ML20217P449

Responds to RAI Related to Power Uprate Facility Operating Licenses & TS Change Requests

ML20217P449
Person / Time
Site:	Farley
Issue date:	03/03/1998
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9803100270
Download: ML20217P449 (17)

Text

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., Dan Morey Southern Nuclear i

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%ce President Operating Company Fadey Project -

P.O. Box 1295 Birmingham. Alabama 35201 Tel 205.992.5131 March 3, 1998 Docket Nos: 50-348 SOUTHERN h 50-364 C,OMM Energy to ServeYourWorld" U. S Nuc. lear Regulatory Commission 10 CFR 50.90 ATIN.: Document Control Desk l Washington, DC 20555 Joseph M. Farley Nuclear Plant  !

Response to Request for Additional Information Related to Power Uprate l Facility Operatina Licenses and Technical Specifications Chanae Reauest i

Ladies and Gentlemen:

By letter dated February 14,1997, Southern Nuclear Operating Company (SNC) proposed to amend the Facility Operating Licenses and Technical Specifications for Joseph M. Farley Nuclear Plant (FNP) Unit I and Unit 2 to allow operation at an increased reactor core power level of 2775 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 and 26,1998 responded to NRC questions resulting from conference calls. During telephone conference calls on February 25,1998 and i March 2,1998, SNC responded to additional NRC Staff questions. Attachment I prosides the SNC responses to these questions. Attachment 11 includes corrections to the power uprate BOP Licensing Report (pages 60,63,64,66, & 86). Attachment 111 provides up-dated information associated with plateout and containment spray system iodine removal rates. ,

if you have any questions, please advise.

Respectfully submitted, ,

D Ulcu  !

Dave Morey Sworn to and subscribed bef e me this3_kay off]fl998 I M a, 0u D /

k No'aryPublic t (/ f()0I My Commission Expires: & /, 300l L

MGFJmaf: pwrup32. doc Attachments cc: Mr. L. A. Reyes, Region 11 Administrator Mr. J.1. Zimmerman, NRR Project Manager Mr. T. M. Ross, Plant Sr. Resident inspector 9803100270 990303

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i ATTACHMENT I  !

SNC Response to NRC Request For Additional Information Related To Power Uprate Submittal- Joseph M. Farley Nuclear Plant, Units 1 and 2 SNC RESPONSES TO NRC QUESTIONS RESULTING FROM NRC/SNC CONFERENCE CALLS ON FEBRUARY 25,1998 AND MARCH 2,1998 I

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SNC Response to NRC Request For Additional Information Related To Power Uprate Submittal- Joseph M. Farley Nuclear Plant, Units 1 and 2 NRC Ouestion No.1 (Reference February 25.1998 NRC/SNC Con.fgrence Call For evaluation of LOCA radiological consequences, there are differences between the iodine removal rate functions calculated by the NRC Staff and those provided by the licensee. In particular, the elemental iodine spray removal A and deposition A are different. Provide tle basis for your values.

SNC Response No.1 The spray removal and deposition A values associated with FSAR Table 15.4-14A, " Parameters Used in the LOCA Analysis (Unit 1)," and the power uprate evaluation were calculated in accordance with WCAP-11611, " Methodology for Elimination of the Containment Spray Additive," March 1988, as indim~l in the current revision of the Farley BOP Licensing Report and pertinent pages of SCS Calculation No. SM-95-8931-002 (Reference Attachment III of SNC letter dated February 26,1998). Based on discussions with the staff related to thejustification for using WCAP-11611 deposition velocity, the SCS calculation for FNP "Offsite and Control Room Dose for Uprate with TSP for pH Control" has been revised to incorporate the deposition velocity described in NUREG-0800. Updated BOP Licensing Report pages (60,63,64,66 & 86) showing the resultant spray removal and deposition A values and offsite and control room dose revisions are included in Attachment II. The revised pages of the referenced calculations are provided in Attachment Ill. These results meet the guidelines of 10 CFR 100.11 and GDC 19.

SCS/ jaw 3/1/98 NRC Ouestion No. 2 (Reference February 25.1998 NRC/SNC Conference Call)

With regard to the Containment Mini-Purge System and its incremental offsite dose contribution, what is the flow rate used in the dose analysis calculations and what is the assumed time of system isolation?

SNC Resoonse No. 2 The purpose of the incremental dose calculation for LOCA is to determme the contribution to offsite doses at the site boundary and LPZ that result from containment mini-purge operation prior to isolation. De ba6 for the incremental dose calculation results presented in FS AR Table 15.4- ,

15 are summarized m-part in APCo letter (F. L. Clayton, Jr.) to the NRC (Albert Schwencer), j dated February 5,1979. This calculation was not changed for power uprate because of the conservative nature of the input assumptions with respect to the present Farley design.

The calculation assumed two I8 inch vent lines and no filtration; whereas, the present mini-purge consists of an 8 inch supply line and an 8 inch exhaust line, which exhausts through the containment purge filter unit. In addition, no credit is taken for iodine removal by plateout; and the mini-purge isolation valves are assumed fully open for a duration of 6 seconds following accident initiation.

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4 The calculation is incremented to I second intervals over the 6 second period. Only the iodines associated with the incremental steam portion of the blowdown are assumed to have been released to the containment along with 100 percent of the noble gases in the incremental portion of the entire primary coolant blowdown. (Incremental is defined as the difference between the maximum value of the blowdown for the present time interval minus the maximum value for the previous time interval.) Source terms for the primary coolant blowdown were based on the reactor coolant equilibrium fission and corrosion product activities from FSAR Table 11.1-2. These source terms were based on parameters from FSAR Table 11.1-1, which includes I percent fuel clad defects.

The flow from mini-purge is postulated as two 18 inch holes in the containment wall. The method

- of analysis determines if the flow from containment is sonic or subsonic based upon the pressure ratio between atmosphere and containment pressure. Once the flow is classified, the velocity is calculated, and the vent cross-sectional area is applied to determine the flow rate through the two 18 inch vent lines. As the containment pressure increases, the flow rate through mini-purge increases The mirdmum and maxunum flow determined during the 6 second time interval is 120,000 cfm and 143,600 cfm, respectively.

B/dm - 3/02/98 NRC Ouestion No. 3 (Reference March 2.1998 NRC/SNC Conference Call)

In the response to Question No. 7 in SNC letter dated November 19,1997, SNC indicated that the principal codes used in support of the uprate licensing submittal are up-to-date with respect to applicable codes and standards and that these codes were used in conformance with their limitations and restrictions. Is this response also applicable to the previous license amendments  !

refarenced in the Farley uprate amendment request? l SNC Resnonse No. 3 The SNC position for the Farley uprate licensing submittal is that the principal codes used in  !

support of or referenced in the submittal are up-to-date with respect to applicable codes and i standards and have been used in conformance with their limitations and restrictions. This includes the codes used in the safety analyses for previous amendments such as upgrad.go VANTAGE 5 fuel, steam generator level tap relocation, and revised OTAT and OPAT reactor trip setpoint changes, which are referenced in the uprate licensing submittal.

W/rgm & SNC/mja & mge - 03/02/98 NRC Ouestion No. 4 (Reference March 2.1998 NRC/SNC Conference Call)

In the response to Questica Nos. 20 and 33 in SNC letter dated November 19,1997, SNC indicated that the limiting single failures have been established for FNP on an event specific basis and that for the steam generator tube rupture (SGTR) event minimum auxiliary feedwater (AFW) was assumed, consistent with a AFW system single failure. Clarify whether this assumption is the most limiting single failure for SGTR.

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$NC Response No. 4

'Ihe Farley Nuclear Plant licensing basis does not require a single failure in the steam genera +or tube rupture (SGTR) calculations. No limiting single failure case study has ever been performed for FNP SGTR. The licensing basis SGTR analysis consists of a simple mass and energy balance to calculate the primary to mandary break flow and atmospheric steam releases from the ruptured a xl intact steam generators for use in calculating the offsite radiological doses. No single failure was ccasidered in the original analysis. Starting in about 1991, the conservative assumption of a failure of the turbine-driwn pump in the AFW system was included in the Farley analysis. 'Ihis failure assumption was retamed for the power uprate SGTR analysis.

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ATTACllMENT 11 SNC Response to NRC Request For AdditionalInformation Related To Power Uprate Submittal- Joseph M. Farley Nuclear Plant, Units 1 and 2 CORRECTED PAGE NOS. 60,63,64,66 & 86 "FARLEY NUCLEAR PLANT UNITS 1 AND 2 POWER UPRATE PROJECT BOP LICENSING REPORT" (ATTACHMENT 6 TO SNC SUBMITTAL DATED FEBRUARY 14,1997) l 1

The potential impact of uncovery of the steam generator tubes during the event (s) was also evaluated for uprated conditions. Assuming technical specification limits for RCS activity (0.5 pCi/gm) and leak rate (150 gpd per generator) and release directly to the environment ( l.e., no mixing with the secondary side water) for the first 30 minutes, the offsite doses remain a small fraction of the 10 CFR 100 guidelines.

2.16.7.3.4 Evaluation of the Radiological Consequences of a Loss of Coolant Accident The radiological consequences of the large break LOCA were evaluated for the uprated core inventory utilizing the methods described in NUREG-0800, section 6.5.2, for the containment model and Farley l specific assumptions as shown in Tables 2.16-1 through 2.16-3 The results shown in Tables 2.16-4 and 2.16-5 include the contribution from ECCS recirculation loop leakage outside containment as described in FSAR Section 6.3. Evaluation of the offsite doses from hydrogen venting of 35 cfm starting 18 days after the LOCA (see Section 2.14) indicate that the total offsite doses and control room doses from all sources continue to meet 10 CFR 100 guidelines and GDC 19 requirements respectively.

l A small break LOCA which results in 100% failure of the fuel clad and release of 100% of the gap i activity, but does not result in a containment pressure high enough to initiate containment spray, was also evaluated. The assumptions shown in Tables 2.16-1 through 2.16-3, except the release source l term and spray removal constants for iodine cleanup in the containment, were used. The resultant offsite doses (from containment leakage without minipurge or hydrogen venting contributions) are:

Thyroid Dose Whole Body Beta Skin (Rem) Dose (Rem) Dose (Rem)

EAB 41.2 0.4 0.3 LPZ 29.0 0.2 0.1 l I

2.16.7.3.5 Evaluation of the Radiological Consequences of a Fuel Handling Accident The radiological consequences of the Fila were evaluated for the uprated core inventory. Two cases were  ;

considered, an accident in the Auxiliary Building and one in the Containment. The accident in the '

Auxiliary Building conforms to the guidelines of Regulatory Guide 1.25 and Standard Review Plan Section 15.7.4, except dose conversion factors from ICRP 30 are used in lieu of those from Regulatory Guide 1.25.

The accident in the containment is evaluated assuming the entire containment airborne source is exhausted via the containment purge filter. The releases result in ofTsite doses that are well within the 10 CFR 100 guidelines, which meets the acceptance criteria.

BOP UPRATE LICENSING REPORT 60 FNP - UNITS 1 AND 2 03/02/98 l

TABLE 2.16-1 (REVISED FSAR TABLE 6.2-29)

SPRAY EVALUATION PARAMETERS Power (MWt) 2831 Contamment pressure (psia) 68.7 C.xitamment temperature ( F) 276 Spray flowrate(gal / min) 2175 pH (sprayinjection) 4.5 (spray recirculation) 7.5 Contamment sprayed vol me(ft') 1.62 x 10' Minimum spray fall height (ft) 110  !

1 A, (h) i Elemental 10 (DF < 14) l 0.0 (DF 214)

Organic 0.0 i Particulate 4.8 (DF < 100)

O.48 (100sDF<1000) 0.0 (DF21000)

BOP UPRATE LICENSING REPORT 63 FNP - UNITS 1 AND 2 03/02/98 l

4 TABLE 2.16-2 (REVISED FSAR TABLE 15.4-14)

PARAMETERS USED IN TIIE LOCA ANALYSIS Core thermal power 2831 MWt (2775 x 1.02)

Containment free volume 2.03 x 10' ft' Vol une fractions Sprayed 0.822 Unsprayed 0.178 Mixing rate between sprayed and 40,500 ft'/ min unsprayed containment volemes Core fission product inventories See Table 15.1-4 Activity released to containment Noble gases 100% of core inventory loc'ines 50% ofcoreinventory Plateout of elemental iodine activity 2.7 li' DF < 100) released to containment 0.27 h

0.0h"

((100 $ DF s 1000)

(DF > 1000)

Form ofiodine activity in containment available for release Ele. mental 95.5 %

Organic 2.0 %

Particulate 2.5 %

Spray removal constants El: mental 10 h (DF <l4) l 0.0 h (DF 214)

Methyl 0.0 h

Partic.ilate 4.8 h (DF < 100) 0.48 li' 0.0 h ' ((DF 2 1000)100 s DF <1000 Time to reach decontamination factor Elemental 17 min l Methyl N/A Particuhte 7.3h ,

Containment leak rate 0-24 h 0.15'/dday 1-30 days 0.07f*/dday Atmospheri ; dilution estimates See Table 15.B-2 BOP UPPATE LICENSING REPORT 64 FNP - UNITS 1 AND 2 03/02/98 l

i TABLE 2.16-4 (REVISED FSAR TABLE 15.4-15) ,

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OFFSITE DOSES FROM LOCA Thvroid Dose. (REM)

LOCA w/o Mini-Purge 10 CFR 100 Mini-Purge Incremental Total Limit l Site boundary 165 5.7 171 300 Low-population zone 101 2.1 103 3;0

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Waole Body. (REM)

LOCA w/o Miri-Purge 10 CFR 100 Mini-Purae Incremental T9tal Limit Site boundary 3.1 8.7 x 10 ' 3.1 25 Low-population zone 1.6 2.2 x 10-3 1.6 25 I

l Offsite doses are based on a LOCA plus the incremental dose resulting from purging prior to isolation of the .

mini-purge system 6 seconds following a LOCA. ]

TABLE 2.16-5 (REVISED FSAR TABLE 15.4-17)

CONTROL ROOM DOSES FOLLOWING A LOCA (Dose (rem)

Thyroid 25.3 l Whole body 0.7 Beta-Skin 17.2 i

i BOP UPRATE LICENSINO REPORT 66 FNP - UNITS 1 AND 2 03/02/98 l 1

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-4.0 References 2.1 ' WCAP-14723. Farley Nuclear Plant Units I and 2 Power Uprate Project NSSS Licensing Report, January 1997.

7.1 Wyle Laboratory MSSV test reports 43516/QP-1169 (Unit I/1987) and 42539/QP-7294 (Unit 2/1993), Alabama Power, Farley Nuclear Plant, Dothan, AL.

10.1 Letter to the NRC dated 7/15/91,

Subject:

VANTAGE-5 Fuel Design Amendment; and NRC letter dated 12/30/91,

Subject:

Tech. Spec. Amendments 91 and 64 13.1 EPRl/NAl 8907-09. Rev. 3, " GOTHIC Containment Analysis Qualification Repori,' Version 5. )

13.2 WCAP-140.)], " Joseph M. Farley Nuclear Station, Units I and 2 Main Steam Valve Room Temperature '

Response to Supe heated Steam." March,1994.

16.1 Deleted.

l 22.1 Joseph M. Farley Nuclear Plant Environmental Report - Operating License Stage 22.2 Joseph M. Farley Nuclear Plant Final Environmental Statement (NUREG 0727) l 22.3 Joseph M. Farley Nuclear Plant Environmental Protection Plan (EPP), Appendix B to Facility Operating Licenses NPF -2 and NPF-8

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BOP UPRATE LICENSING RPORT 86 FNP - UNITS 1 AND 2 03/02/98 l

ATTACHMENT III SNC Response to NRC Request For Additional Information Related To Power Uprate Submittal- Joseph M. Farley Nuclear Plant, Units 1 and 2 EXCERPT FROM FARLEY NUCLEAR PLANT SCS CALCULATION SM-95-8931-002, Rev. 4 OFFSITE AND CONTROL ROOM DOSE FOR UPRATE WITH TSP FOR pH CONTROL (Sheets 2,3,4,5, & 10) i k

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• Southern Company Services Project Calculation Number Farley Nuclear Plant SM-95-8931-002 Subject /ritle Sheet Offsite and Control Room Dese for Uprate with TSP for pH Control 2 of 178

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Major Equations :

The calculations were performed using the TACT 5 computer program running on an NEC pentium desktop computer. The performance of TACT 5 on the NEC machine was verified in reference 1.

The test problems of reference I were successfully rerun on the computer to verify proper execution of the program. A directory listing of the TACT 5 .EXE and library files is included in Attachment 1.

These files were compared to the listings in reference 1 to verify the proper files are installed.

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

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 pH 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 reflected ir. the decontamination factor and removal process cutoffs will be determined in accordance with the guidance of references 4 and 8.

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2. References 8 and 9 indicate an expected value for contaminated boric acid spray solution may be approximately 10 / hr. Additional research (references 13 and 14) indicates that injection of fresh, uncontaminated by iodine, spray solutions are effective with or without additives. The spray removal consta 3t for calculated values using additives is typically limited to 20 per reference 4. For similar conservatism, the assumed value (calculated to be 13.7 per reference 13) will be limited to 10 / hr.

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 for this calculation based on the deposition velocity in reference 4. The deposhion velocity used to derive these values (0.137 cm/sec for zinc /zine coated surfaces and for epoxy coated surfaces) is conservative with respect to the data presented in reference 8, Table 5 for Dimetcote, Corbo-zinc and Amercoat 66 which are similar to the coatings used at FNP. This is assumed to reduce to approximately 10% 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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, Calculat on Number Farley Nuclear Plant SM-95-8931-002 Subjectriitic Sheet Offsite and Control Room Dose for Uprate with 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 iodine below the assumed removal cutoff of 1000. Cutoff times are determined based on I:3i concentrations, i.e. ignoring decay of short lived isotopes. Removal coefficients and removal process cutoffs used per assumptions 1-4 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):he 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% orgva, and 2.5% particulate (references 7,13,20).

6. The sump pH reduction (from previous NaOH addition levels) does not impact the containment pressure / temperature response; thus the sump (recirculation fluid) temperature and flashing fraction, and the ECCS leakage contribution to the total dose, are modeled as described in reference 10. ECCS 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, ECCS 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.

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4 Southern Company Services Project + , Calculation Number Farley Nuclear Plant SM-95-8931-002 hubje Ifiltle Sheet Offsite and Control Room Dose for Upra:e with TSP for pH Control 4 of 178 Rcferences :

1 Nuclear Support Calculation number N-94-02, " Verification of TACT 5," revision 0.

2 NUREG'CR-510( , SAIC-88/3023 " User's Guide for the TACT 5 Computer Program."

3 FNP Final Safety Analysis Report

a. Table 6.3-8

b. Table 15.4-14

c. Table 15.4-16

d. Table 15.4-20

e. Table 15B-2

f. Figure 3.7-20

g. Table 6.2-2

h. Table 6.2-5

i. Table 15.4-18 4 NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan," Section 6.5.2, Resision 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 Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," June,1974.

8 NUREG-CR-0009," Technological Bases for Models of Spray Washout of Airborne Contaminants in Containment Vessels," October 1978.

9 WASH-1329,"A review of Mathematical Models for Predicting Spray Removal of Fission Products in Reactor Containment Vessels," June, 15,1974.

10 NUIEG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan," Section 15.6.5, Revision 1.

I1 Letter AP-21370, dated February 6,1996,"Up-date Control Room Dose Assessment."

12 Murphy, K.G. and Campe, Dr. K.M.,13th AEC AiCicaning Conference," Nuclear Power Plant Control Roem Ventilation System Design for Me:tiat; Gen :ral Design Criterion 19."

4 Southern Company Services Project ,

, Calculation Number Farley Nuclear Plant SM-95-8931-002 hubject/I'itle bheet Offsite and Control Room Done for Uprate with TSP for pil Control 5 of 178 13 WCAP-11611, March 1988," Methodology for Elimination of the Containment Spray Additive."

14 Davis, R.E., et al," Fission Product Removal Effectiveness of Chemical Additives in PWR Containment Sprays," Technical Renort A-3788. 8/12/86, attached to proposed revision 2 to Standard Review Plan 6.5.2, with AIF letter of 5/11/87.

15 Letter ALA-95-756, dated 12/15/95, Analysis Input Assumption List 16 Letter ALA-99-508, dated February 1,1966," Final Core Inventory Source Terms."

17 Calculation 40.05, Revision 3, Post Accident Hydrogen Generation Analysis.

18 FNP Calculations

a. Mechanical calculation 4.2

b. Mechanical calculation 4.1

c. SM-93-0121-001

d. SM-95-8931-001 19 FNP Technical Specifications

a. 3/4.6.1.2 20 NIIREG-0588 ,

21 Regulatory Guide 1.52," Design, Testing and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power Plants," Rev. O.

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Southern Company Services Project * , Calculauvn Number Farley Nuclear Plant SM-95-8931-002 hubject/lTc Sheet Offsite and Control Room Dose for Uprate with TSP for pH Control 10 of 178 The minimum pH is maintained at 7.5 as discussed in assumption 1. The partition factor between liquid and l

gas phases is based on reference 4, Figure 6.5.2-1. With a pH of 7.5, the partition coefficient is 440.

The elemental iodine spray removal coefficient is 10 hf' per assumption 2. With a partition coefficient of 440, the decontamination factor (DF) limit is based on reference 4:

b DF= 1 + (4.92E+4 f* '440)/(1.67E+6 ft3) = 14.0 where the sump and containment volumes are as provided above.

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

5 2 Zinc / zinc painted surfaces from heat sinks 1 and 4-15 = 2.70 x 10 ft 4 2 Epoxy surfaces from heat sinks 2 and 3 = 6.47 x 10 ft Then per references 8 and 1.,

A = 118 I(Deposition velocity x Area / Volume) 5 4

= 118 x 0.137 x (2.70 x_lD + 6.47 x 10 ) = 2.7 hf' 6

2.03 x 10 This decreases to approximately 10% of the initial value or about 0.27 hf' after reducing the original gl concentration by 100, and to 0 after a reduction of 1000.

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

A = 3(100 ftV2175 gpmVO.1 cm d )_ x 60 min x 30.5 cm = 4.77 hf' 3

2(1.669E6 ft )(7.5 gal /ft ) hr fl where 0.1 cm is a conservative washout parameter (E/d) from reference 8, section 5.3.1 (p 34), until 1.he particulate DF = 100. After 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 ft based on reference 3f, and spray flow of 2175 gpm is based on references 3h and 18c.

These values are input to TACT 5 which is run iteratively to determine the cutoff times as described in assumptions 1-5 above. The removal rates and cutoff times are shown below:  ;

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