Addl Info to Continue Review of 860108, 0211,0401 & 0904 Submittals Re Snupps Steam Generator Tube Rupture Analysis.Info Requested within 45 Days of Ltr Receipt

ML20214A224
Person / Time
Site:	Callaway
Issue date:	11/12/1986
From:	Oconnor P Office of Nuclear Reactor Regulation
To:	Schnell D UNION ELECTRIC CO.
References
NUDOCS 8611190312
Download: ML20214A224 (5)

Text

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,4 Docket No.: .50-483 Mr. Donald F. Schnell Vice President - Nuclear NOV 121986 Union Electric Company Post Office Box 149 St. Louis,flissouri 63166

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE SNUPPS STEAM GENERATOR TUBE RUPTURE ANALYSIS The staff is centinuing its review of your submittals dated January 8,1986, February 11, 1986, April 1, 1986, and September 4, 1986. To permit us to continue cur 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.

Sircerely, Paul W. O'Connor, Project Manager PllR Project Directorate #4 Division of PWR Licensing-A

Enclosures:

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( Mr. D. F. Schnell Callaway Plant Union Electric Company Unit No I cc:

Mr. Nicholas A. Petrick Lewis C. Green, Esq.

Executive Director - SNUPPS Green, Hennings & Henry 5 Choke Cherry Road Attorney for Joint Intervenors Rockville, Maryland 20850 314 N. Broadway, Suite 1830 St. Louis, Missouri 65251 -

Gerald Charnoff, Esq.

Thomas A. Baxter, Esq. Ms. Marjorie Reilly Shaw, Pittman, Potts & Trowbridge Energy Chainnan of the League of 2300 N Street, N. W. Women Voters of Univ. City, M0 Washington, D. C. 20037 7065 Pershing Avenue University City, Missouri 63130 Mr. J. E. Birk Assistant to the General Counsel Mr. Donald Bollinger, Member Union Electric Company Missourians for Safe Energy Post Office Box 149 6267 Delmar Boulevard St. Louis, Missouri 63166 University City, Missouri 63130 U. S. Nuclear Regulatory Commission Mr. Dan I. Bolef, President Resident Inspectors Office Kay Drey, Representative RR#1 Board of Directors Coalition Steedman, Missouri 65077 for the Environment St. Louis Region Mr. Donald W. Capone, Manager 6267 Delmar Boulevard Nuclear Engineering University City, Missouri 63130 Union Electric Company Post Office Box 149 St. Louis, Missouri 63166 Chris R. Rogers, P.E.

Manager - Electric Department 301 W. High Post Office Box 360 Jefferson City, Missouri 65102 Regional Administrator U. S. NRC, Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Mr. Ronald A. Kucera, Deputy Director Department of Natural Resources P. O. Box 176 Jefferson City, Missouri 65102 Mr. Glenn L. Koester Vice President - Nuclear Kansas Gas and Electric Company 201 North Market Street Post Office Box 208 Wichita, Kansas 67201

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' ENCLOSURE 1

• REQUEST FOR ADDITIONAL IflFORMATION SliUPPS STEAf: GENEPATOR TUBE RUPTURE (SGTR) A!!ALYSIS CALLAWAY & WOLF CREEK

References:

II) Letter from Nicholas A. Petrick, SNUPPS, to Harold R.

Denton, Ni:C " Steam Generator Tube Rupture Analysis -

ShtPPS" September 4, 1986 (2) SNUPPS Report " Steam Generator Single-Tube Perttre Analysis for SNUPPS Plants" December 1985

1. In Reference (1), item I, you indicated an average break flow rate of 35 lba/sec. for the design basis SGTR, based on the Burnell correlation, which would result in a rargin to solid secondary water relief of 271 cubic feet. You also indicated that an increase in average break flow rate greater than 12% of the calculated flow could result in liquid secondary relief. The staff believes that, baseo on this response, there is sufficient uncertainty in the break flow calculations, as well as in operator action times (see Enclosure 2) and interval hetween safety injection (SI) termination and break flow termination (see question 2 below), to warrant assuming that safety valve (SV) liquid relief can occur. Since the SVs are not specifically designed for liquid or two chase flow, this could result in valve malfunction. During the Ginna SGTR event, licuid relief through a SV prevented proper valve reseating, with consequent continued secondary leakage.

Therefore please perform an analysis which assumes a design basis SGTR with loss of offsite power (LOOP), closure of the ruptured SG main steam isolation valve (14SIV), and SG overfill resulting in liquid relief through one SV. The analysis should further assume that the SV fails partially open with an effective flow area of approximately 5% of total SV flow area, with conscouent 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 methodelegy as outlined in NUREG 0800 Section 15.6.3, and ICRP-P Dose Conversion Factors ("T!D Factors'), in order to demonstrate that the cuide11oes of 10 CFR Part 100 are not exceeded.

2. The information in Peference (1), item 6, as well as the figures in Reference (2) indicate that the analysis was only carried out to the time of Si tenaination. The estimate of additional leakage between 51 ternination and primary to secondary pressere equilibration do not appear to have sufficient bases. Therefore, please extend your analyses to the pressure equilibration time, or provide edcitional bases for the estimates provided.

0 EhCLOSURE 2 RE0 VEST FOR APPITIONAL INFORMATION SNUPPS TOPICAL REPORT:

Steam Generatcr Tube Rupture Analysis for SNUPPS PLANTS 1.

The discussions in Sections ?.1 ar.d 2.2 differentiate between identifying SGTR occurrence and identifying which steam generators (SGs) have ruptured tube (s), and suppest indicitions for these identifications.

Mcwever, it is not clear that this distinction between the two different diegnostic activities is corsidered in the analyses reportec. The synptoms you have identified, their citrms. and operator responses may not be appropriate for the event scenario assumptions. Include explicit consiceration of the symptoms, ind.cating instrumentation, alems or procedural directives, and operater responses, including times for each in the timetables and analysts fcr 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 Westinghouse Emergency Response Guidelines (ERGS) to identify which operator actions are performed in retrense to SGTP. scenarios. There does not seen,to be acequate plant. specific information for the ERGS to provide acequate guidance, to adequately identify and quantify SGTR diagnostics. For instance, since the ERGS ao ret identify radioactivity control as a critical safety functicn, it it r,ct clear that the operator would be properly directed to consult radiationIdentify timely. monitoring) equipment or (1 instrunentation thot and such consultation controls would be which the aralyses assune the operator will use for diagnostic purpcses (2) the procedures that will be signalled by each; (3) alarms or procedural directives which will alert the operator to use them.

Discuss the sensitivity, responsivenest, availability, and qualification of these instruments and controls.

3. The discussion of Section 2.0 refers to operator actions based on ERG Section E-3, SGTR; however, this section does not discuss ever.t milestones and operator actions prior to entry into the ERGS and actiers besed on ERG Section E-0. Address this portion of the SGTR scenaric.

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The operator actions assumed in the analyses of the report are predireted on a predetermined course of operator action which is not detailed er

.iustified. Observations repcrted in kCAP-10599 indicate that durirg tFe ERG validatien progrou operator uncertainty and incorrect interpretations have occurred. Specific incidents cited in this report are rerticularly applicable tc SGTR events. Justify for the SGTR Analysis scenarios the assumed course of operator actions, given NUREG-CECC assuretions and their consequent indicatiers end ERG-instructed responses to those potential indications.

5. Scenarios postulated in the SHIPPS SGTR analyses presure the identification by the operater that a SGTR event is in progress and ti.et he has transitioned to the E-3 procedure. This presumption is not ecequately justified for a NUPIG 0800 scenario.

Fcr a NUREG 0800 scenario itemize step-by-step, from time of tubc rurture to time of event termination, i.e., cold shutdown, all events accoreanying symptoms, alarms operator actions, and times associated with each. This description should include details prior to entry into the EFGs and all transitions in the ERGS. All operator behavior should be justified, including assumptions that the operator would ret rake errorecus transitions. Assumptions required by huREG 0800, e.g., loss of offsite power, stuck rod, ano their impact on operator actions should be considered. Also, other activities eppropriate to operation during SGTP scenaries (e.g., interaction with Emergency Plan Emergency Action Levels) should be accounted for.

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