ML18039A271

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Proposed Tech Specs Re Power Uprate Operation
ML18039A271
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 03/16/1998
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TENNESSEE VALLEY AUTHORITY
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ML18039A270 List:
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NUDOCS 9803250175
Download: ML18039A271 (27)
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ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 and 3 PROPOSED I I TECHN CAL SPEC FI CATION ( TS ) CHANGE TS 384 ~ SUPPLEMENT 1 DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE DESCRIPTION OF THE PROPOSED CHANGE TVA is requesting a change to modify the Unit 2 Primary Containment Isolation Instrumentation function for the Reactor Water Cleanup (RWCU) System and to modify for Units 2 and 3 the primary containment peak temperature value following high energy line breaks inside primary containment. In doing so, the following will be implemented:

1. The Allowable Value is being reduced for the Unit 2 instruments which initiate RWCU isolation based on high temperature in the main steam valve vault (MSVV).
2. The maximum post-accident drywell temperature value for Units 2 and 3 has been increased from 322'F to 336'F based on the results of the containment response analyses at power uprate conditions.

The specific changes are described as follows:

1. Current Unit 2 Table 3.3.6.1-1, item 5.a:

Table 3.3.6.1-1 Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABL FUNCTION CONDITIONS SYSTEM ACTION C.l REQUIREMENTS E VALUE

5. Reactor Water Cleanup (RWCU) System Isolation
a. Main Steam Valve 1,2,3 SR 3.3.6.1.2 g 2PIoF Vault Area SR 3.3.6.1.4 Temperature SR 3. 3. 6. 1. 6 Proposed Unit 2 Table 3.3.6.1-1, item 5.a:
a. Mann Steam Va ve 1,2,3 SR 3.3.6.1.2 g IppoF Vault Area SR 3.3.6.1.4 Temperature SR 3.3.6.1.6 98OSa5017S esOS1S PDR ADOCZ OSOOOZSOR PDR Q
2. Current Unit 2 and 3 BASES 3.6.1,4, Drywell Air Temperature,

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pa'ge B 3.6-28:

Analyses assume an initial average drywell air temperature of 150'F. This limitation ensures that the safety analysis remains valid and ensures that the peak drywell temperature does not exceed the maximum allowable temperature of 322'F (Ref. 2). Exceeding this temperature Proposed Unit 2 and 3 BASES 3.6.1.4, Drywell Air Temperature, page B 3.6-28:

Analyses assume an initial average drywell air temperature of 150'F. This limitation ensures that the safety analysis remains valid and ensures that the peak drywell temperature does not exceed the maximum calculated temperature of 336'F (Ref. 2). Exceeding this temperature II . REASON FOR THE PROPOSED CHANGE New high energy line break (HELB) analyses were performed using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) computer program and models of the Unit 2 and 3 reactor buildings. Based on the new analyses, a change is required in the Unit 2 isolation logic allowable values. The GOTHIC computer code is an upgraded computer code compared to that used for the current analysis and accounts for physical phenomena (specifically steam buoyancy)not previously considered. The new analyses were performed in support of the power uprate project; however, as discussed in the Safety Analysis section, the allowable value change is primarily due to the change in the analysis methodology and modeling revisions rather than due to the effects of power uprate.

The high energy line breaks previously analyzed for Browns Ferry Units 2 and 3 for the current license (3293 MWt), were re-analyzed to support operation at 5% power uprate conditions (3458 MWt). From that reanalysis, the only changes to the EQ analysis are those described in this enclosure.

The HELB analyses evaluated the postulated breaks in the reactor buildings and concluded that the existing allowable values were acceptable for break detection and isolation with the exception of the Unit 2 RWCU break in the MSVV.

For Unit 2 only, the analytical limit for RWCU isolation due to high temperature in the MSVV will be reduced based on the new HELB analyses. The revised Allowable Value for this function was determined based on the maximum abnormal operating temperature for the MSVV and the analytical limit in the HELB analyses.

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~ I The primary containment peak temperature response to high

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energy steam line breaks was analyzed at uprated conditions using the GE proprietary code SHEX (Reference 2). The increased reactor pressure required for operation at uprated conditions results in higher mass and energy releases following postulated line breaks inside primary containment.

The maximum post-accident drywell temperature value has been increased from 322'F to 336'F based on the results of the containment response analyses performed at uprated conditions.

III. SAFETY ANALYSIS SYSTEM DESCRIPTION The RWCU high area temperature isolation functions are required to isolate HELBs outside of primary containment.

Multiple instruments are provided to detect high area temperatures resulting from RWCU line breaks outside of primary containment. For each monitored area, four temperature monitoring instruments are provided which initiate isolation of the monitored This system in a one-out-of-two taken twice logic arrangement. logic arrangement prevents a single instrument failure from preventing isolation when necessary and also prevents a single instrument failure from resulting in a spurious isolation signal.

The setpoints and corresponding Allowable Values for the area temperature monitoring instruments are high enough to prevent spurious isolation of the RWCU system during abnormal plant operating conditions (including station blackout), yet prevents exceeding the analytical limits from the HELB analyses for RWCU isolation due to high area temperature.

Primary containment limits the release of radioactive material following a'design basis accident. The peak temperature following postulated high energy line breaks inside containment is one parameter used to evaluate the containment structural design and acceptability of equipment located inside containment which is required to operate for mitigation of high energy line breaks. The maximum post-accident drywell temperature value has been increased from 322'F to 336'F based on the results of the containment response analyses performed at power uprate conditions.

SAFETY EVALUATION The original reactor building HELB analyses for Browns Ferry were performed using TVA's MONSTER computer program; however, due to changes in TVA's computer systems, this computer program is no longer available. MONSTER is based on the CONTEMPT4 MOD2 code with enhancements added by TVA.

The revised HELB analyses for Browns Ferry Units 2 and 3

program and GOTHIC models of the Unit 2 and 3 reactor buildings.

Results of BFN analyses for HELBs outside of containment using the GOTHIC computer program indicated a lower rate of increase for certain compartment temperatures than the results of previous MONSTER analyses for similar pipe break events. Since the mass flow rates for high energy line breaks at power uprate conditions increased slightly, the slower rate of rise in compartment temperatures were not the result of power uprate implementation. Informal analyses were performed which demonstrated that the lower rate of rise for temperatures in certain areas was due more to the effects of steam buoyancy in the GOTHIC computer analyses.

In certain plant areas, the GOTHIC analyses indicated that more steam would exit through openings in the compartment ceilings and walls more than predicted by previous analyses.

The increase in steam flow to the compartments located adjacent to and above the compartment where the break occurs caused changes to the reactor building temperatures which are not related to power uprate implementation.

The results of the new HELB analyses indicate that steam buoyancy affects the temperature response for the MSVV which has a door and pressure relief panels which provide flow paths for significant amounts'f steam to exit the MSVV following a high energy line break. In order to achieve a timely isolation of RWCU line breaks and limit temperatures in the affected reactor building areas, the analytical limit and the corresponding Allowable Value have been lowered for RWCU isolation based on high area temperature in the Unit 2 MSVV. The lower Allowable Value will require a lower setpoint for the associated temperature detectors which will result in detection and isolation sooner and a reduction in the total mass and energy release following an RWCU line break in the MSVV. There are minor physical configuration differences for Unit 2 and Unit 3 in the openings which provide flow paths for steam released following HELBs in the MSVV. The GOTHIC analysis results indicate that no change is required to the Unit 3 analytical limit and corresponding allowable value for RWCU isolation due to high temperature in the MSVV.

The analytical limits for high area temperature isolation are derived from the limiting values in the HELB analysis.

The Allowable Values are derived from the analytical limits, corrected for calibration, process, and the appropriate instrument errors.

The Allowable Value is high enough to prevent spurious isolation due to temperature increases during abnormal operating conditions (including station blackout) and low enough to preclude exceeding the analytical limit in the

HELB analyses for isolation due to high area temperatures.

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Lowering the Allowable Value results in lower setpoints for the break detection instruments. Lower setpoints result in detection and isolation of high energy line breaks sooner which reduces the total mass and energy release following the line break.

The primary containment response to high energy steam line breaks was analyzed at uprated conditions using the GE proprietary code SHEX which has been previously reviewed and accepted by the NRC for use in long term containment analyses (Reference 2). The maximum post-accident drywell temperature value has been increased from 322'F to 336'F based on the results of the containment response analyses performed at uprated conditions.

In accordance with 10 CFR 50.49, Environmental Qualification evaluations based on the power uprate HELB analysis results for BFNP Units 2 and 3 will be performed for safety related electrical equipment located in harsh environment areas 2 .

The primary containment structural design will be evaluated prior to power uprate implementation to ensure the capability to mitigate and withstand HELBs inside primary containment. When approved by the staff, the reactor building EQ thermal profiles and subsequent equipment evaluations will be based on the GOTHIC software.

References:

GOTHIC Containment Analysis Package, Version 5.0c, Volume 3, Qualification Report.

2. Letter to Gary L. Sozzi (GE) from Ashok Thadani (NRC) on the Use of the SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis, July 13, 1993.

2 This commitment was identified as part of the October 1, 1997, letter. As such no new commitment has been established.

IV,. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

'TVA has concluded that operation of BFN in accordance with the proposed change to the TS does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of the three standards set forth in 10 CFR 50 '2(c).

A. The ro osed amendment does not involve a si nificant increase in the robabilit or consequences of an accident reviousl evaluated.

TVA is proposing to modify the Unit 2 RWCU isolation based on high temperature in the MSVV.

RWCU system isolation is required to mitigate the consequences of high energy line breaks outside of primary containment. The revised Allowable Value for the high area temperature isolation function following a line break in the MSVV is based on the HELB analyses performed for the 5% power uprate implementation. The revised Allowable Value for the affected high area temperature isolation function is lower than previously specified which results in a lower setpoint for the associated temperature detection instrumentation. The lower setpoint results in detection and isolation-of HELBs sooner which reduces the total mass and energy release following the line break.

The proposed Allowable Value was determined based on the results on the HELBs analyses performed at power uprate conditions. Equipment required to function for mitigation of HELB events will be evaluated against the results of these HELB analysis results in accordance with 10 CFR 50.49'. RWCU system isolation is not a precursor to any design basis accident analyzed in the BFN UFSAR. Therefore, proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change increases the maximum post-accident drywell temperature value from 322'F to 336'F based on the results of the containment response analyses performed at 5% power uprate conditions. The increase is containment temperature is the result of the HELB analyses and is not a precursor to any accident.

Therefore, the increase in peak containment temperature does not involve a significant increase in the probability or consequences of an accident previously evaluated.

3 This commitment was identified as part of the October 1, 199', letter. As such no new commitment has been established.

The primary containment structural design and the equipment located inside primary containment will be evaluated prior power uprate implementation to ensure the structures and equipment required for mitigation of high energy line breaks are capable of performing their required safety functions. Therefore, this proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.

The proposed change modifies the Unit 2 RWCU isolation function. However, the design function of the RWCU system will not change as a result of the proposed change. The RWCU isolation function is required to mitigate HELB events and is not a precursor to any accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change increases the maximum post-accident drywell temperature value from 322'F to 336'F based on the results of the containment response analyses performed at 5% power uprate conditions. The post-accident drywell temperature is the result of the high energy line break analyses and is not a precursor to any accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The ro osed amendment does not involve a si nificant reduction in the mar in of safet The proposed change to the RWCU area temperature isolation function is based on'he results of the power uprate HELB analyses which will be used to evaluate the ability of plant equipment to mitigate HELB events at power uprate conditions. The revised Allowable Value for the affected high area temperature isolation function is lower than previously specified which results in a lower setpoint for the associated temperature detection instrumentation. The lower setpoint results in= detection and isolation of high energy line breaks sooner which reduces the total mass and energy release following the line break.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The proposed change increases the maximum post-accident drywell temperature value from 322'F to 336'F based on the results of the containment response analyses

performed at 5% power uprate conditions. The containment structural design and the equipment located inside containment will be evaluated prior to power uprate implementation to ensure the structures and equipment required for mitigation of high energy line breaks are capable of performing their required safety functions. Therefore, this proposed change does not involve a significant reduction in a margin of safety.

ENVIRONMENTAL IMPACT CONSIDERATION The proposed change does not involve a significant hazards consideration, a change in the types of, or increase in, the amounts of any effluents that may be released off-site, or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

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ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPEC IFI CATION ( TS ) CHANGE TS 384 SUPPLEMENT 1

, MAFGCED PAGES AFFECTED PAGE LIST The following pages have been revised and an 'X'as been placed in the right hand margin to indicate where the changes occur.

Technical S ecifications Unit 2 3.3-56 Bases Unit 2 Unit 3 B 3. 6-28 B. 3.6-28 II. MARKED PAGES See attached.

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 2 of 3)

Primary Contaiment Isolation tnstrunentation APPLiCABLE CONDITIONS HCOES OR REQUIRED REFEREHCED OTHER CHANNELS FROH SPECIFIED PER TRIP REQUfRED SURVEILLANCE ALLCMABLE FUNCTION COND I T IONS SYSTEH ACTIOH C.1 REQUIREHENTS VALUE

3. HPCI System Isolation (continued)
d. HPCI Steam Line Space HPCI Punp Rocm Area 1,2,3 SR 3.3.6.1.2 S 200'F Tenperature -High SR 3.3.6.1.3 SR 3.3.6.1.6
e. HPCI Steam Line Space 1,2,3 SR 3.3.6.1.2 S 180'F Torus Area SR 3.3.6.1.3 Tenperature -High SR 3.3.6.1.6
4. Reactor Core isolation Cooling (RCIC) System Isolation
a. RCIC Steam Line 1,2,3 SR 3.3.6.1.2 < 15(C rated Flew -High SR 3.3.6.1.5 steam floe SR 3.3.6.1.6
b. RCIC Steam Supply Line 1,2,3 SR 3.3.6.1.2 ? 50 psig Pressure -Lou SR 3.3.6.1.5 SR 3.3.6.1.6
c. RCIC Turbine 1,2,3 SR 3.3.6.1.2 S 20 psig Exhaust Diaphragm SR 3.3.6.1.5 Pressure -High SR 3.3.6.1.6
d. RCIC Steam Line Space 1,2,3 SR 3.3.6.1.2 S 180'F RCIC Pump Room Area SR 3.3.6.1.3 Tenperature -High SR 3.3.6.1.6
e. RCIC Steam Line Space 1,2,3 SR 3.3.6.1.2 5 155'F Torus Area SR 3.3.6.1.3 Tenperature -High SR 3.3.6.1.6
5. Reactor lister Cleanup (RWCU) System Isolation F88'I=
a. Hain Steam. Valve Vault 1,2,3 F SR 3.3.6 1.2 3.3.6.1.4

~F Area Temperature SR SR 3.3.6.1.6

b. Pipe Trench Area 1,2,3 SR 3.3.6.1.2 S 135'F Tenperature . SR 3.3.6.1.4 SR 3.3.6.1.6 ce PLRp Room 2A Area 1,2,3 SR 3.3.6.1.2 S 152'F Tenperature SR 3.3.6.1.4 SR 3.3.6.1.6 (continued)

BFN-UNIT 2 3.3-56 Amendment

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Orywell Air Temperature 8 3.6.i.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Orywell Air Temperature BASES BACKGROUND The drywell contains the reactor vessel and piping, which add heat to the airspace.'rywell coolers remove heat and maintain a.suitable environment. The average airspace temperature affects the calculated response to postulated Design Basis Accidents (DBAs). The limitation on the drywell average air temperature was developed as reasonable, based on operating experience. The limitation on drywell air temperature is used in the Reference 1 safety analyses.

APPLICABLE Primary containment performance is evaluated for a SAFETY ANALYSES spectrum of break sizes for postulated loss of coolant accidents (LOCAs) (Ref. 1). Among the inputs to the design basis analysis is the initial drywell average air

~-.temperature (Ref. 1). Analyses assume an initial average drywell air temperature of 150'F. This limitation ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak dr welt tern erature does oot exceed the maximum mttmuahtecxdc~la~a~)

empera ure o 'F (Ref. 2). Exceeding this temperature may result in the degradation of the primary containment structure under accident loads. Equipment inside primary containment required to mitigate the effects of a OBA is designed to operate and be capable of operating under environmental conditions expected for the accident.

Drywell air temperature satisfies Criterion 2 of the NRC Policy Statement (Ref. 3),

LCO - In- the event of a OBA, with an initial drywell average air

'temperature less than or equal to the LCO temperature limit, the resultant peak accident temperature is maintained below the maximum allowable temperature. As a result, the ability of primary containment to perform its design function is ensured.-

(continued)

BFN-UNIT B 3.6-28 Amendment 2

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s Orywell Air Temperature B 3.6.1.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Orywell Air Temperature BASES BACKGROUND The drywell contains the reactor vessel and piping, which add heat to the airspace. Orywell coolers remove heat and maintain a suitable environment. The average airspace temperature affects the calculated response to postulated Design Basis Accidents (OBAs). The limitation on the drywell average air temperature was developed as reasonable, based on operating experience. The limitation on drywell air temperature is used in the Reference 1 safety analyses.

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APPLICABLE Primary containment performance is evaluated for a SAFETY ANALYSES spectrum of break sizes for postulated loss of coolant accidents (LOCAs) (Ref. 1). Among the inputs to the design basis analysis is the initial drywell average air temperature (Ref. 1). Analyses assume an irritial average

-dpywell air temperature of 150'F. This limitation ensures'hat the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak dr well tern erature does oot exceed the maximum mttsesehte cal~k4<~

temperature of 'F (Ref. 2). Exceeding this temperature f may result in the degradation of the primary containment structure under accident loads. Equipment inside primary containment required to mitigate the effects of a DBA is designed to operate and be capable of operating under environmental conditions expected for the accident.

Drywell air temperature satisfies Criterion 2 of the NRC Policy Statement (Ref. 3).

LCO In the event of a OBA, with an initial drywell average air temperature less than or equal to the LCO temperature limit,

-the resultant peak accident temperature is maintained below the maximum allowable'emperature. As a result, the ability of primary containment to perform its design function is ensured.

(continued)

BFN-UNIT 3 B 3.'6-28 Amendment

ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPECI FICATION (TS) CHANGE TS 384 I SUPPLEMENT 1 REVISED PAGES I. AFFECTED PAGE LIST The following pages have been revised. A revision bar has been placed in the left hand margin to indicate where the changes occur.

Technical S ecifications Unit 2 Unit 3 3.3-56 Bases Unit 2 Unit 3 B 3. 6-28 B 3. 6-28 II . REVISED PAGES See attached.

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 2 of 3)

Primary Contafreent Isolation Instrunentatfon APPLICABLE COND IT IOHS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOIIABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREHEHTS VALUE

3. HPCI System Isolation (continued)
d. HPCI Steam Line Space HPCI Pump Room Area 1,2,3 SR 3.3.6.1.2 S 200'F Teaperature -High SR 3.3.6.1.3 SR 3.3.6.1.6
e. HPCI Steam Line Space 1,2,3 SR 3.3.6.1.2 S 180'F Torus Area SR 3.3.6.1.3 Tcsperature -Hfgh SR 3.3.6.1.6
4. Reactor Core lsolatfon Cooling (RCIC) System I so let i on 3.3.6.1.2 rated a.

Flow - High Line RCIC Stcam 1,2,3 SR SR 3.3.6.1.5 S 150X steam flow SR 3.3.6.1.6

b. RCIC Steam Supply Line 1,2,3 SR 3.3.6.1.2 2 50 psig Pressure Lou SR 3.3.6.1.5 SR 3.3.6.1.6
c. RCIC Turbine 1,2,3 SR 3.3.6.1.2 S 20 psig Exhaust Diaphragm SR 3.3.6.1.5 Pressure -High SR 3.3.6.1.6
d. RCIC Steam Linc Space 1,2,3 SR 3.3.6.1.2 < 1800F RCIC Ptnp Room Area SR 3.3.6.'I.3 Temperature -High SR 3.3.6.1.6
e. RCIC Steam Line Space 1,2,3 SR 3.3.6.1.2 S 155'F Torus Area SR 3.3.6.'I.3 Temperature -High SR 3.3.6.1.6 5~ Reactor Mater Cleanup (RHCU) System Isolation
a. Hain Steam Valve Vault 2.3 SR 3.3.6.1.2 < 188'F Area Temperature SR 3.3.6.1.4 SR 3.3.6.1.6
b. Pfpe Trench Area 1,2,3 SR 3.3.6.1.2 S 135'F Temperature SR 3.3.6.1.4 SR 3.3.6.1.6
c. Pump Room 2A Area 1,2,3 SR 3.3.6.1.2 S 1524F Temperature SR 3.3.6.1.4 SR 3.3.6.1.6 (continued)

BFN-UNIT 2 3.3-56 Amendment

,I rywell Air Temperature B 3.6.1.4 8 3.6 CONTAINHENT SYSTEMS B 3.6.1.4 Drywell Air Temperature BASES BACKGROUND The drywell contains the reactor vessel and piping, which add heat to the airspace. Drywell coolers remove heat and maintain a suitable environment. The average airspace temperature affects the calculated response to postulated Design Basis Accidents (DBAs). The limitation on the drywell average air temperature was developed as reasonable, based on operating experience. The limitation on drywell air temperature is used in the Reference 1 safety analyses.

APPLICABLE Primary containment performance is evaluated for a SAFETY ANALYSES spectrum of break sizes for postulated loss of coolant accidents (LOCAs) (Ref. 1). Among the inputs to the design basis analysis is the initial drywell average air temperature (Ref. 1). Analyses assume an initial average drywell air temperature of 150'F. This limitation ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak drywell temperature does not exceed the maximum calculated temperature of 336'F (Ref. 2). Exceeding this temperature may result in the degradation of the primary containment structure under accident loads. Equipment inside primary containment required to mitigate the effects of a DBA is designed to operate and be capable of operating under environmental conditions expected for the accident.

Drywell air temperature satisfies Criterion 2 of the NRC Policy Statement (Ref. 3).

LCO In the event of a DBA, with an initial drywell average air temperature less than or equal to the LCO temperature limit, the resultant peak accident temperature is maintained below the maximum allowable temperature. As a result, the ability of primary containment to perform its design function is ensured.

(continued)

BFN-UNIT 2 8 3.6-28 Amendment

I 4

rywell Air Temperature B 3.6.1.4 B 3.6 CONTAINMENT SYSTEMS B 3.6. 1.4 Drywell Air Temperature BASES BACKGROUND The drywell contains the reactor vessel and piping, which add heat to the airspace. Drywell coolers remove heat and maintain a suitable environment., The average airspace temperature affects the calculated response to postulated Design Basis Accidents (DBAs). The limitation on the drywell average air temperature was developed as reasonable, based on operating experience. The limitation on drywell air temperature is used in the Reference 1 safety analyses.

APPLICABLE Primary containment performance is evaluated for a SAFETY ANALYSES spectrum of break sizes for postulated loss of coolant accidents (LOCAs) (Ref. 1). Among the inputs to the design basis analysis is the initial drywell average air temperature (Ref. 1). Analyses assume an initial average drywell air temperature of 150'F. This limitation ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak drywell temperature does not exceed the maximum calculated temperature of 336'F (Ref. 2). Exceeding this temperature may result in the degradation of the primary containment structure under accident loads. Equipment inside primary containment required to mitigate the effects of a DBA is designed to operate and be capable of operating under environmental conditions expected for the accident.

Drywell air temperature satisfies Criterion 2 of the NRC Policy Statement (Ref. 3).

LCO In the event of a DBA, with an initial drywell average air temperature less than or equal to the LCO temperature limit, the resultant peak accident temperature is maintained below the maximum allowable temperature. As a result, the ability of primary containment to perform its design function is ensured.

(continued)

BFN-UNIT 3 B 3.6-28 Amendment

4 ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPEC IFI CATION ( TS ) CHANGE TS 384 ~ SUPPLEMENT 1 COMMITMENT The primary containment structural design and the equipment located inside primary containment will be evaluated prior to Power uprate implementation to ensure the structures and equipment required for mitigation of high energy line breaks are capable of performing their required safety functions.

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