ML20056D518

Forwards Comments on Draft NUREG-1477, Voltage-Based Interim Plugging Criteria for SG Tubes, Per NRC Publication in Fr (58FR35985) on 930702.Util Plans to Submit Addl Info Beginning of Sept

ML20056D518
Person / Time
Site:	Catawba
Issue date:	08/11/1993
From:	Rehn D DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
FRN-58FR35985, RTR-NUREG-1477 NUDOCS 9308170027
Download: ML20056D518 (5)

Text

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. DukeIbwer Compaq D L Ruis

( Catawba Nuclear Generation Department nce President 4800 ConcordRoad (8D3)S312205 Office Eurk, SC23745 (803)831326 Fax t DUKEPOWER August 11,1993 -

i U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555

Subject:

Catawba Nuclear Station Docket Nos. 50-413 and 50-414 Comments on Draft NUREG-1477, " Voltage-Based Interim Plugging Criteria For Steam Generator Tubes" On July 2,1993, the NRC published in the Federal Register (58 FR 35985) a notice that -

draft NUREG-1477, " Voltage-Based Interim Plugging Criteria For Steam Generator Tubes," was available for public comment. Attached are Duke Power Company's comments on this draft report which specifically relate to dose analysis methdology and those issues which affect the dose calculations. Duke Power intends to provide further -

comments on this report but will not have this information available for submittal until after August 16,1993, when the public comment period ends. We plan to submit this additional -

information by the beginning of September. Duke Power appreciates the opportunity to comment on this draft report since this is an issue of critical importance to Catawba Nuclear Station as we presently operate with an Interim Plugging Criteria and plan on renewing its use for the next operating cycle.

Very truly yours, f

D. L. Rehn RKS/

Attachment 9308170027 930811 8 160060 f. 1 PDR ADDCK 05000413 $

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U. S Nuclear Regulatory Commission August 11, 1993 Page 2 xc: (w/ Attachment)  ;

Mr. S. D. Ebneter Regional Administrator, Region II U. S. Nuclear Regulatory Commission .

101 Marietta Street, NW, Suite 2900 1 Atlanta, GA 30323 Mr. R. E. Martin, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 14H25 ,

Washington, D.C. 20555  ?

Mr. R. J. Freudenberger NRC Resident Inspector Catawba Nuclear Station

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Attachment 1 Comments on Draft NUREG-1477

" Voltage-Based Interim Plugging Criteria for Steam Generator Tubes" Comment #1 It is our opinion that the NRC Staf 'should use ICRP-30 dose conversio rfactors for both the licensing basis and realistic dose analysis cases. The NRC has used ICRP-2 dose conversion factors for the licensing basis dose analysis cases. However, ICRP-30 states it " supersedes ICRP Publication 2." ICRP-30 was written to update dose models based on more recent information, and hence, the ICRP-2 models are outdated. Additionally, the use ofICRP-30 dose conversion factors would be consistent with publication EPA-400, which is being used for protective action guidelines. Realistic dose analysis cases should include alternate models for iodine spiking and other realistic versus design basis assumptions, but the use ofICRP-30 versus ICRP-2 dose conversion factors does not make a dose analysis case more " realistic;" nor does the use ofICRP-2 dose conversion factors make a dose analysis "more conservative." Conservatisms in an analysis should include factors intentionally included to increase calculated results, rather than outdated information. Committed dose equivalent due to the intake and uptake of radioactive iodine is well understood, the most recent data concerning which is contained in ICRP-

30. Both ICRP-2 and ICRP-30 were written for " reference man," but ICRP-30 incorporates more recent metabolic modeling and physiologic data.

Comment #2 Iodine spiking models have been developed which closely correspond to empirical data in the nuclear power industry, such as EPRI NP-4595 and EGG-NERD-8648. It is our opinion that these sources of information should be utilized to ascertain the realistic affects of the Main Steam Line Break accident on offsite dose. The " realistic" dose analysis cases should not use the Standard Review Plan iodine spiking model for coincident iodine spike (i.e.,500 times the release rate corresponding to the iodine concentration at the equilibrium value stated in Technical Specifications). This has been demonstrated to be a grossly conservative model.

Comment #3 Section 4.1.5 appears to require a control room operator dose analysis case to be evaluated. It is our opinion that this dose analysis case should not be required. The potential increase in primary to_ secondary leak rate with the IPC criteria may indeed be an additional dose contribution compared to the historical assumptions used for the Main Steam Line Break offsite dose analysis (i.e., only pre-existing primary to secondary leakage). However, if control room operator dose is a significant issue, it should have been significant for the cases with Standard Review Plan assumptions and methodology as well. However, the Standard Review Plan does not require a control room operator Page1

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dose case to be evaluated for the Main Steam Line Break because it has not historically been the most limiting case. Offsite dose has historically been the limiting case. The potential increase in primary to secondary leak rate will increase the calculated offsite dose, and various inputs (e.g., initial RCS concentration) should be modified if necessary to envelope the offsite dose. If this is accomplished, then the control room operator dose case should also be enveloped since this case is not usually seen to be limiting. An increase in leak rate would not change the case which is determined to be the most limiting; it would only change the total calculated results for the limiting case.

Comment #4 The increase in primary to secondary leak rate is required to be calculated based on an RCS pressure corresponding to the pressurizer code safety valve setpoint, assuming that the operators fail to terminate or reduce safety injection flow rate. It is our opinion that t this should not be required. Operator actions should be credited for reduction in the RCS ,

pressure used for the calculation ofleak rate. This should be the case not only because of the extensive training and procedural guidance in place, but also because of the following ,

considerations.

Two scenarios are discussed in NUREG-1477. The first scenario discussed is a Main l Steam Line Break, in which the pressurizer PORVs fail, and the operator fails to terminate safety injection flow. Considering the accident probability, the PORV failure probability and the operator non-response probability, this is a highly improbable scenario. The second scenario discussed is a Feedline Break, in which an increase in the -

RCS temperature and pressure occurs due to the event itself(i.e., loss of heat sink). It is ,

concluded that since the pressurizer PORVs may be unavailable, the pressurizer code safety setpoint should again be used to determine the maxiinum RCS pressure for which the primary to secondary leak rate ca!culation is performed.

However, this approach combines a source term which would be present in a Main Steam Line Break (i.e., iodine spike) with an accident (i.e., Feedline Break) which would not  !

cause an iodine spike. In a Main Steam Line Break, the rapid dissipation of heat from i the faulted steam generator steamline will reduce RCS temperature and pressure with no  ;

operator action or intervention (i.e., it occurs due to the Main Steam Line Break itself). i If fuel pins are leaking plenum fission product gases during normal operation, and a  !

sudden depressurization of the RCS occurs, an increase in the rate at which these plenum  ;

gases escape could occur. This is the mechanism which causes the iodine spike which j the Standard Review Plan conservatively models. With the Feedline Break, this j mechanism is not present. The Feedline Break causes a rapid increase in the RCS  ;

t temperature and pressure, leading the NRC Staff to conclude that this increased RCS pressure should be used in the calculation of primary to secondary leak rate. However, i the iodine spike will not be present for this increased RCS pressure because the mechanism which causes it is not present (i.e., rapid reduction of RCS pressure). -

It appears that a worst case iodine spiking model for the Main Steam Line Break has l been combined with a worst case RCS pressure and primary to secondary leak rate from i

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the Feedline Break. It is our opinion that these two considerations should be separated.  !

Calculation of maximum RCS pressure should be based only on the Main Steam Line l Bmak scenario. Since it is highly improbable that this RCS pressure would exist in a ,

Main Steam Line Break, and since the iodine spike would not exist for a Feedline Break, a more realistic RCS pressure value should be used to yield a more realistic calculation of primary to secondary leak rate. I Comment #5 In the calculation of primary to secondary leak rate, the maximum RCS pressure should not be used for the entire duration of the accident. It is our opinion that credit should be allowed for a reduction in primary to secondary differential pressure based on cooldown .

of the RCS from steaming of the steam generators. Extensive operator training and procedural guidance exists to ensure that this action is accomplished. Modeling of RCS pressure decrease versus time could be performed to recalculate primary to secondary j leak rate versus time, yielding more realistic accident analysis results.

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