ML20101N614

Amend 202 to License DPR-52.Amend Revises TS 3.2.C,3.5.K & 4.5.K to Allow Continued Operation When RBM Is Inoperable & MCPR within Specified Limits

ML20101N614
Person / Time
Site:	Browns Ferry
Issue date:	07/02/1992
From:	Hebdon F Office of Nuclear Reactor Regulation
To:
Shared Package
ML20101N617	List: ML20101N614 ML20101N626
References
NUDOCS 9207100077
Download: ML20101N614 (11)

Text

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'n UNITED STATES

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i NUCLEAR REGULATORY COMMISSION o

WASHINGTON. D C. 2J%$

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT. UNIT 2 AffjJ{@iENT TO FACILITY OPERATING LICENSE Amendment No. 202 License No. DPR-52 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated May 13, 1992, complies with the standards and requiremer.ts of the Atomic -Energy Act of 1954,- as amended (the.

Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical. to the common deferte and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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. 2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-52 is hereby amended to read as follows:

(2) Technical Specifications lhe Technical Specifications contained in Appendices A and B, as revised through Amendment No. 202, are hereby incorporated in the license

.he licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

151s amendment is temporary and expires at the end of the current fuel cycle (Cycle 6).

FOR.THE NUCLEAR REGULATORY COMMISSION s..rbxek r

frederick J. Hebdon, Director Project Directorate 11-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulatif,n

Attachment:

Changes to the Technical Specifications Date of Issuance: July 2,1992

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4 ATTACHMENT TO LICENSE AMENDMENT NO. 202 FAClllTY OPERATING LICENSE NO. OPR-52 DOCKET NO. 50-260 Revise the Appendix A Technical Specifications by removing the pages identi-fied below and inserting the enclosed pages.

The revised pages are identified by the ca)tionad amendment number and contain marginal lines indicating the area of c1ang.

REMOVE INSERT 3.2/4.2-26 3.2/4.2-26 3.2/4.2-27 3.2/4.2-27 3.2/4.2-27a*

3.2/4.2-27b*

3.2/4.2-67 3.2/4.2-67**

3.2.4.2-68 3.2/4.2-68 3.5/4.5-18 3.5/4.5-186*

3.5/4.5-19 3.5/4.5-19

• Spillover pages

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. NOTES FOR TABLE 3,1 d

1. The minimum number of OPERABLE channels for each trip function is detailed for the STARTUP and RUN positions of the reactor mode selector switch.

The SRM, IRM, and APRM (STARTUP mode), blocks need not be OPERABLE in "RUN" mode, and the APRM (flow biased) tod blocks need not 4

be OPERABLE in "STARTUP" mode.

With the number of OPERABLE channels less than required by the minimum OPERABLE channels per trip function requirement, place at least one l

inoperable channel in the tripped condition within one bour.

2. W is the recirculation loop flow in percent of design. Trip level setting is in percent of rated power (3293 MWt).

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3. IRM downscale is bypassed when it is on its lowest range.

4. SRMs A and C downscale functions are byp. sed when IRMs A, C, E, and G are above range 2.

SRMs B and D downscale function is bypassed when IRMs B, D, F, and !! are above range 2.

1 SRM detector not in startup position is bypassed when the count rate is i

1100 CPS or the above condition is satisfied.

5. During repair or calibration of equiptent, not more tban_ one SRM or RBM channel nor more than two APRM or IRM channels may b( bypassed.

Bypassed channels are not counted as OPERABLE channels to meet the minimum OPERABLE channel requirements. Refer to section 3.10.B for SRM requirements during core alterations.

6. IRM channels A, E, C, G all in range 8 or above bypasses SkM channels A and C functions.

IRM channels B, F, D, H all in range 8 or above bypasses SRM channels B and D functions.

• 7. The following operational restraints apply to the RBM only.

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Both RBM char.nels are bypassed when reattor power is 130 a.

percent or when a peripheral (edge) control rod is selected.

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b.

The RBM need not be OPERABLE in the "startup" position of the reactor mode selector switch.

c.

Two RBM channels are provided and only'ono of these may be bypassed with the console selector. The other channel may also be defeated only if the conditionc of "e" or "f" are met.

If the inoperable channel cannot be restored within 24 i

hours, and the conditions of "e" or "f" are not met, the

. inoperable channel shall be placed in the tripped condition within one hour,

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• The provisions of Note 7e and 7f are apc scable during unit i cycle 6 only.

i BFN 3.2-4.2-26 Amendment 202

- Unit 2 5

NOTES FOR TABLE 3.2.C (Cont'd) 7.

(Continued) d.

With both RBM channels inoperable, and the conditions of "e" or "f" not met, place at least one inoperable rod block monitor channel in the tripped condition within one hour.

• e.

The RBM need not be OPERABLE when reactor power is 190 percent and MCPR is 11.40.

• f.

The RBM need not be OPERABLE when reactor power is (90 percent and MCPR is 11.70.

8.

This function is bypassed when the mode switch is placed in RUN.

9.

This function is only active when the mode switch is in RUN. This function is automatically bypassed when the IRM instrumentation is OPERABLE and not high.

10. The inoperative trips are produced by the following functions:

a.

SRM and IRM (1) Local " operate-calibrate" switch not in operate.

(2) Power supply voltage low.

(3) Circuit boards not in circuit.

b.

APRM i

(1) Local " operate-calibrate" switch not in operate.

l (2) Less than 14 LPRM inputs.

(3) Circuit boards not in circuit.

c.

RBM (1) Local " operate-calibrate" switch not in operate.

(2) Circuit boards not in circuit.

(3) kBM fails to null.

(4) Less than required number of LPRM inputs for rod selected.

11. Detector trsverse is adjusted to 114 1 2 inches, placing the detector 1 ewer position 24 inches below the lower core plate.

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• The provisions of Note 7e and 7f are applicable during unit 2-1 l

cycle 6 only.

l BTN 3.2/4.2-27 Amendment 202 Unit 2

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NOTES FOR TABLE 3.2.C (Cont'd)

12. This function may be bypassed in the SHUTDOWN or RETUEL mode.

If this j

function is inoperable at a time when OPERABILITY is required the channel shall be tripped or administrative controls aball be inusedir.tely imposed to prevent control rod withdrawal.

13. RBM upscale flow-biased setpoint clipped at 106 percent rated reactor power.

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4 BTN 3.2/4.2-27a Amendment 202 Unit 2 o.

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l BFN 3.2/4.2-27b Amendment 202 Unit 2

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+3.2 BASES (Cont'd) flow instrumentation is a backup to the temperature instrumentation.

In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above'200*F. The temperature increases ca.n cause an unnecessary main steam line isolation and reactor scram.

Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

1 High radiation monitors in the main steam line tunnel have been provided j

to detect gross fuel failure as in the control rod drop accident. With the established nominal setting of three times normal background and main i

steam line isolation valve closure, fission product release is limited so that 10 CTR 100 guidelines are not exceeded for this accident. Reference l

Section 14.6.2 FSAR. An alarm with a nominal setpoint of 1.5 x normal full-power background la provided also.

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steat. line pressure drops below 825 psig.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. -Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1-out-of-2 loric, and all sensors are required to be OPERABLE.

High temperature in ths vicinity of the HPCI equipment is sensed by four sets of four binetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system. Each trip system consists of two elements. Each channel contains one temperature switch located in the pump room and three temperature switches located in the torus area. The RCIC high flow and high area taperature sensing instrument channels are arranged in the same manner as the HPCI system.

The HPCI high steam flow trip setting of 90 paid and the RCIC high steam flow trip setting of 450" H O have been selected such that the trip 2

setting is high enough to prevent spurious tripping during pump startup but low enough to prevent core uncovery and maintain fission product l

releases within 10 CFR 100 limits.

l The HPCI and RCIC steam line space temperature switch trip settings are high enough to prevent spurious isolation due to normal temperature excursions in the vicinity of the steam supply piping. Additionally, these trip settings ensure that the primary containment isolation steam supply valves isolate a break within an acceptabla time period to prevent core uncovery and maintain fission product releases within 10 CTR 100 limits.

High temperature at the Reactor Water Cleanup (RWCU) System in the main steam valve vault, RWCU pump room 2A,.RWCU pump room 28, RWCU heat exchanger room or in the space near the pipe trench containing RWCU piping could indicate a break in the cleanup system. When high temperature i

occurs, *he cleanup sys~ tem is isolated.

BFN 3.2/4.2-67 AMENDMENT NO. I 8 9 Unit 2

r, 3.2 BAlli (Cont'd)

The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.

An exception to this is when logic functional testing is being performed.

She control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 1.07.

The trip logic for this function is 1-out-of-n e.g., any trip on one of six APRMs, eight IRMs, or four SRMs vill result in a rod block.

A General Electric study, GE-NE-770-06-03?2 shows for the unit 2 cycle 6 core that if the initial MCPR is as specified in item 7e or 7f of Table 3.2.C, then no single rod withdrawal error can cause the MCPR to decretse below the MCPR safety limit. When core operating conditions have been verified to be within the limits of items 7e or 7f of Table 3.2.C, the RBM is not required. When the RBM is required, the minimum instrument channel requirements apply. These requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements fcr the RBM may be reduced by one for maintenance, testing, or calibration. This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence. for withdrawal of control rods.

The APRM rod block function is flow biased and prevents a significant reduction in CCPR, especially during operation at reduced flow. The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so tha, MCPR is maintained greater than 1.07.

The RBM rod block function provides local protection of the core; i.e.,

the prevention of critical power in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.

If the IRM channels are in the worst condition of allowed bypass, the sealing arrangement is such that for unbypass7d IRM channels, a rod block signal is generated before the detected neutrons flux has increased by l

more than a factor of 10.

A downscale indication is an indication the instrument has failed or the instrument is not sensitive enough.

In either case the instrument vill nat respond to changes in control rod motion and chus, control rod motion is prevented.

The refueling interlocks also og o se one logic channel, and are required for safety only when the mode sviten is in the refueling position.

For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapid enough to allow cither core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate.

The arrangement of the tripping contacts is such as to pravide this function when necessary and minimize spurious operation. The trip settings given in the specification are BFN 3/2/4.2-68 Amendment 202 Unit 2 1

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J.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 I Average Planar Linear Heat 4.5.I Maximum Average Planar Generation Rats Linear Heat Generation Rate (MAPLHGR)

During steady-state power operation, The MAPLHGR for each type of the Maximum Average Planar Linear fuel as a function of average Heat Generation Rate (MAPLHGR) for planar exposure shall be each type of fuel as a function of determined daily during average planar exposure shall not reactor operation at 1 25%

exceed the limiting value shown in rated thermal power.

Tables 3.5.I-1, 2, 3, and 4.

If at any time during operation it is determined by normal surveillance that the limiting d

value for MAPLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the MAPLHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the COLD SHUTDOWN CONDITION within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

J.

Linaar_ Heat Generation Rate (LHGR)

J.

Linear Heat Generation Rate (LHGR)

During steady-state power operation, The LHGR shall be checked the linear heat generation rate (LHGR) daily during reactor fuel of any rod in any fuel assembly at operation at 1 25% rated any axial location shall not exceed thermal power.

13.4 kW/ft.

If at any time during operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the LHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the COLD SHUTDOWN CONDITION within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and corresponding action shall continue w 311 reactor operation is within the prescribed limits.

i BTN 3.5/4.5-18 AMENDMENT NO. I 72 Unit 2 i

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3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS 1,IMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.5.K Minimum Critical Power Ratio 4,5.K Minimum Critical Power (MCPR)

Ratio (MCPR)

Except when the provisions of Note 1.

MCPR shall be determined daily 7 of Table 3.2.0 are being employed during reactor power operation due to the inoperability of the Rod at 2. 25% rated thermal power Block Monitor, the minimum critical and following any change in power ratio (MCPR) as a function of power level or distribution scram time and core flow, shall be that would cause operation equal to or greater than shown in with a li: siting control rod Figure 3.5.K-1 multiplied by the pattera as described in the Kg shown in Figure 3.5.2, wheret bases for Specification LJ.

Y = 0 or Yave -

, whichever is 2.

Except as provided by Note 7 TA - T B greater of Table 3.2.C, the MCPR limit shall be determined for each YA=0.90sec(Specification 3.3.C.1 fuel type SX8, 8X8R, P8X8R, scram time limit to 20% insertion from Fiture 3.5.K-1, from fully withdrawn) respectively, using

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N lB = 0.710+1.65 2 (0.053) (Ref.2) n n

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Yave= i= 1 T=0.0priortoinitial a.

n l

scram time measurements l

for the cycle, performed n=

number of surveillance rod in accordance with tests performed to date in Specification 4.3.C.1.

cycle (including BOC test).

b.

YasdefinedinSpecifi-Yi =

Scram time to 20% insertion from cation 3.5.K following the th fully withdrawn of the i rod.

conclusion of each scram-time surveillance test re-N=

total number of active rods quired by Specifications measured in Specification 4.3.C.1 and 4.3.C.2.

4.3.C.1 at BOC.

The determination of the If at my time during steady-state limit must be completed l

operation it is determined by normal within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of each surveillance that the limiting scram-time surveillance l

value for MCPR is being exceeded, required by Specification l

action shall be initiated within 4.3.C.

l 15 minutes to restore operation to l

within the prescribed limits.

If the steady-state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

BFN 3.5/4.5-19 Amendment 202 Unit 2