ML20063F325
| ML20063F325 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 02/08/1994 |
| From: | Milano P Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20063F332 | List: |
| References | |
| NUDOCS 9402140250 | |
| Download: ML20063F325 (14) | |
Text
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UNITED STATES I
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E NUCLEAR REGULATORY COMMISSION
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E WASHINGTON, D. C. 20555
+...+
CAROLINA POWER & LIGHT COMPANY. et al.
DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 1 i
I!
AMENDMENT TO FACILITY OPERATING LICENSE
.t Amendment No. 168-l License No. DPR-71 l
1.
The Nuclear Regulatory Commission (the Commission) has found'that:
{
A.
The application for amendment filed by Carolina Power &' Light 1
Company (the licensee), December 31, 1992, as supplemented June 10,
-l 1993, and August 23, 1993,-and the request dated December 8,~1993, j'
comply 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 j
Commission; C.
There is reasonable assurance (1) that the activities' authorized by this amendment can be conducted without endangering the health and I
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 -
1 defense and security or to the health and safety of the public;.and.
1 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.
Ji 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-I i
1 9402140250 940200
- PDR ADOCK 05000324'-
P PDR.
i
h (2) Technical Specifications
-The Technical Specifications contained in Appendices A and B, as revised through Amendment No.168, 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 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/
S. Singh Bajwa, Acting Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: February 8, 1994
i ATTACHMENT TO LICENSE AMENDMENT NO.168 FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-115 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Paaes Insert Paaes 1-2 1-2 3/4 3-29 3/4 3-29 3/4 9-5 3/4 9-5 8 3/4 9-1 B 3/4 9-1 i
i
' DEFINITIONS
[
t
' CHANNEL FUNCTIONAL TEST (Continued) b.
Bistable channels - the injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions.
i CORE ALTERATION f
CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel.
Movement of source range monitors, local power range monitors, intermediate i
range monitors, traversing in-core probes, or special moveable detectors (including undervessel replacement) is not considered a CORE ALTERATION.
In addition, control rod movement with other than the normal control rod drive is not considered a CORE ALTERATION provided there are no fuel assemblies in the associated core cell.
Suspension of CORE ALTERATIGNS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with' Specifications 6.9.3.1, 6.9.3.2, 6.9.3.3, and 6.9.3.4.
Plant operation within these core operating limits is addressed'in individual specifications.
CRITICAL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be the ratio of'that power in.an assembly.
which is calculated, by application of an NRC approved CPR correlation, to cause some point in the assembly to experience boiling. transition, divided by d
the actual assembly operating power.
DOSE EOUIVALENT I-131
.)
l DOSE EQUIVALENT I-131 shall be concentration of I-131, pCi/ gram, which alone
]
would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, 1-134,"and I-135 actually present.- The'followingfis defined equivalent to 1 pCi of I-131 as determined from Table.III.of
. TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites"-
I-132, 28 pCi; I-133, 3.7 pCi; I-134, 59 pCi; I-135, 12 pCi.
]
E -AVERAGE DISINTEGsATION ENERGY E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of;the
'I average beta and gamma energies per disintegration (in MeV) for isotopes with half lives greater than 15 minutes making up at least 95% of the total-non-iodine activity in the coolant.
BRUNSWICK - UNIT 1 1-2 Amendment No, $7,124,13J, H
JA/,168 r-
m 4
'E I
M TABLE 4.3.2-1 (Continued) n ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS k
CHANNEL OPERATIONAL
-i CHANNEL FUNCTIONAL CHANNEL.
CONDITIONS IN WHICH TRIP FUNCTION' CHECK TEST CALIBRATION SURVEILLANCE REQUIRED
- 4. CORE STANDBY COOLING SYSTEMS ISOLATION a.
High Pressure Coolant Injection System' Isolation 1.
HPCI Steam Line Flow - High
)
Transmitter:
a) g R(b) 1, 2, 3 Trip. Logic:
D M
M 1,2,3 ta 2.
- HPCI Steam Line High Flow D
Time Delay Relay NA R
R 1,2,3 3.
HPCI Steam Supply Pressure - Low NA M
R' 1, 2, 3 4.
HPCI Steam Line Tunnel Temperature ~.- High NA SA' R
1, 2. 3 l
S.
Bus Power Monitor NA R
.NA.
1,2,3 6.
HPCI. Turbine Exhaust Diaphragm Pressure High NA M
Q 1,2,3 7.
HPCI Steam Line Ambient Temperature - High
.NA-SA'.
R 1,2,3 8.
HPCI Steam Line Area l
(
.A Temperature - High NA SA R
1,2,3 E
9.
HPCI Equipment Area-l Temperature
.._High-NA SA R.
1, 2.- 3 r+
.10.
Dr.ywell Pressure - High 2
P Transmitter:
NA")
NA-
-R*
1,2,3 g
Trip. Logic:
D.
.M~
M 1,.2, 3 w
l..
(D
. - =
- = -
=.
. =.
. =
m m
m
-_mm...
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REFUELING OPERATIONS 3/4.9.3' CONTROL ROD POSITION LIMITING CONDITION FOR OPERATION 3.9.3 All control rods shall be fully inserted *.
l i
APPLICABILITY:
OPERATIONAL CONDITION 5, during loading of-fuel assemblies.
.i into the core **.
t ACTION:
With all control rods not fully inserted, immediately suspend loading of fuel assemblies into the core.
The provisions of Specification 3.0.3 are not l
applicable.
SURVEILLANCE REOUIREMENTS 4,9.3 Verify all control rods to be fully inserted within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to.
j the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during loading of fuel assemblies into the core.
l i
- Except control rods removed per Specification 3.9.10.1 or~3.9.10.2.
j
- See Special Test Exception 3.10.3.
i
- 1 t
1
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- i BRUNSWICK - UNIT 1 3/4 9-5 Amendment No.168 i
3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 REACTOR MODE SUITCH Incking 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 UNLOAD, and each SRM will be verified operational by raising the count rate to 3 cps prior to the SPIRAL RELOAD by inserting up to four fuel 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 loading of fuel assemblies into the' core ensures that fuel will not be loaded into a cell without a control rod and 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 i
short lived fission products.
This decay time is consistent with the assumptions used in the accident analyses.
?
3/4.9.5 C0KKUNICATIONS The requirement for communications capability ensures that refueling station personnel can bo promptly informed of significant changes in.the j
facility status or care reactivity condition during movement of fuel within l
.the reactor pressure vessel.
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t BRUNSUICK - UNIT 1 B 3/4 9-1 Amendment No /),168 t
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o UNITED STATES
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't NUCLEAR REGULATORY COMMISSION g
<g W ASHINGTON, D. C. 20555 l
CAROLINA POWER & LIGHT COMPANY. et al.
DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 199 License No. DPR-62 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment filed by Carolina Power & Light Company (the licensee), December 31, 1992, as supplemented June 10, 1993, and August 23, 1993, and the request dated December 8, 1993, comply 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; i
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 public; and.
'l E.
The issuance of this amendment is in accordance with 10'CFR Part 51 of the Comissior'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-62 is hereby amended to read as follows:
. (2) Technical Snecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.199, 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 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION S. Singh Bajwa, Acting Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
j Changes to the Technical Specifications i
Date of Issuance: February 8, 1994
-~
ATTACHMENT TO LICENSE AMENDMENT NO.199 FALILITY OPERATING LICENSE NO. DPR-62 U0CKET NO. 50-324 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Paces Insert Paoes 1-2 1-2 3/4 3-29 3/4 3-29 3/4 9-5 3/4 9-5 B 3/4 9-1 B 3/4 9-1 I
P s
t I
e
I; l
+
DEFINITIONS f
CHANNEL FUNCTIONAL TEST (Continued) b.
Bistable channels - the injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions.
CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, reactivity control I
components, or other components affecting reactivity within the reactor vessel with the vessel head-removed and fuel in the vessel.
Movement of source range monitors, local power range monitors, intermediate range monitors, traversing in-core probes, or special moveable detectors j
(including undervessel replacement) is not considered a CORE ALTERATION.
i In addition, control rod movement with other than the normal control rod drive is not considered a CORE ALTERATION provided there are no fuel assemblies in U-the associated core cell.
Suspension of CORE ALTERATIONS shall not preclude j
completion of movement of a component to a safe position.
CORE OPERATING LIMITS REPORT i
The CORE _0PERATING LIMITS REPORT is the unit-specific document that provides
- i core operating lLmits for the current reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.3.1, 6.9,3.2, 6.9.3.3, and 6.9.3.4.
Plant' operation within these core operating limits is addressed in individual specifications.
CRITICAL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in an assembly-which is calculated, by application of an NRC approved CPR correlation, to cause some point in the assembly to experience boiling transition, divided by
~'.,
the actual assembly operating power.
1 DOSE EOUIVALENT I-131 DOSE EQUIVALENT I-131 shall be concentration of I-131, pci/ gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, I-132,~I-133, I-134, and I-135 actually present. The following_is defined equivalent to 1 pCi of I-131 as determined from Table III of j
TID-14844, " Calculation of Distance Factors-for Power and Test Reactor J
Sites":
I-132, 28 pCi; 1-133, 3.7 pCi; I-134,-59 pCi; I-135, 12 pCi.
i E -AVERAGE DISINTEGRATION ENERGY Y
-E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum.of the
'I average beta and gamma energies per disintegration (in MeV) for isotopes with half lives greater than 15 minutes making up at least 95% of the total' non-iodine activity in.the coolant.
i BRUNSWICK - UNIT 2 1-2 JAmendment No. 199 1
E w
TABLE 4.3.2-1 (Continued) n ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREHENTS CHANNEL OPERATIONAL H
CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH m
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED
- 4. CORE STANDBY COOLING SYSTEMS ISOLATION a.
High Pressure Coolant Injection System Isolation 1.
HPCI Steam Line Flow - High Transmitter:
NA")
NA R(*)
1, 2, 3 Trip Logic:
D H
H 1,2,3 w
2.
HPCI Steam Line Flow - High 2
Time Delay Relay NA R
R 1,2,3 3.
HPCI Steam Supply Pressura Low NA H
R 1,2,3 4.
HPCI Steam Line Tunnel Temperature High NA SA R
1,2,3 l
5.
Bus Power Honitor NA R
NA 1,2,3 6.
HPCI Turbine Exhaust Diaphragm Pressure - High NA H
Q 1,2,3 7.
HPCI Steam Line Ambient Temperature - High-NA SA R
1,2,3 8.
HPCI Steam Line Area-L" A Temperature - High NA SA R.
1,2,3 m
h 9.
HPCI Equipment Area g
Temperature - High NA SA R
1,2,3
~
Dr$nsmitter:11 Pressure - High 10.
z NA")
NA R'b' 1, 2, 3 9
Trip Logic:
D H
H 1,2,3 G
e
=
REFUELING OPERATIONS l
3/4.9,3 CONTROL ROD POSITION l
LIMITING CONDITION FOR OPERATION 3,9.3 All control rods shall be fully inserted *.
APPLICABILITY:
OPERATIONAL CONDITION 5,.during loading of fuel assemblies into the core **.
1 ACTION:
With all control rods not fully inserted, immediately suspend loading of fuel-
-j assemblies into the core. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 1
I 4.9.3 Verify all control rods to be fully inserted within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during loading of fuel assemblies-into the core.
- Except control rods removed per Specification 3.9.10.1 or 3.9.10.2.
l.
- See Special Test Exception 3.10.3.
y i
1 i
I l
l BRUNSWICK - UNIT 2 3/4 9-5 Amendment No.199
..--,.. -, =.
4 3/4.9 REFUELING OPERATIONS BASES 3/4.9.I REACTOR MODE SUITCH 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.
i 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 rate to 3 cps prior to the SPIRAL RELOAD by inserting up to four fuel 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 loading of fuel assemblies into the core ensures that fuel will not be loaded into a cell without a control rod and 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.
3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facilicy status or core reactivity condition during movement of fuel within the reactor pressure vessel, BRUNSWICK - UNIT 2 B 3/4 9-1 Amendment No.JJ4,199