ML18036A494

Proposed TS Bases 4.2 for HPCI & Reactor Core Isolation Cooling Sys.Change Clarifies That Automatic Restart Feature Tested During Performance of Logic Sys Functional Tests

ML18036A494
Person / Time
Site:	Browns Ferry
Issue date:	01/14/1992
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18036A493	List: ML18036A491 ML18036A494
References
NUDOCS 9201210286
Download: ML18036A494 (76)

Text

ENCLOSURE 1

PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT (TVA BFN TS-300) 9201210286 920i-14 PDR AGOCK 05000259 P-

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0 LIST OF EFFECTIVE PAGES TVA BFN TS-300 UNIT 1 UNIT 2 UNIT 3 Index ii Index viii 3.2/4.2-73 3 '/4.5-20 3.5/4.5-20a 3.5/4.5-20b 3.5/4.5-24a 3.5/4.5-34 3.6/4.6-12 3.6/4.6-13 3.6/4.6-14 1

3.6/4.6-32 3.6/4.6-33 3.2/4.2-73a Index ii Index iii Index iv Zndex viii 3.2/4.2-72 3.5/4.5-20 3.5/4.5-20a 3.5/4.5-25a 3.5/4.5-35 3.6/4.6-12 3.6/4.6-13 3.6/4.6-14 3.6/4.6-32

gf Section D.

Reactivity Anomalies E.

Reactivity Control F.

Scram Discharge Volume A.

Normal System Availability B.

Operation with Inoperable Components C.

Sodium Pentaborate Solution.

3.5/4.5 Core and Containment Cooling Systems.

A.

Core Spray System (CSS).

3.4/4.4 Standby Liquid Control System

~Pe e No.

3.3/4.3-11 3.3/4.3-12 3.3/4.3-12 3.4/4.4-1 3.4/4.4-1 3.4/4.4-3 3.4/4.4-3 3.5/4.5-1 3.5/4.5-1 B.

Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling)

C.

RHR Service Water and Emergency Equipment Cooling Water Systems (EECWS).

D.

Equipment Area Coolers E.

High Pressure Coolant Injection System (HPCIS)

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3.5/4.5-4 3.5/4.5-9 3.5/4.5-13 3.5/4.5-13 F.

Reactor Core Isolation Cooling System (RCICS)

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Automatic Depressurization System (ADS).

H.

Maintenance of Filled Discharge Pipe I.

Average Planar Linear Heat Generation Rate J.

Linear Heat Generation Rate (LHGR) 3.5/4.5-14 3.5/4.5-16 3.5/4.5-17 3.5/4.5-18 3.5/4.5-18 K.

Minimum Critical Power Ratio (MCPR).

3.5/4.5-19 3.6/4.6 L.

APRM Setpoints M.

Core Thermal-Hydraulic Stability.

Primary System Boundary A.

Thermal and Pressurization Limitations B.

Coolant Chemistry.

C.

Coolant Leakage.

D.

Relief Valves.

3.5/4.5-20 3.5/4.5-20 3.6/4.6-1 3.6/4.6-1 3.6/4.6-5 3.6/4.6-9 3.6/4.6-10 Unit 1

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IST OF LLUSTRAT 0 S

.~F1 are Title

~Pa e No 2.1.1 2.1-2 APRM Flow Reference Scram and APRM Rod Block Settings APRM Flow Bias Scram Vs. Reactor Core Flow 1.1/2.1-6 1.1/2.1-7 4.1-1 4.2-1 3.5.K-1 3.5.M-l 3.5.2 3.6-1 Graphical Aid in the Selection of an Adequate Interval Between Tests System Unavailability.

MCPR Limits.

BFN Power/Flow Stability Regions Kf Factor.

Minimum Temperature F Above Change in Transient Temperature.

3.1/4.1-13 3.2/4.2-64 3.5/4.5-24 3.5/4.5-24a 3.5/4.5-25 3.6/4.6-24 3.6-2 4.8.1.a 4.S'il.b Change in Charpy V Transition Temperature Neutron Exposure Gaseous Release Points and Elevations.

Land Site Boundary Vs.

3.6/4.6-25 3.8/4.8-10 3.8/4.8-,11 O

BFN Unit 1 viii

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4.2

~S S (Cont'd) 0 The conclusions to be drawn are these:

l.

A 1-out-of-n system may be treated the same as a single channel in terms of choosing a test interval; and 2.

more than one channel should not be bypassed for testing at any one time.

The radiation monitors in the refueling area ventilation duct which initiate building isolation and standby gas treatment operation are arranged in two 1-out-of-2 logic systems.

The bases given for the rod blocks apply here also and were used to arrive at the functional testing frequency.

The off-gas post treatment monitors are connected in a 2-out-of-2 logic arrangement.

Based on experience with instruments of similar design, a testing interval of once every three months has been found adequate.

The automatic pressure relief instrumentation can be considered to be a

1-out-of-2 logic system and the discussion above applies also.

The criteria for ensuring the reliability and accuracy of the radioactive gaseous effluent instrumentation is listed in Table 4.2.K.

The criteria for ensuring the reliability and accuracy of the radioactive liquid effluent instrumentation is listed in Table 4.2.D.

V The RCIC and HPCI system logic tests required by Table 4.2.B contain provisions to demonstrate that these systems will automatically restart on a RPV low water level signal received subsequent to a RPV high water level trip.

Unit 1 3.2/4.2-73

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CORE AND CONTAINMENT COOLING SYSTEMS

'"LIMITING CONDITIONS FOR OPERATION SURVEIL'LANCE RE UIREMENTS 3.5 Core and Containment Coolin S stems 4.5 Core and Containment Coolin

~Sseems L.

APRM Set pints L.

APRM Set pints 1.

Whenever the core thermal power is g 25% of rated, the ratio of FRP/CMFLPD shall be 2 1.0, or the APRM scram and rod block setpoint equations listed in Sections 2.1.A and 2.1eB shall be multiplied by FRP/CMFLPD as follows:

Sg (O.66W + 54%) >>P CMFLPD FRP/CMFLPD shall be determined daily when the reactor is g 25% of rated thermal power.

SRB< (0.66W + 42%)

(FRP

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CMFLPD 2.

When it is determined that 3.5.L.l is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to correct the condition.

3. If 3.5.L.1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to g 25% of rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

M.

Core Thermal-H draulic Stabilit M.

Core Thermal-H draulic Stabilit 1.

The reactor shall not be operated at a thermal power and core flow inside of Regions I and II of Figure 3.5.M-l.

2. If Region I of Figure 3.5.M-1 is entered, immediately initiate a manual scram.

3. If Region II of Figure 3.5.M-1 is entered:

1.

Verify that the reactor is outside of Region I and II of Figure 3.5.M-l:

a.

Following any increase of more than 5% rated thermal power while initial core flow is less than 45% of

rated, and O

BFN Unit 1 3.5/4.5-20

4 CORK AND CO AINMENT COOLING SYSTEMS LIMITING CONDITIO S

FOR OPERATION SURVEILLANCE RE UIREMENTS 3'.5.M.

Core Thermal-H draulic Stabilit 4.5.M.

Core Thermal-H draulic Stabilit 3.5.M.3 (Cont'd) a.

Immediately initate action and exit the region within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by inserting control rods or by increasing core flow (starting a

recirculation pump to exit the region is not an appropriate action),

and 4.5.M.1 (Cont'd) b.

Following any decrease of more than 10% rated core flow while initial thermal power is greater than 40% of rated.

b.

While exiting the region, immediately initiate a manual scram if thermal-hydraulic instability is

observed, as evidenced by APRM oscillations which exceed 10 percent peak-to-peak of rated or LPRM oscillations which exceed 30 percent peak-to-peak of scale.

If periodic LPRM upscale or downscale alarms occur, immediately check the APRM's and individual LPRM's for evidence of thermal-hydraulic instability.

BFN Unit 1 3.5/4.5-20a

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THIS PAGE INTENTIONALLYLEFT BLANK Unit 1 3.5/4.5-20b

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3.5 BASES (Cont'd)

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,3.5.M Core Ther a d aulic Stabi it The minimum margin to the onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1.

A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which could challenge the MCPR safety limit.

Because the probability of thermal-hydraulic oscillations is lower and the margin to the MCPR safety limit is greater in Region II than in Region I of Figure 3.5.M-1, an immediate scram upon entry into the region is not necessary.

However, in order to minimize the probability of core instability following entry into Region II, the operator will take immediate action to exit the region.

Although formal surveillances are not performed while exiting Region II (delaying exit for surveillances is undesirable),

an immediate manual scram will be initiated if evidence of thermal-hydraulic instability is observed.

Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regional oscillations).

Periodic LPRM upscale or downscale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated.

During regional oscillations, the safety limit MCPR is not approached until APRM oscillations are 30 percent peak-to-peak or larger in magnitude.

In addition, periodic upscale or downscale LPRM alarms will occur before regional oscillations are large enough to threaten the MCPR safety limit.

Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit will not be violated in the event that core oscillations initiate while exiting Region II.

Normal operation of the reactor is restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur.

3.5.N. geferenees 1.

"Fuel Densification Effects on General Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEIM-10735, August 1973.

2.

Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 14, 1974 (USA Regulatory Staff).

3.

Communication:

V. A. Moore to I. S. Mitchell, "Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974.

4.

Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.

5.

Letter from R. H. Buchholz (GE) to P.

S.

Check (NRC), "Response to NRC Request For Information On ODYN Computer Model," September 5,

1980.

Unit 1 3.5/4.5-34

0

4 PRIMARY SYS BOUNDAR

, i LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 4.6.E.

J~et Pum e

2.

Whenever there is recirculation flow with the reactor in the STARTUP or RUN Mode and one recirculation pump is operating with the equalizer valve closed, the diffuser to lower plenum differential pressure shall be checked daily and the differential pressure of an individual jet pump in a loop shall not vary from the mean of all jet pump differential pressures i'n that loop by more than 10%.

3.6.F Recirculation Pum 0 eration 4.6.F Recirculation Pum 0 eration The reactor shall not be operated with one recirculation loop out of service for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With the reactor operating, if one recirculation loop is out of

service, the plant shall be placed in a HOT SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is sooner returned to service.

1.

Recirculation pump speeds shall be checked and logged at least once per day.

2.

Following one pump operation, the discharge valve of the low speed pump may not be opened unless the speed of the faster pump is less than 50% of its rated speed.

2.

No additional surveillance required.

3.

When the reactor is not in the RUN mode, REACTOR POWER OPERATION with both recirculation pumps out-of-service for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is permitted.

During such interval, restart of the recirculation pumps is permitted, provided the loop discharge temperature is within 75'F of the saturation temperature of 3.

Before starting either recirculation pump during REACTOR POWER OPERATION, check and log the loop discharge temperature and dome saturation temperature.

Unit 1 3.6/4.6-12

4 f

4 PRIMARY SYSTEM BOU DA Y

, LIMITI G CONDITIO S FOR OPERATIO 3.6.F Recirculation Pum 0 eration SURVEILLANCE RE UIREMENTS 3.6.F.3 (Cont'd) the reactor vessel water as determined by dome pressure.

The total elapsed time in natural circulation and one pump operation must be no greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUN mode.

Following a trip of both recirculation pumps while in the RUN mode, immediately initiate a manual reactor scram.

3.6.G.

St uctural te rit 4.6.G.

Structural Inte rit The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout, the life of the plant.

a.

With the structural integrity of any ASME Code Class 1

equivalent component, which is part of the primary system, not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or maintain the reactor coolant system in either a Cold Shutdown condition or less than 50'F above the minimum temperature required by NDT considerations, until each indication of a defect has been investigated and evaluated.

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Additional inspections shall be performed on certain circumferential pipe welds as listed to provide additional protection against pipe whip, which could damage auxiliary and control systems.

Feedwater GFW-9, GFW-12, KFW-31, KFW-39, KFW-38, KFW-13 GFW-26, GFW-29, GFW-15, and GFW-32 Inservice inspection of ASME Code Class 1, Class 2,

and Class 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g),

except where specific written relief has been granted by NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

Main steam GMS-6, KMS-24, GMS-32, KMS-104 GMS-15, and GMS-24 Unit 1 3.6/4.6-13

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4 PRIMARY SYS EM BOU DARY LINI'TING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS

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3.6.G Structu a

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3.6.G.l (Cont'd) 4.6.G Structu al nte rit 4.6.G.2 (Cont'd) b.

With the structural integrity of any ASME Code Class 2 or 3 equivalent component not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component from all OPERABLE systems.

RHR DSRHR-4, DSRHR-7, DSRHR-8A Reactor Cleanup DSRWC-4, DSRWC-3 DSRWC-6, DSRWC-5 Core Spray DSCS-12, DSCS-11, DSCS-5, and DSCS-4 HPCI THPCI 152 THPCI 153B THPCI 153 THPCI 154 3.

For Unit 1 an augmented inservice surveillance program shall be performed to monitor potential corrosive effects of chloride residue released during the March 22, 1975 fire.

The augmented inservice surveillance program is specified as follows:

a.

Browns Ferry Mechanical Maintenance Instruction 53, dated September 22,

1975, paragraph 4, defines the liquid penetrant examinations required during the first, second, third and fourth refueling outages following the fire restoration.

b.

Browns Ferry Mechanical Maintenance Instruction 46, dated July 18, 1975.

Appendix B, defines the liquid penetrant examinations required during the sixth refueling outage following the fire restoration.

BFN Unit 1 3.6/4.6-14

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3.6/4.6 BASES 3.6.'E/4.6.E (Cont'd) 0 If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation.

If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.

If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced;

hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure). If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed.

Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.

In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate.

The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured.

This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.

Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.

A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true.

The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.

3.6.F/4.6.F Recirculat on Pum 0 eration Operation without forced recirculation is permitted for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is not in the RUN mode.

And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant

, temperature is less than 75'F.

This reduces the positive reactivity insertion to an acceptably low value.

Requiring at least one recirculation pump to be OPERABLE while in the RUN mode (i.e., requiring a manual scram if both recirculation pumps are tripped) provides protection against the potential occurrence of core thermal-hydraulic instabilities at low flow conditions.

Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50K of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.

3.6.G/4.6.G Structural Inte rit The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling BFN Unit 1 3.6/4.6-32

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. 3.6/4.6 BASES 3.6.G/4.6.G (Cont'd) examination of areas of high stress and -highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.

'a The program reflects the built-in limitations of access to the reactor coolant systems.

It is intended that the required examinations and inspection be completed during each 10-year interval.

The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.

Only proven nondestructive testing techniques will be used.

More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip.

These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.

Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings.

Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.

An augmented inservice surveillance program is required to determine whether any stress corrosion has occurred in any stainless steel piping, stainless components, and highly-stressed alloy steel such as hanger

springs, as a result of environmental conditions associated with the March 22, 1975 fire.

REFERElERcs 1.

Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) 2.

Inservice Inspection of Nuclear Reactor Coolant Systems, Section KI, ASME Boiler and Pressure Vessel Code 3.

ASME Boiler and Pressure Vessel

Code,Section III (1968 Edition) 4.

American Society for Nondestructive Testing No. SNT-TC-1A (1968 Edition) 5.

Mechanical Maintenance Instruction 46 (Mechanical Equipment,

Concrete, and Structural Steel Cleaning Procedure for Residue From Plant Fire Units 1 and 2) 6.

Mechanical Maintenance Instruction 53 (Evaluation of Corrosion Damage of Piping Components Which Were Exposed to Residue From March 22, 1975 Fire) 7.

Plant Safety Analysis (BFNP FSAR Subsection 4.12)

BFN Unit 1 3.6/4.6-33

'lÃwi

0 4.2 BASES (Cont'd)

I The criteria for ensuring the reliability and accuracy of the radioactive gaseous effluent instrumentation is listed in Table 4.2.K.

The criteria for ensuring the reliability and accuracy of the radioactive liquid effluent instrumentation is listed in Table 4.2.D.

The RCIC and HPCI system logic tests required by Table 4.2.B contain provisions to demonstrate that these systems will automatically restart on a

RPV low water level signal received subsequent to a RPV high water level trip.

Unit 2 3.2/4.2-73a

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0 Section 3.4/4.4 3.5/4.5

~Pa e Ne D.

Reactivity Anomalies 3.3/4.3-11 E.

Reactivity Control 3.3/4.3-12 F.

Scram Discharge Volume 3.3/4.3-12 Standby Liquid Control System 3.4/4.4-1 A.

Normal System Availability.

3.4/4.4-1 B.

Operation with Inoperable Components 3.4/4.4-3 C.

Sodium Pentaborate Solution.

3.4/4.4-3 Core and Containment Cooling Systems.

3.5/4.5-1 A.

Core Spray System (CSS).

3.5/4.5-1 B.

Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling) 3.5/4.5-4 C.

RHR Service Water and Emergency Equipment Cooling Water Systems (EECWS).

3.5/4.5-9 D.

Equipment Area Coolers 3.5/4.5-13 E.

High Pressure Coolant Injection System (HPCIS).

3.5/4.5-13 F.

Reactor Core Isolation Cooling System (RCICS).

3.5/4.5-14 G.

Automatic Depressurization System (ADS).

3.5/4.5-16 H.

Maintenance of Filled -Discharge Pipe 3.5/4.5-17 I.

Average Planar Linear Heat Generation Rate 3.5/4.5-18 J.

Linear Heat Generation Rate (LHGR) 3.5/4.5-18 3.6/4.6 K.

Minimum Critical Power Ratio (MCPR).

L, APRM Setpoints M.

Core Thermal-Hydraulic Stability Primary System Boundary 3.5/4.5-19 3.5/4.5-20 3.5/4.5-20 3.6/4.6-1 A.

Thermal and Pressurization Limitations 3.6/4.6-1 B.

Coolant Chemistry.

3.6/4.6-5 C.

Coolant Leakage.

3.6/4.6-9 Unit 3

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Relief Valves.

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3.6/4.6-10 3.6/4.6-11 F.'-

- Recirculation Pump Operation 3.6/4.6-12 G.

Structural Integrity'

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H.

Snubbers 3.7/4.7 iContainment Systems A.

Primary Containment.

3.6/4.6-13 3.6/4.6-15 3.7/4.7-1 3.7/4.7-1 B.

Standby Gas Treatment System 3.7/4.7-13 C.

Secondary Containment.

'C D.

Primary Containment Isolation Valves 3.7/4.7-16 3.7/4.7-17 E.

"Control Room Emergency Ventilation.

3.7/4.7-19 F.

Primary Containment Purge System 3.7/4.7-21 G.

Containment Atmosphere Dilution System (CAD) 3,7/4.7-22 A.

Liquid Effluents B.

Airborne Effluents C.

Radioactive Effluents - Dose D.

Mechanical Vacuum Pump

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Miscellaneous Radioactive Materials Sources H.

Containment Atmosphere Monitoring (CAM)

System H2 Analyzer 3.8/4.8 Radioactive, Materials 3.7/4.7-23a 3.8/4.8-1 3.8/4.8-1 3.8/4.8-2 3.8/4.8-6 3.8/4.8-6 3.8/4.8-7 F.

Solid Radwaste 3.8/4.8-9 3.9/4.9 Auxiliary Electrical System 3.9/4.9-1 A.

Auxiliary Electrical Equipment 3.9/4.9-1 B.

Operation with Inoperable Equipment.

3.9/4.9-8 C.

Operation in Cold Shutdown 3.9/4.9-14 Unit 3

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Core Alterations

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Refueling Interlocks B.

Core Monitoring

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3.10/4.10-1 3.10/4.10-1 3.10/4.10-4 C.

Spent Fuel Pool Water 3.10/4.10-7 D.

Reactor Building Crane 3.10/4.10-8

,E.

Spent Fuel Cask 3.10/4.10-9 F.

Spent Fuel Cask Handling-Refueling Floor 3.10/4.10-9 3.11/4.11 Fire Protection Systems 3.11/4.11-1 A.

Fire Detection Instrumentation 3.11/4.11-1 B.

Fire Pumps and Water Distribution Mains

'<<C.

-Spray and/or Sprinkler Systems D.

C02 Systems

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3.11/4.11-2, 3.11/4.11-7 3.11/4.11-8 E.

Fire Hose Stations 3.11/4.11-9 F.

Yard Fire Hydrants and Hose Houses G.

Fire-Rated Assemblies

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3.11/4.11-11 3.11/4,11-12 5.0 H.

Open Flames, Welding and

,Spreading Room Ma)or Design Features, Burning in the Cable

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3.11/4.11-13 5.0-1 5.1 Site Features 5.2 Reactor

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5.0-1 5.0-1 5.3 Reactor Vessel 5.0-1 5.4 Containment 5.0-1 5.5 Fuel Storage.

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5.0-1 5.6 Seismic Design 5.0-2 Unit 3 iv

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2.1.1

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~Pa e No APRM Flow Reference Scram and APRM Rod Block Settings 1.1/2.1-6 2.1-2 4.1-1 APRM Flow Bias Scram Vs. Reactor Core Flow 1.1/2.1-7 Graphical Aid in the Selection of an Adequate Interval Between Tests 3.1/4.1-12 4.2-1 System Unavailability.

3.5.K-1 MCPR Limits.

3.2/4.2-63 3.5/4.5-25 3.5.M-1 BFN Power/Flow Stability Regions 3.5/4.5-25a 3.5.2 Kf Factor.

3.5/4.5-26 3.6-1 Minimum Temperature F Above Change in Transient Temperature.

3.6/4.6-24 3.6-2 Change in Charpy V Transition Temperature Neutron Exposure Vs.

3.6/4.6-25 4.8.1.a Gaseous Release Points and Elevation 3.8/4.8-10 3.8/4.8-11 Unit 3 viii

4.2 BASES (Cont'd)

The conclusions to be drawn are these:

l.

A 1-out-of-n system may be treated the same as a single channel in

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terms of choosing a test interval; and 2.

more than one channel should not be bypassed for testing at any one time.

The radiation monitors in the refueling area ventilation duct which initiate building isolation and standby gas treatment operation are arranged in two 1-out-of-2 logic systems.

The bases given for the rod blocks apply here also and were used to arrive at the functional testing frequency.

The off-gas post treatment monitors are connected in a 2-out-of-2 logic arrangement.

Based on experience with instruments of similar design, a testing interval of once every three months has been found adequate.

The automatic pressure relief instrumentation can be considered to be a

1-out-of-2 logic system and the discussion above applies also.

The criteria for ensuring the reliability and accuracy of the radioactive gaseous effluent instrumentation is listed in Table 4.2.K.

The criteria for ensuring the reliability and accuracy of the radioactive liquid effluent instrumentation is listed in Table 4.2.D.

The RCIC and HPCI system logic tes'ts required by Table 4.2.B contain provisions to demonstrate that these systems will automatically restart on a

RPV low water level signal received subsequent to a RPV high water level trip.

BFN Unit 3 3.2/4.2-72

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4 'ORE A

D CONT NMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 3.5 Core and Containment Coolin S stems 4.5 Core and Containment Coolin

~sseems L.

APRM Set pints 1.

Whenever the core thermal power is g 25% of rated, the ratio of FRP/CMFLPD shall be g 1.0, or the APRM scram and rod block setpoint equations listed in Sections 2.1.A and 2.1.B shall be multiplied by FRP/CMFLPD as follows:

Sq (O.66W + 54%)

FRP CMFLPD FRP/CMFLPD shall be determined daily when the reactor is g 25% of rated thermal power.

SRB~ (0 66W + 42%)

(FRP

)

CMFLPD 2.

When it is determined that 3.5.L.1 is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to correct the condition.

3. If 3.5.L.1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to g 25% of rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

M.

Core Thermal-H draulic Stabilit M.

Core Thermal-H draulic Stabilit 1.

The reactor shall not be operated at a thermal power and core flow inside of Regions I and II of Figure 3.5.M-l.

2. If Region I of Figure 3.5.M-1 is entered, immediately initiate a manual scram.

3. If Region II of Figure 3.5.M-1 is entered:

1.

Verify that the reactor is outside of Region I and II of Figure 3.5.M-1:

a.

Following any increase of more than 5% rated thermal, power while initial core flow is less than 45% of

rated, and BFN Unit 3 3.5/4.5-20

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3 5 4 CORE A D CO AINMENT COOLI G SYSTEMS t

IMITING CONDITIO S FOR OPERATION 3.5.M.

Core Thermal-H draulic Stabilit 3.5.M.3. (Cont'd) a.

Immediately initiate action and exit the region within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by inserting control rods or by increasing core flow (starting a recircu-lation pump to exit the region is not an appropriate action),

and SURVEILLA CE RE UIREMENTS 4.5.M.

Core Thermal-H draulic Stabilit 4.5.M.1 (Cont'd) b.

Following any decrease of more than 10/ rated core flow while initial thermal power is greater than 40% of rated.

b.

While exiting the region, immediately initiate a manual scram if thermal-hydraulic instability is observed, as evidenced by APRM oscilla-tions which exceed 10 percent peak-to-peak of rated or LPRM oscillations which exceed 30 percent peak-to-peak of scale.

If periodic LPRM upscale or downscale alarms

occur, immediately check the APRM's and individual LPRM's for evidence of thermal-hydraulic instability.

Unit 3 3.5/4.5-20a

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A six-hour time period to achieve this condition is justified since the additional margin gained by the setdown adjustment is above and beyond that ensured by the safety analysis.

3.5.M Core Thermal-H draulic Stabilit The minimum margin to the onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-l.

A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which could challenge the MCPR safety limit.

Because the probability of thermal-hydraulic oscillations is lower and the margin to the MCPR safety limit is greater in Region II than in Region I of Figure 3.5.M-l, an immediate scram upon entry into the region is not necessary.

However, in order to minimize the probability of core instability following entry into Region II, the operator will take immediate action to exit the region.

Although formal surveillances are not performed while exiting Region II (delaying exit for surveillances is undesirable),

an immediate manual scram will be initiated if evidence of thermal-hydraulic instability is observed.

Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regional oscillations).

Periodic LPRM upscale or downscale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated.

During regional oscillations, the safety limit MCPR is not approached until APRM oscillations are 30 percent peak-to-peak or larger in magnitude.

In addition, periodic upscale or downscale LPRM alarms will occur before regional oscillations are large enough to threaten the MCPR safety limit.

Therefore, the criteria for initiating manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit will not be violated in the event that-core oscillations initiate while exiting Region II.

, Normal operation of the reactor is restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur.

3.5.N References 1.

Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3, NEDO-24194A and Addenda.

2.

"BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

3.

Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.

BFN Unit 3 3.5/4.5-35

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4 PRIMARY SYSTE BOU DARY LIMITING CO ITIONS FOR OPERATIO SURVEILLA CE RE UIREME S

4.6.E.

~Jet Pum s 2.

Whenever there is recirculation flow with the reactor in the STARTUP or RUN Mode and one recirculation pump is operating with the equalizer valve closed, the diffuser to lower plenum differential pressure shall be checked daily and the differential pressure of an individual jet pump in a loop shall not vary from the mean of all jet pump differential pressures in that loop by more than 10%.

3.6.F Recircu ation Pum 0 eratio 4.6.F Recirculation Pum 0 eration The reactor shall not be operated with one recirculation loop out of service for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With the reactor operating, if one recirculation loop is out of

service, the plant shall be placed in a HOT SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is sooner returned to service.

1.

Recirculation pump speeds shall be checked and logged at least once per day.

2.

Following one-pump operation, the discharge valve of the low speed pump may not be opened unless the speed of the faster pump is less than 50% of its rated speed.

2.

No additional surveillance required.

3.

When the reactor is not in the RUN mode, REACTOR POWER OPERATION with both recircu-lation pumps out-of-service for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is permitted.

During such interval, restart of the recirculation pumps is permitted, provided the loop discharge temperature is within 75'F of the saturation temperature of the reactor 3.

Before starting either recirculation pump during REACTOR POWER OPERATION, check and log the loop discharge temperature and dome saturation temperature.

Unit 3 3.6/4.6-12

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MARY SYSTEM BOU D

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LIMITI G CO D TIO S

FOR OPERATIO 3.6.F Recirculation Pum 0 eration 3.6.F.3 (Cont'd)

SU VEILLA CE RE UI EMENTS vessel water as determined by dome pressure.

The total elapsed time in natural circulation and one pump operation must be no greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.

The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUN mode.

Following a trip of both recirculation pumps while in the RUN mode, immediately initiate a manual reactor scram.

3.6.G Structura nte rit 4.6.G Structural Inte rit 1.

The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.

a.

With the structural integrity of any ASME Code Class 1 equivalent component, which is part of the primary system, not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or maintain the reactor coolant system in either a Cold Shutdown condition or less than 50'F above the minimum temperature required by NDT considera-tions, until each indication of a defect has been inves-tigated and evaluated.

l.

Inservice inspection of ASME Code Class 1, Class 2,

and Class 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g),

except where specific written relief has been granted by NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

2.

Additional inspections shall be performed on certain circumferential pipe welds as listed to provide additional protection against pipe whip, which could damage auxiliary and control systems.

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PRIMARY SYSTEM BOUNDARY LIM2T'ING CONDITIO S

FOR OPERATIO SURVEILLANCE RE UIREME S

3.6.G Structural Inte rit 3.6.G.1 (Cont'd) b.

With the structural integrity of any ASME Code Class 2 or 3 equivalent component not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component from all OPERABLE systems.

4.6.G Structural Inte rit 4.6.G.2 (Cont'd)

Feedwater GFW-9, GFW-12, KFW-31, KFW-39, KFW-38, KFW-13, GFW-26, GFW-29 GFW-15, and GFW-32 RHR DSRHR-6, DSRHR-7, DSRHR-4 Main steam GMS-6, KMS-24, GMS-32, KMS-104, GMS-15, and GMS-24 Core Spray Reactor Cleanup HPCI TSC-407, TSC-423, TSCS-408, TSC-424 DSRWC-4, DSRWC-3 DSRWC-6, DSRWC-5 THPCI 70 THPCI 70A THPCI 71, and THPCI 72 REFERENCE 1.

Plant Safety Analysis (BFN FSAR subsection 4.12) t BFN Unit 3 3.6/4.6-14

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~ 3.6/4.6 BASES 3.'6.E/4.6.E (Cont'd)

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~area is increased, the system resistance to the recirculation pump is also reduced;

hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure). If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed.

Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.

In

addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate.

The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured.

This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.

Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would bh less than the normal forward flow.

A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true.

The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.

3.6.F/4.6.F Recirculat on Pum 0 eration Operation without forced recirculation is permitted for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is not in the RUN mode.

And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75'F.

This reduces the positive reactivity insertion to an acceptably low value.

Requiring at least one recirculation pump to be OPERABLE while in the RUN mode (i.e., requiring a manual scram if both recirculation pumps are tripped) provides protection against the potential occurrence of core thermal-hydraulic instabilities at low flow conditions.

Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50 percent of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.

3.6.G/4.6.G Structural Inte rit The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the; ASME Boiler and Pressure Vessel Code.

The program reflects the built-in limitations of access to the reactor coolant systems.

It is intended that the required examinations and inspection be completed during each 10-year interval.

The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.

BFN 3.6/4.6-32 Unit 3

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ENCLOSURE 2

BROWNS FERRY NUCLEAR PLANT DESCRIPTION AND JUSTIFICATION FOR THE PROPOSED CHANGES Summar of Chan es for Units 1 and 3:

1.

Add LCO and Actions 3.5.M for Core Thermal Hydraulic Stability as follows:

3.5.M.

Core and Thermal-H draulic Stabilit 1.

The reactor shall not be operated at a thermal power and core flow inside of Regions I and IZ of Figure 3.5.M-l.

2. If Region I of Figure 3.5.M-1 is entered, immediately initiate a manual scram.

3. If Region II of Figure 3.5.M-1 is entered:

a ~

Immediately initiate action and exit the region within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by

'inserting control rods or by increasing core flow (starting a recirculation pump to exit the region is not an appropriate action),

and b.

While exiting the region, immediately initiate a manual scram if thermal-hydraulic instability is observed, as evidenced by:APRM oscillations which exceed 10 percent peak-to-peak of rated or LPRM oscillations which exceed 30 percent peak-to-peak of scale.

If periodic LPRM upscale or downscale alarms occur, immediately check the APRM's and individual LPRM's for evidence of thermal-hydraulic instability.

2.

Add Surveillance Requirements (SR) 4.5.M for Thermal-Hydraulic Stability as follows:

4.5.M.

Core and Thermal-H draulic Stabilit

1. 'erify that the reactor is outside of Region I and II of Figure 3.5.M-1:

a.

Following any increase of more than 5% rated thermal power while initial core flow is less than 45% of rated, and b.

Following any decrease of more than 10%

rated core flow while initial thermal power is greater than 40% of rated.

3.

Add Figure 3.5.M-1 (see attached) 4.

Add Bases 3.5.M to read as follows:

The minimum margin to onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1.

A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which could challenge the MCPR safety limit.

Because the probability of thermal-hydraulic oscillations is lower and the margin to the MCPR safety limit is greater in Region II than in Region I of Figure 3.5.M-1, an immediate scram upon entry into the region is not necessary.

However, in order to minimize the probability of core instability following entry into Region II, the operator willtake immediate action to exit the region.

Although formal surveillances are not performed

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2 while exiting Region II (delaying exit for surveillances is undesirable),

an

"~immediate manual scram will.be. initiated if evidence of thermal-hydraulic instability is observed.

Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10%

during regional oscillations).

Periodic LPRM upscale or downscale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated.

During regional oscillations, the safety limit MCPR is not approached until APRM oscillations are 30 percent peak-to-peak or larger in magnitude.

In

addition, periodic upscale or downscale LPRM alarms will occur before regional oscillations are large enough to threaten the MCPR safety limit.

Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limitwill not be violated in the event that core oscillations initiate while exiting Region II.

5.

Normal operation of the reactor is restricted to thermal power and core flow conditions '(i.e;,

outside Regions I

and II) where thermal-hydraulic instabilities are very unlikely to occur.

Change existing LCO 3.6.F.3 as follows:

Existing LCO 3.6.F.3 reads:

Steady-state operation with both recirculation pumps out-of-service for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is permitted.

During such interval restart of the recirculation pumps is permitted, provided the loop discharge temperature is within 75 F

of the saturation temperature of...

Proposed LCO 3.6.F.3 reads:

When the reactor is not in the RUN mode, REACTOR POWER OPERATION with both recirculation pumps out-of-service for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is permitted.

During such interval, restart of the, recirculation. pumps, is.permitted, provided the loop discharge temperature is within 75 'F of the saturation temperature ofo ~

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6.

Change existing Surveillance Requirement 4.6.F.3 as follows:

Existing SR 4.6.F.3 reads:

Before starting either recirculation pump during steady-state operation, check and log the loop discharge temperature and dome saturation temperature.

Proposed SR 4.6.F.3 reads:

Before starting either recirculation pump during REACTOR POWER OPERATION, check and log the loop discharge temperature and dome saturation temperature.

7 ~

Add proposed LCO 3.6.F.4 to read as follows:

The reactor shall not be operated with both recirculation pumps out-of-service while the. reactor is in the RUN mode.

Following a trip of both recirculation pumps while in the RUN mode, immediately initiate a manual reactor scram.

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Change Bases 3.6.F/4.6.F as follows:

Existing Bases 3.6.F/4.6.F reads in part:

Steady-state operation without forced recirculation will not be permitted for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and core coolant temperature is less is less than 75 F.

This reduces the positive reactivity insertion to an acceptably low value.

Proposed Bases 3.6.F/4.6.F reads in part:

Operation without forced recirculation is permitted for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is not is the RUN mode.

And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and core coolant temperature is less is less than 75

'F.

This reduces the positive reactivity insertion to an acceptably low value.

- Requiring at least one recirculation pump to be.OPERABLE while in the RUN

'mode (i.e.,

requiring a

manual scram if both recirculation pumps are tripped) provides protection against the potential occurrence of core thermal-hydraulic instabilities at low flow conditions.

Summar of Chan es for Units 1

2 and 3

1.

Add the following provision to Bases 4.2:

The RCIC and HPCI system logic tests required by Table 4.2.B contain provisions to demonstrate that these systems will automatically restart on an RPV low water level signal received subsequent to a RPV high water level trip.

JUSTIFICATION FOR THE PROPOSED CHANGES Reason for Chan e

BFN Units 1 and 3 Technical Specifications Sections 3.5.M/4.5.M and Figure 3.5.M-1 are being added and Section 3.6.F/4.6.F is being revised to incorporate Surveillance Requirements (SR) and Limiting Conditions for Operation (LCO) for reactor core thermal-hydraulic stability.

These changes are being proposed to implement the recommendations of NRC Bulletin 88-07, Supplement 1 for Units 1 and 3.

Implementation of these provisions occurred on Unit 2 prior to restart.

Bases 4.2 for BFN Units 1,

2, and 3 are being revised to clarify testing requirements for HPCI and RCIC.

The proposed change to the Bases will describe the automatic restart on a low water level initiation, subsequent to a high water level trip test, as being part of the system logic functional tests in Table 4.2.B.

Back round on Thermal-H draulic Stabilit Chan es General Design Criteria (GDC) 12 requires that reactor power oscillations either be (1) prevented or (2) detected and suppressed.

The stability licensing basis for U. S. Boiling Water.Reactors.(BWRs).has been either that oscillations will

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not occur in allowable operating regions (as demonstrated by decay ratio

=""calculations) or that oscillations can be detected and suppressed by reactor.

operators before protection limits are exceeded.

In the past, BFN demonstrated

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compliance with GDC 12 by performing decay ratio analyses for each core reload

>>>'to. show thaticore.;.thermal-hydraulic,oscillations would not occur'n allowable reactor operating regions.

Instability 'events at LaSalle and Vermont Yankee in 1988 have led to concerns relative to the capability of currently approved analytical methods to adequately predict when instabilities will occur.

These events as well as analysis performed by General Electric (GE) and NRC contractors indicate that instabilities may occur which result in regional oscillations and local power peaking greater than previously analyzed for in-phase core oscillations.

In

addition, calculations performed by GE indicated that under some operating conditions, the safety limit Minimum Critical Power Ratio (MCPR) may be violated during regional power oscillations.

The BWR Owners Group (BWROG) is currently working with GE and the NRC to develop a

long-term resolution to stability concerns.

In November

1988, Interim Corrective Actions (ICAs) to address stability concerns were issued by GE and subsequently adopted by the BWROG.

NRC Bulletin 88-07, Supplement 1 (December 30, 1988), requires all BWR licensees to implement the GE recommendations.

In addition, the bulletin requires some BWRs (BFN included) to initiate a manual

.scram following the., trip,of both recirculation pumps when the reactor is in the RUN mode.

In addition to procedural changes, BFN technical specifications must be modified to address stability concerns before restart of Units 1 and 3.

The BFN Unit 2 Technical Specifications have been modified to be consistent with NRC Bulletin 88-07, Supplement 1.

Justification for the Thermal-H draulic Technical S ecification Chan es The NRC Bulletin and the GE ICAs define regions of concern on the power/flow map referred to as Regions A, B, and C.

Region A includes operating conditions above the 100 percent rod line with core flow less than 40 percent of rated flow.

Region B includes operating conditions between 80 and 100 percent rod lines with core flow less than 40 percent of rated flow.

Region C includes operating conditions above the-80 percent rod line with core flow between 40 and 45 percent of rated flow.

The regions defined by GE cover the high power/low flow corner of the operating domain where stability.margins. are lowest.

,The GE recommended region boundaries are based on plant operating experience, special stability

tests,

'and analytical studies.

Region I of the proposed TS change corresponds to Region A as defined in the ICAs.

Most oscillations have occurred during testing and operation at or above the 100 percent rod line with core flow near natural circulation.

This behavior is consistent with analyses which predict reduced stability margin with increasing power or decreasing flow.

Region I bounds the ma)ority of events and tests where core oscillations have been observed in GE BWRs.

This region represents the least stable conditions on the power/flow map and is therefore considered an excluded region in which normal operation is not allowed.

Because operating experience has demonstrated that oscillations may rapidly develop in this region, operator actions are required to prevent the initiation of core oscillations in the event Region I is entered.

Region II of the proposed TS change includes both Regions B and C defined in the ICAs.

Region B of the ICAs is also considered to be an excluded region (i.e.,

no intentional entry) because of the relatively low core flow.

Even though the probability of core oscillations is lower in Region B than in Region A, several events and tests have demonstrated that oscillations can occur in this region for certain operating conditions.

However, because the power level is lower, the margin to fuel safety limits is greater in Region B.

Region C of the ICAs is

.defined as a buffer zone to the excluded regions.

Although no oscillations~have

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When operation does occur in Region C,

the operator should ensure that adequate surveillance of nuclear instrumentation is performed.

The proposed BFN Units 1 and 3 TS (and the existing Unit 2 TS) combine Region B and C into Region II and will conservatively apply Region B restrictions to Region C.

The potential for core thermal-hydraulic oscillations to occur when operating outside Regions I and II is very small and therefore special restrictions are not required outside these regions.

The region boundaries for the ZCAs: were developed based on plant operating and test experience and analysis of GE, fuel designs.

The regions were chosen to generically apply to all licensed GE fuel designs and operating domains (e.g.,

Extended Load Line, Single Loop Operating, etc.).

The cores of BFN Units 1 and 3

contain only GE fuel and therefore the regions defined in the interim corrective actions are appropriate for both BFN Units 1 and 3.

Proposed LCO 3.5.M.1 restricts normal operation within guidelines of the power-flow map.to conditions outside of Regions I and II.

The excluded regions represent the least" stable conditions for the plant.

Regions I and II are usually entered as a result of plant transients (e.g., recirculation pump trip) and are not part of the normal operating domain.

All events (including test experience) that have resulted in core oscillations have occurred in either Region I or II. Intentional operation is not allowed in these regions in order to minimize the probability of encountering core oscillations and potentially challenging fuel safety limits.

Proposed Action 3.5.M.2 requires the operator to manually scram the reactor if Region I is entered.

Because BFN does not have an unfiltered flow-biased neutron flux scram, automatic scram protection is not provided until APRM oscillations reach a peak magnitude'of 120 percent of rated.

Due to partial cancellation of out of phase LPRM signals during regional oscillations, local neutron flux can be significantly higher than indicated by the APRM signal.

Calculations by GE indicate that during operation in Region I, the safety limit MCPR (SLMCPR) may be violated in some situations when APRM oscillations are approximately 30 percent peak-to-peak (Reference 1).

Because. stability margins are lowest in Region I, the potential exists for oscillations to rapidly increase in magnitude once they initiate.

During transients which cause entry into Region I, the operator may not have sufficient time to manually insert control rods or increase core flow to suppress oscillations before they reach an unacceptable magnitude.

The prompt action of manually scramming the reactor if Region I is entered will ensure adequate protection of the SLMCPR.

Proposed Action 3.5.M.3 requires the operator to take immediate action to exit.

Region II if entered inadvertently.

Because core thermal-hydraulic stability is very sensitive to core power and flow, stability margins are greater in Region-II than in Region I.

The increased margin means that the probability of core oscillations is less and that large magnitude oscillations will not be reached as rapidly as in Region Z.

Also, due to the lower power and/or increased core flow in Region II, the margin to the SLMCPR will be larger than in Region I.

Due to increased stability and SLMCPR margins, the operator will have more time to suppress oscillations in Region II before the SLMCPR is violated.

Tests and operating experience have demonstrated that the insertion of control rods or the increase of core flow will rapidly dampen core thermal-hydraulic oscillations and move the plant into a region of increased stability margin.

At reactor conditions where.core oscillations initiate, the insertion of a

few control rod notches or a 1-2 percent increase in core flow will effectively suppress the oscillations (Reference 1).

When control rod insertion is used to exit the region, a predefined set of control rods will generally be used by the

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operator to ensure an expedient reduction in core thermal power. If one or more recirculation pumps. are*-operational,,

increasing core flow is an acceptable alternative to inserting control rods and is generally simpler to perform.

However, starting a recirculation pump to exit the region is not an appropriate action since it can lead to sudden reactivity insertions and initiate core oscillations.

Also, starting a recirculation pump can potentially distract the operators attention away from the detection of potential oscillations while in the region.

The actions described above will minimize the probability of core thermal-hydraulic oscillations following a transient which places the plant in Region II.

The presence of core thermal-hydraulic oscillations is an indication of the loss of control of the reactor, and if not mitigated might rapidly lead to conditions which violate the 'SLMCPR.

Experience has shown that oscillations can grow rapidly to high levels.

Even though most events have been terminated by operator actions, it cannot be demonstrated that a manual control rod insertion or a core flow increase will always be rapid enough to prevent exceeding the SLMCPR.

Therefore, if oscillations are detected, a rapid power reduction (i.e.,

scram) is the appropriate method of mitigation.

This will ensure that oscillations are rapidly suppressed.

To avoid unnecessary challenges to the reactor protection system, a scram should be initiated only when there is clear evidence of core oscillations.

Increases in the size of the observed "noise" in the APRM or LPRM signals not explained by changes in equipment status are evidence of a possible instability.

APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to 10 percent peak-to-peak APRM oscillations) are definitive evidence of core thermal-hydraulic instability; however, it is not intended that the operator should wait until oscillations reach this magnitude to initiate a scram.

The reactor should be manually scrammed as soon as a thermal-hydraulic instability is identified while the size of the oscillations is still small.

Periodic upscale and downscale LPRM alarms may be indicators of core thermal-hydraulic oscillations.

However, LPRM alarms alone should not be used to initiate a reactor scram because they only provide an indirect indication of

'scillations and may,be indicators of other conditions or equipment failures.

'If any LPRM alarms are

received, the APRMs and individual LPRMs should be immediately evaluated to confirm the presence of oscillations.

Based on GE analyses (Reference 1), the SLMCPR is not approached during regional oscillations until APRM oscillations are greater than approximately 30 percent peak>>to-peak.

In addition, periodic upscale and downscale LPRM alarms willoccur before regional oscillations are large enough to threaten the MCPR safety limit (Reference 1).

Therefore, initiating a

manual scram on evidence of core instability as described above is sufficient to ensure that the SLMCPR will not be violated.

Proposed SR 4.5.M is added to verify that the reactor is operating in the proper region (acceptable region) when reactor power is increased greater than 5 percent rated thermal power (RTP) with initial reactor core flow less than 45 percent or a decrease of 10 percent core flow while initial thermal power is greater than 40 percent of rated.

The addition of Figure 3.5.M-1, "BFN Power/Flow Stability Regions," provides the user of the technical specifications with a

clear illustration to allow determination of which combinations of reactor core power versus core flow are acceptable or unacceptable.

Based on this figure and the appropriate Action Statements in 3.5.M, an operator can"readily identify the steps required to exit an unacceptable region.

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"'"and'the basis 'for~these ~requirements.

The proposed change for LCO 3.6.F.3 will clarify that operation with both re'circulation pumps out-of-service is allowed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, but only if the reactor is not in the RUN mode.

The proposed change for SR 4.6.F.3 provides consistency with proposed LCO 3.6.F.3 and requires this SR to be performed during REACTOR POWER OPERATION before starting either recirculation pump.

New LCO 3.6.F.4 requires the operator to manually scram the reactor following a trip of both recirculation pumps while operating in the RUN mode.

The reactor will enter either Region I or II following a recirculation pump trip from above the 80 percent rod line.

Requiring a manual scram immediately following the loss of both recirculation pumps adds= conservatism to ensure that thermal-hydraulic instabilities do not occur.

The changes proposed to Bases 3.6.F/4.6.F are being made to reflect the changes proposed to the respective specifications.

Justification for Pro osed Chan es to HPCI RCIC Bases:

NUREG 0737, Item II.K'.3.13 requires the RCIC system to have automatic restart capability.

In a letter to the NRC dated August 9,

1989, TVA stated that this automatic restart feature is verified in system logic tests and did not at that time propose to add this specific test provision to the technical specifications.

Both the HPCI and RCIC systems have this automatic restart feature.

The proposed change to Bases 4.2 will help to ensure that the logic system functional tests retain this test feature.

This change also meets the intent, for BWR model technical specifications for NUREG-0737 TMI action plan requirements by including the test provisions in the bases of the technical specifications.

References 1.

NED0-31708, "Fuel Thermal Margin During Core Thermal Hydraulic Oscillations in a Boiling Water Reactor," General Electric, June 1989.

2.

Letter, D. N. Grace (BWROG) to A Thadani (NRC), "Sub)ect:

NRC Bulletin 88-07,'Supplement-1, Power Oscillations in Boiling Water Reactors,"

January 26, 1989.

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ENCLOSURE 3

PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION BROWNS FERRY NUCLEAR PLANT (BFN)

Descri tion of Pro osed Technical S ecification Chan es The proposed TS change applies to BFN Units 1, 2, and 3.

For Units 1 and 3, the proposed changes will add provisions for core thermal-hydraulic stability in 3.5.M/4.5.M, 3.6.F/4.6.F, Bases 3.5.M and Bases 3.6.F/4.6.F.

These changes will implement the requirements of NRC Bulletin 88-07, Supplement 1 by defining the reactor flow and power, regions which are acceptable or unacceptable for operation.

Unit 2 was granted similar technical specification changes for core thermal-hydraulic stability prior to restart.

For Units 1, 2, and 3 the proposed changes will update Bases 4.2 to clarify testing provisions for HPCZ and RCZC.

Basis For Pro osed No Si nificant Hazards Consideration Determination The 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 not 1) involve a significant increase in the probability or consequences of an accident previously evaluated, or 2) create the possibility of a new or different kind of accident from any accident previously evaluated, or 3) involve a significant reduction in a margin of safety.

1.

The proposed change does not involve a

significant increase in the probability or consequences of an accident previously evaluated.

Implementation of the proposed TS change decreases the probability of core thermal-hydraulic "oscillations by precluding operation in regions where instabilities may occur.

Zn addition, the proposed change will provide additional assurance that core oscillations that do occur will be suppressed prior to exceeding fuel integrity limits.

The proposed change does not have an adverse safety effect on any affected safety system nor are the assumptions of the safety analyses

.affected by restricting operation to outside.of Regions I and ZZ.

Therefore, the proposed change reduces the probability and consequences of potential core oscillations and does not increase the probability or consequences of any other previously analyzed event.

2.

The proposed change to the technical specifications does not create the possibility of a

new or different kind of accident from any accident previously evaluated.

Restricting operation to outside of Regions I and IZ does not create any new failure mechanisms.

Plant procedures will preclude normal operation in those regions.

Emergency entry into a'estricted region is permitted to protect plant safety equipment provided that the prescribed actions (i.e.,

scram or exit) for the region entered are performed.

Operator actions to exit Region ZI will be performed in compliance with plant procedures, fuel preconditioning restrictions, and technical specifications.

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

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The proposed change does not involve a significant reduction in a margin of safety.

The proposed changes are conservative in nature and provide increased assurance that the fuel safety limit MCPR will not be violated due to core oscillations.

These changes are consistent with NRC and GE guidelines.

The implementation of this technical specification will actually increase this margin of safety at BFN by not allowing the plant to operate in Regions I or II.

If one of these Regions is entered, specific operator actions are required which will place the plant in a

more conservative and safe condition than current BFN Units l and 3 Technical Specifications require.

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