ML20197B680

Forwards Response to RAI Re Proposed Amend to Licenses, Requesting to Allow Operation at Increased Reactor Core Power Level at 2775 Mwt.Corrected Copy of Figure 13-1, Included in ,Encl

ML20197B680
Person / Time
Site:	Farley
Issue date:	12/17/1997
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9712240020
Download: ML20197B680 (16)

Text

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• . Dave Morey Stuthern Nucient V<e heudent Opereting Company Iadey Preject P0. Box 1235 Birmingham, Alabama 35201 i Tel 205 992 5131 December 17, 1997 SOUTHERN L COMPANY Energy to Serve 1 bur %rld" 10 CFR 50.90 Docket Nos. 50-348 50-364 U.S. Nuclear Regulatory Commission ATTN.: Document Control Desk Washington,DC 20555 Joseph M. Farley Nuclear Plant Response to Request for AdditionalInformation Kelated to Power Uprate Facility Operating Licenses and Technical Soccifications Change Reaygt Ladies and Gentlemen:

By letter dated February 14,1997, Southem Nuclear Operating 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 increased reactor core power level of 2775 megauntts thermal (Mwt). NRC letters dated July 1,1997; August 21,1997; and October 14,1997 requested SNf' provide additional information, and SNC responded by letters dated August 5,1997; September 22,1997; and November 19,1997, respectively. By telephme conference calls on December 9,10,11 and 15,1997, SNC responded to additional NRC Staff questions. Attachment I provides the SNC responses to these questions. Attachment !! provides a corrected copy of Figure 13-1, widch was included in SNC letter dated November 19,1997.

If you have any questions, please ad5ise.

Respectfully submitted, flp }gn Dave Morey ,

Sworn to and subscribed before me this/by of&l997 G3 MLtlahk b - Notary Public V I g

My Commission Expires: 6tv24- o /, d00 /

MGFJmaf:pwrup26.dec . .,

Attachments .judUg cc: Mr.L. A.Reyes, RegionII Administrator Mr. J. I. Zimmerman NRR Project Manager

' Mr. T. M. Ross, Plant Sr. Resident Ins;x:ctor ,g!gg l } *

{"12240020971217

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. l ATTACHMENTI I

SNC Response to NRC Request For Additional Information Related To Power Uprate Subauttal Joseph M. Farley Nuclear Plant, Units 1 & 2 SNC RESPONSES TO NRC QUESTIONS RESULTINO FROM NRC/SNC CONFERENCE CALLS ON DECEMBER 9,10,11 & 15,1997

SNC Response To NRC Request For AdditionalInformation Related To Power Uprate Submittal- Joseph M. Farley Nuclear Pira:t Units 1 & 2 NRC Ouestion No.1 (Reference December 9.1997 NRC/SNC Conference Call)

In your November 19,1997 response to Question No. 26 Supplement, you indicated that Reactor Trip System historical data had been reviewed in suppcrt of the pressurizer h gh pressure reactor trip response time change from 2 soonnds to 1 second What was the longest tested response time for this RTS function and how much historical data was reviewed?

SNC Response No. I

%c !ongest total response time for the pressurizer high pressure reactor trip function at Farley is about 720 milliseconds based on a review of previous test data from 1978 to date from both Farley Urute.

SNC/mge & tma.12/10/97 NRC Ouestion No. 2 (Referepsg December 10.1997 NRC/SNC Conference Call)

Please provide a comparisen of the major parameters which nave changed for tiu: power uprate containment analysis, including decay hem modeling.

SNC Response No.

See attached Table 1.

SCS/ jaw A jam - 12tl2/97 Page1 i.r'

_ _ _ .,. . . _ _ . -- _ . . . ~ . _ -

i.

TABLE 1-CONTAINMENT ANALYSIS KEY PARAMETER CHANGES -

PSAR Acalysis UPRATE

1. Calculation Code COPNITAM/COMPACI* G01EIC r

2. Mass and Energy Relemans

~ LOCA Methodology FSAR Esction 6.2.1.3.4 WCAP 14723 Section 6.4.1 MSLB Methodology WCAP-8822 WCAP-14723 Section 6.5.1 1971+20% M /1979F20" 1979+2e Decay Heat (ANS Std.)  ;

3. RHR Hx (Btu /hr/A'/*F) 330 383*

4. Cooling Water CCW Flow (spm)~ 5600 4755* l-4 Service Water Temp (F') 95 97.5 *

5. Air Cooler Performance (10' Blu/hr) at C~d ' :r'

. AirTemperature("F) (1) (6) 106.2 0.0 0.0 120.4 3.217 3.024 139.1 8.371 5.890 201.0 31.000 16 320 265.8 58.200 29.130 274.6 - 61.850 30.758

6. Initial Conditions Cont. Pres. (psig) 0 - 1.5, + 3.0 Cont. Temp. (F') 120 127*

' 7. Containment Pressure (1) (7)

Setpoints (psig ,

Hi-1 4.5 - 7.0 Hi 2 16 '19.2

- Hi-3 27 30

~ (1) Original FSAR.-

(2) Current FSAR. ; 4 (3) As-builtvtlessuppliedbyvendor. l

~( 4) Cuneet asM-=1 system flow. /

(5) Post LOCA temportture for initial pond temperature at Technical Specifications value c' 95 'F. I

- (6) Limiting condition for degradod cnoter performance,  ;

(7) Analysis value, including conserystism atet Technical Specifications value. -

I 1

Page 2 w , .-, -,

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S NRC Ow6 No. 3 (Refer == C+:=h-v 10.1997 NRC/SNC Conference Call)

In your November 19,1997 response to Question No. 8, you provided a summary of the results of the comparison of the computer codes used in the original containment analysis and the uprate analysis Please provide an explanation of the major diferences in the computer codes, why there is an approximate 10 7 difference in results, and provide companson curves of contamment pressure and t .weare vs. time for the computer codes used. ,

SNC P~am- No. 3 The original Farley containment analyses were performed using the Copa tta computer code. Wer

analyms were pu fi.w.ed using the Comped code. The LOCA and MSLB analyses for uprate were 1 performeri using the Gothie ==d~ code. The diF.wcmoes in coda are A--i below and the

- resultant curves for non uprate conditxes are provided in the attached Figures I through 4.

For the LOCA Case, Gothic prodded a prasure and temperature transient that was wry close to that predicted by Copatta lhe only significant differece in tL: two codes' LOCA pressure and temperature profdes oomis at 55 scoonds whm Copatta pabets a pressure and tcmperature peak well into the superheated rcgion that is not pmheted by GotNc.

Copetta splits the mass and energy in the blowdown between the containmet atmosphere and the sump based upon thermodynamic equilibrium in the catainment at the time step. For a two phase blowdown such as occurs during a LOCA, the effect of Copatta's treatment of the blowdown is to remove the liquid from the vapor space. Compessive heating the increases the dry steam / air mixture into the superheat

- ngion.- In any real bkmdown, the fluid eters the contammmt as a high velocity mixture of steam and water dropicts. The high mixing forecs in the containment will tmd to keep the water dn.pkts suspended in the containmmt m--f-, where their pacemce will prevent the attainment of any superheat Since supedicat canreut occur in a space fdled with water droplets, it is concluded that the peak temperature pmheted by Copatta cannot actually cxur.

1 For the MSLB Case, Gothic predksxi shghtly lower long term L..p.. ares and pressures than Copatta

- A significant differmee in the Gothic and Copatta peak bi peidures was also observed early in the j transient. l i

Ev=WN of the bkmdown revealed that a small amount of the blowdown is eutenng the containment in the liquid phase (Gothic treats this as water dropleu). Copatta splits the mass and energy in the bkmdown bctween the containment atw.u pl.c.a and the sump. The Copana sump does not interact with .

the ou.pl.erc excxpt through the long ter.n recirculation of sump water, and no recirculation oxurred i in the 1800 second MSLB runs. This had the effect of removing the water (including water droplets) entirely ' rom the centammet mlume, thereby, ehminating the cmergy rcmoval pathways (sump water surface interfacal heat and mass transfer and drop evaporation) fium the model In summary, the treatmmt ofwater hoplets in the vapor repon of the containment drives the differences in calculated ie..pu.are and piessure and raults in Gothic WM taupu-ure approximately 10 7 lower than Copetta. ,

l SCS/ jaw & jam.12/16/97 Page 3 a- =

Pressure (PSIA) ,

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LOCA PRESSURE COMPARISON COPATTA - COMPACT - GOTHIC FIGURE 1 Page 4

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LOCA TEMPERATURE COMPARISON COPATTA - COMPACT - GOTHIC FIGURE 2 Page5

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500 1.000 1.500 2.000 Time (Sec)

LOCA MSLB PRESSURE COMPARISON - COPATTA - COMPACT - GOTHIC .

FIGURE 3 Page 6 i

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LOCA MSLB TEMPERATURE COMPARISON - COPATTA - COMPACT - GOTHIC '

FIGURE 4 1

Page 7

NRC Ouestion No. 4 (Reference Dwar 10.1997 NRC/SNC Conference Call) -

'Ihe resuk of the containment analysis indicates that the peak temperature inside contain.nent

- exceeds the containment structural design temperature. Please explain w'ay the containment design remains acceptele.

SNC Response No. 4

- The temperatures reported in Table 2.13-1 of the BOP Licensing Report aie peak air temperatures.

. The containment structure response lags behind the air temperature as shown in current FSAR Figure 6.2-78. *Ihe uprate contaicment wall temperature profile response to a LOCA at various

- times (cquivalent to current FSAR Figure 6.2-78) is shown in a*aAnd Egure 5, and the wall surface temperature as a function of time for the MSLB is shown in attached Figure 6. As shown in these figures, the containment structure temperature remains below the structure design temperature (280 *F; at all times.

SCSrjaw te jnm.12/15/97 9

9 Page 8

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E_e i e e I i,e i e 8.3 8.4 8 8.1 8.2 Relative Distance COTHIC 5.8(QA)-b 81/17/97 14:81:43 CONTAINMENT WALL TEMPERATURE PROFILE VS RELATIVE DISTANCE THROUGH WALL FOR LOCA FIGURE 5 Page 9 l

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CONTAINMENT WALL SURFACE TEMPERATURE ,

VS TIME FOR MSLB FIGURE 6 Page 10

NRC Ouestion No. 5 (Reference December 11.1997 NRC/SNC Conference Call) ,

in Table B of Additional Quertion No. 7 of the SNC response dated August 5,1997, the maximum range of stress intmsity for the Steam Generator divider plate entries were reported as

>3S.. Provide the calculaicd stress vslue and clarify why this result is acceptable.

SNC Reso<nse No. $

la the original stress report [ Reference 1], the divider plate analysis based on elastic analysis showed that the maximum stress range exceeded the 3S. limit. Calculatims for 11ydrotest conditions were raA included since work on previous model steam generators indicated a plastic analysis would be required for the liydrotest conditions. De clastically calculated maximum range of stress intensity was 121 ksi between the ambient and less of Load at 15 seconds conditions, which &_cW the maximum allowable value of 69.9 ksi This would increase to 129.5 ksi for the power uprate conditions. Section NB-3228.l(b) of Reference 2 states that the 3S limit may be exceeded if the analysis is done plastically and if shakedown occurs. Because the divider plate loading originates from the imposed deformations derived from the tubesheet and channel head displacements, the amount by which it can deform is limited by these displacements.

Therefore, shakedown does occur.

Plastic analyses had ba.cn performed in Reference I for the !ess of Load at 15 monds and the liydrotest conditions. ne equivalent plastic strains were multiplied by a str .s concentration factor of 1.5 ami the clastic modulus to obtain the stresses used in the fatigue evaluation. %c stresses obtained from the plar.ic analysis were combined with the clastic stresses to perform the >

fatigue evaluation of the divider plate.

For the uprate evaluation the scale factors due to the uprate were applied to the appropriate stresses and the fatigue evaluation was performed. The resised fatigue usage is less than 1.0, satisfying the ASME Code requirements.

Resoonse No. 5 References

1. "$1 Series Steam Generator Stress Report, Section 3.3, Divider Plate Analysis," Westinghouse Tampa Division, Tampa, FL, April 1972.

2, ASME Boiler and Pressure Vessel Code Section III, Nuctcar Pour Plant Components,1971 Edition.

W/ alt & tmi.12/16/97 Page11 l

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NRC. Question No. 6 (Reference Desunktr 15.1997 NRC/SNC Conference Call)

The Farley uprate proposed Technical Specifications changes for ECCS Subsystem Surveillance 4.5.2.g. include an increase in RiiR Pump recirculation flow differential pressure (AP) from 136 to 145 psid. Why did RiiR pump AP increase and how did this affect the RIIR System MOV's?

SNC Responss.]iL6 As discussed in Section 4.1.3.1 of WCAP-14723, the Low Ilcad Safety Injection (LilSI) System performance modeled in the uprate analyses was based on a new (i.e., revised) pump curve with

, 10% flow degradation. The revised pump curve resulted from physical modifications to the RIIR h pump impeller (i.e., underfiling), widch were previously implemented in Farley Units 1 & 2. Since the uprate analyses credited 11. R11R pump performance improvements, the Farley Technical Specifications ECCS surveillance requirements for RIIR pump AP had to be revised.

As discussed in the response to Additional Question No.12 in SNC letter datal August 5.1997, the maximum pump shutoff head is typically used in d:temiining MOV thrust / torque requirements.

For the Faricy LIISI System MOVs, including the RIIR pump miniflow valves, the MOV operating requirements are based on the RilR pump design shutoft . cad, which bounds both the original pump performance curve and the revised pump performance cur ; maximum head values.

Since the pump modification did not change the design shutoff head, the LilSI system MOV's were not impacted.

SNC4nge 12/I7/97 l

Page 12

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ATTACHMENT II SNC Response to NRC Request For Additional Information Related To Power Uprate Submittal - Joseph .M. Farley Nuclear Plant, Units ; & 2 CORRECTED FIGURE 13-1 REFER TO QUESTION NO.13 (PAGE AI.28)

SNC LETTER TO NRC DATED NOVEMBER 19,1997 s

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