ML20053B067
| ML20053B067 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/07/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Carolina Power & Light Co |
| Shared Package | |
| ML20053B068 | List: |
| References | |
| DPR-62-A-070, DPR-71-A-047 NUDOCS 8205270725 | |
| Download: ML20053B067 (19) | |
Text
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UNITED STATES h,
lt NUCLEAR REGULATORY COMMisslON j.. T y
WASHINGTON, D. C. 20555 k, -
gI CAROLINA POWER & LIGHT COMPANY DOCKET N0. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Arendment No. 47 License No. DPR-71 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Carolina Power & Light Company dated April 30, 1982, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act),
and,the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common 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 Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications 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:
(2) Technical Soecifications The Technical Specifications contained in Appendices A a'nd B, as revised through Amendment No. 47, are hereby incorporated in the license. The licensee shall operate the facility in i
l accordance with the Technical Specifications.
l 07 %_
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. 3.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR RE" TORY COMMISSION l
~.
Domenic. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specificctions Date of Issuance: May 7,1982 4
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ATTACHMENT TO LICENSE AMENDMENT NO. 47 FACILITY OPERATING LICENSE N0. 'DPR 71 DOCKET NO. 50-325 Remove the following pages and replace with identically numbered pages.
II XII l-6 3/49-3 3/4 9-4 B3/4 9-1 B3/4 9-2 4
h a
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INDEX DEFINITIONS SECTION 1.0 DEFINITIONS (Continued)
PAGE PRE S S URE BOUNDARY LEAKAGE........................................ 1-4 PRIMARY CONTAINMENT INTEGRITY.................................... 1-4 RATED THERMAL P0WER..............................................
1-5 REACTOR PROTECTION SYSTEM RESPONSE TIME..........................
1-5 REPORTABLE OCCURRENCE............................................
1-5 ROD DENSITY......................................................
1-5 SECONDARY CONTAINMENT INTEGRITY.................................. 1-5 SHUTDOWN MARGIN..................................................
1-6 SPIRAL REL0AD....................................................
1-6 SPIRAL UNL0AD....................................................
1-6 STAGGERED TEST BASIS.............................................
1-6 THERMAL P0WER....................................................
1-6 TOTAL PEAKING FACT 0R............................................. 1-6 U N I D E NT I F I E D LEAKAG E............................................. 1-6 F REQU EN CY NOTATIO N, TAB LE 1.1.................................... 1-7 OPERATIONAL CONDITIONS, TABLE 1.2................................
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l BRUNSWICK - UNIT I II Amendment No. 47
INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS (Continued) 3/4.7.3 FLOOD PROTECTION..................................
B 3/4 7-1 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM.............
B 3/4 7-1 3/4.7.5 HYD RAU LIC S NUB B E RS................................
B 3/4 7-2 3/4.7.6 SEALED SOURCE CONTAMINATION.......................
B 3/4 7-3 3/4.7.7 FIRE SUPPRESSION SYSTEMS..........................
B 3/4 7-3 3/4.7.8 FIRE BARRIER PENETRATIONS.........................
B 3/4 7-4 3/4.8 ELECTRICAL POWER 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 P0SITION..............................
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 HOIST OPERABILITY.......................
B 3/4 9-2 l
3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE P00L..............
B 3/4 9-2 3/4.9.8 WATER LEVEL-REACTOR VESSEL, and 3/4.9.9 WATER LEVEL-REACTOR FUEL STORAGE P00L.............
B 3/4 9-2 3/4.9.10 CO NT ROL ROD REM 0 VAL...............................
B 3/4 9-2
, 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY.....................
B 3/4 10-1 3/4.10.2 R0D SEQUENCE CONTROL SYSTEM.......................
B 3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS....................
B 3/4 10-1 3/4.10.4 RECIRCULATION L00PS...............................
B 3/4 10-1 3/4.10.5 PLANT SERVICE WATER...............................
B 3/4 10-5 i
BRUNSWICK - UNIT 1 XII Amendment No.,32",36(367 47 I
DEFINITIONS SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor would be subcritical assuming that all control rods capable of insertion are fully inserted except for the analytically determined highest worth rod which is assumed to be fully withdrawn, and the reactor is in the shutdown condition, cold, 68'F, and Xenon-free.
SPIRAL RELOAD 1.31 A SPIRAL RELOAD is the reverse of a SPIRAL UNLOAD. Except for two diagonal fuel bundles around each of the four SRMs, the fuel in the interior of the core, sysmetric to the SRMs, is loaded first.
i SPIRAL UNLOAD 1.32 A SPIRAL UNLOAD is a core unload performed by first removing the fuel from the outermost control cells (four bundles surrounding a control blade).
Unloading continues in a spiral fashion by removing fuel from the outermost periphery to the interior of the core, symmetric about the SRMs, except for two diagonal fuel bundles around each of the four SRMs.
STAGGERED TEST BASIS 1.33 A STAGGERED TEST BASIS shall consist of:
a.
A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals.
b.
The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
THERMAL POWER l
1.34 THERMAL POWER shall be the total reactor core heat transfer rate to the l
TOTAL PEAKING FACTOR 1.35 The TOTAL PEAKING FACTOR (TPF) shall be the ratio of local LHGR for any specific location on a fuel rod divided by the average LHGR associated with the fuel bundles of the same type operating at the core average bundle power.
UNIDENTIFIED LEAKAGE 1.36 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
BRUNSWICK - UNIT 1 1-6 Amendment No.,32', 47
REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 During CORE ALTERATIONS, the requirements for the source range monitors (SRMs) shall be:
a.
Two SRMs* shall be OPERABLE, one in the core quadrant where fuel is being moved and one in an adjacent quadrant. For an SRM to be considered OPERABLE, it shall be inserted to the normal operating level and shall have a minimum of 3 cps except as specified in d and e below.
b.
The SRMs shall give a continuous visual indication in the Control
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Room.
c.
The " shorting links" shall be removed f rom the RPS circuitry prior to and during the time any control rod is withdrawn ** and shutdown margin demonstrations, d.
During a core SPIRAL UNLOAD the count rate may drop below 3 cps, e.
Prior to a core SPIRAL RELOAD, two diagonally adjacent fuel assemblies shall be loaded into different control cells containing control blades around each SRM to obtain 3 cps. Until these assemblies have been loaded, the 3 cps count rate is not required.
APPLICABILITY:
CONDITION 5 ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods. The provisions of Specification 3.0.3 are not applicable.
- The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detectors are connected to the normal SRM circuits.
- Not required for control rods removed per Specifications 3.9.10.1 or 3.9.10.2.
BRUNSWICK - UNIT 1 3/4 9-3 Amendment No.,36', 47
SURVEILLANCE REQUIREMENTS 4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:
a.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; 1.
Performance of a CRANNEL CHECK, 2.
Verifying the detectors are inserted to the normal operating
- level, l
3.
During CORE ALTERATIONS, verifying that the detector of an OPERABLE SRM channel is located in the core quadrant where CORE ALTERATIONS are being performed and one is located in the adjacent quadrant, l
4.
During CORE ALTERATIONS, verifying that the channel count rate is at least 3 cps (except as noted in Specification 3.9.2.d and 3.9.2.e),
5.
During a core SPIRAL UNLOAD or SPIRAL RELOAD, verifying that the fuel movement sheet is being followed.
b.
Verifying prior to the start of a SPIRAL RELOAD that the SRMs have been raised to a count rate of at least 3 cps by the insertion of adjacent fuel assemblies.
c.
Performance of a CilANNEL FUNCTIONAL TEST:
1 1.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of CORE ALTERATIONS, and 2.
At least once per seven days.
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1 BRUNSWICK - UNIT 1 3/4 9-4 Amendment No. 47
3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 REACTOR H0DE SWITCH Locking the reactor mode switch in the refuel position ensures that the restrictions on rod withdrawal and refueling platform movement during the refueling operations are properly activated. These conditions reinforce the refueling procedures and reduce the probability of inadverter.t criticality, damage to reactor internals, fuel assemblies and exposure of personnel to excessive radioactivity.
3/4.9.2 INSTRUKENTATION The OPERABILITY of the source range monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
During a SPIRAL UNLOAD, the count rate of the SRM will decrease 'below 3 cps before all of the fuel is unloaded. The count rate of 3 cps is not necessary since there will be no reactivity additions during the spiral unload. The SRMs will be required to be OPERABLE prior to the SPIRAL UNLOAD, and each SRM will be verified operational by raising the count rnte to 3 cps prior to the spiral reload by inserting two assemblies around each SRM.
This will ensure that the SRMs can be relied upon to monitor core reactivity during the reload.
3/4.9.3 CONTROL ROD POSITION The requirement that all control rods be inserted during CORE ALTERATIONS ensures that fuel will not be loaded into a cell without a control rod and prevents two positive reactivity changes from occurring simultaneously.
i 3/4.9.4 DECAY TIME The minimum requirement for reactor subcriticality prior to fuel movement ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assuuptions used in the accident analyses.
3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during movement of fuel within the reactor pressure vessel.
l BRUNSWICK - UNIT 1 B 3/4 9-1 Amendment No. 47 l
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REFUELING OPERATIONS BASES 3/4.9.6 CRANE AND HOIST OPERABILITY The OPERABILITY requirements of the cranes and hoists used for movement of fuel assemblies ensures that:
- 1) each has sufficient load capacity to lift a fuel element, and 2) the core internals and pressure vessel are protected from excessive lif ting force in the event they are inadvertently engaged during lifting operations.
3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE POOL The restriction on movement of loads in excess of the weight specified provides some assurance that with the failure of the lifting device.the fuel pool would not be damaged to such a degree that the irradiated fuel would be subjected to a loss-of-coolant.
3/4.9.8 and 3/4.9.9 WATER LEVEL-REACTOR VESSEL AND SPENT FUEL STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of irradiated fuel assembly. This minimum water depth is consistent with the assumptions of the accident analysis.
3/4.9.10 CONTROL ROD REMOVAL This specification ensures that maintenance or repair on control rods cc control rod drives will be performed under conditions that limit the probability of inadvertent criticality. The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain suberitical with only one l
control rod fully withdrawn.
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l BRUNSWICK - UNIT 1 B 3/4 9-2 Amendment No. 47 l
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UNITED STATES
['y fiUCLEAR REGULATORY COMMisslON j. ( 'Cf )j tj W ASHINGTON, D. C. 20555
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CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE P
Amendment No. 70 License No. DPR-62 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for. amendment by Carolina Power & Light Company dated April 30, 1982, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operaty 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 comon defense and security or to the health and safety of the l
public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of 'the Commission's regulations and all applicable requirements have been satisfied.
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2.
Accordingly, the license is amended by changes to the Technical Spec-l ifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-62 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.70, 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..
FOR THE NUCLEAR REGULATORY COMMISSION
/f Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical i
Specificctions l
Date of Issuance: May 7, 1982 I
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ATTACHMENT TO LICENSE AMENDMENT NO. 70 FACILITY OPERATING LICENSE ~ NO. 'DPR-62 DOCKET NO. 50-324 Remove the following pages and replace with identically numbered pages.
II 2
XII l
l-6 3/49-3 3/4 9-4 B3/4 9-1 B3/4 9-2 l
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INDEX DEFINITIONS SECTION 1.0 DEFINITIONS (Continued)
PAGE PHYSICS TESTS................................................
1-4 P RES S U RE BOUNDARY LEAKAGE....................................
1-4 PRIMARY CONTAINMENT INTEGRITY................................
1-4 RATED THERMAL P0WER..........................................
1-5 REACTOR PROTECTION SYSTEM RESPONSE TIME......................
1-5 REPORTABLE OCCURRENCE........................................
1-5 ROD DENSITY..................................................
1-5 SECONDARY CONTAINMENT INTEGRITY..............................
1-5 SHUTDOWN MARGIN..............................................
1-6 S P I RAL RE L0 AD................................................
1-6 SPIRAL UNL0AD................................................
1-6 STAGGERED TEST BASIS.........................................
1-6 THERMAL P0WER................................................
1-6 TOTAL PEAKING FACT 0R.........................................
1-6 UNIDENTIFIED LEAKAGE.........................................
1-6 FREQUENCY NOTATION, TABLE 1.1................................
1-7 OPERATIONAL CONDITIONS, TABLE 1.2............................
1-8 BRUNSWICK - UNIT 2 II Amendment No. 70
INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS (Continued) 3/4.7.3 FLOOD PROTECTION...................................
B 3/4 7-1 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM..............
B 3/4 7-1 3/4.7.5 HYD RAULIC S NUB BERS.................................
B 3/4 7-2 3/4.7.6 SEALED SOURCE CONTAMINATION........................
B 3/4 7-3 3/4.7.7 FIRE SUPPRESSION SYSTEMS...........................
B 3/4 7-3 3/4.7.8 FIRE BARRIER PENETRATIONS..........................
B 3/4 7-4 3/4.8 ELECTRICAL POWER 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 P0SITION...............................
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 HOIST OPERABILITY........................
B 3/4 9-2 l 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE P00L...............
B 3/4 9-2 3/4.9.8 WATER LEVEL-REACTOR VESSEL, and 3/4.9.9 WATER LEVEL-SPENT FUEL STORAGE P00L................
B 3/4 9-2 3/4.9.10 CONTROL ROD REM 0 VAL................................
B 3/4 9-2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY......................
B 3/4 10-1 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM........................
B 3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.....................
B 3/4 10-1 3/4.10.4 RECIRCULATION L00PS................................
B 3/4 10-1 3/4.10.5 PLANT SERVICE WATER................................
B 3/4 10-5 BRUNSWICK - UNIT 2 XII Amendment No.JFfdEP,70
DEFINITIONS SHUTDOWN MARGIN 1.31 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor would be subcritical assuming that all control rods capable of insertion are fully inserted except for the analytically determined highest worth rod which is assumed to be fully withdrawn, and the reactor is in the shutdown condition, cold, 68'F, and Xenon-free.
SPIRAL RELOAD 1.32 A SPIRAL RELOAD is the reverse of a SPIRAL UNLOAD. Except for two diagonal fuel bundles around each of the four SRMs, the fuel in the interior of the core, symmetric to the SRMs, is loaded first.
SPIRAL UNLOAD 1.33 A SPIRAL UNLOAD is a core unload performed by first removing the fuel from the outermost control c 11s (four bundles surrounding a control blade).
Unloading continues in a spiral fashion by removing fuel from the outermost periphery to the interior of the core, symmetric sbout the SRMs, except for two diagonal fuel bundles around each of the four SRMs.
STAGGERED TEST BASIS 1.34 A STAGGERED TEST BASIS shall consist of:
a.
A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b.
The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
THERMAL POWER 1.35 THERMAL POWER shall be the total reactor core heat transfer rate to the l
l TOTAL PEAKING FACTOR 1.36 The TOTAL PEAKING FACTOR (TPF) shall be the ratio of local LHGR for any specific location on a fuel rod divided by the average LHGR associated with the fuel bundles of the same type operating at the core average bundle power.
UNIDENTIFIED LEAKAGE 1.37 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
BRUNSWICK - UNIT 2 1-6 Amendment No. pep, JFr,70
REFUhlING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 During CORE ALTERATIONS, the requirements for the source range monitors (SRMs) shall be:
a.
Two SRMs* shall be erMRABLE, one in the core quadrant where fuel is being moved and one in an adjacent quadrant. For an SRM to be considered OPERABLE, it shall be inserted to the normal operating level and shall have a minimum of 3 cps except as specified in d and e below.
b.
The SRMs shall give a continuous visual indication in the Control Room.
c.
The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn ** and shutdown margin demonstrations, d.
During a core SPIRAL UNLOAD the count rate may drop below 3 cps.
e.
Prior to a core SPIRAL RELOAD, two diagonally adjacent fuel assemblies shall be loaded into different control cells containing control blades around each SRM to obtain 3 cps. Until these assemblies have been loaded, the 3 cps count rate is not required.
APPLICABILITY: CONDITION 5 ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods. The provisions of Specification 3.0.3 are not applicable.
- The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detectors are connected to the normal SRM circuits.
- Not required for control rods removed per Specifications 3.9.10.1 or 3.9.10.2.
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BRUNSWICK - UNIT 2 3/4 9-3 Amendment No.,50',
70 I
f SURVEILLANCE REQUIREMENTS 4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:
a.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; 1.
Performance of a CHANNEL CHECK, 2.
Verifying the detectors are inserted to the normal operating
- level, 3.
During CORE ALTERATIONS, verifying that the detector of an OPERABLE SRM channel is located in the core quadrant where CORE ALTERATIONS are being performed and one is located in the adjacent quadrant, 4.
During CORE ALTERATIONS, verifying that the channel count rate is at least 3 cps (except as noted in Specification 3.9. 2.d and 3.9.2.e),
5.
During a core SPIRAL UNLOAD or SPIRAL RELOAD, verifying that the fuel movement sheet is being followed.
b.
Verifying prior to the start of a SPIRAL RELOAD that the SRMs have been raised to a count rate of at least 3 cps by the insertion of adjacent fuel assemblies.
c.
Performance of a CllANNEL FUNCTIONAL TEST:
1.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of CORE ALTERATIONS, and 2.
At least once per seven days.
BRUNSWICK - UNIT 2 3/4 9-4 Amendment No. 70
3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 REACTOR MODE SWITCH Locking the reactor mode switch in the refuel position ensures that the restrictions on rod withdrawal and refueling platform movement during the refueling operations are properly activated. These conditions reinforce the refueling procedures and reduce the probability of inadvertent criticality, damage to reactor internals, fuel assemblies and exposure of personnel to excessive radioactivity.
3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
During a SPIRAL UNLOAD, the count rate of the SRM will decrease below 3 cps before all of the fuel is unloaded. The count rate of 3 cps is not necessary since there will be no reactivity additions during the spiral unload. The SRMs will be required to be OPERABLE prior to the SPIRAL Uh10AD, and each SRM will be verified operational by raising the count rate to 3 cps prior to the spiral reload by inserting two assemblies around each SRM.
This will ensure that the SRMs can be relied upon to monitor core reactivity during the reload.
3/4.9.3 CONTROL ROD POSITION The requirement that all control rods be inserted during CORE ALTERATIONS ensures that fuel will not be loaded into a cell without a control rod and l
prevents two positive reactivity changes from occurring simultaneously.
3/4.9.4 DECAY TIME The minimum requirement for reactor suberiticality prior to fuel movement ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.
l l
3/4.9.5 COMMUNICATIONS l
The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the f acility status or core reactivity condition during movement of fuel within the reactor pressure vessel.
l i
BRUNSWICK - UNIT 2 B 3/4 9-1 Amendment No. 70 l
REFUELING OPERATIONS BASES 3/4.9.6 CRANE AND HOIST OPERABILITY The OPERABILITY requirements of the cranes and hoists used for movement of fuel assemblies ensures that:
- 1) each has sufficient load capacity to lift a fuel element, and 2) the core internals and pressure vessel are protected f rom excessive lif ting force in the event they are inadvertently engaged during lifting operations.
3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE POOL The restriction on movement of loads in excess of the weight specified provides some assurance that with the failure of the lifting device the fuel i
pool would not be damaged to such a degree that the irradiated fuel would be subjected to a loss-of-coolant.
3/4.9.8 and 3/4.9.9 WATER LEVEL-REACTOR VESSEL AND SPENT FUEL STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of irradiated fuel assembly. This minimum water depth is consistent with the assumptions of the accident analysis.
3/4.9.10 CONTROL ROD REMOVAL This specification ensures that maintenance or repair on control rods or control rod drives will be performed under conditions that limit the probability of inadvertent criticality. The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.
BRUNSWICK - UNIT 2 B 3/4 9-2 Amendment No. 70
-_