ML20217H872

Forwards Response to Request for Addl Info Re Power Update Submittal from NRC Staff.Addl Info Requested in Supplemental Request Provided in Attachment II

ML20217H872
Person / Time
Site:	Farley
Issue date:	08/05/1997
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9708130310
Download: ML20217H872 (73)

Text

_ _..

Dave Morey 5:inhorn Nucle:r We Pitudent Op:t:ttg C:mpany f arley Pt0Jett PO Boni?%

i e ihrmirgham. Alatiama 35201 bl205 997 5131 August 5, 1997 SOUTHERN COMPANY Energy to Servel' ur World*

o Docket Nos.

50 348 10 CI R 50.90

$0 364 U. S. Nuclear Regulatory Commission NITN.: Document Control Desk Washington, DC 2055$

Joseph hl. l'atley Nuclear Plant t

Response to Request for Additional Information l

Related to Power Uprate l'acility Operating Licenses mLTeshriitalSluificatieJitDianxc. Request Ladies and Gentlemen:

Ily letter dated I'cbruary 14,1997, Southern Nuclear Operating Company (SNC) submitted a request to amend the l'acility Operating Licenses and Technical Specifications for Farley Nuclear Plant Units I and 2 to allow an inercase in the licensed thermal power from 2652 h1Wt to 2775 htWt. On July 7, 1997, SNC received a request for additional information (RAl), dated July 1,1997, related to the Farley power uprate submittal from the NRC staff. On July 28,1997, SNC received a supplement, dated July 24,1997, to the July 1,1997 RAl The SNC response to the RAI is providalin Attachment 1. 'the additional infbrmation requested in the supplemental request is provided in Attachment 11, if you have any questions, please advise.

Respectfully submitted.

l fl ')?lft i

Dave hiorey Sworn to and subscribed be >re me this 5 <Tay of 1997 1

OM Notary Pubb'c V hty Commission Expires:_ _

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i ATTACllMENT I I

I SNC llesponse To NRC Request For Additional Infonnation i

Related To Power Uprate Submittal. Joseph M. Farley Nuclear Plant, Units 1 & 2 l

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i SNC Response to NHC " Request for Additional Information Related to Power Uprate Submittal-Joseph M. Farley Nuclear Plant, Units I and 2" GENERAL QUESTIONS REGARDING WCAP.14723 NRC_QualloitRomi Please provido a discussion of the adequacy of the primary and secondary overpressure protection given the relative relieving capacity has gone down (relative to rated power). Include a Standard Review Plan Section 5.2.2 analysis.

SNC Response No.1 The adequacy of the primary and secondary userpressure protection at uprated conditions was assessed as part of the power uprate analyses. The assessment was perfonned by analyzing the l

FS AR limiting transients (primarily the Loss of Loadtrurbine Trip event) and showing that the primary and secondary pressure limits continue to be met at uprated power. The results of the FSAR limiting transient analyses are provided in Section 6.0 of the Power Uprate NSSS Lived ng Report, with the results of the Loss of Load /rurbine Trip event provided in Section 6.2.7. liased on the analyses for the Loss of Loadfrurbine Trip event. Section 6.2.7.6 concludes that the peak primary and secondary pressures remain below I 10% of design at all titnes.

Standard Review Plan Section 5.2.2 addresses reactor coolant system overpressure protection. As discussed in Farley FSAR Section 5.2.2.3 (Report on Overpressure Protection), analysis was performed during the initial licensing process to show the continued integrity of the reactor coolant system during the umimum transient pressure. The conclusions of this section are confismed as part of the power uprate pro,icet by analyzing the FSAR limiting transients (primarily the Loss of Loadfrutbine Trip event) and showing that the primary and secondary pressure limits continue to be met.

RCS overpressure protection under low temperature conditions is provided at Farley by the RilR relief valves. These valves have been analyzed and their capability to mitigate the cold overpressure transients has been confmned. For power uprate, none of the overpressure pump start transients (worst case mass input event) have been affected since the pumps of concem have not been changed by power uprate. 'Ihc inadvertent start of an RCP at low temperature conditions with the plant cooling down on RilR (worst case heat input event) produces a transient in w hich the stored energy in the steam generator water is transferred to the RCS. Curn:nt Farley Technical Specifications and plant procedures limit the steam generator temperature to no more than 50 F above RCS temperature. Since this limit is not changed by power uprate, the transient is not affected by the uprating.

W/rgm 7/22/97 A SCS/dmn.70I/97 8/5/97 PageI purupl8. doc j

NRChuulier1 Nom 2 The submittal indicates that the large break loss-ef-coolant accident evaluation model is being changed and that selected other new/ improved methods will be used Please give a description of all the other new or impromi methods used to suppon this license amendment and indicate whether they have reccind staff approval.

SNCJkeenscRt2 1hc power uprate project was structured consistent with the methodology established in WCAP-10263, "A Review Plan for Uprating the Licensed Power of a Pressurized Water Reactor Power Plant." Inherent in this methalology are key points that include the use of currently appromi analytical techniques (e g., methmlologies and computer cales) and the use of currently applicable liceruing criteria and standards Consistent with this methodology, the overall approach established for power uprate analyses was to use the current analysis methods except in select areas where new/ improved methods are appropriate. Using the FSAR as a reference for comparison, new or improved methods were used in the following analysis areas.

As described in Section 6.1.1.2 of the Power Uprate NSSS Licensing Repon, the large break Loss of Coolant Accident (LOCA) analysis used the 13 cst Estimate LOCA (llELOCA) methodology and the WCollRA/ FRAC computer code, This methodology was recently approved on a generic basis by the NRC and hat i en used on several plants which have made submittals to the NRC for approval. Rek.cnces are provided in Section 6.1.1.5 of the licensing report. Subsequent to submittal of the licensing report for power uprate, the NRC appromi the first time application of the llELOCA methodology to Indian Point Nuclear Generating Unit 2.

As described in Section 6.4.1.2 of the Power Uprate NSSS Licensing Report, the LOCA mass and energy (hi&E) long term releases analysis used the 1979 version of the LOCA mass and energy release model(including the 1979 ANS-5.1 standard decay heat matcl) for containment design.1his methodology was approved by the NRC in 1983 and has been used by Westinghouse and approved by the NRC on many plant specine dockets. Rcrcrences are provided in Section 6.4.3 of the licensing report.

As described in Section 6.5 2.2 of the Power Uprate NSSS Licensing Report, hiain Steamline 11reak (htSLll) ht&E releases analysis for outside containment used the 1979 ANS-5.1 standard decay heat model. This decay heat model has been previously used by Westinghouse in other analyses for Farley including the htSLil ht&E releases inside containment and has been approved by the NRC. References are provided in Section 6.5.4 of the licensing report, As described in Section 6 6.3 of the Power Uprate NSSS Licensing Report, the e

analysis for LOCA hydraulic forces used the NRC approved h1ULTIFLEX computer code which is the current Westinghouse analytical tool for use in analyzing LOCA hydraulic forces. This cale was previously used for LOCA hydraulic forces analysis as part of the Farley Unit i Up00w Conversion Project. The use of hlULTIFLEX as part of the Power Uprate Project constitutes its Srst application for Farley Unii 3 8/5/97 Page 2 imrupl8 doe

i LOCA hydraulic forces. References are provided in Section 6.6.6 of the licensing report.

The neutron Ouence analy sis at power uprate conditions was performed in accordance with the NRC approved methmlology described in Section 2.2 of WCAP 14040 NP.

A " Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS llcatup and Cooldown Limit Curves." By letter dated July 23,1997, SNC submitted a Technical Specifications amendment ruluest to tekicate the pressure temperature limit curves to the Pressure Temperature Limits Report (PTLR). Table

$.4 of the proposed PTLRs for Unit I and Unit 2 provides the reactor vessel end-of-hfe Cuer.cc (EOL) puijections at 36 effective full power years (EFPY) based on the l

methods of WCAP 14040 NP A and the fluence associated with uprated power.

%ese Huence values were used to detennine the projected LOL properties for the reactor vessel beltline materials. Additionally, the surveillance capsule withdrawal schedule provided in Table 3 1 of the proposed PTLR tc0ccts the use of the methods described in WCAP 14040 NP A at the Cuence associated with uprated power.

As described in Section 7.6.2.2 of the Power Uprate NSSS Licensing Report, the l

ORIGEN2 cale with current code libraries derived from ENDF/Il V was uscd in the I

source tenns analysis for the uprated thennal power level. This version of the code with current libraries is an updated version of the code and libraries used in the original development of source terms for Farley Updated versions of the code with libraries have been used by Westinghouse to calculate source tenns for other uprating projects (most recently Turkey Point Units 3 and 4) which have been approved by the NRC. References are providalin Section 7.6.5 of the licensing report.

As described in the Power Urate NoSS Licensing Report, analyses for LOCA hydraulic forces (for use in P.ictural analyses of the reactor internals, reactor vessel, steam generator, reactor coolant loop piping and fuel a:semblies) and LOCA mass and energy short tenn releases (for use in subcompartment structural evaluations) took credit for the leak-before break (LHB) exemption which has been previously approved by the NRC for Parley but not previously applied to most of these structural analysis arcas. References, including the NRC approval letter for the LDB exemption, are provided in Section 5.5.1 of the licensing report, The structural piping analysis for the reactor coolant loops, as described in Section 5.5 e

of the Power Up ate NSSS IJ nsing Repott, used the PS4 CAEPIPE computer code which is the current Westinght.c analytical tool for use in analyr.ing reactor coolant kop piping (i.e., this cale is used in lieu of WESTDYNE). Although this was the first Westinghouse application of this code to the Farley reactor coolant loop piping, it has previously been used on Farley by Southern Company Senices (SCS), and its use is documented in the Parley FSAR Appendix 3F," Computer Progran s Used in Structural Analyses." SCS and Westinghouse veri 6ed the cale against benchmark problems as required by the NRC.

The containment pressure and temperature analyses for the LOCA and MSLD crents, as described in Sectien 2.13 of the Power Uprate BOP Licensing Report, were perfonned using the GOTillC code. GOTillC has not been submitted for NRC K/5/97 Pagc3 pu rupt8. doc l

__..._-__..m_--__.-,

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e approval. GOTillC was developed by EPRI from the older NitC code, FATilOMS, under a fully quahned quality assurance program and has undergone extensisc peer l

review. GOTillC has bacn validated for safety.related applications at Southern l

Company Services, l

Offsite dose calculations for accident release were prepared using the multi n(xle e

TACT 5 computer code as described in Section 2.16 of the Power Uprate 110P Licensing iteport The TACT 5 code was developed by the NitC (reference NUllEO/ Cit.5106).

To evaluate the impact ofload changes due to power uprate on the Station Auxiliary Electrical Distribution Sptem, computer simulations were tun using the Station Auxiliary (STAUX) Program. The STAUX pro 6 ram provides the capability to perform comprehensive station auxiliary reviews including load now, short circuit, and mctor starting calculations at all buses, load centers, and MCCs The program uas designed to conform with all applicable industry standards, practices, and design criteria. The accuracy of the STAUX computer matel was validated by performing a test case and comparing the analytical results of the STAUX computer model with Held measurements. While the STAUX program has not been submitted for NRC approval, the program was reviewed by the NRC as part of the liDSFis at the SNC nuclear plant sites following implementation of power uprate, where applicable, the FS All wdl be revised to incorporate descriptions of these new or improved methodologies.

W/rgm.7/31/97 & SCS/ dam WO4/97 A SNC/tu-7/31/97 NRCIAtutipatNoJ The submittal mentions the boron injection tank (bit) in a few different locations. Please indicate the current function of the IllT and/or if there are plans to remove the tank.

SNC.Reputndo 3 As discussed in Section 6.0 of the Power Uprate NSSS Licensmg fleport, the power uprate analpes were pe formed to suppoit deletion and removal of the IllT. In this context, deletion signi6cs deletion of concentrated boric acid solution nom the illT (i.e., the litT remains in the piping system, but the contents of the illT are assumed to be at the same boric acid concentration as the piping system in which it is located), itemoval of the 111T signines removal of the tank from the piping system.

At the onset of the uprate analyses, the illT was scheduled for removal. As a contingency, the power uprate analyses were conservatively performed to support either con 6guration The tilt was physically bypassed (i c., removed from the ECCS pipmg sptem) during the 1997 Unit I and 1946 Unit 2 refueling outages Wirgm. 7/22/97 & SNC/nye - 7/24/97 8/5/97 page 4 pwrupl8 doc

4 s

t NRCJhntionNo 4 The specification for the charging pump discharge pressure is being reduced. Are there any circumstances where inject:an flow would be necessary or beneficial for make-up or boration (i e.,

perhaps an A'lWS event) at the pressurizer code safety valve relief pressure that would no longer be available with the new specification?

S K RopstE E L4 As part of power uprate analyses, ECCS flow analysis was redone. The ECCS flow analysis incorporated an increase in the allowable head degradation input assumption Ibr the charging pumps from 8% to 10% of design head. All of the FSAR safety analyses which use ECCS flows as an input assumption were then analyzed or evaluated to show that the associated acceptance criteria were satisfied with the revised ECCS flows at power uprate conditions. The results of the FSAR safety analyses confirmed that associated acceptance criteria were satisfied.

The increase in the allowable pump head degradation assumed in the power uprate analyses does not actually alter any pump specification but would permit a pump to remain in ser ice with a slightly reduced pump head. Should a charging pump degrade to the new 10% degradation limit, slightly less flow would be available at a given RCS pressure than for the previous 8% degradation limit. Ilowever, this slight reduction in ECCS flow has been analyzed and can be accommodated within the current analysis acceptance limits.

Whpn 7/22N7 NECRunt!cIWLi Please provide a description of the transition from (hel with zirealoy cladding to Zirlo cladding.

The submittal trierences both the topical reports for the Vantage 5 and the Vantagc+ fuel designs.

What fuel design will be used and referenced and describe how any transition core elTects will be evaluated. Please provide references for any NRC approvals related to the use of Zirlo cladding at Farley.

SNC RepomeEo. 5 The primary elTect of transitioning from fuel with zircaloy cladding to ZlRLO cladding is the impact of the ZlRLO properties on the LOCA analysis. As described in Reference 1, the VANTAGE-5 and VANTAGE + designs are mechanically and hydraulica'ly equivalent, so there are no additional transition core etTects. The Farley Units are currently operating with zirealoy clad fuel and ZlRLO clad fuel (VANTAGE +). LOCA analyses have been performed to support the use of both cladding materials. The Farley Units will initially operate at uprated power with zircaloy clad fuel and ZlRLO clad fuel until the transition to ZlRLO is completed. The LOCA analyses which were performed to support the uprated conditions address the use of both cladding materials. The non-LOCA analyses for power uprate also addressed the efTects of zircaloy cladding and ZlRLO cladding as appropriate. NRC approval for the use of ZlRLO cladding in the Parley Units was obtained in Reference 2, below.

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pwmpt8 doc

e Ermome.NoJLBrktrucc3

1. Davidson, S. L Nuhfer, D. L," VANTAGE + Fuci Assembly Reference Core Report,"

WCAP 12610 P.A, April 1995.

2.

Farley Unit i License Amendment 110 and Farley Unit 2 License Amendment 101, letter from NRC to SNC, dated September 8,1994.

W/rgm. 7/22/97 NRCttestio11.NLfi Please describe and justify the Dow streaming efTects that would permit a -l

• F bias to the temperature measurements (page 710 of topical).

SNC2rsponse No. 6 in 1991, Westinghouse identined to plants that cold leg temperature measurements with two cold leg RTDs in different circumferential kications demonstrated a temperature gradient. The cold leg temperature gradient is primarily due to difTerent lengths of steam generator tubing and resulting differences in heat transfer rates. 'llic magnitude of the temperature gradient was plant specine and was affected by: 1) the reactor coolant system (RCS) loop configuration; 2) the reactor coolant pump (RCP) model; and 3) the location of the cold leg RTD(s). Tims, with a temperature gradient in the cold leg and depending on the circumferential placement of the RTD, the cold leg RTD measurement will be either higher or lower than the bulk T-cold tempwature.

'lhe narrow range T-cold measurement is electronically combined with the T hot measurement to form Tavg. Farley uses a hiedian Signal Selector to select the median Tavg value of the three RCS hops. The median Tavg is used as an input to the Reactor (i.e., Rod) Control Sy stem to position the control rods. Therefore, a temperature gradient in the cold leg results in an increased uncertainty in the indicated Tavg of the Reactor Control System. This additional uncertainty is conservatively treated in the calculation of the Reactor Control System uncertainty and in the Farley safety analysis.

lletween June 1992 and June 1993 T-ccid data war collected on a monthly basis from Farley Units I and 2. In all three kiops of both units, readings were obtained Dom both the narrow range and wide range RTDs. A subsequent evaluation was performed by Westinghouse to determine if the cold leg streaming penalty used in the Farley uncertainty analysis was su0icient. The evaluation concluded that the cold leg streaming allowance used in the uncertainty analysis was conservative for Farley operation at 2652 htwt. core power. The cohl leg streaming penalty will remain bounding for the Farley Up ate to 2775 h1wt core power.

W/wm 7/3I/97 NB.C_Q1!f31iMMo _7 The analysis Tave window used is $67.2 577.2*F; however, the allowable window in the technical specifications is larger. Describe why the analysis window is not used in the technical specifications.

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i SNC Reemdn _2 Prior to the perfonnance of NSSS analyses for power uprate, the hSSS analyses for Farley Nuclear Plant were based on a full power design Tavg value of $77.2*F and applicabic uncertainties. %c Farley units have been operated at full power at Tavg values less than but close to this maximum design Tavg value. Ac current Farley Technical Specifications encompassed the design Tavg value as the nmimum steady state Tavg value for full power operation in DNI)

Parameters Table 3.2 1 (i.e., indicated Reactor Coolant System Tavg s $80.7'F). He Technical Specifications DNil Parameter Tavg limit includes an uncertainty allowance that is based on both the uncertaintics assumed in the NSSS analyses and the uncertainties associatal with the instrumentation used to perform the periodic Tavg surveillance.

Operation at the uprated power level will potentially result in slightly lower values of full power Tavg than operation at the current power level. To accommodate this potential reduction in full power Tavg, power uprate analyses have been perfonned for a range of full power Tavg values (i.c,,567.2*F to $77.2'F ), which will bound the full power Tavg value(s) for operation at the uprated power level. Applicable Tavg uncertainties were also included in the uprate analyses. He information provided in the Power Uprate NSSS Licensing Report describes the analyses perfonaed for the range of full power Tavg values and the assumed uncertainties. The proposed.

Technical Specifications for uprate will continue to specify a single value for indicated Reactor Coolant System Tavg (i.e., s 580.3'F), which is based on the maximum design Tavg assumed in the power uprate analyses and applicable uncertainties.

W/rgm.7/22/97 & SNC/mpe 7/3I/97 NRCRussion No. 8 ne submittal indicates that using the lowest reactor coolant system (RCS) flow is always used in the analysis. In some analysis, like the main steamline break, higher How can be more limiting.

Please describe how RCS flow ir modekxl when higher How is limiting.

SNC Ikeme No. 8 The analysis of the FSAR Chapter 15 steamline break transient in support of the Parley Nuclear Plant power uprate models minimum RCS How (%ermal Design Flow) as noted above. Reactor coolant now can affect the results of the steamline break, both directly through the DNil ratio calculations and indirectly through the system transient. The impact ofincreasing the reactor coolant flow by 10%, in both the system transient and in the DNil evaluation, was discussed generically in WCAP 9226, Revision 1," Reactor Core Response To IIxcessive Secondaiy Steam Releases " January 1,1978 (Proprietary). The results of the case assuming more reactor coolant flow were slightly better than the reference case, with the additional reactor coolant flow being a slight penalty for the system transient but a large benefit with respect to DNil. All of the non.

LOCA Chapter 15 analyses performed for thv Farley plant uprate model either the Minimum Measured Flow or Thermal Design Flow. The LOCA analyses model Thermal Design Flow. As such, no non LOCA or LOCA analyses model a maximum RCS flow,-which is consistent with the current Farley plant licensing basis analyses.

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e Power uprate evaluations for NSSS systems or con tonents used higher RCS flow (e.g., best estimate flow or mechanical design Dow) if appropriate.

W/gs.7/22/97 A W/rs 7/24/97 NRCRyestion NoJ Please provide an evaluation of your ability to shut the plant down considering all the changes to the main steam pressure, steam flow, RCS How, residual heat removal Dow, and component cooling water temperatures. On page 4-14 of the topical report, a 50'F/hr cooldown rate is assumed. Please evaluate the ability to achieve this cooldown rate for aff cvxi scenarios in the Farley Licensing Basis (i.e., single train cooldown, natural circulation cooldown, etc.).

SNC_RtentnNad Section 4.2.1.2 (page 4-14) of the Power Uprate NSSS Licensing Report describes the evaluation perfonned to assess the ability of the steam generator Atmospheric Felief Valves (ARVs) to cooldown the plant from no load (hot standby) conditions to hot shutdown conditions. This evaluation addressed the increase in NSSS thermal power and concluded that the ARVs are aiquate based on the range of operating conditions for power uprate. Consequently, the steam generator ARVs are adequate to achieve a 50 F/ hour cooldown rate from hot standby to hot shutdown as assumed in the Farley licensing basis in support of two train and single train cooldowns. Under natural circulation conditions, cooldown from hot standby to hot shutdown is limited by operating procedures to a nonnal cooldown rate less than 50*F/imur; consequently, natural circulation cooldown is not limiting relative to the adequacy of the steam generator ARVs for power uprate conditions. Note that power uprate related changes to full power main steam pressure and main steam now do not impact the ability of the ARVs to cooldown the plant from hot standby to hot shutdown conditions.

Section 41.4 of the Power Uprate NSSS Licensing Report describes the evaluation performed to assess the ability of the Reddual Ileat Removal System (RilRS) to cooldown the plant from hot shutdown to cold shutdown mdct refueling conditions. In addition to addressing the increase in NSSS thermal power, the evaluation incorporated conservative assumptions regarding the performance of the Component Cooling Water (CCW) System and the Senice Water (SW)

System. The evaluation included analyses for the two train (design oasis) cooldown and the single train cooldown scenarios as described in the Farley FSAR. For both two train and single train cooldowns, the evaluation showed that the time required to cooldown from hot shutdown (350 F) to cold shutdown (200 F) arJ/or refueling (140 F) is lengthened under power uprate conditions, liowever, the evaluation showed that for both the two train and single train cooldown scenarios, the plant possessed the ability to coc!down from 350 F to 200 F within the '.J.nical Specifications time requirement of 30-hours. The extension in cooldown times associated with power uprate conditions was identified as ar. economic consideration and not a safety-related consideration.

Furthermore, the extension in cooldown times calculated for power uprate conditions was shown to be primarily dependent on the conservative input assumptions for CCW and SW performance and not on the increase in NSSS thermal power.

Since the radiological dose analysis for several of the FSAR transients (e.g., steam generator tube rupture) include the assumption that RHRS operation can be initiated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after event J

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4 initiation, an evaluation was also perfonned to show that this assumption is valid for power uprate conditions.

Whgm. 7Dl/97 NRQusitiep Nst[0 Page 4 20 indicates that the analysis of a partial load rejection caused an oscillating plant response.

Please provide greater detail regarding the calculated results and any associated effects. Include details regarding the magnitude and length of time that the oscillations occurred S.NC._Rr_sponse No.10

• lhis question pertains to the 50% load rejection transient which was analyzed as a part of the Condition I analyses for the Farley power uprate, This transient was analyzed at full power for both high and low Tavg conditions and assumed automatic rod control and steam dump control systems. The conservative core conditions for this transient is beginning of core life (DOL), and therefore, a DOL moderator temperature coefficient (hiTC) was assumed.

The results of the 50% load rejection at high Tavg was acceptable, but at low Tavg (full power l

Tavg of 567,2'F), the results showed slightly oscillatory plant responses. The plant responses were acceptable at DOL with a full power Tavg of 570 F The oscillatory plant response at low Tavg and DOL conditions is due to the combined effect of the less negative hiTC and smaller proportional band of the steam dump loss ofload controller (6,l'F). With a smaller steam dump proportional band and rods in automatic control, the steam dump valves were demanded open'close l

for a relatively small change in the RCS temperature. The small changes in the RCS temperature are caused by the automatic action of the rod control system (rods move in or out) coupled with the hiTC cfTect. The oscillatory plant responses at these conditions were observed up to 400 seconds.

Aner 400 seconds, the plant responses were stable.

As stated in the report, an oscillatory plan response during and following a 50% load rejection transient is not due to the power uprate, tather it is due to a combined effect of h1TC, smaller steam dump proportional band, and automatic rod control system actions. The plant became stable aner 400 seconds. Therefore, these oscillations would not lead to an unsafe plant condition following a 50% load rejection transient.

W/sa - 7/22/97 NRC Ouestion No. II No methodologies are presented for many evaluations performed in the topical report Chapter 5.

Please reference the methodologies used in Chapter 5 of the topical report calculations (i.e., rod drop times, core bypass flows, and flow induced vibration).

SNC Response No.1I Using the Farley FSAR for comparison purposes, the SNC response to general question No. 2 above ducribes the areas where the NSSS analyses for power uprate used ditTerent (i.e., new or 8/5/97 Page 9 pu rupl8. doc

[4-

=

f improved) methodologies. In the other NSSS analyses areas, the power uprate analyses used the -

same basic methodologies as the analyses currently described in the Farley FSAR.-

The SNC response to additional question No 10 on page 59 provides information describing the methodologies used in the structural evaluations performed for the reactor vessel and reactor internals.

The following provides additional information regarding the methodologies used for the analysis areas (i.e., rod drop time, core bypass flows, and flow induced vibration) cited in the example to this question.

Rod Drop Times The RCCA scram performance assessment for power uprate involved the following general steps.

1.

Adjust the current analytical model (consisting of values for parameters that describe geometry of driveline components, component mechanical interaction relationships, hydraulic resistances of flow paths, RCCA/ drive rod assembly weight, and system operating conditions) to account for the new system operating conditions being considered due to power uprating.

2.

Assess the impact of such changes in primary system operating conditions on the limiting RCCA scram characteristics used in the plant accident analyses.

Core Bypass Flow For power uprate, the THRIVE computer code was used to determine the hydraulic behavior of coolant flow within the reactor internals system (i.e., vessel pressure drops, core bypass flows, RPV fluid temperatures and hydraulic lift forces) by solving the mass and energy balances for the

- Farley Nuclear Plant reactor internals fluid system.

Bypass flow is the total amount of reactor coolant flow bypassing the core region and is not considered effective in the core heat transfer process.

The principal core bypass flow paths are:

1.-

Baffic/ Barrel Region; 2.

Vessel Head Cooling Spray Nozzles; 3.

Core Barrel - Reactor Vessel Outlet Nozzle Gap; 4.

Fuel Assembly - Baffle Plate Cavity Gap; and -

5.

Fuel AssemblyThimbleTubes.

Fuel assembly hydraulic characteristics and system parameters, such as inlet temperature, reactor coolant pressure and flow, were used in conjunction with the THRIVE code to determine the impact of the new uprated conditions on the total core bypass flow.

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t Flow Induced Vibration Flow-induced vibrations (FIV) of pressurized water reactor internals has e included in-plant tests, scale model tests, as well as tests in fabricators' shop and bench tests of components, along with various analytical investigations. For power uprate, the vibration response of the Farley reactor internals were obtained using the principle of dimensional analysis and scaling laws. The test results of a 3 loop plant, similar in design to the Farley units, were scaled to the Farley uprated operating parameters.

W/rgm - 7/31/97 N11C_ Question No.12 Please provide a reference for the NRC approval of the use of the Westinghouse revised thermal design procedure at Farley and discuss how the transition core effects will be addressed with this thermal design approach.

SNC Rc.sponse No.12 The use of the Westinghouse revised thermal design procedure was approved for use at Farley as part of the approval for implementation of VANTAGE-5 fuel (Reference 1). The transition core effects for the VANTAGE-5 fuel were addressed by the NRC approved methodology specified in References 2 through 6. These references are applicable for the uprating.

Btsponse No.12 Referenen 1.

Letter from S. T. Iloffman (NRC) to W. G. Hairston 111 dated h1 arch 11,1992, regarding Farley Unit i License Amendment 92 and Farley Unit 2 License Amendment 85.

2. Davidson, S. L., and lorii, J. A., " Reference Core Report - 17xl? Optimized Fuel As:,einbly," WCAP-9500-A, hlay 1982.

3.

Letter from E. P. Rahe (Westinghouse) to Miller (NRC), dated March 19,1982, NS-EPR-2573, WCAP-9500 and WCAPs-9401/9402, "NRC SER Mixed Core Compatibility items."

4.

Letter from C. O. Thomas (NRC) to E. P. Rahe (Westinghouse), " Supplemental Acceptance No. 2 for Referencing Topical Report WCAP-9500," January 1983.

5. Schueren, P. and McAtee, K. R., " Extension of Methodology for Calculating Transition Core DNBR Penalties," WCAP-11837-P-A, January 1990.

6.

Letter from S. R. Tritch (Westinghouse) to R. C. Jones (NRC) " VANTAGE 5 DNB Transition Core Effects." ET-NRC-91-3618, September 1991.

W/rgm 7/22/97 NBC_QuCE. tion No,13 is Southern Nuclear requesting staff approval of the moderator temperature coefficient limit curve presented in Chapter 7 of the submittal (Figure 7.2-1) or merely showing the currently approved limit curve?

8/5/97 Page1I pwrup18. doc

SNClopense No.13 The moderator temperature coefEcient limit curve presented in Figure 7.2-1 is the currently approved limit for the Farley Units. His limit was approved as pan of the approval for VANTAGE 5 fuel (Reference 1).

RopensdoJ1Refersaces

1. Letter from S. T. Iloffman (NRC) to W. G. Ilairston til (SNC) dated March 11,1992, regarding Failey Unit i License Amendment 92 and Farley Unit 2 License Amendment 85.

Whpn 7/22/97 NRC Ouction No,14 Chapter 7.3 presents a number of fuct rod design acceptance limits. For each, please describe w here the limit is derived or referenced and if the limit has been accepted by the NRC generically or for Farley specifically.

l SM'Ennens3 No.14 The acceptance limits for the key fuel rod design criteria presented in Chapter 7.3 were established in the following references.

l Rod Intemal Pressure - The NRC-approved rod internal pressure acceptance limit was generically defined in WCAP-8963 P-A, Reference 1. This acceptance limit has been applied and approved for subsequent generic topicals addressing extended bumup with zircaloy cladding (WCAP-10125-P-A, Reference 2) and VANTAGE + (WCAP-12610-P-A, Reference 3).

Clad Corrosion - The NRC-approved acceptance hmits for clad temperature (metal oxide interface temperature) and hydrogen pickup were generically dermed for zircaloy in WCAP-10125-P-A, Reference 2. The NRC-approved acceptance limits for clad temperature (metal oxide interface temperature) and hydrogen pickup were generically defined for ZIRLO in WCAP 12610-P A, Reference 3.

Clad Stress and Strain - The NRC-approved acceptance hmits for fuel rod clad stress and strain were generically defined for zircaloy in WCAP-10125-P-A, Reference 2 and for ZlRLO in WCAP-12610-P-A, Refesence 3.

Besnonse No.14 Refersnqqs

1. Risher, D. H. (Editor), " Safety Analysis for the Revised Fuel Rod internal Pressure Design Basis,"WCAP-8963-P A, August 1978.

2.

Davidson, S. L. (Ed.), et al., " Extended Burnup Evaluation of Westinghouse Fuel,"

WCAP-10125 P-A, December 1985.

8/5/97 Page 12 pwrupl8. doc I

1

1 s

I.

1 l

3. Davidson, S. L., Nuhfer, D. L., " VANTAGE + Fuel Assembly Reference Core Report,"

WCAP-12610-P A, April 1995.

W/rpn - 7/22/97 NRC Ounjion No.15 Please verify that the fluence value used to support the technical specification pressure / temperature limit curves (effective through 16 and 14 effective full power years for Units 1 and 2, respectively) will not be exceeded at the higher full power limit.

SNC Rupsnse No.15 The results of the analyses performed to assess the impact of power uprate on reactor vessel integrity are summarized in Section 5.1.2 of the Power Uprate NSSS Licensing Report. As stated in this seciion, new heatup and cooldown curves were calculated for 36 EFPY at the new uprated power conditions. These calculations included consideration of the increased neutron fluence due to power uprate. The revised heatup and cooldown curves are included in the proposed Unit I and Unit 2 Pressure Temperature Limits Repoits (PTLRs) submitted to the NRC by keer dated July 23,1997. NRC approval of the PTLRs is required prior to startup to support implementation of power uprate.

SNC/tws 7/24/97 8/5/97 Page 13 purupl8. doc

s -_

n QUESTIONS REGARDING COh1PLIANCE WITil 10 CFR PART 50, AFPENDIX G, AND 10 CFR PART50, APPENDIX 11 NRC Ountion No. I Provide the projected maximum end of-life (EOL) fluences at the inner diameter of the Joseph ht.

Farley Nuclear Plant (Farley) reactor pressure vessels (RPVs) based on the new uprated power conditions and the revised adjusted reference temperature values for the Farley Units I and 2 RPV beltline materials.

SNC Resp 9me No.1 Consistent with Paragraph $.l.2 of the Power Uprate NSSS Licensing Report, revised heatup and cooldown curves have been calculated for 36 EFPY at power uprate conditions. The naised

-- heatup and cooldown limits and related information are included in the Technical Specifications amendment request associated with the PTLR which was provided to the NRC by letter dated July 23,1997, The maximum ond-of-life fluences at the inner diameter of the Farley RPVs, based on

- the new uprated power, wem provided in Table 5-4 of the proposed Unit I and Unit 2 PTLRs. The corresponding adjusted reference temperatures associated with the new uprated power are provided in Table 5-5 of the proposed Unit I and Unit 2 PTLRs.

SNChws 7/24/97 NRC Ouestion No. 2 Provide an assessment of how the proposed power uprate will affect the current pressure-

. temperature (P T) limit curves in the Farley Unit I and Unit 2 Technical Specifications. If the uprated power conditions will change (increase) the adjusted reference temperatures for the most

- limiting beltline materials in the Parley RPVs, new P-T limit curves should be submitted based on the new uprated conditions and fluences.

SNC Response No. -2 Revised pressure-temperature limit curves valid to 36 EFPY, based on the new uprated power, were provided in Figures 2-1 and 2-2 of the proposed Unit I and Unit 2 PTLRs.

SNCAws 7/24/97 NRC Ouestion No. 3 Provide an assessment of how the proposed thermal uprate will affect the EOL upper-shelf energies

- for the Farley Units I and 2 RPV beltline materials. Include appropriate calculations and figures based on the guidelines of Regulatory Guide 1.99, Rev. 2, " Radiation Embrittlement of Reactor Vessel hiaterial," dated hiay 1988 SNC Resoonse No. 3 The EOL Upper Shelf Energies (USE) for Farley Units 1 and 2 beltline materials, based on the fluence associated with the new uprated power, were determined using the methods described in

~

8/5/97 Page 14 pwrupl8.dx I

I Regulatory Guide 1.99, Rev. 2. As shown in Tables A and D, which follow this section, the USE projected at EOL (36 EFPY) are greater than 50 ft lb and continue to meet the twiuirements of 10 CFR 50, Appendix G.

SNC/tws. 7/24/97 NRC OusslinrtNg 4

- Will the resised neutron fluences as a result of the uprated conditions alTect the surveillance capsule withdrawal schedule for the Farley Units I and 2 RPVs?

SEC_Essponte No 4 The surveillance capsule withdrawal schedules have been revised to reflect the increased fluence associated with the new uprated power using the NRC-approved methods described in WCAP-14040-NP-A. The revised surveillance capsule withdrawal schedules were provided in Table 3-1 of the proposed Unit I and Unit 2 PTLRs.

SNC/tws - 7/25/97 NB1 Question No. 5 The staffis providing copies of the Pressurized Thermal Shock (PTS) Summary Files and the Upper Shelf Energy (USE) Sununary Files for the Farley Units 1 and 2 RPV beltline materials, as obtained from the NRC Reactor Vessel Integrity Database (RVID), Version 2.0.2 Update the Summary Files to the extent possible based on the most current data for the Parley RPVs, and using the uprated fluence values for the plants. The updated Sununary Files may be used to assist you in your responses to items 1. - 3. listed above.

SNC Response No 5 SNC has reviewed the RVID PTS and USE Summary Files provided by NRC letter dated July 24, 1997. Attachment 11 provides the requested information based on the most current data for the -

Farley RPVs using the uprated fluence values.

8/5/97 Page 15 pwrupl8. doc j

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TABLE A Predicted End of License (36 EFPY) Upper Shelf Energy Values for the Farley Unit 1 Reactor Vessel Beltline Materials Beltlinc Material.

Wt %.

- 1/4T Fluence

Unirradiated Decrease -

Projected Cu (n/cm')

USE in USE ')

EOL USE -

l Inter. ShcIl Plate H6903 2 0.13 1 2.48 x 10

99 n lb

- 27 n-lb

- 72 n lb Inter. Shell Plate B6903 3 _

0.12

- 2,48 x 10

87 ft-lb - '

23 ft lb 64 ft lb1

=- Lower ShcIl Plate B69191 0.14-2,48 x 10

86 ft lb -

- 24 ft-lb 62 ft lb -

2.48 x 10" 86 ft lb 20 ft lb 66 ft lb using S/C data

< Lower Shell Plate B6919 2 0.14 2.48 x 10

86 ft Ib

. 24 ft lb 62 ft-lb -

. Inter. Shell Longitudinal Weld 0.24 7,67 x 10'8 149 Q-lb 54 ft lb 95 ft lb Scams19-894 A & B

-(IIcat # 33A277) using S/C data 7.67 x 10

149 ft lb 34 ft-lb i15 ft lb f

Circumferential Weld Il 894 -

0,21

'2.48 x 10

104 fi lb 46 fi lb

$8 ft lb -

(llcat # 6329637)

Lower Shell Longitudinal Weld 0.20 7.67 x 10 -

82.5 ft lb 27 ft lb 55 ft lb Scams20-894 A & B (llcat # 90099)

Lgqq a) Per Regulatory Guide 1.99, Revision 2.

SNC/tws - 7/24/97 8/5/97 Page 16 pwrupl8. doc

TABLEB Predicted End of License (36 EFPY) Upper Shelf Energy Values for the Farley Unit 2 Reactor Vessel Beltline Materials f

Beltline Material W1. %

1/4T Fluence Unitradiated Decrease Projected

-=

Cu (n/cm )

USE in USE*

EOL USE 2

Inter. Shell Plate B7203-1 0.14 2.34 X 10" 100 ft-lb Inter. Shell Plate B7212-1 28 n lb 72 fl-lb 0.20 2.34 X 10" 100 ft-lb 35 f1-lb 65 n-Ib using S/C data 2.34 X 10" 100 ft-lb Lower ShcIl Plate B7210-1 39 n lb 61 ft lb 0.13 2.34 X 10" 103 ft-lb Lower Shell Plate B7210-2 28 ft lb 75 n-lb 0.14 2.34 X 10" 99 ft-lb 28 ft-lb 7I ft lb Inter. Shell Longitudinal Weld 0.02 Scam 19-923 A 7.48 X 10" 131 A-lb 23 ft lb 108 ft-lb (lleat # llODA)

Inter. Shell Longitudinal Wcld 0.03 Scam 19-923 B 7.48 X 10" 148 ft-lb 26 ft-lb 122 ft-lb (llcat # BOLA) using S/C data i

7.48 X 10" 148 ft lb 13 n lb 135 ft-lb Circumferential Weld Il 923 0.14 (llcat # SP5622) 2.34 X 10" 10211 lb 35 ft-lb 67 ft-lb Louct Shell Longitudinal Weld 0.05 Scams20-923 A & B 7.48 X 10" 126 ft-lb 23 ft lb 103 ft-lb (IIcat # 83640)

NOTEA a) Per Regula<ory Guide 1.99, Revision 2.

l SNC/tws - 7/2.IN7 I

l

\\

8/5/97 Page 17 pwrupl8. doc

i 0

QUESTIONS REGARDING STEAM GENERATOR INTEGRITY NRC_QtaslionELI Summarize the results of the assessment that evaluars the effect of the power uprate on (1) the

{

minimum wall thickness of steam generator tubes, (2) the number of steam generator tubes l

susceptible to anti vibration bar wear, and (3) susceptibility of the steam generator tubing to j

various forms of degradation mechanismse SN_C_ Response No.1 The results of the assessments of the impact of power uprate on these areas are summarized below.

I

1. The minimum required wall thicknese, following the guidance of Regulatory Guide 1.121, was determined to be 0.022 inches (44% of wall thickness) based on maintaining a safety margin of 3 against burst during normal operations. This minimum wall thickness is acceptable for meeting the loading criteria of Regulatory Guide 1.121, including a postulated accident concurrent with an SSE.

2. The anti-vibration bars (AVil) have been replaced on Farley Units I and 2. Since the AVil replacement, AVil wear has not been observed as an active degradation mechanism at Farley. Power uprate is not expected to introduce AVI3 wear.

3.

The susceptibility of steam generator tubing to various fonns of degradation is described in Section 5.7.4 of the Power Uprate NSSS Licensing Report with additional infonnation provided in response to questions wgarding steam generator integrity Nos.

5 and 6 ubich follow. The conclusions of the tube degradation evaluation is that power uprate will not have a dgnincant impact on the susceptibility of steam generator tubes to various forms ofdegradation (ODSCC and PWSCC). As described in Section 5.7.4, power uprate will not have a significant impact on T-hot, the most important parameter with respect to steam generator tube degradation. Consequently, 3

a signi6 cant increase in steam generator tube degradation due to power uprate is not i

expected.

W/gw & rs 7/31/97 A SNC/ rem - 7/1t/97 NRXQucstion No. 2 It is not clear to the staff whether the Southern Nuclear Operating Company, Inc. (SNC), has assessed the structural integrity of the Farley steam generator tubing under uprated power conditions in accordance with Regulatory Guide 1.121 methodology. Clarify and provide the basis for your conclusions.

8/5/97 Page 18 pwrupl8. doc

e SNC Resan_se No. 2 The structural integrity of the SG tubes was evaluated and determined to be acceptable using Regulatory Guide 1.121 methodology Evaluations were performed for minimum wall thickness (as discussed in the response to part (1) of question No. I above) under the loading conditions prescribed by Regulatory Guide 1.121.

W/gw A rs 7/31/97 A SNC/ rem 7/31/97 NRC Ouenion No. 3 Clarify whether SNC has considered performing any additional surveillance methods to monitor for changes in steam generator degradation as a result of the uprated power conditions. Provide the basis for your conclusions.

SEC Response No. 3 Prior to each steam generator inspection, an assessment is made as to what degradation mechanisms are active in the Farley steam generators and in similar steam generators throughout the industry. Inspection plans are developed which will ensure adequate detection ability for the degradation mechanism in the affected location in the steam generator Consequently, any new degradation mechanisms and any significant increase in degradation rate should be detectable by l

the planned steam generator inspections.

The repair criteria contained in the Technical Specifications will continue to apply to power uprate conditions with the exception of the Unit 2 criteria for F* The F' distance will be revised from 1.54 inches to 1.6 inches as a result of the increased normal ditierential pressure between t e h

primary and secondary (as described in the response to question No. 8 below regarding steam generator integrity). Although a 40% repair criteria does exist in the Farley Technical Specifications, Farley does not use the 40% repair criteria unless a qualified sizing technique exists for the mechanism of concern. The voltage-based alternate repair criteria will continue to be used at Farley in accordance with current Technical Specifications and Generic Letter 95 55.

W/gw & rs - 7/31/97 & SNC/ rem 7/31/97 NRC Ouggion No. 4 Section 5.7.1 discussed the structural evaluation of steam generator intemals. Provide a list of components that were evaluated and results of the evaluation.

SEC Resnonse Np 4 The steam generator components which were evaluated for their structural adequacy at the power uprate conditions were: the tubc/tubesheet weld; tubes; channel head /tubesheet junction; tubesheet/shelljunction; divider plate; feedwater nozzle; secondary manway opening and bolts; and steam nozzle. With the exception of the secondary manway bolts, the structural analysis showed that all of the components experienced maximum stresses and fatigue usage factors less than the allowable limits. In terms of maximum stresses, the tube /tubesheet weld yielded the greatest stress compared to the allowable limit, with the ratio of calculated / limit being 0.976. With regard to the 8/5/97 Page 19 pwrupl8. doc

""L J

'l fatigue usage, the secondary manway bolts had a calculated fatigue usage of 1.18 at 40 years it was determined that, in order to obtain an acceptable fatigue usage value, the bolts would need to I ' replaced before the 34* year of operation. SNC plans on replacing the bolts which have not already been replaced to comply with this requirement. The d vider plate had the second highest fatigue usage, with a value of 0.944 under normal and upset conditions. Additional infoimation regarding the structural evaluation performed for the steam generator internals is provided in the St4C response to additional question No. 7 below.

W4s 7/22/97 NRC Ouggion No. 5 Section 5.7.3 discussed the fatigue evaluation of U-bends from a fluid vibration viewpoint. It is not clear to the staff whether SNC has evaluated the small radius (rows I and 2) U-bends for degradation from stress corrosion cracking. Clarify and provide the basis for your conclusions.

i SNC Rusa.9mg No. 5 l

The evaluations perfonned to assess the impact of power uprate on stress corrosion cracking are described in Section 5.7.4 and included an assessment of the impact of PWSCC, including the effects of power uprate on the kinetics of PWSCC for the steam generator heat transfer tubing. At l

both Farley Unit ! and Unit 2, the small radius U-bends were given a thermal stress relief to reduce the residual manufacturing stresses. As described in Section 5.7.4, the only stress that is efTected by the power uprate conditions is the throughwall pressure stress which increases moderately due 1

to the increase in normal primary to-secondary AP from approximately 1435 psi to 1463 psi. In combination with the reduction in residual stress, the modest increase in throughwall pressure stress due to power uprate is negligible in terms of enhaucing the initiation and propagation of PWSCC in the small radius U-bends.

In order to ensure stress corrosion cracking has not incressed in the row I and 2 U-bends, all row I and 2 U-bends will be inspected at the refueling outage following implementation of power uprate.

W/gw A rs 7/31/97 & SNC! rem. 7/31/97 NRC Ouesion No 6 Section 5.7.4,2 stated that the power uprate will not significantly affect outside diameter stress corrosion cracking (ODSCC). Clarify which regions of the steam generator tubes were assessed with respect to ODSCC, including whether the power uprate would affect ODSCC at tube support plates. Provide the basis for your conclusions.

SNC Response No. 6

- The impact of changes to primary and secondary side pressures and temperatures due to power uprate was evaluated with respect to corrosion in the tube support plate (TSP) crevices, corrosion within the sludge pile (SP) at the top of the tubesheet, and corrosion on the tubing free span (FS).

The beneficial effect oflowering secondary temperature tends to be stronger than the deleterious effect ofincreasing the applied stress in the tube support plate (TSP) crevices and on tubing free 8/5/97 Page 20 pwrup18. doc

spans (FS). No credit is taken for the lower secondary temperatures within the sludge pile (SP);

hence, a small increase in the expected ODSCC rate in the SP is predicted. Ilowever, the SP -

region in all Farley steam generators is routinely inspected during each refueling outage. Any significant increase in the magnitude or rate of degradation should be readily detectable. 'Ihere is little difference in the predictions for the two units.

W/gw & rs - 7/31/97 & SNC/ rem 7/31/97 NRC_Qnestion No. 7 SNC has implemented voltage based alternate tube repair criteria in the Technical Specifications for Farley Units 1 and 2. Discuss whether the uprated power conditions would afTect the structural and leakage analyses that are recommended in Generic Letter 95-05. Provide the basis for your conclusions.

SNC Response No. 7 As discussed in Section 5.7.5, the impact of power uprate on the repair criteria contained in the Technical Specifications, including the voltage-based alternate repair criteria (ARC) for the tube support plate intersecticas, was evaluated. The evaluation showed that the 2 volt alternate repair criteria value in the Technical Specifications is not sensitive to the increase in nonnal primary-to-secondary AP associcted with power uprate since it is a limit set by Generic Letter 95-05 based on steam generator tube size. Furthermore, the upper voltage repair limit, determined specifically for each operating cycle, ;s based on a safety margin of 1.4 X steamline break differential pressure.

The steamline break difTerential pressure is based on "an assumed difTerential pressure across the tube walls equal to the pressurizer safety valve steeping plus 3 percent for the valve accumulation, less atmospheric pressure in faulted steam generators," per Generic Letter 95-05. Since the steamline break difTerential pressure is not being changed due to power uprate, the upper voltage repair limit is not directly afTected by power uprate, llowever, a change to the upper voltage repair limit may be required due to possible changes in the structural limit or flaw growth rate. The structural limit may change due to changes in the NRC database for establishing the voltage corresponding to the tube structural limit. As stated earlier, the upper voltage repair limit is determined for each individual operating cycle.

W/gw & rs 7/3I/97 & SNC/ rem - 7/31/97

- NRC Ouestion No. 8 Section 5.7.5 stated that an analysis was performed to revise the F* criteria in the Farley Unit 2 Technical Specifications to bound the best estimate steam generator outlet pressure at 2785 MWt.

It is not clear to the staff whether SNC will submit for staff review a license amendment to revise the F* criteria specified in the Farley Unit 2 Technical Specifications. Please clarify.

SNC Response No 8 The change to the Farley Unit 2 Technical Specification F* criteria is included in the proposed Technical Specifications changes for power uprate. Specifically, Technical Specification 3/4.4.6 value for "F* Distance" has been revised from 1,54 inches to 1.6 inches.

W/rs - 7/22/97 8/5!97 Page 21 pwrupl8 doc 1

QUESTIONS REGARDING ATTACilMENT 6, SECTION 2.15 - SAFETY.RELATED ELECTRICAL EQUIPMENT QUALIFICATION NRC Ouestion No. I Provide the list of required and qualified radiological doses of the individual safety related electrical equipment before and afler power uprate. In your submittal, it is stated that "for safety, related electrical equipment with uprate doses not bounded by the original design basis, radiological doses at uptate conditions were compared against the dose threshold limits used for the individual components or equipment." We believe that the doses should be bounded by the test report values, not by the dose threshold limits. Explain the difTerences and why your method is acceptable.

SNC.Acm.osc_NL1 In the context of the EQ evaluation prepared for power uprate, the terminology of dose threshold limits and test report values means the same. The limiting qualification radiation dose for each room was determined from the System Component Eva!uation Worksheet for each Environmental Qualification Package which documents the qualification or tested radiation dose.

IWk-7/16/97 NRC Qucation No,2 Furnish composite loss-of-coolant accident / main steamline break containment temperature profiles before and after power uprate case on the same plot that extends to 30 days. Identify where the composite temperature power uprate profiles'are not enveloped by the design basis profile.

SNC Response No 2 The power uprate composite temperature profile was superimposed on the existing composite temperature profile. This plot, "FNP Composite LOCA/MSLB Containment Temperature Profile," is attached to the end of this section. Differences between the power uprate composite profile and the existing composite profile are discussed in detail in response to question 3 below.

IWk - 7/31/97 NRC Ouestion No. 3 Explain why the (power) uprated temperature that exceeds the existing design basis profile by a few degrees (i.e.,5 F) toward the end of the composite temperature profiles (greater than 30,000 seconds)is acceptable by having enough margin between 70 seconds and 10,000 seconds. Should the end of the composite temperature profiles be longer or shorter than 30,000 seconds (8.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />)?

SNC Response No 3 The powc. uprate composite temperature profile was compared to the existing composite temperature profile and to the applicable equipment qualification test profiles. The review 8/5/97 12 age 22 pwrupl8. doc

c 4

concluded that the power uprate composite temperature profile will not have any signi6 cant impact to the environmental quali6 cation of the EQ components at FNP.

As is customary, the 30-day coinposite temperature profiles (pour uprate and existing) were plotted on a semi-log graph. The pIot, "FNP Composite LOCA/MSLB Containment Temperature l

profile," is attached. For discussion purposes, the composite temperature pro 61e has been divided into three sections. Section 1 is the initial 150 seconds of the postulated accident; Section 2 is from 150 seconds to 7000 seconds; and, Section 3 is from 7000 seconds to 30 days. Ilowever, to aid in viewing, Sections I and 2 were also plotted on a linear scale frorn 0 to 7000 seconds; this plot, "FNP Composite LOCA/MSLB Containment Temperature Profile (0 - 7000 seconds)."is also attached. Each plot section is discussed telow.

Section I is the temperature ramp-up portion of the pro 61c. This ramp-up is due to the postulated main steam line breaks and is for a short duration (150 sec.). Inspection of the ramp-up section of plot with linear scale suggests that the existing ramp-up and the power uprate ramp-up would lead to similar heat transfer to the equipment within the containment. The ramp-up results in exposing the equipment to a high-temperature for a short duration. Since the equipment mass does not heat up instantaneously due to thermal transfer from the environment to the equipment surface, the equipment does not reach thermal equilibrium for short-duration events. Based on engineering experience with transient thermal he.at transfer analysis (thermal lag analysis), de sho t-duration ambient temperature excursion is covered by the existing test data. Herefore, the initial power uprate temperature ramp-up is enveloped by the applicable equipment qualification test data.

For Section 2, the power uprate composite temperature profile is enveloped by the existing composite temperature profile.

Although the power uprate composite temperature pro 61e in Section 3 exceeds the existing composite temperature pro 61e by approximately $*F, a review of the test profiles for EQ equipment inside containment indicates that there is sufficient margin in the test pro 61es to envelop the power uprate composite temperature profile. In addition, the EQ equipment has been quali6ed for the peak temperature of 38 PF which exceeds the power uprate peak temperature of 383 F.

Further, the duration at the higher temperatures (i.e., >250 F) is longer for the existing pro 61e than for the power uprate pro 6le.

Based on the above discussion, the FNP designers concluded that the existing equipment quali6 cation was not impacted by the power uprate composite pro 6te.

IWk Ajl.7/31/97 8/5/97 Page 23 pwrupl8. doc

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QUESTIONS REGARDING ATTAClkhtENT 6, SECTION 2.20 - hilSCELLANEOUS ELECTRICAL REVIEWS -

NRC Quotion No.l1; Proside the impact of the load, voltage, and short circuit values for power uprate conditions at all levels of the station auxiliary electrical distribution system (i.e., the onsite power system, the main generator, and its step up transfonner).

SNC Rnponse No 1 As briefly discussed in HOP Licensing Report, Section 20,0,"htiscellaneous Electrical Reviews,"

the plant electrical distribution system was evaluated for potential impact associated with the Farley power uprating. Additional information pertaining to this engineering evaluation is

_ presented herein; in order to support power uprate, the Reactor Coolant Pump (RCP) motors and Condensate Pump motors will be required to deliver slightly higher horsepower (llP) over their current operating

. values. The additional HP for the RCP motors is required to support the uprate RCS Tavg temperature range and the reduced RCS loop flow rates assumed for full power operation. The additional llP requirement for the Condensate motors is due to the increased feedwater How rate.

There are no other electrical load changes on the plant electrical distribution system as a result of power uprate.

. To evaluate the impact of the RCP and Condensate Pump HP increases on the Farley Electrical Distribution System, computer simulations were performed using the Southern Company Services (SCS) Station Auxiliary (STAUX) Program. The STAUX program provides load flow, voltage, and short circuit values at all buses, load centers, and hiCCs.

Impact to 4160V System Imnact to Loading - The increase in loading on the non-1E 4160V buses as a result of power uprate represents only 2% or less of the total loading on each 4160V bus. There is no impact on the Class IE 4160V buses. The total loading after power uprate does not execed the continuous current ratings of the breakers and transformers.

Voltage ImpAcj - Steady state and starting voltages did not decrease more than 0.4% as a result of power uprate.

- Impact to Short Circuit Valugs - No change in short circuit values occurred on the 4160V bus as a result of power uprate.

' Impact to 600V System impact to Loading - No change in loading on the 600V buses is required as a result of power uprate.

Voltage impact - Steady state and starting voltages did not decrease more than 0.4% as a result of power uprate.

8/5/97 Page 26 pwrupl8. doc

D Impagtto Short CircuiLVahtes No change in short circuit values occurred on the 600V bus as a result of power uprate.

Impact to 208/120V System impacticl&Miing - No change in loading on the 208/120V buses are required as a result of power uprate.

Voltage impas.1 - Steady state and starting voltages did not decrease more than 0.4% as a result of power uprate, impact to Short Circuit Values - No change in short circuit values occurred on the 208/120V bus as a result of power uprate.

Impact to Main Generator lomilnJnacj - The generator capability was reviewed to evaluate the impact ofincreased generator load for power uprate. A detailed discussion of this review is provided in the BOP Licensing Report, Section 2.6, " Main Generator and Auxiliaries." This review confirmed that the generator is capable of operation at the specified uprated MW value.

.Vojtagejmpact - Generator bus voltage range is maintained within a range of 95% - 105%

of generator rated voltage (22kV) by design. The actual bus voltage is dictated by system conditions and will not be significantly impacted by power uprate. Calculations perfonned in support of power uprate show that, for the generator minimum voltage of 95%, the Reactor Coolant Pump buses A, B, and C (which are normally powered from the unit auxiliary transformers) have sufficient voltage in addition, the bus loads on buses A, B, and C will not be subjected to unacceptably high voltages when the generator voltage is at 105%.

Short Circuit impact - Calculated short circuit values on the generator bus do not change as a result of power uprate.

Impact to Generator Step-Up Transfonner

1. cad impa_gt - The increase in MW loading is within the design temperature and load ratings of the transfonner, although operating temperature may increase.

Voltage impact - The additional MW output from the generator will increase the voltage drop through the main power transfonner, The actual high (transmission) side winding voltage is set by the system voltage schedule. The low (generator) side winding voltage is adjusted to maintain the high side voltage schedule and kept within operating and equipment limits and therefore is not significantly impacted as a result of power uprate.

- Short Circuit Imnact - Calculated short circuit values on the generator hus and the 230kV and 500kV buses do not change as a result of power uprate.

SCS/wb A tle & jnn-7/31/97 8/5/97 Page 27 pwrupl8 doc

NBCQuctigetNo 2 Provide the result of an analysis which was used to conclude that: (1) the bounding steady-state voltages and motor starting voltages remain within acceptable limits, (2) emergency diesel generator loadings are within the design ratings, and (3) there are no impacts on relay trip set points for loss of voltage or degraded grid voltage protective scheme due to power uprate.

SEC_Enpsnic No,2 To evaluate the impact on the Farley Electrical Distribution System, computer simulations were pe formed using the SCS STAUX Program. The STAUX program provides load Dow, steady state voltage, and short circuit values for all buses, load centers, MCCs and selected equipment. It also provides motor starting voltage dip for selected motors.

(1) The steady-state and motor starting voltages did not decrease more than 0.4%. These voltages are within the minimum required voltages for plant equipment important to safety.

Important non-Class !E equipment voltages were also verified to remain within acceptable limits.

(2) The Farley Diesel Generator (DG) load calculation was reviewed to determine the potential impact of pmver uprate on the ability of the DGs to perform their safety-related function; in addition, the impact of power uprate on the ability of the DGs to perform their safety-related function under Station Blackout (SBO) scenarios was also evaluated. No requirements to add or change out safety-related plant equipment or to increase loading of existing safety-related plant equipment beyond the equipment ratings already analyzed in the DG load calculation were identined. The engineering reviews determined that the DGs are not impacted by power uprate and that the DGs will continue to perform their intended safety-related function.

(3) The increased electrical load (i.e., for the RCP and Condensate Pump motors) associated with the Parley power uprate occurs only on non-lE 4160V buses. The corresponding load increase to each startup transformer represents a very small percentage of the transfonner total load rating. As a result, the additional voltage drop through the transformer is very small (< 0A%), and therefore, Class IE bus voltages are not signincantly impacted. This small change is not sulTicient to impact the voltage setpoints of the Farley LOSP and degraded grid protection scheme.

SCS/wb & tic & jms-7/31/97 NRC Ouestion Nol State what would be the negative impact on the stability of the units by increasing Farley generation to 920 hlWe per unit.

SNC Response No. 3 The increase in power does not impact the stable operation of the Faricy Units for expected design basis conditions. Under normal expected operating conditions, the Farley Units are stable and 8/5/97 Page 28 pwrupl8 doc

safety related buses will continue to be supplied by the off site preferred power source for single contingency events and faults, in general, as the megawatt loading of a generating Unit increases for a given system load level, the margin of stability for that Unit will decrease. Thus, for abnormal system alignments such as an outage of a critical transmission line, increasing the MW output of the Units results in a slight decrease in the margin of stability at a given system load lesel. Load limitations with a transmission line out of senice are currently addressed by plant operating procedures with consideration given to how long the line will be out of service, system load requirements, and operational status of the Units.

SCS/wb & tle A jms-7/31/97 MRCRuntion No. 4 Clarify the statement, "There is a slight decrease in the margin of stability for limited faults during salley load conditions. Normal system growth offsets the slight decrease in margin of stability within 3 to 5 years." Please elaborate on how the generation increase due to its power uprate will decrease the stability margin, but the stability will improve later on when the system load grows, l

SNC lluponse NRJ As discussed in the response to miscellaneous electrical question No. 3 above, in general, as the megawatt loading of a generating Unit increases for a given system load lesel, the margin of stability for that Unit will decrease. 'Ihis is because, as the unit load increases, the Unit will tend to become less stable with respect to the system. Increasing the Farley generation to 920htW results in a slight decrease in the margin of stability for any given system load level when compared to the existing generation at Far!ey, On this same basis, as the system load increases (for a given Unit output), the Unit will tend to become more stable with respect to the system.

1hus, normal system load growth will have a favorable impact on the stability margin for the Farley Units resulting in an increase in the stability margin.

SCS/wb A ttc &jms-7/31/97 N/5/97 Page 29 pu rupl8 doc

4 QUESTIONS REGARDING ATTACilhtENT 6, SECTION 2 - HALANCE OF PLANT PROGRAh!

DESCRIPTION NRC Question No.1 The increase in the probability of turbine overspeed and associated turbine missile production due to plant operations at the proposed uprated power level have not been addressed. Please demonstrate that plant operations at the proposed uprated power level will not increase the probability of turbine overspeed and associated turbine missile production.

SNC Response No.1 The degree of overspeed protection for the turbine is a function of the entrapped energy at the time of trip, the system design and the turbine speed at the time when the trip is initiated, if the anal speed of the turbine fc! lowing an overspeed trip does not exceed the design overspeed, there is no increased probability of missile production.

Governor and interceptor valves throttle closed when turbine speed is 2103% of rated speed and may reopen to maintain rated speed. At 111% of rated speed the mechanical overspeed mechanism functions to trip the turbine.

The design overspeed trip points were set such that the unit should not achieve a fmal overspeed greater than the design overspeed of 120%. He energy available to the turbine l

immediately after a trip will carry the turbine speed beyond that of the trip device setting. For the Farley units, it has been calculated that a mechanical overspeed trip at 111% will not allow the unit to achieve a fir.al overspeed greater than 120%. This calculation was based upon the maximum calculated (htax Calc) design points for the unit. Although the new hiax Calc throttle Dow will exceed the old htax Cale throttle Cow, the level of change in the -

parameters critical to overspeed (entrapped energy) was within the tolerance of the original overspeed trip setting calculation. The increase in throttle Dow was within 0.8%, and the new throttle pressure and LP inlet pressures were less than the original hiax Cale point. The enthalpy was virtually the same. Therefore, the i 11% calculated trip point for the mechanical trip device is still valid at the uprated conditions and will not result in a fmal speed greater than the design speed of 120% and therefore will not increase the probability of turbine overspeed and associated turbine missile pnxluction. In addition, secondary backup overspeed protection from the DEH control system is provided by a turbine trip at i11.5% of rated speed.

SCS/mp 7/22/97 & jms Ril/97 NRC Ouestion No. 2 With regard to spent fuct pool (SFP) decay heat loads and cooling, provide the following information:

The heat load and corresponding peak calculated SFP temperature for each case a.

analyzed.

8/5/97 Page 30 pwrupl8. doc

=

h?

.ls full core omoad a general practice for routine refueling? Ifit is, how many trains of

- the SFP cooling system will be available/ operable prior to refueling operation?

SNC Respg_nse No. 2

.The following table provides the requested information. Note that temperatures were only ac calculated for cases 13 to demonstrate compliance with the Standard Review Plan.

Temperatures were not calculated for the Best Estinute Full Core Omoad or post-refueling cases. The total heat load for the Best Estinute Full Core Omoad case is bounded by the BOC and EOC Full Core Omoad cases. The post-refueling cases were analyzed only for heat loads used in evaluating component cooling water system performance and the ultimate heat sink maximum post accident temperature.

Cnw #

1 2

3 4

5A

$11 SC Partial Core Full Core Full Core Full Core iWt l'ost Post Ca.se Ollioud Otlioad Otlload

. Otlload Refueling Refueling Refueling Description (lK1C)

(IXX')

(llest listtmate)

(2$ dap)

(40 dap)

(65 days)

(1/ 2 trains)

(I / 2 trains) (1/ 2 trains)

. Ileat Load 22.1/11.05 37.0 / 18.5 36.5 / IH.25 30.3 15 4 13.6 12.0 (MirIV/hr) l Max. SFP 147/126 17$/140 174 / 140 Temperature

(*F)

- b.

Full core omoad is a general plant practice, Plant procedures require a minimum of one operable cooling train. Plant procedures further require that fuel handling operations be suspended and actions taken to restore cooling upon receipt of the high temperature alann.

SCS/jvi-7/24/97 -

8/5/97 Page 31 purupl8. doc

QUESTIONS REGARDING ATfACllMENT 5, SECTION 6 - NSSS ACCIDENT ANALYSES NROhestieriELI in order to evaluate the impact of future plant changes, equipment problems, or other issues for the power uprate, please provide the doses for the control room operator, EAU, and Lp2 for the five accidents listed in Attachment 6. Please demonstrate that the doses for the control room operators comply with the regulatory criteria for control room doses given in 10 CFR Part 50, Appendix A, General Design Criterion 19.

SEGlefponse No.1 EAB and LPZ doses were reported in Attachment 6, Section 2.16, of the license amendment request. Control room doses have historically been provided in FS AR Table 15.4 17 for the limiting LOCA in accordance with NUREG 75/087; revised control room doses for the limiting LOCA at power uprate conditions were provided in Attachment 6, Section 2.16, of the license amendment request. These results meet 10 CFR 50, Appendix A, General Design Criterion 19.

Note that control room dose calculations were not perfonned for the five accidents referenced above because the NRC SRP does not require such dose calculations.

SCS/ jaw 7/9/97 & SNC/mge - 7/31/97 1

MRC Ouution No. 2 For all of the accidents listed in Attachment 6, provide the assumptions along with the calculational methodology to support the dose analysis results, i.e., the modeling, assumptions, input data, and results of the dose analysis for each postulated accident should be provided. What power level was used for the accidents listed in Attachment 6. What is the core radionuclide inventory based on.

What meteorological data are the X/Q calculations based on.

SNC Response No. 2 loput parameters for LOCA were provided in Tables 2.16-1 through 2.16-3 of Attachment 6 of the license amendment request, and the results were provided in Tables 2.16-4 and 2.16-5. Input parameters for the remaining accidents are provided in the following tables (A - 1). The results for the remaining accidents were provided in Attachment 6, Section 2.16, of the license amendment request.

The power level considered for radiological consequence calculations was 2831 MWt. The core inventory calculations were based on the methods and parameters discussed in Attachment 5, Section 7.6 (and the associated tables) of the license amendment request.

EAB and LPZ atmospheric dispersion factors are based on the data in FS AR Section 2.3 and are shown in FSAR Table 2.3-12. Revised control room atmospheric dispersion factors (X/Qs) were developed based on the meteorological data listed in the FSAR for the period of April 1972 through March 1973. Thejoint frequency distribution tables for the data set were obtained from FS AR Table 2.3-8A, and the results were shown in Table 2.16-3 of Attachment 6 of the license amendment request.

SCS/ jaw - 7/9/97 K/5/97 Page 32 pwrup1M. doc

l TABLE A PARAMETERS USED IN RCP LOCKED ROTOR ANALYSES Core thermal power (MWI) 2831 OfTsite power Lost Fuel defects (%)

NA")

Steam generator tube leak rate 150 prior to and during accident (cpd/ generator)

Activity released to RCS 20% of gap inventory Secondary side iodine activity O.I Ci/gm DElm lodine partition factor in 0.01 steam generators Duration of plant cooldown by M

secondary system aller accident (h)

Steam release from threc 427,000 (0-2 h) steam generators (Ibs) 820,000 (2 8 h)

Feedwater flow to three 574,000 (0-2 h) steam generators (lbs) 908,000 (2-8 h)

RCS activity (including iodine spike) is negligible compared to 20% gap release, a,

SCS/ jaw.7/9/97 8/5/97 Page 33 purupI8. doc

i t

TABLEIl PARAMETERS USED IN LOSS OF ac POWER ANALYSES Core thermal power 2831 MWt Steam generator tube leak rate 150 gpd per generator prior to and during accident Offsite power

' ost Fuel defects 1M*)

lodine partition factor in steam 0,1 generators prior to and during accident Secondary side iodine activity 0.1 pCilgm dose equivalent Im Duration of plant cooldown 8h by secondary system aner accident Steam release from 448,000 lb (0-2 h) three steam generators 861,000 lb (2 8 h)

Feedwater How to three 603,000 lb (0 2 h) steam generators 953,000 lb (2-8 h)

Meteorology Accident (see FSAR Appendix 15B)

A pre-existing iodine spike of 30 Ci/gm dose equivalent im is assumed.

a.

SCS/ jaw-7/I1/97 8/5/97 Page 34 pwrupt 8. doc

TABLE C PARAMETERS USED IN WASTE G AS DECAY TANK RUPTURE ANALYSES Core thennal power 2831 MWt Plant load factor 1.00 Activity released Contents of one tank from GWPS Tank contents See attached table Number of tanks 6.00 (normal operation) lodine partition 0.01 factor in volume control tank Time of accident immediately aller isolation of tank from GWPS Meteorology Accident (see FSAR Appendix 15B)

SCS/ jaw -7/I IN7 8/5N7 Page 35 pu rupl8 doc

TAllLE D WASTE GAS DECAY TANK INVENTORY (Technical Specification Limit for Conservative Analysis)

Activity

_lstles.

(Ci)

Xc.133 6.77 x 10' Xc-133m 1.02 x 10' Xc135 6.77 x 10 2

1 Xc135ri 2.88 x 10*

Xc138 2.63 x 10 '

Kr 85 (a)

Kr 85m 8.03 x 10' Kr 87 9.15 x 10" Kr 88 8.53 x 10' The dose conversion factor for Krn is much less than the other isotopes, and it accumulates a.

much slower than the other isotopea, thus it is conservatively ignored.

SCS/ jaw 7/l1/97 8/5/97 Page 36 pwnipl8. doc

0 TAlli.li II PARAht!!TIIRS USl!D IN STIIAh! LINii!!Rl!AK ANALYSIIS Core thermal power (htWt) 2831 Steam generator tube leak rate prior to 150 accident and initial 8 h following accident (gpd/per generator) l Offsite power Lost Fuel defects (%)

1("

lodine partition factor for initial steam 1.0 release from defective steam generator lodine partition factor in non defective 0.1 stiam generators prior to and during accident Time to isolate defective steam generator (h) 8 initial steam release from defective steam 473,000 (0 30) generator (lb) (min)

Steam release from two non defective steam 339,000 (0 2) generators (Ib)(h) 730,000 (2 8)

Fecdwater flow to two non defective steam 442,000 (0 2) generators (Ib)(h) 791,000 (2 8) hicteorology Accident (see FSAR Appendix 15b)

A pre-existing iodine spike of 30 pCi/gm or an accident initiated iodine spike 500 times the a.

nonnal appearance rate is assumed.

SCS/ jaw 7/1t/97 R/5/97 Page 37

O 4

TAIll.li I PAllAMiiTlik3 US!!D IN STl!AM Gl:N1!ItATOlt TUlli!110PTURl! ANAINSliS Core thermal power (MWt) 2831 Steam generator tube leak rate prior to and during 150 accident (gpd/ generator)

Offsite power 1.ost Fuci defects (%)

I"'

lodine partition factort in non defective 0.1 steam generators prior to and during accident lodinc partition factor in defective steam-0.1 generator prior to and during accident Time to isolate defective steam generator (min) 30 Duration of plant cooldown by secondary 8

system after accident (h)

Steam release from defective steam 73,300 (0 30) generator (ib)(min)

Steam release from two non defective 400,000 (0 2) steam generators (lb)(h) 889,000 (2 8) 17cedwater flow to two non defective 320,000 (0 2) steam generators (ib)(h) 936,000 (2 8)

Itcactor coolant released to the

-150,000 defective steam generator (Ib)

Metec ology Accident (see I'S All Appendix 1513)

Pre accident iodine spike of 30 pCi/gm or accident initiated iodine spike a.

500 times the normal appearance rate, SCS/ jaw 7/ll/97 8/5/97 Page 38

O TAllt.E G ACTIVITIES IN ll!GilEST RATED ASSEMillN AT TIME OF 1 UEL llANDI.ING ACCIDENT l'uct Cladding Gap (100 hr. af ter Reactor Shutdown) holeps (Cif) 1 131 6.99 x 104 1132 l133 6.22 x 10' l.134 l135 5.34 x 10' Kr 85m Kr 85 2.39 x 10' Kr 87 Kr 88 Xc 131m 7.31 x 102 Xc 133 9.57 x 104 Xc-133m 1.52 x 10' Xc 135 2.06 x 10' Xe 135m Xc 138 Total core Ci x gap fraction x radial peaking factor /157 assemblies.

a.

SCS/ jaw 7/l1/97 R/5/97 Page 39

TABLE H PARAMETERS USED IN FUEL HANDLING ACCIDENT ANALYSIS Accident in Spent-Fuel Accident in Refueling Parameter

- Poo* (Auxiliary Building)

Canal (Containment)

Core thermal power 2831 MWt 2831 MWt Time between plant shutdown and accident

'I00 h 100 h

. Minimum water depth between tops of damaged.

23 ft 23 ft fact rods and water surface Dzmage to fuel assembly All rods ruptured All rods ruptured Feel assembly activity Highest powered fuel assembly Highest powered fact assembly in core region discharged in core region discharged Activity release from assembly Gap activity in ruptured rods Gap activity ruptured rods Radial peaking factor 1.7 (maximum) 1.7 Decontamination factor in water

' Elemental iodine (99.75%)

133 133 Organic lodine ( 0.25?4) 1 I

Noble gases 1

I 5 ft' 6.6 X 10 ft' 5

. Amount of mixing in building 1.0 X 10 lodine filtration system Penetration room -

Containment purge system filtration system Filter efficiencies Elementary iodine 90 %

90?&

Organic iodine 70?4 30%

" Atmospheric dilution factors '

Accident (see FSAR Accident (see FSAR Table 15B-2)

Table 15B-2)

SCSTjaw-7/11/97 8/5/97 Page 40

0 Tall!E 1 pal (AMb"IT.l(S USED IN 1(OD lui:C110N ACCIDENT ANALYSI:S Core thermal power (MWt) 2R31 l

Containment free volume (ft')

2.03 x 10' l'uel defects (%)

l'"

Steam generator tube leak rate prior to and 150 during accident (gpd per generato+)

Failed fuel (%)

10 of fuel rods in core Activity released to reactor coolant from failed fuel and available for release Noble gases (%)

10.0 of gap inventory lodines (%)

10.0 of gap inventory Melted fuel (%)

0.2$ of core inventory Activity released to reactor coolant from melted fuel and available for release Noble gases (%)

0.2$ of core inventory lodines (%)

0.25 of core intentory lodine partition factor in steam generators prior 0.1 to and during accident I'latcout of iodine activity released to containment-

$0

(%)

Form of iodine activity in containment available for release Elemental iodine (%)

91 Methyl iodine (%)

4 Particulate iodine (%)

Containment leak rate (%/ day) 0.15 (0 24 b) 0.075 (130 days)

Offsite power Lost Steam dump from relief valves (ib) 426,000

- Duration of dump from relief valve (s) 98 Time between accident and equalization of 2500

. primary and secondary system pressures (s)

Steam dump to condenser (Ib) 0,0 Meteorology Accident (FSAll App.1511

n. lodine activity at _0.5 and 0.1 pCilgm DEliti in the 1108 and secondary systems resrectively.

SC8/ jaw. 7/11/97 8/5/97 Page 41

NRCJAtestion_Ned For the Control Rod lijection Accident, explain why releases from the secondary side were not included in your evaluation.

SNCAmetud9J Attachment $, Section 6 of the license amendment request addressed only the core and NSSS response to a Control Rod IIjection Accident and provided core releases to the RCS. Attachment 6 of the license amendment request addressed the radiological consequences of the accident and l

included both RCS leakage through the steam generators and activity contained in the secondary l

side water Also see SNC response to NSSS accident analyses question No. 2 above.

SCS/pw. 7/l1/97 M/5/97 Page 42 mm

ADDITIONAL QUESTIONS NliC OuntiojtBo_l in regard to Sections $ I.1 and 5.2.3 of Reference 2, provide the masimum calculated stress and cumulative fatigue usage factor (Cult) at the critical locations of the reactor pressme vessel (Iti'V) and internal components (such as RPV noules, core plates, core barrel, bafIle/ barrel, and fuel assembly, etc.). Also, provide the allowable Cale limits, and the Code and Code edition used in the evaluation for the power uprate. If different from the Code of record, provide the necessary justification.

Sbhrpmse No. I llegarding Section $.l.l. and the reactor vessel, the maximum CUF at all of the limiting locations, except those in the CitDM housings, the bottom head instrumentation nonles, the closure head flange and the closure studs, increase from the previously calculated values. Ilowever, the increases are generally minimal, and all of the CUI'!. remain under the 10 limit with margm. The greatest increase in CUF is 0.157 at the core support pads. The Code edition used was the Code of record, which is the ASME il&PV Code, Section lit,1968 Edition through Summer 1970 Addenda.

The stress intensities and CUl7s for the reactor vessel critical locations are provided in Table A, which follows.

i 8/5/97 Page 43

a l

TAHLE A

{

STRESS INTENSITIES AND FATIGUE

}

USAGE FACTORS FOR Tile REACTOR VESSEL Location P.+ I% + Q Range U,

i Outlet Nonles -

Nonle: 54.7 ksi < 3 S. = 80.1 ksi Nonle: 0.1299 < l.0 Safe End: 40.5 ksi < 3 S. = 49.2 ksi Support: 0.2871 < l 0 Inlet Nonles Nozzle: 47.83 ksi < 3 S. = 80.1 ksi Nonle: 0.0276 < l.0 Safe End: 45.16 ksi (Unit 1) < 3 S. = 49.8 ksi Support: 0.1693 < l.0 43.77 ksi (Unit 2) < 3 S. = 49.8 ksi Main Closure Flange Region

1. Closurelicad 62.08 ksi < 3 S. = 80.1 ksi 0.2259 < l.0 Flange

2. Vessel Flange 77.98 ksi < 3 S. = 80.1 ksi

3. Closure Studs 104.4 ksi < 3 S. = 110.3 ksi 0.9211 < 1,0 CRDM llousir.gs 45.95 ksi < 3 S. = 69 9 ksi 0.5688 < l.0 Vessel Wall Transition 31,72 ksi < 3 S.= 80.1 ksi 0.0603 < l.0 llottom llead to Shell 36.53 ksi < 3 S = 80.1 ksi 0.0052 < l.0 Juncture 110ttom licad 57.28 ksi < 3 S. = 69.9 ksi 0.2201 < l.0 Instrumentation Tubes Core Support Pads 42.79 ksi < 3 S.= 80.1 ksi 0.2224 < l.0 R/5/97 Page 44

e Regarding Section 5.2.3 and the reactor internals, since the Farley reactor internals were designal prior to the introduction of Subsection NG of the ash 1E !!&PV Code Section lil, a plant specinc stress report on the reactor internals was not ratuired. Ilowever, the criteria described in Subsection NG of the ASME Code were utilized. Component quali0 cations were based on the use and extension of existing analyses which had been performed for similar plants which were designed and built strictly in accordance with ASME Subsection NO rajuirements. In addition, a new analysis of the lower core plate was performed in order to remove conservatism.

hittiintunLCalculainLSktndllamthltaniCIUEALCri!!callacatian_s.nUltatioLintettiaLCsmp_cuents REAC10R MAX CODl!

MAX CODE MAX CODl!

CUF INTERNAL STRESS LIMIT STRESS LIMIT STRESS LIMIT COMPONENT (3)

Pm Sm Pm+Pb 1.5Sm Pm+Pb4Q 3Sm Lower Core Plate 3.2 ksi 16.2 ksi 5.85 ksi 24.3 ksi 31.6 ksi 48.6 ksi

.046 (5.2.3.2) lla01c Darrel lloit NA (1)

NA (1)

NA (1)

NA (1)

NA (1)

NA (1)

.917 (5.2.3.3.3)

Upper Core Plate 6.5ksi 16.4 ksi 13.2 ksi 24.6 ksi 70.2 ksi (2) 49.2 ksi (2)

.08 (2)

(5.2.3.3.4)

Notes.

(1) The baffle barrel bolts were quali0cd by test.

(2) The combined primary and secondary stresses exceed the limit. An clastic plast c fatigue analysis i

in accordance with NG 3228.3 was performed to demonstrate that the cumulative fatigue usage attributed to the combination of the low cycle events plus all other cyclic events did not execed the value of 1.0.

(3) He review of other components, such as the Core Harrel (5.2.3.3.1) and lla01c Plate (5.2.3.3.2),

showed very high margins and the upratal conditions did not signincantly effect the structural integrity of the component.

W/rs 7/31/97 NILChugstion No 2 In regard to Section 5.2.2.2 of Reference 2, provide an assessment of Dow-induced vibration of the reactor internal components due to power uprate, SNC Resprnsr_No. 2 For uprated conditions, it was determined that the Dow induced vibration (FIV) loads on the guide tubes and the upper support columns increase by approximately 1.9% Previous FIV ana'yses on the guide tubes and the upper support columns show that there exist suf0cient margins to K/5/97 Page 4$

l

acconuncdate this small increase in the FIV loads. Consaluently, the structural integrity of the Farley reactor internals remains acceptable with regard to dow induced vibrations.

W/rs 7/22/97 MRC.mtclion No. 3 in reference to Section 5.4 of Reference 2, provide an evaluation of the control rod drive mechanism with regard to the stress and fatigue usage as a result of the power uprate. Also, provide the allowable Cale limits for the critical components evaluated, and the Cale and Cale edition used for the evaluation. If different from the Code of recorri, justify and reconcile the differences.

SEClapsnse No 3

'lhe evaluation performed for the control rod drive mechanisms (CRDhis) addressed the AShil!

Code structural considerations for the pressure boundary components of both the part length CRDhis, which are not in use but the pressure boundary components remain present, and the full-length CRDhis. The cale and code edition applicable to the Farley CRDhis is the 1470 Sunuucr Addenda of the AShill Section lil Nil Cnie.

lhe evaluation performed fbt the CRDhis used as input the NSSS PCWG Parameters for power uprate as shown in Table 2.12 of the Power Uprate NSSS Licensing Report and the NSSS design transients for power uprate as discussed in Section 3.0 of the Power Uprate NSSS Licensing Report. During normal operation, the CRDhts experience the RCS pressure and RCS " vessel outlet" temperature as shown in Table 2.12 of the Power Uprate NSSS Licensing Report. The PCWG parameters for power uprate show that RCS pressure does not change but that " vessel outlet" temperature increases slightly (i.e., to a maximum of 613.3*F). Despite the slight increase in temperature, the nonnal operating conditions for power uprate remain bounded h the original

)

generie analysis (as documented in the generie code reports) which used the maximum design temperature of 650'F.

With respect to part-length CRDht stress and fatigue, the current lowest margin of safety on the primary plus secondary stress intensity as shown in the generic analysis is 32.6% for the analyzed case of 2500 psi and 650 F. Review of the design transients for power uprate showed that the maximum transient pressure could be approximately 6% above the 2500 psi used in the generic analysis. Since the generic analysis showed a margin of 32.6%, it was concluded that adequate margin existed at power uprate conditions relative to the code stress intensity limit of 3Sm.

Furthermore, it was concluded that all APs due to power uprate design transients were less than the Code definition of "significant fluctuation" value and that no fatigue consideration is required since the generie waiver remains unchanged.

With respect to full length CRDh1 stress and fatigue, the slight increase in " vessel outlet" temperature will not change thermal stress results and has a negligible efTect on CRDh1 material properties, To assess the impact of power uprate on the full length CRDhis, the location of maximum stress and fatigue (i e., canopy of the upperjoint) was chosen for conservative numerical evaluation. The results of this conservative evaluation showed that the generic evaluations of primary plus secondary stress ranges including the simplified clastic plastic analy ses performed in the generic report remain applicable. The poiver uprate evaluation regarding fatigue also used the R/5/97 Page 46

I upper canopy since it has the largest fatigue usage factor. The conservative power uprate evaluation showed a total usage factor of 0.672 which was less than the conservative fatigue usage factor of 0.H58 calculated in the generic report and which is less than the Code fatigue usage limit of 1.0. liased on the numencal comparison at the location of maximum fatigue, it was concluded that the results of the generic analysis are valid for power uprate conditions.

In summary, the power uprate evaluations for the part length and full length CRDh1 pressure boundary components showed that the Code criteria are satis 0cd at power uprate conditions.

W/rs 7/22/97 NILCJQuestien No. 4 in reference to Section 5.$ of Reference 2, piovide the methalology and assumptions used for evaluating the reactor coolant piping systems for the power uprate. Also, provide the calculated maximum stress, critical locations, allowable stress limits, and the Code and Code edition used in the evaluation for the power uprate. If difTerent from the Code of record, justify and reconcile the -

difTerences.

SE C l vspons d o.4 The methodologies and assumptions used in evaluating the impact of power uprate on the Heactor Coolant Loop (RCL) piping and supports are consistent with the original naalyses As discussed in the response to general question No 2, the PS4 CAEplPE computer code was used in the power uprate evaluation The evaluation addressed changes to the PCWG parameters (as shown in Table 2.12 of the Power Uprate NSSS Licensing Report), NSSS design transients, and LOCA interface loads.

In all cases, except for the RCL crossover leg usage factor, the existing evaluations as documented in the original analyses remained unchanged. The crossuer leg usage factor experienced a minor increase from 0.051 I to 0.1319 due to the design transients for power uprate; however, the usage factor remained below the specified acceptance limit of 0.2 for break postulation. This assessment was done in compliance with the ash 1E D&pV Code Section 111,1971 Edition and all addenda thru Summer 1971. This is the same Cale version and addenda used in previous evaluations, in summary, the results of the power uprate evaluation for RCL piping and supports (including primary equipment nozzles, primary equipment supports, pressurizer safety and relief and piping and piessurizer surge lhe) showed only one variation from previously calculated values and satisfy the requirements of the identified ash 1E cale.

W/rs 7/24/97 NRC Question N_od Discuss the analytical methodology and assumptions used in evaluating pipe supports, nozzles, penetrations, guides, valves, pumps, heat exchangers, and anchors at the power uprate conditions Were the malytical computer codes used in the evaluation difTerent from those used in the original i

. design-basis analy sis? If so, identify the new codes and providejnstification for using the new codes and state how the codes were qualified for such applications.

K/5/97 Page 47

l SNC ResponsdM The response to general question No. 2 regarding WCAP 14723 describes the areas where the l

NSSS analyses for power uprate used different (i c., new or improved) methmiologies. In the other NSSS analyses areas, the power uprate analyses used the same basic methodologies as the analyses currently described in the FSAR.

The SNC responses to general question No.1I and additional question No.10 provide additional infonnation describing the methmlologies used in evaluations performed for the reactor vessel and reactor internals.

W/rs 7/22/97 Ni[C_ Question No. 6 in regard to Section 5.6 of Reference 2, provide a comparison of the design parameters and transients for the reactor coolant pump (RCP) against the power uprate condition, Also, provide the maximum calculated stress and CUF for the RCP, the allowable Code limits, and the Code and Code edition used in the evaluation for the power uprate. If difTerent from the Cale of record, provide a justification.

SNC RunonidAh i

The evaluation perfonned for the RCPs used as input the NSSS PCWG Parameters fer power uprate as shown in Table 212 of the Power Uprate NSSS 1.icensing Report and the NSSS design transients for power uprate as discussed in Section 3.0 of the Power Uprate NSSS Licensing Report. This table provides a comparison of the design parameters. A comparison of the design transients (i c., type and number of occurrences during the 40 year license perimi) for Farley power uprate to the original design transients for Farley is provided in Table A, which follows, With respect to RCP stress and fatigue, the APs and ATs associated with the power uprate design transients were resicwed to detennine if there were any changes that would qualify as a "signincant fluctuation" per the Code dennition and thus require consideration relative to fatigue. It was concluded that all APs due to the power uprate design transients were less than the Cale dennition of"signi0 cant fluctuation" value and that no fatigue consideration is required. The design transients were then reviewed to identify the maximum pressure to which the RCP could be exposed. For Farley, this maximum pressure was determined to be approximately 2650 psia for the loss ofload transient. A review of RCP analyses performed for other plants showed that increases to 2725 psia have been analyzed in detail and shown to be acceptable. It was concluded that the pressure transients are acceptable.

The effect of power uprate on the various RCP generie reports was then assessed. The AP and AT values were used m these assessments. For the most part, the assessments of AP and AT values were sufTicient to show continued applicability of the generic reports to power uprate conditions.

The increase in AT was suf6cient to merit analysis for the casing weir plate. Evaluations showed a stress intensity range of 47,325 psi for the power uprate conditions. Comparison of this value to the Code primary plus secondary stress limit of 3Sm = 50,700 psi showed that the Code limit is satisfied. CUF requirements for the weir plate were satis 6ed by the fatigue waiver evaluation, M/5/97 page 48 l

I 1

i

The evaluation perfbrmed for the RCPs addressed the ASMl! Code structural considerations for the RCP casing, main flange, main flange bolts, theimal barrier, easing foot, casing discharge and suction nonles, casing weir plate, seal housing and auxiliary nonles. The emle and cale edition applicable to the Farley RCPs is the ASME Section lil Nil Code and Appendices,1971 Edition with Addendum (Unit 1) and through 1972 Sununct Addendum (Unit 2).

In summary, the results of the power uprate assessments showed that the Code criteria are satisfied at power uprate conditions, W/rs 8/5/97 & SNC/mja 8/4/97 1

8/5/97 Page 49 i

TAllt.E A APPLICAllLE DESIGN 1RANSIEN1S CYCLE COUNT COMPARISON l~OR REAC10R COOLANT PUMP Plant Condition Power Uprate

• RCP G neric Analpli Number of Occurrences Number of Occurrenttiin in 40 Year Operating 40 Year Operating IJtente

1) cense Period Period UEutal Condition 1.

Plard !!catup 200 200 2.

Plant Cooldown 200 200 1 Unit lxmding at $%/ min 18,300 18.300 4.

Unit Unloading at 5%/ min lH,Mul 1 R,300 5.

Small Step lxad increase 2JHK) 2/KK) 6.

Small Step lxed Decrease 2,(KK) 2,(KK) 7.

1.ntge Step lamd Decrease 200 200 8.

Steady Ftate 11uctuations Infinite Infinite 9 feedwater Cyling/llof Standby -

2JMKP NA Operation

10. 1 mbine Ra'i Test 10 10 (1est Condition) llPxtCrnditiel) i 1.oss of1xmd 80 80 2.

l.oss of Power 40 40 3

loss of f low KO ho 4.

Reactor Trip from Full Power 400 400 Inadvedent RCS Depressunation 10 to htultedf1'nddieu 1.

Reactor Ctulant l'ipe litcak 1

I 2.

Design llanis Ikuthipmle (DitE) l I

3 Stetun 1.ine lireak i

1 4.

Steam Gwnerator Tube Rupture (included atore in reactor trip from full i

pmer)

InLCelt4ttiens 1.

Primary Side flydrostatic Test 5

5 2.

Secondary Side Ilydrostatic Test 10

• 5 3.

Primary Side leak Test 50 50 Neter (1) The transient descriptions contained in the appropriate component E specs remained applicable for the power uprate unless it has been nulified for power uprate conditions.

(2) Feedwater Cycling /Ilot Standby Operation transient included in power uprate for consistency with Steam Generator E Spec. It need not be addressed for power uprate ifit was not in the original design basis.

(3) The number of occurrences for the secondary side hydrostatic test transient increased to 10 from the original design requirement of 5.

8/5/97 Page 50

NRCRuntieriA'nd in regard to Section 3.7.1 ofl(cretence 2, provide a comparison of the design parameters and transients for the Parley steam generators (SGs) Model 51 against the power uprate condition.

Also, provide the maximuh. calculated stress and CUF for the SGs vessel shell and nonles, the allowable Code limits, and the Code and Cafe edition used in the evaluation for the power uprate.

If difTerent from the Cale of record, provide a justification.

SNC Itcspemdu._2 A comparison of the design parameters for power uprate to the original design parameters for Parley is provided in Table 2.12 of the Power Uprate NSSS Licensing iteport A summary of the design transients (i.e., type and number of occurrences during the 40 year license period) for Fmley power uprate it provided in Table A, which follows, r

'lhe power uprate evaluations performed for the SGs addressed the ASME unde structural considerations for the critical SG primary side components (including tube to tubesheet weld, tubes, channel head, tubeeheet, stubbarrel, tubeshect/shell junction and divider plate) and the critical 50 secondary side components (including feedwater nonic, secondary manway opening, secondary manway bolts, and steam nonle). 'ihe cale and cale cdstion applicable to the Farley SGs is the ASME Section til Code,1971 Edition.

Summaries of the maximum stress / allowable ration for steam generator components, fatigue usage in steam generator primary side components, and fatigue usage in steam generator secondary side components are provided in Tables 11, C and D, respectively, which follow.

l M/5/97 Page$1 I

-.-,--w y

....,.--m.

.%y n

,.~ -

.-v w-.e--

TAllLE A APPLICAllLE IDESIGN TRANSIENTS CYCLE COUNT F0". STEAM GENERATORS Plant Condition Number of Occurrences in 40 Year Operating Licenne Period Normal Condition Plant hcatup 2(K)

Plant cooldown 2(KI Unit loading @ $9Vminute 1x,300 Unit unloading (M $%' minute 18,300 Small step load increase 2(HN)

Small step load decrease 2(MK)

Large step load decreaf f 200 Feedunter cyclina"..ot standby oper.ition 2(xx)*

Turbinc : oil test to Upect Condition Loss ofload 80 Loss of power 40 Loss of flow 80 Reactor trip from full power 400 Faulted Coadition Reactor coolant pipe break i

Design ksis carthquake i

Steam line break i

Test Condition Primary side hydrostatic test 5

Secondary side hydiostatic test 1o**

• The number of occurrences has been reduced to 2(KM) from the original design requirement of 18,300.

• *The number of occurrences has been increased to 10 from the original design 4

requirement of 5.

M/5/97 Page 52 i

l

TAllt.E 11

SUMMARY

Ol? MAXIMUM STRESS /Al.I,0WAllLE RATIOS IN STEAM GENERATOR COMPONENTS 1,oad Condition Component Ratio Ratio Current 100%

Low temp Power Uprate Normal Tubesheet/shell 0.725 0.820 Design Tubes 0.79 0.79 i

Divider plate

>3Sm

>3 Sm Normal & Upset Tubesheet 0.733 0.733 Tube /tubesheet 0.872 0.976 weld TAllLE C

SUMMARY

Ol' FATIGUE USAGE FOR STEAM GENERATOR COMPONENTS Load Condition Component Current 100%.

Low. Temp Power Uprate Tubesheet (cold leg) 0.186 0.282 Normal and Upset Tube /tubesheet weld 0.063 0.099 Divider plate 0,791 0.944 8/5/97 Page 53

TAlli,E D SUMMAIW OF FATIGUE USAGE OF SECONDAlW SIDE COMI'ONENTS Coinponents Fatigue Usage (Uprated l.ow Temperature)

Feedwater no721e 0.779 Secondary manway opening 0.031 Secondary manway bolts 1.1803 Steam nor21e 0.590 Note: The bolts are to be replaced prior to 34 years of operation.

i i

M/5/97 Page $4

c NIACDuntionNoJ In reference to Section 5.7.3 of Referenec 2, provide a detailed evaluation of the now induced vibration of the steam generator U bend tubes due to power uprate regarding the analysis methodology, vibration level, computer cales used in the analysis and the calculated cross How velocity. Ihplain why the tube repair would not be required for at least 13.7 years at the propose I power uprate.

SMGupEnaNpJ A complete evaluation of potential U bend vibration and fatigue was perfonned for Farley in 1988.

11ic results of that evaluation were reported in WCAp l 1876 (Reference 1) and concluded that no tubes in either unit required preventive action. The evaluation was updated in 1990 to support a license amendment request to increase the steam generator tube plugging level and reduce the RCS nermal Design Flow. %c update evaluated potential U bend flow conditions, vibration potential and fatigue usage based on changes in steam pressure and steam now which resulted from the increased steam generator tube plugging and reduced RCS Dow. The results of the updated evaluation were reported in WCAp 12M4 (iteference 2) and concluded that fatigue usages for cach susceptible tube remained within acceptable ranges, and therefore, no tt.bes required preventive action in support of power uprate, the U bend fatigue evaluation was updated using the same evaluation methalology established in Reference I and used in Reference 2 in order to establish whether power uprate conditions (i.e., changes in steam pressure and steam flow) cause any of the inner row U-bend tubes to become susceptible to fatigue. Changes to vibration levels and cross Cow velocity due to power uprate were addressed in the evaluation.

- A detaded description of the U bend fatigue evaluation methmlology and computer cales is provided in Reference 1. Infonnation and equations include presentation of the one dimensional methmlology used to account for changes in operating conditions such as ihr power uprating. This information is used to calculate the level of stress resulting from the increased flow induced vibration response of the limitmg tube (s). The total fatigue usage (including the fatigue usage at the previous operating conditions and also the fatigue usage at the future operating conditions) is then determined. This inforn,ation is then used to determine how many years of operation would be required to obtain a fatigue usage of 1.0. The number of years of operation required to obtain a fatigue usage of 1.0 for the limiting tube (s) is then documented.

As described in Section 5.7.3 of the power Uprate NSSS Licensing Report, the ll bend fatigue evaluation for power uprate showed that no preventive tube repair is required to support full power operation of the Farley units at the anticipated steam generator outlet pressure (i.e.,787 psia) for power uprate. This is because at this steam generator outlet pressure, no steam generator tube in either unit is Aawn to accumulate a total fatigue usage of 1.0 prior to the end of the operating license perial. Also as described in Section 5.7.3, analysis peribrmed for lower steam generator outlet pressures and considering the potential for asymmetric steam 00w from the different steam generators showed that under these conservative conditions no tube would accumulate a total fatigue usage that would exceed the acceptance limit of 1.0 for at least 13.7 years of operation after implementation of power uprate on Unit I and Unit 2 in the 1998 outages. Consequently, no tube would require preventive repair prior to that time. As noted in Section 5.7.3 of the power Uprate NSSS 1 icensing Report, following the implementation of power uprate, the steam generator operating conditions (i.e., steam flow and pressure) can be documented on a cycle-speci6c basis

- for use in any future update to the U-bend fatigue evaluation.

8/5/97-page 55

c 4

Ropene No 8 Refylgnyn 1.

WCAP il876," Joseph Farley Unit 1 & 2 Evaluation for Tube %bration Induced Fatigue,"

July,1988,

2. WCAP 12694," Alabama Power, Joseph M. Farley Unit 1 Increased Steam Generator Tube Plugging and Reduced Thermal Design Flow Licensing Report," August,1990.

W/rs 7/24/97 NRC Ountion No. 9 in regard to Section 5.8 of Reference 2, provide a comparison of the design parameters and transients for the pressurizer against the power uprate condition Also, provide the maximum calculated stress and CUF at the critical kications (such as surge nozzle, skirt support, spray nozzle, safety and relief nozzie, upper head / upper shell and instrument nozzle) of the pressurizer, the allowable Calc limits, at d the Cale and Code edition used in the evaluation for the power uprate. If difTerent from tht Cale of record, provide ajusti0 cation.

SEChupangNo9 A comparison of the design parameters for power uprate to the original design parameters for Farley is provided in Table 2.12 of the Power Uprate NSSS Licensing Report. A comparison of the design transients (i e., type and number of occuriences during the 40 year license perial) for Fmley power uprate to the original design transients for Farley is provided in Table A, which follows.

The limiting operating conditions of the pressurizer occur when the RCS pressure is high and the RCS hot leg and cold leg temperatures are low This maximizes the AT that is experienced by the pressurizer The RCS pressure is unchanged for the uprate, but the minimum cold leg temperature decreases by 13 degrees, with respect to the original design conditions. The CUFs at the critical locations are potentially affected by the uprate conditions, and the new values are provided in Table 11, which follows. The maximum calculated stress at the critical locations is unchanged, except for the surge nozzle. For the surge nozzle, Pla Ph+ Q stress intensity range is 49,972 psi, and the allowable stress limit of 3Sm is 80,I00 psi.

The evaluation was performed by modifying the original Farley Units I and 2 pressuriter stress reports, which were performed to the requirements of the ASME Iloiter and Pressure Vessel Cale, Section 111,1968 Edition, Summer of 1970 Addendum, for Farley Unit I and 1968 Edition, Winter of 1970 Addendum, for Farley Unit 2. No new Cale versions were used for the uprate evaluations.

8/5/97 Page $6

o TAllLE A APPLICAllLE DESIGN TRANSIENTS AND CYCLE COUNT COMPARISON FOR PRESSURIZER Plant Condition Power Uprate ("

Original Design Number of Occurrences in Number of Occurrences in 40 Year Operating 40 Year Operating License Period License Period Nero1:d_Ceriditinn

1. Plant lleatup 200 200

2. Plant Cooldown 200

-200-

3. Unit Loading at $Wmin 18,300 18,300 l

4. Unit Unloading at SWmin 18,300-18,300

5. Small Step lead increase 2,000 2,000

6. Small Step lead Decrease 2,000 2,000 l

7. Large Step Load Decrease 200 200 8.

Steady State Fluctuations Innnite infinite 9 Feedwater Cycling /llot Standby 2,000 m NA Operation

10. Turbine Roll Test 10 10 (Test Condition) llpset Condition

1. Loss of Load 80 80

2. Loss of Power 40 40 3.

loss of Flow 80 80

4. Reactor Trip from Full Power 400 400

5. Inadvertent Auxiliary Spray 10 10 Dudlod Conditicu

1. Reactor Coolant Pipe 11reak i

1

2. Steam Line lircak 1

1 InLCenditions

1. Primary Side llydrostatic Test 5

5

2. Primary Side Leak Test 50 50 Notes (1) The transient descriptions contained in the appropriate component E specs remained applicable for the power uprate unless it has been nulined for power uprate conditions.

(2) Feedwater Cycling /llot Standby Operation transient included in Power Uprate for consistency with the Steam Generator E-Spec. It need not be addressed for power uprate ifit was not in the original design basis.

M/5/97 page 57 1

TAHLEH FARLEY UNITS 1 AND 2 PRESSURIZER FATIGUE USAGES Component Calculated Fatigue Usage Surge Noule

<0,17 Spray Novje

<0,94 Safety and Relief Nonic

<0.15 Lower licad,lleater Well

<0.01 Lower llead, Perforation

<0.07 Upper llcad and Shell

<0.78 Support Skirt / Flange

<0.01 Manway Pad 0.0 Manway Cover 0.0 Manway Holts 0.0 Support Lug

<0.05 Instrument Nor21c

<0.11 Immersion lleater

<0,01 Valve Support 11 racket 0.01 9

- 8/5/97 Page 58

NiiC_QuclionEe.10 in reference to Sections 3.1 and 5.2 ofiteference 2, provide the methodology, assumptions, and loading combinations used for evaluating the reactor vessel and internal components with regard to the stress and CUF for the power uprate. Were the analytical computer emies used in the evaluation difTerent from those used in the origmal design basis analysis? If so, identify the new emics used and provide justification for using the new codes and state how the cales were qualified for such applications.

SECJ1onem.NoJD Reactor Vessel With respect to the methodology used in the uprate evaluation for the reactor vessel, the NSSS Design Transients for the uprate were reviewed to determine the most severe transient temperature variations and magnitudes of pressure variations. The original stress report was modified to reficet the changes to the NSSS Design Transients incurred by the uprate.

For regions of the vessel operating at temperatures near T hot (outlet nozzles, main closure, CRDM housings), the most severe transients were identified to be:

Unit loading at 5% of full power; Unit unloadmg at 5% of full power; e

Step load inercase of 10% of full power; Large step load decrease; r

e less ofload; e

loss of power; and e

Feedwater cycling at hot shutdown.

For the remaining regions of the vessel, which operate at temperatures near T-cold, the most ses ere transients were identified to be the same ones as listed above for T hot, except that Step Load increase is not included and Loss of Flow in One Loop is added.

Calculations were performed to account for the changes in stress due to the malified temperature and pressure variations, There were no analytical emles used in this evaluation, new or otherwise.

The methodology used to evakiate the vessel thermal stresses is provided in Document Pil-151487,

" Tentative Structural Design llasis for Reactor Pressure Vessels and Directly Associated Components (Pressurized, Water Cooled Systems)," U.S. Dept. of Commerce,1 December 1958 Revision with Addendum No I dated 27 February 1959. Pressure stresses were scaled from previous results.

Input assumptions that were modified for the Uprate (in addition to the revised transients) were new LOCA reactor vessel / internals interface loads, which were compared to the loads previously considered and found to be enveloped by the previous loads 8/5/97 Page 59

Reactor Internals Mctiejelegy l

Structural integrity evaluations v ere performed to demonstrate that the stnictural integrity of the reactor componema is not adre sely affected by the change in RCS conditions and transients and/or by secondary effects of the change on reactor thennal hydraulic or structural perfonnance. The presence of heat generated in reactor internal cc;nponents, along with the various fluid temperatures, results in thennal gradients within and between components. These thennal gradients result in thermal stresses and thennal growth which must be accounted for in the design and analy sis of the various components. The approaches (i c., meth'dologies) used to evaluate the thennal stresses included.

detennination of temperature distributions in the component; e

detennination of stresses in the component; and e

determination of margin of safety and fatigue usage factor for the most severely stressed e

location of the component.

In addition to the thermal loads, the mechanical loads due to the following conditions were considered: pressure differentials due to coolant flow; weight of the structure, superimposed loads from other components; carthquake (or seismic) loads; loss of coolant accident (LOCA) loads; vibratory loads; and preloads.

la0LComhinationunilamis Nonnal operation (Senice level A) conditions include any condition in the course of system startup, operation in the design power range, hot standby and system shutdown, other than Upset, Emergency, Faulted or Testing conditions.

Upset (Senice Level B) occurrences include any deviations from Nonnal conditions anticipated to occur often enough that the design should include a capability to withstand the conditions without operational impairment.

Emergency (Senice Level C) conditions include those deviations from nonnal conditions which require shutdown for correction of the condition or repair.

Faulted (Senice level D) conditions include those combinations of conditions associated with extremely low probability postulated events u hose consequence., are such that the integrity and operability of the system may be impaired to the extent that consideration of the public health and safety are involved. Under faulted conditions, LOCA (loss of coolant accident) and SSE (safe shutdown earthquake) loads were considered without secondary loadings.

Anablical ComskrfaitsliDirrnLErem OriginalJssia3 mis The structural analysis for the reactor internals lower core plate (as described in Section 5.2.3.2 of the Power Uprate NSSS Licensing Report) used the ANSYS fmite element computer code which is the cunent Westinghouse analytical tool for use in thennal and stress analy ses of lower core plates.

Although this was the first Westinghouse application of this code to the Farley lower core plate, it R/$/97 page 60 i

has previously been u.ed by Westinghouse on Farley, and its use is documented in the Farley FSAit Section 5.2.1.11. " Analysis Method for Faulted Condition."

W/rs 7/22/97 NRCJQuntieda._11 l

Discuss how the calculated CUFs for the reactor vessel and piping components compared to the CUFs resulting from the actual loading cycles based on the data recorded during plant operation.

SNC.RupadoJ.1

• lhe power uprate project was structured consistent with the metinxiology established in WCAP-10263, A Review Plan for Uprating the Licensed Power of a Pressurized Water Reactor Power Plant. Inherent in this methodology are key points that include the use of currently approved I

analytical techniques (e g., methxtologies and computer codes) and the use of currently applicable licensing criteria and standards. Consistent with this metixxiology, the approach used to assess the impact of power uprate on NSSS components and to show the acceptability of the NSSS components for operation at power uprate conditions was to (1) revise the NSSS design transients (i c., temperature / pressure profiles) to be applicable to power uprate conditions and (2) using the revised NSSS design transient pro 6les, evaluate the NSSS components to determine the fatigue usage factors for power uprate conditions. 'lhe fatigue usage factors were then compared relative to the code acceptance limits to show that the NSSS components comply with ASME Code acceptance criteria and can operate acceptably at power uprate conditions.

CUFs resulting from actual loading cycles based on the data recorded during plant operation were not calculated as part of power uprate anal > ses and were not compared to the CUFs based on the NSSS design transients for power uprate. This comparison is not required fbr compliance with the methodology in WCAp.10263.

Whs. 7/24/97 NRCRuntion No.12 Discuss the operability of safety related mechanical components (i.e., valves and pumps) alTected by the power uprate to ensure that the perfonnance specifications and technical specincation requirements (e g., flow rate, close and open times) will be met for the proposed power uprate.

Confirm that safety-related motor operated valves (MOVs) will be capable of performing their intended functions following the power uprate including such affected parameters as Guid now.

temperature, pressure and difTerential pressure, and ambient temperature conditions. Identify mechanical components (br which operability at the uprated power les el could not be con 6nned SNERupadoJ2 The only physical pump modi 6 cations required for uprate were for the condensate pumps (i.e, there were no safety-related pump mods) All other 110P system evaluations were performed with pump Dows based on current operation and'or established pump design ranges. Assumed RilR -

pump and charging pump performance was degraded to provide operational margin and was explicitly modeled in the ECCS flow analyses. Acceptable pump performance was therefore M/5/97 page 61

6 demonstrated by the various analpes (e g., l.OCA and non l.0CA analpes) which met applicable acceptance criteria and technical speciGeation requirements.

'lhe methodology used for the FNP MOV Program to detennme the design basis difTerential pressure (AP's) and line pressure included the most restricted conditions for elevation head, maximum pump shut-off bead, upstream and downstream pressures (where applicable). The clevation heads were determined by assuming that the elevation head of the tank or sump were at their maximum / minimum operating levels to provide the greatest head difTerectial. In most cases the downstream pressure was assumed to be rero to yield the greatest difTerential pressure across a valve in some cases, where sptem conditions dictated, the downstream or back pressure was used.

The maximum pump shut-off head was used in most cases escept for the Rllit pump and charging pump mini Dow, service water, and component cooling water MOVs. For the minimum flow valves, the pump head was based on flow data lbr normal and accident conditions. The pump head for the Senice Water and CCW MOVs was based on data obtained from flow models developed specincally for those two systems. 'Ihc existing flow conditions are not alTected by uprate, and therefore, the existing design basis assumptions used in FNP motor-operated valve program are still valid.

In addition to the above parameters the following assumptions were included Pressure drop due to piping loss were neglected, escept in cases uhere the flow models (primarily SW and CCW system) were employed The value used for the density of water was 62.4 lbs/fl' at 60'F. This value provided the most conservative value (highest pressure) and any fluid temperatme above 6mF was considered negligible.

Relief valve setpoints were conservatively assumed to be 103% oflift setting.

in cases w here the elevation of sister valves varied from train to train or from one unit to the other,

+

the most conservative elevation value fbr the group was used to produce the highest difTerential 4

pressure or line pressure.

The uprate design basis containment pressures and temperatures for normal and accident conditions are bounded by the current P/T analysis.

SCS/gid A,iva 7/24/97 NRC Outstion No. I3 in reference to Reference 3, list the balance of-plant (llop) piping systems that were evaluated for the power uprate. Discuss the methodology and assumptions used for evaluating IlOP piping, components, and pipe supports, nonles, penetrations, guides, valves, pumps, heat exchangers, and anchorage for pipe supports. Were the analytical computer codes used in the evaluation different from those used in the original design basis analysis? If so, identify the new codes and provide justification for using the new codes and state how the codes were qualified for such applications.

SNC Response No 13 The fbliowing HOP piping systems were evaluated for power uprate conditions; mMr, s'eam; extraction steam; condensate and feedwater. The evaluation involved the review of the effect of new temperature, pressure and flow rates on those piping systems. This includes the evaluation of 8/5/97 Page 62

thermai tapansion due to new temperature, increase ot' hoop and longitudinal stress due to pressure, and steam and water hammer effects (if any) due to new flow rates.

In addition to the secondary system piping evaluations discussed below, the piping most susceptible to flow accelerated corrosion (FAC) has been replaced with FAC resistant piping.1hc remaining

$cctions of FAC-susceptible piping are being monitored in accordance with the FAC program. The FAC program uses CilECWORKS as a predictive model for selecting inspection kications.

Main Steam l

For the main steam piping system, the effect of the uprate temperature is not significant. The uprate operating temperatures are less than those used in the current base calculation; therefore, the thennal expansion effect of the main steam is less for the uprate condition.1he uprate pressure is also less than the pressure used in the piping analysis; however, the uprate flow ate is approximately 5% greater than the current flow rate.1he uprate flow rate is bounded by the flow rate utilized in the current piping analysis. Another component vMeh requires evaluation in this sy stem are the relief valves. the dynamic thrust load resulting from the opening of a valve is dependent on the flow area and the relief set prersure. As neither one of them changed due to the power uprate, the thrust load remains the same and therefore there is no change in the dynamic response of the piping. Ilased upon these comparisons, the current calculations for main steam piping bound the uprate condition. Table A shows the parameters of the main steam and reheat

steam, TAllLl! A COhipARISON OF ORIGINAL DESIGN AND UpRATED STEAh! pARAhtETERS UNIT I, UNIT 1 UNIT 2 UNIT 2, 13A4261 13A4261 13A5031 13 A5031 DESIGN UPRATED DESIGN UPRATED hiain Steam Pressure (psin) 750 740 4 750 746 2 Temperature ("F) 510.9 509.4 Sin.9 510.3

,Enthalpy (Blu/ito i197.9 1199I i197.9 l 199. 5 Sp= Vol. (Q'/lb) 060714 061632 0.60714 0.6117 hiass Flow Rate (Ib/hr) 11,710,47H 11,M64,730 11,710,47N 11,818,x70 Vol Flow Rate (D'/hr) 7,109,900 7,312,470 7,109,900 7,229,603 Extraction Steam For extraction steam piping systems, the uprate temperature is less than the current operating temperature except for the extraction steam to No. 6 lleater which has a 5 F increase in temperature. Therefore, the thennal expansion effect of the new temperature on the extraction piping system is negligible. The pressures of the extraction steam lines are low; therefore, the change from the current operating pressures to the uprate operating pressures are judged to le insignificant in the piping analysis. Since the extraction steam piping systems are traditionally not analyzed for steam hammer load, the change in the flow rate of the extraction steam due to the uprate condition does not have any effect in the current analysis. liased upon these reviews, the current calculations for extraction steam piping bounds the uprate condition.

8/5/97 Page 63

4 Condensate and Feedwater Systems Similar to the extraction steam, the uprate operating temperature is less than the current operating temperature except the feedwater from No. 611 eater to the steam generator, w here the temperature increases less than 5'F, Since.his temperature change is small, the current piping analysis is judged to be acceptable for the new operatiry temperature Deadhead pressure of the current condensate pump and the modified condensate pump is 595 psi and 628 psi, respectively. The increase in pressure is approximately 5%, anc.3.idged to be insignificant in the piping stress analysis. There is no increase of pressure in the feedwater system, similar to that of main steam.

The increase in flow rate of the condensate and feedwater is listed in Table B, in the design basis calculations, there is no dynamic analysis. (waterhammer analysis) for the condensate ud r dwater system.

ee TABLEB Feedwater Flow - Uprate Condition vs. Design and Current (Unit I and Unit 2) i Design Current Uprate "A DifTerence Heauer #

(Ib/hr)

(Ib/hr)

(Ib/hr)

(clesign-uprate) /

(current-uprate) 1 8,246,610 8,411.077 K,907,340 8.0 / 5.9 2

8.246,610 8,411,077 8,907,340 8.0 / 5.9 3

8,246,610 8,411,077 8,907,340 8.0 / 3,9 4

8,246,6 to 8,411,077 8,907,340 8.0 / 5.9 5

8,246,610 8,411,077 8,907,340 8.0 / $.9 6

11,650,000 l

l 1,691.080 12.303,I80 5.5 / 5.2 Based upon these reviews, the piping stress analyses for the condensate and feedwater systems are judged to be acceptable for the uprate condition. No new computer codes were used for the stress analysis of the piping systems for the purpose of evaluating the power uprate conditions, SCSlan - 7/24/97 NRC Ouestion No.14 Provide the calculated maximum stresses for the critical BOP piping systems, the allowable limits, the Code of record and Code edition used for the power uprate conditions. If different form the Code of record, justify and reconcile the differences.

S.NC Response No.14 As stated in the SNC response to additional question No.13, there was no new piping stress analysis performed specially for power uprate. He current piping stress analyses are based on parameters such as pr0 sure, temperature, and flow rate, which have been either reduced or changed insignificantly; therefore, the current analysis results are also acceptable for the uprate conditions,.

[

8/5/97 Page 64

O The code of record for the applicable sections of the main steam arid feedwater piping is ASME Section 111,1974 Edition with Summer '75 Addenda. The analyses were perfbrmed to the same code.

SCS/an. 7/24/97 NE_QQuestion No.15 Discuss the potential for flow-induced vibration in the balance of plant heat exchangers following the poiver uprate.

SNC Response No.15 Vibration in feedwater heaters can be predicted by comparing the cross flow velocity to the critical flow velocity of the drain cooling zone. The cross flow velocity is determined based on the original design flews by the feedwater heater manufacturer. The critical velocity is determined by the heater manufacturer based on drain cooling zone design. For this evaluation the uprate cross flow velocities of the drain cooling zones were calculated using the ratio of the original drain flow to the uprate flow which was multiplied by the original design cross flow. SCS feedwater heater standard specifications dictate that the critical flow of the drain cooling zone must be at least 1.35 times the actual cross flow pf the drain cooling zone. Each of the six feedwater heaters met this criteria at power uprate cceditions. Vibration of the drain cooling zone is not a concern based on drain cooling velocity requirements being met with significant margin.

Main condenser tube vibration is a function of steam velocity. For a fixed exhaust area, such as the condenser, steam velocity is a function of eteam volumetric flow, Therefore, the higher the volumetric flow, the more susceptible the tubes are to steam vibration. Volumetric flow rate was calculated for uprate and evaluated against the original condenser design. The evaluation concluded that the steam volumetric flow rate under uprate conditions (992 x 10' cf/hr) would only be slightly greater (0.6%) than current operation (986 x 10' cf/hr) and signincantly less than the original design conditions (1349 x 10' cf/hr), therefore condenser tube vibration should not be a problem.

The moisture separator reheaters cycle steam inlet volumetric flows increased by approximately 0.5 percent (1 ff/sec) over the current steam flowe. Cycle steam flow is below the maximum allowable chevron separator flow of 2.57 x 10' lb/hr. Low pressure heating steam volumetric flows increased by a range of 1.4 to 1.9 percent (~1 ff/sec) over the current steam flows. Due to the slight cycle steam flow increase, there should be no calculable loss in margin for tube vibration and no observable increase in tube wear due to the increased heating steam flow.

No physical changes are required by power uprate to the Service Water (SW) System or its components such as the system supplied heat exchangers. Therefore, no changes to the current design SW flowrates are expected with no increase in the potential for flew-induced vibration in any SW supplied heat exchanger.

No physical changes are required by power uprate to the Component Cooling Water (CCW)

System or its components. Herefore, no changes to the current CCW flowrates are expected with no increase in the potential for flow-induced vibration in any CCW supplied heat exchanger.

8/5/97 Page 65

-.9 To summarize, no signincant increase in flowrates to any BOP heat exchanger is expected due to power uprate.- Therefore, the potential of flow-induced vibration in these heat exchangers following power uprate is minimal.

SCS/dm 7/lk/97 REFERENCES FOR Tile NRC ADDITIONAL OUESTIONS 1.

Letter, Southern Company to the NRC, '? Joseph M. Farley Nuclear Plant Units I and 2, License Nos.- DPR-58 and DPR 74, Proposed Facility Operating Licenses and Technical Speci0 cation Change Request for Power Uprating," dated February 14,1997, with attachments.

2.

Westinghouse Electric Corporation, WCAP-147239 " Joseph M. Farley Nuclear Plant Units I and 2 Power Uprate Project NSSS Licensing Report," dated Jam,ary 1997 (Attachment V to Reference 2).

3.

"Farley Nuclear Plant Units I and 2, Power Uprate Project BOP Licensing Report" (Attachment 6 to Reference 1).

K/5/97 Page 66 m

o a

a ATTACitMENT 11 1

SNC Response To NRC Supplement Request For Additional Information l

l Related To Power Uprate Submittal - Joseph M. Farley Nuclear Plant, Units 1 & 2 1

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