ML20217A027

Forwards Response to 980408,09 & 13 Telcon RAIs Re 970214 License Amend Request to Allow Operation at Increased Reactor Core Power Level of 2775 Mwt

ML20217A027
Person / Time
Site:	Farley
Issue date:	04/13/1998
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9804220131
Download: ML20217A027 (7)

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,0 ave Mor:y Siuthe'm Nuclear l Mce President Op:r: ting Compny brieyProject . P.O. Box 1295

. Birmingham, Alabama 35201 Tel 3.992.5131 1

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SOUTHERN co==.

April 13,1998 l

Docket Nos.: 50-348 10 CFR 50.90 50-364 I

U. S. Nuclear Reed =*my Commission l A'ITN.: Document Control Desk Washmgton,DC 20555 Joseph M. Farley Nuclear Plant l Response to Request for Additional Information Related to Power Uprate l Facility Operatina Licenses and Technical Specifications Chance Reauest Ladies and Gentlemen:.

i By letter dated February 14,1997, Southern Nuclear Operatmg Company (SNC) proposed to

. amend the Facility Operating Licenses and Technical Specifications for Joseph M. Farley Nuclear

! Plant (FNP) Unit I and Unit 2 to allow operation at an increa.wd reactor core power level of 2775 megawatts thermal (MWt). NRC letters dated July 1,1997; August 21,1997; and October 14, 1997 requested SNC provide additional information. SNC responded by letters dated August 5, 1997; September 22,1997; and November 19,1997 respectively. SNC letters dated December 17 and 31,1997; January 23,1998; February 12 and 26,1998; and March 3,6 and 16,1998

, responded to NRC questions resulting from conference calls. During telephone conference calls on April 8,9 and 13,1998, the NRC Staff requested responses to additional questions. The l Attachment provides the SNC responses to these questions; the response to Question No. 3 l includes the qualitative assessment requested by the NRC Staff. i i

! SNC performed a qualitative aawaament of the RCP locked rotor and the control rod ejection I events with respect to the Main Steam Line Break (MSLB) radiological analysis and steam generator Alternate Repair Criteria (ARC). The assessment considered the transient differential  !

pressure between the primary and ==dary side of the steam generators, Mma*M the primary j leakage that potentially'could occur, and then calculated offsite doses. 'Ibe assessment results i support the validity of the conclusions of WCAP-12871, Revision 2, "J. M. Farley Units 1 and 2 SG Tube Plugging Criteria for ODSCC at Tube Support Plates," i.e., the MSLB is the most limiting event for ARC. (

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! As agreed to by the NRC Staff, the transient analysis, evaluations and calculations used to support i

this assessment have not been forr= lim The assessment is based on power uprate conditions and conservative engineeringjudgment For the locked rotor and rod ejection events, some radiological s assumptions were revised as compared to those prevhusly submitted for power uprate. The uprate l calculations previously submitted for Staff review will continue to remain the calculation of record. l The assessment is being provided to the Staff form' for:.ation only. I I

l If you have any questions, please advise. j Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY b hW l Dave Morey 1 Sworn to andsubscribed befo me this/__ o & l998 j

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' NotaryPublic' U cLab b My CommissionErpires: f?dWW bu l,c9CO{

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l MGE/ cit:pwrup39. doc i Attachment  !

l cc: Mr. L. A. Reyes, Region II Admmistrator '

Mr. J. I. Zinunerman, NRR Project Manager Mr. T. M. Ross, Plant Sr. Resident Inspector t

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ATTACHMENT SNC Response to NRC Request For Additional Information Related to Power Uprate Submittal - Joseph M. Farley Nuclear Plant, Units 1 and 2 SNC RESPONSE TO NRC QUESTIONS RESULTING FROM NRC/SNC TELEPHONE CONFERENCE CALLS ON APRIL 8,9 & 13,1998

. SNC Response to NRC Request For Additional Information Related to Power Uprate Submittal- Joseph M. Farley Nuclear Plant, Units 1 and 2 NRC Ouestion No.1 (Reference Anril 8 & 13.1998 NRC/SNC Conference Call) l Enclosure 1 of SNC letter dated March 7,1997, " Steam Generator Tube Support Plate Voltage-Based Repair Criteria," indicates the intact steam generators mass release is 479,000 lbs for the ARC main steam line break radiological analysis. However, Table E of SNC letter dated August 5,1997, " Response to Request for Additional Information Related to Power Uprate Facility Operating Licenses and Technical Specifications Change Request," indicates the intact steam generators mass release is 339,000 lbs for the uprate main steam line break radiological analysis.

Why are the steam releases different in these two Farley licensing submittals?

SNC Response to Ouestion No.1 Prior to 1984, typical W@ingbase steam release calculations used for the dose calculation inputs were very conservative and generic in nature, because the calculations were performed using plant conditions that would bound multiple plants of similar design. 'Ihis conservative approach also assured that calculation revisions would not be required for minor plant changes. In 1984, Westinghouse developed a safety analysis standard for calculation of steam releases for radiological considerations, so that there would be uniformity of calculations by the analysts and to ensure that calculations accounted for all applicable plant-specific conservatisms (e.g., power level, temperatures, thick metal effects, etc.).

Using the generic approach, for Farley, Weinghase determined a steam release value of 479,000 lbs for the intact steam generators mass release input to radiological analysis. (Reference FSAR Table 15.4.23.) This conservative steam release value was used in various steam line break radiological analyses prior to the Farley power uprate project. For the power uprate, Farley-specific calculations resulted in a steam release value of 323,000 lbs for the intact steam generators. This value was conservatively increased by 5% (i.e.,339,000 lbs) for evaluation of radiological consequences. (Reference response to RAI Question No. I1 in SNC letter dated i February 26,1998.)

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W/rjm-4/10/98 NRC Ouestion No. 2 (Reference April 9 & 13.1998 NRC/SNC Conference Call)

With respect to Farley power uprate analyses and alternate repair criteria (ARC) and the steam generator tube uncovery program analyses and conclusion presented in WCAP-13247 for the RCP j locked rotor and control rod ejection events, is Farley considered to be a representative plant?  ;

SMC Response to Ouestion No. 2  !

l The issue of tube bundle uncovery was considered in a generic Westinghouse Owners Group (WOG) program as presented in WCAP-13247, " Report on the Methodology for Resolution of the Steam Generator Tube Uncovery Issue," March 1992. 'Ihe program concluded that the effect of tube uncovery is essentially negligible for the limiting SGTR transient. It also concluded that for 1

non-SGTR events, such as locked rotor and rod ejection, the probability that an event could result in off-site radiological consequences that exceed the acceptance limits was estimated to be sufficiently low so as to place this issue on the exclusion category as defmed in NURrG-0933.

Therefore, the program concluded that the steam generator tube uncovery issue could be closed without any further investigation or generic restrictions. The NRC review of the WOG program concluded "the Westmghouse analyses i.w i. hate that the effects of partial steam generator tube uncovery on the iodme release for SGTR and non-SGTR events is negligible. Wrefore, we agree with your position on the matter and consider tlJs issue resolved." (Reference NRC letter from Robert C. Jones to Lawrence A. Walsh, "WMinghause Owners Group Steam Generator Tube UncoverIssue," dated March 10, 1993.)

The conclusions of the WOG program apply to the Farley units. The implementation of a power uprating and ARC (in accordance with NRC Generic Letter 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking," dated August 3,1995) have no impact on the conclusions of the WOG program. h effects of partial steam generator tube uncovery on the radiological consequences of SGTR and non-SGTR events, including locked rotor and rod ejection, are negligible and do not present a safety concern for Farley.

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NRC Ouestion No. 3 (Reference April 9 & 13.1998 NRC/SNC Conference Call) a) With respect to the 10 CFR Part 100 off-site and on-site dose limits, the Farley power uprate analyses, and the assumptions for steam generator tube ARC, are the RCP locked rotor and control rod ejection events (due to potentially large increases in source terms resulting from fuel claMmg or pellet damage) more limiting than the main steam line break event?

l b) Based on short-term transient parameters, provide a qualitative assessment for each of these j

events, which includes the potential for accident-induced steam generator tube leakage (similar to that Mima *M for application of ARC to the Farley steam generators) and the resultant impact on the off-site and on-site radiological doses, i l SNC Resnonse to Ouestion. No. 3 '

a) For the Farley power uprate with ARC, the Farley steam line break event radiological consequences are more limiting than the radiological consequences of postulated locked rotor and rod ejection events. The licensing basis for this statement is provided by the non-steam line break evaluations presented in WCAP-12871, Revision 2, "J. M. Farley Units 1 and 2 SG Tube Plugging Caiteria for ODSCC at Tube Support Plates," February 1992. WCAP-12871, Section 11.3, concludes that the increased source terms associated with these events are offset by: reduced primary-to-aeaM=y differential pressure; decreased flashing and increased nuxing in the steam generator; and continued coverage of the steam generator tubes at the tubesheet and tube support plate interfaces & assumption of no long-term tube uncovery is supported by WCAP-13247,

" Report on the Methodology for Resolution of the Steam Generator Tube Uncovery Issue," March 1992. This licensing basis is supported by the qualitative assessment presented below.

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1 b) An assessment of the primary and secondary systen pressure transient data associated with the l locked rotor and rod ejectioc analyses deternuned that the locked rotor pressure transients are more l challenging than the control rod ejection pressure transients. Based on the pressure and temperature transient data from a locked rotor analysis performed especially to address this

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question, the primary pressure peaks in about 3 seconds, and it levels off after about 25 seconds l To estimate primary-to-seconday leak rate in the Farley steam generators, scoping calculations 4 l were performed to deternune the leak rate ratio for a steam line break versus a locked rotor l

transient, h calculation =*hulalogy is similar to that applied to adjust measured leak rates for pulled tube specimens Based on the pressure and temperature transient results discussed above, the leak rate ratio was calculated separately for 0 to 25 seconds and >25 seconds time periods.

1 i i Ratio O to 25 seconds >25 seconds Leak Rate at SLB Condition 7.0 9.1 Leak Rate at Locked Rotor Condition l

The above leak rate ratios are based on the equation for vanation of crack openmg area with l

pnmary-to -aaday differential pressure as described in EPRI Report NP-7480-L, Volume 1, Revision 1, " Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube i

Support Plates - Database for Alternate Repair Limits, Volume 1: 7/8 Inch Diameter Tubing,"

Appendix B, using leak rate test data for 5 pulled tube specimens from FNP Units 1 and 2 in the )

ARC database It is also noted that since significant TSP movement is not expected during a i locked rotor event, packed TSP crevices would preclude any significant leak through ODSCC indications as discussed in WCAP-14707, Revision 1, "Model 51 Steam Generator Limited Tube l Support Plate Displacement Analysis for Dented or Packed Tube to Tube Support Plate Crevices," January 1997.

As an illustration of applying the above ratios the following example is presented. The steam line l break leak rate calculated for the limiting SG in Farley Unit 2 (including UOA indications with PIs and applying the latest ARC ddah==a) is 9.34 gpm (at room temperature). The correspondmg leak rate estimates for a locked rotor condition obtained by applying the above leak rate ratios are 1.3 l gpm (0 to 25 seconds) and 1.0 gpm (>25 seconds). It is noted the steam line break leak rate value used (9.34 gpm) was obtained assunung that leak rate is indanandent of bobbin voltage. With the recent NRC clarification on the requirements for a leak rate correlation, a leak rate vs. bobbin

! voltage correlation can now be applied for 7/8" tubes, which leads to a significant reduction in the l estimated leak rate.

l l An assessment of the potential impact on the off-site doses based on tim leak rate ratios presented above follows. Since the accident induced leakage adimwad for the locked rotor accident bounds that for the control rod ejection, this leak rate will be used for both assessments. For similar steam generator tube conditions (i.e., those which result in the limiting leakage of 24 gpm for a main steam line break), the locked rotor leakage is 1/7 of the limiting leak rate (i.e., 3.4 gpm) for the 0 to ,

25 seconds period and 1/9.1 of the limiting leak rate (i.e.,2.6 gpm) for the >25 seconds period. ]

'Ihese leakage rates are assumed to exist in all three steam generators, and the long-term leakage is i assumed to last until the RCS and steam generator pressures equalize. For the control rod ejection event, the duration, as described in the Farley Power Uprate NSSS Licensing Report (WCAP-I l

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l 14723), is 25.00 =a~=A For the locked rotor event, it is assumed that the RCS pressure remains constant to the RHR cut-in at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In addition, for the locked rotor, the estunated fuel failure has been reduced from the power uprate radiological assumption of 20% of the gas gap to the

Farley-specific value of 6.3% (His is a conservatively large number for the " rods in DNB" l calculated for the Farley uprate locked rotor analysis; e.g., the Farley Unit 2 Cycle 13 value is l <0.135%) ne rod ejection event source tenn assumption is the same as used in the power uprate radiological analysis. Comparison of the control room X/Q with iodine protection factor and the off-site X/Q indicates the control room thyroid doses for these accidents with accident-induced leakage is not limiting. %c limiting off-site thyroid doses compare to the acceptance limits as follows.

Partition Total Acceptance Percent Event Factor La6a Duration Limit mEM) ofLimit MSLB (ARC) 1 24 gpm 8 hr 30 100 Locked Rotor 0.01 10.2 gpm 8 hr 30 67 (t <25 sec) 7.8 gpm (t >25 sec)

Rod Ejection 0.01 10.2 gpm 2500 sec 75 63 (t <25 sec) 7.8 gpm (t >25 sec)

These qualitative results demonstrate that the accident-induced leakage limit detennined for the main steam line break n:nmins Ihniting.

W/dh & wjs & vs - 4/11/98 & SCS/ jaw - 4/12/98 l

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