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Rev. 21 l                                                             Effective Date       Ir/ Sdp I
Rev. 21 l
Effective Date Ir/ Sdp I
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((               Document Section NFORMAT!ONONL                                 j l_                   C. R. Nuc}sef
((
_ "j ,
Document Section NFORMAT!ONONL j
l l
l_
l SURVEILLANCE PROCEDURE SP-406 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 REFUELING OPERATIONS DAILY DATA REQUIREMENTS THIS PROCEDURE ADDRESSES SAFETY RELATED COMPONENTS l
C. R. Nuc}sef
l l
_ "j l
l l
SURVEILLANCE PROCEDURE SP-406 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 REFUELING OPERATIONS DAILY DATA REQUIREMENTS THIS PROCEDURE ADDRESSES SAFETY RELATED COMPONENTS APPROVED BY:
APPROVED BY:       Inter   ation     tact
Inter ation tact f/ O
    ~
~
f/ O (SIGNATUR[ ON FILE DATE:           /       _
(SIGNATUR[ ON FILE DATE:
INTERPRETATION CONTACT:       Supervisor Operations Engineering 9603110384 960302 PDR ADOCK 05000302 P               PDR
/
INTERPRETATION CONTACT:
Supervisor Operations Engineering 9603110384 960302 PDR ADOCK 05000302 P
PDR


TABLE OF CONTENTS SECTION                                                                                         EaHE 1.0       PURPOSE . . . . . . . . -, . . . . . . . . . . . . . . . . . . .                         I
TABLE OF CONTENTS SECTION EaHE 1.0 PURPOSE........ -,...................
I


==2.0       REFERENCES==
==2.0 REFERENCES==
    ..........................                                                  1 2.1       IMPLEMENTING REFERENCES . . . . . . . . . . . . . . .                         I 2.2       DEVELOPMENTAL REFERENCES               ..............                        1 2.2.1         Technical Specification References                 ....        1 3.0       PERSONNEL INDOCTRINATION           ...................                                    2 3.1       SETP0INTS . . . . . . . .               . . . . . . . . . . . . . . 2
1 2.1 IMPLEMENTING REFERENCES...............
I 2.2 DEVELOPMENTAL REFERENCES 1
2.2.1 Technical Specification References 1
3.0 PERSONNEL INDOCTRINATION 2
3.1 SETP0INTS......................
2


==3.2       DESCRIPTION==
==3.2 DESCRIPTION==
. . . . . . .              .  .  .  .  .  . . . . . .  .  .  . 2 3.3       DEFIN1f!0NS . . . . . . .               . . . . . . . . . . . . . . 2 3.3.1         Core Alteration         . . . . . . . . . . .   . . . 2 3.4       RESPONSIBILITIES ..................                                            3 3.5       LIMITS AND PRECAUTIONS ...............                                        3 3.6       ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . .-                         3 3.7       PREREQUISITES . . ... . . . . . . . . . . . . . . . .'                         3 4.0       INSTRUCTIONS     .........................                                                4 4.1       PRIOR TO ENTERING N00E 6 ..............                                        4 O                 4.2 4.3 4.4 DURING MODE 6 . . . . . . . . . . . . . . . . . . . .
2 3.3 DEFIN1f!0NS.....................
PRIOR TO CORE ALTERATIONS . . . . . . . . . . . . . .
2 3.3.1 Core Alteration..............
DURING CORE ALTERATIONS OTHER THAN FUEL OR 4
2 3.4 RESPONSIBILITIES 3
4 CONTROL R0D NOVEMENT             ................                            4 4.5       DURING FUEL OR CONTROL R0D MOVEMENT WITHIN THE REACTOR VESSEL ...................                                            5 4.6       DEFUELED       ......................                                          5 5.0       FOLLOW-UP' ACTIONS . . . . . . . . . . . . . . . . . . . . . . .                         5 5.1       RESTORATION INSTRUCTIONS ..............                                        5 5.2       CONTINGENCIES . . . . . . . . . . . . . . . . . . . .                           5 ENCLOSURE I     Refueling Operations Shift Data Requirements. . . . . . . . . .                           7
3.5 LIMITS AND PRECAUTIONS 3
                                                                                                    \
3.6 ACCEPTANCE CRITERIA.................-
SP-406                                     Rev. 21                                             Page i J
3 3.7 PREREQUISITES.....................'
3 4.0 INSTRUCTIONS 4
4.1 PRIOR TO ENTERING N00E 6 4
O 4.2 DURING MODE 6....................
4 4.3 PRIOR TO CORE ALTERATIONS..............
4 4.4 DURING CORE ALTERATIONS OTHER THAN FUEL OR CONTROL R0D NOVEMENT 4
4.5 DURING FUEL OR CONTROL R0D MOVEMENT WITHIN THE REACTOR VESSEL 5
4.6 DEFUELED 5
5.0 FOLLOW-UP' ACTIONS.......................
5 5.1 RESTORATION INSTRUCTIONS 5
5.2 CONTINGENCIES....................
5 ENCLOSURE I
Refueling Operations Shift Data Requirements..........
7
\\
SP-406 Rev. 21 Page i J


l                                                           1.0   PURPOSE l
l 1.0 PURPOSE l
l                                                                 The purpose of this procedure is to provide a checklist of Technical               -
l The purpose of this procedure is to provide a checklist of Technical i
i Specification requirements that must be met prior to and during Mode 6, and when Refueling Operations are in progress.
Specification requirements that must be met prior to and during Mode 6, and when Refueling Operations are in progress.
 
==2.0    REFERENCES==


==2.0 REFERENCES==
l l
l l
2.1   INPLEMENTING REFERENCES None i
2.1 INPLEMENTING REFERENCES None i
2.2   DEVELOPMENTAL REFERENCES                                                             ,
2.2 DEVELOPMENTAL REFERENCES 2.2.1 Technical Specification Referencet l
l 2.2.1 Technical Specification Referencet l
LC0/Other Applicable Sury. Perf.
LC0/Other Applicable           Sury. Perf.     Requirements Sury. Freq. Mode References           Durina Modes     Durino Modes Ergg. Notes Notes 3.9.1.1                   5,6             6         SP-1 39 3.9.1.1                     6               6         SP-2           29 3.9.1.1                   5,6             6         SP-3 39 3.9.1.1                   5,6             6         SP-4 3.9.2.1                   5,6             6           S   39 3.9.4.1                   5,6             6           S   .' 9     47 l           3.9.5.1                     6               6           S   39       48 3.9.6.1                     6               6           D   14       29 l
Requirements Sury.
l I
Freq. Mode References Durina Modes Durino Modes Ergg.
SURVEILLANCE FREQUENCY:
Notes Notes 3.9.1.1 5,6 6
SP-1 39 3.9.1.1 6
6 SP-2 29 3.9.1.1 5,6 6
SP-3 39 3.9.1.1 5,6 6
SP-4 3.9.2.1 5,6 6
S 39 3.9.4.1 5,6 6
S
.' 9 47 l
3.9.5.1 6
6 S
39 48 3.9.6.1 6
6 D
14 29 SURVEILLANCE FREQUENCY:
S - At least once per 12 hrs.
S - At least once per 12 hrs.
1 0 - At least once per 24 hrs.
0 - At least once per 24 hrs.
Q - At least once per 92 days.
Q - At least once per 92 days.
SP Prior to removing or unbolting the Reactor Vessel head.
SP Prior to removing or unbolting the Reactor Vessel head.
l SP Prior to the withdrawal of the first safety or regulating rod (Groups 1 thru 7) in excess of 3 ft. from its fully i
l SP Prior to the withdrawal of the first safety or regulating i
I                                                                            inserted position.
rod (Groups 1 thru 7) in excess of 3 ft. from its fully I
inserted position.
SP At least once per 72 hours.
SP At least once per 72 hours.
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l SP-406 Rev. 21 Page 1


i-i                       SP Special frequency or special case when analyses are inadequate.
i-i SP Special frequency or special case when analyses are inadequate.
FREQUENCY NOTES:
FREQUENCY NOTES:
i                       14 -   Within 2 hrs. prior to movement of fuel assemblies or control j                               rods.
i 14 -
4                                                                                     .
Within 2 hrs. prior to movement of fuel assemblies or control j
39 -   Establish surveillance prior to descent into applicable mode.
rods.
;                        MODE NOTES:
4 39 -
29 -   During core alterations.
Establish surveillance prior to descent into applicable mode.
47 -   With the water level 2156 ft.
MODE NOTES:
l       48 -   With the water level < 156 ft.
29 -
2.2.2 FSAR 9.6.2.3 3.0   PERSONNEL IND0CTRINATION 3.1   SETPOINTS None 1                
During core alterations.
 
47 -
==3.2    DESCRIPTION==
With the water level 2156 ft.
l 48 -
With the water level < 156 ft.
2.2.2 FSAR 9.6.2.3 3.0 PERSONNEL IND0CTRINATION 3.1 SETPOINTS None 1


3.2.1 This procedure provides a means for monitoring specific plant conditions and equipment availability prior to and during Refueling Operations as required by Technical Specifications.
==3.2 DESCRIPTION==
3.3   DEFINITIONS                                                                 i
3.2.1 This procedure provides a means for monitoring specific plant conditions and equipment availability prior to and during Refueling Operations as required by Technical Specifications.
                                                                                                    )
3.3 DEFINITIONS
3.3.1 Core Alteration
)
                                                                \
3.3.1 Core Alteration
CORE ALTERATION shall be the movement of any fuel, sources, or other reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
\\
CORE ALTERATION shall be the movement of any fuel, sources, or other reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel.
Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
i
i
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t 3.4   RESPONSIBILITIES O
t 3.4 RESPONSIBILITIES O
<        3.4.1 The CR-3 Operations section shall be responsible for the content of         -
3.4.1 The CR-3 Operations section shall be responsible for the content of l
l              this procedure.
this procedure.
l 3.4.2 The CR-3 Nuclear Shift Supervisor on Duty shall ensure that this l               procedure has been completed as specified by Step 2.2.1.
l 3.4.2 The CR-3 Nuclear Shift Supervisor on Duty shall ensure that this l
l 3.4.3 The CR-3 Reactor Engineer shall be responsible for providing Operations with the most restrictive boron concentration as specified in Step 3.7.2.
procedure has been completed as specified by Step 2.2.1.
3.4.4 This procedure is designed and written to be performed by Nuclear Operators reporting directly to the Shift Supervisor. There are no additional skills required.
l 3.4.3 The CR-3 Reactor Engineer shall be responsible for providing Operations with the most restrictive boron concentration as specified in Step 3.7.2.
3.5   LINITS AND PRECAUTIONS 3.5.1 IE at any time the requirements of Enclosure I cannot be met, IHfH immediately refer to the applicable Technical Specifications (TS) action statement in the TS Action Column.
3.4.4 This procedure is designed and written to be performed by Nuclear Operators reporting directly to the Shift Supervisor. There are no additional skills required.
3.5.2 For work located in Radiation Controlled Areas, due consideration must be given to the ALARA program. This may result in a O'          determination that special preparations and/or precautions are necessary.
3.5 LINITS AND PRECAUTIONS 3.5.1 IE at any time the requirements of Enclosure I cannot be met, IHfH immediately refer to the applicable Technical Specifications (TS) action statement in the TS Action Column.
3.5.3 Boron concentration analysis that deviate   25 ppm from the last analysis should be reverified.
3.5.2 For work located in Radiation Controlled Areas, due consideration O'
3.6   ACCEPTANCE CRITERIA 3.6.1 The data recorded on Enclosure 1 is within the tolerance listed under the " Required" column.
must be given to the ALARA program. This may result in a determination that special preparations and/or precautions are necessary.
3.7   PREREQUISITES 3.7.1 A DHR loop shall be determined to be operating and circulating RC at a flow rate 2 2700 gpm at least once every 12 hours.
3.5.3 Boron concentration analysis that deviate 25 ppm from the last analysis should be reverified.
                                                                                    /
3.6 ACCEPTANCE CRITERIA 3.6.1 The data recorded on Enclosure 1 is within the tolerance listed under the " Required" column.
Initial /Date 3.7.2 Boron concentration is per the Core Operating Limits Report.       Concur i              with Reactor Engineering.
3.7 PREREQUISITES 3.7.1 A DHR loop shall be determined to be operating and circulating RC at a flow rate 2 2700 gpm at least once every 12 hours.
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/
Initial /Date 3.7.2 Boron concentration is per the Core Operating Limits Report.
Concur with Reactor Engineering.
i SP-406 Rev. 21 Page 3


3.7.3 Determine that the boron concentration analyses meet the Acceptance Criteria established by Step 3.7.2.
3.7.3 Determine that the boron concentration analyses meet the Acceptance Criteria established by Step 3.7.2.
                                                                                    /
/
Initial /Date
Initial /Date 3.7.4 Notify the Nuclear Shift Supervisor on Duty prior to the performance of this procedure.
,        3.7.4 Notify the Nuclear Shift Supervisor on Duty prior to the performance of this procedure.
/
                                                                                    /
Initial /Date 3.7.5 Personnel Indoctrination section has been read and understood.
Initial /Date 3.7.5 Personnel Indoctrination section has been read and understood.
/
                                                                                    /
Initial /Date 4.0 INSTRUCTIONS 4.1 fBLQR_ID_IEIf.RIllA_ TIDE _f 4.1.1 Complete Section "A" of Enclosure 1 prior to entering Mode 6.
Initial /Date 4.0   INSTRUCTIONS 4.1   fBLQR_ID_IEIf.RIllA_ TIDE _f 4.1.1 Complete Section "A" of Enclosure 1 prior to entering Mode 6.
4.2 DURING NODE 6 4.2.1 Perform Section "A" of Enclosure 1 once per shift while in Mode 6.
4.2   DURING NODE 6 4.2.1 Perform Section "A" of Enclosure 1 once per shift while in Mode 6.
4.3 PRIOR TO CORE ALTERATIONS 4.3.1 Complete Section "B" of Enclosure I withi9 ONE hour prior to CORE ALTERATIONS.
4.3     PRIOR TO CORE ALTERATIONS 4.3.1 Complete Section "B" of Enclosure I withi9 ONE hour prior to CORE ALTERATIONS.
4.3.2 Verify that the Refueling Canal water level is greater than the 157 ft, elevation within TWO hours prior to FUEL or Control Rod movement.
4.3.2   Verify that the Refueling Canal water level is greater than the 157 ft, elevation within TWO hours prior to FUEL or Control Rod movement.
4.3.3 Complete Section "C" of Enclosure 1 prior to FUEL or Control Rod movement within the Reactor Vessel.
4.3.3   Complete Section "C" of Enclosure 1 prior to FUEL or Control Rod movement within the Reactor Vessel.
4.4 DURINF CORE ALIERATIONS OTHER THAN FUEL OR CONTROL R0D NOVENENT 4.4.1 Perform Sections "A" and "B" of Enclosure 1 once per shift during CORE ALTERATION 3.
4.4     DURINF CORE ALIERATIONS OTHER THAN FUEL OR CONTROL R0D NOVENENT 4.4.1   Perform Sections "A" and "B" of Enclosure 1 once per shift during CORE ALTERATION 3.
SP-406 Rev. 21 Page 4
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4.5     DURING FUEL OR CONTROL R00 NOVENENT WITHIN THE REACTOR VESSEL 4.5.1     Perform Sections "A", "B", and "C" of Enclosure 1 once per shift     ..
4.5 DURING FUEL OR CONTROL R00 NOVENENT WITHIN THE REACTOR VESSEL 4.5.1 Perform Sections "A", "B", and "C" of Enclosure 1 once per shift during FUEL or Control Rod movements within the Reactor Vessel.
;                  during FUEL or Control Rod movements within the Reactor Vessel.
I 4.6 DEFUELED 4.6.1 Shift readings, per this procedure, may be suspended when the i
I         4.6     DEFUELED 4.6.1     Shift readings, per this procedure, may be suspended when the           i 1
1 Reactor is defueled and no fuel elements are in the fuel transfer canal.
Reactor is defueled and no fuel elements are in the fuel transfer canal.
5.0 FOLLOW-UP ACTIONS 1
5.0       FOLLOW-UP ACTIONS                                                       1 l
5.1 RESTORATION INSTRUCTIONS 1
.          5.1       RESTORATION INSTRUCTIONS                                                 1 l
None 5.2 CONTINGENCIES 5.2.1 Refer to Technical Specification Actions under the Action column on.
;                    None 5.2       CONTINGENCIES 5.2.1     Refer to Technical Specification Actions under the Action column on Enclosure 1.
4 i
4 i
P I
P I
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                                                                                                                                                                                . e REFUELING OPERATIO:         T DATA REOUIREMENTS                                         ENCt05tRE 1 SECTION A Pem YO MODE 6 A MD DURANC naODE 6 00-08         08-16       16-24 TS ACTION                                               DESCRIPTION                                       REQUIRED       ACTUAL         ACTU AL     ACTUAL l
e REFUELING OPERATIO:
3.92.1                             NI-1 Source Ranse Test Modde *On Test
T DATA REOUIREMENTS ENCt05tRE 1 SECTION A Pem YO MODE 6 A MD DURANC naODE 6 00-08 08-16 16-24 l
* tamo                             Dim fA 192.1                             N1-2 Source Ranae Test Mod 2le *On Testa tamo                             Dim fA 3.92.1                             N1-1-NI Flux Indication at Control Console                               > 0 4 cos l
TS ACTION DESCRIPTION REQUIRED ACTUAL ACTU AL ACTUAL 3.92.1 NI-1 Source Ranse Test Modde *On Test
392.1                             NI-2-N! Flux InAc2rian at Control Con =ala                               > 0.4 cos l
* tamo Dim fA 192.1 N1-2 Source Ranae Test Mod 2le *On Testa tamo Dim fA l
39.2.1                             NI-14-Ni Flux Indcation at Control Console                               > 10 ' cos""
3.92.1 N1-1-NI Flux Indication at Control Console
l 192.1                             NI-15-NI Flux fn4eation at Control Copenia                               > 10 ' cos*"*
> 0 4 cos l
l 392.1                             NI-1-DNI Rate indication at Control Console                               -0.1 to + 0.1 dom l
392.1 NI-2-N! Flux InAc2rian at Control Con =ala
1921                               Nf-2-DNl Rate Indication at Control Cnnenle                               -0.1 to + 0.1 dom l
> 0.4 cos l
39.2.1                             Source Range Counts Audible in Containment                                             fA 392.1                             Source Ranae Counts AnAhle in Control eaam                                             fA 3 9.4.1: IOC NL94-0039             Decav Heat flow A or B and Circulatina RC                                 > 2700 enm" 19.4.1: IOC NL94M39               Decav Heat Flmv A or B and Cirmlarine RC                                 > S00 enm 3 9.4.1: IOC NL94-0039             Reactor Coolant System Temocrature (DHP %<tian Temocrature)               s 140*F l
39.2.1 NI-14-Ni Flux Indcation at Control Console
Heat Train A M[D 156' lev Q 8 operable OR Refuehng Water Level 2                         g I     19.S.1 39.1.1                             Too of Re6 '" Canal (RB) Baron                                           >    "*                                    N/A         N/A 39.1.1                             Reactor Vessel Boron                                                     2     '"                                    N/A         N/A SECTION E WITHIN 1 HOUR P*EN :TO CORE AsTrmATION 00 3           08-16       16-24 DESCRIPTION                                       REQUIRED       ACTUAL         ACTUAL       ACTUAL FSAR 9 6 2.4                       Main Control Room - Refueline Stations Communications Check                             (A l
> 10 ' cos""
SECHON C Pam TO AND DURING MOWaa :NT Of FtXL OR CONTROL RNS 00-0B         08-16       16-24 TS ACTION                                               DESCRIPTION                                       REQUIRED       ACTUAL         ACTUAL       ACTUAL FSAR 9.6.2.3 19.6.1                             Refuelins Water Level                                                     >157' elev. fA Soent Fuel Pool Boron                                                    "*                                          N/A         N/A i
l 192.1 NI-15-NI Flux fn4eation at Control Copenia
i FSAR 9 6 2.4                       Last Criticality                                                           > 72 hours f4                             N/A         N /A PERFORMED BY:
> 10 ' cos*"*
l NOTES:
l 392.1 NI-1-DNI Rate indication at Control Console
-0.1 to + 0.1 dom l
1921 Nf-2-DNl Rate Indication at Control Cnnenle
-0.1 to + 0.1 dom 39.2.1 Source Range Counts Audible in Containment fA 392.1 Source Ranae Counts AnAhle in Control eaam fA 3 9.4.1: IOC NL94-0039 Decav Heat flow A or B and Circulatina RC
> 2700 enm" 19.4.1: IOC NL94M39 Decav Heat Flmv A or B and Cirmlarine RC
> S00 enm l
3 9.4.1: IOC NL94-0039 Reactor Coolant System Temocrature (DHP %<tian Temocrature) s 140*F Heat Train A M[D 8 operable OR Refuehng Water Level 2 g
Q I
19.S.1 156' lev 39.1.1 Too of Re6 '" Canal (RB) Baron N/A N/A 39.1.1 Reactor Vessel Boron 2
N/A N/A SECTION E WITHIN 1 HOUR P*EN :TO CORE AsTrmATION 00 3 08-16 16-24 DESCRIPTION REQUIRED ACTUAL ACTUAL ACTUAL l
FSAR 9 6 2.4 Main Control Room - Refueline Stations Communications Check (A
SECHON C Pam TO AND DURING MOWaa :NT Of FtXL OR CONTROL RNS 00-0B 08-16 16-24 TS ACTION DESCRIPTION REQUIRED ACTUAL ACTUAL ACTUAL FSAR 9.6.2.3 19.6.1 Refuelins Water Level
>157' elev. fA N/A N/A Soent Fuel Pool Boron i
i FSAR 9 6 2.4 Last Criticality
> 72 hours f4 N/A N /A l
PERFORMED BY:
NOTES:
* Establish surveillance within 2 hours prior to movement. 23 ft. above fuel in the Vessel is equal to 148' Elevation: 23 ft. above fuel in the Spent Fuel Pool is 156' Elevation; 8 ft. above active fuel level of a raised assembly is 157' elevation which is more restrictive.
* Establish surveillance within 2 hours prior to movement. 23 ft. above fuel in the Vessel is equal to 148' Elevation: 23 ft. above fuel in the Spent Fuel Pool is 156' Elevation; 8 ft. above active fuel level of a raised assembly is 157' elevation which is more restrictive.
                ** Demonstrate flow > 2700 gpm once per 12 hours when water level > 156 ft.
** Demonstrate flow > 2700 gpm once per 12 hours when water level > 156 ft.
              *** Boron Concentration to be determined by Reactor Enoineer and as per the COLR.
*** Boron Concentration to be determined by Reactor Enoineer and as per the COLR.
            ****    LF using the Gama-Metrics System (NI-14-NI or NI-IS-NI) to take the place of the BF3 System (NI-1-NI or NI-2-NI),
LF using the Gama-Metrics System (NI-14-NI or NI-IS-NI) to take the place of the BF3 System (NI-1-NI or NI-2-NI),
1N the BF3 System is expected to indicate 10 to 30 times greater than the Gama-Metrics System.
1N the BF3 System is expected to indicate 10 to 30 times greater than the Gama-Metrics System.
1t not, contact Systems Engineering (
1t not, contact Systems Engineering (
Line 151: Line 210:
==Reference:==
==Reference:==
Technical Support letter NPTS 96-0107).
Technical Support letter NPTS 96-0107).
t SP-406                                                                             Rev. 21                                               Page 6 (LAST PAGE)}}
t SP-406 Rev. 21 Page 6 (LAST PAGE)}}

Latest revision as of 06:25, 13 December 2024

Rev 21 to Surveillance Procedure SP-406, Refueling Operations Daily Data Requirements
ML20100P856
Person / Time
Site: Crystal River 
Issue date: 03/01/1996
From: Vogel K
FLORIDA POWER CORP.
To:
Shared Package
ML20100P811 List:
References
SP-406, NUDOCS 9603110384
Download: ML20100P856 (8)
v • d • e


Text

.

Rev. 21 l

Effective Date Ir/ Sdp I

i

((

Document Section NFORMAT!ONONL j

l_

C. R. Nuc}sef

_ "j l

SURVEILLANCE PROCEDURE SP-406 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 REFUELING OPERATIONS DAILY DATA REQUIREMENTS THIS PROCEDURE ADDRESSES SAFETY RELATED COMPONENTS APPROVED BY:

Inter ation tact f/ O

~

(SIGNATUR[ ON FILE DATE:

/

INTERPRETATION CONTACT:

Supervisor Operations Engineering 9603110384 960302 PDR ADOCK 05000302 P

PDR

TABLE OF CONTENTS SECTION EaHE 1.0 PURPOSE........ -,...................

I

2.0 REFERENCES

1 2.1 IMPLEMENTING REFERENCES...............

I 2.2 DEVELOPMENTAL REFERENCES 1

2.2.1 Technical Specification References 1

3.0 PERSONNEL INDOCTRINATION 2

3.1 SETP0INTS......................

2

3.2 DESCRIPTION

2 3.3 DEFIN1f!0NS.....................

2 3.3.1 Core Alteration..............

2 3.4 RESPONSIBILITIES 3

3.5 LIMITS AND PRECAUTIONS 3

3.6 ACCEPTANCE CRITERIA.................-

3 3.7 PREREQUISITES.....................'

3 4.0 INSTRUCTIONS 4

4.1 PRIOR TO ENTERING N00E 6 4

O 4.2 DURING MODE 6....................

4 4.3 PRIOR TO CORE ALTERATIONS..............

4 4.4 DURING CORE ALTERATIONS OTHER THAN FUEL OR CONTROL R0D NOVEMENT 4

4.5 DURING FUEL OR CONTROL R0D MOVEMENT WITHIN THE REACTOR VESSEL 5

4.6 DEFUELED 5

5.0 FOLLOW-UP' ACTIONS.......................

5 5.1 RESTORATION INSTRUCTIONS 5

5.2 CONTINGENCIES....................

5 ENCLOSURE I

Refueling Operations Shift Data Requirements..........

7

\\

SP-406 Rev. 21 Page i J

l 1.0 PURPOSE l

l The purpose of this procedure is to provide a checklist of Technical i

Specification requirements that must be met prior to and during Mode 6, and when Refueling Operations are in progress.

2.0 REFERENCES

l l

2.1 INPLEMENTING REFERENCES None i

2.2 DEVELOPMENTAL REFERENCES 2.2.1 Technical Specification Referencet l

LC0/Other Applicable Sury. Perf.

Requirements Sury.

Freq. Mode References Durina Modes Durino Modes Ergg.

Notes Notes 3.9.1.1 5,6 6

SP-1 39 3.9.1.1 6

6 SP-2 29 3.9.1.1 5,6 6

SP-3 39 3.9.1.1 5,6 6

SP-4 3.9.2.1 5,6 6

S 39 3.9.4.1 5,6 6

S

.' 9 47 l

3.9.5.1 6

6 S

39 48 3.9.6.1 6

6 D

14 29 SURVEILLANCE FREQUENCY:

S - At least once per 12 hrs.

0 - At least once per 24 hrs.

Q - At least once per 92 days.

SP Prior to removing or unbolting the Reactor Vessel head.

l SP Prior to the withdrawal of the first safety or regulating i

rod (Groups 1 thru 7) in excess of 3 ft. from its fully I

inserted position.

SP At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

l SP-406 Rev. 21 Page 1

i-i SP Special frequency or special case when analyses are inadequate.

FREQUENCY NOTES:

i 14 -

Within 2 hrs. prior to movement of fuel assemblies or control j

rods.

4 39 -

Establish surveillance prior to descent into applicable mode.

MODE NOTES:

29 -

During core alterations.

47 -

With the water level 2156 ft.

l 48 -

With the water level < 156 ft.

2.2.2 FSAR 9.6.2.3 3.0 PERSONNEL IND0CTRINATION 3.1 SETPOINTS None 1

3.2 DESCRIPTION

3.2.1 This procedure provides a means for monitoring specific plant conditions and equipment availability prior to and during Refueling Operations as required by Technical Specifications.

3.3 DEFINITIONS

)

3.3.1 Core Alteration

\\

CORE ALTERATION shall be the movement of any fuel, sources, or other reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel.

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

i

O SP-406 Rev. 21 Page 2

t 3.4 RESPONSIBILITIES O

3.4.1 The CR-3 Operations section shall be responsible for the content of l

this procedure.

l 3.4.2 The CR-3 Nuclear Shift Supervisor on Duty shall ensure that this l

procedure has been completed as specified by Step 2.2.1.

l 3.4.3 The CR-3 Reactor Engineer shall be responsible for providing Operations with the most restrictive boron concentration as specified in Step 3.7.2.

3.4.4 This procedure is designed and written to be performed by Nuclear Operators reporting directly to the Shift Supervisor. There are no additional skills required.

3.5 LINITS AND PRECAUTIONS 3.5.1 IE at any time the requirements of Enclosure I cannot be met, IHfH immediately refer to the applicable Technical Specifications (TS) action statement in the TS Action Column.

3.5.2 For work located in Radiation Controlled Areas, due consideration O'

must be given to the ALARA program. This may result in a determination that special preparations and/or precautions are necessary.

3.5.3 Boron concentration analysis that deviate 25 ppm from the last analysis should be reverified.

3.6 ACCEPTANCE CRITERIA 3.6.1 The data recorded on Enclosure 1 is within the tolerance listed under the " Required" column.

3.7 PREREQUISITES 3.7.1 A DHR loop shall be determined to be operating and circulating RC at a flow rate 2 2700 gpm at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

/

Initial /Date 3.7.2 Boron concentration is per the Core Operating Limits Report.

Concur with Reactor Engineering.

i SP-406 Rev. 21 Page 3

3.7.3 Determine that the boron concentration analyses meet the Acceptance Criteria established by Step 3.7.2.

/

Initial /Date 3.7.4 Notify the Nuclear Shift Supervisor on Duty prior to the performance of this procedure.

/

Initial /Date 3.7.5 Personnel Indoctrination section has been read and understood.

/

Initial /Date 4.0 INSTRUCTIONS 4.1 fBLQR_ID_IEIf.RIllA_ TIDE _f 4.1.1 Complete Section "A" of Enclosure 1 prior to entering Mode 6.

4.2 DURING NODE 6 4.2.1 Perform Section "A" of Enclosure 1 once per shift while in Mode 6.

4.3 PRIOR TO CORE ALTERATIONS 4.3.1 Complete Section "B" of Enclosure I withi9 ONE hour prior to CORE ALTERATIONS.

4.3.2 Verify that the Refueling Canal water level is greater than the 157 ft, elevation within TWO hours prior to FUEL or Control Rod movement.

4.3.3 Complete Section "C" of Enclosure 1 prior to FUEL or Control Rod movement within the Reactor Vessel.

4.4 DURINF CORE ALIERATIONS OTHER THAN FUEL OR CONTROL R0D NOVENENT 4.4.1 Perform Sections "A" and "B" of Enclosure 1 once per shift during CORE ALTERATION 3.

SP-406 Rev. 21 Page 4

4.5 DURING FUEL OR CONTROL R00 NOVENENT WITHIN THE REACTOR VESSEL 4.5.1 Perform Sections "A", "B", and "C" of Enclosure 1 once per shift during FUEL or Control Rod movements within the Reactor Vessel.

I 4.6 DEFUELED 4.6.1 Shift readings, per this procedure, may be suspended when the i

1 Reactor is defueled and no fuel elements are in the fuel transfer canal.

5.0 FOLLOW-UP ACTIONS 1

5.1 RESTORATION INSTRUCTIONS 1

None 5.2 CONTINGENCIES 5.2.1 Refer to Technical Specification Actions under the Action column on.

4 i

P I

O SP-406 Rev. 21 Page 5

e REFUELING OPERATIO:

T DATA REOUIREMENTS ENCt05tRE 1 SECTION A Pem YO MODE 6 A MD DURANC naODE 6 00-08 08-16 16-24 l

TS ACTION DESCRIPTION REQUIRED ACTUAL ACTU AL ACTUAL 3.92.1 NI-1 Source Ranse Test Modde *On Test

  • tamo Dim fA 192.1 N1-2 Source Ranae Test Mod 2le *On Testa tamo Dim fA l

3.92.1 N1-1-NI Flux Indication at Control Console

> 0 4 cos l

392.1 NI-2-N! Flux InAc2rian at Control Con =ala

> 0.4 cos l

39.2.1 NI-14-Ni Flux Indcation at Control Console

> 10 ' cos""

l 192.1 NI-15-NI Flux fn4eation at Control Copenia

> 10 ' cos*"*

l 392.1 NI-1-DNI Rate indication at Control Console

-0.1 to + 0.1 dom l

1921 Nf-2-DNl Rate Indication at Control Cnnenle

-0.1 to + 0.1 dom 39.2.1 Source Range Counts Audible in Containment fA 392.1 Source Ranae Counts AnAhle in Control eaam fA 3 9.4.1: IOC NL94-0039 Decav Heat flow A or B and Circulatina RC

> 2700 enm" 19.4.1: IOC NL94M39 Decav Heat Flmv A or B and Cirmlarine RC

> S00 enm l

3 9.4.1: IOC NL94-0039 Reactor Coolant System Temocrature (DHP %<tian Temocrature) s 140*F Heat Train A M[D 8 operable OR Refuehng Water Level 2 g

Q I

19.S.1 156' lev 39.1.1 Too of Re6 '" Canal (RB) Baron N/A N/A 39.1.1 Reactor Vessel Boron 2

N/A N/A SECTION E WITHIN 1 HOUR P*EN :TO CORE AsTrmATION 00 3 08-16 16-24 DESCRIPTION REQUIRED ACTUAL ACTUAL ACTUAL l

FSAR 9 6 2.4 Main Control Room - Refueline Stations Communications Check (A

SECHON C Pam TO AND DURING MOWaa :NT Of FtXL OR CONTROL RNS 00-0B 08-16 16-24 TS ACTION DESCRIPTION REQUIRED ACTUAL ACTUAL ACTUAL FSAR 9.6.2.3 19.6.1 Refuelins Water Level

>157' elev. fA N/A N/A Soent Fuel Pool Boron i

i FSAR 9 6 2.4 Last Criticality

> 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> f4 N/A N /A l

PERFORMED BY:

NOTES:

  • Establish surveillance within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to movement. 23 ft. above fuel in the Vessel is equal to 148' Elevation: 23 ft. above fuel in the Spent Fuel Pool is 156' Elevation; 8 ft. above active fuel level of a raised assembly is 157' elevation which is more restrictive.
    • Demonstrate flow > 2700 gpm once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when water level > 156 ft.
      • Boron Concentration to be determined by Reactor Enoineer and as per the COLR.

LF using the Gama-Metrics System (NI-14-NI or NI-IS-NI) to take the place of the BF3 System (NI-1-NI or NI-2-NI),

1N the BF3 System is expected to indicate 10 to 30 times greater than the Gama-Metrics System.

1t not, contact Systems Engineering (

Reference:

Technical Support letter NPTS 96-0107).

t SP-406 Rev. 21 Page 6 (LAST PAGE)

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