ML20056F816

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Proposed Tech Specs 3.9.3 Re Definition of Core Alterations & Use of Terminology
ML20056F816
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 08/23/1993
From:
CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20056F813 List:
References
NUDOCS 9308310135
Download: ML20056F816 (8)
v • d • e


Text

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. i ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKET NOS. 50-325 & 50-324 OPERATING LICENSE NOS. DPR-71 & DPR-62 ,

REQUEST FOR LICENSE AMENDMENT CORE ALTERATIONS l

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TYPED TECHNICAL SPECIFICATION PAGES - UNIT 1 i

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1 El-1 9308310135 930823 iJ' PDR ADOCK 05000324 6j P PDR Li .

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D'E FINITIONS i i

CHANNEL FUNCTIONAL TEST (Continued)  !

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b. Bistable channels - the injection of a simulated signal into the i channel sensor to verify OPERABILITY including alarm and/or trip i

functions.

CORE ALTERATION 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  !

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 ALTERATIONS shall not preclude completion of movement of a component to a safe position.

I 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  ;

I these core operating limits is addressed in individual specifications.

CRITICAL POWEk RATIO

[

The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in an assembly -i 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.

l DOSE EOUlVALENT I-131 l

! DOSE EQUIVALENT I-131 shall be concentration of I-131, pCi/ gram, which alone l would produce the same thyroid dose as the quantity and isotopic mixture of l I-131, I-132. I-133, I-134, and I-135 actually present. The following is l defined equivalent to 1 Ci of I-131 as determined from Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites": 1-132, 28 pCi; I-133, 3.7 pCi; I-134, 59 Ci; I-135, 12 Ci.

E -AVERAGE DISINTEGRATION 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 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.

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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 5

APPLICABILITY: OPERATIONAL CONDITION 5, during loading of fuel assemblies into the core **.

ACTION: r i

With all control rods not fully inserted, immediately suspend loading of fuel l assemblies into the core. The provisions of Specification 3.0.3 are not  ;

applicable. i I

SURVEILLANCE REQUIREMENTS l 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 I

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. i I

3

  • Except control rods removed per Specification 3.9.10.1 or 3.9.10.2. -
    • See Special Test Exception 3.10.3.
  • l  !

l l

i BRUNSWICK - UNIT 1 3/4 9-5 Amendment No.

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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 l 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. {

I 3 /4 . 9 . 2 INSTRUMENTATION I

The OPERABILITY of the source range monitors ensures that redundant I monitoring capability is available to detect changes in the reactivity  !

condition of the core.. j 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 l SRMs will be required to be OPERABLE prior to the SPIP.AL UNLDAD, and each SRM >

will be verified operational by raising the count rate to 3 cps prior to the l 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. ,

t 3/4.9.3 CONTROL ROD POSITION  ;

i l 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 l without a control rod and prevents two positive reactivity changes from j occurring simultaneously. l 3/4.9.4 DECAY TIME ff e

The minimum requirement for reactor suberiticality prior to fuel movement I ensures that sufficient time has elapsed to allow the radioactive decay of the j short lived fission products. This decay time is consistent with the ,

l assumptions used in the accident analyses. l 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the l facility status or core reactivity condition during movement of fuel within l the reactor pressure vessel.

BRUNSWICK - UNIT 1 B 3/4 9-1 Amendment No.

i I.

I

ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKET NOS, 50 325 & 50-324 OPERATING LICENSE NOS. DPR-71 & DPR 62 REQUEST FOR LICENSE AMENDMENT CORE ALTERATIONS i

TYPED TECHNICAL SPECIFICATION PAGES - UNIT 2 l i

l l l

r i

t f

E2-1

.  ?

DNFINITIONS CHANNEL FUNCTIONAL TEST (Continued)

b. Bistable channels - the injection of a simulated signal into tho channel sensor to verify OPERABILITY including alarm and/or trip r functions. l CORE ALTERATION =

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. i Movement of source range monitors, local power-range monitors, intermediate f range monitors, traversing in-core probes, or special moveable detectors  !

(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. t the associated core cell. . Suspension of CORE ALTERATIONS shall not preclude  ;

completion of movement of a component to a safe position. l

\

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 i Specifications 6.9.3.1, 6.9.3.2, 6.9.3.3, and 6.9.3.4. Plant operation within i these core operating limits is addressed in individual specifications.  !

I CRITICAL POWER RATIO '

i 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 assenbly to experience boiling transition, divided by i l

the actual assembly operating power. i DOSE EOUIVALENT I-131  !

90!E EQUIVALENT I-131 shall be concentration of I-131, pCi/ gram, which alone

<suld produce the same thyroid dose as the quantity and isotopic mixture of  ;

l I-131, 1-132, 1-133, 1-134, and I-135 actually present. The following is i 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 l

Sites": I-132, 28 pCi; I-133, 3.7 Ci; I-134, 59 pCi; I-135, 12 pCi.

E -AVERAGE DISINTEGRATION 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 l 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.

r l

i BRUNSWICK - UNIT 2 1-2 Amendment No.

l l

. . - . _ . _ -- __~ , , _ - - , - ., - . ,-- , . _ , . . .

A R5 FUELING OPERATIONS  !

i I

3/4.9.3 CONTROL ROD ?OSITION LIMITING CONDITION FO : OPERATION  !

3.9.3 All control rods shall b'e fully inserted *.

APPLICABILITY: OPERATIONAL CONDITION -5, during loading of fuel assemblies  !

into the core **. I 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 I applicable. j i

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 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 l

l l

l l *Except control rods removed per Specification 3.9.10.1 or 3.9.10.2.

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    • See Special Test Exception 3.10.3.

BRUNSWICK - UNIT 2 3/4 9-5 Amendment No

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1 3/4.9 REFUELING OPERATIONS BASES l

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;  :

damate to reactor internals, fuel assemblies and exposure of personnel to exces;ive radioactivity.

3/4.9.2 INSTRUMENTATION  :

The OPERABILITY of the source range monitors ensures that redundant j monitoring capability is available to detect changes in the reactivity }

condition of the core.  !

l I During a SPIRAL UNLOAD, the count rate of the SRM will decrease below 3 cps l before all of the fuel is unloaded. The count rate of 3 cps is not necessary i since there will be no reactivity additions during the spiral unload. The l

SRMs will be required to be OPERABLE prior to the SPIRAL UNLOAD, and each SRM i will be verified operational by raising the count rate to 3 cps prior to the l SPIRAL RELDAD 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 j the reload.

3/4.9.3 CONTROL ROD POSITION The requirement that all control rods be inserted during loading of fuel j assemblies into the core ensures that fuel will not be loaded into a cell i without a control rod and prevents two positive reactivity changes from occurring simultaneously.  ;

l 3/4.9.4 DECAY TIME I 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 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 i I

facility status or core reactivity condition during movement of fuel within the reactor pressure vessel.

BRUNSWICK - UNIT 2 B 3/4 9-1 Amendment No.

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