ML18036A848

Proposed Tech Specs Removing Temporary Changes Re CREVS Operation

ML18036A848
Person / Time
Site:	Browns Ferry
Issue date:	09/10/1992
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18036A847	List: ML18036A846 ML18036A848
References
NUDOCS 9209170184
Download: ML18036A848 (27)

Text

ENCLOSURE 1

PROPOSED TECHNICAL SPECIFICATION REVISIONS BROWNS FERRY NUCLEAR PLANT (BFN)

(TVA BFN TECHNICAL SPECIFICATION AMENDMENT 323) 9209170i84 9209i0 PDR ADOCK,05000259 P

PDR'

1 l

I E

4 CONTAINMENT STEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 3.7.E. Control Room Emer enc Ventilation 4.7.E Control Room Emer enc Ventilation

l. Except as specified in Specification 3.7.E.3 below, both control room emergency pressurization systems shall be OPERABLE at all times when any reactor vessel contains irradiated fuel.

1. At least once every 18 months, the pressure drop across the combined HEPA filters and charcoal adsorber banks shall be demonstrated to be less than 6 inches of water at system design flow rate (+ 10%).

2. a.

The results of the inplace cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show 299%

DOP removal and g99% halogenated hydrocarbon removal when tested in accordance with ANSI N510-1975.

2. a.

The tests and sample analysis of Specification 3.7.E.2 shall be performed at least once per operating cycle or once every 18 months, whichever occurs first for standby service or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

b. The results of laboratory carbon sample analysis shall show 290% radioactive methyl iodide removal at a velocity when tested in accordance with ASTM D3803 (130'C, 95% R.H.).

b. Cold DOP testing shall be performed after each complete or partial replacement of the HEPA filter bank or after any structural maintenance on the system housing.

BFN Unit 1 3.7/4.7-19

3 4

CONTAINMENT YSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 3.7.E. Control Room Emer enc Ventilation 4.7.E.

Control Room Emer enc Ventilation

c. System flow rate shall be shown to be within +10%

design flow when tested in accordance with ANSI N510-1975.

c. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing.

d. Each circuit shall be operated at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.

3. From and after the date that one of the control room emergency pressurization systems is made or found to be inoperable for any reason, REACTOR POWER OPERATIONS or refueling operations are permissible only during the succeeding 7 days unless such circuit is sooner made OPERABLE.

3. At least once every 18 months, automatic initiation of the control room emergency pressurization system shall be demonstrated.

4. If these conditions cannot be met, reactor shutdown shall be initiated and all reactors shall be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for REACTOR POWER OPERATIONS and refueling operations shall be terminated within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4. During the simulated automatic actuation test of this system (see Table 4.2.G), it shall be verified that the necessary dampers operate as required.

BFN Unit 1

3. 7/4. 7-20

3.7/4.7 BASES (Cont These valves are highly reliable, have low service requirements and most are normally closed.

The initiating sensors and associated trip logic are also checked to demonstrate the capability'or automatic isolation.

The test interval of once per operating c~cle for automatic initiation results in a failure probability of 1.1 x 10 that a line will not isolate.

More frequent testing for valve OPERABILITY in accordance with Specification 1.0.MM results in a greater assurance that the valve will be OPERABLE when needed.

The main steam line isolation valves are functionally tested per Specification 1.0.MM to establish a high degree of reliability.

The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system.

Each instrument line contains a 0.25-inch restricting orifice inside the primary containment and an excess flow check valve outside the primary containment.

3.7.E/4.7.E Control Room Emer enc Ventilation The control room emergency ventilation system is designed to filter the control room atmosphere for intake air and/or for recirculation during control room isolation conditions.

The control room emergency ventilation system is designed to automatically start upon control room isolation and to assist other sources of pressurization in maintaining the control room at a positive pressure.

High efficiency particulate absolute (HEPA) filters are installed prior to the charcoal adsorbers to prevent clogging of the iodine adsorbers.

The charcoal adsorbers are installed to reduce the potential intake of, radioiodine to the control room.

The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates.

The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 90 percent for expected accident conditions. If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power

Plants, Appendix A to 10 CFR Part 50.

Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

If the system is found to be inoperable, there is no immediate threat to the control room and reactor operation or refueling operation may continue for a limited period of time while repairs are being made. If the system cannot be repaired within seven days, the reactor is shutdown and'rought to Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or refueling operations are terminated.

BFN Unit 1 3.7/4.7-51

4 7 CONTAINMENT YSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 3.7.E. Control Room Emer enc Ventilation 4.7.E Control Room Emer enc Ventilation

l. Except as specified in Specification 3.7.E.3 below, both control room emergency pressurization systems shall be OPERABLE at all times when any reactor vessel contains irradiated fuel.

l. At least once every 18 months, the pressure drop across the combined HEPA filters and charcoal adsorber banks shall be demonstrated to to be less than 6 inches of water at system design flow rate

(+ 10%).

2. a.

The results of the inplace cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show 299%

DOP removal and g99% halogenated hydrocarbon removal when tested in accordance with ANSI N510-1975.

2. a.

The tests and sample analysis of Specification 3.7.E.2 shall be performed at least once per operating cycle or once every 18 months, whichever occurs first for standby service or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

b. The results of laboratory carbon sample analysis shall show g90% radioactive methyl iodide removal at a velocity when tested in accordance with ASTM D3803 (1304C, 95% R.H.).

b. Cold DOP testing shall be performed after each complete or partial replacement of the HEPA filter bank or after any structural maintenance on the system housing.

BFN Unit 2 3.7/4.7-19

4.

CONTAINMENT YSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 3.7.E. Control Room Emer enc Ventilation 4.7.E.

Control Room Emer enc Ventilation

c. System flow rate shall be shown to be within +10%

design flow when tested in accordance with ANSI N510-1975.

c. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing.

d. Each circuit shall be operated at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.

3. From and after the date. that one of the control room emergency pressurization systems is made or found to be inoperable for any reason, REACTOR POWER OPERATIONS or refueling operations are permissible only during the succeeding 7 days unless such circuit is sooner made OPERABLE.

3. At least once every 18 months, automatic initiation of the control room emergency pressurization system shall be demonstrated.

4. If these conditions cannot be met, reactor shutdown shall be initiated and all reactors shall be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for REACTOR POWER OPERATIONS and refueling operations shall be terminated within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4. During the simulated automatic actuation test of this system (see Table 4.2.G), it shall be verified that the necessary dampers operate as required.

BFN Unit 2 3.7/4.7-20

3.7/4.7 BASES (Cont

)

These valves are highly reliable, have low service requirements and most are normally closed.

The initiating sensors and associated trip logic are also checked to demonstrate the capability for automatic isolation.

The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10 that a line will not is'olate.

More frequent testing for valve OPERABILITY in accordance with Specification 1.0.MM results in a greater assurance that the valve will be OPERABLE when needed.

The main steam line isolation valves are functionally tested per Specification 1.0.MM to establish a high degree of reliability.

The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system.

Each instrument line contains a 0.25-inch restricting orifice inside the primary containment and an excess flow check valve outside the primary containment.

3.7.E/4.7.E Control Room Emer enc Ventilation The control room emergency ventilation system is designed to filter the control room atmosphere for intake air and/or for recirculation during control room isolation conditions.

The control room emergency ventilation system is designed to automatically start upon control room isolation and to assist other sources of pressurization in maintaining the control room at a positive pressure.

High efficiency particulate absolute (HEPA) filters are installed prior to the charcoal adsorbers to prevent clogging of the iodine adsorbers.

The charcoal adsorbers are installed to reduce the potential intake of radioiodine to the control room.

The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates.

The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 90 percent for expected accident conditions. If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power

Plants, Appendix A to 10 CFR Part 50.

Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

If the system is found to be inoperable, there is no immediate threat to the control room and reactor operation or refueling operation may continue for a limited period of time while repairs are being made.

If the system cannot be repaired within seven days, the reactor is shutdown and brought to Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or refueling operations are terminated.

BFN Unit 2 3.7/4.7-46

4 7 CONTAINMENT YSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 3.7.E. Control Room Emer enc Ventilation 4.7.E Control Room Emer enc Ventilation

1. Except as specified in Specification 3.7.E.3 below, both control room emergency pressurization systems shall be OPERABLE at all times when any reactor vessel contains irradiated fuel.

1. At least once every 18 months, the pressure drop across the combined HEPA filters and charcoal adsorber banks shall be demonstrated to to be less than 6 inches of water at system design flow rate

(+ 10%).

2. a.

The results of the inplace cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show g99%

DOP removal and y99% halogenated hydrocarbon removal when tested in accordance with ANSI N510-1975.

2. a.

The tests and sample analysis, of Specification 3.7.E.2 shall be performed at least once per operating cycle or once every 18 months, whichever occurs first for standby service or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

b. The results of laboratory carbon sample analysis shall show g90% radioactive methyl iodide removal at a velocity when tested in accordance with ASTM D3803 (130'C, 95% R.H.).

b. Cold DOP testing shall be performed after each complete or partial replacement of the HEPA filter bank or after any structural maintenance on the system housing.

BFN Unit 3 3.7/4.7-19

4 CONTAINMENT YSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 3.7.E. Control Room Emer enc Ventilation 4.7.E.

Control Room Emer enc Ventilation

c. System flow rate shall be shown to be within +10%

design flow when tested in accordance with ANSI N510-1975.

c. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing.

d. Each circuit shall be operated at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.

3. From and after the date that one of the control room emergency pressurization systems is made or found to be inoperable for any reason, REACTOR POWER OPERATIONS or refueling operations are permissible only during the succeeding 7 days unless such circuit is sooner made OPERABLE.

t

4. If these conditions cannot be met, reactor shutdown shall be initiated and all reactors shall be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for REACTOR POWER OPERATIONS and refueling operations shall be terminated within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

3. At least once every 18 months, automatic initiation of the control room emergency pressurization system shall be demonstrated.

4. During the simulated automatic actuation test of this system (see Table 4.2.G), it shall be verified that the necessary dampers operate as required.

BFN Unit 3 3.7/4.7-20

3.7/4.7 BASES (Con 'd)

\\

These valves are highly reliable, have low service requirements and most are normally closed.

The initiating sensors and associated trip logic are also checked to demonstrate the capability'or automatic isolation.

The test interval of once per operating c~cle for automatic initiation results in a failure probability of 1.1 x 10 that a line will not isolate.

More frequent testing for valve OPERABILITY in accordance with Specification 1.0.MM results in a greater assurance that the valve will be OPERABLE when needed.

The main steamline isolation valves are functionally tested per Specification 1.0.MM to establish a high degree of reliability.

The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system.

Each instrument line contains a 0.25-inch restricting orifice inside the primary containment and an excess flow check valve outside the primary containment.

3.7.E/4.7.E Control Room Emer enc Ventilation The control room emergency ventilation system is designed to filter the control room atmosphere for intake air and/or for recirculation during control room isolation conditions.

The control room emergency ventilation system is designed to automatically start upon control room isolation and to assist other sources of pressurization in maintaining the control room at a positive pressure.

High efficiency particulate absolute (HEPA) filters are installed prior to the charcoal adsorbers to prevent clogging of the iodine adsorbers.

The charcoal adsorbers are installed to reduce the potential intake of radioiodine to the control room.

The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates.

The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 90 percent for expected accident conditions.

If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power

Plants, Appendix A to 10 CFR Part 50.

Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

If the system is found to be inoperable, there is no immediate threat to the control room and reactor operation or refueling operation may continue for a limited period of time while repairs are being made. If the system cannot be repaired within seven days, the reactor is shutdown and brought to Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or refueling operations are terminated.

BFN Unit 3 3.7/4.7-49

ENCLOSURE 2

BROMNS FERRY NUCLEAR PLANT (BFN)

SUMMARY

OF CHANGES Summar of Chan es:

Technical Specification Amendment 323 1.

Proposed revision to Limiting Conditions for Operations (LCOs) 3.7.E.1, 3.7.E.3, and 3.7.E.4.

The current LCOs are annotated with an asterisk and the associated note states:

"CREVS is considered inoperable only because it does not meet its design basis for essentially zero unfiltered inleakage.

REACTOR POWER OPERATION and fuel movement are acceptable until just PRIOR TO STARTUP for unit 2 cycle 7.

During cycle 6, CREVS must be demonstrated to be functional by performing all applicable surveillances.

In the event that the applicable surveillances are not successfully performed, the actions required by the LCO's must be complied with."

The proposed Technical Specification deletes these asterisks and their associated notes.

2.

Proposed revision to Bases 3.7.E/4.7.E, Control Room Emergency Ventilation.

The current Bases states:

"The control room emergency ventilation system is designed to automatically start upon control room isolation and to maintain the control room pressure to the design positive pressure so that all leakage should be out leakage."

The proposed Technical Specification revises this sentence to state:

The control room emergency ventilation system is designed to automatically start upon control room isolation and to assist other sources of pressurization in maintaining the control room at a positive pressure.

3.

Proposed revision to Bases 3.7.E/4.7.E, Control Room Emergency Ventilation.

The current Bases also states:

"During cycle 6, CREVS has been declared inoperable only because it does not meet its design basis for essentially zero unfiltered inleakage.

Reactor power operations and fuel movement are acceptable until just prior to startup for unit 2 cycle 7.

During cycle 6, CREVS must be demonstrated to be functional by performing all applicable surveillances.

In the event that the applicable surveillances are not successfully performed, the actions required by the LCOs must be complied with."

The proposed Technical Specification deletes this statement.

W

ENCLOSURE 2

(CONTINUED)

BROWNS FERRY NUCLEAR PLANT (BFN)

SUMMARY

OF CHANGES Page 2 of 2 4.

Proposed revision to Surveillance Requirement 4.7.E.4.

The current Surveillance Requirement states:

"During the simulated automatic actuation test of this system (see Table 4.2.G), it shall be verified that the following dampers operate as indicated:

Close FCO 150 Bg Dg Eg F~

and G

Open:

FC0-151, FCO-152" The proposed Technical Specification revises this Surveillance Requirement to state:

During the simulated automatic actuation test of this system (see Table 4.2.G), it shall be verified that the necessary dampers operate as required.

ENCLOSURE 3

BROMNS FERRY NUCLEAR PLANT (BFN)

REASON AND JUSTIFICATION FOR THE PROPOSED CHANGES Reason for Chan es:

1)

TVA is revising Limiting Conditions for Operations (LCOs) 3.7.E.1, 3.7.E.3, and 3.7.E.4. to remove the temporary changes that were in place only for Unit 2 Cycle 6.

Failure to remove these temporary changes after Unit 2, Cycle 6 increases the probability of a failure to properly interpret the requirements of the Technical Specifications.

2)

TVA is revising Bases 3.7.E/4.7.E, Control Room Emergency Ventilation, to reflect the revised design basis for the Control Room Emergency Ventilation System (CREVS).

3)

TVA is revising Bases 3.7.E/4.7.E to remove the description of the temporary changes that were in place only for Unit 2 Cycle 6.

Failure to remove these temporary changes after Unit 2, Cycle 6 could result in the CREVS design basis not being properly interpreted.

4)

TVA is revising Surveillance Requirement 4.7.E.4 to remove the list of dampers necessary for the isolation of the Control Bay Habitability Zone and the proper alignment of the CREVS.

Removal of this component list from the Technical Specifications reduces the need for future Technical Specification amendment requests while still assuring changes to the list receive appropriate review.

Justification for Chan es:

1)

Removal of the temporary changes from LCOs 3.7.E.1, 3.7.E.3, and 3.7.E.4 does not effect any LCOs, surveillance, or other requirements for the operation CREVS for support of either of the three units after the Unit 2, Cycle 6 outage.

This is an administrative change to remove expired temporary requirements from the Technical Specifications.

2)

TVA's plan for resolving the concern regarding unfiltered inleakage from ductwork into the CBHZ and the )ustification for revising the CREVS design basis was provided by TVA's July 31, 1992 submittal and is under separate NRC review.

In summary, TVA has determined that the unfiltered inleakage is acceptable in terms of meeting the requirements of GDC 19.

Therefore, Bases 3.7.E/4.7.E are being revised to reflect the new design basis for the CREVS.

3)

Removal of the description of the temporary changes from Bases 3.7.E/4.7.E is an administrative change.

This change deletes the description of the temporary CREVS configuration.

After implementation of the CREVS corrective action plan, this description will no longer be accurate.

Page 2 of 3 ENCLOSURE 3

(CONTINUED)

BROMNS FERRY NUCLEAR PLANT (BFN)

REASON AND JUSTIFICATION FOR THE PROPOSED CHANGES 4)

Generic Letter 91-08 provides guidance for preparing a request for a license amendment to remove component lists from Technical Specifications.

This guidance provides an acceptable alternative to identifying every component by its plant identification number as it is currently listed in the Technical Specifications.

Generic Letter 91-08 requires the following issues be addresses when requesting the removal of a component list from the Technical Specifications:

A)

Each Technical Specification should include an appropriate description of the scope of the components to which the Technical Specification requi.rements apply.

Components that are defined by regulatory requirements or guidance need not

. be clarified further.

However, the Bases Section of the Technical Specification should reference the applicable requirements or guidance.

The current Technical Specification Surveillance Requirement (3.7.E.3) requires the automatic initiation of the control room emergency pressurization system be demonstrated at least once every 18 months.

This requirement meets the intent of the description of the components to which the Technical Specification requirements apply.

There are no other regulatory requirements or guidance that applies to the list of dampers necessary for the isolation of the CBHZ and the proper alignment of the CREVS..

B) If the removal of a component list results in the loss of notes that modify or provide an exception to the Technical Specification requirements, the specification should be revised to incorporate that modification or exception.

The modification or exception should be stated in terms that identify any group of components by function rather than by plant identification number, if practical.

There are no notes, modifications, or exceptions to the Technical Specification requirements that are effected by this change.

C)

Licensees should confirm that the lists of components removed from the Technical Specifications are located in appropriately controlled plant procedures.

The list of components may be included in the next update of the Final Safety Analysis Report.

The Bases Section of individual specifications also may reference the plant procedures or other documents that identify each component list.

The Bases Section for the containment isolation valve Technical Specification should be updated to describe the intent of opening valves under administrative control.

Page 3 of 3 ENCLOSURE 3

(CONTINUED)

BROMNS FERRY NUCLEAR PLANT (BFN)

REASON AND JUSTIFICATION FOR THE PROPOSED CHANGES The list of dampers required necessary for the isolation of the Control Bay Habitability Zone is included in the control room isolation and pressurization functional test procedure.

The present Technical Specification Administrative Controls Section 6.8.1.1 requires written procedures be established, implemented, and maintained for surveillance and test activities of safety related equipment.

Therefore, this procedure is subject to the change control provisions for plant procedures in the Administrative Controls section of the Technical Specifications.

ENCLOSURE 4 BROWNS FERRY NUCLEAR PLANT (BFN)

PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Descri tion of Pro osed Technical S ecification Amendment The proposed Technical Specification changes for Units 1, 2, and 3 will remove temporary requirements from LCOs 3.7.E.1, 3.7.E.3, and 3.7.E.4.

Bases 3.7.E/4.7.E are being revised to reflect the new design basis for the Control Room Emergency Ventilation System (CREVS) and to deletes the description of the configuration of the CREVS that will no longer be accurate.

The revised CREVS Bases reflects a change in BFNs previously reviewed approach to meeting General Design Criterion (GDC) 19 Control Room.

This new method of compliance with GDC 19 is currently under separate review by NRC.

The proposed changes also remove a list of dampers from the Technical Specifications.

These dampers are necessary for the isolation of the Control Bay Habitability Zone and the proper alignment of the CREVS.

Generic Letter 91-08 provides an acceptable alternative to identifying components listed in the Technical Specifications.

This list of dampers is included in the control room isolation and pressurization functional test procedure that is subject to the change control provisions for plant procedures in the Administrative Controls section of the Technical Specifications.

Basis for Pro osed No Si nificant Hazards Consideration Determination NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c).

A proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would notp (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

This proposed change does not significantly increase the probability or consequences of an accident previously evaluated.

The proposed Technical Specification change to remove the temporary revisions, which were in place only for CREVS operation during Unit 2 Cycle 6, is administrative.

The Limiting Conditions for Operation and Surveillance Requirements will be as they were prior to Unit 2 Cycle 6.

Therefore, the proposed changes will not significantly increase the consequences of an accident previously evaluated.

The removal of the temporary changes does not reflect any significant change to any precursor for the design basis events or operational transients that are analyzed in the Browns Ferry Final Safety Analysis Report.

Therefore, the probability of an accident. previously evaluated is not significantly increased.

Page 2 of 3 ENCLOSURE 4 (CowvxwueD)

BROWNS FERRY NUCLEAR PLANT (BFN)

PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The relocation of the list of dampers necessary for the isolation of the Control Bay Habitability Zone and the proper alignment of the CREVS from the Technical Specifications to a functional test procedure conforms to current NRC guidelines as cited in Generic Letter 91-08.

The proposed Technical Specification changes result in an acceptable alternative to identifying these dampers by their plan identification number as currently listed in the present BFN Technical Specifications.

The functional test procedure is subject to the change control provisions of Administrative Controls Section 6.8.1.1 of the current Technical Specifications.

This section provides an adequate means to control changes to these component lists without processing a license amendment.

Therefore, the proposed changes will not significantly increase the consequences of an accident previously evaluated.

The proposed relocation of the list of dampers does not reflect any significant change to any precursor for the design basis events or operational transients that are analyzed in the Browns Ferry Final Safety Analysis Report.

Therefore, the probability of an accident previously evaluated is not significantly increased.

2 ~

This proposed change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

The proposed Technical Specification changes to remove the temporary changes that were in place only for CREVS operation during Unit 2 Cycle 6 are administrative.

The Limiting Conditions for Operation and Surveillance Requirements will be as they were prior to Unit 2 Cycle 6.

The relocation of the list of dampers necessary for the isolation of the Control Bay Habitability Zone and the proper alignment of the CREVS from the Technical Specifications to a functional test procedure is a Technical Specification line item improvement and is in accordance with the guidance contained in Generic Letter 91-08.

The list of dampers will move from one controlled location, the Technical Specifications, to another controlled location, a plant procedure.

This plant procedure is subject to the change control provisions of Administrative Controls Section 6.8.1.1 of the current Technical Specifications.

Therefore, these proposed changes do not create the possibility of a new or different kind of accident from an accident previously evaluated.

3 ~

These proposed temporary and permanent changes do not involve a significant reduction in a margin of safety.

The proposed Technical Specification changes to remove the temporary changes that were in place only for Unit 2 Cycle 6 are administrative.

The Limiting Conditions for Operation and Surveillance Requirements will be as they were prior to Unit 2 Cycle 6.

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

Page 3 of 3 ENCLOSURE 4 (CONTINUED)

BROWNS FERRY NUCLEAR PLANT (BFN)

, PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The relocation of the list of dampers necessary for the isolation of the Control Bay Habitability Zone and the proper alignment of the CREVS from the Technical Specifications to a functional test procedure is a

Technical Specification line item improvement and is in accordance with the guidance contained in Generic Letter 91-08.

The relocation of the list of dampers will permit administrative control of changes to these lists while still assuring changes to the list receive appropriate review.

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

Determination of Basis for Pro osed No Si nificant Hazards Since the application for amendment involves a proposed change that is encompassed by the criteria for which no significant hazards consideration

exists, TVA has made a proposed determination that the application involves no significant hazards consideration.

0 0

A