Text

Amends 133,129 & 104 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Changing Tech Specs Re Rod Worth Minimizer & Rod Sequence Control Sys Requirements
	ML20214H851
Person / Time
Site:	Browns Ferry
Issue date:	05/13/1987
From:	Zwolinski J; NRC OFFICE OF SPECIAL PROJECTS
To:	;
Shared Package
ML20214H831	List: ML20214H830; ML20214H851; ML20214H909;
References
	TAC-61960, TAC-61961, TAC-61962, NUDOCS 8705270523
	Download: ML20214H851 (63)
	v • d • e

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ga merg UNITED STATES i*

g fj kg NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20066 y E

\....,/

TENNESSEE VALLEY AUTHORITY 4

DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 133

' License No. DPR-33

1. The Nuclear Regulatory Comission (the Commission) has found that:

i A .- The application for amendment by Tennessee Valley Authority (the i

licensee) dated June 4,1986, complies with the standards and l

requirements of the Atomic Energy Act of 1954, as amended (the Act),

' and the Comission'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 .

Comission; l ,

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 Comission's regulations; D. The issuance of this amendment will not be inimical to the comon defense 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 Comission's regulations and all applicable requirements have been satisfied.

i 2. Accordingly, the license is amended by c'hanges to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-33 is hereby amended to-read as follows:

4 8705270523 870513 hDR ADOCK 05000259 PDR i

.' l (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.133, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance and shall be implemented within 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY C MMISSION

(-

John Zwolinski, Assistant Director for rojects Division of TVA Projects Office of Special Projects

Attachment:

Changes to the Technical Specifications Cate of Issuance: May 13, 1987

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. l ATTACHMENT TO LICENSE AMENDMENT NO.133 FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the areas of changes. Overleaf pages are provided to maintain document completeness.*

REMOVE INSERT i i ii ii 3.3/4.3-5 3.3/4.3-5*

3.3/4.3-6 3.3/4.3-6 3.3/4.3-7 3.3/4.3-7 3.3/4.3-8 3.3/4.3-8 3.3/4.3-9 3.3/4.3-9 3.3/4.3-10 3.3/4.3-10**

3.3/4.3-11 3.3/4.3-11**

Bases 3.3/4.3-12 3.3/4.3-12**

Bases 3.3/4.3-13 Bases 3.3/4.3-13**

Bases 3.3/4.3-14 Bases 3.3/4.3-14**

Bases 3.3/4.3-15 Bases 3.3/4.3-15**

Bases 3.3/4.3-16 Bases 3.3/4.3-16**

Bases 3.3/4.3-17 Bases 3.3/4.3-17**

Bases 3.3/4.3-18 Bases 3.3/4.3-18**

Bases 3.3/4.3-19 Bases 3.3/4.3-19**

-- Bases 3.3/4.3-20**

- Pagination change only

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TABLE OF CONTENTS Section Paae No.

1.0 Definitions. . . . . . . . . . . . . . . . . . .. 1.0-1 SAFETY LINITS AND LIMITING SAFETY SYSTEN SETTINGS 1.1/2.1 Fuel Cladding Integrity. . . . . .. . . . .. .. 1.1/2.1-1 1.2/2.2 Reactor Coolant System Integrity . . . . . . . . . 1.2/2.2-1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRENENTS 3.1/4.1 Reactor Protection System . . . ... ....... 3.1/4.1-1 3.2/4.2 Protective Instrumentation. . . . . . . . . . . . . 3.2/4.2-1 A. Primary Containment and Reactor Building Isolation Functions. . . ....... ... 3.2/4.2-1 B. Core and Containment Cooling Systems -

Initiation and Control . . .-. . . . . . . . 3.2/4.2-2 I

C. Control Rod Block Actuation. . . . . . . . . . 3.2/4.2-2 D. Radioactive Liquid Effluent Monitoring Instrumentation. . . . . . . . . . . . . . . 3.2/4.2-3 E. Drywell Leak Detection . . . . . . . . . . . . 3.2/4.2-4

, F. Surveillance Instrumentation . . . . . . . . . 3.2/4.2-4 G. Control Room Isolation . . . . . . . . . . . . 3.2/4.2-4 H. Flood Protection . . . . . . . . . . . . . . . 3.2/4.2-4 I. Noteorological Nonitoring Instrumentation. . . 3.2/4.2-4 J. Seismic Monitoring Instrumentation . . . . . . 3.2/4.2-5 K. Radioactive Liquid Effluent Monitoring Instrumentation ... . .. . ... . .. . 3.2/4.2-6 3.3/4.3 Reactivity Control . . . . . . . .'. . . . . . . . 3.3/4.3-1 A. Reactivity Limitations . . .'. . . . . . . . . 3.3/4.3-1 i

I B. Control Rods . . . . . . . .. . .. . . .. . 3.3/4.3-5 i l

C. Scram Insertion Times. . . . . . . . . . . . . 3.3/4.3-10 i

BFN-Unit 1 Amendment No. 133 i

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Section Pane No. .

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D. Reactivity Anomalies'. . . . . . . . . . . . . 3.3/4.3-11 l E. Reactivity Control . . . . . . . . . . . . . . 3.3/4.3-12 F. Scram Discharge Volume . . . . ........ 3.3/4.3-12 3.4/4.4 Standby Liquid Control System. . . . . . . . . . . 3.4/4.4~-1 A. Normal System Availability . . . . . . . . . . 3.4/4.4-1

8. Operation with Inoperable Components . . . . . 3.4/4.4-2 C. Sodium Pentaborate Solution. . . . . . . . . . 3.4/4.4-3 3.5/4.5 Core and Containment Cooling Systems . . . . . . . 3.5/4.5-1 A. Core Spray System (CSS). ........... 3.5/4.5-1

8. Residual Heat Removal System (RNRS)

(LPCI and Containment Cooling) ....... 3.5/4.5-4 C. RHR Service Water System and Emergency Equipment Cooling Water System (EECWS) ... 3.5/4.5-10 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). ... 3.5/4.5-15 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 K. Minimum Critical Power Ratio (MCPR). ..... 3.5/4.5-19 L. APRN Setpoints . . . . ............ 3.5/4.5-20 '

N. Reporting Requirements . . ..'........ 3.5/4.5-20 N. ' References . . . . . . ............ 3.5/4.5-34 3.6/4.6 Primary System Boundary. . . . . . . ....... 3.6/4.6-l' A. Thermal and Pressurization Limitations . ... 3.6/4.6-1

8. Coolant Chemistry. . . ............ 3.6/4.6-5 8FN-Unit 1 11 Amendment No. 133 i

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3.3/4.3 REACTIVITY CONTROL p:

I.lMITING CONDITIONS FOH OPERATION SURVEILLANCE REQUINEMENTS 3.3.B. Control Rods 4.3.B. Control Rods

1. Each control rod shall be I

1. The coupling integrity coupled to its drive or shall be verified for completely inserted and the each withdrawn' control control rod directional rod as follows:

control valves disarmed electrically. This a. Verify that the requirement does not apply control rod is in the refuel condition following the drive when the reactor is vented, by observing a Two control rod drives may response in the be removed as long as nuclear instru-Specification 3.3.A.1 mentation each time is met. a rod is moved when the reactor is operating above the preset power level of the RSCS.

b. When the rod is fully withdrawn the first time t after each refueling outage "7 or after

' maintenance, i observe that the drive does not go j to the overtravel ,

position.

2. The control rod drive 2. The control rod drive !

housing support system shall housing support system f be in place during reactor shall be inspected power operation.or when the after reassembly and ;

reactor coolant system is and the results of the i pressurized above atmospheric inspection recorded. !

pressure with fuel in the reactor vessel, unless all control rods are fully l inserted and specification 3.3.A.1 is met.

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Bpw 3.3/4.3-5 Unit 1

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3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS .

3.3.8. Control Rods 4.3.8. Control Rods 3.a Whenever the reactor is in 3.a.1 The Rod Sequence Control j the startup or run modes System (RSCS) shall be 1 below 20% rated power, the demonstrated to be OPERABLE f Rod Sequence Control System for a reactor startup~by )

(RSCS) shall be OPERABLE, the following checks:

except that the RSCS constraints may be a. Performance of the suspended by means of the comparator check of individual rod bypass group notch circuits switches for within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior 1 - special criticality to control rod tests, or withdrawal for the purpose of making 2 - control rod scram the reactor critical i timing per 4.3.C.1.

When RSCS is bypassed on b. Selecting and

! individual rods for these attempting to exceptions, RWN must be withdraw an out-of-operable per 3.3.B.3.b and sequence control rod v l

a second party verification after withdrawal of the first insequence !

l may not be used in lieu of l RWN. control rod. :

e. Attempting to withdraw a control rod more than one notch prior to other control rod movement after the group notch mode is automatically initiated.

3.a.2 The Rod Sequence Control System (RCS) shall be demonstrated to be OPERABLE for a reactor shutdown by the following checks:

a. Performance of the comparator check of the group notch circuits within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to automatic initiation of the group notch mode.

BFN 3.3/4.3-6 Unit 1 Amendment No. 133

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3.3/4.3 8 'TIVITY CONTROL LINITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRENENTS l Control Rods 4.3.B. Control Rods 3.3.B. I l i

3.a.2 (Cont'd)

b. Attempting to insert l a control rod more I

than one notch prior l

' to other control rod j

' ! movement after the j

' group notch mode is ;

' automatically l initiated. }

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c. Selecting and ,

attempting to move an ;

out-of-sequence l control rod after !

insertion of the i first insequence l control rod after j reaching a black and [

white rod pattern. j l

3.b. Whenever the reactor is 3.b.1 The Rod Worth Ninimizer !

in the startup or run modes (RWN) shall be demonstrated ,

below 20% rated power, the to be OPERABLE for a i

Rod Worth Minimizer (RWM) reactor startup by the !

j shall be OPERABLE. With l following checks: ;

the RWN INOPERABLE, verify ,

control rod movement .

s. By demonstrating that i and compliance with the

' the control rod {

prescribed control rod patterns and sequence l pattern by a second ,

input to the RWN :

! licensed operator or ;

computer are correctly ;

other technically loaded following any j l

qualified member of the loading of the program !

plant staff who is into the computer. !

present at the reactor control console. ,

b. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior Otherwise, control rod I to withdrawal of I movement may be only by I control rods for the i actuating the manual purpose of making the {

li reactor critical scram or placing the '

i reactor mode switch in - verify proper- i the shutdown position. annunciation of the j selection error of at !

I least one out-of- {

! sequence control rod.

BFN 3.3/4.3-7 l Amendment No. 133 j Unit 1 ,

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1 3.3/4.3 REACTIVITY CONTROL i LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS .

3.3.8. Control Rods 4.3.8. Control Rods 3.b.1 (Cont'd)

c. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to withdrawal of ,

control rods for the purpose of making the reactor critical, the rod block function of the RWN shall be verified by moving an

-ac- out-of-sequence control rod.

3.b.2 The Rod Worth Minimizer W (RWM) shall be demonstrated to be OPER&BLE for a reactor shutdown by the following checks:

a. By demonstrating that the control rod patterns and sequence input to the RWN computer are correctly loaded following any loading of the program into the computer.

b. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to RWN automatic initiation when reducing thermal power, verify proper annunciation of the -

selection error of at least one out-of-sequence control rod.

c. Within one hour after RWN automatic initiation when reducing thermal power, the rod block function of the RWN shall be verified by moving an out-of-sequence control rod.

BFN 3.3/4.3-8 Unit 1 Amendment No. 133

3.3/4 3 REACTIVITY CONTROL

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS Control Rods 4.3.B. Control Rods 3.3.B.

3.c. If Specifications 3.3.B.3.a 3.b.3 When the RWN is not through .b cannot be met the OPERABLE a second reactor shall not be started, licensed operator or other technically or if the reactor is in the qualified member of run or startup modes at less than 20% rated power, the plant staff shall control rod movement may be verify that the correct only by actuating the rod program is followed manual scram or placing except as specified in the reactor mode switch in 3.3.B.3.a.

the shutdown position.

4 Control rods shall not be 4. Prior to control rod withdrawn for startup or withdrawal for startup refueling unless at least or during refueling, two source range channels verify that at least two have an observed count rate source range channels equal to or greater than have an observed count three counts per second. rate of at least three counts per second. 1 a

5. During operation with 5. When a limiting limiting control rod control rod pattern patterns, as determined by exists, an instrument i the designated qualified functional test of the j personnel, either: RBN shall be performed prior to withdrawal of

a. Both RBN channels shall the designated rod (s) be operable: and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter. l or

b. Control rod withdrawal shall be blocked.

l BFN 3.3/4.3-9 Amendment No. 133 Unit 1

3.3/4.3 REACTIVITY CONTROL ,

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.C. Scram Insertion Times 4.3.C. Scram Insertion Times

1. The average scram 1. After each refueling insertion time, based on outage, all operable rods the deenergization of the shall be scram-time .

scram pilot valve sole- tested from the fully noids as time zero, of withdrawn position with all operable control rods the nuclear system in the reactor power pressure above 800 psis.

operation condition shall This testing shall be be no greater than: completed prior to exceeding 40% power.

Below 20% power, only rods in those sequences (A12 and A34 or B12 and

% Inserted From Ava. Scram Inser- 834) which were fully Fully Withdrawn tion Times (sec) withdrawn in the region from 100% rod density to 5 0.375 50% rod density shall be 20 0.90 scram-time tested. The 50 2.0 sequence restraints 90 3.500 imposed upon the control rods in the 100-50 percent

rod density groups to the preset power level may be removed by use of the individual bypass switches associated with those control rods which are fully or partially withdrawn and are not within the 100-50 percent

, rod density groups. In order to bypass a rod, the actual rod axial position must be known; and the rod must be in the correct in-sequence position. As required by 3.3.8.3.a., a second licensed operator may not be used in lieu of RWM for this testing.

l BFM 3.3/4.3-10

, Unit 1 Amendment No.133

. 3.3/4.3 REACTIVITY CONTROL SURVEILLANCE REQUIRENgNTS LIMITING CONDITIONS FOR OPERATION Scram Insertion Times 4.3.C. Scram Insertion Times 3.3.C.

2. At 16-week intervals, 10%

2. The average of the scram inser-tion times for the three fastest of the operable control operable control rods of all rod drives shall be scram-timed above 800 psig.

groups of four control rods in a two-by-two array shall be no Whenever such scram time 4

measurements are made, an greater than:

' evaluation shall be made Avg. Scram Inser- to provide reasonable

% Inserted From assurance that proper Fully Withdrawn tion Times (see) control rod drive 5 0.398 performance is being 20 0.954 maintained.

50 2.120 90 3.800

a. The maximum scram insertion time for 90% insertion of any operable control rod shall not exceed 7.00 seconds.

Reactivity Anomalies D. Reactivity Anomalies D.

The reactivity equivalent of During the startup test the difference between the program and startup following actual critical rod refueling outages, the critical rod configurations

- configuration and the expected configuration during power will be compared to the operation shall not exceed 1% Ak. expected configurations at If this limit is exceeded, the selected operating conditions, reactor will be shut down These comparisons will be until the cause has been used as base data for determined and corrective reactivity monitoring during actions have been taken as subsequent power operation appropriate. throughout the fuel cycle.

At specific power operating conditions, the critical rod configuration will be campared to the configuration espected based upon.

appropriately corrected past

- data. This comparison will

' be made at least every full power month.

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gry 3.3/4.3-11 Amendment-No. 133 Unit 1 l

3.3/4.3 REACTIVITY CONTROL .

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.E. If Specifications 3.3.C and .D 4.3.E. Surveillance requirements are above cannot be met, an orderly as specified in 4.3.C and .D shutdown shall be initiated and above, the reactor shall be in the shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

F. Scram Discharge 7olume (SDV) F. Scram Discharge Volume (SD7)

1. The ceram discharge volume 1.a. The scram discharge drain and vent valves shall volume drain and vent be operable any time that valves shall be verified the reactor protection open prior to each system is required to be startup and monthly operable except as thereafter. The valves specified in 3.3.F.2. may be closed

~ intermittently for testing not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in any 24-hour period ducir.g operation.

4 1.b. The scram discharge volume drain and vent valves shall be demonstrated operable monthly.

2. In the event any SDV drain 2. When it is determined or vent valve becomes that any SDV drain or inoperable, reactor vent valve is inoperable, operation may continue the redundant drain or provided the redundant vent valve shall be drain or vent valve is demonstrated operable operable. immediately and weekly thereafter.

3. If redundant drain or vent 3. No additional

, valves become inoperable, surveillance required, the reactor shall be in hot standby within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 BFN 3.3/4.3-12 Unit 1 Amendment No. 133

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3.3/4.3 8_M_gS, A. Reactivity Limitation

1. The requirements for.the control rod drive system have been identified by evaluating the need for reactivity control via c'ntrol o rod movement over the full spectrum of plant conditions and events. As discussed in subsection 3.4 of the Final Safety Analysis Report, the control rod system design is intended to provide sufficient control of core reactivity that the cere.

could be made suberitical with the strongest rod fully withdrawn. This reactivity characteristic has been a basic assumption in the analysis of plant performance. Compliance ;

i with this requirement can be demonstrated conveniently only at the time of initial fuel loading or refueling. Therefore, the I demonstration must be such that it will apply to the entire subsequent fuel cycle. The demonstration shall be performed with the reactor core in the cold, zenon-free condition and will show that the reactor is suberitical by at least R + 0.38 percent Wk with the analytically determined strongest control I rod fully withdrawn.

The value of "R", in units of percent Wk, is the amount by which the core reactivity, in the most reactive condition at any time in the subsequent operating cycle, is calculated to be greater than at the time of the demonstration. "R", therefore, is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated

, beginning-of-life core reactivity. The value of "R" must be positive or zero and must be determined for each fuel cycle.

! The demonstration is performed with a control rod which is calculated to be the strongest rod. In determining this

" analytically strongest" rod, it is assumed that every fuel i

assembly of the same type has identical material properties.

In the actual core, however, the control cell material properties vary within allowed manufacturing tolerances, and j i the strongest rod is determined by a combination of the control )

cell geometry and local kx. Therefore, an additional l margin is included in the shutdown margin test to account for the fact that the rod used for the demonstration (the

" analytically strongest") is not necessarily the strongest rod in the core. Studies have been made which campare experimental criticals with calculated criticals. These studies have shown that actual criticals can be predicted within a given tolerance band. For gadolinia cores the additional margin required due to control cell material manufacturing tolerances and i calculational uncertainties has experimentally been determined to be 0.38 percent Wk. When this additional margin is demonstrated, it assures that the reactivity control requirement is set.

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gg, 3.3/4 3-13

! Unit 1 Amendment No. 133 l

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3.3/4.3 BASES (Cont'd)

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2. Reactivity martin - inoperable control rods - Specification ,

3.3.A.2 requires that a rod be taken out of service if it cannot be moved with drive pressure. If the rod is fully inserted and disarmed electrically *, it is in a safe position of maximum contribution to shutdown reactivity. If it is disarmed electrically in a nonfully inserted position, that position shall be consistent with the shutdown reactivity limitations stated in Specification 3.3.A.1. This assures that the core can be shut down at all times with the remaining control rods assuming the strongest operable control rod does

. not insert. Also if damage within the control rod drive mechanism and in particular, cracks in drive internal housings, cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled out. Circumferential cracks resulting from stress-assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs. This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented-in the affected rods.

Limiting the period of operation with a potentially severed rod after detecting one stuck cod will assure that the esactor will not be operated with a large number of rods with failed collet housings. The Rod Sequence Control System is not automatically bypassed until reactor power is above 20 percent power.

Therefore, control rod movement is restricted and the single notch exercise surveillance test is only performed above this power level. The Rod Sequence Control System prevents movement of out-of-sequence rods unless power is above 20 percent.

B. Control Rods

1. Control rod dropout accidents as discussed in the FSAR can lead to significant core damage. It coupling integrity is maintained, the possibility of a rod dropout accident is eliminated. The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides a verification that the rod is following its drive.

Absence of such response to drive movement could indicate an uncoupled condition. Rod position indication is required for proper function of the Rod Sequence Control System and the rod worth minimizer. ~

.

To disarm the drive electrically, four amphenol type plus connectors are removed from the drive insert and withdrawal solenoids rendering the rod incapable of withdrawal. This procedure is equivalent to valving out the drive and is preferred because, in this condition, drive water cools and minimizes crud accumulation in the drive. Electrical disarming does not eliminate position indication.

l BFN 3.3/4.3-14 Unit 1 Anendment No. 133

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3.3/4.3 BASgS,-(Cent'd)

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. 2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of.a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system. The design basis is given in subsection 3.5.2 of the FSAR and the safety evaluation is given in subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing. Additionally, the support is not required if all control rods are fully inserted and if an adequate shutdown margin with one control rod withdrawn has-been demonstrated, since the reactor would remain suberitical even in the event of complete ejection of the strongest control rod.

3. The Rod Worth Minimizer (RWM) and the Rod Sequence Control System (RSCS) restrict withdrawals and insertions of control I rods to prospecified sequences. All patterns associated with these sequences have the characteristic that, assuming the worst single deviation from the sequence, the drop of any control rod from the fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in any pellet average enthalpy in excess of 280 calories per gram. An enthalpy of 280 calories j per gram is well below the level at which rapid fuel dispersal

could occur (i.e., 425 calories per gram). Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly dispersed. Reference Sections 3.6.6, 7.7.A. 7.16.5.3, and 14.6.2 of the FSAR, and NEDO-10527 and ,

supplements thereto.

In performing the function described above, the RWN and RSCS I are not required to impose any restrictions at core power levels in excess of 20 percent of rated. Natorial in the cited reference shows that it is impossible to reach 280 calories per i

gram in the event of a control rod drop occurring at power

greater than 20 percent, regardless of the rod pattern. This

is true for all normal and abnormal patterns including those I which maximize individual control rod worth.

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l'* -15 Amendment No. 133 hkft1

3.3/4.3 8ASJS.(Cent'd)

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At power levels below 20 percent of re,ted, abnormal control rod ,

patterns could produce co'd worths high enough to be of concern i relative to the 280 calorie per gram rod drop limit. In this range the RWN and the RSCS constrain the control rod sequences and patterns to those which involve only acceptable rod worths.

The Rod Worth Minimizer and the Rod Sequence Control System provide automatic supervision to assure that out of sequence

' control rods will not be withdrawn or inserted; i.e., it limits operator deviations fron-planned withdrawal sequences. Ref.

. Section 7.16.5.3 of the FSAR. They serve as a backup to procedure control of control rod sequences, which limit the maximum reactivity worth of control rods. Except during l specified exceptions, when the Rod Worth Minimizer is out of service a second licensed operator can manually fulfill the ;

j control rod pattern conformance functions of this system. In ,

! this case, the RSCS is backed up by independent procedural l controls to assure conformance. ,

I The functions of the RWN and RSCS make it unnecessary to l specify a license limit on rod worth to preclude unacceptable .

consequences in the event of a control rod drop. At low i powers, below 20 percent, these devices force adherence to acceptable rod patterns. Above 20 percent of rated power, no constraint on rod pattern is required to assure that rod drop accident consequences are acceptable. Control rod pattern constraints above 20 percent of rated power are imposed by power distribution requirements, as defined in Sections 3.5.I.

3.5.J. 4.5.I. and 4.5.J of these technical specifications.

I Power level for automatic bypass of the RSCS function is sensed j by first stage turbine pressure.

l Because it is allowable to bypass certain rods in the RSCS 4

during specified testing below 20 percent of rated power in the startup or run modes, a second licensed operator is not an

. acceptable substitute for the RWM during this testing.

! 4. The Source Range Monitor (SRN) system performs no automatic I safety system function; i.e., it has no scram function. It j does provide the operator with a visual indica $1on of neutron i

level. The consequences of reactivity accidents are functions j of the initial neutron fluz. The requirement of at least i 3 counts per second assures that any transient, should it

! occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions.

!- One operable SRN channel would be' adequate to monitor the approach to criticality using homogeneous patterns of scattered 4 control rod withdrawal. A minimum of two operable SRMs are provided as an added conservatism.

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1 3FN 3.3/4.3 16 j Unit 1 Amendment No. 133 3

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- - - . . - - _ - - - - - - - . _ . - . . - - _ . . - . . - . . _ - - . - . . , - - , . , - _ , .---,-.-.-,J

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I 3.3/4.3 R&SgS (Cent'd) o . .

i I - 5. The Rod Block Monitor (RBN) is designed to automatically i prevent fuel damage in the event of erroneous rod withdrawal i from locations of high power density during high power level

! operation. Two RBN channels are provided, and one of these may be bypassed from the console for maintenance and/or testing.

i Automatic rod withdrawal blocks from one of the channels will block erroneous rod withdrawal soon enough to prevent fuel ,

i damage. The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.

A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit, (i.e., NCPR given j by Specification 3.5.k or LNGR of 13.4 kW/ft. During use of l such patterns, it is judged that testing of the RBN system l prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur. It is normally the responsibility of the nuclear engineer to identify i these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to

the occurrence of inoperable control rods in other than

' limiting patterns. Other personnel qualified to perform these functions may be designated by the plant superintendent to 4 perform these functions.

C. Scram Insertion Times 1 The control rod system is designated to bring the reactor suberitical at

the rate fast enough to prevent fuel damage; i.e., to prevent the NCPR from becoming less than 1.07. The limiting power transient is given in Reference 1. Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification provide the requirjd ;

protection, and NCPR remains greater than 1.07. -

On an early BWR, some degradation of control rod scram performance occurred during plant startup and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter. The design of the present control rod drive (Model 7RDB1448) is grossly improved by the relocation of the filter to a 4

location out of the scram drive path; i.e., it can no longer interfere with scram performance, even if completely blocked.

i The degraded performance of the original drive (CRD7RD8144A) under dirty

operating conditions and the insensitivity of the redesigned drive (CRD7AD81448) has been demonstrated by a s'ories of engineering tests

under simulated reactor operating conditions. The successful performance

. of the new drive under actual operating conditions has also been i

demonstrated by consistently good in-service test results for plants i using the new drive and may be inferred from plants using the older model i

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BFN 3.3/4.3-17

-Unit 1 Amendment No. 133 i l

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_ _ . _ , . _ . _ _ _ _ - - _ _ . _ . _ , . _ -. _ _ _ _ __ _- ._..__.__._-__-_..-..J

3.3/4.3 8_AfES E (Cont'd) drive with a modified (larger screen size) internal filter which is less' ,

prone to plugging. Data has been documented by surveillance reports in various operating plants. These include Oyster Creek, Monticello, j Dresden 2 and Dresden 3. Approximately 5000 drive tests have been recorded to date.

Following identification of the " plugged filter" problem, very frequent 4 scram tests were necessary to ensure proper performance. However..the more frequent scram tests are now considered totally unnecessary and i unwise for the following reasons:

1. Erratic scram performance has been identified as due to an obstructed drive filter in type "A" drives. The drives in BFNP are of the new "B" type design whose scram performance is unaffected by

{ filter condition.

I 2. The dirt load is primarily released during startup of the reactor when the reactor and its systems are first subjected to flows and '

pressure and thermal stresses. Special attention and measures are now being taken to assure cleaner systems. Reactors with drives f identical or similar (shorter stroke, smaller piston areas) have

operated through many refueling cycles with no sudden or erratic changes in scram performance. This preoperational and startup testing is sufficient to detect anomalous drive performance.

I 3. The 72-hour outage limit which initiated the start of the frequent scram testing is arbitrary, having no logical basis other than quantifying a " major outage" which might reasonably be caused by an event so severe as to possibly affect drive performance. This '

l requirement is unwise because it provides an incentive for shortcut l

actions to hasten returning "on line" to avoid the additional l

testing due a 72-hour outage. ,

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i BFN 3.3/4.3-18 Unit 1 knendment No. 133

3.3/4.3 B R (C:nt'd)

[.' The surveillance requirement for scram testing of all the control rods after each refueling outage and 10 percent of the control rods at 16-week l

intervals is adequate for determining the operability of the control rod system yet is not so frequent as to cause excessive wear on the control rod system components.

The numerical values assigned to the predicted scram performance are !

based on the analysis of data from other BWRs with control rod drives.the same as those on Browns Ferry Nuclear Plant.

The occurrence of scram times within the limits, but significantly longer j than the average, should be viewed as an indication of systematic problem with control rod drives especially if the number of. drives exhibiting such scram times exceeds eight, the allowable number of inoperable rods. -l In the analytical treatment of the transients which are assumed to scram on high neutron flux, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of control rod motion.

This.is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results. i Approximately the first 90 milliseconds of each of these time intervals result fram sensor and circuit delays after which the pilot scram solenoid doenergizes to 120 milliseconds later, the control rod motion is estimated to actually begin. However, 200 milliseconds, rather than 120 :

milliseconds, are conservatively assumed for this time interval in the

' transient analyses and are also included in the allowable scram insertion j J

times of Specification 3.3.C.

i In order to perform scram testing as required by Specification 4.3.C.1, !

the relaxation of certain restraints in the rod sequence control system

) is required. Individual rod bypass switches may be used as described in i Specification 4.3.C.l.

l The position of any rod bypassed must be known to be in accordance with !

j rod withdrawal sequence. Bypassing of rods in the manner described in i Specification 4.3.C.1 will allow the subsequent withdrawal of any rod ]

scrammed in the 100 percent to 50 percent rod density groups; however, it j will maintain group notch control over all rods in the 50 percent density 1 to preset power level range. In addition, RSCS will prevent movement of f rods in the 50 percent density to preset power level range until the l scrammed rod has been withdrawn.

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i BFN 3.3/4.3-19 Unit 1 Amendment No. 133 l 1

1,

3.3/4.3 BASES

. l D. Reactivity Anomalies ,

During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred l l

from the critical rod configuration. As fuel burnup progresses .

anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons.

Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds.1 percent WK. Deviations in core reactivity greater than 1 percent WK are not expected and require thorough evaluation.

One percent reactivity into the core would not lead to transients '

exceeding design conditions of the reactor system.

E. No BASES provided for this specification F. Scram Discharae Volume The nominal stroke time for the scram discharge volume vent and '

drain valves is < 30 seconds following a scram. The purpose of these valves is to limit the quantity of reactor water discharged after a scram and no direct safety function is performed. The i surveillance for the valves assures that system drainage is not impeded by a valve which fails to open and that the valves are operable and capable of closing upon a scram.

References

1. Generic Reload Fuel Application, Licensing Topical Report, NEDE-240ll-P-A and Addenda.

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1 SFN 3.3/4.3 20 1 Unit 1 /

Amendment No. 133 l l

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pa tt43"o UNITED STATES

-[" ' ,) NUCLEAR REGULATORY COMMISSION

a WASHINGTON, D. C. 20665

% ..+ .

}

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 129 License No. OPR-52

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment by Tennessee Valley Authority (the licensee) dated June 4, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission'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 Comission; 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 Comission's regulations; D. The issuance of this amendment will not be inimical to the common defense 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 Comission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications 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:

i 2- l l

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised thrnugh Amendment No.129, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance and shall be implemented within 90 days from the date of issuance.

FOR THE NUCLEAR REGULATO Y C0!! MISSION John A. Zwolinski, Assistant Director fo Projects Division of TVA Projects Office of Special Projects l

Attachment:

Chances to the Technical Specifications Date of Issuance: May 13, 1987 1

1 I.

ATTACHMENT TO LICENSE AMENDMENT NO.129 FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Revise the Appendix A Technical Specifications by removing the pages-identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the areas of changes. Overleaf pages are provided to maintain document completeness.*

REMOVE INSERT i i ii 11 3.3/4.3-5 3.3/4.3-5*

3.3/4.3-6 3.3/4.3-6 3.3/4.3-7 3.3/4.3-7 3.3/4.3-8 3.3/4.3-8 3.3/4.3-9 3.3/4.3-9 3.3/4.3-10 3.3/4.3-10**

3.3/4.3-11 3.3/4.3-11**

Bases 3.3/4.3-12 3.3/4.3-12**

Bases 3.3/4.3-13 Bases 3.3/4.3-13**

Bases 3.3/4.3-14 Bases 3.3/4.3-14**

Bases 3.3/4.3-15 Bases 3.3/4.3-15**

Bases 3.3/4.3-16 Bases 3.3/4.3-16**

Bases 3.3/4.3-17 Bases 3.3/4.3-17**

Bases 3.3/4.3-18 Bases 3.3/4.3-18**

Bases 3.3/4.3-19 Bases 3.3/4.3-19**

-- Bases 3.3/4.3-20**

- Pagination change only

TABLE OF CONTENTS Section Pane No.

1.0-1 1.0 Definitions. . . . . . . . . . . . . . . . . . . .

l SAFETY LINITS AND LIMITING SAFETY SYSTEN '

SETTINGS 1.1/2.1 Fuel Cladding Integrity. . . . . . . . . . . . . . 1.1/2.1-1 i

1.2/2.2 Reactor Coolant System Integrity . . . . . . . . . 1.2/2.2-1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.1/4.1 Reactor Protection System . . . . . . . . . . . . . 3.1/4.1-1 3.2/4.2 Protective Instrumentation. . . . . . . . . . . . . 3.2/4.2-1 A. Primary Containment and Reactor Building Isolation Functions. . . . . . . . . . . . . 3.2/4.2-1 B. Core and' Containment Cooling Systems -

Initiation and Control . . . . . . . . . . . 3.2/4.2-1 C. Control Rod Block Actuation. . . . . . . . . . 3.2/4.2-2 1

i D. Radioactive Liquid Effluent Monitoring i Instrumentation. . . . . . . . . . . . . . . 3.2/4.2-3 E. Drywell Leak Detection . . . . . . . . . . . . 3.2/4.2-4 I

F. Surveillance Instrumentation . . . . . . . . . 3.2/4.2-4 j

G. Control Room Isolation . . . . . . . . . . . . 3.2/4.2-4 H. Flood Protection . . . . . . . . . . . . . . . 3.2/4.2-4 I. Noteorological Monitoring Instrumentation. . . 3.2/4.2-4 J. Seismic Monitoring Instrumentation . . . . . . 3.2/4.2-5 K. Radioactive Gaseous Effluent Monitoring Instrumentation . . . . . . . . . . . . - . . 3.2/4.2-6 3.3/4.3 Reactivity Control . . . . . . . . . . . . . . . . 3.3/4.3-1 A. Reactivity Limitations . . . . . . . . . . . . 3.3/4.3-1 B. Control Rods . . . . . . . . . . . . . . . . . 3.3/4.3-5 C. Scram Insertion Times. . . . . . . . . . . . . 3.3/4.3-10 BFN-Unit 2 Amendment No. 129 v- ,_m- - _ - - . . -

r--e,, - . , . , , - _ - .-m-, .,---__ - _ ., , ,--, _ - , - - -, ,--e .--,-,y, v ,-- y.y- r -s,...- m., --.--- y-

f Section Page No. .

D. Reactivity Apomalies . . . . . . . . . . . .. 3.3/4.3-11 E. Reactivity Control . . . . . . . . . . . . . . 3.3/4.3-12 F. Scram Discharge Volume . . . ......... 3.3/4.3-12 3.4/4.4 Standby Liquid Control System. . . . . . . . . . . 3.4/4.4-1 1

A. Normal System Availability . . . . . . . . . . 3.4/4.4-1 j B. Operation with Inoperable Components . . . .. 3.4/4.4-2 C. Sodium Pentaborate Solution. . . . . . . . . . 3.4/4.4-3 3.5/4.5 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 4

C. RHR Service Water System and Emergency Equipment Cooling Water System (EECWS) ... 3.5/4.5-9 D. hquipmentAreaCoolers.... ........ 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 i G. Automatic Depressurization System (ADS). ... 3.5/4.5-15 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 K. Minimum Critical Power Ratio (MCPR). ..... 3.5/4.5-19 L. APRM Setpoints . . . . . ........... 3.5/4.5-20 M. Reporting Requirements . . .......... 3.5/4.5-20 i

N. References . . . . . . . ........... 3.5/4.5-32 3.6/4.6 Primary System Boundary. . . ........... 3.6/4.6-1 A. Ihermal and Pressurirstion Limitations . ... 3.6/4.6-1 ;

8. Coolant Chemistry. . . . . .......... 3.6/4.6-5 )

BFN-Unit 2 11 Amendment No.129 i

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l 3.3/4.3 REACTIVITY CONTROL Lm m m,wwuiAAwa EvR vrRI G ?*'""'"""""""""

Control Rods 4.3.B. Control Rods 3.3.B.

1. The coupling integrity

1. Each control rod shall be coupled to its drive or shall be verified for ,

completely inserted and the each withdrawn control l control rod directional rod as follows:

control valves disarmed Verify that the electrically. This a.

control rod is requirement does not apply following the drive in the REFUEL condition by observing a when the reactor is vented. response in the Two control rod drives may -

be removed as long as nuclear instru-Specification 3.3.A.1 mentation each time is met.

a rod is moved when the reactor is operating above the preset power level of the RSCS.

b. When the rod is fully withdrawn the first time after each refueling outage or after maintenance, observe that the drive does not go to the overtravel position.

2. The control rod drive 2. The control rod drive housing support system shall housing support system shall be inspected be in place during reactor after reassembly and power operation or when the reactor coolant system is the results of the pressurized above atmospheric inspection recorded.

pressure with fuel in the reactor vessel, unless all control rods are fully inserted and specification 3.3.A.1 is met.

a BrW 3.3/4.3-5 Unit 2

3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.8. Control Rods 4.3.8. Control Rods c

3.a Whenever the reactor is in 3.a.1 The Rod Sequence Control the startup or run modes System (RSCS) shall be below 20% rated power, the demonstrated to be OPERABLE Rod Sequence Control System for a reactor startup-by (RSCS) shall be OPERABLE, the following checks:

except that the RSCS constraints may be a. Performance of the suspended by means of the comparator check of individual rod bypass group notch circuits switches for within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior 1 - special criticality to control rod tests, or withdrawal for the purpose of making 2 - control rod scram the reactor critical timing per 4.3.C.l.

When RSCS is bypassed on b. Selecting and individual rods for these attempting to exceptions, RWN must be withdraw an out-of-operable per 3.3.8.3.b and sequence control rod a second party verification after withdrawal of may not be used in lieu of the first insequence RWN. control rod.

c. Attempting to withdraw a control rod more than one notch prior to other control rod movement after the group notch mode is automatically initiated.

3.a.2 The Rod Sequence Control System (RCS) shall be '

demonstrated to be OPERABLE for a reactor shutdown by the following checks:

a. Performance of the comparator check of

' - the group notch circuits within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to automatic initiation of the group notch mode.

8FN 3.3/4.3-6 Unit 2 Amendment No. 129 l

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1 3-3/4 3 REACTIVITY CONTROL I 1

SURVEILLANCE REQUIRENENTS

. LIMITING CONDITIONS FOR OPERATION J Control Rods 4.3.B. Control Rods 3.3.B.

3.a.2 (Cont'd)

b. Attempting to insert a control rod more than one notch prior to other control rod movement after the group notch mode is automatically initiated.

c. Selecting and attempting to noie an out-of-sequence .

4 control rod after insertion of the first insequence control rod after reaching a black and white rod pattern.

3.b. Whenever the reactor is 3.b.1 The Rod Worth Minimizer in the startup or run modes (RWN) shall be demonstrated below 20% rated power, the to be OPER&BLE for a reactor startup by the Rod Worth Minimizer (RWM) shall be OPERABLE. With following checks:

the RWN INOPERABLE, verify control rod movement a. By demonstrating that and compliance with the the control rod prescribed control rod patterns and sequence pattern by a second input to the RWK .

licensed operator or computer are correctly l l loaded following any 1 j

other technically I qualified member of the loading of the program.

plant staff who is into the computer.

f

! present at the reactor control console. b. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior Otherwise, control rod to withdrawal of control rods for the movement may be only by actuating the manual purpose of making the

]

! scram or placing the reactor critical i reactor mode switch in verify proper l the shutdown position. annunciation of the selection error of at I least one out-of-sequence control rod. i l

l BFN 3.3/4.3-7 Amendment No. 129 Unit 2 ,

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3.3/4.3 REACTIVITY CONTROL .

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRENENTS .

3.3.8. Control Rods 4.3.8. Control Rhes 3.b.1 (Cont'd)

c. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to withdrawal of control rods for the i purpose of making the

reactor critical, the <

rod block function of the RWN shall be 4

verified by moving an out-of-sequence j

control rod.

3.b.2 The Rod Worth Minimizer (RWN) shall be demonstrated to be OPERA 8LE for a reactor

~

shutdown by the following checks:

i

a. By demonstrating that the control rod patterns and sequence i input to the RWN computer are correctly i loaded following any loading of the program into the computer.

b. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to RWN automatic 1

initiation when reducing thermal power, l

verify proper annunciation of the selection error of at least one out-of-sequence control rod.

i . c. Within one hour after

! RWN automatic i

initiation when reducing thermal power, the rod block function of the RWN shall be i verified by moving an I out-of-sequence control rod.

BFN 3.3/4.3-8

Unit 2 Amendment No. 123

.y. - , . , . , , _ , - , , , _ _ , - _ _ , , _ . . ..w .. , , . - , _--____,,.m , , , _ , . . - - _ . _ _ _ , _ _ . _ _ _ , . _ _ , _ _ . _ . , _ _ . , _ _ ~ _ . . , - . _ . - _

,,.y, -

-3.3/4.3 RELCTIVITY CONTROL

~

SURVEILLANCE REQUIRENENTS LIMITING CONDITIONS FOR OPERATION Control Rods 4.3.B. Contro.1 Rods 3.3.B.

3.c. If Specifications 3.3.B.3.a 3.b.3 When the RWN is not through .h cannot be met the OPERABLE a second l reactor shall not be started, licensed operator i or other technically l or if the reactor is in the qualified member of i run or startup modes at less than 20% rated power, the plant staff shall ]

control rod movement may be verify that the correct only by actuating the rod prog:sm is followed manual scram or placing except as specified in the reactor mode switch in 3.3.B.3.a.

the shurriawn position.

i 4. Jontrol rods shall not be 4. Prior to control rod ,

withdrawn for startup or withdrawal for startup or during refueling, J

refueling unless at least two source range channels verify that at least two have an observed count rate source range channels equal to or greater than have an observed count three counts per second. rate of at least three counts per second.

During operation with S. When a limiting 5.

limiting control rod control rod pattern patterns, as determined by exists, an instrument the designated qualified functional test of the personnel, either: RBN shall be performed prior to withdrawal of

a. Both RBN channels shall the designated rod (s) be operable: and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

or l

b. Control rod withdrawal l shall be blocked.

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i BFN 3.3/4.3-9 Amendment No. 129 Unit 2

3.3/4.3 REACTIVITY CONTROL .

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.C. Scram Insertion Times 4.3.C. Scram Insertion Times

1. The average scram 1. After each refueling i insertion time, based on outage, all OPERABLE rods the deenergization of the shall be scram-time scram pilot valve sole- tested from the fully noids as time zero, of withdrawn position with all OPERABLE control rods the nuclear system in the reactor power pressure above 800 psig.

operation condition shall This testing shall be be no greater than: completed prior to exceeding 40% power.

Below 20% power, only rods in those sequences (A12 and A34 or 812 and

% Inserted From Ava. Scram Inser- 834) which were fully Fully Withdrawn tion Times (sec) withdrawn in the region from 100% cod density to

5 0.375 Sp% rod density shall be 20 0.90 scram-time tested. The 50 2.0 sequence restraints 90 3.500 imposed upon the control rods in the 100-50 percent rod density groups to the l preset power level may be removed by use of the individual bypass switches associated with those control rods which are tully or partially withdrawn and are not within the 100-50 percent rod density groups. In I order to bypass a rod, the ;

actual rod axial position j must be known; and the rod I must be in the correct ,

in-sequence position. l l

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g[{ 3.3/4.3-10 Amendment No.129 1

3.3/4.3 REACTIVITY CONTROL SURVEILLANCE REQUIRENENTS

. LINITING CONDITIONS FOR OPERATION 3.3.C. Scram Insertion Times 4.3.C. Scram Insertion Times

2. At 16-week intervals,10%

2. The average of the scram inser-tion times for the three fastest of the OPERA 8LE control OPERABLE control rods of all rod drives shall be scram-timed above 800 psig.

groups of four control rods in a two-by-two array shall be no Whenever such scram time ;

measurements are made, an greater than:

evaluation shall be made 1 Inserted From Avg. Scram Inser- to provide reasonable tion Times (sec) assurance that proper Fully Withdrawn control rod drive 5 0.398 performance is being 20 0.954 maintained.

50 2.120 90 3.800

a. The maximum screa insertion time for 90% insertion of any OPERABLE control rod shall not exceed 7.00 seconds.

Reactivity Anomalies D. Reactivity Anomalies D.

l The reactivity equivalent of During the STARTUP test the difference between the program and STARTUP following actual critical rod refueling outages, the configuration and the expected critica*5 rod configurations configuration during power will bei compared to the operation shall not exceed 1% Ak. expected configurations at If this limit is exceeded, the selected operating conditions, reactor will be shut down These comparisons will be until the cause has been used as base data for

determined and corrective reactivity monitoring during actions have been taken as subsequent power operation appropriate. throughout the fuel cycle.

l At specific power operating conditions, the critical rod i

configuration will be

' compared to the configuration i

expected based upon appropriately corrected past l data. This e.omparison will be made at least every full l power month.

8FN 3.3/4.3-11 Unit 2 Amendment No. 129

-,-- - , , , ,,-- ,,- , - , , , , - , , , , , , ---.-,,--.,--,,,-,-g,--m-.g,,-,--. . - ,,,y-, - , .,,,-,,-,-

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3.3/a.3 REACTIVITY CONTROL .

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.E. If Specifications 3.3.C and .D 4.3.E. Surveillance requirements are above cannot be met, an orderly as specified in 4.3.C'and .D shutdown shall be initiated and' above.

the reactor shall be in the shutdown condition within ' '

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

F. Scram Discharge Volume (SD7) F. Scram Discharge Volume (SD7) f 1. The scram discharge volume 1.a. The scram discharge j drain and vent valves shall volume drain and vent be OPERA 8LE any time that valves shall be verified ;

the reactor protection open prior to each system is required to be STARTUP and monthly OPERA 8LE except as thereafter. The valves

specified in 3.3.F.2. may be closed intermittently for-testing not to exceed I hour in any 24-hour j

period during operation.

1.b. The scram discharge volume drain and vent valves shall be demonstrated OPERABLE

! monthly.

2. In the event any SDV drain 2. When it is determined or vent valve becomes that any SDV drain or INOPERA8LE, reactor vent valve is INOPERABLE, i operation may continue the redundant drain or provided the redundant vent valve shall be

drain or vent valve is demonstrated OPERA 8LE f OPERABLE. immediately and weekly 2 thereafter.-

3. If redundant drain or vent 3. No additional valves become INOPERA8LE, surveillance required, the reactor shall be in hot standby within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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3. 3 / a . 3 -12 gFyt2 Amendment No.129 l

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3.3/4.3 R&SES

. A. Reactivity Limitation 1.- The requirements for the control rod drive system have been identified by evaluating the need for reactivity control via control rod movement over the full spectrum of plant conditions and events. As discussed in subsection 3.4 of the Final Safety Analysis Report, the control rod system design is intended to provide sufficient control of core reactivity that the core could be made suberitical with the strongest rod fully withdrawn. This reactivity characteristic has been a basic assumption in the analysis of plant performance. Compliance with this requirement can be demonstrated conveniently only at the time of initial fuel loading or refueling. Therefore, the demonstration must be such that it will apply to the entire subsequent fuel cycle. The demonstration shall be performed with the reactor core in the cold, zenon-free condition and will show that the reactor is suberitical by at least R + 0.38 percent Wk with the analytically determined strongest control rod fully withdrawn.

The value of "R", in units of percent Wk, is the amount by which the core reactivity, in the most reactive condition at any time in the subsequent operating cycle, is calculated to be greater than at the time of the demonstration. "R", therefore, is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of "R" must be positive or zero and must be determined for each fuel cycle.

The demonstration is performed with a control rod which is calculated to be the strongest rod. In determining this

" analytically strongest" rod, it is assumed that every fuel assembly of the same type has identical material properties. In the actual core, however, the control cell material properties vary within allowed manufacturing tolerances, and the strongest I rod is determined by a combination of the control cell geometry and local kx. Therefore, an additional margin is included in the shutdown margin test to account for the fact that the rod used for the demonstration (the " analytically strongest") is not necessarily the strongest rod in the core. Studies have been made which compare experimental criticals with calculated criticals. These studies have shown that actual criticals can be predicted within a given tolerance band. For gadolinia cores the additional margin required due to control cell material manufacturing tolerances and calculational uncertainties has experimentally been determined to be 0.38 percent Wk. When this additional margin is demonstrated, it assures that the reactivity control requirement is met.

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OYt2 knendment No. 129

3.3/4.3 gggi,(Cont'd)

= 2. Reactivity Narain - INOPERABLE Control Rods - Specification 3.3.A.2 requires that a rod be taken out of service if it cannot be moved with drive pressure. If the rod is fully inserted and i disarmed electrically *, it is in a safe position of maximum contribution to shutdown reactivity. If it is disarmed electrically in a nonfully inserted position, that position shall be consistent with the shutdown reactivity limitations stated in i Specification 3.3.A.1. This assures that the core can be shut down at all times with the remaining control rods assuming the

' strongest OPERABLE control rod does not insert. Also if damage within the control rod drive mechanism and in particular, cracks in drive internal housings, cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled out.

Circumferential cracks resulting from stress-assisted i intergranular corrosion have occurred in the collet housing of i drives at several BWRs. This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the

affected rods. Limiting the period of operation with a potentially severed rod after detecting one stuck rod will assure i that the reactor will not be operated with a large number of rods l

with failed collet housings. The Rod Sequence Control System is not automatically bypassed until reactor power is above 20 i

percent power. Therefore, control rod movement is restricted and the single notch exercise surveillance test is only performed above this power level. The Rod Sequence Control System prevents j movement of out-of-sequence rods unless power is above 20 percent.

j j B. Control Rods

-! 1. Control rod dropout accidents as discussed in the FSAR can lead l to significant core damage. It coupling integrity is maintained, the possibility of a rod dropout accident is eliminated. The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides a verification i that the rod is following its drive. Absence of such response to drive movement could indicate an uncoupled condition. Rod. :

position indication is required for proper function of the Rod i i Sequence Control System and the rod worth minimiter.

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To disarm the delve electrically, four amphenol type plus connectors are l I removed from the drive insert and withdrawal solenoids rendering the rod

- incapable of wi.thdrawal. This procedure is equivalent to valving out the drive and is preferred because, in this condition, drive water cools and i minimizes crud accumulation in the drive. Electrical disarming does not

eliminate position indication.

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ggy 3.3/4.3-14 Amendment No. 129 l l

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3.3/4.3 JAfig,.(C:nt'd)

- 2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the estremely remote event of a housing failure. The amount of reactivity which could be j added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system. The design basis is given in subsection 3.5.2 of the FSAR and the safety evaluation is given in subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric j pressure since there would then be no driving force to rapidly

eject a drive housing. Additionally, the support is not required j' if all control rods are fully inserted and if an adequate shutdown margin with one control rod withdrawn has been

! demonstrated, since the reactor would remain suberitical even in the event of complete ejection of the strongest control rod.

The Rod Worth Minimizer (RWM) and the Rod Sequence Control System

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j 3.

i (RSCS) restrict withdrawals and insertions of control rods to prospecified sequences. All patterns associated with these -

sequences have the characteristic that, assuming the worst single

deviation from the sequence, the drop of any. control rod from the >

i fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power escursion l

resulting in any pellet average enthalpy in excess of 280 l

calories per gram. An enthalpy of 280 calories per gram is well j below the level at which rapid fuel dispersal could occur (i.e.,

( 425 calories per gram). Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly dist orsed. Reference Sections 3.6.6, 7.7.A. 7.16.5.3, and 14.6.2 i

of the FSAR, and NEDO-10527 and supplements thereto, i

j In performing the function described above, the RWN and RSCS are

]

not required to impose any restrictions at core power levels in j ,

excess of 20 percent of rated. Material in the cited reference

! shows that it is impossible to reach 280 calories per gram in the event of a control rod drop occurring at power greater than i

20 percent, regardless of the rod pattern. This is true for all j normal and abnormal patterns including those which maximize individual control rod worth.

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I g 3.3/4.3-15 Amendment No. 129 l

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l 3.3/4.3 BASES (Cent'd) ,

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At power levels below 20 percent of rated, abnormal control rod patterns could produce ro'd worths high enough to be of concern relative to the 280 calorie per gram rod drop limit. In this range the RWM and the RSCS constrain the control rod sequences and patterns to those which involve only acceptable rod worths.

The Rod Worth Minimizer and the Rod Sequence Control System provide automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences. Ref.

Section 7.16.5.3 of the FSAR. They serve as a backup to procedure control of control rod sequences, which limit the maximum reactivity worth of control rods. Except during specified exceptions, In the event that the Rod Worth Minimizer is out of service, when required, a second licensed operator can manually fulfill the control rod pattern conformance functions of this system. In this case, the RSCS is backed up by independent procedural controls to assure conformance.

The functions of the RWM and RSCS make it unnecessary to specify a license limit on rod worth to preclude unacceptable consequences in the event of.a control rod drop. At low powers, below 20 percent, these devices force adherence to i acceptable rod patterns. Above 20 percent of rated power, no constraint on rod pattern is required to assure that rod drop accident consequences are acceptable. Control rod pattern constraints above 20 percent of cated power are imposed by power distribution requirements, as defined in Sections 3.5.I.

3.5.J. 4.5.I. and 4.5.J of these technical specifications.

Power level for automatic bypass of the RSCS function is sensed by first stage turbine pressure.

Because it is allowable to bypass certain rods in the RSCS during scram time testing below 20 percent of rated power in the STARTUP or RUN modes, a second licensed operator is not an acceptable substitute for the RWM during this testing, j 4 The Source Range Monitor (SRM) system performs no automatic l safety system function; i.e., it has no scram function. It ,

does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are functions of the initial neutron fluz. The requirement of at least 3 counts per second assures that eny transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions.

One OPERABLE SRM channel would be adequate to monitor the

- approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two OPERABLE SRMs are provided as an added conservatism.

l 3.3/a.3 16 SFN Unit 2 Amendment No. 129

3.3/4.3 8MES, (Co:t'd)

5. The Rod Block Monitor (R8N) is designed to automatically I prevent fuel damage in the event of erroneous rod withdrawal

' from locations of high power density during high power level operation. Two RBN channels are provided, and one of these may be bypassed from the console for maintenance and/or testing.

Automatic rod withdrawal blocks from one of the channels will i

block erroneous rod withdrawal soon enough to prevent fuel i damasc. The specified restrictions with one channel out of i service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.

i A limiting control rod pattern is a pattern which results in

the core being on.a thermal hydraulic limit, (i.e., NCPR given by Specification 3.5.k or LNGR of 13.4 kW/ft. During use of such patterns, it is judged that testing of the R8N system prior to withdrawal of such rods to assure its operability 4 will assure that improper withdrawal does not occur. It is l

normally the responsibility of the nuclear engineer to identify

these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to

the occurrence of INOPERABLE control rods in other than limiting patterns. Other personnel qualified to perform these functions may be designated by the plant superintendent to perform these functions.

i C. Scram Insertion Times

! The control red system is designated to bring the reactor suberitical at the rate fast enocish te :: event fuel damage; i.e., to prevent the NCPR from becoming less than 1.07. The limiting power transient is given in Reference 1. Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given.in the above specification provide the required protection, and NCPR remains greater than 1.07.

On an early BWR, some degradation of control rod scram performance

occurred during plant STARTUP and was determined to be caused by particulate material (probably construction debris) plugging an internal

! control rod drive filter. The design of the present control rod drive 4 (Model 7RDB1448) is grossly improved by the relocation of the filter to a 4 location out of the scram drive path; i.e., it can no longer interfere with scram performance, even it. completely blocked.

! The degraded performance of the original drive (CRD7RDB144A) under dirty operating conditions and the insensitivity of the redesigned drive (CRD7ADB1448) has been demonstrated by a series of engineering tests i under simulated reactor operating conditions. The successful performance i

of the new drive under actual operating conditions has also been demonstrated by consistently good in-service test results for plants I using the new drive and may be inferred from plants using the older model 1

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l nFN 3.3/4.3-17 Unit 2 Amendment No. 129

3.3/4.3 BASES (Cant'd) drive with a modified (larger screen size) internal filter which is less prone to plugging. Data has been documented by surveillance reports in !

various operating plants. These include Oyster Creek, Monticello, i Dresden 2 and Dresden 3. Approximately 5000 drive tests have been I recorded to date. 1

! Following identification of the " plugged filter" problem, very frequent scram tests were necessary to ensure proper performance. However, the more frequent scram tests are now considered totally unnecessary and unwise for the following reasons:

1. Erratic scram performance has been identified as due to an obstructed drive filter in type "A" drives. The drives in SFNP are of the new "B" type design whose scram performance is unaffected by f filter condition.

2. The dirt load is primarily released during STARTUP of the reactor when the reactor and its systems are first subjected to flows and pressure and thermal stresses. Special attention'and measures are now being taken to assure cleaner systems. Reactors with drives j identical or similar (shorter stroke, smaller piston areas) have operated through many refueling cycles with no sudden or erratic changes in scram performance. This preoperational and STARTUP testing is sufficient to detect anomalous drive performance.

3. The 72-hour outage limit which initiated the start of the frequent scram testing is arbitrary, having no logical basis other than quantifying a " major outage" which might reasonably be caused by an event so severe as to possibly affect drive performance. This requirement is unwise because it provides an incentive for shortcut

actions to hasten returning "on line" to avoid the additional testing due a 72-hour outage.

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SFN 3.3/4.3-18 Unit 2, Amendment No. 129 an

. . ~ _ _ , . . . . _ _ , . _ , _ _ _ _ _ _ _ _ _ . - _.._...--__,.___.-,._______--_._-.____..___m.. , _ . . _.

d 3.3/4.3 Bg[S,(C:nt'd)

The surveillance requirement for scram testing of all the control rods

after each refueling outage and 10 percent of the control rods at 16-week

. intervals is adequate for determining the operability of the control rod system yet is not so frequent as to cause excessive wear on the control rod system components.

l The numerical values assigned to the predicted scram performance are based on the analysis of data from other BWRs with control rod dri.ves the-same as those on Browns Ferry Nuclear Plant.

The occurrence of scram times within the limits, but significantly longer

'_ than the average, should be viewed as an indication of systematic problem with control rod drives especially if the number of drives exhibiting such scram times exceeds eight, the allowable number of INOPERABLE rods.

In the analytical treatment of the transients which are assumed to scram on high neutron flux, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of control rod motion.

r This is adequate and conservative when compared to the typical time delay

, of about 210 milliseconds estimated from scram test results.

Approximately the first 90 milliseconds of each of these time intervals result from sensor and circuit delays after which the pilot scram solenoid deenergizes to 120 milliseconds later, the control rod motion is

' estimated to actually begin. However, 200 milliseconds, rather than 120 milliseconds, are conservatively assumed for this time interval in the

' transient analyses and are also included in the allowable scram insertion i times of Specification 3.3.C.

In order to perform scram testing as required by Specification 4.3.C.1, the relaxation of certain restraints in the rod sequence control system is required. Individual rod bypass switches may be used as described in Specification 4.3.C.l.

The position of any rod bypassed must be known to be in accordance with rod withdrawal sequence. Bypassing of rods in the manner described in Specification 4.3.C.1 will allow the subsequent withdrawal of any rod i scrammed in the 100 percent to 50 percent rod density groups; however, it

! will maintain group notch control over all rods in the 50 percent density to preset power level range. In addition. RSCS will prevent movement of rods in the 50 percent density to preset power level range until the scrammed rod has been withdrawn.

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1 b!Yt2 Amendment No. 129

i

, 3.3/4.3 BASES D. Reactivity Anomalies .

.During each fuel cycle excess operative reactivity varies ~as fuel depletes and as any burnable poison in supplementary control is

- burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of 1 the critical rod pattern at selected base states to the predic'ted' rod

! inventory at that state. Power operating base conditions provide the j most sensitive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons.

t Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1 percent WK. Deviations in core reactivity greater than i 1 percent WK are not expected and require thorough evaluation. One percent reactivity into the core would not lead to transients i exceeding design conditions of the reactor system.

I E. No BASES provided for this specification l F. Scram Discharae Volume I

j The nominal stroke time for the scram discharge volume vent and drain j valves is $ 30 seconds following a scram. The purpose of these i valves is to limit the quantity of reactor water discharged after a j scram and no direct safety function is performed. The surveillance j for the valves assures that system drainage is not impeded by a valve j which fails to open and that the valves are OPERA 8LE and capable of

closing upon a scram.

I I References i

1. Generic Reload Fuel Application.

Licensing Topical Report, NEDE-24011-P-A and Addenda.

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! 8FN 3.3/4.3 20 Unit 2 Amendment No. 129

1 e

' # UNITE 3 STATES f %,, -

NUCLEAR REGULATORY COMMISSION

. y g E

WASHINGTON, D. C. 20066

/

l TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.104 License No. DPR-68

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment by Tennessee Valley Authority (the licensee) dated June 4, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR a Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D. The issuance of this amendment will not be inimical to the common defense 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 Comission's regulations and all applicable requirements have been satisfied. *

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-68 is hereby amended to read as follows:

2 (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.104, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

i

3. This license anendment is effective as of the date of issuance and shall be implemented within 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY OMMISSION

(.

John . Zwolinski, Assistant Director for Projects Division of TVA Projects Office of Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: May 13, 1937 I

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ATTACHMENTTOLICENSEAMENDMENTNO.154 i

FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by l

the captioned amendment number and contain vertical lines indicating'the areas of changes. Overleaf pages are provided to maintain document completeness.*

REMOVE INSERT j i i ii 11 i

3.3/4.3-5 . 3.3/4.3-5*

I 3.3/4.3-6 3.3/4.3-6 3.3/4.3-7 3.3/4.3-7 3.3/4.3-8 3.3/4.3-8 3.3/4.3-9 3.3/4.3-9 i 3.3/4.3-10 3.3/4.3-10**

3.3/4.3-11 3.3/4.3-11** .

Bases 3.3/4.3-12 3.3/4.3-12**

Bases 3.3/4.3-13 Bases 3.3/4.3-13** '

Bases 3.3/4.3-14**

> Bases 3.3/4.3-14 Bases 3.3/4.3-15 Bases 3.3/4.3-15**

i 1 Bases 3.3/4.3-16 Bases 3.3/4.3-16**

i Bases 3.3/4.3-17 Bases 3.3/4.3-17**

j Bases 3.3/4.3-18 Bases 3.3/4.3-18**

Bases 3.3/4.3-19 Bases'3.3/4.3-19**

-- Bases 3.3/4.3-20**

i i ** Pagination change only ,

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TABLE OF CONTENTS Pane No.

Section I 1.0-1 1.0 Definitions. . . . . . . . . . . . . . . . . . . .

SAFETY LIMITS AND LIMITING SAFETY SYSTEN SETTINGS 1.1/2.1-1 1.1/2.1 Fuel Cladding Integrity. . . . . . . . . . . . . .

1.2/2.2-1 1.2/2.2 Reactor Coolant System Integrity . . . . . . . . .

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.1/4.1-1 3.1/4.1 Reactor Protection System . . . . . . . . . . . . .

3.2/4.2-1 3.2/4.2 Protective Instrumentation. . . . . . . . . . . . .

A. Primary Containment and Reactor Building 3.2/4.2-1 Isolation Functions. . . . . . . . . . . . .

B. Core and Containment Cooling Systems -

3.2/4.2-1 Initiation and Control . . . . . . . . . . .

3.2/4.2-2 l C. Control Rod Block Actuation. . . . . . . . . .

D. Radioactive Liquid Effluent Monitoring 3.2/4.2-3 Instrumentation. . . . . . . . . . . . . . .

Drywell Leak Detection . . . . . . . . . . . . 3.2/4.2-4 E.

3.2/4.2-4 F. Surveillance Instrumentation . . . . . . . . .

3.2/4.2-4 G. Control Room Isolation . . . . . . . . . . . .

3.2/4.2-4 H. Flood Protection . . . . . . . . . . . . . . .

Meteorological Nonitoring Instrumentation. . . 3.2/4.2-4 I.

Seismic Monitoring Instrumentation . . . . . . 3.2/4.2-5 J.

E. RadioactiveGaseous Rffluent'Nonitoring i 3.2/4.2-6 Instrumentation ..............

Reactivity Control .......'......... 3.3/4.3-1 3.3/4.3 Reactivity Limitations . . . . . . . . . . . .

3.3/4.3-1 A.

3.3/4.3-5 2 8. Control Rods . . . . . . . . . . . . . . . . .

3.3/4.3-10 C. Scram Insertion Times. . . . . . . . . . . . .

L Amendment No. 104 8FN-Unit 3 1

Section Pare No.

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

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3.4/4.4 Standby Liquid Control System. . . . . . . . . . . 3.4/4.4-1 A. Normal System Availability . . . . . . . . . . 3.4/4.4-1

8. Operation with Inoperable Components . . . . . 3.4/4.4-2 C. Sodium Pentaborate Solution. . . . . . . . . . 3.4/4.4-3 3.5/4.5 Core and Containment Cooling Systems . ...... 3.5/4.5-1 A. Core Spray System (CSS). ........... 3.5/4.5-1

8. Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling) ....... 3.5/4.5-4 C. RHR Service Water System and Emergency i Equipment Cooling Water System (ERCWS) ... 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 i

i F. Reactor Core Isolation Cooling System (RCICS). . . . . . . . ...,........ 3.5/4.5-14 I

G. Automatic Depressurization System (ADS). ... 3.5/4.5-15 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 K. Minimum Critical Power Ratio (MCPR). ..... 3.5/4.5-19 L. APRM Setpoints . . . . . . . . . ....... 3.5/4.5-20 M. Reporting Requirements . . . . . . . . . . . . 3.5/4.5-20 N. References . . . . . . . ........... 3.5/4.5-35 3.6/4.6 Primary System Boundary. . . . . . . . . . . . . . 3.6/4.6-1 A. Thermal and Pressurization Limitations . . . . 3.6/4.6-1

8. Coolant Chemistry. . . . ...........

3.6/a.6-S 8FN. Unit 3 11 Amendment No. 104 s

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2,,.J.,J a..J. . A. ,a M- _ _ . _ , m_f _ A. -

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).3/4.3 RsACTIVITY CONTROL .

SURVEILLANCE REQUIREMENTS

! 1.IMITING CONDITIONS FOR OPERATION i

Control Rods 4.3.u. gontrol Rods i 3.3.3.

i Each control rod shall be 1. The coupling integrity '

l 1.

coupled to its drive or shall be verified,for f completely inserted and the - each withdrawn contro! :

control rod directional rod as follows:

) control valves disarmed electrically. This a. Verify that the i requirement does not apply control rod is following the drive j

- in the refool condition by observing a when the reactor is wented, I tesponse in the f Two control rod drives may be removed as long as nu? lear instru- .

specification 3.3.A.1 mentation each time [

I a rod is moved when the l is met.

I reactor is operating i' above the preset power '

i 1evel of the RSCS.

{

b. When the rod is fully withdrawn the first '

time after each ,

] refueling outage or '

l after maintenance, t j

,) observe that the drive f does not go to the :

f overtravel position, l

h

2. The control rod drive 2. The control rod drive 1

housing support system shall housing support system lo be in place during reactor shall be' inspected j

power operation or when the after reassembly and ,

i reactor coolant system is the results of the

) inspection recorded.

pressurized above atmospheric i l pressure with fuel in the >

.i' reactor vessel, unless all j i control rods are fully !

i inserted and specification ,

3.3.A.1 is met. ,

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BPW-Unit 3 3.3/4.3-5 1

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- i 3.3/4.3 REACTIVITY CONTROL

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LINITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRENENTS ,

l 3.3.8. Control Rods 4.3.8. Control Rods ,

i 3.a Whenever the reactor is in 3.a.1 The Rod Sequence Control -

the startup or run modes System (23CS) shall be

'r below 201 rated power, the demonstrated to be OPERABLE '

Rod Sequence Control System for a reactor startup by ;

i (RSCS) shall be OPERABLE, the following checks: ,

1 except that the RSCS l constraints may be a. Performance of the

suspended by means of the comparator check of individual rod bypass group notch circuits

! switches for within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior 1 - special criticality to control rod , ,

tests, or withdrawal for the l purpose of making 2 - control rod scram the reactor critical timing per 4.3.C.1.

! When RSCS is byparsed on b. Selecting and ,

individual rods for these attempting to 3 ,

4 exceptions. RWN sust be withdraw an out-of-operable per 3.3.8.3.b and sequence control rod

a second party verification after withdrawal of l may not be used in lieu of the first insequence RWN. control rod.

o

! c. Attempting to i withdraw a control rod more than one 1

notch prior to other

! control rod movement I after the group notch 'l j mode is automatically j initiated.

il j 3.a.2 The Rod Sequence Control

}

1 System (RCS) shall be 1 1

demonstrated to be l OPERABLE for a reactor shutdown by the following checks:

a. Performance of the comparator check of i the group notch ]

] circuits within 8 l hours prior to

]' automatic initiation of the group notch mode.

l l 8FN 3.3/4.3-6 Unit 3 Amendment No. 104 l

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3.3/4.3 mEACTIVITY CONTROL r

LINITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRENENTS 3.3.8. Control Rods ~

~

4.3.5. Control Rods -

3.a.2 (Cont'd) (

a <

' b. Attempting to insert a control rod more

' than one notch prior to other control rod movement after the

/

group notch mode is automatically initiated.

c. Selecting and attempting to move an out-of-sequence control rod after insertion of the first insequence control rod after reaching a black and white rod pattern.

3 b. Whenever the reactor is 3.b.1 The Rod Werth Minimizer in the startup or run modes (RWN) shall be demonstrated below 20% rated power, the to be OPERABLE for a Rod Worth Ninimizer (RWN) reactor startup by the shall be OPERABLE. With following checks:

the RWN INOPERABLE. verify control rod movement a. By demonstrating that and compilance with the the control rod prescribed control rod patterns and sequence pattern by a second input to the RWN licensed operator.or computer are correctly other technically loaded following any qualified member of the loading of the program l plant staff who is into the computer.

present at the reactor control console, b. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior Otherwise, control rod to withdrawal of l movement may be only by control rods for the l actuating the manual purpose of making the scram or placing the ,

reactor critical reactor mode switch in verify proper the shutdown position. annunciation of the J selection error of at least one out-of-sequence control rod.

BFN 3.3/4.3-7 Unit 3 Amendment No. 104

. 3.3/4.3 RL' ACTIVITY CONTROL LINITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.8. Cent ol Rods 4.3.8. Control Rods 3.b.1 (Cont'd)

c. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to withdrawal of control rods for the purpose of making the reactor critical, the rod block function of the RWN shall be verified by sofing an out-of-sequence control rod.

3.b.2 The Rod Worth Ninimizer (RWM) shall be demonstrated to be OPERABLE for a reactor shutdown by the following checks:

a. By demonstrating that the control rod _

patterns and sequence input to the RWN computer are correctly loaded following any i -

loading of the program into the computer.

b. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to RWN automatic initiation when i reducing thermal power,

! verify proper I

annunciation of the :

l selection error of at !

l least one out-of-sequence, control rod. i l '

c. Within one hour after RWN automatic l

initiation when reducing thermal power, the rod block function of the RWN shall be verified by moving an out-of-sequence control rod.

SFN 3.3/4.2-8 Unit 3 Amendment'~16.104

1 .

i

. 3.3/4.3 REACTIVITY CGuisGL LINITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRENENTS 3.3.8. Control Rods 4.3.8. Control Rods' ,

3.c. If Specifications 3.3.5.3.a 3.b.3 When the RWN is not through .b cannot be met the OPERABLE a second licensed operator.

reactor shall not be started. or other technically 4

or if the reactor is in the qualified member of j

run or startup modes at less f than 20% rated power, the plant staff shall )

control rod movement may be verify that the correct only by actuating the rod program is followed manual scram or placing except as specified in '

the reactor mode switch in 3.3.8.3.a.

the shutdown position.

4. Control rods shall not be 4. Prior to control rod withdrawn for startup or withdrawal for startup refueling unless at least or during refueling, two source range channels verify that at least two have an observed count rate source range channels equal to or greater than have an observed count three counts per second. rate of at least three counts per second.

5. During operation with 5. When a limiting limiting control rod control rod pattern patterns, as determined by exists, an instrument the designated qualified functional test of the i personnel, either; R8N shall be performed prior to withdrawal of

~l _

the designated rod (s)

a. Both R8N channels shall be operable: and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

or l

b. Control rod withdrawal shall be blocked.

l BFN 3.3/4.3-9 Amendment No. 104 Unit 3 i

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3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS .

3.3.C. Scram Insertion Times 4.3.C. Scram Insertion Times

1. The average scram 1. After each refueling insertion time, based on outage, all OPERABLE the doenergization of the rods shall be scram-time scram pilot valve sole- tested from the fully noids as time zero, of withdrawn position with all OPERABLE control rods the nuclear system in the reactor power pressura above 800 psig, operation condition shall This testing shall be be no greater than: completed prior to exceeding 40% power.

Below 20% power, only rods in those sequeness (Al2_and A34 or

% Inserted From . Avg. Scram Inser- Bi g and 834) which Fully Withdrawn tion Times (sec) were fully withdrawn in the region from 100%

5 0.375 rod density to 50% cod 20 0.90 density shall be 50 2.0 scram-time tested. The 90 3.5 sequence restraints imposed upon the control rods in the 100-50 percent rod density groups to the preset power level may be removed by use of the individual bypass switches associated with th,ose control rods which are fully or partially withdrawn and are not within the 100-50 percent rod e

density groups. In order to bypass a rod, the actual rod axial

position must be known; and the rod must be in the correct in-sequence position.

l l[!t3 3 3/* 3-10 Amendment No. 104

,-,..,,,.y----. . , - , - - - . . .

l .

. 3.3/4.3 REACTIVITY CONTROL i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRENENTS

~

3.3.C. Scram Insertion' Times 4.3.C. Scram Insertion Times The average of the scram inser- -2. At 16-week intervals, 10%

2.

tion times for the three fastest of the OPERABLE control-OPERABLE control rods of all rod drives shall be scram-groups of four control rods in timed above 800 psig.'

a two-by-two array shall be no Whenever such scram time measurements are made, an greater than:

evaluation shall be made

% Inserted From Avg. Scram Inser- to provide reasonable Fully Withdrawn tion Times (see) assurance that proper control rod drive 5 0.398 performance is being 20 0.954 maintained.

50 2.120 90 3.800

3. The maximum scram insertion time for 90% insertion of any OPERABLE control rod shall not exceed 7.00 seconds.

Reactivity Anomalies D. Reactivity Anomalies D.

The reactivity equivalent of During the startup test the difference between the program and startup following actual critical rod refueling outages, the configuration and the expected critical rod configurations configuration during power will be compared to the operation shall not exceed 1% at. expected configurations at If this limit is exceeded, the selected operating conditions.

reactor will be shut down These comparisons will be until the cause has been used as base data for determined and corrective reactivity monitoring during actions have been taken as subsequent power operation appropriate. throughout the fuel cycle. At specific power operating conditions, the critical rod configuration will be compared to the configuration expected based upon appropriately corrected past data. This comparison will be made at least every full power month.

BFN 3.3/4.3-11 Amendment No. 104 Unit 3 7 - , , . , . , - - _ - - - . - . - - - --

l 3.3/4.3 REACTIVITY CONT E i 9

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.E. If Specifications 3.3.C and 4.3.E. Surveillance requirements are 3.3.D above cannot be met, as specified in 4.3.C and an orderly shutdown shall be 4.3.D above.

initiated and the reactor shall be in the Shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l l

F. Scram Discharae Volume (SDV) F. Scram Discharae volume (SDV) 1.a. The scram discharge

1. The scram discharge volume drain and vent valves shall volume drain and vent be OPERAELE any time that valves shall be verified the reactor protection open prior to each system is required to be startup and monthly OPERABLE except as thereafter. The valves specified in 3.3.F.2. may be closed intermittently for testing not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in any 24-hour period during operation.

1.b. The scram discharge volume drain and vent valves shall be demonstrated OPERABLE monthly.

2. In the event any SD7 drain 2. When it is determined or vent valve becemes ,c. that any SD7 drain or IN3PERABLE, reactor vent valve is INOPER&BLE, opsration may continue the redundant drain or provided the redundant vent valve shall be drain or vent valve is demonstrated OPERABLE OPERABLE. Immediately and weekly thereafter.

3. If redundant drain or vent 3. No additional valves become INOPERABLE, surveillance required.

the reactor shall be in hot standby within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

J

!!Ye3 Amendment No. 104 1 -

i __. _ __ _ -_ _ ..

3.3/4.3 BASES A. Reactivity Limitation

1. The requirements for the control rod drive system have been identified by evaluating the need for reactivity control via control rod movement over the full spectrum of plant conditions and events. As discussed in subsection 3.4 of the Final Safety Analysis Rep. ort, the control rod system design is intended to provide sufficient control of core reactivity that the core could be made suberitical with the strongest rod fully withdrawn. This reactivity characteristic has been a basic i assumption in the analysis of plant performance. Compliance with this requirement can be demonstrated conveniently only at the time of initial fuel loading or refueling. Therefore, the demonstration must be such that it will apply to the entire subsequent fuel cycle. The demonstration shall be performed with the reactor core in the cold, zenon-free condition and will show that the reactor is suberitical by at least R + 0.38 percent Wk with the analytically determined strongest control rod fully withdrawn.

The value of "R", in units of percent Wk, is the amount by which the core reactivity, in the most reactive condition at any time in the subsequent operating cycle, is calculated to be greater than at the time of the demonstration. "R", therefore, is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of "R" must be positive or zero and must be determined for each fuel l cycle.

The demonstration is performed with a control rod which is calculated to be the strongest rod. In determining this " analytically strongest"

rod, it is assumed that every fuel assembly of the same type has

identical material properties. In the actual core, however, the control cell material properties vary within allowed manufacturing tolerances, and the strongest rod is determined by a combination of the control cell geometry and local k X. Therefore, an additional margin is included in the shutdown margin test to account for the fact that the rod used for the demonstration (the " analytically strongest")

is not necessarily the strongest rod in the core. Studies have been made which compare experimental criticals with calculated criticals.

These studies have shown that actual criticals can be predicted within a given tolerance band. For gadolinia cores the additional margin required due to control cell material manufacturing tolerances and calculational uncertainties has experimentally been determined to be 0.38 percent Wk. When this additional margin is demonstrated, it assures that the reactivity control requirement is met.

8FN 3.3/4.3-13 Unit ~3 Amendment No. 104

3.3/4.3 BASES (Cont'd) .

2. Reactivity margin - INOPERABLE control rods - Specification 3.3. A.2 requires that a rod be taken out of service if it cannot be moved with l drive pressure. If the rod is fully inserted and disarmed electrically *, it is in a safe position of maximum contribution to shutdown reactivity. If it is disarmed electrically in a nonfully inserted position, that position shall be consistent with the shutdown reactivity limitations stated in Specification 3.3.A.l. This assures that the core can be shut down at all times with the remaining control-rods assuming the strongest OPERABLE control rod does not insert.

Also if damage within the control rod drive mechanism and in

' particular, cracks in drive internal housings, cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled out. Circumferential cracks resulting from stress-assisted

. intergranular corrosion have occurred in the collet housing of drives at several BWRs. This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the affected rods.

Limiting the period of operation with a potentially severed rod after detecting one stuck rod will assure that the reactor will not be operated with a large number of rods with failed collet housings. The Rod Sequence Control System is not automatically bypassed until reactor power is above 20 percent power. Therefore, control rod movement is restricted and the single notch exercise surveillance test is only performed above this power level. The Rod Sequence Control System prevents movement of out-of-sequence rods unless power is above 20 percent.

8. Control Rods

1. Control rod dropout accidents as discussed in the FSAR can lead to i significant core damage. If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated. The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides a verification that the rod is following its drive.

Absence of such response to drive movement could ir.dicate an uncoupled condition. Rod position indication is required for proper function of the Rod Sequence Control System and the rod worth minimizer. -

To disarm the drive electrically, four amphenol type plus connectors are removed from the drive insert and withdrawal solenoids rendering the rod incapable of withdrawal. This procedure is equivalent to valving out the delve and is preferred because, in this condition, drive water cools and minimizes crud accumulation in the drive.

Electrical disarming does not eliminate position indication.

l BFN 3.3/4.3-14 Unit 3 Amendment No. 104

l 1.

. 3.3/4.3 BASES (Cont'd)

2. The control rod housing support restricts the. outward movement of a. ~

-we ~ . control rod to less than three inches in the extremely remot'e event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system. The design basis is given in subsection 3.5.2 of the FSAR and the safety evaluation is given in subsection 3.5.4.

This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to j

rapidly eject a drive housing. Additionally, the support is not r required if all control rods are fully inserted and if an adequate shutdown margin with one control rod withdrawn has been demonstrated, {

since the reactor would remain suberitical even in the event of complete ejection of the strongest control rod. I

3. The Rod Worth Minimizer (RWM) and the Rod Sequence Control System (RSCS) restrict withdrawals and insertions of control rods to prospecified sequences. All patterns associated with these sequences have the characteristic that, assuming the worst single deviation from the sequence, the drop of any control rod from the fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in any pellet average

' enthalpy in excess of 280 calories per gram. An enthalpy of 280 calories per gram is well below the level at which rapid fuel-dispersal could occur (i.e., 425 calories per gram). Primary system damage in this accident is not possible unless a significant maount of fuel is rapidly dispersed. Reference Sections 3.6.6, 7.7.A, 7.16.5.3, and 14.6.2 of the FSAR, and NEDO-10527 and supplements thereto.

In performing the function described above, the RWN and RSCS are not required to impose any restrictions at core power levels in excess of 20 percent of rated. Material in the cited reference shows that it is

' impossible to reach 280 calories per gram in the event of a control rod drop occurring at power greater than 20 percent, regardless of the rod pattern. This is true for all normal and abnormal patterns including those which maximize individual control rod worth.

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i I

BFN 3.3/4.3-15 Amendment No. 104 Unit 3 i - .- -.

3.3/4.3 BASE.S,(Cont'd)

ES .

At power levels below 20 percent of. rated. abnorr.al control rod ,

patterns could produce rod worths high enough to be of concern relative to the 280 calorie per gram rod drop limit. In this range the RWN and the RSCS constrain the control rod sequences and patterns to those which involve only acceptable rod worths.

I The Rod Worth Minimizer and the Rod Sequence Control System provide .

automatic supervision to assure that out of sequence control rods will ,

not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences. Reference Section 7.16.5.3 of the 4 FSAR. They serve as a backup to procedure control of control rod :

l sequences, which limit the maximum reactivity worth of control rods.

Except during specified exceptions, when the Rod Worth Minimizer is out of service a second licensed operator can manually fulfill the control rod pattern conformance functions of this system. In this case, the RSCS is backed up by independent procedural controls to assure conformance.

The functions of the RWN and RSCS make it unnecessary to specify a license limit on rod worth to preclude unacceptable consequences in the event of a control rod drop. At low powers, below 20 percent, these devices force adherence to acceptable rod patterns. Above 20 percent of rated power, no constraint on rod pattern.is required to assure that rod drop accident consequences are acceptable. Control rod pattern constraints above 20 percent of rated power are imposed by power distribution requirements, as defined in Sections 3.5.I. 3.5.J.

4.5.I. and 4.5.J of these technical specifications. Power level for automatic bypass of the RSCS function is sensed by first stage turbine

^

pressure. Because the instrument has an instrument error of 310 percent of full power the nominal instrument setting is 30 percent i of rated power.

Because it is allewable to bypass certain rods in the RSCS during sceam-time testing below 20 percent of rated power in the startup or run modes, a second licensed operator-is not an acceptable substitute for the RWN during this testing.

a. The Source Range Monitor (SRN) system performs no automatic safety i system functien; i.e., it has no scram function. It does provide the l

operator with a visual indication of neutron level. The consequences ;

of reactivity accidents are functions of the initial neutron flux. l The requirement of at least three counts per second assures that any transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions. One OPERA 8LE SRN channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two OPERABLE SRMs are provided as an added conservatism.

i 8FN 3.3/4.3-16 unit 3 Amendment No. 104 l

9 3.3/4.3 BASES (Cont'd)

5. The Rod Block Monitor (RBN) is designed to automatically prevent fuel

. damage in the event of erroneous rod withdrawal 'fran'locati'ons of high power density during high power level operation." Two R8N channels are ,

provided, and one of these may be bypassed from the console for maintenance and/or testing. Automatic rod withdrawal blocks from one of the channels will block erroneous-rod? withdrawal soon edssgh to prevent fuel damage. The specified restrictions with one channel'out' of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.

A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit (i.e., NCPR = *** or LNGR = 13.4).

During use of such patterns, it is judged that testing of the R8N system prior to withdrawal of such rods to assure itsItoperability is normally will assure that improper withdrawal does not occur.

the responsibility of the nuclear engineer to identify these limitin5 patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of INOPERABLE control rods in other than limiting patterns. Other personnel qualified to perform these functions may be designated by the plant superintendent to perform these functions.

C. Scram Insertion Times The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from-becoming less than 1.07. Analysis of this transient shows that the negative reactivity rates resulting from the scram (FSAR Figure N3.6-9)

with the average response of all the drives as given in the above specification, provide the required protection, and NCPR remains greater than 1.07.

On an early BWR, some degradation of control rod scram performance occurred during plant startup and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter. The design of the present control rod drive (Model 7RDB144B) is grossly improved by the relocation of the filter to a j location out of the scram drive path; 1.e., it can no longer interfere j

with scram performance, even if completely blocked.

The degraded performance of the original drive (CRD7RD8144A) under dirty

! operating conditions and the insensitivity of the redesigned drive (CRD7RD81448) has been demonstrated by a series of engineering tests under simulated reactor operating conditions. The successful performance of the new drive under actual operating conditions has also been demonstrated by consistently good in-service test results for plants using the new drive and may be inferred from plants using the older model

- - See Section 3.5.K l

l i

RFN 3.3/4.3-17 Amendment No. 104 U2it 3

i s 3.3/4.3 BASES (Cont'd) .

drive with a modified (larger screen size) internal filter which is less .

prone to plugging. Data has been documented by surveillance reports in various operating plants. These include Oyster Creek, Monticello, Dresden 2, and Dresden 3. Approximately 5000 drive tests have been recorded to date.

Following identification of the " plugged filter" problem, very frequen.t scram tests were necessary to ensure proper performance. However, the more frequent scram tests are now considered totally unnecessary and unwise for the following reasons:

1. Erratic scram performance has been identified as due to an obstructed drive filter'in type "A" drives. The drives in BFNP are of the new "B" type design whose scram performance is unaffected by filter condition.

2. Th,e dirt load is primarily released during startup of the reactor when the reactor and its systems are first subjected to flows and pressure and thermal stresses. Special attention and measures are now being taken to assure cleaner systems. Rosetors with drives identical or similar (shorter stroke, smaller piston areas) have operated through many refueling cycles with no sudden or erratic changes in scram performance. This preoperational and startup testing is sufficient to detect anomalous drive performance.

3. The 72-hour outage limit which initiated the start of the frequent scram testing is arbitrary, having no logical basis other than quantifying a " major outage" which might reasonably be caused by an event so severe as to possibly affect drive performance. This requirement is unwise because it provides an incentive for shortcut actions to hasten returning "on line" to avoid the additional testing

a. due a 72-hour outage.

I I

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l BFN 3.3/4.3-18 Unit 3 Amendment No. 104 l

t

i-L .

, e

3.3/4.3 BASES (Cont'd)

. The surveillance requirement for scram testing.of all the control rods

'after each refueling outage and 10 percent of the control. cods at 16-week

'~~~-eintervals is adequate for determining the operability of the control rod - -

~.

system yet is not'so fraquent as to cause excessive wear on the control i

rod system components.

T5e~numericalvaluesassignedtotheTridicted'scramperformancearebased on the analysis of data from other BWRs with control rod drives the same as those on Browns Ferry Nuclear Plant.

The occurrence of scram times within the limits, but significantly longer l than the average, should be viewed as an indication of systematic problem-with control rod drives especially if the number of drives exhibiting such scram times exceeds eight, the allowable number of INOPERABLE rods.

i j

In the analytical treatment of the transients which are assumed to scram '

on high neutron flux, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of control rod motion.

This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results.

Approximately the first 90 milliseconds of each of these time intervals result from the sensor and circuit delays after which the pilot scram solenoid doenergizes and 120 milliseconds later, the control rod motion is estimated to actually begin. However, 200 milliseconds, rather than 120 milliseconds, are conservatively assumed for this time interval in the transient analyses and are also included in the allowable scram insertion times of Specification 3.3.C.

-In order to perform scram time testing as required by Specification 4.3.C.1, the relaxation of certain restraints in the rod sequence control system is required. Individual rod bypass switches may be used as described in Specification 4.3.C.1.

The position of any rod bypassed must be k.nown to be in accordance with ,

rod withdrawal sequence. Bypassing of rods in the manner described in 5 Specification 4.3.C.1 will allow the subsequent withdrawal of any rod scrammed in the 100 percent to 50 percent rod density groups; however, it ,

will maintain group notch control over all rods in the 50-percent to ,

0-percent rod density groups. In addition, RSCS will prevent movement of ,

! rods in the 50-percent density to a preset power level range until the scrammed rod has been withdrawn.

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l BFN 3.3/4.3-19 Amendment No.104 Unit 3

- - _ _ - - - - - _ =

i 3.3/4.3 BASES .

D. Reactivity Anomalies During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavi,or in the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons.

Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1 percent Wk. Deviations in core reactivity greater than 1 percent Wk

, are not expected and require thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.

E. No BASES provided for this specification F. Scram Discharge Volume The nominal stroke time for the scram discharge volume vent and drain valves is 5 30 seconds following a scram. The purpose of these valves is to limit the quantity of reactor water discharged after a scram and no direct safety function is performed. The surveillance for the valves assures that system drainage is not i.peded i by a valve which fails to open and that the valves are OPERABLE and capable of closing upon a scram.

References

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

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) BFN 3.3/s.3-20 i

Unit 3 Amendment No.104

- -

v t e Letter Sequence Other
TAC:61960, (Approved, Closed) TAC:61961, (Approved, Closed) TAC:61962, (Approved, Closed)
Initiation Acceptance... Administration Questions Answers Results Approval, Approval Other: ML20214H851
MONTHYEAR1987-05-13 [Table View]

ML20214H851

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