ML11105A151

Response to Request for Additional Information Regarding Technical Specifications Change TS-474 - TS 3.7.3, Control Room Emergency Ventilation System, Dated March 10, 2011

ML11105A151
Person / Time
Site:	Browns Ferry
Issue date:	04/11/2011
From:	Krich R Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML11105A151 (8)

Text

Tennessee Valley Authority 1101 Market Street, LP 3R Chattanooga, Tennessee 37402-2801 R. M. Krich Vice President Nuclear Licensing April 11,2011 10 CFR 50.4 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296

Subject:

Response to Request for Additional Information Regarding Technical Specifications Change TS-474 - TS 3.7.3, "Control Room Emergency Ventilation System," dated March 10, 2011

References:

1. Letter from TVA to NRC, "Technical Specifications Change TS-474 -

Request to Add a TS 3.7.3, "Control Room Emergency Ventilation (CREV)

System," Action to Address Two CREV Subsystems Inoperable Due to Inoperable CREV System High Efficiency Particulate Air (HEPA) Filter and/or Charcoal Adsorbers," dated August 27, 2010

2. NRC Letter to TVA, "Browns Ferry Nuclear Plant, Units 1, 2, and 3 -

Request for Additional Information Regarding Technical Specification Change TS-474 (TAC Nos. ME4668, ME4669, and ME4670)," dated March 10, 2011 On August 27, 2010, the Tennessee Valley Authority (TVA) submitted Technical Specifications (TS) Change TS-474, TS 3.7.3, "Control Room Emergency Ventilation (CREV) System," requesting approval of a change for the condition when both CREV subsystems are inoperable due to an inoperable High Efficiency Particulate Air (HEPA) filter, or when one or more CREV subsystems are inoperable due to an inoperable AOO' printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 April 11,2011 charcoal adsorber. TVA also requested a completion time of 90 days to restore the HEPA filter and charcoal adsorber to operable status (Reference 1). On March 10, 2011, TVA received a Request for Additional Information (RAI) letter from the NRC (Reference 2) applicable to TS Change TS-474. The NRC requested the response within 30 days, i.e., by April 9, 2011. Since April 9, 2011 is a Saturday, the response due date is no later than April 11, 2011.

The enclosure to this letter provides the TVA response to the NRC RAI. Additionally, an error was identified in TVA's submittal on March 10, 2010, in the second paragraph on page E-5 of Enclosure 1, and a corresponding error in the response to Question 3 of Section 4.3 (i.e., the Significant Hazards Determination) on page E-7 of Enclosure 1 (Reference 1). TVA is providing the corrected text in the enclosure.

This letter does not include any new regulatory commitments. Please direct any questions concerning this matter to Tom Matthews at (423) 751-2687.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on the 1 1 th day of April, 2011.

Respectfully, R. M. Krich

Enclosure:

TVA Response to NRC Request for Additional Information cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant Alabama State Department of Public Health

ENCLOSURE Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications Change TS-474 - TS 3.7.3, "Control Room Emergency Ventilation System" TVA Response to NRC Request for Additional Information

ENCLOSURE Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications Change TS-474 - TS 3.7.3, "Control Room Emergency Ventilation System" TVA Response to NRC Request for Additional Information NRC Request for Additional Information (RAI) Question "The current design basis LOCA radiologicalconsequence analysis assumes credit for the CREV HEPA filter. The applicationchanges this assumption in the LOCA analysis by removing the credit taken for the HEPA filtration. Provide additionalinformation describing all the basic parametersused in the LOCA dose consequence analyses. For each parameter,please indicate the current licensing basis (CLB) value, the revised value where applicable, as well as the basis for any changes to the CLB. Also describe any methodologies that may have changed based on the proposed amendment. In addition,provide the resulting change to the calculatedradiologicalconsequence of the design basis LOCA. The staff requests that the information be presented in a table as was done for the Browns Ferry alternativesource term amendment request."

TVA Response As part of the TS-405 License Amendment Request, dated July 31, 2002 (ML022200382) to adopt the Alternate Source Term (AST) methodology, TVA provided a table of inputs to be used with RADTRAD Computer Code (NUREG/CR-6604, April 1998) to calculate the radiological consequences of the design basis loss-of-coolant accident (LOCA). Table 1 of this enclosure contains the current licensing basis (CLB) values, revised values and the basis for the revised value.

Based on the inputs provided in Table 1, the revised radiological consequences of the design basis LOCA results in a control room dose of 1.94 rem. The CLB control room dose from a design basis LOCA is 1.62 rem. The revised design basis LOCA control room dose was calculated using the same methodology approved as part of the AST amendment issued by NRC letter dated September 27, 2004.

Corrected Text The revised wording in the second paragraph on page E-5 of Enclosure I of the August 27, 2010, letter is provided below.

"The recent analyses show that the post-LOCA 30-day control room dose with no credit for either the HEPA filters or the charcoal adsorbers results in a minimal El of 5

increase in dose consequences (9.5% decrease in margin), with the final dose of 1.94 rem. This remains well below the regulatory limit of 5 rem."

TVA's response to Question 3 of Section 4.3 (Significant Hazards Determination) on page E-7 of Enclosure 1 of the August 27, 2010, letter contained the same error identified above. The revised wording is provided below.

"Analyses associated with the prior approval of Alternate Source Term methodology for design basis accident dose consequences previously did not credit the CREV System charcoal adsorbers. Recent analyses have been performed to assess the post-accident 30-day control room dose removing credit for the CREV System HEPA filter. The results indicate a minimal increase in dose consequences (9.5% decrease in margin) due to removing credit for the CREV System HEPA filter. Even with no credit for either the CREV System HEPA filter or CREV System charcoal filter, the resultant control room dose maintains more than 60 percent margin to the regulatory limit of 5 rem TEDE. As such there is no reduction in a margin of safety for any duration of inoperability of the CREV System HEPA filter or charcoal adsorbers. While the HEPA filter and charcoal adsorbers are not credited for accident mitigation, they remain required by the BFN TS for compliance with the LCO 3.7.3, "Control Room Emergency Ventilation (CREV) System," further minimizing any potential reduction in a margin of safety."

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Control Room Emergency Ventilation (CREV) Intake Flow Rate CREV Makeup Filtered Flow Rate 3000 scfm 0 CREV Unfiltered Inleakage Rate 3717 scfm 6717 Changes made to eliminate credit for the CREV High Efficiency Particulate Air 90% Particulate No credit taken HEPA filter in the CREV System (HEPA) Filter Efficiency CREV Charcoal Adsorption Efficiency No credit taken Control Room Volume 210,000 ftW Standby Gas Treatment (SGT) Flow Rate 24,750 scfm SGT HEPA Filter Efficiency 90% Particulate SGT Charcoal Adsorption Efficiency No credit taken Environment Breathing Rate 0-8 hours: 3.5E-04 ma/sec 8-24 hours: 1.8E-04 m3/sec 1-30 days: 2.3E-04 m3/sec Control Room Breathing Rate 3.5E-04 m3/sec Control Room Occupancy Factors 0-1 day: 1.0 1-4 days: 0.6 4-30 days: 0.4 Fission Products Release Fractions Regulatory Guide 1.183 (Revision 1) Table 1 BWR Core Inventory Fraction Released Into Containment Gap Early In-Release vessel Grou p Phase Phase Tot al Nobt e Gases 0.05 0.95 1.0 Halo gens 0.05 0.25 0.3 AlkalliMetals 0.05 0.20 0.2!5 Tellu rium Metals 0.00 0.05 0.015 Ba, Sr 0.00 0.02 0.0:2 Noble Metals 0.00 0.0025 0.0(025 Ceiuum Group 0.00 0.005 0.0(005 Lantlhanides 0.00 0.0002 0.01 Fission Product Release Timing Regulatory Guide 1.183 (Revision 1) Table 4 LOCA Release Phases BWR Phase Onset Duration Gap Release 2 min 0.5 hr Early In-vessel 1.5 hr 0.5 hr Fission Product Iodine Chemical Form Particulate 95%

Elemental 4.85%

Organic 0.15%

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Control Room Isolation/CREV Initiation 10 minutes Emergency Core Cooling System Ten percent of the radioiodine in the (ECCS) Leakage Release Fractions leaked coolant is assumed to become airborne in the reactor building (secondary containment). Of this activity, 97% is assumed to be elemental iodine and 3% is assumed to be organic iodine.

Flow __ __ __ __ __.. _____.______,s Primary Containment Leak Rate 2% containment air weight/day (30 days)

Secondary Containment Bypass Leak Hardened Wet Well Vent Release =

Rate (30 days) 10 scfh beginning at t>8 hours Assumed ECCS Leak Rate (30 days) 5 gpm 20 gpm Revised to provide additional margin. The CLB Control Room Dose was increased from 1.25 REM to 1.62 REM, in accordance with 10 CFR 50.59.

ECCS Leakage Temperature <212°F Main Steam Isolation Valve (MSIV) Leak 150 scfh total Rate at Test Pressure of 25 psig 100 scfh maximum for one line Leakage at Base of Stack (stack 10 scfm 20 scfm Revised to provide additional margin for bypass) damper testing MSIV Leakage that Bypasses Main 0.5%

Condenser (percentage of total MSIV leakage)

Containment Atmosphere Dilution Vent 139 scfm for 24 hrs Rate @ 10 days, 20 days, 29 days Drywell Airspace 159,000 ftJ (Min value used for dose calculation)

Torus Airspace 119,400 ft3 (Minimum)

Suppression Pool 121,500 ftW (Minimum)

Reactor Building Free Volume 1,931,502 fte 1.1311E6 fte Unit 1 Volume used for conservatism (50% of this value used due to incomplete mixing) (50% of this value used due to incomplete mixing)

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Stack Room 69,120 ft' (50% of this value used due to incomolete mixina)

High Pressure Turbine 568.6 ft3 (No credit taken)

Low Pressure Turbine 51,000 ftW (No credit taken)

Drywell Natural Deposition Particulate: Power's Model, 10th percentile values (conservative compared to SRP 6.5.2 X, Elemental: Same as particulate.

Drywell Accident Conditions (maximum) P = 48.5 psig, T = 295.2 Degrees F Surface Area for Elemental Iodine 3409 mz Deposition in Drywell Condenser Volume 90% of 136,000 ft or 122,400 ft Steam Line Conditions Saturated Conditions at 1050 psia Steam Line Volume: Inboard to 53.7 ftW Outboard MSIV Steam Line Volume: Outboard MSIV to 173.1 ftW Drain Line Sedimentation Height 27.2 ft Removal Efficiency for Removal Efficiency for Aerosol Particles Elemental Iodine Steam Line Leakage 99.87% 99.01%

(Drywell to Main Condenser) (These removal efficiencies applied to a leakage entering the main condenser volume include removal in the condenser downstream)

Main Condenser Bypass 89.33% 16.37%

(Drywell to Environment) I I I E5 of 5