ML20203K117

Forwards Response to 980210 & 13 RAI Re Power Uprate License Amend Request Allowing Operation of Increased Reactor Core Power Level of 2775 Mwt

ML20203K117
Person / Time
Site:	Farley
Issue date:	02/26/1998
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9803040465
Download: ML20203K117 (22)

Text

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Dave Morey Southern Nucl:ar See Presdent Operating Compa:y Farley Project P0, Box 1295 -

Birmingham. Nabama 35201 i

Tel 205 992.5131 February 26,1998 l

Docket Nos:

50-348 SOUTHERN 50-364 COMM Energy to Serve nur%rld'*

U. S. Nuclear Regulatory Commission 10 CFR 50.90 ATFN.: Document Control Desk Washington, DC 20555 Joseph M. Farley Nuclear Plant Response to Request for Additional Information Related to Power Uprate Facility Operating Ljcenses and Technical Specifications Change Request Ladies and Gentlemen:

Dy letter dated February 14,1997, Southern Nuclear Operating Company (SNC) proposed to amend the Facility Operating Licenses and Technical Specifications for Joseph M. Farley Nucicar Plant (FNP) Unit I and Unit 2 to allow operation at an increased reactor core power level of 2775 megawatts thermal (Mwt). NRC letters dated July 1,1997; August 21,1997; and October 14, 1997 requested SNC provide additional information. SNC responded by letters dated August 5, 1997; September 22,1997; and November 19,1997 respectively. SNC letters dated December 17 and 31,1997; January 23,1998; and February 12,1998. responded to NRC questions resulting from conference calls. During telephone conference calls on February 10 and 13,1998, SNC responded to additional NRC Stafiquestions. Attachment I provides the SNC responses to these questions, Attachment II includes corrections to the power uprate BOP Licensing Report (page 62). Attachment 111 provides requested information associated with platcout and containment spray system iodine removal rates.

If you have any questions, please advise.

Respectfully submitted, f}h

]>yw Dave Morey Sworn to andsubscribed before me this %dayo ch 1998

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Notary Public 6

My Commission F.rpires:

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Mr. L. A. Reyes, Region 11 Administrator Mr. J.1. Zimmerman, NRR Project Manager Mr. T. M. Ross, Plant Sr. Resident Inspector fhfflff[llll[llll 9803040465 980226 PDR ADOCK 05000348 P

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ATTACHMENT I I

SNC Response to NRC Request For Additionalinformation Related To Power Uprate Submittal. Joseph M. Farley Nuclear Plant, Units 1 and 2 i

SNC RESPONSES TO NRC QUESTIONS RESULTING FROM NRC/SNC CONFERENCE CALLS ON FEDRUARY 10 AND 13,1998 o

SNC Response to NRC Request For AdditionalInformation Related To Power Uprate Submittal-Joseph M. Farley Nuclear Plant, Units 1 and 2 NRC Ouggion No.1 (Reference February 10.1998 NRC/SNC Conference Call and February 11.

)

1998 NRC Facsimile)

It was indicated that the control room X/Q values were revised and were based upon one year's worth of meteorological data, April 1972 - March 1973. Provide the basis for concluding that the control room atmospheric dispersion factors should be based upon this one year's meteorological data and that this data is representative of the meteorolegical conditions which are anticipated to occur over the 40 year life of the facility, if this is an update of previous values due to re-assessments of the data or correction ofinformation, provide the difference between the present and previous values and the time when the revisions were made and/or any items corrected.

SNC Response No. I From the evaluation documented in Safety Evalntion Report, dated May 2,1975, and the FS.AR, it was concluded that one year of meteorological data was representative of the meteorological data over the 40 year life of the facility for determining the control room atmospheric dispersion factors.

Section 2.3 of the Safety Evaluation Report states the onsitejoint frequency data from April 1971 through March 1973 provided an acceptable basis to make conservative and representative estimates of atmospheric dispersion characteristics for accidental and routine gaseous releases.

FSAR Section 2.2.2 compares the data from April 1,1971 through March 31,1972 with the data from April 1,1972 through March 31,1973. Analysis of this data showed close similarity between the twe data sets.

During the FNP control room ventilation system self assessment, it was identified that the atmospheric dilution factors (X/Q) used in the control room dose assessment were not based upon as-built conditions. To be consistent with the FNP licensing basis as described in FS AR Section 9.4.1.6.2.3, the atmospheric dilution factors were re-assessed based on the meteorological data l

from April 1972 through March 1973, as well as the as-built configuration of the control room intake. Thejoint frequency data obtained from April 1972 through March 1973 is referenced in FSAR Table 2.3-8A.

Original Control Room Atmospheric Dilution Factors 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3.0 x 10s/m 2

2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.9 x 10s/m' 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.2 x 10s/m' 1 - 4 days 8.6 x 10" s/m' 4 - 30 days 4.1 x 10"s/m' Re-assesscd Control Room Atmospheric Dilution Factors 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3.28 x 10s/m' 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.65 x 10s/m' 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2.19 x 10s/m' 1 - 4 days 1.64 x 10s/m' 4 - 30 days 1.08 x 10-3s/m' Page1

he above re-assessed control room atmospheric dilution factors have been included in Resision 14 of the FGAR for Unit I (FSAR Table 15.4 16A).

II/dm 2/23/98 NRC Ouestion No. 2 (Reference February 10.1998 NRC/SNC Conference Call and February 11.

1998 NRC Facsimile)

For the jRCP) locked rotor accident, is there any fuel melting which occurs?

k SNC Response No 2 1

ne review of the RCP locked rotor analysis results confirmed that fuel melting will not occur during this postulated ANS Condition IV event.

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2/I3/98 & SNC/mge 2/I3/98 HRC Ouestion No. 3 (Reference February 10.1998 NRC/SNC Conference Call and February 11.

- 1998 NRC Facsimile) lias it been confirmed by analysis that in the event of a SGTR,'there will be no tube uncovery nor immediate flashing in the steam generator with the rupt : red tube? It is stated on page 58 of the BOP Uprate Licensing Report that uncovery was evaluated, but there is no discussion of the results of the evaluation which was performed relative to uncovery and flashing. Were cases evaluated [where] the PORV of the faulted steam generator failed closed in one case and failed open in another case?

l SNC Response No 3

-Westinghouse confirmed on a generic basis, applicable to the Farley units, that the effects of l

part? steam generator tube uncovery on the iodine release for an SGTR is negligible. This basis is documented in WCAP-13247, " Report on the Methodology for Resolution of the Steam Generator Tube Uncovery issue," dated March 1992, with NRC approval of the submittal and l

- agreement with the conclusion provided in a letter dated March 10,1992. He supporting analyses included considemtion of a stuck open PORV on the ruptured steam generator. Some flashing of break flow would be expeuoi (see SNC response to Question No. 7 below) but is not considered as part of the Farley licensing basis analysis.

A single case evaluation, which considers a stuck open PORV, was performed similar to that presented in the current FSAR, but using power uprate mass flow rates. No studhs of additional single failures were performed. RCS activity was assumed to be at the Technical Specifications l-limit (0.5 Ci/gm, which has since been reduced) with no iodine spike, secondary side activity at the Technical Specifications limit (0.1 pCi/gm), primary to secondary leakage to the intact generators at the Technical Specifications limit (150 gpd/ generator), and flow from the ruptured tube based on power uprate (150,000 lbs in 30 minutes). Leakage from the uncovered tubes is modeled as a direct 100% release to the atmosphere for 30 minutes without partitioning. The results continue to meet the FSAR results, i.e., a small fraction of the 10 CFR 100 limit.

W/ub & jm 2/12/98 & SCS/ jaw - 2/12/98 Page 2 k

NRC Ouestion No. 4 (Reference February 10.1998 NRC/SNC Conference Call and February 11.

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1998 NRC Facsimile) l For the power uprate amendment, the control room volume is indicated as i 14,000 f1. Previous 5

IPC amendments and UFS AR had indicated a volume of 69,000 fi'. What is the basis for the significant increase in volume?

SNC Response Nod Daring the FNP control room ventilation v;r assessment, it was identified that the control room volume did not include other volumes wtn which the control room proper communicates, e g., the volume above the ceiling tiles. As a result, subsequent control room dose calculations evaluated the consequences of the accident based on a control room volume of 114,00011'. This value is included in FSAR Table 15.4 16A, Revision 14, for Unit 1, ll/dm 2/11/98 NRC Ouestion No 5 (Reference February 10.1998 NRC/SNC Conference Call and February 11.

1998 NRC Facsimile)

The doses calculated for a locked rotor accident by the staff are a facor of 10 lower than those calculated by the licensee if a partition factor of 0.01 is used, but approximately the same if a partition factor of 0.1 was used. Was a partition factor of 0.1 actually utilized by the licensee for the locked rotor accident rather than a factor of 0.0l ?

SNC Resnonse No. 5 A partition factor of 0.01 was used in the calculation. However, during the review to verify the partition factor, it was discovered that the source term had been inadvertently used for a core melt in lieu of the correct value for a gas gap release. (See response to Question No. 2 above.) When the correct source term is used, the doses decrease by approximately a factor of ten. As a result of this finding, page 62 of the BOP Licensing Rt u ; has been revised. The revised page is provided in Attachment 11.

SCS/ jaw 2/12/98 NRC Ouestion No. 6 (Reference February 10.1998 NRC/SNC Conference Call and February 11.

1998 NRC Facsimile)

At what time after the accident are the containment sprays initiated, and at what time after the accident is recirculation from the containment sump initiated? It is the staffs understanding that recirculation must be manually initiated by the operators. Is this the case? Is there a period of time between the initial containment spray operation and the initiation of recirculation in which no spraying is occurring?

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1 SNC Pemonse No. 6

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Based on review of the containment analysis results, spray initiation occurs at approximately 56 j

seconds. His is modeled coincident with core melt at 0 secords for radiological consequences evaluation. Based on the draw down rate from the RWST, recirculation from the sump will start i

after approximately 20 minutes.

Following initiation of containment spray, the transfer of spray pump suction from the RWST to the containment sump is an evolution controlled by the plo,:mergency response procedures. The i

operator manually transfers the spray pump suction valve alignment without interrupting spray

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flow.

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SCS/ jaw 2/12/98 A SNC/mge-2/12/98 NRC Ouestion No. 7 (Refertitee February 10.1998 NRC/SNC Conference Call and February i L 1998 NRC Facsimile)

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l Provide the flashing fractions as a function of time in the faulted steam generr. tor for the SGTR I

accident for two cases. In the first case, the PORY for the faulted steam generator is assumed to

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fail closed. In the second case, the PORV for the faulted steam generator is assumed to fail open.

in both cases, the onset of the SGTR event is assumed to occur coincident with a loss of offsite power.

SNC Response No. 7 l

The Farlev SGTR analysis does not include flashing fractions as a function of time. As requested by tk NPC Staff, the attached figures provide examples of the break flow flashing fraction as a function of time for detailed SGTR transient analyses performed for plants which include flashing

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in the licensing basis calculation. These transients model the expected operator actions to terminate primary to secondary break flow, including isolation of tiv ruptured steam generator, cooldown with the intact SGs and depressurization with the presst azer PORV. Figure 1 assumes i

that a PORY cn an intact SG fails closes. Figure 2 assumes that the PORV on the ruptured SG fails open and requires operator action to close the associated PORV block salve. Both analyses assume a loss of offsite power. The loss of ofTsite power assumption for these analyses is the same l

as for Farley, which is discussed in Section 6.3 of WCAP-14723.

W/ub & jm. 2/12/98 f

NRC Ouestion No. 8 (Reference February 10.1998 NRC/SNC Conference Call and February 1 L l

1998 NRC Facsimile)

In Table 2 of the Enclosure of the December 9,1997 letter from D. N. Morey to the Staff, it was

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indicated that the control room pressurization filter flow was 450 cfm for the amendment request involving the spray additive tank removal, the power uprate and the ESF filters. However, the technical specification change proposed for the ESF filters amendment indicates a flow of 300 cfm.

Explain the basis for the differences. Which value is the one which is utilized in the analysis by the licensee? Previous staff analyses had determined that a flow rate of 270 cfm (300 cfm -10%) was more limiting from a dose standpoint.

Page 4 1

SNC Response No. 8 De Technica' Specifications limit remains 300 cfm i 10% 450 cfm is the maximum calculated fan runout capability that was used in our calculations. Previous parametric studies by the licensee, and the lodine Protection Factor (IPF) formula from " Nuclear Power Plant Control Room Ventilation System Design For Meeting General Criterion 19" by K. G. Murphy and Dr. K. M. Campe, indicate that higher intake flow results in lower IPF and higher doses.

SCS/ jaw 2/12/98 NRC Ouestion No. 9 (Reference February 10.1998 NRC/SNC Conference Call and February 11.

1998 NRC Facsimile)

Table 2 of the December 9,1997 letter from D. N. Merey indicates that the ECCS leakage was assumed to be [20) times the leakage presented in Table 6.3-8 of the UFSAR. However, Table I of the June 30,1997 letter on the ESF filter amendment request indicates that ECCS leakage is 10

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times the value in UFSAR Table 6.3 8. Which value should be used in the accident analyses? At -

l what leakage level does the TMI Ill.D.I.1, Leakage Reduction Program, require maintenance actions to repair leaking systems and/or components?

SNC Resoonse No. 9 The ECCS recirculation loop leakage value assumed for the ESF filter amendment radiological evaluations was 10 times the total shown in FSAR Table 6.3 8. His value was doubled (i.e.,20 times the Table 6.3-8 total) in accordance with SRP 15.6.5, Appendix B, for use in accident analysis dose calculations.

The Farley " Borated Water Leakage Assessment and Evaluation Program"is described in manual FNP-0-M 101. The program manual provides guidelines for identificatica of boric acid leaks and.

assessment of repair need and component wastage. Identified active leaks are documented upon -

discovery and subsequently trended. _ If the total trend from all leaks approaches the 3760 mis /hr i

criterion, then corrective actions are initiated by the plant staff. For any one leakage source, significant leakage is defined as approximately 10 drops per minute or greater. Significant leakage requires initiation of a Deficiency Report (DR), which is used to facilitate necessary repairs.

SCS/ jaw - 2/12/98 & SNC/mge & rim - 2/21/98 NRC Ouestion Na 10 (Reference February 10.1998 NRC/SNC Conference Call and February 11.1998 NRC Facsimile)

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i What are the differences in the control room filter unit and the recirculation filter unit and how are

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- they aligned in the emergency mode of operation?

Page 5

i SNC Resnonse No.10 Ebch train of the control room IIVAC system has three safety-related Olters in the emergency mode, the pressurization unit supplies Oltered outside air to pressurize the control room with a Dow rate as described in the response to Question No. 8 above. Also, in the emergency mode, the recirculation and Oltration units provide recirculation filtration in parallel for the air inside the control room with a total flow rate of 3000 cfm i 10%

SCS' jaw 2/12/98 NRC Ouestion No. I1 (Reference February 10.1998 NRC/SNC Conference Call and February 111998 NRC Facsimile)

In Table E of the Attachment 5 of the August 5,1997 letter from the licensee on the power uprate,it is indicated that the steam released from the faulted steam generator in the event of a MSLB accident is 473,000 lbs over 30 minutes. Previous information provided in various IPC amendment requests indicated that the release from the faulted steam generator would be [96,200] lbs plus all primary e secondary leakage. For the power uprate amendment request, how does the increase iri power level affect the teleases from the intact and faulted steam generators and what are the values for releases to these two sources? What values were utilized in the power uprate amendment request?

I-SNC Resnonse No. I1 For the various IPC submittals, the initial steam generator mass of 96,200 lbs was used for dose analyses. This mass corresponds to a pre-power uprate full power value which is appropriate for these calculations. For the power uprate, since the assumed activity levels are enchanged and the MSLB releases are based on a zero power steam generator mass, the increase in power level has insignificant impact on the releases. The power uprate steam releases for the faulted generator are the initial inventory of 168,000 lbs plus main and auxiliary feedwater flow until isolation at 30 minutes of 290,000 lbs. The more conservative steam generator mass co. responds to a hot zero power value. The flow is conservatively increased by 5% for evaluation of radiological consequences [i. e., (168,000) + (1.05) (290,000) a 473,000). The intact generator mass releases for power uprate are 323,000 lbs (0 - 2 hr) and 695,000 lbs (2 - 8 hr), which were also conservatively increased by 5% for evaluation of radiological consequences.

SCS/ jaw 2/25MS NRC Ouestion No.12 (Reference February 10.1998 NRC/SNC Conference Call and February Ili 1998 NRC Facsimile)

What was the basis for [the building volume assumed for] dilution for the fuel haadiing accident?

SEC Resnonse No.12 The building volume considered the open area directly above the spent fuel pool area bounded by concrete walls. It does not include any adjacent areas such as the new fuel storage area, heat exchanger or pump rooms, HVAC equipment rooms, etc. For two trains of exhaust (2 x 4000 cfm assumed), the building volume is exhausted over nine times in two hours, consistent with Regulatory Guide 1.25.

SCS/ jaw 2/12SS Page 6 o

i NRC Ouesdon No.13 tReferrDg+1cbruarv_10._1998 NRC/SNC Conferrassfall) nere are differences betw:en the iodine removal rate functions calcutated by the staff and those provided by the licensee. In particular, the elementaliodine spray removal A and deposition A are I

different. Provide the bases for your values.

SNC Respanne No.13 l

l ne spray remov J A was chosen based on values iom NURE0/CR 0009, This A is considered to be a conservatively small value wh3 h might occur for boric acid spray water contaminated with iodine. De deposition A was cabulated in accordance with the methodology described ir.

NUREG 0800, Section 6.5.2, Revision 1. De total elemental removal rate assumed in our analysis (spray plus deposition) is conservatively lower than the Staff velues discussed in our telephone call of February 10,1998. As requested by the Staff, pertinent pages from the l

calculation are included in Attachment 111.

SCS/ jaw 2/12/98 NRC Outstion No.14 (RefstvDcclebruary 13.1998 NRC/SNC Conference Call)

Regarding modeling assumptions r non LOCA events that kssume LOOP and ESF actuation, what is the sensitivity of the trans

sults to the time of LOOP.

SNC.RcspoMC.B93.4 Previous responses have addressal v:hich non LOCA events are analyzed with a loss of offsite power (LOOP), For Failey, the events analyzed with a LOOP are the loss of normal feedwater, feed line break, steam line break, and locked rotor. With the exception of the locked rotor, there 3

ever.ts are ones that assume a Enginected Safety Features (ESP) system actuation. His response

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will address the sensitivity of the transient results to the time of LOOP, specifically, whether or not j

tiu LOOP occurs at event initiation, at the time of reactor trip, or, as typically assumed in j

Westinghouse non LOCA analyses, at approximately 2 seconds following reactor trip.

De LOOP affects the analysis assumptions primarily in two ways: (1) the time the RCPs are tripped, and (2) the ESF ftmetions are delayed or sometimes diminished (e g., longer Si delcy due to dicsci sequencing, diminished AFW capacity due to single failure consi& rations, etc.). ne assumd time delay of RCP trip is 'ypically the same for non LOCA analyses, his delay time is approximately 2 seconds afler reactor trip. Following RCP trip, core Dow decreases as the RCPs begin to coast down. De other aspects of the LOOP, such as diesel segencing ar,d AFW startup delays and Dow rates tend to be plant specific and are discussed in the Farley FSAR and NSSS Licensing Report: It should be noted, however, that no assumption with respect to LOOP was changed for the Farley Units I and 2 Uprating Program.

Steam Line Dreak (FSAR 15 A.2)

For the steam line break event, the LOOP occuo 3 seconds after event initiation or approximately

-- 2 seconds (actual is 2.2 seconds) after the low pressurizer pressure safety injection setpoint is Page 7

reached. This is consistent with standard Westinghouse steam line break analysis methodology and is repor*cd in WCAP 9226, Rev. l.

For the steam line break core response analytis, the case with offsite power available is thc limiting case. Changing the time of the LOOP to the beginning of the event will not change conclusions regarding the core response and DN!! analysis, nor change the limiting case from that which has offsite power available.

Loss of Nonnal Feedwater (FSAR 15.2.8 and 15.2.9)

Two loss of normal feedwater cases are analyzed: (1) loss of normal feodwater with offsite power (LONF, FS AR l$ 2.8) and (2) loss of ncnnat feedwater with a loss of offsite power (LOOP, FSAR l$.19). The LONF case is analyzed to ensure that the ARY systern van remove the core decay heat and RCP heat. He LOOP case is analyzed to ensure that the ARV system in conjunction with the primary side under a natural circulation How regime can remove the core decay heat. %c analysis applies a conservative acceptance criterion of ensuring that the pressurizer does not fill. Since both FSAR events have the smne ruumptions with respect to the ARV system, the LONF case is the more limiting of the two cases due to the RCP pump heat addition.

The LOOP cvent is not particularly sensitive to the tirne the loss of offsite power occurs. The transient is relatively long in duration with the linuting condition (approach to y essorizer Gil) being reached at 1466 seconds. De reactor trip and ARV initiation occurs as result of the low.

Iow s' cam generator water level protection. The ARV actuation delay is 60 seconds. For the case presented in the FSAR, the loss of offsite power is e.ssumed to occur 2 seconds after reactor trip.

Assuming that the loss of offsite power occurred exactly at the time the trip setpoint is reached would provide essentially the same results.

Analping the event with the LOOP at event initiation would result in reactor trip on low reactor coolant Dow very early in the transient. He pre trip portion of the event analyzed in this manner would be similar to and no more limiting than the complete loss of How event presented in FSAR Section 15.3.4. He reactor trip would occur on low RCS How. ARV initiation continues to be a result of the low low steam generator water level protection As to the post-trip, long tenn cooling elTects, modeling the event with a kss of ofTsite power at event initiation is less limiting than the LOOP event as currently analyzed since the Steam generators enter the post trip (core deca; heat removal) phase of the transient with much more steam generator water inventory.

He evaluation statements made abos c were venfied with several LOITRAN simulations.

(LOFTRAN is an NRC approved non LOCA Code ) The results of these simulations are graphically shown on the attached Sgure, Figure 3 shows a plot of pressurizer volume versus time for the FSAR case (LOOP at reactor trip plus 2 seconds), LOOP at reactor trip, and LOOP at event initiation. Note that the two cases LOOP at RT plus 2 seconds and LOOP at RT are essentially the same. Assuming the LOOP at event initiation is not limiting.

Fmi Line Urcak (FS AR 15.4.2) ne feed line break (FLB) accident, similar to the LONF and LOOP events, is analyzed to ensure that the AFW rystem can remove the core decay heat for the LOOP case and the core decay heat Page 8

d and RCP heat for the case with offsite power. For the case analyzed with a loss of offsite power, the LOOP is assunied to occur 2 seconds aller reactor trip occurs.

Foi reasons similar to the LONF event, the FLB cvent is also not sensitive to the time of the LOOP. In all cases main feedwater is terminated at the time the feed line break occurs, and the AFW actuation delay is relative to the time the low low steam generator water level signal is reached. De limiting conditions are reached at 900 seconds or more after the break occurs.

Derefore, a LOOP at the time of the reactor trip with no delay would result in essentially the same result as die case with the LOOP delayed by 2 seconds. Also similar to the LONF w/ LOOP, should the LOOP occur at event initiation, the transient conditions would be much more favorable than those currently analyzed. This result is primarily due to the steam generators entering the -

post trip (core decay heat removal) phase of the accident with much more steam generator water -

mass available to remove the decay heat.

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4 ATTACilMENT 11 SNC Response to NRC Request For Additional information Related To Power Uprate Submittal. Joseph M. Farley Nuclear Plant Units I and 2 CORRECTED PAGE NO. 62 "FARLEY NUCLEAR PLANT UNITS 1 AND 2 POWER UPRATE PROJECT DOP LICENSING REPORT" (A1TACllMENT 6 TO SNC SUllMITTAL DATED FEBRUARY I4,1997)

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The potential impact of uncovery of the steam generator tubes during the event was also evaluated for uprated conditions. Assuming technical specification limits for RCS activity (0.5 pCi/gm) and l

leak rate (150 gpd per generator) and release directly to the environment (.c. no mixing with the i

secondary side water) the offsite doses remain well within the 10 CFR 100 guidelines.

2.16.7.3.8 Evaluation of the Radiological Consequences of an RCP Locked Rotor l

%c radiological consequences of RCP locked rotor releases assuming 20% of the fuel clad / pellet gas gap is relcased to the RCS with subsequent leakage to the steam generators and secondary side steam relcares were evaluated utilir.ing the assumptions of Standard Review Plan Section 15.3.3.

Dese releases result in offsite doses that are a small fraction of the guidelmes of 10 CFR 100, j

which meets the acceptance criteria.

Thyroid Dose Whole Body Beta Skin (Rem)

Dose (Rem)

Dose (Rem)

EAB 1.62 0.20 0.18 LPZ 3 02 0.11 0.09 2.16.8 Summary of Conclusions No changes or additions to structures, equipment, or procedures are necessary to provide adequate radiation protection for the operators or the public during normal or post accident operations to support the uprate. The existing structures, systems, and componer.ts can safely handle the changes in post accident source terms and releases from the uprate conditions, and resulting onsite and offsite doses are less than the 10 CFR 100.11 guidelines and are within the Standard Resiew Plan guidelines. Therefore the radiological consequences acceptance criteria fc postulated Condition II,111, and IV events are satisned.

1 BOP UPRATE LICENSING REPORT 62 FNP - UNITS 1 AND 2 ov20/98 l

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ATTACllMENT 111 SNC Response to NRC Request For AdditionalInformation Related To Power Uprate Submittal. Joseph M. Farley Nuclear Plant, Units I and 2 EXCERPT FROM FARLEY NUCLEAR PLANT SCS CALCULATION SM 95 8931002 0FFSITE AND CONTROL ROOM DOSE FOR UPRATE WITil TSP FOR pil CONTROL (Sheets 2,3. 4,5, & 10)

Southern Company Services hojn t

( ahulation Nunder i atley Nuclear I'lant SM 95 8931002 suNuehuc sheci Off site and Control Room Dose for Uprate with TSP for pil Control 2 of 17

Major ikluations :

The calculations were performed using the TACTS computer program running on an NIIC pentium desktop computer. The performance of TACTS on the NIIC machine was verified in reference 1.

The test problems of reference I were successfully rerun on the computer to verify proper execution 1

of the program. A directory listing of the TACTS.liXII and library files is included in Attachment 1.

These files were compared to the listings in reference I to verify the proper files are installed.

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The equations used by the TACT 5 computer program are described in reference 2.

i Assumptions :

1. To aid retention ofiodine in the sump, trisodium phosphate (TSP) will be added to the sump solution in sufficient quantity to maintain a pil of 7.5 (Ref 18d). Determination of the quantity of TSP to be added is not within the scope of this calculation. The effectiveness of the iodine retention as tellected in the decontamination factor and removal process cutoffs will be determined in accordance with the guidance of reference 4.

2. Removal of elemental iodine by the boric acid spray injection solution will be assumed to have a removal constant of 1.4/hr. Reference 9 indicates a conservative constant of 0.9/hr and reference 8 (section 6.1.11) indicates a value of 1.44/hr, with an expected value 7.8 times this value. Aoditional research (reference 14) indicates that injection of fresh, uncontaminated by iodine, spray solutions are elTective with or without additives. A value of 1.4/hr is chosen as a conservative estimate.

3. Coatings typically have plateout retention capacity well in excess of the inventory released (reference 8, page 68). Thus all iodine plated out will remain on the plateout surface.

4. The plate out (deposition) removal constants from will be estimated Ibr this calculation. The deposition velocity used to derive these values (0.68cnt see ihr zinc >zine coated surfaces and

/

0.49 cnt'see ihr epoxy coated surfaces) are consistent with the data presented in reference 8, Table 5 for Dimetcote, Corbo.zine and Amercoat 66 which are similar to the coatings used at FNP. This is assumed to reduce to approximately 5% of the estimated rate after reaching 1% of the initial concentration and to stop at 0.1% of the initial concentration (references 8 and 13).

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( miculathm Numhet l'atley Nuclear Plant SM 95 8931002 sutyctc hile sheci of hite and Control f(oom Dose for Uprate 5,ith TSP for pil Control 3 of 178

5. No credit for removal of organic iodine is taken, nor is credit taken for removal of elemental or particulate lodine below the assumed removal cutoff of 1000. Cutoff times are determined based on im concentrations, l.c. Ignoring decay of short lived isotopes. Itemoval coefficients and removal process cutoffs used per assumptions 14 are shown in Table 1. Since time dependent plateout is modeled (in lieu or instantaneous 50% ), and organic iodine is not removed (except via leakage) the core releases to the containment are modeled to maintain the same organic source term as discussed by reference 7, i.e. 50% total release as 95.5% elemental,2% organic, and 2.5% particulate (references 7,13,20).

6. The sump pil reduction (from previous NaOli addition levels) does not impact the containment pressure / temperature response; thus the sump (recirculation fluid) temperature and flashir Taction, and the !!CCS leakage contribution to the total dose, are mode led as described in reference 10. IICCS l

leakage, taken from reference 3a, is assumed to be 10 x 4000 cc/hr. This conservatism is intended to avoid any requirement to closely monitor, or have an explicit Technical Specification on, liCCS i

leakage.

7. Ilydrogen purge may be initiated as a backup to the redundant hydrogen recombiners. The initiation time (18 days)and flow rate required (35 cfm) are taken from reference 17.

l

Southern Company Services hiijcst t alculatum Number l'arley Nuclear Plant SM 95 8931002 EM'litic sheet Olhite and Control Itoorn Dose for Urrate with TSP for pil Control 4 of 178 References :

1 Nuclear Support calculation number N 94 02, " Verification of TACTS," revision 0, 1

2 NURiiG/CR 5106 SAIC-88/3023. " User's Guide for the TACT 5 Computer Program."

i 3

1 NP Final Safety Analysis Report

a. Table 6.3 8

b. Table 15.414

c. Table 15.4-16

d. Table 15.4 20

c. Table 15112

f. Figure 3.7 20 g, Table 6.2-2

h. Table 6.2 5

i. Table 15.418 1

4 NURiiG.0800, "U.S. Nuclear Regulatory Commission Standard Review Plan," Section 6.5.2, f

Revision 1.

5 10 CFR 50, Appendix A, General Design Criterion 19," Control Room."

6 10 CFR 100.11. " Determination of exclusion area, low population zone and population center distance "

7 Regulatory Guide 1.4," Assumptions Used for livaluating the Potential Radiological Consequences of a 1 oss of Coolant Accident for Pressurized Water Reactors," June,1974.

8 NURiiG-CR 0009," Technological llases for hiodels of Spray Washout of Airborne Contaminants in Containment Vessels," October 1978.

9 WASil 1329,"A review of hiathematical hiodels for Predicting Spray Removal of Fission Produc t in Reactor Containment Vessels," June, 15,1974, 10 NURl!G 0800,"U.S. Nuclear Regulatory Commission Standard Review Plan," Section 15.6.5, Revision 1.

Ii 1.etter AP 21370, dated February 6,1996,"Up-date Control Room Dose Assessment."

12 hlurphy, K.G. and Campe, Dr. K.ht.,13th AFC Air Cleaning Crnferenes," Nuclear Power Plant Control Room Ventilation System Design for hiceting General Design Criterion 19."

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( akulation Numtwr l'ariry Nuclear Plant SM 95 8931002 suNea/i nte sheet off$ite and Control Room Dose for Uprate with TSP for pil Control 5 of 178 13 WCAP-11611, hlarch 1988,"hiethodology for Elimination of the Containment Spray Additive."

14 Davis, it.E., et al," Fission Product itemoval liffectiveness of Chemical Additives in PWit Containment Sprays,"Itchnicallkport A-3788. 8/12/86, attached to proposed revision 2 to Standard lleview Plan 6.5.2, with AIF letter of 5/l l/87.

15 1.etter AI.A 95 756, dated 12/15/95, Analysis input Assumption 1.ist 16 1 etter AI.A 99 508, dated February 1,1966," Final Core inventory Source Terms."

17 Calculation 40.05,llevision 3. Post Accident llydrogen Generation Analysis, ifs FNP Calculations

a. hicchanical calculation 4.2

b. hiechanical calculation 4.1

c. Sht 93-0121001

d. Sht 95 8931001 19 FNP Technical Specifications I

a. 3/4.6.1.2 20 NllitEG 0588 21 llegulatory Guide 1,52," Design, Testing and hiaintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units ofl.ight Water Cooled Nuclear Power Plants,"llev. O.

Southern Company Servlees Projest L miculatum Number l'arley Nuclear Plant SM 95 8931002 subg /liile shee Offsite and Control Room Dose for Uprate with TSP for pil Control 10 of 178 i

The minimum pil is maintained at 7.5 as discussed in assumption 1. The partition factor between liquid and gas phases is based on reference 4 Figure 6.5.21. With a pil of 7.5, the partition coefilcient is 440.

The elemental lodine spray removal coefficient is 1.4 hf' per assumption 2. With a partition coefficient of 440, the decontamination factor (DF) limit is based on reference 4.

DF= 1 + (4.921!+4 83)(440)/(1.671?+6 A3) = 14.0 where the sump and containment volumes are as provided above.

lilemental iodine plateout is calculated per assumption 4, with efTective plateout areas taken as the containment heat sinks (Ref 3g):

Zinc /zine painted surfaces from heat sinks 1 and 415 = 2.70 x 10' 0 2 4 2

!?poxy surfaces from heat sinks 2 and 3 = 6.47 x 10 f1 Then per references 8 and 13, A = 118 I(Deposition velocity x Area / Volume)

{

= 118 (0.68 x 2.70 x 10 + 0.49 x 6.47 x 108.) = - 12.5 hf' 3

2.03 A 10' This decreases to approximately 5% of the initial value or about 0.5 hf' after reducing the original concentration by 100, and to 0 aller a reduction of 1000.

4 i

The particulate spray removal coeflicient is calculated as described in reference 8 (page 118):

A = 1(100 lb(2175 epm)(0.1 smk. x fdLmin x 30.5 cm = 4.77 hf' f

2(1.669116 A )(7.5 gal /ft )

hr ft where 0.1 cm is a conservative washout parameter (Fid) from reference 8, section 5.3.1 (p 34), until the particulate DF = 100. Aner this time the value decreases by a factor of ten, until a DF of 1000 is achieved.

Drop fall height is assumed to be 100 8 based on reference 3f, and spray flow of 2175 gpm is based on references 3h and 18e.

These values are input to TACTS which is rtm iteratively to determine the cutoff times as described in assumptions 15 above. The removal rates and cutoff times are shown below:

L h

,,._..,.n...,.,,

.,,.n--,

.. ~., -

--,rn,r, n

,-n....a