ML20059H633
| ML20059H633 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/11/1990 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059H636 | List: |
| References | |
| NUDOCS 9009180128 | |
| Download: ML20059H633 (43) | |
Text
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., is A.
UNITED STATES j
8 NUCLEAR REGULATORY COMMISSION
-!s WASHINGTON, D, C. 20666
- g CAROLINA POWER & LIGHT COMPANY. et al. -
i DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No.144 License No..nPR-71' 1.
The Nuclear Regulatory Comission (the Comission) has found;that:
A.
The application for amendment filed by Carolina Power & Light Company (the licensee), dated March 14,:1990,- as supplemented August 9 and 29, 1990, 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 (1) 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 L
public; and i
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-71 is hereby amended to read as follows:
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'(2) lechnical Specifications l
The Technical Specifications contained in ApperdicesLA and B, as revised through Amendment No.144, are hereby incorporated 'in the -
license.
Carolina Power & Light Company shall, operate the facility in accordance with the Technical Specificatio V.
j 3.
This license amendment is effective as of the dste of its issuance and shall be implemented within 60 days of issuance.
.l FOR THE NUCLEAR REGULATORY COMMISSION LI Original' Signed.By:
Elinor G.-Adensam,' Director i
Project Directorate II I Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the-Technical Specifications 4
1 Date of Issuance: September 11, 1990 y
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A,TTACHMENT TO LICENSE AMENDMENT NO. 144 FACILITY OPERATING LICENSE NO. OPR-71 DOCKET NO. 50-325 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marginal lines..
Remove Pages
. Insert Pages IV IV IX IX XII XII 3/4 1 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-7 3/41-7 3/4 1-8 3/4 1-8 3/4 1-9 3/4 1-9 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12 3/4 1-14 3/4 1-14 3/4 1-14a 3/4 1-15 3/4 1-15 3/4 1-16 3/4 10-2 3/4 10-2 B 3/4 1-3 8 3/4 1-3 8 3/4 1-4 8 3/4 1-4 8 3/4 1-5 8 3/4 1 B 3/4 10-1 8 3/4 10-1 l
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INDEX LIMITING CONDITIONS PC. OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PACE 3/4.0 A P P LI C A B I L I TY..............................................
3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS I
3/4.1.1 SHUTDOWN MARCIN..........................................
'3/4 1-1 3/4.1.2 REACTIVITY AN0 MAL lES.....................................
3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability..................................
3/4 1-3 Control Rod Maximum Scram Insertion Times................
3/4 1-5 Control Rod Ave rage Scram Ins erti on Times................ '3/4 1-6 Four Cont rol kud Group Sc ram Inse rt ion Times.............
3/4 1-7 control Rod Scram Accumulators...........................
3/4 1-8 Con t r ol Rod D ri ve Coupl ing...............................
3/4 1-9 Control Rod Position Indication..........................
3/4 1-11 Control Rad Drive Housing Support........................
3/4 1-13 l
3/4.1.4 CONTROL RM PROGRAM CONTROLS Rod Worth Minimizer......................................
3/4 1-14 I
Rod Sequenc e Cont ro l Syst em ( DELETED)....................
3/4 1-15 l
Rod Block Monitor........................................
3/4 1-17
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3/4.1.5 STAND BY L IQUI D CO NTRO L SYSTEM............................
3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.i AVERACE PLANAR LINEAR HEAT CENERATION RATE...............
3/4 2-1 3/4.2.2 APRM SETP01NTS...........................................
3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATI0.............................
3/4 2-3 BRUNSWICK - UNIT 1 IV Amendment No.
23, 24, 56, 124, 131, 14
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o INDEX LIMITINC CONDITIONS FOR OPERATION AND SURVEILLANCE REQUl?EMENTS SECTION PnCE j
3/4.9 REFUELINC OPERATIONS (Continued) ~
3/4.9.3 CONTROL ROD POSIT10N..................................
3/4 9-5 3/4.9.4 DECAY TIME............................................
3/4 9-6 3/4.9.5 C OM MU N I C A TI ON S........................................
3/4 9-7 3/4.9.6 C RANE AND HOI ST OPE RABI LI TY...........................
3/4 9-8 3/4.9.7 CRANE TRAVELS-SPENT FUEL STORACE P00L..................
3/4 9-9 3/4.9.8 WATER LEVEL-REACTOR VESSEL.............................
3/4 9-10 3/4.9.9 WATER LEVEL-SPENT FUEL STORACE P00L...................
3/4 0-11 3/4.9.10 CONTROL ROD REMOVAL Single Control Rod Removal............................
3/4 9-12 f
Multiple Control Rod Removal..........................
3/4 9-14 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY.........................
3/4 10-1 3/4.10.2 ROD S EQUENCE CONTROL SYSTEM ( DELETED).................
3/4 10-2 l
3/4.10.3 SHUTDOWN MARCIN DEMONSTRATIONS........................
3/4 10-3 3/4.10.4 RECIRCULATION L00PS...........-.......................
3/4 10-4 3/4.10.5 P LANT S ER V I C E W AT ER...................................
3/4 10-5 I
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1 1 BRUNSWICK - UNIT 1 IX Amendment No. [2, 2 2, 3(,
133. 14t m
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INDEX f
BASES i
SECTION P_AC E A
3/4.7 PLANT SYSTEMS (Continued) 3/4.7.3 F LOOD P ROTE CT 10N e.....................................
B 3/4 7-1 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM..................
B 3/4 7-1 3/4.7.5 SNUBBERS...............................................
B 3/4 7 i 3/4.7.6 SEALED SOURCE CONTAMINATION............................
B 3/4 7-3 3/4.7.7 FIRE SUPPRESSION SYSTEMS...............................
B 3/4 7-3 i
3/4.7.8 FIRE BARRIER PENETRATIONS..............................
B 3/4 7-3 t
3/4.8 ELECTRICAL POWLR SYSTEMS...................................
B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH....................................
B 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................
B 3/4 9-1 3/4.9.3 CONTROL ROD POSIT 10N..................................
B 3/4 9-1 3/4.9.4 DECAY TIME............................
B 3/4 9-1 3/4.9.5 COMMUNICATIONS........................................
B 3/4 9-1 3/4.9.6 CRANE AND Hol ST OPERABI LI TY...........................
B 3/4 9-2 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORACE P00L.................
B 3/4 9-2 l
3/4.9.8 WATER LEVEL-REACTOR VESSEL, and 3/4.9.9 WATER LEVEL-REACTOR FUEL STORACE P00L.................
B 3/4 9-2 3/4.9.10 CON TR OL R OD R EMOV AL...................................
B 3/4 9-2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIKARY CONTAlWHENT INTECRITY.........................
B 3/4 10-1 3/4.10.2 ROD S EQUENCE CONT ROL SYSTEM ( DELETED).................
B 3/4 10-2 l
3/4.10.3 SHUTDOWN HARCIN DEMONSTRATIONS.......................-
B 3/4 10-3 3/4.10.4 R EC I R CU LATI ON L00 PS...................................
B 3/4 10-4 3/4.10.5 P LAN1 S ERVI CE W AT ER...................................
B 3/4 10-5 BRUNSWICK - UNIT 1 X11 Amendment No.
gg, fgg, 144 j
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REACTIVITY CONTROL SYSTEMS LIMITINC CONDITION FOR OPERATION (Continued)
ACTIONI (Continued) l 2.
If the inoperable control rod (s) is insertedt a)
Within one hour disarm the associated directional control valves either 1)
Electrically, or 2)
Hydraulically by closing the drive water and exhaust water isolation valves.
b)
Otherwise, be in at least HOT SHUTDOWN within the next 12 hou s.
i With more than 8 control rods inoperable, be in at least HOT SHUTDOWN c.
within t2 hours.
SURVEILLANCE REQUIPEMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE at least once per 31 days byt*
Verifying each valve to be open.
a.
b.
Cycling each valve at least cne complete cycle of full travel.
4.1.3.1.2 All withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notcht a.
At least once per 7 aays when above the preset power level of the RWH and l
b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when above the preset power level of the l
RWM and any control rod is immovable as a result of excessive friction or mechanical interference.
4.1.3.1.3 All withdrawn control rods shcIl be determined OPERABLE by demonstrating the scram discharge volume drain and vent valver OPERABLE, when the reactor protection system logic is tested per Specification 4.3.1.2, by verifying that the drain and vent valves Close within 30 seconds after receipt of a signal for control rods to a.
scram, and b.
Open when the scram signal is reset or the scram discharge volume l
trip is bypassed.
- These valves may be closed intermittently for testing under administrative control.
BRUNSWICK - UNIT 1 3/4 1-4 Amendment No. 48, 51, 144 l
s REACTIVITY CONTROL SYSTEMS CONTROL ROD HAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 6, based on de-energitation of the scram' pilot valve solenoids as time zero, shall not exceed 7.0 seconds.
APPLICABILITY:
CONDITIONS 1 and 2.
ACTION With the maximum scram insertion time of one or more control rods exceeding 7.0 seconds, operation may continue and the provisions of Specification 3.0.4 are not applicable provided thatt The control rod with the slow insertion time is declared inoperable, a.
b.
The requirements of Specification 3.1.3.1 are satisfied and c.
If within the preset power level of the RWH, the requirements of Specification 3.1.4.1.d are also satisfied, and d.
The Surveillance Requirements of Specification 4.1.3.2.c are l
performed at least once per 92 days when operation is continued with three or more control rods with slow scram insertion timest Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demonstrated through measurement:
For all control rods prior to THERMAL POWER exceeding 40% of RATED a.
TSERMAL POWER f ollowing CORE ALTERATIONS or siter a teactor shutdown that is greater than 120 days, b.
For specifically af fected individual control rods following maintena6ce on or modification to the control rod or rod drive system which could af fect the scram insertion time of those specific control rods, and For 10% of the control rods, on a rotating basis, at least once per j
c.
120 days of operation.
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BRUNSWICK - UNIT 1 3/4 1-5 Amendment No.
144 i.
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t REACTIVITY CONTROL SYSTEMS FOUR CONTROL ROD CROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time, f rom the fully withdrawn position, for the three fastest control rods in each group of four control rods arranged in a two-by-two array, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed any of the following Position Inserted From Average Scram inser-l Fully Withdrawn tion Time (Seconds) 46 0.33 36 1.12 26 1.93 6
3.58 i
APPLICABILITY OPERATIONAL CONDITIONS 1 and 2.
ACTION With the average scram insertion times of control rods exceeding the above-limits, operation may continue and the provisions of Specification 3.0.4 are not applicable providedt The control rods with the slower than average scram insertion times a.
are declared inoperable, b.
The requirements of Specification 3.1.3.1 are satisfied, and c.
It within the preset power level of the RWH, the requirements of Specification 3.1.4.1.d are also satisfied, and d.
The Surveillance Requirements of Specification 4.1.3.2.c are l
performed at least once per 92 days when operation is continued with three or more control rods with slow scram insertion' times.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.4 All control rods shall be demonstrated OPERABLE by scram time testing f rom the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.
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BRUNSWICK - UNIT 1 3/4 1 7 Amendment No.
56, 144 1
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e REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS LIMITINC CONDITION FOR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and $*.
ACTION:
a.
In 0." ERAT!0NAL CONDITION 1 or 2 with one control rod scram accumulator inoperable, the provisions of Specification 3.0.4 are not applicable and operation may continue, provided that within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />st 1
1.
The inoperable accumulator is restored to OPERABLE status, or 2.
The control rod associated with the inoperable accumulator is declared inoperable, and the requirements of Specification 3.1.3.1 are satisfied.
3.
And, if within the preset power level of the RWM, the requirements of Specification 3.1.4.1.d are also satisfied.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In >PERATIONAL CONDITION 5* with a withdrawn control rod scram accumulator inoperable, fully insert the affected control rod and electrically disarm the directional control valves within one hour.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.1.3.5 The control rod scra'm accumulators shall be determined OPERABLE:
a.
At least once per 7 days by verifying that the pressure and leak detectors are not in the alarmed condition, and b.
At least once per 18 months by performance of at 1.
CHANNEL FUNCTIONAL TEST of the leak detectors, and i
2.
CHANNEL CALIBRATION of the pressure detectors.
- At least the accumulator associated with each withdrawn control rod. Not applicable to cont rol rods removed per Specification 3.9.10.1 or 3.9.10.2.
i BRUNSWICK - UNIT 1 3/4 1-B Amendment ' No. 33, f 30, 14 4
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1 REACTIVITY CONTROL SYSTEMS I
COWTROL ROD DR)VE COUPLINC
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LIMITINC CONDITION FOR OPERATION 3.1.3.6 All control rods shall be coupled to their drive mechanisms.
APPLICABILITY:
CONDITIONS 1, 2, and $*.
ACTION In CONDITION 1 or 2 with one control rod not coupled to its a.
associated drive mechanism, the provisions of Speeltication 3.0.4 are not applicable and operation may continue provided 1.
Within the preset power level of the RWM, the control rod is l
declared inoperable and fully inserted until recoupling can be at tempted with THERMAL POWER above the preset power level of the RWM and the requirements of Specification 3.1.4.1.d are l
satisfied.
1 2.
Above the preset power level of the RWM, the control rod drive l
is inserted to accomplish recoupline.
If recoupling is not accomplished on the first attempt, declare the control rod inoperable, fully insert the control rod, and electrically disarm the directional control valves.
3.
The requirements of Specification 3.1.3.1 are satistled.
b.
In CONDITION $*, witt. a withdrawn control rod not coupled to its associated drive mechanism, insert the control rod to accomplish recoupling. The provisions of Specification 3.0.3 are not applicable.
l SURVEILLANCE REQUIREMENTS 4.1.3.6 The coupling integrity of a control rod shall be demonstrated by withdrawing the control rod to the f ully withdrawn position and verifying that the rod does not go to the overtravel positions Prior to reactor criticality af ter completing CORE ALTERATIONS that a.
could have affected_the control rod drive coupling integrity,
- At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
BRUNSWICK - UNIT 1 3/4 1-9 Amendment No.
144
6 REACTIVITY CONTROL SYSTEMS CONTROL ROD POSITION INDICATION LIMITINC CONDITION FOR OPERATION 3.1.3.7 All control rod rhed switch position indicators shall be OPERABLE.
APPLICABILITY:
CONDITIONL 1, 2, and $*.
ACTION:
a.
In CONDITION 1 or 2: With one or more control rod reed switch l
position indicators inoperable, including " Full-in" or " Full-out" indication, the provisions of Specification 3.0.4 are not applicable and operation may continue, provided that within one hourt 1)
The position of the control rod is determined by an l
alternate method, or 2)
The control rod is moved to a position with an OPERABLE l
reed switch position indicator, or 3)
The control rod with the inoperable reed switch position l
indicat or is declared inoperable and the requirements of Specification 3.1.3.1 are satisfiedi 4)
And, if within the preset power level of the RWH, the requirements of Specification 3.1.4.1.d are satisfiedi Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
I b.
In CONDITION 5* with a withdrawn control rod reed. switch position indicator inoperable, fully insert the withdrawn control rod.
The provisions of Specification 3.0.3 are not applicable.
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- At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
BRUNSW10K - UNIT 1 3/4 1-11 Anendment No.14 4
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REACTIVITY CONTROL SYSTEMS' f
i SURVEILLANCE REQUIREMENTS l.
I 4.1.3.7 The control rod reed switch position indicators shall be determined
[
OPERABLE by verifyingt i
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, that the position of the control rod is a.
i indicated, b.
That the indicated control. rod position changes duringc the movement.
of the control rod when performing Surveillance Requirement 4.1.3.1.2, and That the control rod reed switch position indicator corresponds to
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c.
the control rod position Indicated by the " Full-out" reed switches
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when perf orming Surveillance Requirement 4.1.3.6.b.
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BRUNSWICK - UNIT 1 3/4 1-12 Amendment No. 76. 144' f
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e REACTIVITY CONTROL SYSTEMS 3/4 1.4 CONTROL ROD PROCR AM CONTROLS ROD WORTH MINIMi2ER LIMITING CONDITION FOR OPERATION 3.1.4.1 The Rod Worth Minimiter (RWM) shall be OPERABLE when THERKAL POWER is less than 10% af RATED THERMAL POWER.
l APPLICABILITY OPERATIONAL CONDITIONS 1 and 2*.
i ACTION:
With the RWM inoperable after the first 12 control rods have been a.
f ully withdrawn on a startup, operation may continue provided that control rod movement and compliance with the prescribed BPWS control rod pattern are verified by a second licensed operator or qualified member of the plant technical staf f.
b.
With the RWM inoperable before the first 12 control rods are withdrawn on a startup, one startup per calender year may be performed provided thae control roa movement and compliance with the prescribed BPWS contrcl rod pattern are verified by a second licensed operator or qualified member of the plant technical staff.
With RWM inoperable on a shutdown, shutdown may continue provided c.
that control rod movement and compliance with the prescribed BPWS control rod pattern are verified by a second licensed operator or qualified member of the plant technical staff.
d.
With RWM operable but individual control rod (s) declared inoperable, operation and centrol rod movement below the present power level of the RWM may continue providedt l.
No more than three (3) control rods are declared inoperable in any one BWS group, and, 2.
The inoperable control rod (s) is bypassed on the RWM and control rod movement of the bypassed rod (s) is verified by a second licensed operator or qualified member of the plant technical
- staff, With RWH inoperable, the provisions of Specification 3.0.4 are not e.
applicable.
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
BRUNSWICK - UNIT 1 3/4 1-14 Amendment No.
727, 144
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REACTIVITY CON'.ROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.4.1.1 The RWH shall be demonstrated OPERABLE in OPERATIONAL CONDITION 2, prior to withdrawal of control rods for the purpose of making the reactor critical and in OPERATIONAL CONDITION 1 when the RWM is initiated during control rod insertion when reducing tilERHAL POWER byt Verif ying proper annunciation of the selection error of at least one a.
out-of-sequence control rod, and b.
Verif ying the rod block function of the RWM by moving an out-of-sequence control rod.
4.1.4.1.2 The RWM shall be demonstrated OPERABLE by verifying the control rod Banked Position Withdrawal Sequence input to the RWM computer is correct following any loading af the sequence program into the computer.
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t BRUNSWICK - UNIT 1 3/4 1-14a Amendment No. 14 4 l
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i REACTIVITY CONTROL SYSTEMS ii ROD SEQUENCE CONTROL SYSTEM j
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Pages 3/4' l-15 through 3/41-16 have been deleted.
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l (Next.page is 3/4 1-17) l BRUNSWICK - UNIT 1 3/4 1 15 -
Amendment No.- 144 j
SPECIAL TEST EXCEPTIONS
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3/4.10.2 ROD SEQUENCE CONTROL SYSTEM i
e Page 3/4 10-2 has been deleted.
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i BRUNSWICK - UNIT 1 3/4 10-2 Amendment' No. 22, ' 14 4 I
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REACTIVITY CONTROL SYSTEM
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BASES l
CONTROL RODS (Continued) on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure.
Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactors.
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in tha FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and, therefore, this check must be performed prior to achieving criticality after each refueling. The subsequent check is performed as a backup to the initial demonstration.
i In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE.
The cont rol rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure.
The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 CONTROL ROD PROCRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum in sequence individual control rod or control rod segments i
which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than or equal to 10% of RATED THERMAL POWER, there is no possible l
rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus, requiring the RWM to be OPERABLE when THERMAL POWER is less than 10% of RATED THERMAL POWER provides adequate control.
Use of the Banked Position Withdrawal Sequence (BPWS) ensures that in the event of a control rod drop accident the peak fuel enthalpy will not be greater than 280 cal /gm (Ref erence 4).
BRUNSWICK - UNIT 1 B 3/4 1-3 Amendment No.
f27, 144
REACTIVITY CONTROL SYSTEM BASES CONTROL ROD PROCRAM COWTROLS (Cont inued)
The RWM as a backup to procedural control provides an automatic control rod pattern monitoring function to ensure adherence to the BPWS control movement sequences f rom 100% control rod density to 10% RATED TilERMAL POWER and, thus, eliminates the postulated control rod drop accident from resulting in a peak fuel enthalpy greater than 280 cal /gm (Reference $).
The requirement that RWH be operable for the withdrawal of the first 12 control rods on a startup is to ensure that the RWM system maintains a high degree of availability.
Deviation f rom the BPWS control rod pattern may be ' allowed for the performance of Shutdown Margin Demonstration tests.
The analysis of the rod drop accident is presented in Section 15.4.6 of the Updated FSAR and the techniques of the analysis are presented in a topical l
report (Reference 1) and two supplements (References 2 and 3).
The RBH is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent f uel damage. This system backs up the written sequence used by the operator for withdrawal of control rods.
3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for maintaining the reactor suberitical in the event that insuf ficient rods are inserted in the core when a scram is called for. The volume and weight percent of poison material in solution is based on being able to bring the reactor to the suberitical condition as the plant cools.to ambient condition. The temperature requirement is necessary to keep the sodium pentaborate in solution. Checking the volume and tem hours assures that the solution is available for use.perature once each 24 With redundant pumps and a highly reliable control rod scram system operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods.of time with one of the redundant components inoperable.
Surveillance requirements are established on a f requency that assures a high reliability of the system. Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not f ail because of detertoration of the charges.
BRUNSWICK - UNIT 1 B 3/4 1-4 Amendment No. 77/,144 l
c
e REACTIVITY CONTROL SYSTEM BASES References 1.
C. J. Paone, R. C. Stirn, and J. A. Woodley, " Rod Drop A..ident Analysis for Large BWRs " C. E. Topical Report NEDO-10527, March 1972.
2.
C. J. Paone, R. C. Stirn, and R. M. Yound, Supplement I to NEDO-10527, July 1972.
3.
J. A. Haum, C. J. Paone, and R. C. Stirn, addendum 2 " Exposed Cores" supplement 2 to NEDO-10527, January 1973.
4.
NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel,"
Revision 6, Amendment 12.
5.
NEDE-20411-P-A, " General Electric Standard Application f or Reactor Fuel,"
Revision 8 Amendment 17.
BRUNSWICK - UNIT 1 B 3/4 1-5 Amendment No. 144
{
o.
3/4.10 SrECIAL TEST EXCEPTIONS BASES 3/4.10.1 ' PRIMARY CONTAINMENT INTECRITY l
The requirement for PRIKARY CONTAINMENT INTECRITY is removed during the period when open vessel tests are being perf ormed during low power PHYSICS TESTS.
3/4.10.2 ROD SEQUENCE CONTROL SYSTEM (DELETED) t 3/4.10.3 SHUTDOWN MARCIN DEMONSTRATIONS f
Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur. These additional restrictions are specified in this LCO.
3/4.10.4 RECIRCULATION LOOPS i
This special test exception permits reactor criticality under no flow conditions and is required to perf orm certain start-up and PHYSICS TESTS while at low THERMAL POWER levels.
3/4.10.5 PLANT SERVICE WATER This Special Test Exception permits securing the Service Water System conventional header when the nuclear header is out of service and is required to permit flange installation in service water system header cross-connect piping.
BRUNSWICK - UNIT 1 B 3/4 10-1 Amendment No. 22, gg,
36, 144 1
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NUCLEAR REGULATORY COMMISSION p!
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l CAROLINA POWER 4 LIGHT COMPANY. et al.
DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.175 License No. DPR-62 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment filed by e.arolina Power & Light Company (the licensee), dated March 14, 1990, as supplemented August 9 and 29, 1990, 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 (1) that the activities authorized 1
by th*s amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common
' 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; andparagraph2.C.(T)ofFacilityOperatingLicenseNo.DPR-62is hereby amended to read as follows:
1
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i n
(2) Technical Specifications The Technical Specifications contained in Appendices A and B as revised through Amendment No.175. are hereby incorporated in the license.
Carelina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
i 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By:
Elinor G. Adensam, Director Project Directorate 11-1 Division of Reactor Projects I/II i
Office of Nuclear Reactor Regulation Attachment Changes tt. a Technical Specifications Date of Issuance:
September 11, 1990 i
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(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.175, are hereby incorporated in the license.
Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By:
Elinor G. Adensam, Director Project Directorate 11-1 Division of Reactor Projects I/fl Office of Nuclear Reactor Reguli tion
Attachment:
Changes to the Technical i
Specifications Date of Issuance:
September 11, 1990 IM OFC : LA: P t
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ATTACHMENT TO LICENSE AMENDMENT NO. 175 FACILITY OPERATING LICENSE NO. 0PR-62 j
DOCKET NO. 50-324 Replace the following pages of the Appendix A Technical Specifications with j
the enclosed pages.
The revised areas are indicated by marginal lines.
Remove Pages Insert Pages IV IV IX IX XII XII 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-7 3/4 1-7 3/4 1-8 3/4 1-B 3/4 1-9 3/4 1-9 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12 3/4 1-14 3/4 1-14 3/4 1-14a 3/4 1-15 3/4 1-15 3/4 1-16 3/4 10-2 3/4 10-2 B 3/4 1-3 8 3/4 1-3 8 3/4 1-4 B 3/4 1-4 8 3/4 1-5 8 3/4 1-5 B 3/4 10-1 B 3/4 10-1 I
9 9
I
i 1
INDEX l
LIMITINC CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PACE 3/4.0 A P P L I C A B I L I TY..............................................3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARC1N..........................................
3/4 1-1 3/4.1.2 R EA CT I V I T Y AN 0 MALI E S.....................................
3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operab(11ty..................................
3/4 1-3 Cont rol Rod Max in.um Scram Ins ertion Times................
3/4 1-5 Control Rod Average Scram Insertion Times................
3/4 1-6 Four Cont rol Rod Group Sc ram Inse rt ion Times.............
3/4 i,-7 Control Rod Scram Accumulators...........................
3/4 1-d Con t rol Rod D ri ve Cou pl ing...............................
3/4 1-9 Control Rod Position Indication..........................
3/4 1-11 Control Rod Drive Housing Support........................
3/4 1-13 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer......................................
3/4 1-14 Rod Sequence Control System (DELETED)....................
3/4 1-15 l
I I:od Block Honitor........................................
3/4 1-17 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................
3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE...............
3/4 2-1 3/4.2.2 MINIMUM CRITI CAL POWER RAT 10.............................
3/4 2-2 i
l BRUNSWICK - UNIT 2 IV Amendment No. gg, f(g, fgf 168, 175
,INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS j
1 SECTION PACE 3/4.9 REFUELINC OPERATIONS (Continued) 3/4.9.3 CONTROL ROD POSIT 10W..................................
3/4 9-5 3/4.9.4 DECAY TIME............................................
3/4 9-6 3/4.9.5 COHHUNICAT!0NS........................................
3/4 9-7 3/4.9.6 C RANE AND HOI ST O PERABI LI TY...........................
3/4 9-8 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORACE P00L..................
3/4 9-9 3/4.9.8 WATER LEVEL-REACTOR VESSEL............................
3/4 9-10 3/4.9.9 WATER LEVEL-SPENT FUEL STORACE P00L...................
3/4 9-11 3/4.9.10 CONTROL ROD REMOVAL Single Control Rod Removal............................
3/4 9-12 Multiple Control Rod Remova1..........................
3/4 9-14 3/4.10 SPECIAL TEST EXCEPTION')
3/4.10.1 PRI KARY CONTAINMENT I NTEGRI TY.........................
3/4 10-1 3/4.10.2 ROD S EQUENCE CONTROL SYSTEM ( DELETED)................. 3/4 10-2' i
3/4.10.3 S HUTDOWN MARCIN DEMON STRATION S........................3/4 10-3 3/4.10.4 R EC I R CU LATI ON L00 0 S...................................3/4 10-4 3/4.10.5 P LANT S ERVI CE W AT ER...................,..<............
3/4 10-5 l
1 l
'1 l
BRUNSWICK - UNIT 2 IX Amendment No. 57, 163, 1
175 m
INDEX
- BASES, SECTION PACE 3/4.7 PLANT SYSTEMS (Continued) 3/4.7.3 F LOOD P ROTE CT 10 N.......................................
B 3/4 7-1 3/4.7.4 REACTOR CORE ISOLATION C00LINC SYSTEM..................
B 3/4 7-1 3/4.7.5 SNUBBERS...............................................
B 3/3 7-2 3/4.7.6 SEALED SOURCE CONTAMINATION............................
B 3/4 7-4 3/4.7.7 FIRE SUPPRESSION SYSTEMS............................
B 3/4 7-4 j
3/4.7.8 FIRE BARRIER PENETRATIONS..............................
B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS...................................
B 3/4 8-1 3/4.9 REFUELINC OPERATIONS 3/4.9.1 REACTOR MODE.CWITCH....................................
B 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................
B 3/4 9-1 3/4.9.3 CONTROL ROD P0SITION..................................
B 3/4 9-1 3/4.9.4 DEC.Y TIME............................................
B 3/4 9-1 3/4.9.5 COMHJNICAT10NS........................................
B 3/4 9-1 3/4.9.6 C RANE AND HOI ST OPERABI LI TY...........................
B 3/4 9-2 3/4.9.7' CRANE TRAVEL-SPENT FUEL STORACE P00L..................
B 3/4 9-2 3/4.9.8 WATER LEVEL-REACTOR VESSEL, and 3/4.9.9 WATER LEVEL-REACTOR FUEL STORACE P00L.................
B 3/4 9-2 3/4.9.10 CONTROL ROD REM 0 VAL....................................
B 3/4 9-2 3/4.10 SPECI AL TEST EXCEPTIONS 3/4.10.1 PRI MARY CONTA INMENT I NTEGR1 TY.........................
B 3/4 10-1 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM ( DELETED).................
B 3/4 10-1 l
3/4.10.3 S HUTDOWN MARGIN D EMONS1 RATI ON S........................
B 3/4 10-1 3/4.10.4 RECIRCULATION L00PS...................................
B 3/4 10-1 3/4.10.5 P LANT S ERV I C E W AT ER...................................
B 3/4 10-1 BRUNSWICK - UNIT 2 XII Amendment No. 51.-57, 70, 163, 175 J
e REACTIVJ7f CONTROL SYSTEMS
,LTHITING CONDITION FOR OPERATI (Continued)
ACTIOWt (Continued) 2.
If the inoperable control rod (s) is insertedt a)
Within one hour disarm the associated directional control valvee eithert 1)
Electrically, or 2)
Hydraulically by closing the drive water and exhaust water isolation valves.
b)
Otherwise, be in at least HOT SilVTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With more than 8 control rods inoperable, be in at least HOT SHUTDOWN c.
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE at least once per 31 days byt*
Verifying each valve to be open, a.
b.
Cycling each valve at least one complete cycle of full travel.
4.1.3.1.2 All withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notcht a.
At least once per 7 days when above the preset power level of the RWM and l
b.'
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when above the preset power level of the RWM and any control rod is immovable as a result of excessive l
friction or mechanical interference.
4.1.3.1.3 All withdrawn control rods shall be determined OPERABLE by I
demonstrating the scram discharge volume drain and vent valves OPERABLE, when the reactor protection system logic is tested-per Specification 4.3.1.2, by verifying that the drain and vent valvest Close within 30 seconde af ter receipt of a signal for control rods to a.
scram, and b.
Open when the scram signal is reset or the scram discharge volume trip is bypassed.
- These valves may be closed intermittently for testing under administrative control.
BRUNSWICK.- UNIT 2 3/4 1-4 Amendment No.
72, 76, 175
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i REACTIVITY CONTROL SYSTEMS
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CONTROL ROD HAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 6, based on de-energization of the scram pliot valve solenoids as time zero, shall not exceed 7.0 seconds.
APPLICABILITY: OPERATIONAL CONDITIONS I and 2.
ACTIONI With the maximum scram insertion time of one or more control rods exceeding 7.0 seconds, operation may continue and the provisions of Spacification 3.0.4 are not applicable provioed that t The control rod with the slow insertion time is declared inoperable, a.
b.
The requirements of Specification 3.1.3.1 are satisfied, and c.
If w' thin the preset power level of the RWM, the requirements cf Specitication 3.1.4.1.d are also satisfied, and d.
The Surveillance Requirements of Specification 4.1.3.2.c are l
performed at least once per 92 days when operation is continued with three or more control rods with slow scram insertion times Otherwise, be in at least 110T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demonstrated through mear rements For all control rods prior to THERMAL POWER exceeding 40% of RAYED a.
THERMAL POWER f ollowing CORE ALTERATIONS or af ter a reactor shutcown that is greater than 120 days, b.
For specifically affected individual control rods following maintenance on or modification to the control rod or. rod drive system which could af fect the scram insertion time of those specific control-rods, and For 10% of the control rods, on a rotating basis, at least once per c.
120 days of operation.
BRUNSWICK - UNIT 2 3/4 1-5 Amendment No. f f 9, 175
1 REACTIVITY CONTROL SYSTEMS FOUR CONTROL ROD CROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time, f rom the fully withdrawn position, for the three fastest control rods in each group of four control rods arranged in a two-by-two array, based on deenergitation of the scram pilot valve solenoids as time zero, shall not exceed any of the following Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 46 0.33 36 1.12 26 1.93 6
3.58 APPLICABILITY OPERATIONAL CONDITIONS I and 2.
ACTION With the average scram insertion times of control rods exceeding the,bove limits, operation may continue and the provisions of Specification 3.0.4 are not applicable provided8 The control rods with the slower than average scram insertion times a.
are declared inoperable, b.
The requirements of Specification 3.1.3.1 are satisfied, and If within the preset power level of the RWH, the requirements of c.
Specification 3.1.4.1.d are also satisfied, and' d.
The Surveillance Requirements of Specification 4.1.3.2.c are performed at least once per 92 days when operation is continued with l
three or more control rods with slow scram insertion times.
Otherwise, be in at least HOT SilVTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS l
4.1.3.4 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.
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BRUNSWICK - UNIT 2 3/4 1-7 Amendment No. 8 3, 175
-___L
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS LIMITINC CONDITION FOR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERABLE.
APPLICABILITY OPERATIONAL CONDITIONS 1, 2, and $*.
ACTIONI a.
In OPERATIONAL CONDITION ! or 2 with one control rod scram accumulator inoperable, the provisions of Specification 3.0.4 are not applicable and operation may continue, provided that within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />st 1.
The inoperable ar.cumulator is restored to OPERABLE status, or 2.
The control rod associated with the inoperable accumulator is declared inoperable, and the requirements of Specification 3.1.3.1 are satisfied.
3.
And, if within the preset power level of the RWM, the requirements of Specification 3.1.4.1.d are also satisfied.
Otherwi se, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In OPERATIONAL CONDITION $* with a withdrawn control rod scram accumulator inoperable, fully insert the af fected control rod and electrically disarm the directional control valves within one hour.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.1.3.5 The control rod scram accumulators shall be determined OPERABLE:
a.
At least once per 7 days by verif ying that the pressure and leak detectors are not in the alarmed condition, and b.
At least once per 18 months by performance of at 1.
CHANNEL FUNCTIONAL TEST of the leak detectors, and 2.
CHANNEL CALIBRATION of the pressure detectors.
- At least the accumulator associated with each withdrawn con'.rol rod. Not l
applicable to control rods removed per Specification 3.S.19.1 or 3.910.2.
BRUNSWICK - UNIT 2 3/4 1-8 Amendment No. 46, 160, 175 1
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j REACTIVITY CONTROL SYSTEMS--
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CONTROL ROD DRIVE COUPLING -
LIMITINC ' CONDITION FOR OPERATION 3.1.3.6 All control rods shall be coupled to their driv'e mechanisms..
APPLICABILITY:
CONDITIONS 1, 2, and $*.
ACTION:.
In r % i? ION 1 or 2 with one control rod not coupled to its-a.
'assos.a w d drive mechanism, the provisions of Specification 3.0.4 are not applicable and. operation may continue provided:
l 1.
Within the preset = power level of the RWH, the control rod is l
f declared inoperable and fully inserted until recoupling can be attempted with THERMAL POWER above the preset power level 'of the RWH and the requirements of Specification.3.1.4.1.d are l
satisfled.
1 2.
Above.the preset power level of the RWH, the control rod drive
_l is inserted to accomptish recoupling.
If'<ecoupling is not j
accomplished on the first attempt, declare the control' rod inoperable,-fully insert the control rod, and electrically 1
disarm the directional control valves.
3.
Tne requirements of Specification 3.1.3.1 are satisfied.
b.
In CONDITION 5*, with a withdrawn control rod not coupled to its i
associated drive mechanism, insert the control' rod to accomplish-recoupling. The provisions of Specification 3.0.3 are not_
I applicable.
SURVEI(LANCE REQUIREMENTS 4.1.3.6 The coupling integrity of a-control rod shall be. demonstrated by withdrawing the control rod to the fully withdrawn position and verifying that the rod does not go to the overtravel position:
- At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
BRUNSWICK - UNIT 2 3/4 1-9
- Amendment No. 175
i 3-
-4; 4
REALTIVITY CONTROL SYSTEMS CONTROL -ROD ' POSITION INDICATION '
LIMITING CONDITION FOR OPERATION 3.1.3.7 AllcontrolrodreedswitchpositionindicatorsshallbeOPERABLE.
APPLICABILITYt-CONDITIONS 1, 2, and 5*.
ACTIONt' a.
.in CONDITION 1 or 2: With one or more ccr rod reed swi tch' l
position indicators inoperable, including " Full-in" or " Full-out" indication, the provisions of Specification 3.0.4 are not applicable-and operation may continue,.provided.that within one hourt 1)
The position of. the control: rod is determined by an l
alternate method or-2)
The control rod is moved to a: position with an OPERABLE l
reed switch position indicator, or :
7 3)
The control rod with the inoperable reed switch position l
indicator is declared inoperable and the requirements of Specification 3.1.3.1 ~4re satisfied; t
4)
And, if within the preset power level of the RWH, the requirements of Specification'3.1.4.1.d are also satisfied; Other ae. be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
I b.
In L.G'f!ON 5* with a withdrawn control' rod reed switch position i' Tic t r inoperable, fully insert the withdrawn control rod. The visions of pecification 3,0.3 are not applicable.
4 4
- At least each witndrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
BRUNSWICK - UNIT 2 3/4 1-11 Amendment No. '175
=+.
a.
9 REAL"f!VITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod reed switch position indicators shall be determined l
OPERABLE by verifying a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, that the position of the control rod is ind8c4 ted i-b.
T:tei th4 sadicated control rod position changes during the movement.
vt the cont rol rod when performing ' Surveillance Requirement 4.i 3.1.2, and-c.
.That the control rod reed switch position indicator corresponds to the control rod position indicatedfby the "Fu l-out" reed switches l
when perf orming Surveillance Requirement 4.1.'.6, b.
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BRUNSWICK - UNIT 2 3/4 1-12
. Amendment No. 102, 175
i o
REACTIVITY CONTROL SYSTEMS 3/4 1.4 CONTROL ROD PROGRAM CONTROLS:
RCD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1 The Rod Worth Minimizer (RWM)'shall be OPERABLE when THERKAL POWER l's l
less than 10% of RATdD THERMAL POWER. _
APPLICABILITY OPERATIONAL CONDITIONS I and 2*.
ACTIOWs 1
~With the RWH inoperable after t'he first 12 control rods have been a.
fully withdrawn on a-startup, operation may continue provided that
. control rod < movement and compliance with the prescribed BPWS control rod pattern are verified by a second licensed operator or qualified member of the plant technical staff.
b.
With the RWH inoperable before:the first 12 control rods are withdrawn on a startup, one startup per calender year may be perfotwed provided that control rod movement and compliance with the prescribed BPWS control rod pattern are verified by a second licensed operator or qualified member of the plant technical staff.
y With RWH inoperable on a shutdown, shutdown may continue provided c.
that control rod movement. and compliance with the prescribed BPWS control rod pattern are verified by a second licensed' operator or qualified member of the plant technical staff.
d.
With RWH operable but individual control rod (s) declared inoperable, operation and control rod movement below the preset power level of-the RWH may continue providedt t
1.
No more than three (3) control rods are declared inoperable in any one BWS group, and, 2.
The inoperable control rod (s) is bypassed on the RWH and control rod movement of the bypassed rod (s) is verified by a second licensed operator or qualified member of the plant technical staff.-
With RWH inoperable, the provisions of Specification 3.0.4 are not e.
applicable.
i
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of = the RWH prior to withdrawal of control -rods for the purpose of bringing the reactor to criticality.
\\
BRUNSWICK - UNIT 2 3/4 1-14 Amendment No.
157, 175 j
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.11.4.1.1 The RWH shall be demonstrated Ort: ABLE in OPERATIONAL CONDITION 2, prior to withdrawal of control rods for the purpose of-making the reactor' j
critical and in OPERATIONAL CONDITION 1 when the RWM is initiated during 1
control rod insertion when reducing THERMAL POWER byt Verifying proper annunciation of the ' selection error of at least'one a.
out-of-sequence control rod, and b.
Verifying the rod block function of the RWM by moving an out-of-sequence control rod.
4.1.4.1.2 The RWM shall be demonstrated.0PERABLE-by-verifying the control rod Banked Position Withdrawal Sequence input to the RWM computer is correct following any loading of the sequence. program-into the computer.
W l
l l
BRUNSWICK - UNIT 2 3/4 1-14a Amendrent No. 175
r
'p1 t
_ REACTIVITY CONTROL SYSTEMS ROD SEQUENCE CONTROL SYSTEM l
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(Next page is 3/4 1-17) i BRUNSWICK -' UNIT 2 3/4 1-15 Amendment No.-'175 1
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o-SPECIAL TEST EXCEPTIONS 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM i
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BRUNSWICK - UNIT 2 3/4 10-2 Amendment No. 175 i
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0 REACTIVITY CONTROL SYSTEM BASES' CONTROL RODS (Continued) on a scram than has been analyzed even though cont rol rods with inoperable' i
accumulators may still be inserted with normal drive water pressure.
Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization' of the reactors.
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident. in the FSAR.'. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after each refueling..The subsequent check is performed as a.
backup to the inittal demonstratlon.
In order to ensure that the control rod patterns can be followed and trierefore that other parameters are within their limits, the control rod posi'. ton indication system must be OPERABLE.-
The control rod housing support restricts the outward movement of a control rod to less than.3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a' driving f orce to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so f requent as. to cause excessive wear on the system components.
3/4.1.4 CONTROL ROD PROCRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum in sequence individual control rod or control rod' segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /ga in the event of'a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than or equal to 10% of RATED THERMAL POWER, there is no possible
-l rod worth which, if dropped'at the design rate of the velocity limiter, could-result in a peak enthalpy of 280 cal /gm. Thus, requiring the RWM to be OPERABLE when THERMAL POWER is less than 10% of RATED THERMAL POWER provides adequate control.
Use of the Banked Position Withdrawal Sequence (BPWS) ensures that in the event of a control rod drop accident the peak fuel enthalpy will not be greater than 280 cal /gm (Reference 4).
BRUNSWICK - UNIT 2 B 3/4 1-3 Amendment No. 757, 175
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-.l e-REACTIVITY CONTROL SYSTEM BASES 1
i CONTROL ROD PROCRAM CONTROLS (Continued)
The RWM as a backup to procedural control provides an Lautomatic control-I
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rod pattern monitoring f unction to ensure adherence to the BPWS control movement sequences f rom 100% control rod density to 10% RATED THERMAL POWER
- J and, thus, eliminates the postulated control rod drop' accident from resulting in a peak fuel enthalpy greater than 280 cal /gm (Reference 5).
The requirement that RWH be operable for the withdrawal of the first'12 control rods on a startup is to ensure that the RWH. system maintains a high i
degree of availability.
Deviation from the BPWS control rod pattern may be allowed for the performance of Shutdown Margin Demonstration tests.
The nalysis of the rod drop accident is presented in -Section 15.4.6 of a
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the Updated FSAR and the techniques of the analysis are presented in a topical l.
re port (Reference 1) and two supplements (References 2 and 3).
The RBH is designed to automatically prevent. fuel. damage in the event of-erroneous rod withdrawal f rom locations of high power density during high power operation. The RBH is only required Leo be operable when the limiting condition described in Specification 3.i.4.3 exists.~ Two channels are provided. Tripping one of the channels.will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods.. Further. discussion of the RBM system is provided in Reference 5.
3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backu maintaining the reactor suberitical in the event that. p capability for tnsufficient rods are i
inserted in the core when a scram is called for. The' volume and weight percent of poison material in solution is based on being able to bring the reactor to the suberitical condition as the plant' cools to ambient condition. The temperature requirement is necesrary.to keep the' sodium pentaborate in solution. Checking the volume and tem hours assures that the solution is available for use.perature once each 24 Vith redundant pumps and a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the' redundant components inoperable.
Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
Replacement of the explosive charges 'in the valves at -regular intervals will assure that these valves will not fail because of detertoration of the charges.
" BRUNSWICK - UNIT 2 B 3/4 1-4 Amendment No.~ 75/,7$
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o REACTIVITY CONTROL SYSTEM j
BASES i
1 Referencest
'1.
C. J. Paone, R. C. Stirn, and J. A. Woodley, " Rod Drop Accident Analysis for,Large BWRs " C. E. Topical Report' NEDO-10527, March 1972.
I 2.
C. J. Paone,- R. C. Stirn, and R. H. Yound, Supplement I to NEDO-10527, July 1972.
3..
J. A. Haum, C. J. Paone, and R. fC. Stirn, addendum 2 " Exposed Cores" supplement 2 L to NEDO-10527, January 1973.
4.
NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel,"
Revision 6, Amendment 12.
5.
NEDC-31654P, " Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant, February 1989.
6.
NEDE-20411-P-A, " General Electric Standard Application for Reactor Fuel,"
Revision 8, Amendment 17.
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BRUNSWICK - UNIT 2 -
B 3/4 1-5 Amendment No. 175 l
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s 3/4.10 SPECIAL TEST EXCEPTIONS
. BASES 3/4.10.1 PRIMARY CONTAINMENT INTECRITY The requirement for PRIMARY CONTAINMENT INTECRITY is removed during the period when open vessel tests are being' performed during low power PHYSICS TESTS.
3/4.10.2 ROD SEQUENCE CONTROL SYSTEM (DELETED) 3/4.10.3 SHUTDOWN MARCIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head i
removed requires additional restrictions in order to ensure that criticality does not occur. These additional restrictions are specified.in this LCO.
3/4.10.4-RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain start-up and. PHYSICS TESTS while =,
at low THERMAL POWER levels.
3/4.10.5 PLANT SERVICE WATER This Special Test Exception permits securing the Service Water System conventional header when the nuclear header is out of service and is required to permit flange installation in service water system header cross-connect piping.
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BRUNSWICK - UNIT 2 B 3/4 10-1 Amendment No. 57, 175
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