ML18038B865

Proposed Tech Specs,Submitting Revised BFN TS Bases Section 3.5.N, References, Reflecting Updated LOCA Analyses for Units 2 & 3

ML18038B865
Person / Time
Site:	Browns Ferry
Issue date:	04/24/1997
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18038B864	List: ML18038B863 ML18038B865
References
NUDOCS 9704300055
Download: ML18038B865 (19)

Text

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

UNITS 2 AND 3 REVISION TO TECHNICAL SPECIFICATION (TS) BASES (TS-389)

MARKED PAGES FFE TED PA E LI T Unit 3.5/4.5-33 3.5/4.5-36 II. MARKED PA E (See Attached)

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3.5 . ~~ (Cont'd) ~ ~ NOv 02 <995 region is ao't ntctssary Hovevtr in order to minimize the probability of core instability folloving eatry into Region II, the operator vill take immediate action to exit the region. Although formal surveillances are not performed vhile exiting Region II (delayiag exit for surveillances is undesirable), an k~"

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reactor is restricted to thermal power and vn01tions ( i. e., outside Regions I and II) vhere tncrmal-hydraulic instabilities are very unlikely to occur.

3.5.N.

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

2- "BWR Transient Analysis Model Utilizing the RETRAH Program,"

TVA-TR81-01-A.

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

4,5 Coo The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operacion. For example, in the case of the HPCI, aucomatic initiation during pover operation vould result in pumping cold vacer into the reactor vessel vhich is not desirable. Complete ADS testing during pover operation causes an undesirable loss-of-coolant. inventory. To increase the availability of the core and containment cooling system, the components vhich make up the system, i.e., instrumentation, pumps, valves, etc., are test tested frequently. The pumps and mocor operated in)ection valves are also BFÃ 3.5/4.5-36 hMENOMENT NO. I9 9 Unit 3

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

UNITS 2 AND 3 REVISION TO TECHNICAL SPECIFICATION (TS) BASES (TS-389)

REVISED PAGES I. AFFECTED PA E LI T QnnJ- 2 3.5/4.5-32* 3.5/4.5-35*

3.5/4.5-33 3.5/4.5-36 3.5/4.5-34* 3.5/4.5-37*

3.5/4.5-35* 3.5/4.5-38*

• Denotes Overleaf Page (See Attached)

3.5 BASES (Cont'd)

The LHGR shall be checked daily during reactor operation at 25 percent power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For LHGR to be a limiting value below 25 percent of rated thermal power, the largest total peaking would have to be greater than approximately 9.7 which is precluded by a considerable margin when employing any permissible, control rod pattern.

3.5.K. um C 't'ca Powe at't core thermal power levels less than or equal to 25 percent, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at and thermal hydraulic this point, operating plant experience analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached orensures that MCPR will be known following a change in power power shape at a (regardless of magnitude) that could place operation thermal limit.

3. 5. L. et Operation is constrained to the LHGR limit of Specification 3.5.J. This limit is reached when core maximum fraction of limiting power density (CMFLPD) equals 1.0. For the case where CMFLPD exceeds the fraction of rated thermal power, operation is permitted only at less than 100-percentSpecification rated power and only with APRM scram settings as required by 3.5.L.1.

The scram trip setting and rod block trip setting are adjusted to ensure that no combination of CMFLPD and FRP the will increase the LHGR transient peak beyond that allowed by 1-percent plastic strain limit. 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. C a 'c tab'l'he minimum margin to the onset of thermal-hydraulic manually instability occurs in Regioninto I of Figure 3.5.M-1.is A

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

BFN 3.5/4.5-32 Unit 2

3 5 .~l~ (Cont ')

Because the probability of thermal-hydraulic oscillations is lower and the margin to the MCPR.safety limit is greater in Region scram II upon than in Region I of Figure 3.5.M-1, an immediate 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 is undesirable), an immediate manual scram will be initiated observed.

if evidence of thermal-hydraulic instability is

'urveillances 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.

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.

1. Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit, 2, NEDO - 24088-1 and Addenda.

2. "BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

3.'eneric- Reload Fuel Application, Licensing Topical Report, NEDE 24011-P-A and Addenda.

4. GE Document NEDC-32484P, Rev. 1, S. K. Rhow and C. T. Young, "Browns Ferry Nuclear Plant Units 1, 2, and 3 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," February 1996.

5. GE Document GE-NE-B13-01755-2, Rev. 1, S. K. Rhow and T. H. Chuang, "Relaxation of Emergency Core Cooling System Parameters for Browns Ferry Nuclear Plant Units 1, 2, and 3 (Perform Program Phase I)," February 1996.

BFN 3.5/4.5-33 Unit 2

4,5 e 'es The testing interval for the core and containment cooling

'systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable. Complete ADS testing during power operati'on causes an undesirable loss-of-coolant inventory. To increase the availability of the core and containment cooling system, the components which make up the system, i.e., instrumentation, pumps, valves, etc., are tested frequently. The pumps and motor operated injection valves are also tested in accordance with Specification 1.0.MM to assure their OPERABILITY. A simulated automatic actuation test once each cycle combined with testing of the pumps and injection valves in accordance with Specification 1.0.MM is deemed to be adequate testing of these systems. Monthly alignment checks of valves that are not locked or sealed in position which affect the ability of the systems to perform their intended safety function are also verified to be in the proper position. Valves which automatically reposition themselves on an initiation signal are permitted to be in a position other than normal to facilitate other operational modes of the system.

When components and subsystems are out-of-service, overall core 'and containment cooling reliability is maintained by OPERABILITY of the remaining redundant equipment.

Whenever a CSCS system or loop is made inoperable, the other CSCS systems or loops that are required to be OPERABLE shall be considered OPERABLE if they are within the required surveillance testing frequency and there is no reason to suspect they are inoperable. If the function, system, or loop u'nder test or calibration is found inoperable or exceeds the trip level setting, the LCO and shall the required surveillance testing for the system or loop apply.

Ave a e R GR and MCP The APLHGR, LHGR, and MCPR shall be checked daily to determine if fuel burnup, or control power distribution. Since rod movement has caused changes in changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate.

BFN 3.5/4.5-34 Unit 2

THIS PAGE INTENTIONALLY LEFT BLANK 3.5/4.5-35 AMBDMEg gP pep BFN Unit 2

3.5 (Cont'd)

The LHGR shall be checked daily during reactor operation at

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> 25 percent power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For LHGR to be a limiting value below 25 percent of rated thermal power, the largest total peaking would have to be greater than approximately 9.7 which is precluded by a considerable margin when employing any permissible control rod pattern.

3.5.K. 'mum C it'ca Powe at't core thermal power levels less than or equal to 25 percent, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative'o MCPR. The daily requirement for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape a.

(regardless of magnitude) that could place operation at thermal limit.

3.5.L. AP Set pints Operation is constrained to the LHGR limit of Specification 3.5.Z. This limit is reached when core maximum fraction of limiting power density (CMFLPD) equals 1.0. For the case where CMFLPD exceeds the fraction of rated thermal power, operation is permitted only at less than 100-percent rated power and only with APRM scram settings as required by Specification 3.5.L.1.

The scram trip setting and rod block trip setting are adjusted to ensure that no combination of CMFLPD and FRP will increase the LHGR transient peak beyond that allowed by the one-percent plastic strain limit. 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. Co e e d au 'c Stabil't 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.

BFN 3.5/4.5-35 Unit 3

3.5 BASES (Cont'd)

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 observed.

if evidence of thermal-hydraulic instability is 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.

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. 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.

4. GE Document NEDC-32484P, Rev. 1, S. K. Rhow and C. T. Young, "Browns Ferry Nuclear Plant Units 1, 2, and 3 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," February 1996.

5. GE Document GE-NE-B13-01755-2, Rev. 2, S. K. Rhow and T. H. Chuang, "Relaxation of Emergency Core Cooling System Parameters for Browns Ferry Nuclear Plant Units 1, 2, and 3 (Perform Program Phase I)," December 1996.

BFN 3.5/4.5-36 Unit 3

V 4i5 a e t C S te u e'a ce e encies

. The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory. To increase the availability of the core and containment cooling system, the components which make up the system, i.e., instrumentation, pumps, valves, etc., are tested frequently. The pumps and motor operated injection valves are also tested in accordance with Specification 1.0.MM to assure their OPERABILITY. A simulated automatic actuation test once each cycle combined with testing of the pumps and injection valves in accordance with Specification 1.0.MM is deemed to be adequate testing of these systems. Monthly alignment checks of valves that are not locked or sealed in position which affect the ability of the systems to perform their intended safety function are also verified to be in the proper position. Valves which automatically reposition themselves on an initiation signal are permitted to be in a position other than normal to facilitate other operational modes of the system.

When components and subsystems are out-of-service, overall core and containment cooling reliability is maintained by OPERABILITY of the remaining redundant equipment.

Whenever a CSCS system or loop is made inoperable, the other CSCS systems or loops that are required to be OPERABLE shall be considered, OPERABLE if they are within the required surveillance testing frequency and there is no reason to suspect they are inoperable. If the function, system, or loop under test or calibration is found inoperable or exceeds the trip level setting, the orLCOloop and the required surveillance testing for the system shall apply.

ve a e G LHG and CP The APLHGR, LHGR, and MCPR shall be checked daily to determine if fueldistribution.

power burnup, or control rod movement has caused changes in Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate.

BFN 3.5/4.5-37 Unit 3

THIS PAGE INTENTIONALLY LEFT BLANK AMENDMENT NO. y g 9 BFN 3.5/4.5-38 Unit 3

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