ML093430196
| ML093430196 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/07/2009 |
| From: | NRC/RGN-II/DRS/OLB |
| To: | Southern Nuclear Operating Co |
| References | |
| 50-321/09-302, 50-366/09-302 | |
| Download: ML093430196 (292) | |
Text
FACILITY NAME: _.....;..H.;;.;.A..:..;T~C;;..;..H.=.....-
_....;;..H.;.;..A..:::..;T...;:C;..:.,H=----_
_ _ _ _ _ _ _ _ ___ Section 6 REPORT NUMBER: 05000321/2009302 AND 05000366/2009302 DRAFT ADMINISTRATIVE JPMS CONTENTS:
D Draft ADMIN JPMs Location of Electronic Files:
O:\Hatch Examinations\lnitial Exam 2009-302 Submitted By: Bruno Caballero Verified By: _ _ _ _ _ _ _ __
DRAFT Southern Nuclear E. I. Hatch Nuclear Plant Operations Training JPM ADMIN!
ADMINl
( TITLE Determine Correct Method of Independent Verification for a System Lineup AUTHOR MEDIA NUMBER TIME F.N.FAGAN HLT5-Admin 1 15 Minutes RECOMMENDED BY APPROVED BY DATE NIR SOUTHERN . .\
COMPANY sM Energy to Serve Your World SM
SOUTHERN NUCLEAR OPERATING COMPANY PLANT E. I. HATCH Page 1 ofl of 1 FORM TITLE: TRAINING MATERIAL REVISION SHEET Program/Course OPERATIONS TRAINING Media Code: Number:
Rev. No. Date Reason for Revision Author's Supv's Initials Initials 00 00 Initial development FNF
(
(
HLT5-Admin 1 UNIT 1 (X) UNIT 2 (X)
TASK TITLE: Determine Correct Method of Independent Verification for a System Lineup JPMNUMBER:
TASK STANDARD: The task shall be complete when the operator has completed identifying the independent verification process for the listed valves.
TASK NUMBER:
OBJECTIVE NUMBER: LT-LP-3004, 300.022A, 300.022.a.02 PLANT HATCH JTA IMPORTANCE RATING:
( KIA CATALOG NUMBER: 2.1.29 CATALOGJTA KIA CATALOG JTA IMPORTANCE RATING:
RO 4.1 SRO 4.0 OPERATOR APPLICABILITY: Reactor Operator (RO)
IGENERAL
REFERENCES:
Unit 1 & 2 34GO-SUV-00 -001-0 "Control And Surveillance Of Locked Valves, Lock Wired Valves, And Locked Breakers" NMP-OS-002 "Verification Policy" IREQUIRED MATERIALS: Unit 1 & 2 34GO-SUV-00I-0 "Control And Surveillance Of Locked Valves, Lock Wired Valves, And Locked Breakers" NMP-OS-002 "Verification Policy" APPROXIMATE COMPLETION TIME: 15 Minutes SIMULATOR SETUP: NIA N/A
UNITl&2 READ AND PROVIDE A COpy TO THE OPERATOR INITIAL CONDITIONS:
perronned on system X.
- 1. A valve checklist was performed
- 2. All valves had a concurrent verifier.
- 3. All valves have no remote position indication.
- 4. The checklist was completed and the system was placed in service
- 5. Then the SS decided that an Independent Verification should be performed perfonned on some
( of the valves.
INITIATING CUES:
IAW 34GO-SUV-OOI-O lAW 34GO-SUV-OOl-O "Control And Surveillance Of Locked Valves, Lock Wired Valves, And Locked Breakers" and IAW NMP-OS-002 "Verification Policy",
lAW select the proper method for Independent Verification of the checklist valves.
(
Examiner Key
- A. Open Air Operated Flow Control Valve.
Answer: #2, Perform Verification by observing flow indication downstream of valve.
- B. Closed Manual valve Answer: #5, Turn the valve handwheel or attempt to turn in the Closed direction.
- c.
- C. Manual Globe valve Throttled (3 turns Open)
Answer: #6, This valve should NOT be Independently Verified (The concurrent verification already performed was adequate).
- D. Locked Closed Knife valve (full stroke in one quarter turn).
Answer: #8, Determine position by visual inspection of the valve handlelhandwheel orientation.
- E Open Manual valve.
Answer: #4, Turn the valve handwheel in the Closed direction and verify the valve stem moves and return the valve to the Open position.
(
Candidate Handout A valve checklist was performed on system X. All valves had a concurrent verifier. All valves have no remote position indication.
The checklist was completed and the system was placed in service. Then the SS decided that an Independent Verification should be performed on some of the valves.
lAW 34GO-SUV-OOI-O 34GO-SUV-001-0 "Control And Surveillance Of Locked Valves, Lock Wired Valves, And Locked Breakers" and NMP-OS-002 "Verification Policy", select the proper method for Independent Verification of the checklist valves.
CHECKLIST VALYES VERIFICATION PROCESS A. Open Air Operated Flow Control Valve 1. Perform Verification by observing Red and Green lights B. Closed Manual valve indication on Control Room panel.
C. Manual Globe valve Throttled (3 turns Open) 2. Perform Verification by observing flow indication downstream of valve.
D. Locked Closed Knife valve (full stroke in one quarter turn). 3. Turn the valve handwheel in the Open direction and verify valve stem does NOT move.
E Open Manual valve.
- 4. Turn the valve handwheel in the Closed direction and verify the valve stem moves and return the A. _ _ valve to the Open position.
For the valves (A-E): S.
- 5. Turn the valve handwheel or attempt to turn in the Closed B.
B. _ _ Enter a single number (1-8)
(I-S) that corresponds direction.
to the correct verification process used to 6. This valve should NOT be Independently Verified (The
- c. _
C. __ _ _ verify the valve position and place it in the concurrent verification already performed was space provided. If valve verification can be adequate).
D. _ _ done with a preferred method and a non- Turn the valve handwheel in the Open direction 7.
preferred method, select the PREFERRED and verify the valve stem moves and return the E. _
__ _ _ method. (The numbers to the right may be valve to the Closed position.
used more than once or not at all).
S.
- 8. Determine position by visual inspection of the valve F. _ _ handlelhandwheel handle/handwheel orientation.
I
Candidate Handout Flow Element Pump Heat b Exchanger l .~--
Valve B (Closed)
Valve E ValveE (Open)
--.,.;-~
,I Valve 0 D /
Closed)
(Locked ClOsed) Valve C (Throttled) r=
I Drain P T
UNITl&2 READ AND PROVIDE A COpy TO THE OPERATOR INITIAL CONDITIONS:
- 1. A valve checklist was perfonned on system X.
- 2. All valves had a concurrent verifier.
- 3. All valves have no remote position indication.
- 4. The checklist was completed and the system was placed in service
- 5. Then the SS decided that an fudependent Independent Verification should be perfonned on some
( of the valves.
INITIATING CUES:
IAW 34GO-SUV-001-0 "Control And Surveillance Of Locked Valves, Lock Wired Valves, And Locked Breakers" and IAW NMP-OS-002 "Verification Policy",
select the proper method for fudependent Independent Verification of the checklist valves.
SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH GENERAL OPERATING PROCEDURE 1 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR G. L. Johnson DATE 02/10/06 DATE:
N/A NPGM / POAGM / 12-11-08 PSAGM N/A DATE N/A 1.0 OBJECTIVE This procedure establishes the controls and surveillance which shall be used to ensure that the following components are maintained in their proper positions:
- LOCKED VALVES
- LOCKED WIRED VALVES
- LOCKED BREAKERS TABLE OF CONTENTS Section 2.0 APPLiCABILITy .................................................................................................................. 2
3.0 REFERENCES
................................................................................................................... 2
( 4.0 REQUIREMENTS ............................................................................................................... 4 5.0 PRECAUTIONS/LIMITATIONS ........................................................................................... 5 6.0 PREREQUISITES ............................................................................................................... 5 7.0 PROCEDURE ..................................................................................................................... 6 7.1 VALVE CONTROL AND SURVEILLANCE INFORMATION .......................................... 6 7.2 ADMINISTRATIVE CONTROL OF VALVE MANIPULATION AND RESTORATION ..... 9 7.2.1 Valve Manipulation ................................................................................................. 9 7.2.2 Restoration .......................................................................................................... 11 7.2.3 Independent Verification ...................................................................................... 12 7.2.4 Locked Valve Logbook ......................................................................................... 13 7.3 LOCKED VALVE SURVEILLANCE ............................................................................. 14 7.4 BREAKER CONTROL AND SURVEILLANCE INFORMATION ................................... 16 7.5 ADMINISTRATIVE CONTROL OF BREAKER MANIPULATION AND RESTORATION .......................................................................................................... 18 7.5.1 Breaker Manipulation ........................................................................................... 18 7.5.2 Restoration .......................................................................................................... 20 7.5.3 Independent Verification ...................................................................................... 21 7.5.4 Locked Breaker Logbook ..................................................................................... 22 7.6 LOCKED BREAKER SURVEILLANCE ....................................................................... 23
(
MGR-0002 Ver. 8
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 2 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 2.0 APPLICABILITY This procedure is applicable to all breakers and valves that are locked in position. Instrument valves and fire protection valves controlled by the following procedures are exempt from this procedure: 51 GM-SPR-001-0, 42SV-FPX-035-1, 42SV-FPX-035-2, 42SV-FPX-016-1, 42SV-FPX-017 -1, and 42SV-FPX-017-2.
42SV-FPX-016-2, 42SV-FPX-017-1, 42SV-FPX-017 -2. This procedure also applies to those valves lock wired in position due to Event Review Team 89-006 recommendations. Locked breaker surveillance shall be performed quarterly. Lock wired valve surveillance shall be performed semi-annually. Locked valve surveillance will be performed at the discretion of Operations Management.
This procedure in itself does NOT constitute authority to reposition any valves OR breakers. It is NOTE: intended to be used as an administrative control and will always be used in conjunction with approved operating procedures, surveillance procedures, system clearances OR other approved plant procedures to reposition any valves OR breakers.
3.0 REFERENCES
3.1 NMP-GM-002, Corrective Actions program
(., 3.2 30AC-OPS-003-0, Plant Operations 3.3 80AC-SEC-002-0, Key and Annunciated Door Control 3.4 42SV-FPX-035-1, Fire Protection Valve Cycling Surveillance 3.5 42SV-FPX-035-2, Fire Protection Valve Cycling Surveillance 3.6 51GM-SPR-001-0, Sealing of Instrument Valves 3.7 Unit 1 FSAR Table 7.3-1 3.8 Unit 2 FSAR Table 6.2-5 3.9 Unit 1 and Unit 2 P&IDs 3.10 10CFR50 Appendix A, Criteria 55, 56 and 57 3.11 ANSI-ANS 59.1-1979, Safety Related Cooling Water Systems in Nuclear Power Plants.
3.12 QATR 3.13 ASME Standard No. 116, Recommended Practices for the Design of Steam Turbine Generator Oil Systems.
3.14 ASME Standard No. TWDPS-1 Part 2, Recommended Practices for the Prevention of Water Damage to Steam Turbines Used for Electrical Power Generation.
I 3.15 IER 80-02
\
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 3 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 3.16 IER 82-10 3.17 IER 83-07 3.18 IER 83-15 3.19 IER 83-26 3.20 SER 57-82 3.21 SER 75-82 3.22 SER 51-83 3.23 LER 76-045 3.24 LER 80-045 3.25 LER 80-061 3.26 LER 81-057 3.27 LER 81-072
( 3.28 LER 83-049 3.29 LER 83-096 3.30 NRC Enforcement Action 82-137 3.31 Full Size Forms
- OPS-0018, Locked Valve Manipulation Log
- OPS-0021, Locked Valve Manipulation Form
- OPS-0071, Unit 1 Accessible Locked Valve Surveillance
- OPS-0072, Unit 2 Accessible Locked Valve Surveillance
- OPS-0073, Unit 1 Inaccessible Locked Valve Surveillance
- OPS-0069, Unit 2 Inaccessible Locked Valve Surveillance
- OPS-0443, Unit 1 Locked Valves
- OPS-0444, Unit 2 Locked Valves
- OPS-0662, Locked Breaker Manipulation Log
- OPS-0659, Locked Breaker Manipulation Log
- OPS-0660, Locked Breaker Surveillance
- OPS-1697, Locked Breakers
- OPS-1046, Non-Power Block Lock Wired Valve Surveillance
- OPS-1047, Control Building And Turbine Building Lock Wired Valve Surveillance
- OPS-1048, Reactor Building And Radwaste Building Lock Wired Valve Surveillance
- OPS-1049, Inaccessible Lock Wired Valve Surveillance
- OPS-1698, Unit 1 Lock Wired Valves
- OPS-1699, Unit 2 Lock Wired Valves MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 4 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 3.32 NQA-1-1994 4.0 REQUIREMENTS 4.1 PERSONNEL REQUIREMENTS Only personnel who have received training in locked valve and system restoration requirements are allowed to perform restoration, independent verification and surveillance per this procedure.
4.2 MATERIAL AND EQUIPMENT 4.2.1 Keys for locked valves 4.2.2 Keys for locked breakers 4.3 SPECIAL REQUIREMENTS 4.3.1 Independent verification as defined in 10AC-MGR-019-0, Procedure Use And Adherence, is required following restoration whenever a locked valve OR breaker, OR a lock wired valve is manipulated.
(
4.3.2 Report Condition Reports as required by NMP-GM-002, Corrective Actions Program.
4.3.3 Obtain a Radiation Work permit AND an HP escort, as required.
4.3.4 IE a locked valve OR breaker is deleted from or added to this procedure, a CR must be initiated specifying that any plant drawing showing the valve or breaker be changed in accordance with approved plant procedures.
4.3.5 Steel wire will be used for locking Lock Wired Valves.
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 5 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 5.0 PRECAUTIONS/LIMITATIONS 5.1 PRECAUTIONS 5.1.1 Observe signs AND warnings of industrial hazards.
5.1.2 Observe signs AND warnings of radiation hazards.
5.1.3 Observe the requirements of Southern Nuclear Safety and Health Manual.
5.1.4 Follow proper radiation protection practices/procedures to maintain exposure ALARA AND limit the spread of contamination. Remain alert for changing conditions which may require additional radiation protection.
5.1.5 Some of these valves and breakers are located in high radiation areas AND areas of infrequent travel. Contact HP for current conditions AND an escort, IF required.
5.2 LIMITATIONS 5.2.1 WHEN significant health/safety hazards (e.g., high radiation, high temperature, etc.) are likely to be encountered during performance of independent verification, the verification
( can be waived by the Shift Supervisor. However, in these situations, an alternate means for independent verification that does NOT involve exposure to hazards shall be used, !E available. Any independent verification that is waived shall be documented per step 7.2.3.1.1.
5.2.2 significant health/safety hazards (e.g., high radiation, high temperature, etc.) are likely WHEN Significant to be encountered during performance of surveillance, surveillance can be waived by the Shift Supervisor. Document waived surveillance per step 7.3.4.1.
6.0 PREREQUISITES PRIOR to checking breaker locking device operability, confirm that the associated valve control switch is positioned to match the current valve position.
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 6 OF 23 60F DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 7.0 PROCEDURE 7.1 VALVE CONTROL AND SURVEILLANCE INFORMATION CONTINUOUS I
- Lock wired valves are subject to the same administrative controls as locked valves. The only difference is the device used to secure the valve handle OR handwheel. Therefore, the requirements of this procedure that apply to locked valves also apply to lock wired valves.
- In addition to the use of Steel Wire to secure "Lock wired valves", Operations Management may elect to use a "keeper" device over a knife valve handle to secure it in the specified position. Where a "keeper" device is used, a footnote will be provided in the Form.
NOTES:
- Valves are locked for a variety of reasons including prevention of unauthorized manipulation AND protection of personnel AND equipment. Other valves are locked for operational concerns AND as required by various commitments. Therefore NOT all locked valves will have the locked designation shown on the P&ID.
- Valve positions given in this procedure reflect the normal configuration as established by
( system valve lineups.
7.1 .1 7.1.1 The Shift Supervisor Su pervisor shall maintain administrative ad ministrative control of the positions of locked valves.
7.1.2 Locked valves shall be maintained locked in the position specified in Forms OPS-0443, and OPS-0444 except as covered by steps 7.1.5, 7.1.6.4, and 7.2.
7.1.3 Lock wired valves shall be maintained lock wired in the position specified in Forms Ops 1698 and Ops-1699 except as covered in steps 7.1.5, 7.1.6.4, and 7.2.
7.1.5,7.1.6.4, 7.1.4 Valve keys are controlled in accordance with 80AC-SEC-002-0.
7.1.4.1 Under normal conditions, valve keys may only be checked out to personnel performing surveillance per this procedure OR those in possession of a Locked Valve Manipulation Form OR as permitted by step 7.2.1.1 approved by the Shift Supervisor.
7.1.4.2 In an emergency, keys may be checked out with verbal approval from the Shift Supervisor OR Shift Manager AND the Locked Valve Manipulation Form initiated later.
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 7 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 7.1.5 Confirmation of valve position for independent verification OR sUNeiliance surveillance shall be performed as follows:
IIlcAUTION:
CAUTION: FOR VALVES REQUIRED TO BE THROTTLED, DO NOT PERFORM CONFIRMATION OF III POSITION. PERFORM ONLY LOCKING DEVICE OPERABILITY PER STEP 7.1.6.
7.1.5.1 lE the valve is a knife valve (full stroke in one quarter turn), determine position from the valve handle orientation. Proceed to step 7.1.5.4.
7.1.5.2 Unlock the valve, as necessary.
7.1.5.2.1 lE the valve is required to be closed, turn the handwheel in the "close" direction. The handwheel will normally turn no more than one quarter turn to seat the valve.
7.1.5.2.2 lE IF the valve is required to be open, turn the handwheel in the "close" direction NO MORE THAN ONE QUARTER TURN, THEN return the valve to the full open position.
(
NOTE: Valves found out of position during surveillance, sUNeiliance, but whose position is accurately reflected on an active Locked Valve Manipulation Form, do NOT constitute a Condition Report.
7.1.5.3 IF the valve is found in a position other than the "Required Position", contact the Shift Supervisor for guidance.
SupeNisor 7.1.5.4 Relock the valve, lE unlocked by 7.1.5.2.
7.1.6 Confirmation of locking device operability shall be performed as follows:
7.1.6.1 Visually check the integrity of the locking device.
7.1.6.2 ObseNe Observe that the locking device is attached securely to both the valve handle/handwheel AND to its anchor point.
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 8 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS DO NOT ATTEMPT TO REPOSITION "LOCKED CLOSED" VALVES TO CONFIRM CAUTION:
LOCKING DEVICE OPERABILITY.
7.1.6.3 !E IF a chain OR lock wire is used, ensure that it is taut AND is attached so that it would NOT loosen !E IF the handwheel were rotated in the direction necessary to reposition the valve.
7.1.6.4 IF the locking device is inoperable, attempt to make it operable by changing the locking device AND repeat steps 7.1.6.1 through 7.1.6.3.
7.1.6.5 Notify the Shift Supervisor of failures of steps 7.1.6.1 through 7.1.6.3.
(
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 9 OF 23
(
DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 7.2 ADMINISTRATIVE CONTROL OF VALVE MANIPULATION AND RESTORATION 7.2.1 VALVE MANIPULATION
[CONTINUOUS CONTINUOUS I 7.2.1.1 A Locked Valve Manipulation Form (LVMF) as shown in OPS-0021 shall be used to document unlocking a valve's locking device except:
- 1) During surveillance required by subsection 7.3
- 2) WHEN another administrative control documents proper manipulation (e.g.
clearance ).
- 3) For the Radwaste Operator in using keys issued him by 80AC-SEC-002-0.
7.2.1.2 Personnel needing a valve unlocked shall request a LVMF from the Shift Supervisor.
NOTE: OIIOC~~ 2, E'II Block key number, will be marked "N/A" for lock wired valves.
L., Key 7 .2.1.2.1 7.2.1.2.1 The requestor shall complete Blocks 2 through 5 on the LVMF AND discuss with the Shift Supervisor the desired manipulation AND the position/condition in which the valve is to be left.
7.2.1.2.2 The Shift Supervisor shall:
7.2.1.2.2.1 Have the present position/condition of the valve determined from the Locked Valve Logbook.
7.2.1.2.2.2 Have amplifying instructions, as necessary, entered on Block 6 of the LVMF.
7.2.1.2.2.3 Approve the manipulation by completing Block 7 of the LVMF.
7.2.1.2.2.4 Have personnel assigned to perform manipulation.
7.2.1.2.2.5 Have the valve added to the Locked Valve Manipulation Log (LVML) by completion of Blocks 1 through 3 of the LVML and Block 1 of the LVMF.
7.2.1.2.2.6 !E IF the assigned personnel is to manipulate AND return the valve to its "Required Position", restoration can also be assigned at this time by performing steps 7.2.2.1 through 7.2.2.1.3.
7.2.1.2.2.7 Have the LVMF given to the assigned personnel.
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 10 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 7.2.1.3 The assigned personnel shall:
7.2.1.3.1 Check out the valve key, IF applicable, by showing the LVMF to the Shift Support Supervisor.
7.2.1.3.2 Manipulate the valve in accordance with the LVMF. Contact the Shift Supervisor for guidance!E unexpected conditions are encountered.
7.2.1.3.3 Complete Blocks 8,9, 8, 9, and 10 of the LVMF.
7.2.1.3.4 IF restoration is approved, proceed to step 7.2.2.2.2.
7.2.1.3.5 Secure the locking device:
7.2.1.3.5.1 By locking the valve, !E the "As Left" position (Block 9) is the same as the "Required Position" (Block 3);
( CAUTION: DO NOT INHIBIT OPERATION OF ADJACENT EQUIPMENT IN PERFORMANCE OF STEP 7.2.1.3.5.2.
7.2.1.3.5.2 By locking the device to adjacent equipment OR piping.
7.2.1.3.6 Return the valve key, !E applicable, to the Shift Support Supervisor.
7.2.1.3.7 Return the LVMF to the Shift Supervisor to assign restoration OR place in the Locked Valve Logbook.
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 11 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 7.2.2 RESTORATION CONTINUOUS I 7.2.2.1 The Shift Supervisor shall:
7.2.2.1.1 Approve restoration by completing Block 11 on the LVMF.
7.2.2.1.2 Have personnel assigned to perform restoration.
7.2.2.1.3 IF the valve is located in an area that could cause ALARA concerns THEN independent verification may also be assigned aSSigned at this time by performing steps 7.2.3.1 through 7.2.3.1.3.
7.2.2.1.4 Have the LVMF given to the aSSigned personnel.
7.2.2.2 The assigned personnel shall:
7.2.2.2.1 Check out the key, IE applicable, by showing the LVMF to the Shift Support Supervisor.
(
7.2.2.2.2 Position the valve in its "Required Position" (Block 3). Contact the Shift Supervisor for guidance IF unexpected conditions are encountered.
7.2.2.2.3 IInstall nstall the locking device so it is operable per step 7.1.6.
7.2.2.2.4 Complete Blocks 12 through 15 on the LVMF.
7.2.2.2.5 IF independent verification is approved, give the LVMF to the assigned personnel.
Proceed to step 7.2.3.2.2.
7.2.2.2.6 Return the key, IF applicable, to the Shift Support Supervisor.
7.2.2.2.7 Return the LVMF to the Shift Supervisor to assign independent verification.
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 12 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 7.2.3 INDEPENDENT VERIFICATION CONTINUOUS I 7.2.3.1 The Shift Supervisor shall:
7.2.3.1.1 Complete Block 16, indicating the applicability of step 5.2.1. Proceed to step 7.2.3.3, IF independent verification is waived.
7.2.3.1.2 Approve independent verification by completing Block 17 on the LVMF.
7.2.3.1.3 Have personnel assigned to perform independent verification.
7.2.3.1.4 Have the LVMF given to the assigned personnel.
7.2.3.2 The assigned personnel shall:
7.2.3.2.1 Check out the key, !E applicable, by showing the LVMF to the Shift Support Supervisor.
( 7.2.3.2.2 Perform independent verification including confirmation of valve position AND locking device operability per subsections 7.1.5 and 7.1.6. Contact the Shift Supervisor for guidance !EIF unexpected conditions are encountered.
7.2.3.2.3 Complete Block 18 and 19 of the LVMF.
7.2.3.2.4 Return the key, !E IF applicable, to the Shift Support Supervisor.
7.2.3.2.5 Return the LVMF to the Shift Supervisor.
7.2.3.3 The Shift Supervisor shall review the LVMF AND complete Block 20 of the LVMF.
7.2.3.4 The Shift Supervisor shall have the LVML updated by completion of Block 4.
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 13 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 7.2.4 LOCKED VALVE LOGBOOK CONTINUOUS I NOTE: There will be two Locked Valve Logbooks; one for Unit 1 AND one for Unit 2.
7.2.4.1 Each Unit's Shift Supervisor shall maintain control of the Locked Valve Logbook for that Unit.
7.2.4.2 This book shall include the Locked Valve Manipulation Log, all active LVMFs organized alpha-numerically by MPL number AND forms OPS-0443 or OPS-0444 and OPS-1698 or OPS-1699 as applicable to the affected Unit.
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 14 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 7.3 LOCKED VALVE SURVEILLANCE CONTINUOUS I 7.3.1 Surveillance required for locked valves AND lock wired valves included in this procedure are as indicated below.
7.3.1.1 Forms OPS-0071, OPS-0072, OPS-1046, OPS-1047, and OPS-1048 contain valves that are locked in normally accessible areas.
7.3.1.1.1 Forms OPS-0071 and OPS-0072, surveillance will be performed at the discretion of Operations Management.
7.3.1.1.2 Forms OPS-1046, OPS-1047, and OPS-1048 surveillance will be performed semi-annually.
7.3.1.2 Forms OPS-0073, OPS-0069 and OPS-1049 contain valves that are located in areas that are normally inaccessible due to health/safety hazards during Plant operation.
7.3.1.2.1 Forms OPS-0073 and OPS-0069 surveillance will be performed at the discretion of
( Operations Management; WHEN they are accessible.
7.3.1.2.2 Form OPS-1049, Inaccessible Lock Wired Valve Surveillance shall be performed semi-annually; WHEN it is accessible.
7.3.1.3 lE Form OPS-1049, IE Inaccessible Lock Wired Valve Surveillance, cannot be completed, the Shift Supervisor shall complete the first page of the form by checking "Unacceptable",
writing a brief explanation in the comments section, AND sign AND date. Transmit the first page of the form to Operations Support Supt. AND defer the Surveillance.
7.3.2 The Shift Supervisor shall have personnel assigned to perform surveillance.
7.3.3 The assigned personnel shall sign AND initial the first page of the required Form.
7.3.4 The assigned personnel shall perform surveillance for each valve by:
7.3.4.1 IE IF step 5.2.2 is applicable, obtain the Shift Supervisor's concurrence to waive surveillance AND write "Waived" in the space provided on the Form. Have the Shift Supervisor acknowledge his concurrence by entering his initials AND date beside the word "Waived",
upon return to the Control Room.
7.3.4.2 IE IF two positions are listed under "Required Position":
7.3.4.2.1 The "Required Position" shall be the as found position.
( 7.3.4.2.2 Indicate the "Required Position" by circling it.
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 15 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
CONTROL AND SURVEILLANCE OF LOCKED VALVES, 34GO-SUV-001-0 27.3 LOCK WIRED VALVES, AND LOCKED BREAKERS 7.3.4.3 Confirming the valve is in its "Required Position" per sUbsection 7.1.5.
7.3.4.4 IF performing Attachments 13, 14, 15, or 16, confirming each lock wired valve has a lock wired label affixed.
7.3.4.5 Confirming the locking device operability per subsection 7.1.6.
7.3.4.6 Document completion of surveillance for each valve by entering initials AND date in the space provided on the Form.
7.3.5 The Shift Supervisor shall review the completed Form AND document on the first page of the Form.
(
(
MGR-0001 Ver. 3
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN COMPANY A. Management Verification Policy Version 3.0 En~"KYI"&r"~
EntrJ!.YI'6Serpt Y6urW'srld" ll;.urWsrlir Procedure Page 1 of 11 Procedure Owner: Paul D. Rushton 1 Fleet Operations Manager 1 Corporate (Print: Name I/ Title I/ Site)
Approved By: Original Signed by Paul D. Rushton on 04/28/2008 04/2812008 (Procedure Owner's Signature I/ Date)
Effective Dates: N/A 6/1312008 6/13/2008 6/13/2008 Corporate FNP HNP VEGP This Standardization Process Control NMP is under the oversight of the Operations Peer Team.
Writer(s): Eric Snell
(
PROCEDURE USAGE REQUIREMENTS SECTIONS Procedure must be open and readily available at the Continuous Use: work location. Follow procedure step by step unless otherwise directed by the procedure.
Procedure or applicable section(s) available at the work Reference Use:
read:i reference by person performing location for ready performin9_ steps.
Information Use: Available on site for reference as needed. ALL
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN COMPANY A
A. Management Verification Policy Version 3.0 l:.Rtv E'n~VJ(}Se,.vt!
to&rw Y,,,.r Y"UI' Wsrltr Ws,-/.r Procedure Page 2 of 11 Procedure Version Description Version Number Version Descri tion 2.0
- Steps 4.2 and 6.1.4, clarified that independent verifier can be involved in the prejob briefing and this does not constitute being "influenced" by the performer.
- Step 6.0, added expectation that guiding document include verification requirements or that requirement is established at the prejob brief.
- Steps 6.1.4 through 6.1.9 provided more specific detail in expectations for independent verification (including timing of the verification, verifier being separately dispatched, etc). Steps 6.1.4, 6.1.6 and 6.1.7 are new and the remaining steps in this section were re-numbered accordingly.
- 4.3 - changed definition of qualified reviewer 3.0 Added NOTE prior to step 6.1.6 to clarify Independent Verification requirement for activities in the main control room.
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN COMPANY A Management Verification Policy Version 3.0 EnergyJ6$nw l~lIr WorJr f:.llergylll$nJIC :llilllr World" Procedure Page 3 of 11 Table of Contents 1.0 Purpose ............................................................................. ,........................................................ 4 2.0 Applicability ................................................................................................................................ 4 3.0 References ................................................................................................................................. 4 4.0 Definitions .................................................................................................................................. 4 5.0 Responsibilities .......................................................................................................................... 5 6.0 Procedure .................................................................................................................................. 5 6.1 IIndependent ndependent Verification ......................................................................................................... 6 6.2 Concurrent Verification ......................................................................................................... 10 6.3 Mispositioned Components Discovered During Verification .................................................. 11 7.0 Records .................................................................................................................................... 11 8.0 Commitments ........................................................................................................................... 11
(
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN COMPANY COMPANY A Management Verification Policy Version 3.0 l'Ouf Wsrld~
t()$UI'1! Yo""
l:.'lItrgyf()&rpl!
i:.'lIergy WfJrlJ~ Procedure Page 4 of 11 1.0 Purpose This procedure establishes policy and provides methods for verifying correct performance of normal operating, testing, and maintenance activities that affect the alignment or status of safety-related and some non-safety related systems or components.
2.0 Applicability This procedure applies to manipulation of power plant equipment where improper operation could create a challenge to plant safety or personnel safety or result in an unintended plant response such as a trip or ESFAS actuation.
3.0 References 3.1 NUREG 0737, Item I.C.6, "Guidance on Procedures for Verifying Correct Performance of Operating Activities" 3.2 USNRC IE Information Notice 84-51, "Independent Verification" 3.3 USNRC IE Information Notice 84-46, "Circuit Breaker Position Verification" 3.4 INPO 01-002 "Guidelines for the Conduct of Operations at Nuclear Power Stations"
( 4.0 Definitions 4.1 Concurrent Verification (CV) - Concurrent verification is the act of checking a condition, such as lifting a lead or installing a jumper, concurrent with the activities related to establishing the component's condition. Concurrent verification is used when an action or manipulation could result in an immediate threat to safe and reliable plant operation or a significant transient. Persons performing concurrent verifications identify the correct unit, train, or component and review the intended actions and expected responses before the task is performed, to prevent an unintended plant response.
4.2 Independent Verification (IV) - Independent verification is the act of checking the condition of a component independently from the individual responsible for establishing the component's condition. Independent verifications are truly independent in that the first and second checkers have no interaction during component manipulation. IV gives added assurance that a component is left in the required position and is used to verify the lineup of safety related equipment being returned to service.
4.3 Qualified Individual -
For CV - An individual possessing knowledge of the activity, systems, and/or components involved and the relationship of these activities, components, and systems to plant safety.
For IV - An individual who has basic knowledge of the type of component involved (valve, breaker, etc.). The individual need not be trained on the activity or system involved.
4.4 Significant Radiation Exposure - As applicable to activities described by this procedure,
( greater than or equal to 10 mrem whole body dose or airborne contamination in excess of ALARA guidelines.
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUIHERNA.
SOUTHERN COMPANY A Management Verification Policy Version 3.0 l:.'II~r£J F6 SeT"': Y6urW'"r/J' EllcrvI6&rv.: ~ur ""WtJrlJ' Procedure Page 5 of 11 5.0 Responsibilities 5.1 Department Managers Department managers shall ensure establishment of provisions within applicable procedures, which implement the policy, described herein.
5.2 Operations Manager The Operations Manager has overall responsibility for plant status control. As such the Operations Manager is responsible for proper implementation of this procedure at each respective SNC nuclear plant. He shall provide direction to other managers on implementation of this procedure, make any interpretations necessary and resolve issues that may arise.
Operations management reinforces site wide expectations that personnel conducting maintenance are responsible to ensure components are aligned properly after maintenance and to question off-normal components.
Establish, clearly communicate, and provide written guidance for routine component position verifications. Ensure that the guidance considers technical specification requirements, mode changes, and other transient conditions.
Ensure maintenance department personnel have a clear understanding of expectations for
( positioning components within the boundary of the tagtagout out and for the need to ensure systems are properly aligned before restoration. A clear process is in place to track components repositioned within a tagout boundary.
5.3 Supervisors, Team Leaders, and Assistant Team Leaders Supervisors have the following responsibilities:
A. Only qualified individuals are assigned to perform verifications.
B. Verifications are performed in accordance with the policy described in this procedure.
6.0 Procedure Guidelines, Instructions or Forms developed to support this procedure (NMP-OS-002) will be reviewed and approved by the Operations Peer Team Champion or designee.
SNC uses two forms of verification: Independent Verification, and Concurrent Verification.
Instructions for documenting independent and concurrent verification shall be provided in applicable procedures.
The practice of verifying throttled valves by shutting and reopening the valves a prescribed number of turns can create valve mispositionings. Instead, use position indicators, scribe marks, or other recognizable indicators that have been designated to determine throttled valve positions. When shutting and reopening a throttled valve is necessary to determine its position, perform concurrent verification rather than having both persons individually shut and reopen the valve.
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUIHERNA SOUTHERN COMPANY A Management Verification Policy Version 3.0 EnertYt6Serw EtltrrJ Ye",,.WiJrlJ" 16Serw Yellll' W()rlJ" Procedure Page 6 of 11 In some situations, functional testing may substitute for normal verification techniques in checking that components are correctly positioned. An example would be a full-flow test to prove the correct positioning of flow control valves. However, surveillance tests frequently will not serve to verify the positions of all components that are important to subsequent system operation. Therefore, use surveillance testing as component verification only if it can be shown that the test conclusively proves the position of the components. The Operations Manager must approve the use of surveillance testing applicability to satisfy component verification requirements.
The instructions for verification techniques describe the methods for verifying items such as manual valves, motor-operated and air-operated valves, solenoid-operated valves, circuit breakers, blank flanges, and removable links and fuses, as well as the status of control power.
During system lineups such as those performed coming out of a refueling outage, it is not necessary to have two people go to each component and check the position. A lineup is by definition a verification. Presumably, the components have already been positioned. At other times such as after system realignments, there should be a positioner and a verifier (Le., two people) to go to each component and make sure it is in the right position.
In most cases, the guiding document for an activity (procedure, tagout, work sequence, etc) should specify whether independent or concurrent verification is required. If not, the supervisor responsible for the activity will designate the type of verification at the prejob briefing.
6.1 Independent Verification
(
6.1.1 Independent Verification is required to be performed for restoration of:
A. Safety related systems or components.
B. Valve positions in liquid or gaseous radioactive waste systems that if mispositioned could lead to unintended or unmonitored radioactivity release.
- c. Other component positioning as determined necessary by the Operations Manager.
6.1.2 Exceptions - IV may be waived for the following reasons:
6.1.2.1 In cases that involve significant radiation exposure.
6.1.2.2 In cases that involve containment entry (PWR) or drywell entry (BWR).
(BWR), while containment integrity is established.
6.1.3 The Independent Verifier must be someone who has been independent of the task and has not been influenced by the positioner. The individual requesting the IV should provide instructions to the verifier regarding the procedural step(s) or tagout pOints to be verified.
6.1.4 Depending on the job scope and complexity, consideration should be given by supervision for the independent verifier to attend the pre-job brief.
(
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN SOUIHERNA COMPANY A Management Verification Policy Version 3.0 EnerD/~Suve ~llr'Wtl,ur i!neW/o$uwY(llttWtlTid' Procedure Page 7 of 11 6.1.5 Independent verification should be performed as soon as practical after the associated task is performed, but can generally wait until completion of the task unless an adverse consequence could result (plant transient, loss of safety function, etc).
NOTE: The following step applies ONLY to Operations personnel that are restricted to the main control room.
6.1.6 When independent verification is specified for activities in the main control room, independence will be maintained to the extent practical (Le. verifier will not directly observe the performance of the step).
6.1.7 For restoration of systems which require IV, careful consideration must be given to the sequence of placing the affected components in service and restoration of the system to operable status. If desired to place a system in service prior to completion of the IV, a peer check should be used to verify critical components are properly aligned (this is to prevent damage to equipment, spilling of water, etc). The system should not be considered operable until completion of the IV.
6.1.8 Independent verification involves the following process:
6.1.8.1 The person performing the component manipulation enters the area, separated from the verifier by time and/or distance.
( 6.1.8.2 The positioner then references the lineup, procedure, tagout, tag out, or caution tag and verifies the proper component, using human performance tools such as STAR.
6.1.8.3 The positioner shall place (or check) the component in the required position per the lineup, procedure, tag out, or caution tag, as applicable.
6.1.8.4 The positioner signs or initials in the prescribed place.
6.1.8.5 The verifier enters the area, separated from the positioner performing the manipulation by time and/or distance.
6.1.8.6 The verifier references the lineup, procedure, tagout, or caution tag and verifies the correct component has been identified, using human performance tools such as STAR.
6.1.8.7 The verifier observes the position of the component and physically checks component position.
6.1.8.8 The verifier signs or initials in the prescribed place.
(
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN SOU'HERNA COMPANY A Management Verification Policy Version 3.0 t4&r..~ :lli.llr Energy 1()$trH lliur World" WodJ" Procedure Page 8 of 11 PaQe 6.1.9 Independent Verification Methods 6.1.9.1 Direct Observation (preferred method) 6.1.9.1.1 Methods of performing direct observation for independent verification of valves or breakers include, but are not limited to, the following:
A. Visual observation of local breaker position indicating lights.
B. Visual observation of local breaker position indicating mechanical "flags."
C. Visual observation of breaker switch or handle position.
D. Manual valves to be independently verified open should be moved slightly in the closed direction and then moved in the open direction until the valve is considered in the fully open position, and, visual observation of the stem, i.e., grease markings indicating normal valve travel, valve stems extended on rising stem valves and mechanical position indication should also be included.
E. Valves required to be positioned slightly off "backseat" to prevent binding should be fully opened and returned to the procedurally established position during independent verification.
(
F. Manual valves to be independently verified closed should be moved, or attempted to be moved, only in the closed direction using normal closing torque and visually observing the stem. i.e., Grease markings indicating normal valve travel, valve stems inserted on rising stem valves, and mechanical position indication.
G. Visual observation and comparison with required stem position, local indicators, or other suitable valve component should be used to independently verify the position of throttled valves. Throttled valves shall not be moved to verify position unless specifically permitted to do so by the Shift Supervisor.
H. Control valve positions should be independently verified by ensuring that power or air, as appropriate, is available to the valve operators and that no physical obstructions which could prevent proper operation are apparent.
(
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN COMPANY A Management Verification Policy Version 3.0 Etlt'V1"Snv.:YourWtJ,1J' E,itrt:J1IJSawYQurWtJrlil' Procedure Page 9 of 11 6.1.9.2 Indirect Observation 6.1.9.2.1 Equipment failures can cause incorrect remote position indicating lights on the main control board. For remotely operated equipment, verification can usually be accomplished from the control room using instruments, annunciators or valve position indications. It is highly desirable to perform initial and independent verification using diverse indications. Initial and independent verification from the control room is permitted using non-diverse methods if alternate control room indicators or methods of local verification of position are not available. Valve stems without stem indicators are not considered a diverse indication. While valve stem markings may provide some information regarding valve position, it is not to be relied on as a verified valve position for purposes of this procedure.
6.1.9.2.2 Problems may occur with remote indicating reach-rod valves, in which the remote indicator does not exactly duplicate the actual valve position. For important reach-rod valves, consider using a local verification of position when possible.
6.1.9.2.3 In some situations, a component's position can be determined by observation of process parameters such as pressure, flow, or voltage.
This, combined with a physical check of a component's position, can constitute an independent verification. However, exercise caution when using process parameters, because alternate flow paths or other factors
( could cause them to be misleading indicators of component position.
6.1.9.2.4 Methods of performing indirect independent verification of breakers, setpoints, and valves include, but are not limited to, the following:
A. Visual observation of remote indicating lights for breaker operation.
B. Visual observation of the actuation of status or indicating lights at the required panel-meter; indicated value, of an established setpoint.
c.
C. Visual observation of flow indicators, as applicable to opening or closing valves, and/or remote valve position indicating lights for valve position.
NOTE: Functional tests used in lieu of independent verification should be examined to ensure they test the entire portion of the system affected by the previous actions.
D. Functional surveillance tests may be used for indirect, independent verification only if plant safety is not compromised and the indications are positive and immediate (Le., annunciator changes status following an action). The Operations Manager must approve the use of functional testing to satisfy component verification requirements.
(
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN COMPANY A Management Verification Policy Version 3.0 EnertYlo&rwY~NrWorltl' I::lltrtylo&rw Y~NrWorur Procedure Page 10 of 11 6.2 Concurrent Verification 6.2.1 Concurrent Verification should be performed for:
A. Removing equipment from service when an action or manipulation could result in an unintended or undesirable condition B. Placement of electrical grounds.
C. Manipulation of valves, breakers, switches, jumpers, lifted wires, blind flanges, plugs, or any other components that, if improperly installed or mispositioned, could degrade a safety function or cause an unnecessary unit trip.
D. Operation of equipment necessary to support operation of important systems, such as electrohydraulic control, instrument air, redundant generator stator cooling water pumps or any other component that, if improperly installed or mispositioned, could degrade a safety function or cause an unnecessary unit trip.
E. Manipulation of valve positions in liquid or gaseous radioactive waste systems that if mispositioned could lead to unintended or unmonitored radioactivity release.
F. Other component positioning as determined necessary by the Operations Manager.
( 6.2.2 Concurrent verification involves the following process:
6.2.2.1 Both individuals involved determine, prior to the verification, who will fulfill the role of the person locating and performing the component manipulations and who will be the verifier of the component. The individuals must rigorously adhere to these roles.
6.2.2.2 The person performing the component manipulation references the lineup, procedure, tagout, or caution tag, locates the component and verbally identifies each unique identifier on the component label to the verifier.
6.2.2.3 The positioner verbalizes the position in which he or she intends to place (or check) the component.
6.2.2.4 The verifier must independently read the lineup, procedure, tag out, or caution tag. The verifier must verify that the correct component is to be manipulated, and verbalize his agreement.
6.2.2.5 The positioner places (or checks) the component in the intended position.
6.2.2.6 The verifier witnesses the positioning (or check) of the component and physically verifies component position, when applicable.
6.2.2.7 Both persons sign or initial in the prescribed place.
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUIHERNA SOUTHERN COMPANY A Management Verification Policy Version 3.0 EH~rrJ,.$.trveKtUJ'w.rhr E'tm"xy .oSuve Your World" Procedure Page 11 of 11 6.3 Mispositioned Components Discovered During Verification NOTE: A component found out of position after all system alignments and verifications have been completed is considered a mispositioning.
A component found out of position during the verification phase is considered a near miss.
6.3.1 If, while performing verification, a component is found to be in a position other than required, the verifier will immediately notify the Shift Supervisor.
6.3.2 A component found out of desired position during the verification shall not be repositioned until the Shift Supervisor is notified and a verification of the mispositioned component is performed.
6.3.3 The Shift Supervisor will determine if the improper position of the component has caused any adverse system condition and if repositioning it to its correct alignment will result in an adverse condition.
6.3.4 If no adverse effects are noted or none could occur, the Shift Supervisor will direct that the component be properly positioned.
6.3.5 If any adverse condition exists or could occur, the affected system will first be placed
{ in a safe condition where the component can be set to its correct position.
6.3.6 A Condition Report shall be written to document any mispositioning or near miss.
7.0 Records None 8.0 Commitments None
DRAFT Southern Nuclear uclear E. I. Hatch Nuclear Plant Operations Training JPM ADMIN 22 SRO Onll:
(
TITLE DETERMINE MINIMUM CREW STAFFING AUTHOR MEDIA NUMBER TIME EN.FAGAN LR-JP-300.1-00 15 Minutes RECOMMENDED BY APPROVED BY DATE N/R
sM Energy to Serve Your World WOrld sM SOUTHERN NUCLEAR OPERATING COMPANY PLANT E. I. HATCH Page 1 of 1 FORM TITLE: TRAINING MATERIAL REVISION SHEET Program/Course Code: OPERATIONS TRAINING Media Number: LR-JP-300.1-00
. . . . . <. *. . . ***.** > . ......... ~ .. ..
Rev. No. Date Date . *. *;ReasonJorRel'ision Reason for Revision
- Autbol"s Author's Su}>v's Supv's
.. .......*............ ; . ........* ...... . . *.*.*. . . lIlidals Initials Illifials***.*****
Initials 00 Initial development FNF
(
LR-JP-300.1-00 Page 1 of 4
( UNIT 1 (X) UNIT 2 (X)
TASK TITLE: IDENTWYADEQUATEnNADEQUATESH~STABANG IDENTIFY ADEQUATEIINADEQUATE SHIFT STAFFING JPMNUMBER: LR-JP-300.1-00 TASK STANDARD: The task shall be complete when the operator has determined that crew staffing is short an HP, SO and appropriate management notified.
TASK NUMBER: H-OPSR300.001 OBJECTIVE NUMBER: H-OP300.001.A.01 H-OP300.001.A.Ol PLANT HATCH JTA IMPORTANCE RATING:
RO N/A SRO 3.03 KIA CATALOG NUMBER: 2.1.S 2.1.5
( KIA CATALOG JTA IMPORTANCE RATING:
RO N/A SRO 3.9 OPERATOR APPLICABILITY: Senior Reactor Operator (SRO)
IGENERAL
REFERENCES:
Unit 1 & 2 10 CFR 50.54(m)(2)(i) 30AC-OPS-003-0 "Plant Operations" DI-OPS-S 1-0501 , "Minimum Planned Crew Staffing" DI-OPS-81-0S01, Tech Specss IREQUIRED MATERIALS: Unit 1 & 2
-'Vr:~~-'J~ u*-vV.J-V. "Plant Operations" 30AC-OPS-003-0, 10 CPR CFR SO.S4(m)(2)(i) 50.54(m)(2)(i)
(note: lOCPR 10CFR requirements covered in 30AC-OPS-003-0)
DI-OPS-SI-0501, "Minimum Planned Crew Staffing" DI-OPS-81-0S01, Tech Specs APPROXIMATE COMPLETION TIME: IS Minutes 15 SIMULATOR SETUP: N/A
UNITl&2 READ TO THE OPERATOR INITIAL CONDITIONS:
- 1. It is 2 AM and both Units are at 100% power.
- 2. The HP Tech has sustained an injury and has been driven to the hospital by one of the on-shift SOs.
- 3. Current shift staffing includes the following:
( 1 - SSS 1- STA 4-NPO 5-S0 O-HP INITIATING CUES:
Determine if shift staffing is adequate and, if not, applicable actions/time limits per Tech Specs, Administrative Procedures or Departmental Instructions.
(
Page 3 of of44 PERFORMANCE STEP STANDARD SATIUNSAT COMMENTS For INITIAL Operator Programs:
For OJT/OJE; ALL PROCEDURE STEPS must be completed for Satisfactory Performance.
For License Examinations; ALL CRITICAL STEPS must be completed for Satisfactory Performance.
NOTE: The use/research of procedures may occur in any order. If the Operator does not verbalize his findings in each procedure, the evaluator may ask if ifhe he found any "applicable actions/time limits".
START TIME: ___ _____
NOTE: S.2.2.d Tech Specs Section 5.2.2.d
- 1. The operator obtains the correct Operator obtains a copy of Tech SAT/UNSAT SAT I UNSAT procedure. Specs.
- 3. Determines required action/time limit Operator determines that: SAT I UNSAT
- Immediate action required to replace HP
- Time limit of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> NOTE: 30AC-OPS-003-0, "Plant Operations" Attachment 1
- 4. The operator obtains the correct The Operator obtains a controlled SAT/UNSAT SAT I UNSAT procedure. copy of 30AC-OPS-003-0, "Plant Operations"
- 5. Determine required staffing. Operator determines staffing SAT/UNSAT requirements are met.
NOTE: DI-OPS-81-0S01, DI-OPS-81-0501, "Minimum Planned Crew Staffing" Section 4.0
- 6. The operator obtains the correct Operator obtains a controlled SAT/UNSAT SAT I UNSAT procedure. copy of DI-OPS-81-0S0 1, ofDI-OPS-81-0501, "Minimum Planned Crew Staffing".
(** Indicates critical step)
Page 4 of4 PERFORMANCE STEP STANDARD SATfJ:J~$A.,)] .......
SAT/UNSAT CQMME:1S'fS*;
COMMENTS
- 8. Detennines required actions Operator detennines the SAT I UNSAT SAT/UNSAT following actions are required:
- Write a condition report.
- Notify the Hatch Duty NotifY Manager
- Notify the Duty Ops NotifY Supervisor.
END TIME:
TERMINATING CUE: We will stop here.
(
{
\
(*
- Indicates critical step)
UNITl&2 READ TO THE OPERATOR INITIAL CONDITIONS:
- 1. It is 2 AM and both Units are at 100% power.
- 2. The HP Tech has sustained an injury and has been driven to the hospital by one of the on-shift SOs.
- 3. Current shift staffing includes the following:
( 1 - SSS 1- STA 4-NPO 5-S0 5-SO O-HP INITIATING CUES:
Determine if shift staffing is adequate and, if not, applicable actions/time limits per Tech Specs, Administrative Procedures or Departmental Instructions.
Edwin 1.
I Hatch Nuclear Plant Technical Specifications Unit 1
(
\
(
TABLE OF CONTENTS 1.0 USE AND APPLICATION APPLiCATION ........................................................................... 1.1-1 1.1 Definitions ...................................................................................................
Definitions...... ................ ........... ... ........ ...... ..... ....... ......... ..... ... ..... ... .... .... .... 1.1-1 1.2 Logical Connectors ...................................... ...............................................
..................................................................................... 1.2-1 1.3 Completion TimesTimes.... .......................................................................................
... ... ..... ...... ... ..... ...... ..... .... ..... ........... ..... ... ............ ... .... 1.3-1 1.4 Frequency ................................................................................................... 1.4-1 2.0 SAFETY LIMITS (SLs) ............................................................................... 2.0-1 2.1 SLs ............................................................................................................. 2.0-1 2.2 SL Violations ............................................................................................... 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY .......... 3.0-1 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ......................... 3.0-3 3.1 REACTIVITY CONTROL SYSTEMS .......................................................... 3.1-1 3.1.1 SHUTDOWN MARGIN (SDM) .................................................................... 3.1-1
( 3.1.2 Anomalies ...................................................................................
Reactivity Anomalies................. ... ........... ....... ......... .............. ........ .... .... ... ... 3.1-4
\ 3.1.3 Control Rod OPERABILITY ........................................................................ 3.1-5 3.1.4 Control Rod Scram Times............
Times. ........... ...............................................................
........ ... ....... ... ..... ............... .... ......... ... ...... 3.1-9 3.1.5 Control Rod Scram Accumulators ............................................................... 3.1-12 3.1.6 Rod Pattern Control....................................................................................
Control............... ................ ..................................................... 3.1-15 3.1.7 Standby Liquid Control (SLC) System ......................................................... 3.1-17 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves ............................. 3.1-22 3.2 POWER DISTRIBUTION LIMITS ............................................................... 3.2-1 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ....... 3.2-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ........................................... 3.2-2 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ............................................ 3.2-4 3.3 INSTRUMENTATION ................................................................................. 3.3-1 3.3.1.1 Reactor Protection System (RPS) Instrumentation ..................................... 3.3-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation ............................................ 3.3-10 3.3.2.1 Control Rod Block Instrumentation .............................................................. 3.3-15 3.3.2.2 Feedwater and Main Turbine Trip High Water Level ................................... 3.3-20 Instrumentation 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation ........................................ 3.3-22 3.3.3.2 Remote Shutdown System .......................................................................... 3.3-25 (continued)
HATCH UNIT 1 Amendment No. 239
TABLE OF CONTENTS (continued) 3.3 INSTRUMENTATION (continued) 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation ............ 3.3-27 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation .............................................. 3.3-30 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation ......................... 3.3-33 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation ................. 3.3-43 3.3.6.1 Primary Containment Isolation Instrumentation ........................................... 3.3-47 3.3.6.2 Secondary Containment Isolation Instrumentation ...................................... 3.3-55 3.3.6.3 Low-Low Set (LLS) Instrumentation ............................................................ 3.3-58 3.3.7.1 Main Control Room Environmental Control (MCREC) System Instrumentation ..................................................................................... 3.3-62 3.3.8.1 Instrumentation .........................................................
Loss of Power (LOP) Instrumentation............ ................ ..... ................ ........ 3.3-64 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ...................... 3.3-67 3.4 REACTOR COOLANT SYSTEM (RCS) ..................................................... 3.4-1 3.4.1 Recirculation Loops Operating .................................................................... 3.4-1 3.4.2 Jet Pumps ................................................................................................... 3.4-4 3.4.3 Safety/Relief Valves (S/RVs) ......................................................................
Safety/ReliefValves 3.4-6 3.4.4 RCS Operational LEAKAGE............
LEAKAGE .......................................................................
...... .... ........... ..... ..... ... .......... ... ............ 3.4-8 3.4.5 RCS Leakage Detection Instrumentation .................................................... 3.4-10 3.4.6 RCS Specific Activity ..................................................................................
Activity........................ ..... ... ....... .... .................. ..... ...... ..... ..... 3.4-12 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System -
Hot Shutdown ....................................................................................... 3.4-14 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System -
Cold Shutdown ...................................................................................... 3.4-17 3.4.9 RCS Pressure and Temperature (PIT) (P/T) Limits .............................................. 3.4-19 3.4.10 Reactor Steam Dome Pressure..................................................................
Pressure .................................................................. 3.4-25 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ........................................ 3.5-1 3.5.1 ECCS - Operating ....................................................................................... 3.5-1 3.5.2 ECCS - Shutdown ....................................................................................... 3.5-6 3.5.3 RCIC System ..............................................................................................
............ .......... ......... ................................. .......... ....... ........ ..... 3.5-9 3.6 CONTAINMENT SYSTEMS ....................................................................... 3.6-1 3.6.1.1 Primary Containment .................................................................................. 3.6-1 3.6.1.2 Primary Containment Air Lock.. Lock ....................................................................
.......... ........ ................. ......... ........ .............. 3.6-3 3.6.1.3 Primary Containment Isolation Valves (PCIVs) ........................................... 3.6-7 3.6.1.4 Drywe/l Pressure .........................................................................................
Drywell 3.6-13 3.6.1.5 Drywel/
Drywell Air Temperature ............................................................................. 3.6-14 (continued)
HATCH UNIT 1 ii Amendment No. 195
(
TABLE OF CONTENTS (continued) 3.6 CONTAINMENT SYSTEMS (continued) 3.6.1.6 Low-Low Set (LLS) Valves ............................................................................ 3.6-15 3.6.1.7 Reactor Building-to-Suppression Chamber Vacuum Breakers ..................... 3.6-17 3.6.1.8 Suppression Chamber-to-Drywell Vacuum Breakers .................................... 3.6-19 3.6.2.1 Suppression Pool Average Temperature ...................................................... 3.6-21 3.6.2.2 Suppression Pool Water Level ...................................................................... 3.6-24 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling ........................... 3.6-25 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray .............................. 3.6-27 3.6.3.1 Containment Atmosphere Dilution (CAD) System ......................................... 3.6-29 3.6.3.2 Primary Containment Oxygen Concentration ................................................ 3.6-31 3.6.4.1 Secondary Containment ................................................................................ 3.6-32 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) ........................................ 3.6-35 3.6.4.3 Standby Gas Treatment (SGT) System ........................................................ 3.6-38 3.7 PLANT SYSTEMS ........................................................................................ 3.7-1 3.7.1 Residual Heat Removal Service Water (RHRSW) System ........................... 3.7-1 3.7.2 Plant Service Water (PSW) System and Ultimate Heat Sink (UHS) ............. 3.7-3 3.7.3 Diesel Generator (DG) 1 1BB Standby Service Water (SSW) System ............... 3.7-6 3.7.4 Main Control Room Environmental Control (MCREC) System ..................... 3.7-8
( 3.7.5 Control Room Air Conditioning (AC) System ................................................ 3.7-12 3.7.6 Main Condenser Offgas ................................................................................ 3.7-16 3.7.7 Main Turbine Bypass System ........................................................................ 3.7-18 3.7.8 Spent Fuel Storage Pool Water Level Level............................................................
.......................................................... 3.7-19 3.8 ELECTRICAL POWER SySTEMS ............................................................... 3.8-1 3.8.1 AC Sources - Operating ................................................................................ 3.8-1 3.8.2 AC Sources - Shutdown ................................................................................ 3.8-20 3.8.3 Diesel Fuel Oil and Transfer, Lube Oil, and Starting Air ............................... 3.8-23 3.8.4 DC Sources - Operating ................................................................................ 3.8-26 3.8.5 DC Sources - Shutdown ................................................................................ 3.8-31 3.8.6 Battery Cell Parameters ................................................................................ 3.8-33 3.8.7 Distribution Systems - Operating ................................................................... 3.8-36 3.8.8 Distribution Systems - Shutdown .................................................................. 3.8-39 3.9 REFUELING OPERATIONS ......................................................................... 3.9-1 3.9.1 Refueling Equipment Interlocks ..................................................................... 3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock ......................................................... 3.9-3 3.9.3 Control Rod Position ..................................................................................... 3.9-4 3.9.4 Control Rod Position Indication ..................................................................... 3.9-5 3.9.5 Control Rod OPERABILITY - Refueling ........................................................ 3.9-7 I
\
(continued)
HATCH UNIT 1 iii Amendment No. 253
TABLE OF CONTENTS (continued) 3.9 REFUELING OPERATIONS (continued) 3.9.6 Reactor Pressure Vessel (RPV) Water LeveL Level. ............................................... 3.9-8 3.9.7 Residual Heat Removal (RHR) - High Water LeveL Level. ...................................... 3.9-9 3.9.8 Residual Heat Removal (RHR) - Low Water Level ....................................... 3.9-11 3.10 SPECIAL OPERATIONS .............................................................................. 3.10-1 3.10.1 Inservice Leak and Hydrostatic Testing Operation ........................................ 3.10-1 3.10.2 Reactor Mode Switch Interlock Testing ......................................................... 3.10-3 3.10.3 Single Control Rod Withdrawal - Hot Shutdown ............................................ 3.10-5 3.10.4 Single Control Rod Withdrawal - Cold Shutdown .......................................... 3.10-8 3.10.5 Single Control Rod Drive (CRD) Removal- Removal - Refueling .................................. 3.10-11 3.10.6 Multiple Control Rod Withdrawal - Refueling ................................................. 3.10-13 3.10.7 Control Rod Testing - Operating ................................................................... 3.10-15 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling ............................................ 3.10-17 4.0 DESIGN FEATURES .................................................................................... 4.0-1 4.1 Site ................................................................................................................ 4.0-1
( 4.2 Reactor Core ................................................................................................. 4.0-1 4.3 Fuel Storage .................................................................................................. 4.0-2 5.0 ADMINISTRATIVE CONTROLS ................................................................... 5.0-1 5.1 Responsibility ................................................................................................ 5.0-1 5.2 Organization .................................................................................................. 5.0-2 5.3 Unit Staff Qualifications ................................................................................. 5.0-5 5.4 Procedures .................................................................................................... 5.0-6 5.5 Programs and Manuals ................................................................................. 5.0-7 5.6 Reporting Requirements ............................................................................... 5.0-18 5.7 High Radiation Area ...................................................................................... 5.0-21 (continued)
HATCH UNIT 1 iv Amendment No. 253
TABLE OF CONTENTS (continued)
LIST OF TABLES 1.1-1 MODES ....................................................................................................... 1.1-6 3.1.4-1 Control Rod Scram Times ........................................................................... 3.1-11 3.3.1.1-1 Reactor Protection System Instrumentation ................................................ 3.3-7 3.3.1.2-1 Source Range Monitor Instrumentation ....................................................... 3.3-14 3.3.2.1-1 Control Rod Block Instrumentation .............................................................. 3.3-19 3.3.3.1-1 Post Accident Monitoring Instrumentation ................................................... 3.3-24 3.3.5.1-1 Emergency Core Cooling System Instrumentation ...................................... 3.3-38 3.3.5.2-1 Reactor Core Isolation Cooling System Instrumentation ............................. 3.3-46 3.3.6.1-1 Primary Containment Isolation Instrumentation ........................................... 3.3-51 3.3.6.2-1 Secondary Containment Isolation Instrumentation ...................................... 3.3-57 3.3.6.3-1 Low-Low Set Instrumentation ...................................................................... 3.3-61 3.3.8.1-1 Loss of Power Instrumentation .................................................................... 3.3-66 3.8.6-1 Battery Cell Parameter Requirements ......................................................... 3.8-31 LIST OF FIGURES 3.1.7-1 Sodium Pentaborate Solution Volume Versus Concentration Requirements ........................................................................................ 3.1-20 3.1.7-2 Sodium Pentaborate Solution Temperature Versus Concentration
(
\ Requirements ........................................................................................ 3.1-21 3.4.1-1 Deleted.......................................................................................................
Deleted ....................................................................................................... 3.4-3 3.4.9-1 PressureiTemperature Limits for Inservice Hydrostatic and Inservice Pressure/Temperature Leakage Tests ....................................................................................... 3.4-22 3.4.9-2 PressureiTemperature Limits for Non-Nuclear Heatup, Low Power Pressure/Temperature Physics Tests, and Cooldown Following a Shutdown ............................ 3.4-23 3.4.9-3 PressureiTemperature Limits for Criticality ..................................................
Pressure/Temperature 3.4-24 4.1-1 Site and Exclusion Area Boundaries and Low Population Zone .................. 4.0-3 HATCH UNIT 1 v Amendment No. 195
Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.
- a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the Plant Hatch Unit 1 FSAR;
- b. The Plant Manager shall be responsible for overall safe operation of the
( plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant;
- c. The Vice President - Hatch shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and
- d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.
5.2.2 Unit Staff The unit staff organization shall include the following:
- a. A total of three plant equipment operators (PEOs) for the two units is required in all conditions. At least one of the required PEOs shall be assigned to each reactor containing fuel.
(continued)
HATCH UNIT 1 5.0-2 Amendment No. 252
Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued)
- b. At least one licensed Reactor Operator (RO) shall be present in the control room for each unit that contains fuel in the reactor. In addition, while the unit is in MODE 1,2, 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.
- c. The minimum shift crew composition shall be in accordance with 10 CFR 50.54(m}(2}(i}.
50.54(m)(2)(i). Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m}(2}(i}
50.54(m)(2)(i) and 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
- d. An individual qualified to implement radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
- e. Administrative procedures shall be developed and implemented to limit
( the working hours of unit staff who perform safety related functions (e.g.,
licensed and non-licensed operations personnel, health physics technicians, key maintenance personnel, etc.).
Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a nominal 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:
- 1. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time;
- 2. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 day period, all excluding shift turnover time;
- 3. A break of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> should be allowed between work periods, including shift turnover time;
- 4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
(continued)
HATCH UNIT 1 5.0-3 Amendment No. 195
Organization 5.2 5.2 Organization 5.2.2 Unit Staff
- e. (continued)
Any deviation from the above guidelines shall be authorized by the Plant Manager or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.
Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.
- f. The Operations Manager or at least one Operations Superintendent shall hold an SRO license.
- g. The Shift Technical Advisor (STA) shall provide advisory technical support to the shift supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the
( unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.
HATCH UNIT 1 5.0-4 Amendment No. 260
§ 50.54 Conditions oflicenses.
of licenses.
§ 50.54 Conditions of licenses.
The following paragraphs with the exception of paragraphs (r) and (gg) of this section are conditions in every nuclear power reactor operating license issued under this part. The following paragraphs with the exception of paragraph (r), (s), and (u) of this section are conditions in every combined license issued under part 52 of this chapter, provided, however, that paragraphs (i), (i-I), 0), (k), (1), (m), (n), (w), (x), (y), and (z) of this section are only applicable after the Commission makes the finding under § 52.1 03 (g) of this chapter.
(a)(1) Each nuclear power plant or fuel reprocessing plant licensee subject to the quality assurance criteria in appendix B of this part shall implement, under § 50.34(b)(6)(ii) or § 52.79 of this chapter, the quality assurance program described or referenced in the safety analysis report, including changes to that report. However, a holder of a combined license under part 52 of this chapter shall implement the quality assurance program described or referenced in the safety analysis report applicable to operation 30 days prior to the scheduled date for the initial loading of fuel.
(2) Each licensee described in paragraph (a)(l)
(a)(1) of this section shall, by June 10, 1983, submit to the appropriate NRC Regional Office shown in appendix D of part 20 of this chapter the current description of the quality assurance program it is implementing for inclusion in the Safety Analysis Report, unless there are no changes to the description previously accepted by NRC. This submittal must identify changes made to the quality assurance program description since the description was submitted to NRC. (Should a licensee need additional time beyond June 10, 1983 to submit its current quality assurance program description to NRC, it shall notify the appropriate NRC Regional Office in writing, explain why additional time is needed, and provide a schedule for NRC approval showing when its current quality assurance program description will be submitted.)
(a)(1) of this section may make a change to a (3) Each licensee described in paragraph (a)(l) previously accepted quality assurance program description included or referenced in the Safety Analysis Report without prior NRC approval, provided the change does not reduce the commitments in the program description as accepted by the NRC. Changes to the quality assurance program description that do not reduce the commitments must be submitted to the NRC in accordance with the requirements of Sec. 50. 71 (e). In addition 50.71(e).
to quality assurance program changes involving administrative improvements and clarifications, spelling corrections, punctuation, or editorial items, the following changes are not considered to be reductions in commitment:
(i) The use of a QA standard approved by the NRC which is more recent than the QA standard in the licensee's current QA program at the time of the change;
(ii) The use of a quality assurance alternative or exception approved by an NRC safety evaluation, provided that the bases of the NRC approval are applicable to the licensee's facility; (iii) The use of generic organizational position titles that clearly denote the position function, supplemented as necessary by descriptive text, rather than specific titles; (iv) The use of generic organizational charts to indicate functional relationships, authorities, and responsibilities, or, alternately, the use of descriptive text; (v) The elimination of quality assurance program information that duplicates language in quality assurance regulatory guides and quality assurance standards to which the licensee is committed; and (vi) Organizational revisions that ensure that persons and organizations performing quality assurance functions continue to have the requisite authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations.
(4) Changes to the quality assurance program description that do reduce the commitments must be submitted to the NRC and receive NRC approval prior to implementation, as follows:
( (i) Changes made to the quality assurance program description as presented in the Safety Analysis Report or in a topical report must be submitted as specified in Sec. 50.4.
(ii) The submittal of a change to the Safety Analysis Report quality assurance program description must include all pages affected by that change and must be accompanied by a forwarding letter identifying the change, the reason for the change, and the basis for concluding that the revised program incorporating the change continues to satisfy the criteria of appendix B of this part and the Safety Analysis Report quality assurance program description commitments previously accepted by the NRC (the letter need not provide the basis for changes that correct spelling, punctuation, or editorial items).
(iii) A copy of the forwarding letter identifying the change must be maintained as a facility record for three years.
(iv) Changes to the quality assurance program description included or referenced in the Safety Analysis Report shall be regarded as accepted by the Commission upon receipt of a letter to this effect from the appropriate reviewing office of the Commission or 60 days after submittal to the Commission, whichever occurs first.
(b) No right to the special nuclear material shall be conferred by the license except as may be defined by the license.
(c) Neither the license, nor any right thereunder, nor any right to utilize or produce special nuclear material shall be transferred, assigned, or disposed of in any manner, either voluntarily or involuntarily, directly or indirectly, through transfer of control of the license to any person, unless the Commission shall, after securing full information, find that the transfer is in accordance with the provisions of the act and give its consent in writing.
(d) The license shall be subject to suspension and to the rights of recapture of the material or control of the facility reserved to the Commission under section 108 of the act in a state of war or national emergency declared by Congress.
(e) The license shall be subject to revocation, suspension, modification, or amendment for cause as provided in the act and regulations, in accordance with the procedures provided by the act and regulations.
(f) The licensee shall at any time before expiration of the license, upon request ofthe ofthe of the Commission, submit, as specified in § 50.4, written statements, signed under oath or affirmation, to enable the Commission to determine whether or not the license should be modified, suspended, or revoked. Except for information sought to verify licensee compliance with the current licensing basis for that facility, the NRC must prepare the reason or reasons for each information request prior to issuance to ensure that the burden to be imposed on respondents is justified in view of the potential safety significance of the issue to be addressed in the requested information. Each such justification provided
( for an evaluation performed by the NRC staff must be approved by the Executive Director for Operations or his or her designee prior to issuance of the request.
(g) The issuance or existence of the license shall not be deemed to waive, or relieve the licensee from compliance with, the antitrust laws, as specified in subsection 1105a05a of the Act. In the event that the licensee should be found by a court of competent jurisdiction to have violated any provision of such antitrust laws in the conduct of the licensed activity, the Commission may suspend or revoke the license or take such other action with respect to it as shall be deemed necessary.
(h) The license shall be subject to the provisions of the Act now or hereafter in effect and to all rules, regulations, and orders of the Commission. The terms and conditions of the license shall be subject to amendment, revision, or modification, by reason of amendments of the Act or by reason of rules, regulations, and orders issued in accordance with the terms of the act.
(i) Except as provided in § 55.13 of this chapter, the licensee may not permit the manipulation of the controls of any facility by anyone who is not a licensed operator or senior operator as provided in part 55 of this chapter.
(i-I) Within 3 months after either the issuance of an operating license or the date that the (i-1)
Commission makes the finding under § 52.103(g) of this chapter for a combined license, as applicable, the
\,
licensee shall have in effect an operator requalification program. The operator requalification program must, as a minimum, meet the requirements of
§ 55.59(c) of this chapter. Notwithstanding the provisions of § 50.59, the licensee may not, except as specifically authorized by the Commission decrease the scope of an approved operator requalification program.
G) Apparatus and mechanisms other than controls, the operation of which may affect the reactivity or power level of a reactor shall be manipulated only with the knowledge and consent of an operator or senior operator licensed pursuant to part 55 of this chapter present at the controls.
(k) An operator or senior operator licensed pursuant to part 55 of this chapter shall be present at the controls at all times during the operation of the facility.
(1) The licensee shall designate individuals to be responsible for directing the licensed activities of licensed operators. These individuals shall be licensed as senior operators oflicensed pursuant to part 55 of this chapter.
(m)(l) A senior operator licensed pursuant to part 55 of this chapter shall be present at (m)(1) the facility or readily available on call at all times during its operation, and shall be present at the facility during initial start-up and approach to power, recovery from an unplanned or unscheduled shut-down or significant reduction in power, and refueling, or
( as otherwise prescribed in the facility license.
(2) Notwithstanding any other provisions of this section, by January 1, 1984, licensees of nuclear power units shall meet the following requirements:
(i) Each licensee shall meet the minimum licensed operator staffing requirements in the following table:
Minimum Requirements! Per Shift for On-Site Staffing of Nuclear Power Units by Operators and Senior Operators Licensed Under 10 CFR Part 55 N urn b erooff Number !I lOne Unit I I Two units II -
Three units I
JI, nuclear power I!!I Position One I One I Two II Two I Three II operating2-units operatini I II I control room I
I control room II control II rooms I I
control I control contro rooms !I rooms I
I, J
--.J ISenior j
I F
None 11 11 11 1111 IOperator I 11 r-I I
I1 I 1, -----lI I ______~I-O_p-er_a-to-r_+-------l~1 fOperator 11 ______21J 2~1____--_2+1-------3~1------~
21 3j 2J F-
\
lOne i
ISenior I
Operator I,Operator 21 21 21 II 21 21 I1 21 I I
I 21I, II I
j
, I!Operator Operator 2121 31I 3 1 i
31 41 ~ 4J
hwo
_ior~~I_ _~21 ~31 1~3' ~
Two Senior 2 3 ~3 3 l~erator ILs
. . e_.n...
Operator I _ _ _ _ I _ _
IOperator I 331 441 lsi
.25 5s '
IThree Three IOperator Senior Senior
!Operator
,I. II I
.--+------t-----;-----
331:
1
--;14 1
I O-pe-ra-to-r-+'----+--I-Operator i-I i I Sl 5
--.l 661 1.
ITemporary deviations from the numbers required Temporary devIatIOns this table shall be In reqUIred by thIS in accordance with criteria established in the unit's technical specifications.
2For the purpose of this table, a nuclear power unit is considered to be operating when it 2For is in a mode other than cold shutdown or refueling as defined by the unit's technical specifications.
3The number of required licensed personnel when the operating nuclear power units are controlled from a common control room are two senior operators and four operators.
(ii) Each licensee shall have at its site a person holding a senior operator license for all fueled units at the site who is assigned responsibility for overall plant operation at all times there is fuel in any unit. If a single senior operator does not hold a senior operator license on all fueled units at the site, then the licensee must have at the site two or more senior operators, who in combination are licensed as senior operators on all fueled units.
(iii) When a nuclear power unit is in an operational mode other than cold shutdown or
( refueling, as defined by the unit's technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. In addition to this senior operator, for each fueled nuclear power unit, a licensed operator or senior operator shall be present at the controls at all times.
(iv) Each licensee shall have present, during alteration of the core of a nuclear power unit (including fuel loading or transfer), a person holding a senior operator license or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person.
(3) Licensees who cannot meet the January 1, 1984 deadline must submit by October 1, 1983 a request for an extension to the Director of the Office of Nuclear Regulation and demonstrate good cause for the request.
(n) The licensee shall not, except as authorized pursuant to a construction permit, make any alteration in the facility constituting a change from the technical specifications previously incorporated in a license or construction permit pursuant to § 50.36 SO.36 of this part.
(0) Primary reactor containments for water cooled power reactors, other than facilities for which the certifications required under §§ SO.82(a)(1) S2.110(a)(1) of this chapter have been submitted, shall 50.82(a)(1) or 52.11O(a)(1) be subject to the requirements set forth in appendix J to this part.
SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH ADMINISTRATIVE CONTROL PROCEDURE 1 OF 8 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
PLANT OPERATIONS 30AC-OPS-003-0 26.4 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR G. R. Brinson DATE 8-08-08 DATE:
N/A DATE 6-19-09 SSM I/ PM W. L. Bargeron 8-12-08 1.0 OBJECTIVE This procedure establishes the requirements and responsibilities to safely operate Plant Hatch.
Included are administrative controls established for operations personnel and types of procedures necessary to control plant operations. This procedure implements the requirements of Unit 1 TS and Unit 2 TS Section 5.2.2.a, 5.2.2.b, 5.2.2.c, and 5.2.2.d.
TABLE OF CONTENTS Section 2.0 APPLICABILITY .................................................................................................................. 1
3.0 REFERENCES
................................................................................................................... 2 4.0 RESPONSIBILITIES ........................................................................................................... 2 5.0 REQUIREMENTS ............................................................................................................... 3
( 6.0 PRECAUTIONS/LIMITATIONS ........................................................................................... 3 7.0 PREREQUISITES ............................................................................................................... 3 8.0 PROCEDURE ..................................................................................................................... 4 8.1 CONDUCT OF OPERATIONS ...................................................................................... 4 8.2 CONDUCT OF PERSONNEL IN THE MAIN CONTROL ROOM (MCR) AND RADWASTE CONTROL ROOMS (RWCR) ................................................................... 4 8.3 REQUIRED MANNING ................................................................................................. 4 8.4 PERSONNEL DUTIES .................................................................................................. 5 8.4.1 Administrative Assistants ....................................................................................... 5 8.5 MEDICAL STATUS CHANGE ....................................................................................... 6 8.6 NO SOLO OPERATIONS LICENSE RESTRICTIONS .................................................. 6 Attachments 1 MINIMUM SHIFT CREW COMPOSiTION ........................................................................... 7 2 WATCH-STANDING SHIFT PROFICIENCY PROFiCiENCy ....................................................................... 8 2.0 APPLICABILITY The administrative controls established by this procedure apply to those personnel involved in the operation and maintenance of Plant Hatch.
(
MGR-0002 Rev 8.1
SNC PLANT E. I. HATCH I Pg 7 of 8 DOCUMENT TITLE:
PLANT OPERATIONS IIDOCUMENT NUMBER:
30AC-OPS-003-0 Ver No:
26.4 ATTACHMENT _1 Att. Pg.
TITLE: MINIMUM SHIFT CREW COMPOSITION 1 of 1 Minimum requirements per shift for On Site Staffing per 10 CFR 50.54(m)(2)(i). Additionally Included is staffing for non licensed operators (TS 5.2.2.a) and STA (TS 5.2.2.g) positions.
Number of nuclear power units Position Minimum number required operating Senior Reactor Operator 1
Nuclear Plant Operator None 2 System Operator 3
Shift Technical Advisor 0
Senior Reactor Operator One 2 Nuclear Plant Operator or 3 System Operator Two 3 Shift Technical Advisor 1
- Temporary deviations from the numbers required by this table shall be in accordance with criteria established in the TS
(
- A nuclear power unit is considered to be operating when it is in a mode other than cold shutdown or refueling as defined by the TS
- This does not include the Licensed Senior Reactor Operator limited to fuel handling, supervising core alterations.
MGR-0009 Rev 5.0
SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH DEPARTMENT DIRECTIVE 1 OF 1 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
MINIMUM PLANNED CREW MANNING DI-OPS-81-0501 3.0 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR G. R. Brinson 01/09/09 DATE:
N/A N/A N/A DATE N/A 1-9-09 SSM / PM SSM/PM 1.0 OBJECTIVE The objective of this department instruction is to provide guidance on minimum staffing levels to ensure POD activities and commitments can be performed as required.
2.0 APPLICABILITY This instruction is applicable to routine Plant operations.
3.0 REFERENCES
3.1 DI-OPS-59-0896, Operations Management Expectations 3.2 AG-MGR-26-0487, Duties of the Plant Hatch Duty Manager 3.3 30AC-OPS-003-0, Plant Operations
/
\ 4.0 PROCEDURE 4.1 Minimum planned crew manning is as follows:
4 Supervisors, 4 NPO's, 6 SO's (4 of which cover Fire Brigade).
4.2 If 5 supervisors are on shift they normally fill the following positions:
SM, U-1 SS, U-2 SS, STA, SSS. The SSS fills the Fire Brigade Leader position. Should only 4 supervisors be on shift, one of the Unit SS's will be the Fire Brigade Leader.
4.3 NPO's can be used to substitute for SO's so that the total of 10 covered personnel is met.
4.4 This staffing level is above Tech Spec minimums but is needed to meet commitments and workload. If this staffing level cannot be met at the beginning of the shift or any time during the shift the following actions will be taken:
- Write a condition report.
- Notify the Hatch Duty Manager per AG-MGR-26-0487, Duties of the Plant Hatch Duty Manager.
Explain where you are short and the impact to the team, if any.
- If you are below Tech Spec minimum or cannot meet safe shutdown or fire brigade minimums then make every effort to call in support. Otherwise, call in support if needed.
- Notify the Duty Ops Supervisor. Get him to assist you calling in support if needed.
{
MGR-0002 Rev 8.1
DRAFT Southern Nuclear E. I. Hatch Nuclear Plant Operations Training JPM Admin 3, RO Only TITLE DETERMINE THE DRYWELL FLOOR DRAIN LEAKAGE RATE
(
AUTHOR MEDIA NUMBER TIME JOHN PENDLEBURY LR-JP-40.05-04 13.0 Minutes RECOMMENDED BY APPROVED BY DATE NIR N/R
SOUTHERN NUCLEAR OPERATING COMPANY
( PLANT E. I. HATCH of 1 Page 1 of1 FORM TITLE: TRAINING MATERIAL REVISION SHEET Program/Course Code: OPERATIONS TRAINING Media Number: LR-JP-40.05 Rev. Author's Supv's Date Reason for Revision No. Initials Initials 00 12118/2002 Initial Development DNM DHG 01 02/04/2003 Correct VI Calculation DNM DHG 02 0611712005 Revised Initial License statement for successful completion RAB RAB 03 05105/2006 05/05/2006 Remove Response Cues RAB RAB 04 09/30108 Modified attachments 1,3,4, and 6 to reflect the tables JWP RAB referenced more closely.
(
(
Page 1 of7 LR-JP-40.05-04
(
UNIT 1 (X) UNIT 2 (X)
TASK TITLE: DETERMINE THE DRYWELL FLOOR DRAIN LEAKAGE RATE JPMNUMBER: LR-JP-40.05-03 TASK STANDARD: The task will be complete when the Drywell Leakage Rate has been detennined determined and any procedural limit violations reported to the Shift Supervisor.
TASK NUMBER: 040.004 OBJECTIVE NUMBER: 040.004.A PLANT HATCH JTA IMPORTANCE RATING:
RO 3.07 SRO 2.81 KIA CATALOG NUMBER: 2.2.12
( KIA CATALOG JTA IMPORTANCE RATING:
RO 3.7 SRO 4.1 OPERATOR OPERA TOR APPLICABILITY: Nuclear Plant Operator (NPO)
IGENERAL
REFERENCES:
Unit 1 34SV-SUV-019-1 current version)
(current version lREQUIRED MATERIALS: Unit 1 34SV-SUV-019-1 (current version)
Calculator APPROXIMATE COMPLETION TIME: 13.0 Minutes SIMULATOR SETUP: NIA
LR-JP-40.05-04 UNITl UNIT 1 READ TO THE OPERATOR INITIAL CONDITIONS:
- 1. Unit One is operating at 100% power.
- 2. The time is 0800 and Unit One has just received a High Drywell Pressure alarm.
- 3. The DryweU Drywell Floor Drain (DWFD) sumps have just been "pumped down."
- 4. 34AB-T23-002-1, "Small Pipe Break Inside Primary Containment," is in progress.
- 5. 34SV-SUV-019-1, "Surveillance Checks," is in progress and lAW IAW section
( 7.25.1, step 8, Drywell Equipment Drain 24 Hour Leakage has been calculated to be 1.20 gpm.
INITIATING CUES:
- Complete Drywell Floor Drain Leakage, section 7.25.2 of 34SV-SUV-019-1, "Surveillance Checks"
- Evaluate the results and report to the Shift Supervisor (SS).
(
ATTACHMENT I (VI)
(Excerpt from 34SV-SVV-OI9-I)
For INITIAL Operator Programs:
For OJT/OJE; ALL PROCEDURE STEPS must be completed for Satisfactory Performance.
For License Examinations; ALL CRITICAL STEPS must be completed for Satisfactory Performance.
START TIME: _ _ __
- 1. Operator identifies the procedure Operator has identified I UNSAT SAT IUNSAT needed to perform the task. 34SV-SUV-019-1 34SV -SUV-019-1 as the correct procedure.
PROMPT: WHEN the Operator addresses getting procedure 34SV-SUV-019-1, Section 7.24, PROVIDE the Operator a copy of Attach I and Attach 2.
PROMPT: IF addressed by the Operator, as the SS, INFORM the Operator that another Operator has already performed the DWED Leakage Rate check.
NOTE: The initial conditions stated the sumps had just been pumped down.
(\
- 2. Select the correct integrator. Operator has selected DWFD SAT/UNSAT leakage integrator 1G11-K601.
IGII-K601.
- 3. Operator reads the integrator. Operator has obtained DWFD SAT/UNSAT reading of 4017.0 (+/-0.5).
- 4. Operator completes Attachment 1. Operator has completed SAT/UNSAT Attachment 1 with the same results as Attachment 3 (+/-0.5 gpm).
PROMPT: IF the Operator addresses verification of calculations, INFORM the operator that will be done after he has completed his analysis.
- 5. Operator determines DWFD 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Operator records step 16 at < 5 SAT IUNSAT SAT/UNSAT leakage. gpm (meets Tech Specs).
- 6. Operator determines DWFD 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Operator records step 22 at > 5 SAT/UNSAT leakage leakage. gpm (exceeds Tech Specs).
- 7. Operator determines DWFD DWFD and Operator records step 16 at < 30 SAT/UNSAT DWED Total Leakage. gpm (meets Tech Specs).
( **8. Operator determines Differential Floor Operator records step 25 at > 2 SAT I UNSAT SAT/UNSAT
\
Drain Leakage gpm (exceeds Tech Specs).
ATTACHMENT 1I (Ul) (UI) 34SV-SUV-OI9-1)
(Excerpt from 34SV-SUV-019-1)
PROMPT: IF addressed by the Operator, as the Shift Supervisor, INFORM the Operator that a second Drywell Floor Drain leakage rate check is not desired at this time.
END TIME:,
TIME:- _ _ __
NOTE: The terminating cue shall be given to the Operator when:
- With no reasonable progress, the Operator exceeds double the allotted time.
- Operator states the task is complete.
( TERMINATING CUE: We will stop here.
ATTACHMENT 1 (Vl)
(Excerpt from 34SV-SVV-Ol9-l) 7.25 PANEL - INSTRUMENT I TECH SPEC. NOTE REAC TIS -OPER 0000 0800 1600 CONTD. MODE LIMIT IHII-P613:
IH11-P613: - IGll-K601, IG11-K601, Floor Drain Leakage 0000 Actual Time (9) F.l 3752.9 Present Reading (10) F.l Yesterday's Reading Time (11) F.l 0000 0800 1600 Yesterday's Reading (12) F.l 3723.9 3732.9 3744.1 29.0 Difference (10) - (12) (13) F.l 290 Difference Conversion (13) X 10 (14) F.l 1440 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> elapsed minutes (15) F.l FF,F.2, .20 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> leakage (14) / (15) (16) :: 5 gpm F3
( Previous 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reading (17) F.l 3744.1 3752.9 7.25.2 1,2,3 1600 0000 Previous 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reading time (18) F.l 8.8 8 Hour difference (10) - (17) (19) F.l 88 Difference conversion (19) X 10 (20) F.l 480 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> elapsed minutes (21) F.l FF, F.2, .18 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> leakage (20) / (21) (22) :: 5 gpm F3
.16 .16 .18 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> leakage from previous (23) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.55 Total Leakage (8) + (16) (24) :: 30 gpm Differential Floor Drain (25) FF .02 Leakage (22)-(23) ::2 gpm in (SR 3.4.4.1) Mode 1 Initials CME Time ARB CalculatIOns CalculatiOns Verified Venfied 0000 0800 1600 DATE
ATTACHMENT 2 DRYWELLLEAKAGEINTEGRATORS DRYWELL LEAKAGE INTEGRATORS IGII-K601, Floor Drain Leakage 3 9 640 1 7:--
(
GII-K603, Equipment Drain Leakage
...,=
396 3 9661-(
Attachment 3: Evaluator Use Only (not a handout)
REAC T/S OPER TIS 7.25 PANEL-INSTRUMENT/TECH P ANEL-INSTRUMENT/TECH SPEC NOTE MODE LIMIT 0000 0800 1600 lH11-P613 IH11-P613 -lGl1-K601,FloorDrain
- IG11-K601, Floor Drain F 1,2,3 Leakage Actual Time 9 0000 0800 Present Reading lO 10 3752.9 4017.0 Yesterday's Reading Time 11 0000 0800 1600 Yesterday's Reading 12 3723.9 3732.9 3744.1 Difference (10) - (12) 13 29.0 284.1 Difference Conversion (13) x 10 14 290 2841 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> elapsed minutes 15 1440 1440 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> leakage (14) /I (15) 16 S; 5 gpm .20 1.97 Previous 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reading 17 3744.1 3752.9 Previous 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reading time 18 1600 0000 8 Hour difference (10) - (17) 19 8.8 264.1 7.25.2 Difference conversion (19) x 10 20 88 2641 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> elapsed minutes 21 480 480 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> leakage (20) I/ (21) 22 S; 5 gpm .18 5.50 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> leakage from previous 23
( 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> .16 .16 .18 Total Leakage (8) + (16) 24 S; 30 gpm 1.55 2.17 Differential Floor Drain 25 S; 2 gpm Leakage (22) - (23) 4.2-lO, item 1 and 2, (Table 4.2-10, T.S.4.6.G) .02 5.34 Initials CME Time 0003 Calculations Verified 0000 0800 1600 DATE
(
Page 7 of7
LR-JP-40.05-04 UNITl READ TO THE OPERATOR INITIAL CONDITIONS:
- 1. Unit One is operating at 100% power.
- 2. The time is 0800 and Unit One has just received a High Drywell Pressure alarm.
- 3. The Drywell Floor Drain (DWFD) sumps have just been "pumped down."
- 4. 34AB-T23-002-1, "Small Pipe Break Inside Primary Containment," is in progress.
- 5. 34SV -SUV-019-1, "Surveillance Checks," is in progress and IAW section 34SV-SUV-019-1,
(
7.25.1, step 8, Drywell Equipment Drain 24 Hour Leakage has been calculated to be 1.20 gpm.
INITIATING CUES:
- Complete Drywell Floor Drain Leakage, section 7.25.2 of34SV-SUV-019-1, "Surveillance Checks" SUV-019-1,
- Evaluate the results and report to the Shift Supervisor (SS).
SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH SURVEILLANCE PROCEDURE 1 OF 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR G. L. Johnson DATE 11/19/02 DATE:
N/A N/A SSM 1 PM SSMI N/A N/A DATE N/A N/A 8-7-09 1.0 OBJECTIVE This procedure contains those surveillance requirements of Technical Requirements Manual (TRM) 1 Technical Specifications (TS) such as channel checks, level, pressure AND temperature records AND other checks performed by the Operators without the need for additional procedures.
This procedure satisfies, in part OR in total, the requirements of the TRM 1 TS listed below:
- SR 3.3.1.1.1 for 3.3.1.1-1 (1.a.),(2.a.),(2.b.),(2.c.),(2.e),(2.f),(3.),(4.),(6.)
- SR 3.3.1.2.1, SR 3.3.1.2.3, SR 3.3.1.2.4, and 3.3.1.2.5 for 3.3.1.2-1 (1.)
- SR 3.3.3.1.1 for 3.3.3.1-1 (1.), (2.a.), (2.b.), (2.c.), (2.d), (3.a.), (3.b.), (4.a.), (4.b.), (4.c.),
(5.), (9), (10), (12)
(
- SR 3.3.4.2.1 for 3.3.4.2(a.)(b.)
- SR 3.3.5.1.1 for 3.3.5.1-1 (1.a.),(1.b.),(1.c.),(1.d.),(2.a.),(2.b.),(2.c.), (2.d.), (2.e.), (2.g.), (2.9.),
(3.a.), (3.b.), (3.c.), (3.e.), (3.f.), (4.a.), (4.b.), (4.d.), (4.e.), (4.f.), (5.a.), (5.b.), (5.d.),
(5.e.), (5.f.)
- SR 3.3.5.2.1 for 3.3.5.2-1 (1.),(2.)
- SR 3.3.6.1.1 for 3.3.6.1-1 (1.a.),(1.c.),(1.e.),(2.a.),(2.b.),(2.c.),(2.d.),
(1.a. ),(1.c.),(1.e. ),(2.a. ),(2.b. ),(2.c.),(2.d.), (2.e.), (3.a.), (3.b.),
(4.9.), (4.h.),
(3.c.), (3.d.), (3.e.), (3.f.), (3.h.), (3.i.), (4.a.), (4.b.), (4.c.), (4.d.), (4.e.), (4.g.),
(5.a.), (5.b.), (5.d.), (6.a.), (6.b.)
- SR 3.3.6.2.1 for 3.3.6.2-1 (1.),(2),(3.),(4.)
- SR 3.3.6.3.1 for 3.3.6.3-1 (1.),(2.)
- SR 3.3.8.1.1 for 3.3.8.1-1 (3.a.)
- SR3.1.3.1, SR 3.1.3.1, 3.1.6.1, 3.1.7.3, 3.1.8.1, 3.3.7.1.1, 3.4.4.1, 3.4.5.1, 3.4.7.1, 3.4.8.1, 3.4.9.7, 3.4.10.1, 3.5.1.2, 3.5.2.1, 3.5.2.2.a., 3.5.2.2.b., 3.5.2.4, 3.5.3.2, 3.6.1.4.1, 3.6.1.5.1, 3.6.1.7.1, 3.6.2.2.1, 3.6.1.8.1, 3.6.2.1.1, 3.6.3.2.1, 3.6.4.3.2, Unit 2 SR 3.6.4.3.2, 3.7.2.1, 3.7.4.2, Unit 2 SR 3.7.4.2, 3.8.1.3, 3.9.2.1, 3.9.4.1, 3.9.5.2, 3.9.7.1, 3.9.8.1, 3.10.2.1, 3.10.2.2, 3.10.8.1 3.10.2.2,3.10.8.1
- TS 5.5.8.a., 5.5.4.f., 5.5.7, Unit 2 TS 5.5.7 MGR-0002 Ver. 8.1
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 20F73
(
DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33
- TSR 3.3.3.1 for T3.3.3-1 (2.),(3.),(4.),(5.),(8.),(9.)
- TSR 3.3.5.1 for T3.3.5-1 (2.),(3.),(5.),(6.),(7.a.),(7.b.)
- TSR 3.3.7.1 for T3.3. 7-1 (1.),(2.),(3.),(4.),(5.)
forT3.3.7-1(1.),(2.),(3.),(4.),(5.)
- TSR 3.3.8.1 for T3.3.8-1 (1.),(2.)
- TSR 3.3.9.1
- TSR 3.3.11.1, TSR 3.3.13, TSR 3.4.1.1
- ODCM, Table 2-1 (2.), (4.)
- ODCM, Table 3-1 (1.a.), (1.d.), (3.a.), (4.a.)
2.0 APPLICABILITY This procedure applies to Unit 1 Control Room instrumentation and other plant instrumentation which is directly related to Control Room surveillance requirements. This procedure is performed daily.
(
3.0 REFERENCES
3.1 Technical Specifications, Unit 1 3.2 ODCM 3.3 Technical Requirements Manual, Unit 1 3.4 DI-OPS-55-0193, Computerized Rounds 3.5 A-16397, Instrument Setpoint Index Drawing, Unit 1 G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 30F73 3 OF 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 4.0 REQUIREMENTS 4.1 PERSONNEL REQUIREMENTS The number AND qualification level of Operations personnel performing this procedure will be determined by the Shift Supervisor.
4.2 MATERIAL AND EQUIPMENT N/A - Not applicable to this procedure 4.3 SPECIAL REQUIREMENTS 4.3.1 All readings identified as unsatisfactory OR NOT meeting the limit in the TS/Operating Limit column must be circled in red, initialed by a Licensed Operator, AND reported to the Shift Supervisor. !E readings were taken using a computer, readings identified as unsatisfactory OR NOT meeting the minimaxmin/max limits in the computer will be reviewed by the Shift Supervisor.
4.3.2 The following notes apply as designated in the NOTES column:
A. WHEN total time reaches 650 hours0.00752 days <br />0.181 hours <br />0.00107 weeks <br />2.47325e-4 months <br />, the filter must be tested per 42SV-T46-003-1, Testing of SBGT Filter Trains OR 42SV-Z41-002-0, Testing of Control Room Habitability Trains.
B. Check (...J) OR enter "SAT" IF item satisfactory. Enter "UNSAT" AND circle in red!E item is unsatisfactory.
C. NOT a TRM I/ TS instrument OR requirement.
D. Check that the green status light is lit, !E installed; the gross failure light is OFF; AND the red tripped light is OFF UNLESS the instrument is tripped due to a valid plant condition (Le., low flow Core Spray pump AlB tripped indication red light ILLUMINATED WHEN (i.e.,
Core Spray pump NOT operating). Circle in red !E item is unsatisfactory.
E. Main power switch on Analyzers A & & B (Panels 1P33-P601A & & B) must be placed in the ANALYZE position one hour before taking readings.
F.1 Enter "NI A" !E the integrator is OR was inoperable due to drifting OR is OR was NOT "N/A" required to be operable. For actual time AND present reading, enter data WHEN integrator is restored to service OR required to be operable.
F.2 Use 34S0-G11-013-1 to calculate leakage!E integrator drifting. Circle the results in red OR, IF using HHC, add note stating integrator drifting.
F.3 IF an abnormal increase in Drywell Floor Drain leakage has been noted AND Drywell average temperature is < 100'F , request Engineering to evaluate the need to secure some drywell cooling fans. This is due to the possibility of excessive moisture
( condensation at the low drywell temperature.
G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 40F73 40F 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 G. IE. Process Computer is unavailable, use Attachment 1, otherwise, attach a Rod
!E Pattern Log (OD 7) printout.
H. Day shift will RECORD the run time of SBGT filter trains AND Control Room Filter trains from 1200 EASTERN Time one day to 1200 EASTERN Time the next day AND THEN calculate the Total Run Times.
I. Operations Personnel initials required.
J. Refer to Attachment 2 for Torus Temperature Monitoring.
K. The Post-LOCA radiation monitors 1D11-K622A AND 1D11-K622B are to be used as a pre-planned alternate for 1D11-K621A AND 1D11-K621 B, High Range Radiation Monitors IF the high range instruments are inoperable. See TS Table 3.3.3.1-1, Item 5.
L. IF Drywell average temperature is > 135'F, confirm OR PLACE Drywell Cooling System in Additional Cooling Operating Mode per 34S0-T47-001-1.
IF Drywell average temperature is < 120'F, confirm OR PLACE Drywell Cooling System in Normal Operating Mode per 34S0-T47-001-1.
( IF Drywell average temperature is < 100'F 1001= AND an abnormal increase in Drywell Floor Drain leakage has been noted, request Engineering to evaluate the need to secure some drywell cooling fans. This is due to the possibility of excessive moisture condensation at the low drywell temperature.
M. For the following ranges of Reactor Coolant temp. with Rx. head installed, utilize the applicable formula to obtain level.
=
L = Actual level
=
LI = Indicated level on 1B21-R605
< 200°F L = .74 LI + 9 2: 200'F to < 300'F
.::: L = .75 L 1+10 L=.75LI+10 2: 300'F to < 400'F
.::: L = .80 L 1+9 2: 400'F to < 500'F
.::: L = .86 L 1+7 With Rx Coolant temp 2: 500'F, no correction is necessary for 1B21-R605 re cord temp.:::
indicated level AND N/A corrected level. Obtain Reactor Coolant temp. from 1 B31-R650AlB, 1 1B31-R650AlB, E41-R605 pt 1 OR2, 1E41-R605 OR 2, OR equivalent.
Corrected Level = indicated level IE.
IF the temporary reference leg is connected.
(
G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 5 OF 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 N. Corrected Level = indicated level-151£ level-15 IF the reactor is in COLD SHUTDOWN.
Corrected Level = indicated level IF > 212'F. N/A corrected level.
O. If required, adjust RtF RlF AND Rx Bldg. to outside air delta pressure using 34S0-T41-006-1 AN D 34S0-T41-005-1 41-005-1..
P. Check that no trips OR alarms are present. Check that Self-Test Status indicates OK.
Q. Alternate the DIW sump pumps in the PTL position on Monday. Check ('..J) (-V) when completed.
R. TRM I TS recorders are required to be advancing, the pen inking, AND the pen must be indicating the parameter value as indicated by a channel check BEFORE it can be considered operable. Acceptable instrument reading is WHEN pen is properly on-scale AND neither upscale OR downscale. Refer to the "Video Graphic Recorders" section of 31 GO-OPS-007-0, GO-OPS-OO? -0, Shift Logs and Relief of Personnel, for additional information concerning proper operation of paperless recorders.
(
(
\
G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 6 OF 73 60F73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 S. IE recorders 1T47-R611 AND/OR 1T47-R612 AND/OR SPDS are INOP, THEN substitute subject reading from SPDS OR from I&C test equipment.
For Torus water temperature readings per following indicated Test Points use Voltage to Temperature Conversion Chart below to determine equivalent temperature. Add note/comment indicating that instrument is INOP, AND source of reading.
Alternate Reading Test Point-VDC Sensor (I nstrumentJPanel) nstrumentiPanel) fiQure below)
(see figure 1T4B-N301A 1T4B-N301 A 1T4B-K610/1H11-P691 1T4B-K610 /1 H11-P691 [ 3 (+) & 3 (-)]
A [3(+)&3(-)]
1T4B-N302A 1T4B-K610/1H11-P691 1T4B-K61 0 /1 H11-P691 B [4(+)&4(-)]
1T4B-N303A 1T4B-K611 /1 H11-P691 1T4B-K611/1H11-P691 A [3(+)&3(-)]
[3 (+) & 3{-)]
1T4B-N304A 1T4B-K611 /1 H11-P691 B [4(+)&4(-)]
1T4B-N305A 1T4B-N30SA 1T4B-K612/ 1H11-P691 1T4B-K612/1H11-P691 A [3(+)&3(-)]
1T4B-N306A 1T4B-K612/1H11-P691 [ 4 (+) & 4[-) ]
B [4(+)&4(-)]
1T4B-N307A 1T4B-K613/1H11-P691 1T4B-K613/1 H11-P691 A [3(+)&3(-)]
[ 3 (+) & 3{-)]
1T4B-N30BA 1T4B-K613 /1 H11-P691 1T4B-K613/1 B [4(+)&4(-)]
[4 (+)&4J-)]
1T4B-N309A 1T4B-K614/1H11-P691 1T4B-K614/1 H11-P691 A [3(+)&3(-)]
[ 3 (+) & 3 (-) ]
1T4B-N310A 1T4B-K614/1H11-P691 1T4B-K614/1 H11-P691 B [4(+)&4(-)]
1T4B-N311A 1T4B-K615/1H11-P691 1T4B-K61S/1 H11-P691 A [3(+)&3(-)]
(
B 8
- A leeel leeel G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 7 OF 73
- l. DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 VOLTAGE TO TEMPERATURE CONVERSION CHART Ii>
I/~
TEST EQUALS TEST EQUALS TEST EQUALS TEST EQUALS TEST EQUALS READING READING READING I}!i) READING i READING
.(VDC)
(VDCJ (OF) j"Ft i> (VDC) (OF) (VDC) (OF) (VDC) (OF) i' (VDC) (OF) 0.00 50 !> 2.00 90 I' 4.00 130 ;; 6.00 170 :' 8.00 210 0.05 51 ", 2.05 91 ".,'.'... 4.05 131 h,. 6.05 171 8.05 211 0.10 52 ."
2.10 92 Ii' I
4.10 132 '. :;,
I>,
6.10 172 '>. 8.10 212 0.15 0.20 0.25 53 54 55 'c.
2.15 2.20 2.25 93 94 95 I*".
1\\
4.15 4.20 4.25 133 134 135 i\'
6.15 6.20 6.25 173 174 175 8.15 8.20 8.25 213 214 215 0.30 56 ~';' 2.30 96 I* 4.30 136 I: 6.30 176 '/ 8.30 216 0.35 57 r:: 2.35 97 I*; 4.35 137 I';:' 6.35 177 ..'; 8.35 217 0.40 58 .,\, 2.40 98 4.40 138 I:. 6.40 178 ; 8.40 218 I"
0.45 0.50 59 60
\:
2.45 2.50 99 100 ,:
1'<, 4.45 4.50 139 140 i.
6.45 6.50 179 180 i.
8.45 8.50 219 220 0.55 61 rr. 2.55 101 It; 4.55 141 (ic' 6.55 181
, 8.55 221 0.60 62 0; 2.60 102 ";:. 4.60 142 :i 6.60 182 8.60 222 0.65 63 '.'" 2.65 103 .* ' 4.65 143 : 6.65 183 8.65 223 0.70 64 ; 2.70 104 i; 4.70 144 ,', 6.70 184 8.70 224 V) 0.75 0.80 0.85 65 66 67
.i
.. ,~.
2.75 2.80 2.85 105 106 107 I' 4.75 4.80 4.85 145 146 147
"'<..'* 6.75 6.80 6.85 185 186 187
/t i'
8.75 8.80 8.85 225 226 227 0.90 68 ........,.;: 2.90 108 4.90 148 ;. 6.90 188 ."; 8.90 228 0.95 69 !' 2.95 109 ,- 4.95 149 ) 6.95 189 " .. ' 8.95 229 1.00 70 ,-.~/ 3.00 110 >:. 5.00 150 ;" 7.00 190 'i' . 9.00 230 1.05 71 'r" 3.05 111 .r 5.05 151 .:) 7.05 191 ;> 9.05 231 i) 1.10 1.15 72 73 .F.
3.10 3.15 112 113 ,:'X 5.10 5.15 152 153 7.10 7.15 192 193
'.:{
9.10 9.15 232 233 1.20 74 ..'.*,'.**.. 3.20 114 ...*'. ',. 5.20 154 ",;';i, 7.20 194 < 9.20 234
" 9.25 1.25 75 ".""""
3.25 115 ',' , 5.25 155 7.25 195 235 1.30 76 3 3.30 116 5.30 156 :
7.30 196 c. 9.30 236 1.35 77 i; ... ; 3.35 117 ; 5.35 157 ". ? 7.35 197 9.35 237 1.40 78 '}C 3.40 118 <-,< 5.40 158 '(i 7.40 198 <, 9.40 238 1.45 79 3.45 119 5.45 159 .;, 7.45 199 .:<.', 9.45 239 1.50 80 *r 3.50 120 i 5.50 160 i;' 7.50 200 I, .* 9.50 240 1.55 81 3.55 121 5.55 161 .( 7.55 201 2 9.55 241 1.60 82 3.60 122 ":i 5.60 162 ;' 7.60 202 :3 9.60 242 1.65 83 [t* 3.65 123 { 5.65 163 ;::1 7.65 203 I> 9.65 243 1.70 84 3.70 124 '.' 5.70 164 l 7.70 204 I'e 9.70 244 1.75 1.80 85 86 I,"' ; 3.75 3.80 125 126
}'
5.75 5.80 165 166
~C 7.75 7.80 205 206 I.,* 9.75 9.80 245 246 87 1"'- 3.85 127 5.85 167 207 I.**.*:..: 9.85 247 1.85
- 7.85 1.90 88 I:; 3.90 128 < 5.90 168 ,.: 7.90 208 1'\ 9.90 248 1.95 89 1,< 3.95 129 **t 5.95 169 '>: 7.95 209 I: 9.95 249 10.00 250 T. !E 1T48-R635 Pt 1 OR 2 is INOP, THEN substitute value from SPDS by subtracting Torus pressure from Drywell pressure.
The numerical value of the instrument listed is to be recorded for those instruments NOT designated with note "B" UNLESS otherwise specified.
(
G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 8 OF 73 80F DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 U. Swap Suction Source to Torus, obtain O2 reading, WHEN stabilized, THEN swap back to the Drywell per 34S0-P33-003-1, Drywell & & Torus Atmosphere Oxygen Analyzer System.
V. Min. and Max. limits may be added to the computerized round when not specified by procedure. This note will only appear in the computerized round when applicable. If reading is abnormal due to these added limits, a Condition Report (CR) is NOT required to change the limit. If changes to these limits are needed, notify (Le., email)
Software Support personnel so that the limits may be investigated and changed. The following notes are used to specify when either the Min or Max. limit has been added.
V.1 Minimum limit not specified by procedure.
V.2 Maximum limit not specified by procedure.
W. IF the specified component OR reading is NOT available (Le., OCR DCR work in progress),
THEN substitute values for the reading as follows:
- Record value from 1B21-R623A Fuel Zone Range (Uncompensated). IF 1B21-R623A Fuel Zone Range is NOT available, THEN substitute value from 1B21-N685A H11-P925 (1 H 11-P925 panel) AND initiate RAS for inoperable instruments.*
- Record value from 1B21-R623B Fuel Zone Range (Uncompensated). IF 1B21-R623B
( Fuel Zone Range is NOT available, THEN substitute value from 1B21-N685B (1 H11-P926 H 11-P926 panel) AND initiate RAS for inoperable instruments.*
instruments. *
- IF 1 B21-R623A Fuel Zone Range (Compensated) is NOT available, THEN record - 317 (downscale).*
- IF 1B21-R623B Fuel Zone Range (Compensated) is NOT available, THEN record - 317 (downscale).*
- IF 1B21-R623A Wide Range (Uncompensated) is NOT available, THEN substitute value from 1B21-R604A (1 H11-P603 panel) AND initiate RAS for inoperable instruments.*
- IF 1 B21-R623B Wide Range (Uncompensated) is NOT available, THEN substitute value from 1B21-R604B (1 H11-P603 panel) AND initiate RAS for inoperable instruments.*
- IF 1 B21-R623A Wide Range (Compensated) is NOT available, THEN record - 150 (downscale).*
- IF 1 B21-R623B Wide Range (Compensated) is NOT available, THEN record -150 (downscale).*
- IF 1B21-R623A RPV pressure is NOT available, THEN record 1B21-N690A (1H11-P927 panel) value.*
- IF 1 B21-R623B RPV pressure is NOT available, THEN record 1 B21-N690D (1H11-P928 1B21-N690D (1 H11-P928 panel) value.*
- Enter note / comment for substituted value(s) recorded.
recorded .
.!E the operating condition specified for a reading
!E does NOT match the current plant operating mode, THEN place "N/A" as the value for the reading.
{
\,
G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 9 OF 73 90F73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 X. For instruments in which an indication of "Upscale" OR "Downscale" is expected based on the value of the parameter measured under the current operating mode AND the scale of the instrument, enter the min or max value that may be read on the gauge PRIOR to the reading being Upscale or Downscale. For example, the new Fuel Zone RWL bargraph displays on 1B21-R623A & B normally indicate upscale at 100% power. The max value for this RWL instrument range is -17 and therefore a value of -17 will be entered if upscale.
Y. !E 1P41-R603 IF 1 P41-R603 is INOP OR INVALID, THEN substitute value from one of the following:
- 3. 2P41-R372A!E 2P41-R372A IF Unit Two RBCCW HX A is in service,
- 4. 2P41-R372B !E IF Unit Two RBCCW Hx B is in service, or
- 5. I&C installed test equipment. !E IF alternate indication is used add Note/comment indicating source of reading.
Z. The operator will enter the Self Test page to check for error messages.
A FAILURE detected by the APRM self-test feature will be classified as a "CRITICAL" error OR "NON-CRITICAL" error.
A "CRITICAL" failure affects the operability of the instrument, AND causes alarm
( 603-210, APRM/OPRM TRIP, to annunciate.
A "NON-CRITICAL" failure, (which will NOT cause the above annunciator), does NOT render the APRM inoperable, but requires a CR to be written for I&C to investigate.
AA !E IF any LED fails to illuminate, initiate a CR. Failure of an LED does NOT render the APRM Voter inoperable.
BB. The value listed in the TS or Operating limit OR Min.! Max. limit column is the acceptable operating value that is more conservative than EITHER the TS / TRM allowable value, if applicable, OR its actual instrument setpoint for that associated plant parameter. !E IF an out of spec reading exists, initiate a CR AND notify (Le., email) Operations computer support personnel to see if there is justification for changing the value listed in the TS or Operating limit OR Min.! Max. limit column.
CC. IF either 1Z41-C009 or 1Z41-C010, Cable Spreading Room Fan, is found not running, then secure the other fan, 1 1Z41-C010 Z41-C01 0 or 1 1Z41-C009.
Z41-C009.
Initiate compensatory action to confirm twice per shift that the temperature in the cable spreading room is less than 105 degrees F. An Engineering Evaluation is required if the temperature reaches 105 degrees F.
(
G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 10 OF 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 DD.
DO. Reset Total Nitrogen Usage, 1T48-R602 Pt. 38 on 1H11-P654, on Saturday Day Shift by performing the following:
- 1. Press the right arrow to reach the "TOTALIZER" screen.
- 2. Open the front access panel on the recorder.
- 3. Press "STOP".
- 4. Press "Mem + Math"
- 5. Press "ENTER".
- 6. Press "Func".
- 7. Scroll to the next screen by pressing the top right most button.
- 8. Press "MATH RESET".
- 9. Press "START".
EE From "Control" screen, SELECT "pressure xmittr"xmittr" CONFIRM the following:
- 1) Throttle Transmitter #1, Throttle Transmitter #2, Throttle Transmitter #3 read "okay".
- 2) Confirm the following alarms are not present:
P1_P33 <R> PROCESSOR OFFLINE P1_P34 PROCESSOR OFFLINE P1_P35 <T> PROCESSOR OFFLINE S1_P264 <R> PROCESSOR OFFLINE
( S1_P265 PROCESSOR OFFLINE S1_P266 <T> PROCESSOR OFFLINE FF IF DWFD leakage is >0.5 GPM, perform the following:
- 1. Notify Management (Operations Duty Management AND Plant Hatch Duty Manager)
AND include DWFD leakage on the Morning Report under Major Equipment Impacting Operational Focus.
- 2. Each Operating crew will review 34AB-T23-002-1 at the beginning of shift AND log the DWFD leakage rate each shift.
- 3. Initiate a Condition Report and make a Control Room Log entry.
- 4. Engineering Support will include a graph of U-1 DWFD leakage in the Daily Status Report each Friday.
IF DWFD leakage is >1.5 GPM, perform the following:
- 1. Notify Management to schedule a Unit 1 shutdown based on DWFD leakage trend.
- 2. Initiate a Condition Report and make a Control Room Log entry.
- 3. Every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, calculate DWFD leakage and make a Control Room Log entry of the leakage rate.
IF DWFD leakage is >2.5 GPM, perform the following:
- 1. Perform a controlled shutdown of Unit 1 to repair the leakage.
- 2. Initiate a Condition Report and make a Control Room Log entry.
(
G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE E. I. HATCH PLANT E.I. 11 OF 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 GG If the alarm is inop, OR illuminated, the actual DP for the associated Core Spray loop must be confirmed to be more negative than the alarm setpoint once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or the associated loop must be declared inoperable.
HH If the Core Spray DP is greater than the limit (less negative), OR if 1E21-N004A(B) is inoperable, declare the associated Core Spray loop inoperable AND initiate a CR for I&C and Engineering to investigate.
II Prior to exceeding 10 degrees F difference, place RHR in torus mixing per 34S0-E11-010-1.
4.3.3 Readings are required to be obtained WHEN in the Reactor Mode listed as follows:
- 1. RUN
- 2. STARTUP & HOT STANDBY
- 3. HOT SHUTDOWN
- 4. COLD SHUTDOWN
- 5. REFUEL
- 6. At all times 4.3.4 Readings are required at the following frequencies as noted in the FREQ column
( a. Once I shift
\ b. Once 14 I 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
- c. Once I 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
- d. Once 17 I 7 days (day will be designated)
- e. Once every 31 days (day will be designated) 4.3.5 Surveillances required once 112 I 12 hr shift must be taken between 1200 and 1500 EASTERN Time on Day Shift AND between 0000 and 0300 on Night Shift.
4.3.6 ECCS Status Check is to be performed WITHIN 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of shift change per 34SV-SUV-018-1.
4.3.7 Surveillances required once I 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> must be taken within the time constraints established in step 4.3.5 for the once I 12 hr surveillances.
4.3.8 DIW Sump reading AND Fission Product reading are taken at times 0000, 0800, & 1600.
Surveillances required once/8 hours must be taken WITHIN 1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> of the designated time.
4.3.9 "N/A" will be entered for all surveillances NOT required by existing plant conditions.
4.3.10 Independent Verification, as described in 1 OAC-MGR-019-0, Procedure Use and Adherence, 10AC-MGR-019-0, will be required on each shift by a licensed operator for portions of this procedure where a "Calculations Verified" sign-off is given.
Verification of calculations is NOT required !E IF data is taken using the handheld computer.
G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 12 OF 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 4.3.11 !E IF an Operator fails to complete any signofflinitial spaces, an SS OR SOS may sign off/initial the space.
4.3.12 Night shift begins at 1900 EASTERN Time, Le., where a reading is specified for Saturday night shift, it will be taken from 1900 EASTERN Time Friday to 0700 EASTERN Time Saturday.
4.3.13 !E IF using the paper copy of this procedure, Attachment 3 is NOT required.
IF a word such as "INOP" is entered into a station in place of a value, any calculation NOTE: performed with that station will sUbstitute substitute a numeric value of zero for the word. IF the calculation is a channel check, appropriate action must be taken for the inoperable instrument AND to ensure adequate channel check of remaining operable instruments.
4.3.14 !E the procedure is to be performed using the computerized rounds software, the procedure beginning with 7.1 may be completed by completing the following:
( NOTE: 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> tour is NOT applicable in modes 4 AND 5. Performance AND printouts are NOT required.
4.3.14.1 Perform 8 and 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> tours for 34SV-SUV-019-1, in accordance with Attachment 5.
4.3.14.2 Complete Condition 5 Surveillance Checks sheet as required by plant conditions.
4.3.14.3 Complete Operating Orders sheet, as required by plant conditions.
4.3.14.4 Complete Safety Parameter Display System Status Checks sheets.
4.3.14.5 Complete Attachment 3, Computerized Surveillance Documentation.
4.3.15 IF more than one Operator collects data for a tour using the computer, one of the following actions will be taken:
4.3.15.1 Each Operator will log into the computer and will enter the data collected, OR 4.3.15.2 Information will be placed in the Notes section that identifies the other operator who collected the data. The note will include name and area of data, if applicable.
(Le., Cooling tower data taken by Terry Doe).
G16.030 MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 13 OF 73
(
DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
SURVEILLANCE CHECKS 34SV-SUV-019-1 33.33 4.3.16 The paper copy of the procedure will be used:
- IF the computer will NOT perform required calculations.
- IF the output from the computer is obviously incorrect AND represents a serious degradation of the computer system operation.
- IF the version number specified in the computerized round for the procedure is incorrect AND the Software Support Personnel can NOT resolve the problem.
AutoTour will NOT display all required outstanding surveillances for a Unit Mode change, therefore a paper copy of this procedure MUST be performed. (e.g., IF Unit is in Mode 4 AND NOTE: Mode 2 is inputted into AutoTour, at that time the Mode 2 surveillances will be displayed on AutoTour but some Mode 4 surveillances will NOT be displayed. In addition, surveillances that shift, day of the week OR month will NOT be displayed.
- IF a pre-planned Mode change for the Unit will take place during the shift that may require performance of outstanding surveillances.
( 4.3.17 IF the HHC fails in some manner, every effort is to be made to perform the surveillance However, lE full completion of the surveillance is impossible, a using the paper copy. However,!E CR is to be written documenting the incomplete performance of the surveillance. Notify Operations Support for assistance with trouble shooting.
4.3.18 IF the version number specified in the computerized round for the procedure is incorrect,
!E IF possible, contact the Software Support Personnel for assistance.
4.3.19 Any abnormal condition identified will be marked (Le., red circled) and documented/explained in the notes section of the computer.
5.0 PRECAUTIONS/LIMITATIONS 5.1 PRECAUTIONS None applicable to this procedure.
5.2 LIMITATIONS
!E IF items being compared to each other fail to meet the limits of a comparison check, AND it is NOT readily discernible which item has an operability concern, THEN a CR must be initiated to investigate the condition in order to assist in making the operability determination.
6.0 PREREQUISITES None applicable to this procedure.
G16.030 MGR-0001 Ver. 4
ATTACHMENT 1I (VI) (Ul)
(Excerpt from 34SV-SUV-Ol9-l) 34SV-SVV-OI9-1) 7.25 PANEL -INSTRUMENT
- INSTRUMENT II TECH SPEC. NOTE REAC TIS --OPER OPER 0000 0800 1600 CONTO.
CONTD. MODE LIMIT IHII-P613: - IG11-K601, Floor Drain Leakage 1H11-P613:
0000 Actual Time (9) F.l F.1 3752.9 Present Reading (10) F.l F.1 Yesterday's Reading Time (11 )
(11) F.l F.1 0000 0800 1600 Yesterday's Reading (12) F.l F.1 3723.9 3732.9 3744.1 29.0 Difference (10) - (12) (13) F.l F.1 290 Difference Conversion (13) X 10 (14) F.l F.1 1440 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> elapsed minutes (15) F.l F.1 FF,F.2, .20 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> leakage (14) / (15) (16) ~5gpm
- 5 gpm F3
( Previous 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reading (17) F.l F.1 3744.1 3752.9 7.25.2 1,2,3 1600 0000 Previous 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reading time (18) F.l F.1 8.8 8 Hour difference (10) - (17) (19) F.l F.1 88 Difference conversion (19) X 10 (20) F.l F.1 480 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> elapsed minutes (21) F.l F.1 FF,F.2, .18 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> leakage (20) / (21) (22) ~5gpm
- 5 gpm F3
.16 .16 .18 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> leakage from previous (23) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.55 Total Leakage (8) + (16) (24) ~30
- 30 gpm Differential Floor Drain (25) FF .02 Leakage (22)-(23) ~2 gpmin
- 2 (SR 3.4.4.1) Mode Model1 Initials CME Time ARB CalculatIons Venfied CalculatlOns 0000 0800 1600 DATE
DRAFT Southern Nuclear uclear E. I. Hatch Nuclear uclear Plant Operations Training JPM Admin 4, SRO Only TITLE REVIEW/APPROVE EQUIPMENT DANGER TAGOUT
(
A:tJ'fIIoB,..
AUTHOR . MEDIA NUMBER ****TIME*
TIME EFAGAN F. FAGAN LR-JP-25020-05A LR-JP-2S020-0SA 40 Minutes REcoMNffiNDEJ)kY'*.*
RECOMMENDED BY .* <.* . . . (.*
.iiArllR.ovE~B¥ APPROVED BY DATE N/R Energy to Serve Your World Wo rid SM SM
/
IISOUTHERN NUCLEAR OPERATING COMPANY PLANT E. I. HATCH Page 1 of 1 FORM TITLE: TRAINING MATERIAL REVISION SHEET Program/Course Code: OPERATIONS TRAINING Media Number: LR-JP-25020A
. . . . . . . . . . . *******.J2..C>A~ .*. . . . ;h.l.< .. . . *. . . .......\\ *. . . . *.*.A.lltb.()r~~;Supv'.~
Author's Supv's Rev. No. Date f.*.*.**.*..Reason>,"":7; for Revision Q~.....i} ~Z.Ini~bd~** Initials . Initials
(
(
LR-JP-25020-05 LR-JP-2S020-0S
( UNITt () UNIT2 UNIT 2 (X)
TASK TITLE: REVIEW/APPROVE EQUIPMENT DANGER TAGOUT JPMNUMBER: LR-JP-25020-05 LR-JP-2S020-0S TASK STANDARD: The task shall be completed when the operator has reviewed a Danger Tagout for RBCCW Pump 2A per procedure NMP-AD-003 "Equipment Clearance and Tagging".
TASK NUMBER: 300.016 OBJECTIVE NUMBER: 300.016.0 PLANT HATCH JTA IMPORTANCE RATING:
RO 3.20 SRO 2.74 KIA CATALOG NUMBER: 2.2.13 KIA CATALOG JTA IMPORTANCE RATING:
( OPERATOR APPLICABILITY: Senior Reactor Operator (SRO)
GENERAL
REFERENCES:
IUnit 2 NMP-AD-003 Equipment Clearance and Tagging (current version)
UJ-VVoJ NMP-AD-003-FOI Tagout Cover Sheet (current version)
NMP-AD-003-F02 Tagout Tag Listing (current version)
NMP-AD-003-F06 Clearance Tagout Log Sheet NMP-AD-003-F08 Tagout Preparation And Approval Checklist (current version) 34S0-P42-001-2 Reactor Building Closed Cooling Water System (current version)
NMP-OS-002 Verification Policy (current version)
NMP-OS-OO2 H-26054 Plant Drawing H-260S4 TRM REQIDREDMATERIALS: I Unit 2 NMP-AD-003 Equipment Clearance and Tagging (current version)
NMP-AD-003-FOI Tagout Cover Sheet (current version)
NMP-AD-003-F02 Tagout Tag Listing (current version)
NMP-AD-003-F06 Clearance Tagout Log Sheet NMP-AD-003-F08 Tagout Preparation And Approval Checklist (current version) 34S0-P42-001-2 Reactor Building Closed Cooling Water System (current version)
NMP-OS-002 Verification Policy (current version)
H-26054 Plant Drawing H-260S4 TRM
( APPROXIMATE COMPLETION TIME: 40 Minutes SIMULATOR SETUP: N/A
(
UNIT 2 READ TO THE OPERATOR INITIAL CONDITIONS:
- 1. The 2A RBCCW pump has developed high vibrations.
- 2. 2B and 2C RBCCW pumps are operating normally.
- 3. The Shift Manager has directed that 2A RBCCW pump be tagged out for Maintenance disassembly and inspection.
- 4. The following is being generated:
- Work Order
(
- Condition Report
- 5. Maintenance has requested door 2RW02, "East Corridor 112 El.
Radwaste", be left open for running of temporary air and electrical lines.
- 6. The E-Soms computer tagging system is out of service.
- 7. The BOP operator has manually prepared a Tagout.
- 8. You are the Shift Supervisor.
INITIATING CUES:
- 1. Review the Tagout and, if acceptable, sign as the reviewer and authorize the Tagout.
- 2. As a result of the Tagout, determine what, if any, are other administrative requirements.
For INITIAL Operator Programs:
O.JT/O.JE; ALL PROCEDURE STEPS must be completed for Satisfactory For O.IT/O.IE; Performance.
For License Examinations; ALL CRITICAL STEPS must be completed for Satisfactory Performance.
NOTE: Steps of this JPM may be completed in any order.
NOTE: Provide the candidate the attached Tagout Sheet and Cover Sheet.
START TIME:.
TIME: ___ _____
- 1. Operator identifies the procedure Operator has IDENTIFIED the SAT/UNSAT needed to perform the task. correct procedure as NMP-AD-003.
- 2. Verify the Tagout points and Operator determines the Tagout SAT/UNSAT boundary isolations points and boundary isolations are acceptable.
- 3. Verify and assess all potential Operator determines there are no I UNSAT SAT/UNSAT SAT automatic actions and/or effects on the automatic actions or effects on plant. the plant.
- 4. Review the Tagout points for the Operator reviews and determines
, correct: the following:
(
- Tag type used ** Caution and Danger Tag(s) SAT/UNSAT are the correct type.
- Component number and name ** 2P42FOO4A 2P42F004A should be SAT/UNSAT labeled as the SUCTION valve.
- Position ** 2P42FD005 drain should SAT/UNSAT be tagged OPEN.
- Sequence ** RBCCW Brkr, 2R23S003 SAT/UNSAT FR 2B, should be before the valve manipulations.
- Placement verifications required Operator determines there are no SAT I UNSAT SAT/UNSAT secondary verification requirements.
- 5. Review impact on equipment Operator determines that the open SAT/UNSAT (including Tech Specs) fire door requires a tracking Fire Protection RAS.
- 6. Sign Tagout as Reviewer and Operator determines that the SAT/UNSAT
(~'VTE:
Authorizer. signed..
Tagout can NOT be signed
(
dvTE:
- Operator may add special instructions to the Tagout.
- Operator may reposition the control switch tag also.
END TIME: _
( NOTE: The terminating cue shall be given to the operator when:
- With no reasonable progress, the operator exceeds double the allotted time.
Operator states the task is complete.
TERMINATING CUE: We will stop here.
NOTE: ** Indicates a Critical Task.
(
(
Southern Nuclear Operating Company Nuclear NMP-AD-003-F06
-~HERN~
-"'UTHERNA. Management Clearance Tagout Log Sheet Version 1.0 COMPANY
,Sa"JItYo"rWor/d" JStTUeYourWorltr Form Page 1I of 1I UNIT 2 Clearance Tagout Installed Released Type Number Number Component Reason for Tagout Date Initials Date Initials Affected 2-DT- 2P33-DT 2P33-F003 Repair air leak 1/24/09 BKW 1126109 1126/09 BKW 09 30000 2-DT- 2Ell-DT 2Ell-F007A Quarterly PM 2/05/09 ARB 2/06/09 MMG 09 30001 2-DT- 2T46-DT 2T46-D001B DOP Test DOPTest 3/15/09 CME 3/16/09 ARB 09 30002 DT 2-DT- 2P42- 2P42COOIA Repair Pump 09 30003
(
r'- ..
Southern Nuclear Operating Operatin2 Company Nuclear NMP-AD-003-POl NMP-AD-003-FOI
- "HERN...\.
HERNA COMPANY Management Tagout Cover Sheet Version 1.0
, Serve Your
&rJle Wdrt.r
}'Dill!" WDrUl" Form Page lof 1 Clearance: 2-DT-09 Tagout: 2P42-30003 Component Affected: 2P42COOIA
Description:
RBCCWPmp2A OPS Instructions: Insure pump vented and drained.
Holder Instructions: None
References:
Drawin~ H-26054, 34S0-P42-001-2 Plant Drawin2 Ta~out Attributes:
Tagout Attribute Attribute Description Value W orkDoc Holder List:
Number 1 Equipment ID Description 1st Verified 2M Verified T agout
~gout V'f' V en erI'fiIcatlOn:
Status Description Name Verification Date Prt)pared Prepared Prepared Jack Jones 10126/09 Reviewed Reviewed Authorized Authorized T~gs Verified Hung Tags Tags Verified Hung Removal Prepared Removal Prepared Removal Reviewed Removal Reviewed Removal Authorized Removal Authorized T~gJ Verified Removed Tags Tags Verified Removed Records Forwarded Records Forwarded Pagel of_
Southern Nuclear \:' ...~rating Company
\"yerating I Nuclear NMP-AD-003-F02 SOUTHERN COMPANY A Management Tagout Tag Listing Version 1.0 l:imvtQStr"i Ji,lIYWfm.t" EneTDwStTN YourWwlti' Procedure ----------
Page 1 of 1 2-DT-09 2-DT-09-2P42-30003 Tag Equipment Eguinment Placement Restoration Num Type Equipment ill ID Verif Seq Configuration 1st 2M 2nd Verif Seq Configyration 1st 20a na Description/Location DescriptioniLocation Notes Verif Verif Notes Verif Verif 2P42F004A HW 2P42FOO4A CLOSED Danger CV 1 RBCCW Pump 2A Disch V1v Vlv 2P42F005A HW 2P42FOO5A CLOSED Danger CV 2 RBCCW Pump 2A Disch V1v Vlv ,
2P42FD005 HW CLOSED Danger CV 3 RBCCW Pmp 2A Dis Line drn I 2P42F044A HW OPEN Danger CV 4 RBCCW Pmp 2A casing vent 2P42COOIA CS 2P42C001A OFFIPTL Danger CV 5 RBCCW Pmp 2A, 2H11P650 2R23S003 FR 2B RACKOUT Danger RBCCW Pmp 2A, 130 U2 CV 6 CIB Caution 2RW02Door CV 7 OPEN East Corridor 112 El.
Radwaste ---------- - - - - - - -
For Training Use Only - JPM Reference Page 7 of7
(
UNIT 2 READ TO THE OPERATOR OPERA TOR INITIAL CONDITIONS:
- 1. The 2A RBCCW pump has developed high vibrations.
- 2. 2B and 2C RBCCW pumps are operating normally.
- 3. The Shift Manager has directed that 2A RBCCW pump be tagged out for Maintenance disassembly and inspection.
- 4. The following is being generated:
- Work Order
(
- Condition Report
- 5. Maintenance has requested door 2RW02, 2RW02, "East Corridor 112 El.
Radwaste", be left open for running of temporary air and electrical lines.
- 6. The E-Soms computer tagging system is out of service.
- 7. The BOP operator has manually prepared a Tagout.
- 8. You are the Shift Supervisor.
INITIATING CUES:
- 1. Review the Tagout and, if acceptable, sign as the reviewer and authorize the Tagout.
- 2. As a result of the Tagout, determine what, if any, are other administrative requirements.
(
1>909~-H lP.!I~{.._ ~~ ~
~P ----tII,oos""'". t~ l:"_2.!~.¥~t_-,~ /"...--"'\;------,
... ' DIMiNo WTR.
11."".
y._---' ,
TIJReIN£
,pa'MAE-t ~__ '_ , ALL E E \ LOCATED INSIlE
_DING EXCEPT INSOE I THE R, .OING
, I' 6OO1(Ah BOO'(I\I. AJOI(A,6,4 C)*
"OOI(A,Il!4 C) f---.~J (*vt..,._~
I ~-
~ __ J IQ -t r ..A:'\ NOTES E 410;; '\Y - - -
~
I I l~1J. tflM NUMeERS AlE 16. PRlFlKED IV.2""1 au; NOTE 8 _______ .,
i G'\f.-
1-0'" R TO COMPLETE JIll'/. f.IS"N4 UHI..E65 'HOWIf
+
/ ~o .oTHIh'I&I, L MY JlDDITIONN. KI'li miNT VENTS ~D leA\!
~-_;.PiNGaV ----..,~ l' PDINT DR"INS 1ll BL ADDEtI BY FI[LI)
AS REqtJII\£O.
Ot'HS'R' 10
,J,;pUN, l,PU"P COMNECTION$
COt.lNECTIONS "US $IZES W'LL WILL SitSE 8PttIFIEO SPE.CIFIED III rN TNI iY THi PUMP MANUFAC.TUUR.
MANI.JFA(.TURU.
I otVAL'ItS FOO',
':VALV£6 FOOl, FOCI~FtlJ.9;IF'OU FOOl, FCOI. FOO!l,FOS~,F03d ,1'051 ,FOS'1 I rQ24, F024, ANO F03I f:j~L 81 lotATIi:D 1'036 SHALL.Bl LOCA.1!iD IN I~ AN
.... N ARU
"LA I CHRW CNRW' 2P"l:*HIE"4~ Iii'" 1& AtCU&I15L1 TIl ...T I. Act;(fo&I"Ll JlURINEt P(JRINq HOItMAL NORMAL PLANT P!.A~T I ,,'£lATION.
OPUATION. HANPW'H!!L.
HANt)WJ.4t!L. FOil FOil. ~I.YE I\(LVt trOll, FlO!'
-'H"J.U. B£
$!fALL. ..ce tiSl8LE B~ 11.<<: t&&18I.E WRINlt DURIN, NORMAL NORNAl PLANT PI-ANT OPiRATION OPIRATION Al.iO. ALiO.
UINt S,LIN.Ii: ItOOTI"P I'.OUT£9 10 iW4~TI: SUR",
TO WAftn SUR41i: TANK T",WK ""Y MAo"!' 8£ S£ CON"N£CTtD TO IlI-ofR CONNECTED RJ.fR lift" DJtAiN LINE.
&V.sT£.M ORAIN L.INE..
'."~L ,lLAAMS AND INI)IC.\TINCi t:AU.lLARMS !NDJCA.T1NG WiHT$ tlCiHTS AlE A.1lE.
LOCAT!O IN THE MAIN LOc:ATEC) MAIH" CON.TROI..
CONTROL AOOM. ROOM.
'1'.*A.WNUt.lCIA.iO~ 1.0c.
'7.*'NNUNCIA'I'OJ5, LOtAn!)...Ti.P ON MAIN HAlN CONTROL CONTROL.
CONTROL CONTROL. ROOM ROON PANt PANtL .. It, A.
a THI
- 8. .sERV1~1L WATER THE SEIU'It.E WIT1L1t PlfIlnullt.
p/U.nu~!. AT THii
--e THE PlitHAl4E DlStHARq£ OF THE HEAT !-lEAl tXC:HAN4U:S U*CHANijERS MUST MLIfIT 8ESf.
"f"SI(NIN.)
"f':JI(HI/oI.} HIIilMeR HIri;HCR THAN TI'IAti THEnoi[ Rettll' R&celll INl.lT INion too. PRI.~!iiURE ~
PRE!!oSURE. O¥ 'THE 1-1 ' UC~N"U'.
il-lE" HI'" 'W;CWAN(fERI.
- 9. DLtTED
!t. Oe:UTE.O PI $.'
p, 10,'"
- 10. .,- ITENS FUMI~tD WITK EQUIPMENT.
ITEMS FURlfI'5fl.eo EQuiPMENT.
"ili I I
n.
II_ VAU'E:S VALVES ON FOOOI THft,u f\'OOI FvOOI THRU T"l' THIS OW&.
TI1RU ;0020 THRIJ nozo OWG. ARE ME NUM8ERED "FD020 FOR ORI\IJt FVOW FOR NUI/\5f.RED ORI\IN VM.VES
~"'L~ES,
~""l.VE.S rOf\ VENT yl\LyE$
i L__ _
~ ... ", .Ilrururu REFERENCES
~tJl'II.£NCli:i RE'£RENtU ~KN. 'niH NO. ~UO'J!tj,1/O..
- 731 I. PtlO R\NCU '5'($,
LPiIORt.lCU<S.
2.P{II~
l,P(ID DUWI WolT~<< Irs.
OEI1H.j 'lMl£lt if$..
~ l~HO'"
tpl'*1010 2:.0"'1010 l{.~
\o\.~
14..2:Gooa.,
",,*%f.Oo\1
~ £ 6
- ..... ------~-,
i ,J,'ptJD
+, P(ID if. P~IO 'UEL CRD. IYIi.
J:pfIDC,IU,IYr.,
FUEL POQL C.OOt,IN~ JoYS.
POot. eOOlIMt bY5.
2(.11*1010 z(.n-IOIO llttHOIO Z,,-41-1010 J.I.tGA%
J.4-1t-COGo
~.u.o~
I-I*u.oae i'i ~I l.,tlD P:r:tlll(.~Y!!i.&Hl:n S.pflo M""(.~!.flHln tall-iC>IO
&IUI"IOIO H-Z600!
H'!6001 I I U'jlD PUNT
""ffO PlANT "Ul¥'"..me Wit W"ltR .s.ys.
lit ioY.li. ZP4r~lOla lPa{~1010 H,ZIO!.'
H,ZIQ'!'
.....,!. Ll 1:5 i
I P65~,
! ~
11'110*'"
4.DFiHjN
!:~f~N~~~~:;:M!a£~~~
U'LANT IKQUtUMIWT6 IO.P(ID RADW.u.rr.
RAlI, MON.
1, Pi/O P'.()::1i:,S RAD.I'ION.
II:E'O~HnJtlATjl)W~
DE5h1N RECOMKtNPATIQN60
'Y'.
~l)II-IOIO tDlHOIO UAll04(GrENt)
WlJ04(Gt£HI)
- .~~-!~O 111RU&aD>>
I:AlI*4OlO II-U02to 1~fi.IClIO If-tiel!
H-2r..OIl
~~~~
r--t2 IL"IP!)lQ.(J~!'TRU~[NT iY",*_
ILPIP1N4CIH5TRUHENT jt,llHPL.l~ .TATION
,UIIMPUNf! "TATION
\'2:..~LLI'I-JEU,~'1'5,P4\!)
I'&.DR't'WIU.I'tJlIJ.1HS.P4\()
ayl1.BOl" MZ~1OI0 Mt*IOIO tn~*'ZOlO
~~'!OIO 1+?i.O!&
ZP'!O~IOIO M.z"Oi6 ZP'lQw.IOIO "50~1
'51,c;.,-,
I-I.~
1-(-l'0'6
'r\"'o (4, MAC'tOf\
{4, RECti\!. ~~.
Re,/;,ClOf\ MetA:. ~I(", M-6.
M,(,.
M:t: ",pii~~iro\.-=
1"P.4I.c..QlDP.f.D~ P~ID
~e.iL\J5.tOlI.. 'e~ P410
~$~:::~
t~'2A-IOIO IN6M4 1G5I-IOIctI020fl-2fp04Z DAAl~o\GE M4D PURlfIC,t..'tlat M.
ORr\IN,I.,G£.o\HDPl.iRIFIC'-'lICflm.
~.w:.~!~~J:;o\~~~.
IMea IT...~ ......... -ro,..
It.&¥P>TiMI£O. 2n~-IO\O 2"....10 ~~c.m.
~~.""
~'f"
'IJN't,OJI
{'I l~W~ "'"
r.w-737 i
.t i
~ ,
F302.~
"')1.'>
("jy-n'J
'4 I"4K K
~p4a-Ht[-IlI:
iP42,'HEE- rE!
DW6. IH~~6if
.lP42HEr*"u . aP'41*Jof';'.I6~
CIWG. ",..a"o",
- wnl
'8£'
J--
TURStNE. BLI RtAtTOR' ILK..
LVN~~YOO1 WN-~"'OO1
~ Utl-rDO!>>
LVN~r(lOO!o lVli'~Wt tVN- ~I!'
L. __ ~ CRITICAL CCICUMIONT CCICUMENT
.-£ MPL NO. 2P42-1OiO OVY20 "'so..
H26DS' SOUTHERN .\..
COMPANY IVm!on:
V'Blon: 22.oIDaIt:
22,0 0011: 10-11-07 to-tll-07 I'REYlS[O pta ABN-HD07$7, REVl5EOPER4.BH-H00737, [OWIN I, HATCH NUCLEAR PLANT UNIT No.2 VER.I.o VER.l,O RU.CTOR BLDG, CLOSEO COOUNG WATLR SVS. PI:ID SHT.1 w,.
77 (, ,.
10 orv /4WtIWlN Prv!NWO[WLN 11
=
~-ao-7Z 1.
NN
- Ji![
.. no Itont 10-502 10*502
.n ,,~
12*
~11 I F., ....
v.cu,. o....0 AUTO START 1 ~-J ~ ~Il I~IIMEE
\ ," VE~1
... ,- I
~ I 2K75'MUH(I!~ I ~11~A I TO R5Tcld C~ FL.
~1:I:
~ Hll'f~lD50: (REF. fl6 ~ H*c. ~ -., ~
~ RM5 __ .... __
I , ------,
I ;
I ~
4_ EQUIP, EQUIP. SUMP t
~ fDO~A \2,0 ~02 \ ~ '" F
~ ~
I Br*
I ,., lONY d l
~
8 ~~ ~
C J..W.l'1
~ ~
rn- ~ ~ CODl~
~
~ ~ ~ ~ ~ --
t TO O]fW
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA SOUTHERN COMPANY COMPANY A Management Equipment Clearance and Tagging Version 11.0
{ ~iI'rW;,"U" Encro1t'lSerNY",,,,rWorl.r i!lu1't:fIIlSn-W Procedure Page 1 of 34
\
Procedure Owner: Paul D. Rushton 1 Fleet Operations Manager 1 Corporate (Print: Name I Title I Site)
Approved By: Original signed by Paul D. Rushton 1 8/2112008 8121/2008 (Procedure Owner's Signature I Date)
Effective Dates: N/A 8/22/2008 8/22/2008 8/22/2008 Corporate FNP HNP VEGP This Standardization Process Control NMP is under the oversight of the Operations Peer Team.
Writer(s): Eric Snell
(
PROCEDURE USAGE REQUIREMENTS SECTIONS Continuous Use: Procedure must be open and readily available at the work location. Follow procedure step by step unless otherwise directed by the procedure.
Reference Use: Procedure or applicable section(s) available at the work location for ready reference by person performing steps.
Information Use: Available on site for reference as needed.
ALL
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0
( ER~TDltJSer" YoMrWor/,r EMro1flSnlte YDMr Worlrr Procedure Page 2 of 34 Procedure Version Description Version Number Version Description 6.0 See applicability determination or change summary in Documentum.
7.0 Editorial Changes Revised steps 6.1.1 and 6.1.3.3 (added clarification to both steps).
Inserted new step 6.1.4 and renumbered the remaining steps. (These steps were moved from NMP-SH-003) 8.0 Editorial Change Added step 4.17 - Single Point Vulnerability definition - renumbered remaining steps 9.0 Editorial Change - Revised to add NOTE to require a training element for future revisions to this NMP if the revision potentially affects employee protection 10.0 Editorial Change to remove several duplicated step numbers. Each of the affected steps had the step number listed twice before the step language.
11.0 Editorial Change to step 6.23.4 to evaluate tag outs approaching 90 days using tagouts NMP-AD-008, Applicability Determinations, and have a 10 CFR 50.59 or other appropriate screening performed. This step didn't previously reference the NMP to be used for performing appropriate screening, but only required that a 10 CFR 50.59 screening be performed on tagouts approaching 90 days.
(
(
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 En~rol"s.:r"eYo""rWQrlJ" En~rJ(YIt) &r"l! Yo,vr WorlJ" Procedure Page 3 of 34 Table of Contents 1.0 Purpose ...................................................................................................................................... 4 2.0 Applicability ................................................................................................................................ 4 3.0 References ................................................................................................................................. 4 4.0 Definitions .................................................................................................................................. 5 5.0 Responsibilities .......................................................................................................................... 8 6.0 Procedure ................................................................................................................................ 12 6.1 General ................................................................................................................................ 13 6.2 Clearance Tagout Types ...................................................................................................... 16 6.3 Referenced Tagouts ............................................................................................................. 17 6.4 Outage Tagouts .................................................................................................................... 18 6.5 Tagout Placement Preparation ............................................................................................. 19 6.6 Tagout Placement Review .................................................................................................... 20 6.7 Tagout Placement Authorization ........................................................................................... 21 6.8 Tag Placement ..................................................................................................................... 21 6.9 Tag Placement Verification ................................................................................................... 22
{ 6.10 Tagout Holder (Sign On) ...................................................................................................... 22
\
6.11 Work Document Holder (Sign On) ........................................................................................ 23 6.12 Supplemental Worker (Sign On) ........................................................................................... 23 6.13 Supplemental Worker (Sign Off) ........................................................................................... 23 6.14 Work Document Holder (Sign Off) ........................................................................................ 23 6.15 Tagout Holder (Sign Off) ...................................................................................................... 24 6.16 Alternate Release Authorization ........................................................................................... 24 6.17 Prepare Tagout Removal ..................................................................................................... 25 6.18 Review Tagout Removal ...................................................................................................... 25 6.19 Authorizing Tagout Removal ................................................................................................ 25 6.20 Tagout Removal ................................................................................................................... 25 6.21 Tagout Removal Verification ................................................................................................. 26 6.22 Temporary Lifts .................................................................................................................... 26 6.23 Tagging Audits ..................................................................................................................... 29 6.24 Revisions .............................................................................................................................. 30 6.25 Adding Work Documents to a Tagout ................................................................................... 30 6.26 Clearance Requirements for Vendors ................................................................................... 30 7.0 Records .................................................................................................................................... 31 8.0 Commitments ........................................................................................................................... 31
(
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA.
SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 EtI~rt:1Jq&r"y.,urW{JrlJ" EMro1tl&r.. AttrWilyur Procedure Page 4 of 34 NOTE: Any revision to this NMP or its associated Guidelines, Instructions, or Forms that potentially impacts employee protection must include a re-training element for all authorized and affected employees. This re-training should normally occur prior to the effective date of the new version. Present draft revision to Training Peer Team prior to establishing a new effective date to allow time for the development of training material and completion of the required training prior to the new version becoming effective. Ref. 29CFR1910.269(d)(2)(viii)(A).
29CFR191 0.269(d)(2)(viii)(A).
1.0 Purpose This procedure establishes administrative controls for protection of personnel and plant equipment during operation, maintenance, inspection, modification and testing activities.
Further, it establishes a method for providing special instructions or cautions as necessary.
Additionally, provisions of the procedure ensure that the status of safety-related and other important equipment is verified when the equipment is removed from and restored to service.
2.0 Applicability This procedure applies to all SNC Nuclear Power Plants and shall be implemented for the equipment under the control of the individual Plant Managers.
3.0 References
( 3.1 NMP-OS-002, "Verification Policy" 3.2 NMP-SH-003, "Electrical Work Practices" 3.3 29CFR1910.269, Electric Power Generation, Transmission and Distribution Standard 3.4 NEI 96-07 3.5 NMP-AD-003-001, "Tag Standards" 3.6 NMP-AD-003-002, "Tagout Standards" 3.7 NMP-AD-003-003, "Tagout "Tag out Restorations" 3.8 NMP-AD-003-004, "General Techniques for Venting and Draining" 3.9 NMP-AD-003-005, "Tags/Maintenance Lock use with Operation Permit Tags" 3.10 NMP-AD-003-006, "PDT use with a PDT Documentation Sheet" 3.11 NMP-AD-003-007, "Farley Nuclear Plant Special Considerations" 3.12 NMP-AD-003-008, "Hatch Nuclear Plant Special Considerations" 3.13 NMP-AD-003-009, "Vogtle Nuclear Plant Special Considerations"
(
3.14 NMP-AD-003-010, "Reptask Template for Tagouts" 3.15 NMP-AD-003-GL01, "Example Tagout Tags"
Southern Nuclear O~erating Operating Company Nuclear NMP-AD-003 SOUIHERNA SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 e"~'X7lo&rwYo"'rW"rf,r EtuTD'ltlSnw Htwr\Vdr/,r Procedure Page 5 of 34 3.16 NMP-AD-003-F01, "Tag out Cover Sheet" 3.17 NMP-AD-003-F02, "Tag out Tag Listing" 3.18 NMP-AD-003-F03, "Temp Lift Sheet" 3.19 NMP-AD-003-F04, "PDT Documentation Sheet" 3.20 NMP-AD-003-F05, "Work Doc / Tagout Holder List" 3.21 NMP-AD-003-F06, "Clearance Tagout Log Sheet" 3.22 NMP-AD-003-F07, "Supplemental Worker List" 3.23 NMP-AD-003-F08, "Tag out Preparation and Approval Checklist" 3.24 NMP-AD-003-F09, "Tag out Temporary Lift or Revision Checklist" 3.25 NMP-AD-003-F10, "Clearance Tagout Audit Form" 4.0 Definitions 4.1 Alternate Boundary - A boundary equivalent to that formed by tagging the Main Steam
( Isolation Valves/Plugs closed. This "alternate" boundary is established administratively by maintaining the reactor <200 'F at Vogtle or Farley and <150'F at Hatch, and by also maintaining reactor vessel level below the Main Steam Lines at Plant Hatch.
4.2 Alternate Release - A release authorized by the Shift Manager in the event that Tagout release is required and a Work Document Holder cannot be contacted and is not on site.
4.3 Authorizer - An individual qualified to initiate implementation of Tagouts on an operating plant.
4.4 Authorized Employee - An individual designated as the "lead" person responsible for the overall direction of a crew performing maintenance when using a PDT. Usually this will be workers who hang their Personal Danger Tags to perform work.
4.5 Boundary Point - Those components such as isolation valves and/or electrical isolation devices that are required to be tagged to provide plant and personnel safety during work activities or procedural performance.
4.6 Caution Tag - A tag used to provide special operating instructions for equipment. It may also be used to provide general information on equipment condition. Caution Tags are similar to that shown on Figure 2 of NMP-AD-003-GL01, and are yellow with black lettering.
4.7 Clearance - A folder containing Clearance types addressed in Section 6.2 of this procedure. The Clearance number will be a 3-part alpha numeric designator using the unit number, Tagout type, and the year issued. Example: 1-DT-04 a unit one Danger Tag issued in 2004. The unit 1 designator will be used for unit one and common equipment.
4.8 Danger Tag - A tag which, when attached to a component, prohibits the operation of that component in all circumstances except as specified in NMP-AD-003-001 section 6.2.1.1.
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUTHERN SOUIHERNA COMPANY A Management Equipment Clearance and Tagging Version 11.0
~rt:fI.SNnHtNrw"rhr En~rx.rI()Svn~NrWorlJ* Procedure Page 6 of 34 Danger tags are similar to that shown on Figure 1 of NMP-AD-003-GL01, and are red with black lettering.
4.9 INFO ONLY Step - A Tagout step in which no Configuration position is given. This type step may be used to provide additional information or initiate a Tagout control (such as MSIV alternate boundaries).
4.10 Maintenance Lock - An individually keyed lock used to isolate electric power to a component being tested or repaired in conjunction with an Operating Permit.
4.11 Non-Power Plant Equipment - Any equipment that has no affect on the operation of the plant, such as equipment located in out buildings. Personnel should consult with the Shift Supervisor if in doubt when making this determination. Equipment not meeting this definition should be considered "Power Plant Equipment".
4.12 Operating Permit Tag (Op Tag) - A tag which, when attached to a component, identifies the component as being released to an Operations Permit Tagout Holder for component position alignment, testing or maintenance. Op Tags are similar to that shown on Figure 3 of NMP-AD-003-GL01, and are blue with white lettering.
4.13 Personal Danger Tag (PDT) - A tag, hung by an approved PDT user, and used in conjunction with either an Operating Permit Tagout or with a PDT Documentation Sheet (NMP-AD-003-F04). Personal Danger tags are similar to that shown on Figure 5 of NMP-( AD-003-GL01, and are red with black lettering.
4.14 Power Plant Equipment - Any equipment that can have an effect on the operation of the plant as designated per Site procedures. Personnel should consult with the Shift Supervisor if in doubt when making this determination.
4.15 Preparer - An individual qualified to prepare Tagouts.
4.16 Reviewer - An individual qualified to review Tagouts.
4.17 Single Point Vulnerability (SPV) - IIndividual ndividual components and equipment which, upon failure, will result in a reactor or turbine trip. The effects of performing a Tagout involving an SPV must be considered in all phases of the tagging process.
4.18 Status Point - A tag or other means used to communicate the impact of a Tagout on system or component functionality (Vogtle AI 2006205179).
4.19 Supplemental Worker - Any individual who requires the protection of a Tagout when performing their work but does not have ability to sign on the electronic tagging system as a Work Document Holder. All supplemental workers must be signed onto the Supplemental Worker List while performing work under the protection of a Tagout. The individuals signing on/off the Tagout will do so by entry of their name and date/time on the Supplemental Worker List, NMP-AD-003-F07.
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 En~'XY J,,&rn YONrWtJ"I;I"
&~'X1If1&nt ¥ollf WorM~ Procedure Page 7 of 34 PaQe 70f34 4.20 Supplemental Worker Supervisor - An individual or a "Plant Position" identified as a Tagout Holder. Typically, this will be the job supervisor. This individual shall sign on to a Tagout electronically as a Work Document Holder and shall maintain the Supplemental Worker List for the associated work document. This individual shall be responsible for briefing the Supplemental Workers and keeping them informed of Tagout boundaries and protections provide by the Tagout. This individual may not sign off the Tagout as Work Doc Holder until all supplemental workers have signed off the Supplemental Worker List.
4.21 Tagger - A qualified person, who repositions equipment per a Tagout and hangs, removes or verifies tags.
4.22 Tagging Official Official-- The Tagging Desk Operator (TDO), Unit Shift Supervisor C& C&TT (USS C& T), Shift Support Supervisor (SSS), Shift Technical Advisor (STA) or Shift Supervisor C&T),
that may perform all Clearance and Tagging functions outlined in this procedure. The Tagging Official can serve as designee for the Shift Supervisor, provided that person is cognizant of the plant's status/configuration and the Shift Supervisor is made aware of all resultant changes to the plant configuration.
4.23 Tagout - A tool used to uniquely identify and authorize a collection of data to remove equipment from service, track component changes, track activities associated with the return the equipment to service. A Tagout is a unique document that is used entity and retum once and only once and is then stored as a completed document.
( 4.24 Tagout Holder - An individual or a "Plant Position" identified on a Tagout Holder List. This individual is usually the "lead" person responsible for the overall direction of a crew performing maintenance or an individual requiring administrative hold on a Tagout. It can also be the "Authorized Employee" or "lead" person responsible for performing a test or maintenance under an Operating Permit Tagout.
4.25 Tagout Lockout - A Tool used on Tagouts to Preclude Sign-on of Tagout Holders and Work Document Holders.
4.26 Tagout Number - A unique control number assigned to a Tagout document. The Tagout Number will be a 6-part alpha numeric designator using the unit number, Clearance Tagout Type, year or outage identifier, system number, consecutive or assigned number and the revision number. Example: 1-DT-04-1208(E11 )-00014(002) would be the second revision of the fourteenth Danger Tagout document issued in 2004. The unit 1 designator will be used for unit one and common equipment. Computer generated Tagouts will be assigned the next sequential number available. Manual or hand written Tagouts will be numbered sequentially starting with the number 30,000 proceeded by the unit, Tagout type, system number and year. An index will be maintained to prevent duplication of manually generated Tagout numbers and can be discarded at the end of the calendar year.
4.27 Tagout Point - Any device, valve, breaker, switch, etc. that is positioned by a step on a Tagout.
Tag out.
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 E1jerx.r/~&rw Y()""1*Wor/rr E1IUXY1oS",wYoMrWor/tI" Procedure Page 8 of 34 4.28 Tag Numbers - The number placed on each Tag. The Tag Number will be a 4-part alpha numeric designator using the unit number, Clearance Tagout Type, year issued, and the consecutive number. Example: 1-DT-04-00014 would be the fourteenth Danger Tag issued in 2004. Outage Tags will be identified by the Outage number replacing the year issued field.
Example 1-DT-R14-00014 the R14 would identify the Outage code. Manual or hand written tags will be numbered sequentially starting with the number 30,000 proceeded by the unit, Clearance Tagout type, and year. An index will be maintained to prevent duplication of manually generated tag numbers and can be discarded at the end of the calendar year.
4.29 Tags - See NMP-AD-003-001, "Tag Standards" for Tag definitions.
4.30 Temporary Lift (Temp Lift) - The act of releasing one or more Tagout points with the possibility or intent to reinstall the Tagout points at a later time. This is performed after the component or subsystem has been placed in a configuration that assures personnel safety and the safe operation of plant equipment. Temporary Lift tags are similar to that shown on Figure 4 of NMP-AD-003-GL01.
4.31 Verification - The "second check" of a component's condition/position. Performed by an individual other than the one who performed the initial check/positioning. Requirements for verification can be found in procedure NMP-OS-002, "Verification Policy".
4.31.1 NV - No Verification used for position of steps such as NO TAG steps that will only be used for configuration control or as specified by Operations Management
( 4.31.2 SC - Single Check is normally used for placing plaCing and restoring components such as NON Safety related items that do not meet the Purpose or Applicability of Verification Policy per NMP-OS-002 4.31.3 CV - Concurrent Verification to be used per NMP-OS-002, "Verification Policy" 4.31.4 IV - Independent Verification to be used per NMP-OS-002, "Verification Policy" 4.32 Work Document Holder - An individual, who Signs-On a Tagout to identify that they are working under the protection of the Tagout. The individuals who Sign-on to the Tagout will do so by entry of the work document, user name, and date/time on to the Work Document/Holder List. This must be done by .ill! workers who are afforded protection from by.ill!
energy sources or other hazards by the Tagout. Workers who are Work Document Holder qualified in the electronic system shall use the electronic system. Others who do not have electronic system access are called Supplemental Workers per this procedure and should sign on using the Supplemental Worker List, NMP-AD-003-F07.
5.0 Responsibilities 5.1 Plant Manager Nuclear Plant Ensuring that plant personnel are informed of their individual responsibilities regarding the Equipment Clearance and Tagging Procedure 5.2 Department Managers (all Departments)
- Ensures all department personnel (including contractors) working for their department are trained on and comply with this procedure
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 IJn~rt.rltJ&rfleYo-.Io'rWorlj{"
En~rtr UI &rf'f: Your Wort,/" Procedure Page 9 of 34
- Defining Departmental Titles "by Plant Position" that may be Tagout Holders, and a list of personnel in the department which can sign on as Tagout Holder by those Positions
- Will designate which individuals or groups can perform the tagging functions specified in this procedure
- Approve individuals that may use Personal Danger Tags 5.3 Operations Manager - In addition to the Department Manager responsibilities listed above, the Operations Manager is responsible for:
- Implementation of this procedure
- Ensuring clearance audit requirements specified in this procedure are met 5.4 Training Manager - In addition to the Department Manager responsibilities listed above, the Training Manager is responsible for:
- Providing Equipment Clearance and Tagging training when requested by Department Managers
- Providing Department Managers with a list of their personnel who successfully complete Equipment Clearance and Tagging training 5.5 Shift Manager
- Authorizes release of a Tagout, for a Work Document Holder when the Holder is not
( on site and cannot be contacted to gain their approval per the Alternate Release provisions of section 6.16 of this procedure.
- Provides approval when using a check valve as a boundary point
- Making decisions in situations not clearly covered by this procedure 5.6 Shift Supervisor
- Evaluates the impact of a Tagout on plant operations and configuration control. The evaluation should ensure unaffected systems or components are not impacted
- Ensuring Technical Specifications, ODCM, TRM and Fire Protection Program requirements are met when implementing Tagouts
- Authorization of all Tagouts
- Authorizing Temp Lifts
- Ensuring Tagout audits and reviews are performed as specified in this procedure The Tagging Desk Operator (TOO),
(TDO), Unit Shift Supervisor C&T (USS C&T), Shift Support Supervisor (SSS) or Shift Technical Advisor (STA) may perform all Clearance and Tagging functions outlined above, provided that person is cognizant of the plant's status/configuration and the Shift Supervisor is made aware of all resultant changes to the plant configuration.
5.7 Department Supervisor
- Provide approval for use of a check valve as a boundary point
- Contacting a Work Document Holder when release of a Tagout is required and the Holder is not on site but can be contacted
- Instruct Work Doc Holder to contact Tagging Official to release Tagout per Section 6.1.6
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 elle~
/:.'u~1'f:IIe&rw tf1&r~ Your Y"'Yr Wtu,ltr
'W6r/,i' Procedure Page 10 of 34 5.8 All Personnel (plant personnel and contractors)
- Observing equipment for the presence of Tagout tags
- Complying with the warnings/instructions on Tagout tags
- Reporting of misplaced, loose or misSing missing tags to the Tagging Official - Personnel shall not attempt to reattach misplaced or loose tags without proper authorization 5.9 Tagout Holder
- Overall responsibility for the workers and the work activities under their control and performed under the protection of a Tagout (Le. the Job Supervisor or Job Lead).
- Review the Tagout for applicable precautions and/or limitations.
- Verify Tagout points required for work are adequate for the work activities under their control before allowing their personnel to work.
- Verifies Tagout boundaries remain adequate for the work activities under their control for the duration of holding the Tagout.
- Ensuring that they sign on as Tagout Holder before allowing their crew to commence work.
- Conducts pre job briefing for crew members under their direction of the scope and limits of the Tagout when directing a crew to perform work under that Tagout
- Verifies workers are clear and the equipment is ready to admit energy prior to signing off as the Tagout Holder
- If holding for Administrative purposes, ensure boundaries remain adequate to support
( the Administrative need.
- When employing workers who are utilizing the Supplemental Workers List, sign on the Tagout as a Work Doc Holder/ Supplemental Worker Supervisor.
- Notify Tagging Official when work is complete and when signing off as a Tagout Holder.
- A Tagout Holder may sign off another Tagout Holder or accept a Temp Lift by verbal authorization on a case by case basis. When doing so, they shall be cognizant of the work in progress and assume all tagging responsibilities of that Tagout Holder.
5.10 Work Document Holder NOTE: Work Document Holder Sign On and Sign Off should normally be performed by the individual Holder:.
Holder.:.
- Review or obtain an understanding of the Tagout to satisfy for themselves that it is safe to perform work, before commencing work.
- Review the Tagout for applicable precautions and/or limitations
- Sign on the Tagout for all applicable Work Documents prior start of work.
- Sign on Tagout if personnel protection or equipment protection is required for work performance.
- The Work Document Holders shall accept the Temporary Lift for Authorization of the Temporary Lift, if the work scope and status of work will allow it, Work Doc holders may be added to a tagout while a temp lift is in place provided that the temp lift does not effect the work and that the work doc holder as well as supplemental workers are
( made aware of the temp lift.
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 YowrWaI'Itl' EIIf:TDJo$n-wIDNrWoy!;l' EIUt1XYZI1&1"w Procedure Page 11 of 34
- If the Work Document Holder can not accept the Temporary Lift due to the status of work or plant conditions, then they shall notify the Tagout Holder of the work status/condition and the Temporary Lift will not be Authorized until status/conditions will allow.
- Sign off as Work Document Holder when protection under the Tagout is no longer required.
- When work extends past the end of shift, the Work Document Holder should sign off the Tagout prior to leaving site (preferred).
- If Work Document Holders remain signed on the Tagout they should ensure that they can be contacted off-site, in the event the Tagout may need to be Released or Modified.
- Work Document Holders who are Supplemental Worker Supervisors will also sign off the Tagout only after all workers under their control have completed their work and signed off the Supplemental Worker List and it is safe to return the component to service.
5.11 Supplemental Worker Supervisor
- Maintains overall responsibility for the workers and the work activities under their control performed under the protection of a Tagout.
- Sign on as Work Document Holder for work being performed.
- Brief Workers on the Tagout work scope and applicable precautions and/or limitations
- Brief Workers on the Tagout points and ensure they are adequate for the work before performing work each day.
- Facilitates Sign on the Supplemental Worker List for all applicable Work Documents prior start of work.
- Brief Workers on the changes to Tagout points and ensure Workers agree to changes before accepting changes to the Tagout.
Tagout.
- If the performance of a Temporary Lift may affect worker safety, the responsible Work Supervisor, shall ensure that all workers under their control sign off the Supplemental Worker List and stop work.
- When work extends past the end of shift, facilitate the worker sign off of the Supplemental Worker List prior to leaving site (preferred).
- If Workers remain signed on the Supplemental Worker List coordinate the contact of off-site workers, in the event the Tagout may need to be Released or Modified.
- Facilitates Workers Sign off of Supplemental Worker List when protection under the Tagout is no longer required.
- Sign off as Work Document Holder after work is complete and Workers have signed off Supplemental Worker List.
- Maintain NMP-AD-003-F07 "Supplemental Worker List" and ensure it is transmitted with the applicable Work Document for document retention.
5.12 Supplemental Worker NOTE: Supplemental Worker Sign on and Sign off should normally be performed by the individual Worker.
- Understand and have been briefed on the Tagout for applicable precautions and/or limitations
- Understand and have been briefed on the Tagout points and agree they are adequate for the work before performing work each day.
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA.
SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0
( E1I~r!.71t)&rI'~ Yo-ur Worlrl" E"ut."IlfJSnnY~lIrWorl,r Procedure Page 12 of 34
- Sign on Supplemental Worker List to all applicable Work Documents prior start of work.
- Understand and have been briefed on the changes to Tagout points and agree to changes under a Temporary Lift before accepting changes to the Tagout.
- If the performance of a Temporary Lift may affects worker safety, Workers utilizing the Supplemental Work List shall sign off the Supplemental Worker List and stop work.
- Sign off Supplemental Worker List when protection under the Tagout is no longer required.
- When work extends past the end of shift, the Workers should sign off the Supplemental Worker List prior to leaving site (preferred).
- If Workers remain signed on the Supplemental Worker List they should ensure that they can be contacted off-site, in the event the Tagout may need to be Released or Modified.
6.0 Procedure NOTES: NMP-AD-003-007, 008, & 009 contains special considerations for each site that lists additional requirements unique to that site. The appropriate INSTRUCTION should be reviewed when determining requirements for any given situation.
Guidelines, Instructions, and Forms developed to support this procedure (NMP-AD-003) will be reviewed and approved by each plant site Operations Manager (or designee) and by the Fleet
( Operations Manager (or designee) unless noted below.
- Instruction NMP-AD-003-001 outlines Tag Standards
- Instruction NMP-AD-003-002 outlines Tagout Standards
- Instruction NMP-AD-003-003 outlines Tagout Restorations
- Instruction NMP-AD-003-004 outlines General Techniques for Venting and Draining
- Instruction NMP-AD-003-005 outlines Tags/Maintenance Lock use with Operation Permit Tags
- Instruction NMP-AD-003-006 outlines PDT use with a PDT Documentation Sheet
- Instruction NMP-AD-003-007 outlines Farley Nuclear Plant Special Considerations (Review and approval by Farley Operations Manager or designee only)
- Instruction NMP-AD-003-008 outlines Hatch Nuclear Plant Special Considerations (Review and approval by Hatch Operations Manager or designee only)
- Instruction NMP-AD-003-009 outlines Vogtle Nuclear Plant Special Considerations (Review and approval by Vogtle Operations Manager or designee only)
- Instruction NMP-AD-003-010 outlines Reptask Templates for Tagouts
- Guideline NMP-AD-003-GL01 outlines Example Tagout Tags
- Form NMP-AD-003-F01 - Tagout Cover Sheet
- Form NMP-AD-003-F02 - Tagout Tag Listing
- Form NMP-AD-003-F03 - Temp Lift Sheet
- Form NMP-AD-003-F04 - PDT Documentation Sheet
- Form NMP-AD-003-F05 - Work Doc / Tagout Holder List
(
O~erating Company Southern Nuclear Operating Nuclear NMP-AD-003 sou'HERNA SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0
( E1I~rg I,,&TN RJ.",.
Eanol"Snw ¥"Nr Wo,t ..
Wo"JJ' Procedure Page 13 of 34
- Form NMP-AD-003-F06 - Clearance Tagout Log Sheet
- Form NMP-AD-003-F07 - Supplemental Worker List
- Form NMP-AD-003-F08 - Tagout Preparation and Approval Checklist
- Form NMP-AD-003-F09 - Tagout Temporary Lift or Revision Checklist
- Form NMP-AD-003-F10 - Clearance Tagout Audit Form 6.1 General 6.1.1 Any employee that may be exposed to hazardous energy of any type shall be protected by Danger tags and signed on the Tagout. An Operating Permit Tag with a PDT or Maintenance Lock or PDT tags per NMP-AD-003-006 may be used to provide individual personnel protection for certain activities.
6.1.2 Any questions or doubt in connection with use of Tagout tags or associated safe working conditions should be referred to the Shift Supervisor or Shift Manager until the matter is resolved.
6.1.3 Work Activities that do not require a Tagout 6.1.3.1 Work that presents NO hazard to personnel performing the work.
6.1.3.2 Examples of work activities involving no hazard would be tasks such as:
(
- Work on plug-in devices, such as recorders
- Installation of jumpers to perform testing When these activities are performed without a Tagout, the equipment must be removed and returned to service in accordance with approved procedures.
6.1.3.3 Work performed where controlled plant conditions are maintained with minimal risk to workers or equipment. These are typically outage type activities where Operations is controlling the plant conditions. Examples of this are, but are not limited to, activities such as removal of pressurizer code safety valves or steam generator primary manways, and work performed inside the reactor cavity while the reactor is shutdown.
6.1.4 All circuits and equipment operating at greater than (>> 50 volts to ground should be DEENERGIZED and tagged prior to beginning work unless:
6.1.4.1 Additional or increased hazards are introduced by DEENERGIZING.
6.1.4.2 DEENERGIZING is infeasible due to equipment design, operational limitations, or troubleshooting activities.
6.1.4.3 The work is on equipment operating at voltages (>> 50 VAC where the current is
(~) 1 milliamp].
[low (::;;)
6.1.4.4 The work is on 4-20 maDC or 10-50 maDC instrument transmitter loops.
(
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA SOUfHERNA. Management Equipment Clearance and Tagging Version 11.0 COMPANY EJI~rol"&r"eYo,..rWor/J" E1U!'X7I.Snn YHrWorltri" Procedure Page 14 of 34 6.1.5 Maintenance Issues 6.1.5.1 Personnel performing work on mechanical systems should initially assume the system may not be drained when breaching the pressure boundary. This can happen due to problems with drain lines clogging or drain valves stuck closed etc.
Extreme care should be utilized when first breaching the system pressure boundary. Techniques such as slowly loosening studs, partial loosening of fittings, etc. should be utilized to minimize the potential for spillage or pressure release in the event the system has not been adequately depressurized and drained.
6.1.5.2 Work will not be performed on a valve that has a Danger tag on its hand wheel other than repacking the valve on the backseat or performing inspections which will not alter the configuration of the valve. These are limited to the work activities listed below. Other work activities can be considered and approved by the Shift Manager.
- Repacking the valve on the backseat
- Lubes / MOV Grease Inspections
- Limit Switch work/wiring inspections 6.1.5.3 Components will not be removed from the system (Le. physically cut out of the system, removed from breaker cubicles, etc) with Danger Tagout tags still attached. If necessary, the Tagging Official should be contacted and the tag removed in accordance with the requirements of this procedure. Temporary Lift
( Tags may be used to allow for removal from system. Electrical breaker tags may be transferred to the breaker cubicle door.
6.1.5.4 Personnel performing work on electrical systems should initially assume the circuit is energized. The worker will verify the circuit is de-energized using an appropriate circuit potential testing device before commencing work.
6.1.5.5 Performance of work activities on tagged breakers/cubicles 6.1.5.5.1 The intent of tagging electrical breakers is to electrically isolate the load side to establish a safe boundary for load side work.
6.1.5.5.2 Breaker removal may be performed on tagged breaker, as long as the tag has not been placed on the operating breaker mechanism.
6.1.5.5.3 Prior to re-installation of a tagged breaker, ensure breaker being installed is verified open.
6.1.5.5.4 Electrical maintenance or other work activities which cannot energize the load side are permitted inside this boundary, the only exception allowed which may allow energization is for motor electrical testing which is addressed below.
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 En~'XrJII&rJ'eY""lfrWorl.r E#<<D/(;SctrIttY.,.rw.,.ur Procedure Page 15 of 34 6.1.5.5.5 Motor electrical testing (Megger) can be performed on a tagged component only when the group/individuals performing the testing are the only Holders on the Tagout and other Holders are locked out from Sign On. Otherwise, a Temporary Lift will have to be performed or, the testing will have to be performed after the Tagout has been restored.
6.1.5.5.6 See individual Site Appendices for further information when performing work on breakers.
6.1.5.6 A Maintenance Lock may be used to isolate power to components during performance of maintenance or testing when periodic de-energization is required provided:
- The power supply has been tagged by a Danger Tagout and a Temporary Lift has been issued or,
- The power supply has been tagged by a Personal Danger Tag or,
- An Operating Permit Tagout has been installed.
All persons afforded protection for the work must affix their own lock or PDT. The Maintenance Department is responsible for control and installation of Maintenance Locks. The use of Maintenance Locks is determined by the Tagout Holder.
( 6.1.6 Computer/Software Related Issues 6.1.6.1 Computer software may be used to generate and administer Tagouts. If required however, Tagouts may be written manually on forms similar to the computer generated forms, provided the general process requirements (Preparation, Review, Authorization, Placement, Verification, Signing on as Tagout Holder, Work Document Holder, etc.) remain the same.
6.1.6.2 Tagout Lock Out tool may be used to prevent Tagout Holder and Work Document Holder Sign On to a Tagout. Tagout Lock Outs are preformed by a Tagging Official from the Tagout Detail Tab of the Tagout. The Tagging Official clicks on the Preclude Sign-On button to lock it out and Allow Sign-On button to unlock it.
6.1.6.3 The forms described in this procedure are typical. A current copy of the blank forms used to support the manual tagging process shall be maintained by Operations.
6.1.6.4 The computer file will normally be the "official version" or QA record copy of each Tagout. In the event a manual Tagout is generated, the hardcopy will be the QA record copy.
(
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 En~rol,,&r,,~Y",..r\W)rl,r EIU'XY1O$e1'I'tYoNrw"rM' Procedure Page 16 of 34 6.1.6.5 Users performing Verifications (Le. Preparers, Reviewers, Authorizers, and Taggers) shall document their actions associated with a Tagout. This should be done by entry of their User 10 and Password. When this documentation can not be performed by the performer of the action another individual(documenter) individual( documenter) may document the action provided:
- The documenter enters their own User 10 and Password, and,
- An entry is made in the Tagout Change Log to that effect (for example, "Step 5, Placement Verification, signed off by AI Covington with permission from Hans Bishop per telecon, Time & & Date").
6.1.6.6 Work Doc Holders should release Tagouts and accept Temp Lifts in eSOMS WorkDoc using their own User 10 and Password, however, when computer access is NOT available (user is not on site and has no RAS capability, worker is dressed out in the drywell waiting for a Temp Lift but no computer is available in the drywell, etc.), the worker may release clearance or accept a Temp Lift over the phone to a Tagging Official as follows:
- The Tagging Official will serve as the Documenter and must communicate directly with the WorkDoc Holder.
- The WorkDoc Holder must agree to give his/her permission to release the Tagout or accept the Temp Lift.
- The Documenter will Sign Off (using their own User 10 and Password) for
( the release of the Tagout or acceptance of the Temp Lift for the WorkDoc
\ Holder.
- An entry should be made in the Tagout Change Log to that effect:[Temp Lift 2006-0101 accepted by T. Jones (Work Doc Holder working in the drywell),
per telecon, No local computer available.]
6.2 Clearance Tagout Types 6.2.1 Danger Tagouts 6.2.1.1 Danger Tagouts shall be used whenever a tag is to be placed to provide personnel protection.
6.2.1.2 Danger Tags will be hung using the electronic tagging program, POTs or manually if the electronic system is unavailable.
6.2.1.3 Danger Tagouts are the only Tagouts used to provide protection for personnel.
6.2.1.4 Two tag types are allowed under Danger Tagouts.
- Danger tags when required for personnel protection
- Caution tags when required to provide additional information 6.2.2 Caution Tagouts 6.2.2.1 A Caution Tagout is used to provide special operating instructions for equipment
( that may be out of its normal alignment or operating mode. It may also be used to provide general information on equipment condition or configuration.
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA.
SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0
( Euul:/l<1&rfJe Y(JllrWtJ,lJ" 8url.:r/(JSUP(! X>ttrWW,IJ" Procedure Page 17 of 34 6.2.2.2 The only tag types allowed are Caution tags. Caution Tagouts may not have Danger tags listed on them.
6.2.2.3 A required component configuration may be specified. If no configuration is required at Placement, the configuration may be assigned as "N/A".
6.2.2.4 If a Caution Tag is used to reposition a safety related component then that Caution Tag shall be issued using a Danger Tagout. This ensures that Tagout reviews and hanging/releasing verifications take place for safety related components. Otherwise, Caution Tagouts do not require Review.
6.2.2.5 The Tagging Official will determine if verification is required.
6.2.3 Operating Permit Tagouts 6.2.3.1 Operating Permit Tagouts may be used to track components turned over to a Tagout Holder to allow for testing, minor maintenance or configuration control.
6.2.3.2 Only a single Tagout Holder is allowed to hold an Operating Permit Tagout.
6.2.3.3 The Operating Permit Tagout approval process requires at a minimum Placement Authorization and Removal Authorization from the Shift Supervisor.
(, 6.2.3.4 Operating Permit tags shall not be placed on Danger tags.
6.2.3.5 If plant conditions require "Boundary Point" isolation of the equipment to perform testing or minor maintenance a Danger Tagout shall be created. Both Tagouts must cross reference the other Tagout when this is performed.
6.2.3.6 Operating Permit Tagouts will annotate the reason for the Tagout. Each Operating Permit tag will state the reason for the Tagout. (Le. MOV, LLRT, Config Cont. etc) 6.2.4 Personal Danger Tag (PDT)/Maintenance Lock 6.2.4.1 PDT or Maintenance Lock programs will be administered as described in NMP-AD-003-005 & NMP-AD-003-006. This program will not be performed electronically.
6.3 Referenced Tagouts 6.3.1 If when preparing a Tagout, it is found that the work scope requires the use of more than one Tagout to fully isolate a work area, an additional Tagout may be created to work in conjunction with the first to cover the specific work activity. Each Tagout will contain a cross-reference to the other. The work order should also have a reference to all Tagouts necessary for performance of the applicable work scope.
6.3.2 The Tagging Official reviewing the Tagout for completeness shall ensure all Tagouts are cross-referenced on all associated Tagouts. The Tagout Holder is responsible for signing on all Tagouts necessary to ensure his work is protected.
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA Management Equipment Clearance and Tagging Version 11.0 COMPANY EtUrt;y EtU'O tllM" YO/f'r World'"
to&rfJt Your W(}fl;l" Procedure Page PaQe 18 of 34 6.4 Outage Tagouts 6.4.1 Administrative Tagout 6.4.1.1 During outages, an Administrative Tagout may be used as necessary to control certain physical boundaries or process parameters. The Tagout does not require physical tags; instead it requires Operations to control the boundaries or parameters. An Administrative Tagout will provide worker assurance that fluid or gas will not enter their work area 6.4.1.2 This type of Tagout must be clearly identified as an "Administrative Tagout" on the Tagout. Administrative Tagouts may only be used with Operations and Maintenance Management concurrence.
6.4.1.3 Blocking Tagouts or the use of MSIV Alternate Boundaries are methods of control that may be used. These controls will be identified by the use of "INFO ONLY" step. The control used will be clearly identified in the Placement Notes field.
6.4.1.4 Reference Tagouts listed on Blocking Tagout will be referenced on the Administrative Tagout as well.
6.4.1.5 TagouUWork Tagout/Work Document Holders may sign on to the Administrative Tagout.
( 6.4.2 Blocking Tagouts 6.4.2.1 A Blocking Tagout provides for a physical boundary to be in place, when working with an Administrative Tagout. The Blocking Tagout will require tags, but workers will not be allowed to sign on as Holders of the Blocking T Tagout.
agout. The Outage Shift Manager position (or higher) will be the only Tagout Holder for a Blocking Tagout.
6.4.2.2 Blocking T Tagouts agouts may be used to control certain physical boundaries or process (i.e. MSIV Alternate Boundaries). The isolation of the boundary may parameters (Le.
require installation of tags on a physical boundary or it may require Operations to control the conditions or parameters of the plant.
6.4.2.3 Blocking Tagouts may use Referenced Tagouts to completely isolate the work area. The Referenced Tagout must be listed on the both the Blocking and the Administrative Tagouts.
6.4.3 MSIV Alternate Boundaries 6.4.3.1 It is permissible to establish an Alternate Boundary in lieu of the MSIVs under the following conditions.
- At Plants Farley and Vogtle Main Steam temperatures are verified to be less than 200 'F
- At Plant Hatch, Main Steam temperatures are verified to be less than 150
( 'F
- At Plant Hatch, Reactor Vessel level is verified to be below the MSLs
O~erating Company Southern Nuclear Operating Nuclear NMP-AD-003 SOUTHERN'\
SOUTHERN'\' Management Equipment Clearance and Tagging Version 11.0 COMPANY II1&rW Your EturtJlt:lSnw Eturr/ Yl1/1r Wsd,r WSf'It/' Procedure Page 19 of 34
- No pressure such as Nitrogen over pressure exists in the main steam system
- The main steam header has been drained as necessary to ensure no adverse water flow will occur when the MSIV's are opened 6.4.3.2 With Shift Supervisor concurrence and satisfactory completion of the above prerequisites, tags may be placed on appropriate components which will prevent heat input into the main steam system from the reactor. Consideration should be given to the following when selecting appropriate components to tag.
- Nitrogen supply to the main steam lines.
- Auxiliary steam heat from opposite Unit.
- Pressurizer heater input to the RCS.
6.4.3.3 Tagging of components associated with RCS Pump and Pressurizer Heater heat input to the RCS is not required if the following conditions are met:
- Administrative controls are verified in place to ensure the Reactor Vessel level is maintained below the MSLs at Plant Hatch.
- Administrative controls are verified in place to ensure the RCS is 0
maintained less than 200 'F (FarleyNogtle) or 150 0 F (Hatch).
- Administrative controls are in place to ensure MSIVs are re-tagged prior to
( RCS Temperature exceeding 200 'F (FarleyNogtle) or 150'F (Hatch), if a MSIV boundary is required.
6.5 Tagout Placement Preparation NOTE: Action shall be initiated to correct any database error found such as equipment name or location, etc. during preparation, review, approval and hanging of a Tagout by the individual discovering the error 6.5.1 A Preparer will use the guidance provided in NMP-AD-003-F08, "Tag out Preparation and Approval Checklist" (or equivalent) and prepare the Tagout as follows:
6.5.1.1 Review the scope of work to ensure a thorough understanding of the protection needed for personnel and equipment.
6.5.1.2 Using approved documents, determine hazardous energy sources and isolations necessary to provide a safe work boundary for each work activity.
6.5.1.3 Identify and assess all potential hazards, automatic actions and/or effects on the plant which may result due to execution of the Tagout. These hazards should be identified in the Tagout Instructions when possible. Guidance should be provided as appropriate to inform and/or prevent any unwanted occurrences.
6.5.1.4 If approved documents do not exist for equipment to be tagged, then a physical walk down may be performed.
6.5.1.5 Enter the required information on the Tagout.
O~erating Company_
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0
( ~'XY I~&rl'e Y~Mr Worltl' ERU"-'ID&rwYo.wrWorltl" Procedure Page 20 of 34 6.5.1.6 Components that are removed from the system or having maintenance performed on them should be identified with a "No-tag" tag type to ensure that it is positioned properly upon placement and removal. Do not hang any other type of tag on that component.
6.5.1.7 For components that are positioned but not tagged, mark the tag as a "No-tag" tag type.
6.5.1.8 For Tagout points that are to be used for Information only an "INFO Step" may be used.
6.5.1.9 Utilizing the Tagout points, prepare the Tagout:
- Indicate type of tag
- Indicate component number and name
- List the required position and sequence
- Indicate verification requirements
- If special instructions are applicable, annotate in appropriate section 6.5.1.10 Sign the Tagout as Preparer.
6.5.1.11 Sign the Work Documents listed on the Tagout as 151 51 Verified.
( 6.6 Tagout Placement Review 6.6.1 A Reviewer will perform an independent review of the Tagout.
6.6.2 Verify the Tagout points and boundary isolations selected provide adequate plant and personnel safety for work activities listed.
6.6.3 Verify and assess all potential automatic actions and/or effects on the plant which may result due to execution of the Tagout. Ensure these items are identified and are properly documented on the Tagout when possible.
6.6.4 Review the Tagout points for the correct:
- Tag type used
- Component number and name
- Position and sequence PosWonandsequence
- Placement verifications required 6.6.5 Review impact on equipment (including Tech Specs).
6.6.6 Review or add any special instructions that apply to the Tagout.
6.6.7 Sign the Tagout as Reviewer.
6.6.8 Sign the Work Documents listed on the Tagout as 2nd nd Verified.
(
Southern Nuclear Operating Company Nuclear NMP-AD-003 SOUIHERNA SOUTHERN COMPANY A Management Equipment Clearance and Tagging Version 11.0 J:.'n~1'D I()&r.,~ YO-liT Wot/tf"
&~'XT'tlSnftHtM,.UT.r/'" Procedure Page 21 of 34 6.7 Tagout Placement Authorization 6.7.1 Review the Tagout to ensure:
- Tagging this system or component does not have an unacceptable impact on current plant operation.
- Tech Spec and other administrative commitments are satisfied.
- Tagout hang sequence has been specified
- Concurrent Verification is specified for any step that requires Verification by a second individual.
6.7.2 Sign the Tagout as Authorizer.
6.7.3 If placement verification of any component is to be waived, then N/A applicable "Verified By" blocks.
6.7.4 Print the Tagout Cover Sheet, Hang List and tags associated with this Tagout.
6.7.5 Verify a Tagout hang sequence has been specified.
6.7.6 Assign and brief a Tagger to hang the Tagout. If needed, assign and brief a Concurrent Verifier to accompany the Tagger.
(, 6.7.7 Ensure appropriate Control Room personnel are notified of the status of the Tagout.
6.8 Tag Placement NOTE: Tags shall not be hung on components that are not properly labeled. If a component label is missing or in error, consult the Tagging Official before proceeding.
6.8.1 Personnel hanging or verifying tags shall have in their possession the applicable Tag Hang List.
6.8.2 The Tagger will hang the tags using STAR, as follows:
6.8.2.1 Verify all required authorizations have been obtained.
6.8.2.2 Review the T agout detail.
Tagout 6.8.2.3 Ensure the Tagout tagging sequence is followed.
6.8.2.4 Perform the following for each component listed on the Tag Hang List:
- Verify component label matches component identification on the Tag.
- Place or verify components in the required positions.
- Ensure expected response is received.
- Hang the tag on the component (see NMP-AD-003-001, step 6.3).
- Utilize place keeping on the Tag Hang List.
- Sign the Tagout as Placement 1st verification.
SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH SYSTEM OPERATING PROCEDURE 1 OF 54 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM 34S0-P42-001-2 34S0-P42-00 1-2 15.9 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR G. L. Johnson DATE 01-31-03 DATE:
N/A SSM/PM 5-4-09 N/A DATE N/A 1.0 OBJECTIVE This procedure gives instructions for operation of the Reactor Building Closed Cooling Water System.
TABLE OF CONTENTS Section 2.0 APPLiCABILITy .................................................................................................................. 2
3.0 REFERENCES
................................................................................................................... 2 4.0 REQUIREMENTS ............................................................................................................... 3 5.0 PRECAUTIONS/LIMITATIONS ........................................................................................... 3 5.1 PRECAUTIONS ............................................................................................................ 3 5.2 LIMITATIONS ............................................................................................................... 3
( 6.0 PREREQUISITES ............................................................................................................... 4 7.0 PROCEDURE ..................................................................................................................... 5 7.1 SYSTEM STARTUP AND OPERATION ....................................................................... 5 7.2 SYSTEM SHUTDOWN ................................................................................................. 8 7.3 INFREQUENT OPERATIONS ....................................................................................... 9 7.3.1 RBCCW SYSTEM FILL AND VENT ....................................................................... 9 7.3.2 SWAPPING RBCCW HEAT EXCHANGERS, PLACING 'A' INTO SERVICE AND REMOVING 'B' .................................................................................................... 11 7.3.3 SWAPPING RBCCW HEAT EXCHANGERS, PLACING 'B' INTO SERVICE AND REMOVING 'A' .................................................................................................... 13 7.3.4 SWAPPING RBCCW PUMPS .............................................................................. 15 7.3.5 DRAINING AN RBCCW HEAT EXCHANGER ..................................................... 16 7.3.5.1 Draining RBCCW Heat Exchanger 2P42-B001A 2P42-B001A............................................
.......................................... 16 7.3.5.2 Draining RBCCW Heat Exchanger 2P42-B001B 2P42-B001 B........................................... 17 7.3.6 FILLING A DRAINED RBCCW HEAT EXCHANGER ........................................... 18 7.3.6.1 Filling Drained RBCCW Heat Exchanger, 2P42-B001 2P42-B001A A ................................ 18 7.3.6.2 Filling Drained RBCCW Heat Exchanger, 2P42-B001 B ................................ 20 7.3.7 SINGLE RBCCW PUMP OPERATION ................................................................ 22 7.3.8 ADJUSTING HEAT EXCHANGER PSW/RBCCW DIFFERENTIAL PRESSURE. 24 7.3.9 FILL, VENT AND UNISOLATE AN RBCCW PUMP ............................................. 25 7.3.9.1 Fill, Vent And Unisolate RBCCW Pump 2P42-C001A ................................... 25 7.3.9.2 Fill, Vent AND Unisolate RBCCW Pump 2P42-C001 B .................................. 27 7.3.9.3 Fill, Vent AND Unisolate RBCCW Pump 2P42-C001 2P42-C001C C .................................. 29 7.3.10 LOWERING RBCCW SURGE TANK WATER LEVEL ......................................... 31
(
MGR-0002 Ver. 8.1
SOUTHERN NUCLEAR PAGE
,/ PLANT E. I. HATCH 2 OF 54 20F54 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM 34S0-P42-001-2 15.9 Attachments 1 RBCCW SYSTEM ELECTRICAL LINEUP ........................................................................ 32 2 RBCCW SYSTEM VALVE LINEUP ................................................................................... 35 3 RBCCW SYSTEM INSTRUMENT LINEUP ....................................................................... 49 2.0 APPLICABILITY This procedure applies to applies to the Unit Two Reactor Building Closed Cooling Water System.
3.0 REFERENCES
3.1 H-26054 and H-26055, Reactor Building Closed Cooling Water Sys P&ID 3.2 H-27750 and H-27751, Reactor Building Closed Cooling Water System 2P42 3.3 H-23367, Single Line Diagram Turbine Building 120/20BV 120/208V Station Service System 2R20D MPL's 2R25-S039 and 2R25-S041 3.4 H-26062, Reactor Building North Side Interruptible Inst. Air P&ID
(
3.5 H-23454, Wiring Diagram Essential and Instrument 120/20BV 120/208V Cabinets 2R25-S036, S037, S064,
& S065
&S065 3.6 120/208V Essential AC System 2R20N MPL's 2R25-S064 and H-23369, Single Line Diagram 120/20BV 2R25-S065 3.7 FSAR, Unit Two, Section 9.2.2 3.B 3.8 A-26415-P42-B, Instrument I nstrument Level Setting Diagram Diag ram
(
MGR-0001 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 3 OF 54
(
DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM 34S0-P42-001-2 15.9 4.0 REQUIREMENTS 4.1 PERSONNEL REQUIREMENTS The number AND qualification level of Operations personnel performing the sections in this procedure will be determined by the Unit Two Shift Supervisor.
4.2 MATERIALS AND EQUIPMENT Hoses AND connectors for draining OR filling a Heat Exchanger.
4.3 SPECIAL REQUIREMENTS 4.3.1 Independent verification, as described in 10AC-MGR-019-0, Procedure Use AND Adherence, will be required for portions of this procedure, as indicated by either a step directing to perform Independent verification OR a "VERIFIED" sign-off blank.
5.0 PRECAUTIONS/LI PRECAUTIONS/LIMITATIONS MITATIONS 5.1 PRECAUTIONS
( 5.1.1 Notify laboratory supervision PRIOR to draining system OR any part of system to Radwaste.
5.1.2 Performance of valve lineups AND verifications as required by this procedure may require entry into high radiation areas.
Follow proper radiation protection practices/procedures to maintain personnel exposure ALARA.
5.2 LIMITATIONS N/A - NOT applicable to this procedure
(
MGR-0001 Ver.4 Ver. 4
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 4 OF 54 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM 34S0-P42-001-2 15.9 6.0 PREREQUISITES 6.1 AC Power available 6.2 DC Power available 6.3 Instrument Air System in operation per 34S0-P51-002-2, Instrument AND Service Air Systems 6.4 Makeup Demineralizer System in operation per 64CH-OPS-005-0, Makeup Demineralizer Sys.
NOTE: System valve, electrical, AND instrument valve lineups provided on the attachments are to be completed as required by 34GO-OPS-003-2, Startup System Status Checklist.
6.5 RBCCW Electrical Lineup complete per Attachment 1 6.6 RBCCW Valve Lineup complete per Attachment 2 6.7 RBCCW Instrument Lineup complete per Attachment 3
(
6.8 Plant Service Water is in operation per 34S0-P41-001-2, Plant Service Water System 6.9 Process Radiation Monitoring System in operation 6.10 An RWP may be required to perform parts of this procedure.
(
MGR-0001 Ver. 4
SNC PLANT E. I. HATCH I Pg 32 of 54 DOCUMENT TITLE: I I DOCUMENT NUMBER: Version No:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I 34S0-P42-001-2 15.9 ATTACHMENT _1 Att. Pg.
TITLE: RBCCW SYSTEM ELECTRICAL LINEUP 1 of 3 PERSON(S) PERFORMING OR VERIFYING TASK (PRINT NAME) INITIALS
(
LINEUP COMPLETED: TIME :_ _ _ _ _ ET DATE :__
DATE: _~ _ ______
_ _ _~ __ __
(mm I/ dd I/ yy)
REVIEWED BY: _ _ _ _ _ _ _ _ _ ____
Shift Supervisor Time (ET) Date (mm I/ dd I/ yy)
COMMENTS: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _
OPS-0679 Ver. 1 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 33 of 54 DOCUMENT TITLE:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I I
I DOCUMENT NUMBER:
34S0-P42-001-2 Version No:
15.9 ATTACHMENT J.. _1 Att. Pg.
TITLE: RBCCW SYSTEM ELECTRICAL LINEUP 2of3 2 of 3 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION 2R23-S003 600V Station Serv Swgr 2C, 130TET14 RACK IN Frame 2B RBCCW Pump 2A 2P42 COO1A OPEN Frame 3B RBCCW Pump 2C 2P42 C001 C RACK IN OPEN 2R23-S004 600V Station Serv SWGR 2D, 130TCT14 RACK IN Frame 4B RBCCW Pump 2B 2P42-C001 B OPEN 2R25-S065120/208 2R25-S065 120/208 DIST CAB 2C Instrument Bus 2B, 130TGT12 Turb. BUild.
Build. Vent Sys., Drain. Pumps N2 Inert.
BRKR10 CLOSED (RBCCW Suct. Temp. Switch 2P42-TS-R600)
Cont. Rm Rx Bldg DrywelllTorus Drywell/Torus Ventilation (RBCCW BRKR15 CLOSED Pressure Transmitter 2P42-NOO3) 2R25-S041 205/120 VAC DIST CAB, Water Analysis Room 112TBT13
( BRKR9 P11, P21 AND P42 Valve Control Panel Pane12H11-P650 2H 11-P650 CLOSED BRKR BRKR1717 RBCCW Pumps 2P42-C001A AND C Motor Heaters CLOSED BRKR18 RBCCW Pumps 2P42-C001 B Motor Heater CLOSED 2R24-S012 600/208V MCC 2B ESS Div 2, 130RHR24 RACK IN Frame 9A RBCCW Drwllnlet Isolation Valve, 2P42-F051 CLOSED Frame 14C RACK IN RBCCW Drwl Outlet Isolation Valve, 2P42-F052 CLOSED 2R25-S102 Miscellaneous Power Panel 130RHR24 Drywell/Torus Flow/Pressure Indication(RBCCW Temp BRKR2 CLOSED Indicators)
Indicators)
BRKR3 Surge Tank Level Control Valve, 2P42-F054 CLOSED 2R25-S105 Miscellaneous Power Panel, 185RHR23 RBCCW Chemical Addition Tk - Level Switch Control BRKR 11 CLOSED for 2P42-COO2 BRKR 39 RBCCW Metering Pump 2P42-COO2 CLOSED BRKR 40 BRKR40 RBCCW Chemical Addition Tank Agitator, 2P42-DOO4 CLOSED
(
OPS-0679 Ver. 1 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 34 of 54
(
DOCUMENT TITLE:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I I
I DOCUMENT NUMBER:
34S0-P42-001-2 Version No:
15.9
\
ATTACHMENT _1 Att. Pg.
TITLE: RBCCW SYSTEM ELECTRICAL LINEUP 3 of 3 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION 2R25-S107 Distribution Panel203RLR21 BRKR6 RBCCW Surge Tank Alarm Logic 2H21-P350 CLOSED 2H11-P650 Control Switches 2P42-F051 RBCCW DIW Inlet Isolation, 2P42-F051 OPEN 2P42-F052 RBCCW DIW Outlet Isolation, 2P42-F052 OPEN PULL TO 2P42-S1 RBCCW Pump 2A LOCK PULL TO 2P42-S2 RBCCW Pump 2B LOCK PULL TO 2P42-S3 RBCCW Pump 2C LOCK 2R25-S064 Instrument Bus 2A, 130TGTT13
( BRKR 11 RBCCW Suct. Temp Switch 2P42-TS-R600 / RBCCW Pressure Transmitter CLOSED 2P42-N003 //2P42-PS-R601 2P42-PS-R601 OPS-0679 Ver. 1 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I ~g 35 of 54 Pg DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I 34S0-P42-001-2 15.9 ATTACHMENT .1-
..£. Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 1 of 14 PERSON(S) PERFORMING OR VERIFYING TASK (PRINT NAME) INITIALS
(
LINEUP COMPLETED: TIME ::, _
_ ET DATE: _ _______ __ __ ____
(mm I/ dd I/ yy)
REVIEWED BY: ___
Shift Supervisor Time (ET) Date (mm I/ dd I/ yy)
COMMENTS: __________________________________________________ ____
(
OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 36 of 54 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I 34S0-P42-001-2 15.9 ATTACHMENT 2. 2- Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 2 of 14 VALVE NORMAL DESCRIPTION CHECKED NUMBER POSITION 112TCT15 RBCCW Pump AND Heat Exchanger Area 2P42-FVOO2 RBCCW Heat Exchanger, 2P42-B001 B, Inlet Header Vent CLOSED 2P42-FOO2B RBCCW Heat Exchanger, 2P42-B001 2P42-B001B,B, Inlet Isolation OPEN RBCCW Heat Exchanger, 2P42-R001 B, Inlet Pressure Indicator, 2P42-F3047B OPEN 2P42-R002B, Root Valve RBCCW to Service Water Differential Pressure Sw.,
2P42-F3046B OPEN 2P42-dps-N065B, RBCCW Side Root Vlv RBCCW to Service Water Diff. Press. Switch 2P42-dps-N065B 2P42-F3045B OPEN Service Water Side Root Vlv 2P42-F108B RBCCW Heat Exchanger 2P42-B001 B, Service Water Side Drain CLOSED 2P42-F041 B RBCCW Heat Exchanger, 2P42-B001 B, RBCCW Side Drain CLOSED 2P42-F102B RBCCW Heat Exchanger, 2P42-B001 B, Service Water Side Vent CLOSED
(
2P42-F040B RBCCW Heat Exchanger, 2P42-B001 B, RBCCW Side Vent CLOSED RBCCW Heat Exchanger, 2P42-B001 B, Outlet Pressure Indicator 2P42-F3049B OPEN 2P42-PI-R005B Root Valve 2P42-FOO1B RBCCW Heat Exchanger, 2P42-B001 B, Outlet Isolation OPEN 2P42-FDOO2 RBCCW Heat Exchanger, 2P42-B001B, 2P42-B001 B, Outlet Header Drain CLOSED 2P42-FVOO3 RBCCW Heat Exchanger, 2P42-B001 B, Outlet Header Vent CLOSED 2P42-FVOO1 RBCCW Heat Exchanger, 2P42-B001A, Inlet Header Vent CLOSED 2P42-FOO2A RBCCW Heat Exchanger, 2P42-B001A, Inlet Isolation OPEN RBCCW Heat Exchanger, 2P42-B001A, Inlet Pressure Indicator 2P42-F3047A OPEN 2P42-R002A Root Valve RBCCW to Service Water Diff. Press. Switch, 2P42-dps-N065A, 2P42-F3046A OPEN RBCCW Side Root Valve RBCCW to Service Water Diff. Press. Switch, 2P42-dps-N065A, 2P42-F3045A OPEN Service Water Side Root Valve
(
OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
HATCH I SNC PLANT E. I. HATCH! Pg 37 of 54 DOCUMENT TITLE: I
! DOCUMENT NUMBER: Version No:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM SYSTEM!I 34S0-P42-001-2 15.9 ATTACHMENT 2. 2- Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 3 of 14 VALVE NORMAL DESCRIPTION CHECKED NUMBER POSITION 112TCT15 RBCCW Pump AND Heat Exchanger Area 2P42-F108A RBCCW Heat Exchanger, P24-B001A, Service Water Side Drain CLOSED 2P42-F041A RBCCW Heat Exchanger, 2P42-B001A, RBCCW Side Drain CLOSED 2P42-F102A RBCCW Heat Exchanger, 2P42-B001A, Service Water Side Vent CLOSED 2P42-F040A RBCCW Heat Exchanger, 2P42-B001A, RBCCW Side Vent CLOSED RBCCW Heat Exchanger, 2P42-B001A, Outlet Pressure Indicator 2P42-F3049A OPEN 2P42-PI-R005A Root Valve 2P42-FOO1A RBCCW Heat Exchanger, 2P42-B001A, Outlet Isolation OPEN 2P42-FDOO1 RBCCW Heat Exchanger, 2P42-B001A, Outlet Header Drain CLOSED 2P42-PX-N059, RBCCW Pumps Suction Header Pressure 2P42-F3025 CLOSED Connection, Root Valve 2P42-PX-N017, RBCCW Pumps Suction Header Sample
( 2P42-F3000 Connection, Root Valve CLOSED RBCCW Pumps Suction Header, Sample Coupon AND Filter 2P42-F118 OPEN Connection 2P42-FOO4C RBCCW Pump, 2P42-C001 C, Suction Valve OPEN RBCCW Pumps Discharge Header, Sample Coupon AND Filter 2P42-F119 OPEN Connection 2P42-PX-N064C, RBCCW Pump 2P42-C001C Suction Pressure 2P42-F3027C CLOSED Connection, Root Valve 2P42-PX-N058C, RBCCW Pump 2P42-C001C 2P42-C001 C Suction Pressure 2P42-F3024C CLOSED Connection, Root Valve 2P42-F044C RBCCW Pump, 2P42-C001 C, Casing Vent CLOSED RBCCW Pump, 2P42-C001 C, Discharge Pressure Indicator, 2P42-F3048C OPEN 2P42-PI-R003C, Root Valve 2P42-FOO5C RBCCW Pump, 2P42-C001 C, Discharge Valve OPEN 2P42-FDOO3 RBCCW Pump, 2P42-C001 C, Discharge Line Drain CLOSED 2P42-FOO4B RBCCW Pump, 2P42-C001 B, Suction Valve 2P42-C001B, OPEN
(
OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
HATCHjI SNC PLANT E. I. HATCH Pg 38 of 54 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:
\
\
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I 34S0-P42-001-2 15.9 ATTACHMENT .1.. Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 4 of 14 NORMAL VALVE NUMBER DESCRIPTION CHECKED POSITIO 112TCT15 RBCCW Pump AND Heat Exchanger Area 2P42-PX-N064B, RBCCW Pump 2P42-C001 B Suct. Press.
2P42-F3027B CLOSED Connection, Root Valve 2P42-PX-N058B, RBCCW Pump 2P42-C001 B Suct. Press.
2P42-F3024B CLOSED Connection, Root Valve 2P42-F044B RBCCW Pump, 2P42-C001 B, Casing Vent CLOSED RBCCW Pump, 2P42-C001 B, Disch. Press. Indicator 2P42-2P42-F3048B OPEN PI-R003B Root Valve 2P42-F005B RBCCW Pump, 2P42-C001 B, Discharge Valve OPEN 2P42-FD004 RBCCW Pump, 2P42-C001 B, Discharge Line Drain CLOSED 2P42-F004A RBCCW Pump, 2P42-C001A, Suction Valve OPEN 2P42-PX-N064A, RBCCW Pump 2P42-C001A Suct. Press.
2P42-F3027 A CLOSED Connection, Root Valve 2P42-PX-N058A, RBCCW Pump 2P42-C001A Suct. Press.
( 2P42-F3024A Connection, Root Valve CLOSED 2P42-F044A RBCCW Pump, 2P42-C001A, Casing Vent CLOSED RBCCW Pump, 2P42-C001A, Disch. Press. Indicator, 2P42-PI-2P42-F3048A OPEN R003A, Root Valve 2P42-F005A RBCCW Pump, 2P42-C001A, Disch. Valve OPEN 2P42-FD005 RBCCW Pump, 2P42-C001A, Disch. Line Drain CLOSED 2P42-FV004 RBCCW Pumps Discharge Header Vent CLOSED RBCCW Pumps Disch. Header Press. Switches 2P42-PS-NOO5A, 2P42-F3029 OPEN N005B, N005C AND Press. Xmitter, 2P42-PT-N003, Root Vlv RBCCW Pumps Disch. Header Flow Indic., 2P42-FI-R031, High 2P42-R031- RV-1 OPEN Pressure Root Valve RBCCW Pumps Disch. Header Flow Indic., 2P42-FI-R031, Low 2P42-R031- RV-2 OPEN Pressure Root Valve 2P42-SX-N018, RBCCW Heat Exchangers Inlet Header Sample 2P42-F3001 CLOSED Connection, Root Valve 2P42-F135 RBCCW Corrosion Test Loop Isolation Valve OPEN OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
HATCHjI SNC PLANT E. I. HATCH Pg 39 of 54 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:
( REACTOR BUILDING CLOSED COOLING WATER SYSTEM I 34S0-P42-001-2 15.9 ATTACHMENT £ 2. Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 5 of 14 VALVE NORMAL DESCRIPTION CHECKED NUMBER POSITION 112TCT15 RBCCW Pump AND Heat Exchanger Area 2P42-F136 RBCCW Corrosion Test Loop Inlet Valve OPEN 2P42-F137 RBCCW Corrosion Test Loop Outlet Valve OPEN 2P42-F138 RBCCW Filter, 2P42-D021, Inlet Valve CLOSED 2P42-F139 RBCCW Filter, 2P42-D021, Outlet Valve CLOSED RBCCW Filter, 2P42-D021, Inlet Pressure Indicator, 2P42-R037, 2P42-F140 OPEN Root Valve RBCCW Filter, 2P42-D021, Outlet Pressure Indicator, 2P42-R036, 2P42-F141 OPEN Root Valve 2P42-F142 RBCCW Corrosion Test Loop Vent Valve CLOSED 2P42-F143 RBCCW Corrosion Test Loop Vent Valve CLOSED I 2P42-F144 RBCCW System Sample Valve CLOSED
\
2P42-F145 RBCCW Filter, 2P42-P021, Drain Valve CLOSED 203RHR19 Demineralized Water to RBCCW Chemical Addition Tank, 2P42-F028 CLOSED 2P42-A002, Isolation Valve 2P42-F029 RBCCW Chemical Addition Tank, 2P42-A002, Drain Valve CLOSED RBCCW Chemical Addition Metering Pump, 2P42-C002, Suction 2P42-F082 OPEN Valve RBCCW Chem Add Metering Pmp, 2P42-C002, Disch Press Indic, 2P42-F3050 OPEN 2P42-R006, RV RBCCW Chemical Addition Metering Pump, 2P42-C002, Discharge 2P42-F031 CLOSED Valve 203RHR21 Demineralized Water to RBCCW Surge Tank 2P42-A001, Isolation 2P42-F027 CLOSED Valve RBCCW Surge Tank, 2P42-A001, Level Control Valve, 2P42-F054, 2P42-F055 CLOSED Bypass Valve RBCCW Surge Tank, 2P42-A001, Level Control Valve, 2P42-F054, 2P42-F100 OPEN Downstream Isol.
(
OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 40 of 54 DOCUMENT TITLE:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I I
I DOCUMENT NUMBER:
34S0-P42-001-2 Version No:
15.9 ATTACHMENT .£ 2- Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 6 of 14 VALVE NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION 203RHR21 2P42-F032 RBCCW Surge Tank, 2P42-A001, Drain Valve CLOSED 2P42-F054, RBCCW Surge Tank, 2P42-A001, Level 2P52-F701 OPEN Control Valve, Instr. Air Header Isol 2P42-F054, RBCCW Surge Tank, 2P42-A001, Level 2P42-F054-AS-1 OPEN Control Valve, Instr. Air Supply RBCCW Surge Tank, 2P42-A001, Outlet to RBCCW LOCKED 2P42-F057 System Isolation Valve OPEN RBCCW Surge Tank, 2P42-A001, Outlet to RBCCW 2P42-FV006 CLOSED System Vent RBCCW Surge Tank, 2P42-A001, Level Glass 2P42-F3030 OPEN 2P42-LG-D020 Upper Root Valve RBCCW Surge Tank, 2P42-A001, Level Glass 2P42-F3031 OPEN 2P42-LG-D020 Lower Root Valve RBCCW Surge Tank, 2P42-A001, Level Glass 2P42-F134 CLOSED 2P42-LG-D020 Drain Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F3035 OPEN 2P42-LS-N033 Lower Root Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F129 CLOSED 2P42-LS-N033 Drain Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F3034 OPEN 2P42-LS-N033 Upper Root Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F128 CLOSED 2P42-LS-N033 Vent Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F3039 OPEN 2P42-LS-N060 Lower Root Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F131 CLOSED 2P42-LS-N060 Drain Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F3038 OPEN 2P42-LS-N060 Upper Root Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F130 CLOSED 2P42-LS-N060 Vent Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F3043 OPEN 2P42-LS-N044 Lower Root Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F133 CLOSED 2P42-LS-N044 Drain Valve
(
OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
HATCH!I SNC PLANT E. I. HATCH Pg 41 of 54 DOCUMENT TITLE: I! DOCUMENT NUMBER: Version No:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM! SYSTEM I 34S0-P42-001-2 15.9 ATTACHMENT ..£.
- 2. Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 7 of 14 VALVE NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION 203RHR21 RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F3042 OPEN 2P42-LS-N044 Upper Root Valve RBCCW Surge Tank, 2P42-A001, Level Switch 2P42-F132 CLOSED 2P42-LS-N044 Vent Valve 194RAR23 SBGT Rm Above Dividing Wall RBCCW Surge Tank, 2P42-A001, Line To RBCCW System 2P42-FV005 2P42-FVOO5 CLOSED Vent Valve Drywell 2P42-FD021 RBCCW to Drywell Supply Header Inboard Drain Valve CLOSED 2P42-FD028 RBCCW to Drywell Supply Header Outboard Drain Valve CLOSED 1/
//
RBCCW to Drywell Supply Header Pressure Indicator, 2P42-F3051A OPEN 2P42-PI-R013A, Root Valve
( 2P42-F3056A 2P42-PI-R021A, Reactor Recirc Pump, 2B31-COO1A RBCCW Outlet Press Indic, RV 2B31-C001A OPEN RBCCW to Drywell Supply Header Pressure Indicator, 2P42-F3051 B OPEN 2P42-PI-R013B, Root Valve 2P42-PI-R021 B, Rx Recirc Pump, 2B31-C001B 2P42-PI-R021B, 2B31-C001 B RBCCW 2P42-F3056B OPEN Out Press Indic, RV 2P42-F042 Drywell Sump Cooler, 2G11-B001, RBCCW Inlet Isolation OPEN 2P42-PX-N043, Drywell Sump Cooler, 2G11-BOO1, 2G11-B001, 2P42-F3015 CLOSED RBCCW Inlet Press Conn, RV 2P42-F068 Drywell Sump Cooler, 2G11-B001, RBCCW Inboard Drain CLOSED Drywell Sump Cooler, 2G11-B001, RBCCW Outboard 2P42-FD034 CLOSED Drain 2P42-PI-R032, Drywell Sump Cooler, 2G11-B001, RBCCW 2P42-F3061 OPEN Outlet Press Indic, RV 2P42-F010 Drywell Sump Cooler, 2G11-B001, RBCCW Outlet Isolation OPEN 2P42-FD022 RBCCW to Drywell Return Header Inboard Drain CLOSED 2P42-FD029 RBCCW to Drywell Return Header Outboard Drain CLOSED
(
OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 42 of 54 DOCUMENT TITLE:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I I
I DOCUMENT NUMBER:
34S0-P42-001-2 Version No:
15.9 ATTACHMENT 2- 2... Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 8 of 14 VALVE NORMAL DESCRIPTION CHECKED NUMBER POSITION 087RAR14 NW Corner Room 2P42-F104 Sample Cooler in 2P33-P101 RBCCW Outlet Isolation OPEN 2P42-F105 Sample Cooler in 2P33-P101 RBCCW Outlet Isolation OPEN 2P42-F111 Sample Cooler in 2P33-P101 RBCCW Outlet Isolation OPEN 2P42-SX-N050, Sample Cooler in 2P33-P1 01 RBCCW Outlet Press 2P42-F3019 CLOSED Conn, Root Valve 087RAR24 SW Corner Room 2P42-F013 Reactor Building Sump Cooler, 2G11-B002, RBCCW Inlet Isolation OPEN N42-PX-N040, Reactor Building Sump Clr, 2G11-B002, Inlet Press 2P42-F3013 CLOSED Conn, Root Valve 2P42-F061 Reactor Building Sump Cooler, 2G11-B002, 2G 11-B002, RBCCW Side Drain CLOSED 2P42-SX-N049, Reactor Building Sump Clr, 2G11-B002, RBCCW Out
( 2P42-F3018 Sample Conn, RV CLOSED P42-PI-R017, Reactor Building Sump Cooler RBCCW Outlet Pressure 2P42-F3054 OPEN Indicator Root Valve 2P42-F014 Reactor Building Sump Cooler, 2G11-B002, RBCCW Outlet Isolation OPEN 118RAR24 SW Corner Room 2P42-F011A CRD Pump, 2C11-C001A, RBCCW Inlet Isolation OPEN 2P42-PX-N041A, CRD Pump, 2C11-C001A, RBCCW Inlet Press Conn, 2P42-F3014A CLOSED Root Valve 2P42-F097A CRD Pump, 2C11-C001A, Oil Cooler RBCCW Outlet OPEN 2P42-F098A CRD Pump, 2C11-C001A, Inboard Bearing Cooler Outlet OPEN 2P42-F099A CRD Pump, 2C11-C001A, Outboard Bearing Cooler Outlet OPEN 2P42-PI-R016A, CRD Pump, 2C11-C001A, RBCCW Outlet Pressure 2P42-F3053A OPEN Indicator, RV 2P42-F012A CRD Pump, 2C11-C001A, RBCCW Outlet Isolation OPEN OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 43 of 54 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I 34S0-P42-001-2 15.9 ATTACHMENT 2- Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 9 of 14 NORMAL VALVE NUMBER DESCRIPTION CHECKED POSITION 118RAR24 SW Corner Room 2P42-F011B CRD Pump, 2C11-C001B, 2C11-C001 B, RBCCW Inlet Isolation OPEN 2P42-PX-N041A, CRD Pump, 2C11-C001B, RBCCW Pressure 2P42-F3014B CLOSED Connection, Root Valve 2P42-F097B CRD Pump, 2C11-C001 B, Bearing Cooler Oil Cooler Outlet OPEN 2P42-F098B CRD Pump, 2C11-C001 B, Inboard Bearing Cooler Outlet OPEN 2P42-F099B CRD Pump, 2C11-C001 B, Outboard Bearing Cooler Outlet OPEN 2P42-PI-R016B, CRD Pump, 2C11-C001B, 2C11-C001 B, RBCCW Outlet Press 2P42-F3053B OPEN Indicator, Root Valve 2P42-F012B CRD Pump, 2C11-C001B, RBCCW Outlet Isolation OPEN 130RBR20
( 2P42-F033 RBCCW Supply to 158' Elevation AND above Isolation OPEN 2P42-FV033 RBCCW Supply to 158' Elevation AND above Header Vent CLOSED 2P42-FV034 RBCCW Retum Return from 158' Elevation AND above Header Vent CLOSED 2P42-FD101 RBCCW Supply to 158' Elevation AND above Header Drain CLOSED 2P42-FD102 RBCCW Return from 158' Elevation AND above Header Drain CLOSED OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
HATCHII SNC PLANT E. I. HATCH Pg 44 of 54 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:
( REACTOR BUILDING CLOSED COOLING WATER SYSTEM I 34S0-P42-001-2 15.9 ATTACHMENT .£ 2.. Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 10 of 14 NORMAL VALVE NUMBER DESCRIPTION CHECKED VERIFIED POSITION 087RAR19 Torus Area LOCKED 2P42-F067A RBCCW Supply Header Drain to Radwaste CLOSED 2P42-F067B LOCKED RBCCW Return Header Drain to Radwaste CLOSED 2P42-F072 RBCCW Supply to Sample Panel in 2P33-P101 OPEN RBCCW Supply to RBCCW Loads Below 158' Elevation LOCKED 2P42-F034 Isolation OPEN 2P42-FT-N032, RBCCW Flow to Drywell, High Side Root 2P42-FT-N032, 2P42-N032-RV-1 OPEN Valve 2P42-FT-N032, RBCCW Flow to Drywell, Low Side Root 2P42-N032-RV-2 OPEN Valve TORUS BAY 6, INNER CATWALK 2P42-F150 Drywell Inlet Test Valve RBCCW Drywelliniet CLOSED 2P42-F151 RBCCW Drywell Outlet Test Valve CLOSED LOCKED 2P42-F152 RBCCW Drywell Inlet Isolation Valve OPEN 2P42-F153 LOCKED RBCCW Drywell Outlet Isolation Valve OPEN 168RBR17 2P42-F079 Drywell Pneumatic System RBCCW Inlet Isolation OPEN 2P42-PX-N061, Drywell Pneumatic System RBCCW Inlet 2P42-F3026 CLOSED Pressure Connection, Root Vlv Drywell Pneumatic System RBCCW System RBCCW Inlet 2P42-F081 CLOSED Drain 2P42-PI-R020, Drywell Pneumatic System RBCCW Outlet 2P42-F3055 OPEN Press Indicator, Root Valve 2P42-F103 Drywell Pneumatic System RBCCW Outlet Vent CLOSED 2P42-F080 Drywell Pneumatic System RBCCW Outlet Isolation OPEN OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 45 of 54 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:
REACTOR BUILDING CLOSED COOLING WATER SYSTEMi SYSTEM I 34S0-P42-001-2 15.9 ATTACHMENT .£ .2... Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 11 of 14 NORMAL VALVE NUMBER DESCRIPTION CHECKED POSITION 158RJR21 2P42-F073 RWCU Pump Coolers RBCCW Inlet Isolation OPEN 2P42-F024 RWCU Pump Coolers RBCCW Outlet Isolation OPEN 2P42-PI-R022, RWCU Pump Coolers RBCCW Outlet Press Indicator, 2P42-F3057 OPEN Root Valve Recirc A50 2B31-5002A 158RJR15 RecircASO 2831-S002A 2P42-F576 Rx Bldg Interruptible Instrument Air Header Isolation CLOSED 2P42-FV021 ASD's, 2B31-S002A & B, RBCCW Supply Header Vent CLOSED 2P42-FV022 ASD's, 2B31-S002A & B, RBCCW Supply Return Header Vent CLOSED 158RFR13 Recirc ASO A50 2B31-5002B 2831-S0028
( 2P52-F577 Rx Bldg Interruptible Instrument Air Header Isolation CLOSED 158RFR23 RBCCW to Non-Regenerative Heat Exchange, 2G31-B002, Flow 2P42-R034-RV-1 OPEN Indic High Side Root Valve RBCCW to Non-Regenerative Heat Exchange, 2G31-B002, Flow 2P42-R034-RV-2 OPEN Indic Low Side Root Valve RWCU Non-Regenerative Ht Exchangers, 2G31-B002 RBCCW 2P42-F038 **OPEN OPEN Outlet Throttle Vlv Non-Regenerative Heat Exchanger, 2G31-B002, Vent 2G31-F041A CLOSED (these are vents for RBCCW side of Hx)
Non-Regenerative Heat Exchanger, 2G31-B002, Vent 2G31-F041 B CLOSED (these are vents for RBCCW side of Hx)
{
\.
OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 46 of 54 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:
REACTOR BUILDING CLOSED COOLING WATER SYSTEMJ SYSTEM I 34S0-P42-001-2 15.9 ATTACHMENT .1.. Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 12 of 14 NORMAL VALVE NUMBER DESCRIPTION CHECKED POSITION 158RFR23 2G31-F026 Non-Regenerative Heat Exchanger, 2G31-B002, Drain CLOSED RWCU Non-Regenerative Heat Exchangers, 2G31-B002, RBCCW 2P42-F022 OPEN Outlet Isolation 2P42-F122 RBCCW Return From Cobalt Deposition Facility OPEN 2P42-SX-N029, RWCU Non-Regen. Ht Exch, 2G31-B002, RBCCW 2P42-F3005 CLOSED Out Smpl Conn, Isol 2P42-PX-N037, RWCU Non-Regen Ht Exch, 2G31-B002, RBCCW 2P42-F120 OPEN Inlet Press Conn, RV 2P42-F121 RBCCW Supply to Cobalt Deposition Facility CLOSED RWCU Non-Regenerative Heat Exchanger, 2G31-B002, Inlet 2P42-F021 OPEN Isolation 2P42-SX-N029, RWCU Non-Regen Ht Exch, 2G31-B002, RBCCW 2P42-F123 CLOSED Out Smpl Conn in Cobalt Deposition Facility, Isolation
( 158RFR13 2P42-PI-R015, RWCU Non-Regen Ht Exch, 2G31-B002, RBCCW 2P42-F3052 OPEN Out Press Indic, RV RWCU Non-Regenerative Heat Exchanger, 2G31-B002, RBCCW 2P42-F077 CLOSED Outlet Header Drain 2P42-PX-N037, RWCU Non-Regen Ht Exch, 2G31-B002, RBCCW 2P42-F3010 CLOSED Inlet Press Conn, Isol 158RFR19 RWCU Pump Room A 2P42-PX-N025A, RWCU Pump, 2G31-C001A, RBCCW Outlet Press 2P42-F3003A CLOSED Conn, Root Valve 2P42-F088A RWCU Pump, 2G31-C001A, RBCCW Outlet Valve OPEN 2P42-F089A RWCU Pump, 2G31-C001A, RBCCW Cooling Water Supply Valve OPEN 2P42-F090A RWCU Pump, 2G31-C001A, RBCCW Cooling Water Supply Valve OPEN 2P42-F091A RWCU Pump, 2G31-C001A, RBCCW Cooling Water Supply Valve OPEN 2P42-F117A RWCU Pump, 2G31-C001A, RBCCW Cooling Water Supply Valve OPEN
(
OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 47 of 54 DOCUMENT TITLE:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I I
I DOCUMENT NUMBER:
34S0-P42-001-2 Version No:
15.9 ATTACHMENT .£ Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 13 of 14 NORMAL VALVE NUMBER DESCRIPTION CHECKED POSITION 158RFR19 RWCU Pump Room A 2P42-F087A RWCU Pump, 2G31-C001A, RBCCW Supply Valve OPEN 2P42-F116A RWCU Pump, 2G31-C001A, Pedestal Drain to CHRW CLOSED 2P42-F064A RWCU Pump, 2G31-C001A, RBCCW Supply Line Vent CLOSED 2P42-PX-N038A, RWCU Pump, 2G31-C001A, RBCCW Sply Line 2P42-F3011A CLOSED Press Conn, RV 158RHR19 RWCU Pump Room B 2P42-F124 RWCU Pump, 2G31-C001B, RBCCW Outlet Valve OPEN 2P42-F125 RWCU Pump, 2G31-C001B, RBCCW Supply Valve OPEN 2P42-F127 RWCU Pump, 2G31-C001B, RBCCW Supply Line Drain CLOSED
( 2P42-F126 RWCU Pump, 2G31-C001B, RBCCW Discharge Line Vent CLOSED 185RFR24 2P42-F106 Sample Cooler, 2P33-P102, RBCCW Inlet Isolation OPEN 2P42-F3028 Sample Cooler, 2P33-P102, Rack Inlet Isolation OPEN 2P42-F107 Sample Cooler, 2P33-P102, Rack Outlet Isolation OPEN 2P42-F112 Sample Cooler, 2P33-P102, RBCCW Outlet Isolation OPEN 2P42-SX-N048, Sample Cooler, 2P33-P102, RBCCW Outlet Sample 2P42-F3017 CLOSED Conn, Root Valve 185RFR23 Fuel Pool Cooling Heat Exchanger Room 2P42-FV028 FPC Heat Exchanger, 2G41-B001, RBCCW Supply Header Vent CLOSED 2P42-FD023 FPC Heat Exchanger, 2G41-B001, RBCCW Supply Header Drain CLOSED 2P42-F023 FPC Heat Exchanger, 2G41-B001, RBCCW Inlet Isolation OPEN 2P42-PX-N036, FPC Heat Exchanger, 2G41-B001, RBCCW Inlet 2P42-F3009 CLOSED Press Conn, Root Valve
(
OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HA TCHLI HATCH Pg 48 of 54 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:
REACTOR BUILDING CLOSED COOLING WATER SYSTEM I 34S0-P42-001-2 15.9 ATIACHMENT .£.
ATTACHMENT 2- Att. Pg.
TITLE: RBCCW SYSTEM VALVE LINEUP 14 of 14 VALVE NORMAL DESCRIPTION CHECKED NUMBER POSITION 185RFR23 Fuel Pool Cooling Heat Exchanger Room 2P42-F075 FPC Heat Exchanger, 2G41-B001, RBCCW Side Drain CLOSED 2P42-SX-N028, FPC Heat Exchanger, 2G41-B001, RBCCW Out 2P42-F3004 CLOSED Sample Conn, Root Valve 2P42-PI-R311, FPC Heat Exchanger, 2G41-B001, RBCCW Outlet 2P42-F3062 OPEN Press Indic, Root Valve 2P42-F076 FPC Heat Exchanger, 2G41-B001, RBCCW Side Vent CLOSED 2P42-FV027 FPC Heat Exchanger, 2G41-B001, RBCCW Return Header Vent CLOSED 2P42-F039 FPC Heat Exchanger, 2G41-B001 RBCCW Return Outlet Throttle OPEN *
- Normal position of 2P42-F039 is OPEN, however it may require throttling lAW 34S0-G41-003-2 in order to maintain the Fuel Pool temperature greater than 68° F.
(,
(
OPS-0680 Ver. 9.3 G16.030 MGR-0009 Ver. 4
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN COMPANY A Management Verification Policy Version 3.0 y.,.,. W},rlJ' Entrv loServe Your EM"DtIlSnw ~,ur Procedure Page 1 of 11 Procedure Owner: Paul D. Rushton 1 Fleet Operations Manager 1 Corporate (Print: Name I Title I Site)
Approved By: Original Signed by Paul D. Rushton on 04/28/2008 (Procedure Owner's Signature I Date)
Effective Dates: N/A 6/1312008 6/13/2008 6/13/2008 Corporate FNP HNP VEGP This Standardization Process Control NMP is under the oversight of the Operations Peer Team.
Writer(s): Eric Snell
(
PROCEDURE USAGE REQUIREMENTS SECTIONS Procedure must be open and readily available at the Continuous Use: work location. Follow procedure step by step unless otherwise directed by the procedure.
Procedure or applicable section(s) available at the work Reference Use:
location for ready reference by by: person performing steps.
Information Use: Available on site for reference as needed. ALL
(
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUIHERNA SOUTHERN A Management Verification Policy Version 3.0
( COMPANY
~IU WtJ,IJ-ElIt'rJ!YllfSU/l.:~urWorld*
EnergyillSnn Procedure Page 2 of 11 Procedure Version Description Version Number Version Descri tion 2.0
- Steps 4.2 and 6.1.4, clarified that independent verifier can be involved in the prejob briefing and this does not constitute being "influenced" by the performer.
- Step 6.0, added expectation that guiding document include verification requirements or that requirement is established at the prejob brief.
- Steps 6.1.4 through 6.1.9 provided more specific detail in expectations for independent verification (including timing of the verification, verifier being separately dispatched, etc). Steps 6.1.4,6.1.6 6.1.4, 6.1.6 and 6.1.7 are new and the remaining steps in this section were re-numbered accordingly.
- 4.3 - changed definition of qualified reviewer 3.0 Added NOTE prior to step 6.1.6 to clarify Independent Verification requirement for activities in the main control room.
((
l\
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUIHERNA.
SOU'HERNA. Management Verification Policy Version 3.0 COMPANY EnerD IlJ$el"1If: Your Worur EnerC/t4l$erwYourWo,ur Procedure Page 3 of 11 Table of Contents 1.0 ....... ............................................................................................................................... 4 Purpose ......................................................................................................................................
2.0 .. .............................................................................................................................. 4 Applicability ................................................................................................................................
3.0 References ................................................................................................................................. 4 4.0 Definitions .................................................................................................................................. 4 5.0 Responsibilities .......................................................................................................................... 5 6.0 Procedure .................................................................................................................................. 5 6.1 Independent Verification ......................................................................................................... 6 6.2 Concurrent Verification ......................................................................................................... 10 6.3 Mispositioned Components Discovered During Verification .................................................. 11 7.0 Records ....................................................................................................................................
.................................................................................................................................... 11 8.0 Commitments ...........................................................................................................................
.... ....................................................................................................................... 11
(
\
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUIHERNA SOUTHERN COMPANY A Management Verification Policy Version 3.0 Enerx.y I()&rp~ ~1t1'w.,.1J" EMrtyl#Snw ~lIr Ws,hr Procedure Page 4 of 11 1.0 Purpose This procedure establishes policy and provides methods for verifying correct performance of normal operating, testing, and maintenance activities that affect the alignment or status of safety-related and some non-safety related systems or components.
2.0 Applicability This procedure applies to manipulation of power plant equipment where improper operation could create a challenge to plant safety or personnel safety or result in an unintended plant response such as a trip or ESFAS actuation.
3.0 References 3.1 NUREG 0737, Item I.C.6, "Guidance on Procedures for Verifying Correct Performance of Operating Activities" 3.2 USNRC IE Information Notice 84-51, "Independent Verification" 3.3 USNRC IE Information Notice 84-46, "Circuit Breaker Position Verification" 3.4 INPO 01-002 "Guidelines for the Conduct of Operations at Nuclear Power Stations"
( 4.0 Definitions 4.1 Concurrent Verification (CV) - Concurrent verification is the act of checking a condition, such as lifting a lead or installing a jumper, concurrent with the activities related to establishing the component's condition. Concurrent verification is used when an action or manipulation could result in an immediate threat to safe and reliable plant operation or a significant transient. Persons performing concurrent verifications identify the correct unit, train, or component and review the intended actions and expected responses before the task is performed, to prevent an unintended plant response.
4.2 Independent Verification (IV) - Independent verification is the act of checking the condition of a component independently from the individual responsible for establishing the component's condition. Independent verifications are truly independent in that the first and second checkers have no interaction during component manipulation. IV gives added assurance that a component is left in the required position and is used to verify the lineup of safety related equipment being returned to service.
4.3 Qualified Individual -
For CV - An individual possessing knowledge of the activity, systems, and/or components involved and the relationship of these activities, components, and systems to plant safety.
For IV - An individual who has basic knowledge of the type of component involved (valve, breaker, etc.). The individual need not be trained on the activity or system involved.
4.4 Significant Radiation Exposure - As applicable to activities described by this procedure,
( greater than or equal to 10 mrem whole body dose or airborne contamination in excess of ALARA guidelines.
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUIHERNA.
SOUTHERN COMPANY A. Management Verification Policy Version 3.0 l:~ltrut4SU"i! }~urWorlr Elu,ygyl#Su"e }~urWorl,r Procedure Page 5 of 11 5.0 Responsibilities 5.1 Department Managers Department managers shall ensure establishment of provisions within applicable procedures, which implement the policy, described herein.
5.2 Operations Manager The Operations Manager has overall responsibility for plant status control. As such the Operations Manager is responsible for proper implementation of this procedure at each respective SNC nuclear plant. He shall provide direction to other managers on implementation of this procedure, make any interpretations necessary and resolve issues that may arise.
Operations management reinforces site wide expectations that personnel conducting maintenance are responsible to ensure components are aligned properly after maintenance and to question off-normal components.
Establish, clearly communicate, and provide written guidance for routine component position verifications. Ensure that the guidance considers technical specification requirements, mode changes, and other transient conditions.
Ensure maintenance department personnel have a clear understanding of expectations for
( positioning components within the boundary of the tagout and for the need to ensure systems are properly aligned before restoration. A clear process is in place to track components repositioned within a tagout boundary.
5.3 Supervisors, Team Leaders, and Assistant Team Leaders Supervisors have the following responsibilities:
A. Only qualified individuals are assigned to perform verifications.
B. Verifications are performed in accordance with the policy described in this procedure.
6.0 Procedure Guidelines, Instructions or Forms developed to support this procedure (NMP-OS-002) will be reviewed and approved by the Operations Peer Team Champion or designee.
SNC uses two forms of verification: Independent Verification, and Concurrent Verification.
Instructions for documenting independent and concurrent verification shall be provided in applicable procedures.
The practice of verifying throttled valves by shutting and reopening the valves a prescribed number of turns can create valve mispositionings. Instead, use position indicators, scribe marks, or other recognizable indicators that have been designated to determine throttled valve positions. When shutting and reopening a throttled valve is necessary to determine its position, I perform concurrent verification rather than having both persons individually shut and reopen the
\
valve.
Southern Nuclear Operating Compaf'!}'
Company Nuclear NMP-OS-002 SOOTHERNA SOUTHERN A. Management Verification Policy Version 3.0
( COMPANY l:'ner1.1 II1&rJle RtNI'UfJ,Id" e-'X'Jto&rw nur War/J' Procedure Page 6 of 11 In some situations, functional testing may substitute for normal verification techniques in checking that components are correctly positioned. An example would be a full-flow test to prove the correct positioning of flow control valves. However, surveillance tests frequently will not serve to verify the positions of all components that are important to subsequent system operation. Therefore, use surveillance testing as component verification only if it can be shown that the test conclusively proves the position of the components. The Operations Manager must approve the use of surveillance testing applicability to satisfy component verification requirements.
The instructions for verification techniques describe the methods for verifying items such as manual valves, motor-operated and air-operated valves, solenoid-operated valves, circuit breakers, blank flanges, and removable links and fuses, as well as the status of control power.
During system lineups such as those performed coming out of a refueling outage, it is not necessary to have two people go to each component and check the position. A lineup is by definition a verification. Presumably, the components have already been positioned. At other times such as after system realignments, there should be a positioner and a verifier (Le., two people) to go to each component and make sure it is in the right position.
In most cases, the guiding document for an activity (procedure, tagout, work sequence, etc) should specify whether independent or concurrent verification is required. If not, the supervisor responsible for the activity will designate the type of verification at the prejob briefing.
6.1 Independent Verification
(
6.1.1 Independent Verification is required to be performed for restoration of:
A. Safety related systems or components.
B. Valve positions in liquid or gaseous radioactive waste systems that if mispositioned could lead to unintended or unmonitored radioactivity release.
- c. Other component positioning as determined necessary by the Operations Manager.
6.1.2 Exceptions - IV may be waived for the following reasons:
6.1.2.1 Significant radiation exposure.
In cases that involve significant 6.1.2.2 In cases that involve containment entry (PWR) or drywell entry (BWR), while containment integrity is established.
6.1.3 The Independent Verifier must be someone who has been independent of the task and has not been influenced by the positioner. The individual requesting the IV should tagout provide instructions to the verifier regarding the procedural step(s) or tag out points to be verified.
6.1.4 Depending on the job scope and complexity, consideration should be given by supervision for the independent verifier to attend the pre-job brief.
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN COMPANY A
A. Management Verification Policy Version 3.0 Ent!rgyltJ&r~t:
EneI'D leSer-v.: Yeut World' Y8uI'WDrU' Procedure Page 7 of 11 6.1.5 Independent verification should be performed as soon as practical after the associated task is performed, but can generally wait until completion of the task unless an adverse consequence could result (plant transient, loss of safety function, etc).
NOTE: The following step applies ONLY to Operations personnel that are restricted to the main control room.
6.1.6 When independent verification is specified for activities in the main control room, independence will be maintained to the extent practical (i.e.
(Le. verifier will not directly observe the performance of the step).
6.1.7 For restoration of systems which require IV, careful consideration must be given to the sequence of placing the affected components in service and restoration of the system to operable status. If desired to place a system in service prior to completion of the IV, a peer check should be used to verify critical components are properly aligned (this is to prevent damage to equipment, spilling of water, etc). The system should not be considered operable until completion of the IV.
6.1.8 Independent verification involves the following process:
6.1.8.1 The person performing the component manipulation enters the area, separated from the verifier by time and/or distance.
( 6.1.8.2 The positioner then references the lineup, procedure, tag out, or caution tag and verifies the proper component, using human performance tools such as STAR.
6.1.8.3 The positioner shall place (or check) the component in the required position per the lineup, procedure, tagout, tag out, or caution tag, as applicable.
6.1.8.4 The positioner signs or initials in the prescribed place.
6.1.8.5 The verifier enters the area, separated from the positioner performing the manipulation by time and/or distance.
6.1.8.6 The verifier references the lineup, procedure, tagout, or caution tag and verifies the correct component has been identified, using human performance tools such as STAR.
6.1.8.7 The verifier observes the position of the component and physically checks component position.
6.1.8.8 The verifier signs or initials in the prescribed place.
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUIHERNA SOUTHERN COMPANY A Management Verification Policy Version 3.0 EllewrQ~:NJ(! ){jAIl' EneW/o~r".e y",wl'w",/d' Wr'Jdd' Procedure Page 8 of 11 6.1.9 Independent Verification Methods 6.1.9.1 Direct Observation (preferred method) 6.1.9.1.1 Methods of performing direct observation for independent verification of valves or breakers include, but are not limited to, the following:
A. Visual observation of local breaker position indicating lights.
B. Visual observation of local breaker position indicating mechanical "flags."
C. Visual observation of breaker switch or handle position.
D. Manual valves to be independently verified open should be moved slightly in the closed direction and then moved in the open direction until the valve is considered in the fully open position, and, visual observation of the stem, i.e., grease markings indicating normal valve travel, valve stems extended on rising stem valves and mechanical position indication should also be included.
E. Valves required to be positioned slightly off "backseat" to prevent binding should be fully opened and returned to the procedurally established position during independent verification.
(
F. Manual valves to be independently verified closed should be moved, or attempted to be moved, only in the closed direction using normal closing torque and visually observing the stem. i.e., Grease markings indicating normal valve travel, valve stems inserted on rising stem valves, and mechanical position indication.
G. Visual observation and comparison with required stem position, local indicators, or other suitable valve component should be used to independently verify the position of throttled valves. Throttled valves shall not be moved to verify position unless specifically permitted to do so by the Shift Supervisor.
H. Control valve positions should be independently verified by ensuring that power or air, as appropriate, is available to the valve operators and that no physical obstructions which could prevent proper operation are apparent.
(
\
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN COMPANY A Management Verification Policy Version 3.0
£,urty fq&rw f:.'nUD ~Ul' WtJrl,r t#SU/1t lliul' W'tJr/,r Procedure Page 9 of 11 6.1.9.2 Indirect Observation 6.1.9.2.1 Equipment failures can cause incorrect remote position indicating lights on the main control board. For remotely operated equipment, verification can usually be accomplished from the control room using instruments, annunciators or valve position indications. It is highly desirable to perform initial and independent verification using diverse indications. Initial and independent verification from the control room is permitted using non-diverse methods if alternate control room indicators or methods of local verification of position are not available. Valve stems without stem indicators are not considered a diverse indication. While valve stem markings may provide some information regarding valve position, it is not to be relied on as a verified valve position for purposes of this procedure.
6.1.9.2.2 Problems may occur with remote indicating reach-rod valves, in which the remote indicator does not exactly duplicate the actual valve position. For important reach-rod valves, consider using a local verification of position when possible.
6.1.9.2.3 In some situations, a component's position can be determined by observation of process parameters such as pressure, flow, or voltage.
This, combined with a physical check of a component's position, can constitute an independent verification. However, exercise caution when using process parameters, because alternate flow paths or other factors
( could cause them to be misleading indicators of component position.
6.1.9.2.4 Methods of performing indirect independent verification of breakers, setpoints, and valves include, but are not limited to, the following:
A. Visual observation of remote indicating lights for breaker operation.
B. Visual observation of the actuation of status or indicating lights at the required panel-meter; indicated value, of an established setpoint.
- c. Visual observation of flow indicators, as applicable to opening or closing valves, and/or remote valve position indicating lights for valve position.
NOTE: Functional tests used in lieu of independent verification should be examined to ensure they test the entire portion of the system affected by the previous actions.
D. Functional surveillance tests may be used for indirect, independent verification only if plant safety is not compromised and the indications are positive and immediate (Le., annunciator changes status following an action). The Operations Manager must approve the use of functional testing to satisfy component verification requirements.
Compan~
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN COMPANY A Management Verification Policy Version 3.0 tl)&r~ Y6lit Enero 16&rw EllUl[! Ydur Wild"'"
w"u' Procedure Page P,!ge 10 of 11 6.2 Concurrent Verification 6.2.1 Concurrent Verification should be performed for:
A. Removing equipment from service when an action or manipulation could result in an unintended or undesirable condition B. Placement of electrical grounds.
C. Manipulation of valves, breakers, switches, jumpers, lifted wires, blind flanges, plugs, or any other components that, if improperly installed or mispositioned, could degrade a safety function or cause an unnecessary unit trip.
D. Operation of equipment necessary to support operation of important systems, such as electrohydraulic control, instrument air, redundant generator stator cooling water pumps or any other component that, if improperly installed or mispositioned, could degrade a safety function or cause an unnecessary unit trip.
E. Manipulation of valve positions in liquid or gaseous radioactive waste systems that if mispositioned could lead to unintended or unmonitored radioactivity release.
F. Other component positioning as determined necessary by the Operations Manager.
6.2.2 Concurrent verification involves the following process:
6.2.2.1 Both individuals involved determine, prior to the verification, who will fulfill the role of the person locating and performing the component manipulations and who will be the verifier of the component. The individuals must rigorously adhere to these roles.
6.2.2.2 The person performing the component manipulation references the lineup, procedure, tag out, or caution tag, locates the component and verbally identifies each unique identifier on the component label to the verifier.
6.2.2.3 The positioner verbalizes the position in which he or she intends to place (or check) the component.
6.2.2.4 tag out, or caution The verifier must independently read the lineup, procedure, tagout, tag. The verifier must verify that the correct component is to be manipulated, and verbalize his agreement.
6.2.2.5 The positioner places (or checks) the component in the intended position.
6.2.2.6 The verifier witnesses the positioning (or check) of the component and physically verifies component position, when applicable.
6.2.2.7 Both persons sign or initial in the prescribed place.
Southern Nuclear Operating Company Nuclear NMP-OS-002 SOUTHERN'\.
SOUTHERN COMPANY A Management Verification Policy Version 3.0 Energy t/j&r~ Your w.,rfr
£ntrgyllJ&rwYIJIII'WiJrlff' Procedure Page 11 of 11 6.3 Mispositioned Components Discovered During Verification NOTE: A component found out of position after all system alignments and verifications have been completed is considered a mispositioning.
A component found out of position during the verification phase is considered a near miss.
6.3.1 If, while performing verification, a component is found to be in a position other than required, the verifier will immediately notify the Shift Supervisor.
6.3.2 A component found out of desired position during the verification shall not be repositioned until the Shift Supervisor is notified and a verification of the mispositioned component is performed.
6.3.3 The Shift Supervisor will determine if the improper position of the component has caused any adverse system condition and if repositioning it to its correct alignment will result in an adverse condition.
6.3.4 If no adverse effects are noted or none could occur, the Shift Supervisor will direct that the component be properly positioned.
6.3.5 If any adverse condition exists or could occur, the affected system will first be placed
( in a safe condition where the component can be set to its correct position.
6.3.6 A Condition Report shall be written to document any mispositioning or near miss.
7.0 Records None 8.0 Commitments None
SECTION 9.2 APPENDIX B FIRE PROTECTION SURVEILLANCE REQUIREMENTS
(
(
Edwin I. Hatch Nuclear Plant Units 1 and 2 Fire Protection Program and Fire Hazards Analysis Appendix 8 Active Page Listing 9.2-8-1 (Rev 22) 9.2-8-42 (Rev 25) 9.2-8-2 (Rev 22) 9.2-8-43 (Rev 25) 9.2-8-3 (Rev 22) 9.2-8-44 (Rev 22) 9.2-8-4 (Rev 22) 9.2-8-45 (Rev 22) 9.2-8-5 (Rev 23A) 9.2-8-46 (Rev 25) 9.2-8-6 (Rev 22) 9.2-8-47 (Rev 25) 9.2-8-7 (Rev 23A) 9.2-8-48 (Rev 25) 9.2-8-8 (Rev 22) 9.2-8-49 (Rev 25) 9.2-8-9 (Rev 26A) 9.2-8-50 (Rev 25) 9.2-8-10 (Rev 22) 9.2-8-51 (Rev 25) 9.2-8-11 (Rev 22) 9.2-8-52 (Rev 22) 9.2-8-12 (Rev 22) 9.2-8-13 (Rev 22) 9.2-8-14 (Rev 22) 9.2-8-15 (Rev 23) 9.2-8-16 (Rev 22) 9.2-8-17 (Rev 22)
( 9.2-8-18 (Rev 22) 9.2-8-19 (Rev 22) 9.2-8-20 (Rev 22) 9.2-8-21 (Rev 22) 9.2-8-22 (Rev 22) 9.2-8-23 (Rev 22) 9.2-8-24 (Rev 26A) 9.2-8-25 (Rev 22) 9.2-8-26 (Rev 22) 9.2-8-27 (Rev 22) 9.2-8-28 (Rev 22) 9.2-8-29 (Rev 22) 9.2-8-30 (Rev 22) 9.2-8-31 (Rev 22) 9.2-8-32 (Rev 22) 9.2-8-33 (Rev 22) 9.2-8-34 (Rev 22) 9.2-8-35 (Rev 26) 9.2-8-36 (Rev 22) 9.2-8-37 (Rev 25) 9.2-8-38 (Rev 22) 9.2-8-39 (Rev 25) 9.2-8-40 (Rev 25) 9.2-8-41 (Rev 25)
FIRE PROTECTION EQUIPMENT OPERATING AND SURVEILLANCE REQUIREMENTS Fire protection systems are required to protect safety related or safe shutdown components from the effects of offire.
fire. Consistent with nuclear safety objectives, minimum operating requirements and surveillance requirements for these systems have been developed. These requirements, formerly embodied in the plant Technical Specifications, have been incorporated into this Appendix.
Definitions for the ACTION statement as used in this Appendix are provided below to ensure uniform and consistent interpretation of the Appendix is achieved. Regulatory separation required by 10 CFR 50 Appendix R protects at least one safe shutdown path to remain free of fire damage, thereby ensuring safe shutdown capability of the unites). The fire protection systems, equipment, and components ensuring safe shutdown capability are better refined and identified within respective tables as "CATEGORY I" (SSD) and "CATEGORY II" (Non-SSD),
aligned with their corresponding compensatory action.
As required under the provisions of Criterion 3 of 10 CFR 50 Appendix A and BTP 9.5-1 Appendix A, the fire protection systems, equipment, and components necessary to respond to fires without safe shutdown concern are included under a compensatory category designated as CATEGORY II (Non-SSD). Fire protection systems and components are included in the following paragraphs and tables. These fire protection systems, equipment, and components, considered important to safety but not associated with safe shutdown function, provide
( compliance to 10 CFR 50 Appendix A, Criteria 3 and 5 and BTP 9.5-1 Appendix A. The defense in depth features are aligned to the CATEGORY II compensatory action during impairment periods.
This strategy is continued throughout FHA section 9.2 (Appendix B) to ensure the graded response is applied consistently for those systems and components during periods of planned or unplanned impairments or inoperable condition.
Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension of25% of the surveillance interval; this 25% extension is known as the grace period.
9.2-B-1 9.2-B-l RevO 06/86,Rev4 Rev 0 06/86, 07/89,RevlOD Rev 4 07/89, Rev IOD 07/95, Rev lID 07/96, 07/96,RevI2B Rev l2B 7/97, Rev 22 9/04
FIRE-RATED ASSEMBLIES REQUIREMENTS OPERATING REOUIREMENTS 1.1.1 Fire-rated assemblies and sealing devices in fire-rated assembly penetrations separating fire areas or separating portions of redundant systems important to safe shutdown within a fire area shall be OPERABLE. Fire-rated assemblies are walls, floor/ceilings, cable tray enclosures and other fire barriers. Sealing devices in fire-rated assembly penetrations consist of fire doors; fire dampers; and cable, piping, and ventilation duct penetration seals. Tables 1.1-1 and 1.1-2 contain the Unit 1 and Unit 2 fire door listings to which this Specification applies.
APPLICABILITY: When fuel is in the reactor vessel for the affected unit.
ACTION:
With one or more of the above required fire-rated assemblies and/or sealing devices inoperable or with the required surveillance interval (including grace period) exceeded, within 1 hour:
CATEGORY I:
Establish a continuous fire watch on at least one side of the affected fire rated assembly
( and/or sealing device(s) or verify the OPERABILITY of fire detectors on at least one side of the inoperable assembly(s) and sealing device(s), and establish an hourly fire watch patrol and notify onshift fire brigade leader.
CATEGORY II:
Assign a tracking Fire Action Statement (F (FAS)
AS) for the affected fire rated assembly and/or sealing device(s). Ifnot If not returned to operable status within 45 days, establish an hourly fire watch patrol and notify onshift fire brigade leader.
SURVEILLANCE REQUIREMENTS 2.1.1 Each of the above required fire-rated assemblies and penetration sealing devices shall be verified OPERABLE at least once per 24 months by performing a visual inspection of:
- a. The exposed surfaces of each fire-rated assembly.
- b. Each fire damper and associated hardware.
9.2-B-2 Rev 0 06/86, Rev 3 07/88, Rev 4 07/89, Rev 6B 01191, 0l/91, Rev 8D 07/93, Rev 10D 07/95, Rev lID liD 07/96, Rev 12B 7/97, Rev 217/03, 21 7/03, Rev 22 9/04
(
SURVEILLANCE REQUIREMENTS (Continued)
- c. At least 10 percent of each type of sealed penetration. If apparent changes in appearance abnormal degradations are found, a visual inspection of an additional 110 or abnonnal 0 percent of each type of sealed penetration shall be made. This inspection process shall continue until a 10-percent lO-percent sample with no apparent changes in appearance or abnonnal abnormal degradation is found. Samples shall be selected such that each penetration seal will be inspected at least once per 15 years.
2.1.2 Each of the required fire doors (i.e., the doors in Tables 1.1-1 and 1.1-2) shall be verified QPERABLE by:
- a. Verifying that each locked-closed fire door is closed at least once per 7 days.
- b. Verifying that doors with automatic hold-open and release mechanisms are free of obstructions at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and by perfonning performing a functional test of these mechanisms at least once per 18 months.
- c. Verifying that each unlocked fire door without electrical supervision is closed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- d. Inspecting the automatic hold-open, release and closing mechanism and latches at least once per 6 months.
(
9.2-B-3 RevO 06/86, Rev 2 07/87,Rev4 07/87, Rev 4 07/89,Rev6B 01l91,RevllA 01/91, Rev IIA 10/95,RevI2B 10/95, Rev 12B 7/97, Rev 22 9/04
FIRE DOOR TABLES LEGEND DOOR NUMBER:
1 C 38 I
UNIQUE DOOR NUMBER BLDG DESIGNATION: C ...... CONTROL BLDG D ...... DIESEL GEN BLDG R. .....REACTOR BLDG RADWASTE BLDG RW .. RADWASTE T........
T .... TVRBINE TURBINE BLDG UNIT NUMBER: O.*. ... COMMON 1...UNIT 1 1...UNIT UNIT 2 2 ...UNIT2 COMPENSATORY ACTION:
- "Category I": Safe Shutdown Capability Compensatory Action
- "Category II": Non-Safe Shutdown Associated Compensatory Action DOOR TYPE:
EX...... EXCEPTION EX ROLL UP RU ...ROLLUP
( SL.. .. SLIDING SW ... SWINGING DOOR STAT:
AHO ...... AUTOMATIC HOLD OPEN LKCL. ... LOCKED-CLOSED UNLK. ..UNLOCKED DOOROPER:
AIRL......
AIRL. ... LOCKED BY AIRLOCK MECHANISM CARD ...REQUIRES BADGE CARD FOR ENTRY FL...........
FL. ........HELD OPEN BY FUSIBLE LINK KEY.......
KEY .....REQUIRES KEY FOR ENTRY MAG .....HELD OPEN BY MAGNETIC MECHANISM AREA NO.1 / AREA NO.2: THE FIRE AREAS THAT THE DOOR SEPARATES
- Fire Areas designated "Dominant Fire Risk" by the IPEEE Risk Analysis DETECTION ZONES: XL DETECTION ZONES FOR APPLICABLE FIRE AREAS ELEV: FLOOR ELEVATION ON WHICH THE DOOR IS LOCATED DWG: FHA DRAWING (FHA SECTION 8.0) ON WHICH THE DOOR CAN BE FOUND LOCATION: BRIEF DESCRIPTION OF DOOR LOCATION 9.2-B-4 Rev 0 06/86, Rev 1 10/86, 10186, Rev 3 07/88, Rev 4 07/89, Rev 6B 01191, Rev 6C 07/91, Rev 12B 7/97, Rev 22 9/04 9104
TABLE 1.1-1 (SHEET 1I OF 4)
UNIT 1I AND COMMON FIRE DOORS FIRE ACTION DOOR DOOR DOOR AREA DETECTION AREA DETECTION FHA DOOR CATEGORY TYPE STATUS OPER. No.1 ZONE No. 1I No.2 ZONE No. 2 ELEV. DRAWING LoCATION LOCATION 1COI 1C01 I SW UNLK 0007A 1Z43112D14 1006 None 112 H-11811 U1 UI WATER ANALYSIS ROOM 1C03 IC03 II SW AHO MAG 0007A 1Z43112D12 IZ43112D12 0001 2Z43112D12 2Z43112DI2 112 H-11811 EAST CORRIDOR 1C06 IC06 I EX LKCL 1005 1Z43112D04 IZ43112D04 0001 2Z43112D12 2Z43112DI2 112 H-11811 STATION BATTERY ROOM 1B IB 1C07 IC07 II SW UNLK 1010 IZ43112D06 1Z43112D06 0001 2Z43112D12 2Z43112DI2 112 H-11811 RPSBATTERyN RPS BATTERY N ROOM lC08 IC08 II SW UNLK 1009 lZ43112D06 IZ43112D06 0001 2Z43112D12 112 H-11811 RPS BATTERY S ROOM lC09 IC09 II SW UNLK 2009 2Z43112D09 0001 2Z43112D12 2Z43112DI2 112 H-I1811 H-11811 RPS BATTERY BATTERyNN ROOM lClO 1C1O II SW UNLK 2010 2Z43112D09 0001 2Z43112D12 112 H-11811 RPS BATTERY S ROOM lCll ICII I SW UNLK 1008 lZ43112DOS IZ43112D05 0001 2Z43112D12 112 H-11811 UNIT 1I AC INVERTER ROOM lC12 ICI2 II EX LKCL 1004* 1Z43112D03 IZ43112D03 0001 1Z43112D01 IZ43112DOI 112 H-11812 STATION BATTERY ROOM 1A IA 1C13 II SL AHO FL 1003 1Z43112W02 0001 lZ431l2DOl 1Z43112D01 112 H-118l2 H-11812 OIL STORAGE TANK ROOM lC13A ICI3A II SL AHO FL 1003 lZ43112W02 IZ43112W02 0001 1Z43112D01 I Z43 II 2DOI 112 H-11812 OIL STORAGE TANK ROOM IC14 ICI4 II SW UNLK 0002A NONE 0001 IZ43112DOI 112 H-11812 CONTROL BLDG STAIRWELL ICI5 ICIS II SW UNLK 21011 2101I NONE 0001 1Z43II2D01 IZ43112DOI 112 H-118l2 H-11812 WEST CABLEWAY CAB LEW AY ICI7 1C17 II SW UNLK 2008 IZ43112D08 1Z43112D08 0001 2Z43112D12 112 H-11811 UNIT 2 AC INVERTER ROOM IC21 II SW UNLK 2104* 2U43130D02 0014B IZ43130D09 130 H-IJ814 H-11814 UNIT 2 - EAST CABLEWAY CABLEWAY IC22 1C22 I SW AHO MAG 1105* IU43130D02 1U43130D02 0014K IZ43130D09 130 H-IJ814 H-11814 EASTCABLEWAY EAST CABLEWAY 1C29 lC29 II SL AHO FL 1020 lZ43130D08 IZ43I30D08 0014K IZ43130D02 1Z43130D02 130 H-IJ814 H-11814 EAST DCDCSWGRROOM 1B SWGR ROOM IB 1C31 IC31 II SL AHO FL 1017 IZ43130DOS 1Z43130D05 0014K IZ43130D02 130 H-IJ814 H-11814 EAST 600VOLT SWGRROOM SWGR ROOM lD ID IC3S 1C35 II SL AHO FL 1016 IZ43130D04 1Z43130D04 0014K IZ43I30D02 IZ43130D02 130 H-IJ814 H-II814 WEST 600VOLT SWGRROOM IC lC38 IC38 II SL AHO FL 1018 lZ43130D06 IZ43130D06 0014K IZ43130D02 130 H-11814 WEST DC SWGR ROOM IA IC46 II SL AHO FL 1015* IZ43I30DIO IZ43130D1O 0014K IZ43130D02 1Z43130D02 130 H-IJ8IS H-11815 UNIT 1I - ANNUNCIATOR ROOM IC47 II SL AHO FL 2015 2Z43130DI5 2Z43I30DIS 0014K 2Z43I30D17 2Z43130D17 130 H-11815 H-118lS ANNUNCIATOR ROOM IC48 II SW UNLK 10I3 1013 lZ43I30DIJ I Z43130DI1 00 14K 0014K 2Z43l30DI7 2Z43130D17 130 H-IJ8IS H-11815 RPSMGSETRooM RPS MG SET ROOM IC49 lC49 I SW UNLK 2013 2Z43130DI4 2Z43130D14 0014K 2Z43I30D17 2Z43130D17 130 H-11815 H-ll8lS RPS MG SET ROOM ICSO IC50 II SW UNLK 0014K IZ43130D02 0002A NONE 130 H-IJ81S H-11815 CB STAIRWELL/130' STAIRWELL! 130' ELEVATION ICS2 1C52 II SW AHO MAG SRVBG NONE 0014K IZ43130D02 1Z43130D02 130 H-IJ81S H-11815 SERVICE BLDG/ CONTROL BLDG lCS3 1C53 II SL AHO FL 1023 IZ43130W03 0014K 1Z43130D02 IZ43130D02 130 H-1181S H-11815 OIL CONDITIONER ROOM lCS3A IC53A II SL AHO FL 1023 IZ43130W03 0014K 1Z43130D02 IZ43130D02 130 H-IJ81S H-11815 OIL CONDITIONER ROOM 1CS4 IC54 II SL AHO FL 1019 1Z43130D07 IZ43130D07 0014K 1Z43130D02 IZ43130D02 130 H-11814 __ TRANSFORMER ROOM L1:itANSFORMER 9.2-B-5 Rev 3 07/88, Rev 4 07/89, Rev 6B 01/91, 01191, Rev 6C 07/91, Rev 9C 04/94, Rev lOA 1O/94,Rev 10/94,Rev lOB 01/95, lID 07/96, Rev 12B 7/97, 01195, Rev liD 7197, Rev 22 9/04, Rev 23A 5/06
,/'~'
TABLE 1.1-1 (SHEET 2 OF OF44)
FIRE ACTION DOOR DOOR DOOR AREA DETECTION AREA DETECTION FHA DOOR CATEGORY TYPE STATUS OPER. No.1 ZONE No. I No.2 ZONE No. 2 ELEV. DRAWING LOCATION LoCATION IC58 II SW UNLK 0014K IZ43130D02 0002A NONE 130 H-1l815 H-118I5 CB STAIRWELL! 130' ELEVATION IC60 I SW AHO MAG 2104* 2U43130D02 1105* IU43130D02 1U43130D02 130 H-1I814 UNIT 2 - EAST CABLEWA CABLEWAY Y IC61 I SW LKCL CARD 0025 NONE 0024A* IZ43147D04 147 H-11816 CABLE SPREADING ROOM IC62 I SW UNLK 0025 NONE 0024B IZ43147D06 IZ43 I 47D06 147 H-11816 COMPUTER ROOM IC63 II SW UNLK 0025 NONE 0002A NONE 147 H-1l816 H-11816 CB STAIRWELL! 147' ELEVATION IC64 I SW LKCL CARD 0028 lZ43 I 47D08 IZ43147D08 0024A* IZ43147D04 147 H-11816 LPCI INVERTER ROOM IC66 I RU AHO FL OlOlG OIOIG IZ43164D02 IZ43 I 64D02 0024C* IZ43164DOI 164 H-11817 CHART STORERM & HALLWAY lC71 IC71 II SW UNLK 010lA OIOIA NONE 0002A NONE 164 H-1l817 H-11817 CB STAIRWELL STAIRWELL!UI
/ Ul TuRBINE DECK lC82 IC82 II SW UNLK OIOIA NONE 0024D NONE 164 H-1l817 H-11817 MAIN CONTROL ROOM ENTRYWA ENTR YWA Y lC83 IC83 II SW LKCL KEY 0031 NONE 0002B NONE 186 H-11818 MAIN CONTROL RM ROOF IC84 II SW UNLK 0002A NONE 0002B NONE 186 H-11818 MAIN CONTROL RM ROOF IC85 II SW UNLK 0031 NONE 0002A NONE 180 H-11818 MAIN CONTROL RM ROOF IC86 II RU AHO FL 010lA OIOIA NONE 0024C* IZ43130DOI 164 H-11817 CONTROL ROOM / UI U 1 TuRBINE DECK IC87 II RU AHO FL OIOlJ 0101J NONE 0024C* IZ43164DOI 164 H-11817 MAIN CONTROL / U2 TuRBINE DECK lC88 IC88 II SW UNLK 2013 2Z43130D14 2Z43130DI4 0040* IZ43130D13 130 H-11815 UNIT 2 - RPS MG SET ROOM lC89 IC89 II SW UNLK 1013 IZ43130Dl1 IZ43130DII 0040* IZ43130D13 130 H-11815 UNIT I - RPS MG SET ROOM IC97 II SW LKCL KEY OIOIA NONE 0024D NONE 164 H-11817 MAIN CONTROL ROOM ENTRYWA ENTRYWAY Y ICI60 II SW AHO MAG 1104* 1U43130D02 1105* 1U43130D02 130 H-11814 EASTCABLEWAY EAST CAB LEWAY 10134 lDI34 I RU AHO FL 2403 2X43130C05 0401 NONE 130 H-11846 DIESEL GENERATOR GENERA TOR RM 2A 10135 lD135 I SW UNLK 2403 2X43130C05 0401 NONE 130 H-11846 DIESEL GENERA TOR RM 2A 10136 lD136 II SW UNLK 2401 2X43130C05 0401 NONE 130 H-1l846 H-11846 DGDAY DG DAY TANK ROOM 2A TANKRoOM2A 10137
!D137 II RU AHO FL 2407 2X43130C06 0401 NONE 130 H-11846 DIESEL GENERA GENERATOR TOR RM 2C 10138 lD138 I SW UNLK 2407 2X43130C06 0401 NONE 130 H-11846 GENERATOR DIESEL GENERA TOR RM 2C 10139 lD139 I SW UNLK 2405 2X43130C06 0401 NONE 130 H-11846 DGDAyTANKRoOM 2C lDI40 10140 I RU AHO FL 1411 IX43130C02 0401 NONE 130 H-11846 DIESELGENERATORRM DIESEL GENERA TOR RM IA IA 10141 lDI41 I SW UNLK 1411 IX43130C02 0401 NONE 130 H-1l846 H-11846 DIESEL GENERATOR RM IA 10142 lDI42 II SW UNLK 1409 IX43130C02 0401 NONE 130 H-11846 DGDAyTANKRoOM IA 10143
!D143 II RU AHO FL 1407 IX43130C03 0401 NONE 130 H-11846 DIESEL GENERATOR RM 1B I B (SWING DG) 10144
!D144 II SW UNLK 1407 IX43130C03 0401 NONE 130 H-11846 GENERA TOR RM IIB DIESEL GENERATOR B (S WING DG)
(SWING 10145 lDI45 II SW UNLK 1405 IX43130C03 0401 NONE 130 H-11846 DGDAyTANKRoOM IB (SwINGDG) 10146 IDl46 II RU AHO FL 1403 IX43130C04 0401 NONE 130 H-11846 H-I1846 DIESEL GENERATOR RM I C 10147 lDI47 I SW UNLK 1403 IX43130C04 0401 NONE 130 H-1I846 H-11846 DIESEL GENERATOR RM IIC C 9.2-B-6 Rev 3 07/88, Rev 6C 07/91, Rev lOA 10/94, Rev lOB JOA 10194, 01195. Rev lID 07/96, Rev 12B 7/97, Rev. 14A 1/99, JOB 01/95, 1199, Rev 22 9104 9/04
TABLE 1.1-1 (SHEET (SHEET33 OF 4)
FIRE DOOR ACTION CATEGORY DOOR TYPE DOOR DOOR STATUS I OPER.
DOOR AREA No.
No.1I DETECTION ZONE No. I ZONENo.
AREA No.2 DETECTION ZONE No. 2 ELEV.
FHA DRAWING LoCATION 10148 I1 SW UNLK 1401 lX43130C04 IX43130C04 0401 NONE 130 H-11846 DO DAY TANK ROOM lC DGDAY IC 10149 II SW UNLK 2404* 2X43130CIO 2403 2X43130C05 130 H-11846 DO DG SWGR ROOM 2E 10150 II SW LKCL 2408* 2X43130CII 2402 2X43130CIl 2X43130CII 130 H-Il846 H-11846 DO DG BATTERY ROOM 2F 10151 I1 SW UNLK 2408* 2X43130CII 2X43130C11 2407 2X43130C06 130 H-II846 H-11846 DO DG SWGR ROOM 2F 10152 II SW LKCL 2409* 2X43130C12 2406 2X43130C12 130 H-11846 DO DG BATTERY ROOM 20 2G 10153 1 SW UNLK 2409* 2X43130C12 1411 IX43130C02 130 H-Il846 H-11846 DO DG SWGR ROOM 20 2G 10154 II SW UNLK 1412* IX43130C07 14II 1411 IX43130C02 130 H-II846 H-11846 DOSWGRRoOM DGSWGRRoOM IE 10155 II SW LKCL 1412* IX43130C07 1410 IX43130C07 130 H-Il846 H-11846 DO DG BATTERY ROOM IE 10156 II SW LKCL 1408* lX43130C08 IX43130C08 1407 lX43130C03 IX43130C03 130 H-Il846 H-11846 DO DG BATTERY ROOM IF 10157 II SW UNLK 1408* IX43130C08 1406 lX43130C08 IX43130C08 130 H-11846 DOSWGRRooM DGSWGRRoOM IF 10158 II SW UNLK 1404* IX43130C09 1403 IX43130C04 130 H-11846 DO DG SWGR ROOM IG IG 10159 II SW LKCL 1404* lX43130C09 IX43130C09 1402 IX43130C09 130 H-11846 DO DG BATTERY ROOM 10 IG 10214 II RU AHO FL 0702B IX43130W13 0703 IX43130W13 130 H-11848 FIRE PROTECTION PUMP HOUSE 10215 II RU AHO FL 0703 lX43130W13 IX43130W13 0704 IX43130W13 lX43130W13 130 H-11848 FIRE PROTECTION PUMP HOUSE IR28 1 SW LKCL AIRL 1203C IT43130D02 1105* 1U43130D02 IU43130D02 130 H-11814 EASTCABLEWAY lUI RB IR35 II SW UNLK 1211 IT43158WOI 1203K NONE 158 H-11828 EAST RECIRC MOMG SET B lR36 IR36 II SW UNLK 1211 IT43158WOI IT43158WOl 1203K NONE 158 H-11828 EAST RECIRC MOMG SET B IR37 II SW UNLK 1210 IT43158W02 1203K NONE 158 H-11828 WEST RECIRC MOMG SET A IR38 II SW UNLK 1210 IT43158W02 1203K NONE 158 H-11828 WEST RECIRC MOMG SET A IR40B 1 SW UNLK 1205N IT43 I 64D02 IT43164D02 12031 NONE 164 H-11828 HVAC ROOM EL 164' HVACRoOMEL IR41 II SW LKCL AIRL 12031 NONE OIOIA NONE 164 H-11828 Ul UI RB STAIRWELLlUl STAlRWELLlUI TB lR42 IR42 1 SW LKCL AIRL 22031 NONE 12031 NONE 164 H-11828 UI RB STAIRWELLlU2 TB VIRB IR51 1 SW UNLK 1205U IT43 I 85W03 IT43185W03 12031 NONE 185 H-11829 UI RB STAIRWELL I SW CORNER VI IR52A 1 SW LKCL AIRL 22031 NONE 12031 NONE 185 H-II829 H-11829 VIRB UI RB STAIRWELLlU2RB STAIRWELL IU2 RB IR61 1 SW UNLK 1205X NONE 12031 NONE 203 H-11830 STACK MONITORING ROOM IR64 1 SW LKCL AIRL 12031 NONE 020lA NONE 228 H-II831 H-11831 VI UI RB STAIRWELL EL 228' IRWOl IRWOI II SW UNLK 1101H 110lH NONE 130IG 1301G NONE 112 H-11805 UNIT I RADw ASTE EAST CORRIDOR ENTRY RADWASTE lRW21 IRW21 II SW UNLK 1301I 130 II NONE 1104* IU43130D02 1U43130D02 132 H-11839 UI RWBLDG/Ul Ul TuRBINE BLDG.
RWBLDG/UI Tt!RBINEBLDG.
IRW30 1 SW LKCL AIRL 130lJ NONE 1205F* IT43130DIO IT43130010 132 H-11839 UI RWBLDG/Ul Ul RWBLDG/UI REACTOR BLDG.
lT07 II SW UNLK 110IC 110lC NONE 1102 NONE 112 Il2 H-Il804 H-11804 UNIT 1I TuRBINE BLDG EAST STAIRWELL ITlO II SW UNLK 110lH II0lH NONE 1103 NONE 112 H-11804 UNIT I TuRBINE BLDG NORTH STAIRWELL ITlI II SW UNLK IlOlH II0lH NONE 0007A lZ43112D14 IZ43112014 Il2 112 H-11804 CORRlDOR CONTROL BLDG EAST CORRIDOR ITl7 II SW UNLK 110lJ II0lJ lU43130D05 1U43130D05 1102 NONE 130 H-1l805 H-11805 UNIT 1I TuRBINE BLDG EAST STAIRWELL 9.2-B-7 01/91, Rev 6C 07/91, Rev lOA 10/94, Rev lID 07196, Rev 3 07/88, Rev 4 07/89, Rev 6B 01191, 07/96, Rev 12B 7/97, Rev 22 9/04, Rev 23A 5/06
TABLE 1.1-1 (SHEET 4.......
~ OF 4)~ .
FIRE ACTION DOOR DOOR DOOR AREA DETECTION AREA DETECTION FHA DOOR CATEGORY TYPE STATUS OPER. No.1I No. ZONENo.
ZONE No. I1 No.2 ZONE No. 2 ELEV. DRAWING LoCATION ITl8
!T18 II SW UNLK II0lJ 1U43130D05 1102 NONE 130 H-11805 UNIT 1I TuRBINE BLDG NORTH STAIRWELL ITI9 ITl9 II SW UNLK 1104* 1U43130D02 1103 NONE 130 H-11805 EASTCABLEWAY INE STAIRWAY IT23 II SW UNLK 110lM 1l01M 1U43147DOI 1U43147DOl 1103 NONE 147 H-11806 UNIT 1I TuRBINE BWG BLDG EAST STAIRWELL IT24 II SW UNLK 1I0lN IIOIN 1U43I 47D03 1102 NONE 147 H-11806 UNIT I TuRBINE BLDG NORTH STAIRWELL IT33 II SW UNLK 1103 NONE OIOIA NONE 164 H-11807 UNIT I TuRBINE BLDG EAST STAIRWELL IT34 II SW UNLK 1102 NONE OIOIA NONE 164 HI1807 UNIT I TuRBINE BLDG NORTH STAIRWELL 9.2-B-8 Rev 6C 07/91, 07/9\, Rev 8B 01/93, Rev lOA 10/94, Rev lOB 01/95, Rev 12B
\2B 7/97, Rev 22 9/04
~..
TABLE 1.1-2 (SHEET 1I OF 2)
UNIT 2 FIRE DOORS FIRE ACTION DOOR DOOR DOOR AREA DETECTION AREA DETECTION FHA DOOR CATEGORY TYPE STATUS OPER. No.1 ZONE No. 1I No.2 ZONE No. 2 ELEV. DRAWING LocATION LOCATION 2C01 2COI II SW UNLK 2006 NONE 0007A 1Z43112D14 IZ43112DI4 112 H-11811 U2 WATER ANALYSIS ROOM 2C02 II EX LKCL 2005 2Z43112D11 2Z43112DII 0001 2Z43112D12 112 H-11811 H-II811 BAITERY ROOM 2B STATION BATTERY 2C03 II EX LKCL 2004* 2Z43112D1O 2Z43112DIO 0001 2Z43112D12 2Z43112DI2 112 H-11811 H-118II STATION BATTERY BAITERY ROOM 2A 2COS 2C05 II SL AHO FL 2003 2Z43112W13 0001 1Z43112D01 IZ43112DOI 112 H-11812 OIL STORAGE TANK ROOM 2C05A 2COSA II SL AHO FL 2003 2Z43112W13 2Z43112WJ3 0001 IZ43112DOI 1Z43112D01 112 H-11812 OIL STORAGE TANK ROOM 2C06 II SL AHO FL 2020 2Z43130D22 2014 2Z43130D17 2Z43J30Dl7 130 H-11814 EAST DC SWGR ROOM 2B 2C07 2e07 II SL AHO FL 2017 2Z43130D19 2014 2Z43130D17 130 H-11814 EAST 600V SWGR ROOM 2B 2C08 II SL AHO FL 2019 2Z43130D21 2014 2Z43130D17 2Z43I30DI7 130 H-11814 TRANSFORMER ROOM 2C09 I SL AHO FL 2016* 2Z43130D18 2Z43130DI8 2014 2Z43130D17 130 H-11814 WEST 600V SWGR ROOM 2C 2C1O 2CIO I SL AHO FL 2018 2Z43130D20 2014 2Z43I30Dl7 2Z43130D17 130 H-11814 WEST DC SWGR ROOM 2C 2C11 2CII II SW UNLK 2023 2Z43130W23 0014K 2Z43130D17 130 H-1181S H-11815 OIL CONDITIONER ROOM 2C33 II SW UNLK 2021 2Z43130D30 2014 2Z43130D17 2Z43I30D17 130 H-11814 AC DISTRIBUTION ENCWSURE ENCLOSURE 2C34 II SW UNLK 2021 2Z43130030 2Z43130D30 2014 2Z43130D17 2Z43I30D17 130 H-11814 AC DISTRIBUTION ENCWSURE ENCLOSURE 2R23 II SW LKCL AIRL 2203F* 2T43130D02 2104* 2U43130D02 2U43I30D02 130 H-11833 RB/u2 EAST CABLEWAY U2 RBlU2 CAB LEWAY !
2R26 II SW LKCL AIRL 230lJ 23011 NONE 2205F* 2T43130D04 130 H-11843 DRY WASTE STORAGE AREA I 2R32 I SW UNLK 220SN 2205N 2T43164D02 22031 NONE 164 H-11834 ClULLER ClllLLER ROOM 2R52 II SW UNLK 2205T NONE 22031 NONE 185 H-11835 RB EXHAUST FILTER ROOM 2R53 II SW UNLK 2205U 2T43185W05 2T43185WOS 22031 NONE 185 H-11835 U2 RB STAIRWELUNW CORNER 2R61 I SW UNLK 2205V NONE 22031 NONE 203 H-11836 U2RBSTAIRWELIiFLTR U2 RB STAIRWELUFLTR 2R63 II SW UNLK 2205X NONE 22031 NONE 203 H-11836 U2 RB STAIRWEWNW STAIRWELUNW CORNER 2R71 II SW UNLK 22031 NONE 0201B NONE 228 H-11837 U2 RB STAIRWELL EL 228' 2RW02 II SW UNLK 210lH NONE 2301A NONE 112 H-11820 EAST CORRIDOR 112 ELRAowASTE EL RADWASTE 2RW53 II SW UNLK 01011 010lJ NONE 2301 NONE 164 H-11823 U2 TBDECK/RAowASTEENTRY TB DECK I RADWASTE ENTRY 164' EL.
2RW57 II SW UNLK 01011 OlOlJ NONE 2301 NONE 164 H-11823 U2 TB DECK I RAoWASTE RADWASTE ENTRY 164' EL.
2T01 2TOI II SW UNLK 210lH NONE 2103 NONE 112 H-11820 EAST CORRIDOR I U2 TuRBINE EAST STAIR 9.2-B-9 l2B 7/97, Rev 22 9/04, Rev 23A 5106, Rev 0 06/86, Rev 3 07/88, Rev 4 07/89, Rev 7C 07/92, Rev 10D 07/95, Rev lID 07/96, Rev 12B 5/06, Rev 26A 3/09
TABLE 1.1-2 (SHEET ... --2 OF 2)./
I TYPE
~. ~
FIRE ACTION DOOR DOOR DOOR DOOR AREA DETECTION AREA DETECTION FHA DOOR !:;ATEGORY CATEGORY TYPE STATUS OPER. No.1 ZONE No. 1 No.2 ZQN!lNo.2 ZONE No. 2 ELEV. DRAWING LOCATION 2T02 II SW UNLK 2101C NONE 2102 NONE 112 H-11820 U2 TuRBINE BLDG SOUTH STAIR 2T19 II SW UNLK 2104* 2U43130D02 2103 NONE 130 H-11821 EASTCABLEWAY IISESTAIR SE STAIR 2T23 II SW UNLK 210lJ 2T43130D05 2102 NONE 130 H-11821 WORKING FLOOR II SOUTH STAIR 2T29 II SW UNLK 2101N 2U43147DOI 2U43147D01 2102 NONE 147 H-11822 U2 TB SOUTH STAIR/U2 TB SWGR 147' EL..
2T30 II SW UNLK 2101M 2U43147D03 2103 NONE 147 H-11822 U2 TB EAST STAIR I U2 TB SWGR 147' EL.
STAIR/U2 2T40 II SW UNLK OlOlJ 010lJ NONE 2102 NONE 164 H-11823 U2 TB SOUTHSTAIR/U2 TB DECK 164' EL.
2T41 II SW UNLK 010lJ NONE 2103 NONE 164 H-11823 U2 TB EAST STAIR/U2 STAIR I U2 TB DECK 164' EL.
Rev 3 07/88, Rev 4 07/89, Rev 7C 07/92, Rev IOD 07/95, Rev 12B 7/97, Rev 22 9/04 9104 9.2-B-IO
/
DRAFT
(
Southern Nuclear E. I. Hatch Nuclear Plant Operations Training JPM Admin 5, ALL TITLE EVALUATE AN RWP AND SURVEY MAP
( AUTHOR MEDIA NUMBER TIME FRANK FAGAN LR-JP-I0030-00 10 Minutes iEc(jMMJ!jl§))i~'ll~r RECOMMENDED BY .. APPROVED BY DATE NIR sM Energy to Serve Your World sM
(
SOUTHERN NUCLEAR OPERATING COMPANY PLANT E. I. HATCH of 1 Page 1 of1 FORM TITLE: TRAINING MATERIAL REVISION SHEET Program/Course Code: OPERATIONS TRAINING Media Number: LR-JP-10030 LR-JP-I0030 Rev. No. Date Reason for Revision Author's Supv's Initials Initials
(
LR-JP-I0030-00 LR-JP-10030-00 UNIT 1 ( X) UNIT 2 (.)
TASK TITLE: Comply with radiation work permit requirements during normal or abnormal conditions.
JPMNUMBER: LR-JP-I0030-00 TASK STANDARD: The task shall be completed when the operator has determined:
location of a valve, dress out requirements, time before dosimetry alarm occurs, actions if an alarm occurs and when a brief is required.
TASK NUMBER: N/A OBJECTIVE NUMBER: NIN/AA TYPE N/A PLANT HATCH JTA IMPORTANCE RATING:
RO N/A c(
SRO N/A KIA CATALOG NUMBER: G2.3.7 KIA CATALOG JTA IMPORTANCE RATING:
RO 3.5 SRO 3.6 OPERATOR APPLICABILITY: Nuclear Plant Operator (NPO)
I*GENERAL
REFERENCES:
Unit 2 R WP 09-0004 for Operations RWP V7-,-"'VV""
HP survey N.E. Diagonal HP survey S.E. Diagonal 60AC-HPX-004-0, Radiation & Contamination Control 60AC-HPX-002, Personnel Dosimetry 34S0-EII-0I0-2, Attachment 3 34S0-EII-01O-2 IREQUIRED MATERIALS: Unit 2 R WP 09-0004 for Operations HP survey N.B.N.E. Diagonal HP survey S.E. Diagonal 60AC-HPX-004-0, Radiation & Contamination Control 60AC-HPX-002, Personnel Dosimetry 34S0-EII-0I0-2, Attachment 3 34S0-EII-Ol
( APPROXIMATE COMPLETION TIME: 10 Minutes SIMULATOR SETUP: NI N/AA
UNIT 1 READ AND GIVE A COPY TO THE OPERATOR INITIAL CONDITIONS:
1 Unit 1 is at 100% power with no significant problems.
- 2. Health Physics reports a seal leak on lEll-C002B, RHR Pump B.
- 3. performing nonnal The Shift Supervisor wants the rounds operator, while perfonning normal round activities, to report back on the extent of the packing leak.
- 5. The RRWP WP and HP Survey Maps are available.
- 5. HP approval has been granted to use minimum requirements for entry and inspection of the leak.
( INITIATING CUES:
You are to do a pre-job brief with the rounds operator and provide the information:
following infonnation:
- What are the minimum dress requirements for entry.
- Assuming the highest current General Area Dose Rate, calculate the maximum stay time before the DAD alarms on dose accumulated.
- Required actions if the DAD alarms on dose accumulated.
- determine the minimum Assuming plant conditions change, detennine General Area Dose Rate which would REQUIRE an HP brief prior to entry.
(
LR-JP-10030-00 PERFO~ANCESTEP STANDARD SATJUNSAT COMMENTS For INITIAL Operator Programs:
For OJT/OJE; ALL PROCEDURE STEPS must be completed for Satisfactory Performance.
For License Examinations; ALL CRITICAL STEPS must be completed for Satisfactory Performance.
START TIME:_ _ __
- 1. Determine the appropriate survey Operator determines that survey SAT I UNSAT SAT/UNSAT map. map 57769, U2 D.E. Diag. 87 is the correct map.
NOTE: If operator selects the incorrect map, Critical Step #3 will be incorrectly calculated.
- 2. Determine the minimum dress Operator determines that lab coat, SAT I UNSAT SAT/UNSAT
( requirements for entry. booties, and gloves are required when entering this area NOTE: An answer of "Full Dress" makes the step UNSAT but no longer Critical (i.e. Full Dress is more conservative).
- 3. Determine the maximum stay time Operator determines the max stay SAT/UNSAT before the DAD alarms on dose time is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 15 minutes.
accumulated.
NOTE: Per RWP, DAD set at 25 mr for rounds. Per Survey Map, the max general area dose rate is 20 mr/hr.
- 5. Determine the minimum General Area Operator determines that> .1 SAT I UNSAT SAT/UNSAT Dose Rate which would REQUIRE an (l00 mrlhr)
Remlhr (100 Rem/hr mr/hr) requires a HP brief prior to entry brief.
NOTE: Per RWP, a briefing is required prior to entry into a High Rad Area which is defined as > .1 Rem/hr.
END TIME:- _- _-_- _
LR-IP-I0030-00 LR-JP-lO030-00 PERFO~ANCESTEP STANDARD SAT/UNSAT
... COMMENTS NOTE: The terminating tenninating cue shall be given to the Applicant when:
- With no reasonable progress, the Applicant exceeds double the allotted time.
- Applicant states the task is complete.
TE~INATING CUE: That completes this IPM.
TERMINATING JPM.
(
(
Plant Hatch Unit Radiation
. Radiation***
Work Permit 09-0004 Rev ACTIVE o Job Operations Inspection, Surveillance and Fire Watch - This RWP not for entries into Locked High Rad or Very High Rad 1~~~~~Areas Description Areas I Location IGENERAL PLANT LOCATION HP Coverage Authorization Briefing :,
Start Date 1/1/2009 11112009 End Date 1/1/2010 INTERMITTENT WORK GROUP CONDITIONAL Job Supv. SOS Ext. 5959 ;,
, ......_ v r
~" " "" " '
~,.
'~"
Radiological Conditions CondItions j~1_____________ T_as_ks____________~1 Tasks Description DAD Alarms Dose(rnr)
Dose (mr) Rate (rnr/h)
(mrlh)
" Refer to current survey of work area. OPS Rounds, Clearances, Surveil. 25 500
" OPS Control Room Activities 10 50 Dosimetry " Supervision 1
/ Observation 20 100 DIGITAL ALARMING DOSIMETER (DAD) . JPMs ITraining Activities 10 100
-'THOLE BODY TLD
(
Protective Clothing Requirements REFER TO WORKER/SPECIAL INSTRUCTIONS Respirators RESP Usage is Conditional per HP I Instructions IDAD's must be accessible for visual monitoring. Monitor DAD periodically while in the RCA.
Lab Coats, Booties & Gloves allowed for training, inspections, surveillances, Step Off Pad maintenance or Light Work (with HP approval)
IUnless otherwise specified by HP, full dress is required for contaminated area entry.
Use Cameras in lieu of entry, when possible, to reduce exposure.
Entries into Locked High Rad or Very High Rad areas are not permitted on this RWP.
Briefing required prior to entering a High Radiation Area Health Physics 12/26/2008 08;25 12/26/200808:25 Tenninated Prepared Staff Approved by DPHOBBS Suspended Terminated
UNIT 1 READ AND GIVE A COpy COPY TO THE OPERATOR INITIAL CONDITIONS:
1 Unit 1 is at 100% power with no significant problems.
- 2. IE11-C002B, RHR Pump B.
Health Physics reports a seal leak on lE11-C002B,
- 3. The Shift Supervisor wants the rounds operator, while performing normal round activities, to report back on the extent of the packing leak.
- 5. HP approval has been granted to use minimum requirements for entry and inspection of the leak.
( INITIATING CUES:
You are to do a pre-job brief with the rounds operator and provide the following information:
- What are the minimum dress requirements for entry.
- Assuming the highest current General Area Dose Rate, calculate the maximum stay time before the DAD alarms on dose accumulated.
- Required actions if the DAD alarms on dose accumulated.
- Assuming plant conditions change, determine the minimum General Area Dose Rate which would REQUIRE an HP brief prior to entry.
(
"t
,t Hatch Radiological Information Survey # 57773 Radiologicallnfonnation 1012512009 16:45 10/25/2009 Page 1
( U2 N.E. Diag.
Diag. 87 (2RX87NE Diag)
Pos~c @ El1tto RCA.:
k£]2
=="i
'------i
- rEZ1 2fi21 F001
( UNIT 2 N.E. DIAGONAL
- 87' Status: Approved Rx Reactor Power: 100%
Performed By: HendrIx, David Reactor Mode: 1 Griffis, Don H2 Injection Level: 10 Max Dose Rate: 6mrem/hr 6 mrem/hr Void Level: 0 MaxCntm: <MDA dpm/100 GITl2 cm2 System Running: No Approved By: Griffis, Don 1012512009 16 :45 Survey Dose: 0
Purpose:
Routine Survey Remarks: Di Rad 03-1087N Component: o A$R#s ASR#s RWP#s 09-0001 Instrument Description Comment 2368 RO-2A 725 RM-25 Postings' Postings:
l\ \Type
- JType Description ILow Dose IJ *"ILow 8aution - Rad Sign Dosimetry Required for Entry j Radiation Area
,t Hatch Radiological Information Survey # 57773 101251200916:45 Page 2 Radiation Controlled Area RWP required for entry Point Data:
- Point lType Type ~alue Value Units Level Comments 1 91 Dose Rate - Gamma G/A 3 mrem/hr Low
....ow Dose Rate - Gamma Contact 5 mrem/hr Low-Low 2 92 Dose Rate - Gamma G/A 5 mrem/hr Low Dose Rate - Gamma Contact 8 mremlhr Low-Low 3 Flex Dose Rate - Gamma G/A <2 mrem/hr Low-Low Flex ~ontamination Contamination - BIG <1000 dpm/100 cm2 Low-Low
....ow-Low 4 Flex Dose Rate - Gamma G/A 6 mrem/hr Low Flex Dose Rate - Gamma Contact 10 mrem/hr Low-Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 5 Flex Dose Rate - Gamma G/A ~.
4 mrem/hr Low Flex Dose Rate - Gamma Contact 5 mrem/hr Low-Low 6 Flex Dose Rate - Gamma G/A 2 mrem/hr Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 7 Flex Dose Rate - Gamma G/A <2 mrem/hr Low-Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 8 Flex Dose Rate - Gamma G/A <2 mrem/hr Low-Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 9 Flex Dose Rate - Gamma G/A 3 mrem/hr Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low Flex Dose Rate - Gamma G/A 2 mrem/hr Low
(, Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low
'11 11 Flex Dose Rate - Gamma G/A 2 mrem/hr Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 12 Flex Dose Rate - Gamma G/A 2 mrem/hr Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 13 Flex Dose Rate - Gamma G/A 2 mrem/hr Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 14 Flex Dose Rate - Gamma G/A <2 mrem/hr Low-Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 15 Flex Dose Rate - Gamma GIG/A A <2 mrem/hr Low-Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 16 Flex Dose Rate - Gamma G/A 2 mrem/hr Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 17 Flex Dose Rate - Gamma G/A 5 mrem/hr Low Flex Contamination - BIG <1000 dpmf100 cm2 dpm/100 Low-Low 18 Flex Dose Rate - Gamma G/A 3 mrem/hr Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 19 Flex Dose Rate - Gamma G/A ~4 mrem/hr Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 20 Flex Dose Rate - Gamma G/A 3 mrem/hr Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low 21 Flex Dose Rate - Gamma G/A 2 mrem/hr Low Flex Contamination - BIG <1000 dpm/100 cm2 Low-Low
(
- ,r
"-. '+ Radiological Information Survey # 57769 10/25/2009 16:45
,+ Hatch Radiologicallnfonnation Page 1
( U2 S.E. Diag. 87 (2RX87SE Oiag)
OIag)
RX 13(1 posted 1IJ1 N
UNIT 2 S.E. DIAGONAL -.. 81' 87'
(
Status: Approved Rx Reactor Power: 100%
Performed By: Griffis, Don Reactor Mode: 1 Aycock, Brett Aycock,Brett H2 Injection Level:
H21njectioli 10 Max Dose Rate: 20 mremfhr mremlhr Void Level: o0 MaxCntm: 232 dpmf100 dpml100 cm2 System Running: No Approved By: Griffis, Don 101251200916:45 Survey Dose: 1
Purpose:
Routine Survey Remarks: R.as 03-1087N Di Ras Component:
ASR#s RWP#s 09.0001 09-0001 Instrument Description Comment 3487 RO-2A 42123-1 TENNELEC Postings:
- Type Description
- ~ Caution - Rad Sign Radiation Area
(
+ Hatch Radiological Information Survey # 57769 10/25/2009 16:45 1012512009 Page 2 Data:
Data" Type
- Point !rype ~alue lV'alue Units Level Comments 1 103 Dose Rate - Gamma G/A 5 mrem/hr rnremlhr Low
/-Ow Dose Rate - Gamma Contact 10 mrem/hr Low-Low
/-Ow-Low Contamination - Alpha <20 f<20 dpm/100 cm2 Low-Low Contamination - BIG <MDA f<MDA ~pm/100 dpm/100 cm2 Low-Low
/-Ow-Low 2 104 Dose Rate - Gamma G/A 20 mrem/hr rnremlhr Low
/-ow Dose Rate - Gamma Contact ~O 50 rnrem/hr rnremlhr ~ow Contamination - Alpha <20 f<20 dpm/100 cm2 Low-Low
/-Ow-Low Contamination - BIG <MDA I<MDA dpm/100 cm2 Low-Low 3 Flex Dose Rate - Gamma Contact 2 mremlhr rnremlhr Low-Low Flex Contamination - Alpha <20 \:Ipm/100 dpm/100 cm2 ~ow-Low Low-Low Flex Contamination - BIG <MDA dpm/100 cm2 dpml100 ~ow-Low Low-Low 4 Flex Dose Rate - Gamma Contact ~2 mrem/hr Low-Low Flex Contamination - Alpha <20 dpm/100 cm2 Low-Low Flex Contamination - BIG <MDA ~pm/100 dpm/100 cm2 Low-Low 5 Flex Dose Rate - Gamma Contact 2 mrem/hr mremlhr Low-Low Flex ~ontamination Contamination - Alpha <20 1<20 dpm/100 cm2 Low-Low Flex Contamination - BIG <MDA I<MDA dpm/100 cm2 Low-Low 6 Flex Dose Rate - Gamma Contact 2 mrem/hr Low-Low Flex Contamination - Alpha <20 ~pm/100 dpm/100 cm2 Low-Low Flex Contamination - BIG <MDA dpm/100 cm2 Low-Low 7 Flex Dose Rate - Gamma Contact ~2 mrem/hr mremlhr Low-Low
/-Ow-Low
~<<<<<
Flex Contamination - Alpha <20 f<20 dpm/100 dpml100 cm2 Low-Low
(.,
Flex Contamination - BIG <MDA dpm/100 cm2 Low-Low I'-~< Flex Dose Rate - Gamma G/A 10 mrem/hr Low Flex Dose Rate - Gamma Contact 20 mrem/hr Low-Low Flex Contamination - Alpha <20 dpm/100 cm2 Low-Low Flex Contamination - BIG ~32 232 dpm/100 cm2 dpml100 Low-Low 9 Flex Dose Rate - Gamma G/A 12 mremfhr mremlhr Low Flex Dose Rate - Gamma Contact ~O 20 mremfhr mremlhr Low-Low Flex Contamination - Alpha <20 1<20 dpm!100 cm2 dpm/100 Low-Low Flex Contamination - BIG <MDA dpm!100 dpm/100 cm2 Low-Low
(
SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH ADMINISTRATIVE CONTROL PROCEDURE 1 OF 13
\. DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION PERSONNEL DOSIMETRY PROGRAM 60AC-HPX-002-0 NO: 12.8 EXPIRATION fA,PPROVALS:
fA.PPROVALS: EFFECTIVE DATE: DEPARTMENT MANAGER WBK DATE 2-26-02 DATE:
NIA NPGM/POAGM/PSAGM for DRM JAB forDRM DATE 2-26-02 11/18/08 1.0 OBJECTIVE This procedure establishes the requirements and responsibilities for the Personnel Dosimetry Program. It includes requirements for personnel dosimetry issuance and use, and maintenance of exposure records.
TABLE OF CONTENTS Section Page 8.1 PERSONNEL MONITORING .............................................................................................. 4 8.2 ISSUANCE OF PERSONNEL DOSIMETRY ....................................................................... 5 8.3 MULTIPLE BADGING/EXTREMITY MONITORING ............................................................ 5 8.4 WEARING AND USE OF PERSONNEL DOSiMETRy ........................................................ 6
( 8.5 EXPOSURE TRACKING AND WORK RESTRICTIONS ..................................................... 7 8.6 PERSONNEL WORK COMPLETIONITERMINATION ........................................................ 8 8.7 REPORTING REQUIREMENTS/NOTIFICATIONS .................. ;.......................................... 8 8.8 DOSIMETRY RECORDS .................................................................................................... 9 Attachments 1 DEFINITIONS ................................................................................................................... 11 2.0 APPLICABILITY The controls established by this procedure are applicable to all personnel whose duties require access to Radiation Control Areas (RCA).
3.0 REFERENCES
3.1 DEVELOPMENTAL REFERENCES 3.1.1 Regulatory Guide 1.33 Appendix A, Quality Assurance Program Requirements 3.1.2 ANSI N13.6 - 1966 (R1972), Practice for Occupational Radiation Exposure Records Systems, Section 3 & 4 3.1.3 Southern Nuclear Operating Company (SNC) Quality Assurance Topical Report (QATR)
MGR-0002 Rev 8
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 2 OF 13 20F DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION PERSONNEL DOSIMETRY PROGRAM 60AC-HPX-002-0 NO:12.8 3.2 IMPLEMENTING REFERENCES 3.2.1 10CFR20, Standards for Protection Against Radiation, Section 1502 and 2103, 2201, 2202,2203.
3.2.2 62RP-RAD-001-0, Dosimetry Issuance and Tracking 3.2.3 ANSI N13.6 - 1966 (R1972), Practice for Occupational Radiation Exposure Records Systems, Sections 3 & 4 3.2.4 FSAR, Unit 2, Sections 12.1 and 12.5 3.2.5 TS, Units 1 and 2, Section 5.4.1 3.2.6 N.E.1. 95-03, 95-05 4.0 RESPONSIBILITIES 4.1 HEALTH PHYSICS (HP)
The Health Physics Section, under the supervision of the Health Physics Manager:
( 4.1.1 Establishes and maintains an external dosimetry program which includes appropriate procedures.
4.1.2 Issues, controls and maintains personnel dosimetry equipment.
4.1.3 Maintains exposure records, prepares exposure reports and performs exposure trend analyses.
4.1.4 Provides notification when individuals approach applicable exposure limits.
4.2 PLANT SUPERVISORS AND FOREMAN 4.2.1 Ensure that all personnel under their direction comply with all rules, regulations, and procedures governing radiation safety and:
- Are informed of the exposure status of their employees.
radiological 4.3 PLANT PERSONNEL 4.3.1 Practice radiation safety and maintain individual radiation exposure As Low As Reasonably Achievable (ALARA).
4.3.2 Comply with all radiation protection rules, regulations, and procedures.
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 3 OF 13 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION PERSONNEL DOSIMETRY PROGRAM 60AC-HPX-002-0 NO:12.8 5.0 REQUIREMENTS 5.1 PERSONNEL REQUIREMENTS 5.1.1 Personnel requiring fully-qualified status must have successfully completed the Health Physics portion of General Employee Training.
5.1.2 Visitors and other less then fully-qualified radiation workers may be exempted from the training requirements of 5.1.1 provided:
5.1.2.1 The appropriate form (HPX-1124, Escorted Visitor or HPX-1125, Escorted Radiation Worker) is completed.
5.1.2.2 For escorted radiation workers, HP Supervision specifies the level of training required to ensure the individual's safety.
5.1.2.3 Training Department personnel will document completion of training requirements from 5.1.2.2, if applicable.
5.1.2.4 The individual MUST be escorted at all times while in RCAs by an authorized escort.
escort .
\, 5.1.2.5 The individual MUST complete GET radiation worker training if the visit/work period exceeds 30 days.
5.1.3 Health Physics personnel performing dOSimetry dosimetry functions will receive indoctrination and training in accordance with 70AC-TRN-001-0, Plant Training Program, and in external dosimetry requirements and procedures.
5.2 MATERIAL AND EQUIPMENT N/A - Not applicable to this procedure 5.3 SPECIAL REQUIREMENTS N/A - Not applicable to this procedure 6.0 PRECAUTIONS/LIMITATIONS 6.1 PRECAUTIONS N/A - Not applicable to this procedure 6.2 LIMITATIONS 6.2.1 No person under 18 years of age will receive an occupational dose at Plant Hatch.
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 40F 4 OF 13 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION PERSONNEL DOSIMETRY PROGRAM 60AC-HPX-002-0 NO:12.8 6.2.2 Employees who have received AND continue to retain internal medical isotopes will generally be prohibited from RCAs until elimination of the intake allows the employee to pass through a portal monitor without causing the monitor to alarm. This is a temporary restriction necessitated by the need to detect personal contamination or occupational intakes without interference from medically administered isotopes.
7.0 PREREQUISITES N/A - Not applicable to this procedure 8.0 PROCEDURE 8.1 PERSONNEL MONITORING 8.1.1 HP shall supply appropriate personnel monitoring equipment. The primary dosimeter will normally be a thermo luminescent dosimeter (TLD), and the secondary dosimeter will normally be a Digital Alarming Device (DAD) or other device deemed appropriate by HP.
Persons required to wear this personnel monitoring equipment are:
8.1.1.1 Each individual who enters a High Radiation Area or Very High Radiation Area.
( 8.1.1.2 Those individuals who enter RCAs AND are likely to receive, in one year, from sources external to the body, a dose in excess of:
- A total effective dose equivalent of 100 mrem, OR
- The sum of the deep-dose equivalent and the committed dose equivalent of 1,000 mrem, OR
- A lens dose equivalent of 300 mrem, OR
- A shallow-dose equivalent of 1,000 mrem to the skin or any extremity, OR
- A total effective dose equivalent in excess of 50 mrem to a declared pregnant woman.
8.1.2 The TLD or primary dosimeter will normally be used to provide a legal record of exposure for gamma, beta, and neutron radiations. Internal dose monitoring will be performed by air sampling, DAC-hour tracking, and/or bioassay.
8.1.3 The secondary dosimeter will be used to monitor daily accumulations of gamma radiation.
8.1.4 An Area Monitoring Program has been established outside the RCA to ensure occupational dose to individuals not having permanently assigned dosimetry is below the limits described in 8.1.1.2.
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E.E./.I. HATCH 50F 5 OF 13 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION PERSONNEL DOSIMETRY PROGRAM 60AC-HPX-002-0 NO:12.8 8.1.4.1 The Area Monitoring Program includes, but is not limited to, posted TLDs, and radiological surveys.
8.1.4.2 Posted TLD locations will be determined by HP.
8.2 ISSUANCE OF PERSONNEL DOSIMETRY An exit whole body count performed at fleet facilities (VEGP, FNP, or HNP) can NOTE: serve as an incoming WBC when transfer is directly between facilities and it is verified that no other nuclear facility has been visited.
8.2.1 New personnel will provide previous occupational radiation exposure history to the HP Section. The Training Department will provide documentation of successful completion of the Health Physics portion of General Employee Training.
8.2.2 HP shall evaluate the individual's previous exposure history, assign exposure limits, and prepare appropriate records.
(
\ 8.2.3 Qualified HP personnel shall issue primary dosimetry and/or secondary dosimetry in accordance with approved radiation protection procedures.
8.2.4 For the purpose of dosimetry issuance, visitors will be categorized as follows:
IF THE INDIVIDUAL INDIVIDUAL....., THEN THE INDIVIDUAL
.. IS AN;;,
ISAN ...
Is expected to receive >100 mrem/year OR plans to enter:
- a High Radiation Area OR Escorted Radiation Worker
- a Contaminated Area OR
- an Airborne Radioactivity Area Will only tour the RCA, observe work activities, OR perform minor work tasks Escorted Visitor that involve minimal use of radioactive materials (as determined by HP) 8.2.5 Visitor dosimetry will be issued in accordance with 62RP-RAD-001-0, Dosimetry Issuance and Tracking.
( 8.3 MULTIPLE BADGING/EXTREMITY MONITORING MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 6 OF 13
{ DOCUMENT NUMBER: REVISIONNERSION DOCUMENT TITLE:
PERSONNEL DOSIMETRY PROGRAM 60AC-HPX-002-0 NO:12.8 8.3.1 HP/ALARA will normally determine the need and specifications for multiple badging and extremity monitoring on a job-by-job basis.
8.3.2 Multiple badging, includes, but is not limited to, the issuance of TlDs TLDs and personnel monitoring devices for whole body locations other than normal issue, and extremity monitoring (Le., finger and/or toe rings). They will be issued and returned daily OR as specified by HP or Dosimetry.
8.4 WEARING AND USE OF PERSONNEL DOSIMETRY 8.4.1 Personnel will check their personnel monitoring device and/or their DAD for an initial reading before entering any RCA.
8.4.2 The TlD TLD and personnel monitoring devicelDAD device/DAD will normally be worn between the shoulders and waist and on the front of the body or as otherwise specified by HP or Dosimetry.
8.4.3 The face of the TlD TLD will normally be worn facing outward from the body.
8.4.4 When protective clothing is worn, the TlD TLD and personnel monitoring device/DAD devicelDAD MUST be worn in such a way as to be readily accessible for monitoring accumulated exposure.
( Wearing dosimetry inside protective clothing is permitted ONLY if both of the following conditions exist:
- authorization by HP Supervision has been obtained, AND
- provision has been made to allow the worker to maintain an awareness of accumulated exposure CAUTION: DO NOT CONTINUE WORKING WITH A DOSE ALARM ON A DAD.
8.4.5 If any of the following occurs, leave the area IMMEDIATELY and notify HP:
- Problems with personnel monitoring device
- DAD actuates the dose accumulated alarm
- DAD malfunctions
(
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 7 OF 13
( DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION PERSONNEL DOSIMETRY PROGRAM 60AC-HPX-002-0 NO:12.8 NOTE: If dose rate alarms are anticipated en route to the work location AND this is discussed with the worker during a pre-job briefing, it is NOT necessary to notify HP, provided the alarm clears.
8.4.6 If an unanticipated dose rate alarm occurs en route to the work location which has NOT been discussed during a pre-job briefing, retreat to a low dose area where the alarm ceases, and notify HP.
NOTE: Personnel should check their personnel monitoring device!device/ DAD while in RCAs.
A good "rule of thumb" would be to perform this check every 15-20 minutes.
8.4.7 While in Radiation, High Radiation, or Very High Radiation Areas, personnel will examine their personnel monitoring device /DAD periodically to determine their accumulated dose, unless other means are provided to keep individuals informed of their dose and the dose rate.
(
8.4.7.1 IF there is an accumulated dose alarm on the Digital Alarming Dosimeter, THEN the individual must notify HP immediately.
8.4.7.2 Personnel must immediately report the loss of any personnel monitoring device to HP.
8.4.7.3 Personnel must immediately report to HP any instance in which one individual wears another individual's dosimetry.
8.4.7.4 Personnel must notify HP whenever erratic readings are noted on their DAD.
8.4.8 Upon leaving RCAs, personnel will check their personnel monitoring device/DAD for an exit reading and may leave their dosimetry at the designated access points when work activities in the RCA are complete.
8.5 EXPOSURE TRACKING AND WORK RESTRICTIONS 8.5.1 HP shall maintain current exposure records for personnel monitored for radiation exposure.
8.5.1.1 Exposure records will be updated to incorporate TLD data as it becomes available.
8.5.1.2 Neutron exposure calculations will normally be made (pending TLD data updates) for personnel who enter a Radiation Area within the RCA where neutron radiation is present in excess of 5 mrem in one hour at 30 cm or in excess of 2 mrem in one hour at 30 cm for outside areas.
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 8 OF 13 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION PERSONNEL DOSIMETRY PROGRAM 60AC-HPX-002-0 NO:12.8 8.5.1.3 (:2: 0.3 DAC).
DAC-hours will be tracked for individuals working in posted airborne areas (;;::
At such time that an individual accrues 4 DAC-hours in a 7 day period, that individual will normally receive a whole body count.
8.5.2 HP shall investigate and evaluate dose discrepancies, abnormal exposures, and instances of lost or damaged dosimetry.
8.5.3 HP shall impose work restrictions upon personnel whose radiation exposure approaches applicable limits.
8.5.4 HP shall provide notification to appropriate personnel whenever work restrictions are imposed.
8.5.5 Skin dose assignments will be completed when applicable, in accordance with NMP-HP-004, Skin Dose Assessment.
8.6 PERSONNEL WORK COMPLETION/TERMINATION 8.6.1 Terminating personnel will normally turn in their TLD at the Dosimetry office.
8.6.2 Dosimetry personnel will complete the individual's files and records in accordance with
( approved radiation protection procedures.
8.7 REPORTING REQUIREMENTS/NOTIFICATIONS Federal reporting requirements are found in 10 CFR 20, sections 2201, 2202, and 2203.
Conditions which warrant notification or report have specific time constraints and must be considered, even in situations when limits have NOT been exceeded but conditions exist which have POTENTIAL to cause limits to be exceeded. Reports are to be made in accordance with procedure 00AC-REG-001-0, Federal and State Reporting Requirements.
8.7.1 IMMEDIATE notification shall be made when any of the following conditions exist:
8.7.1.1 Any occurrence of lost, stolen or missing licensed material in a quantity equal to or greater than 1000 times the quantity specified in 10 CFR 20, Appendix C under such circumstances that an exposure could result to persons in unrestricted areas.
8.7.1.2 Any event involving radioactive material that may have caused OR threatens to cause
- The release of radioactive material inside OR outside of a restricted area where an individual present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> could have received an intake five times the annual limit.
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 9 OF 13
( DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION PERSONNEL DOSIMETRY PROGRAM 60AC-HPX-002-0 NO:12.8 8.7.2 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery, notification shall be made when any of the following conditions exist:
Any event involving loss of control of radioactive material that may have caused OR threatens to cause
- An individual to receive, in a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (including any committed exposures), a total effective dose equivalent greater than 5 rem, OR an eye dose equivalent greater than 15 rem, OR a shallow-dose equivalent to the skin or extremities greater than 50 rem.
- The release of radioactive material inside OR outside of a restricted area in which an individual present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> could have received an intake greater than one occupational annual limit.
8.8 LOSS/MISUSE OF DOSIMETRY EQUIPMENT 8.8.1 The loss or misuse of dosimetry equipment causes special problems for the HP staff:
- it prevents proper monitoring radiation exposure received.
- it requires assignment of the maximum credible radiation exposure to the individual,
( potentially limiting an individual's access to Radiation Control Areas.
8.8.2 Examples of misuse of dosimetry equipment include:
- deliberately exposing dosimetry to unnecessary radiation.
- wearing another individual's or tampering with yours or another individual's dosimetry.
- willful destruction of or damage to any piece of dosimetry equipment.
8.8.3 Replacement of lost dosimetry requires the approval of HP. Replacement after repeated losses will require the approval of the HP Manager, and may be the basis for disciplinary action.
8.9 DOSIMETRY RECORDS Personnel dosimetry program records shall be maintained in accordance with 20AC-ADM-002-0, Plant Records Management. These shall include, but are not limited to, the following:
8.9.1 Current year occupational radiation exposure received at other installations and Plant Hatch 8.9.2 Current occupational radiation exposure received at Plant Hatch 8.9.3 Current occupational radiation exposure received at other installations for those individuals
(, that maintain active access at other facilities as well as Plant Hatch MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 10 OF 13
( DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION PERSONNEL DOSIMETRY PROGRAM 60AC-HPX-002-0 NO:12.8 8.9.4 Records relevant to exposure limits (Le., shallow dose equivalent, lens dose equivalent, committed dose equivalent, and total effective dose equivalent) 8.9.5 Records of unusual exposures, incidents and investigations, and planned special exposures 8.9.6 Results of surveys used to determine exposures received in the absence of personnel monitoring devices 8.9.7 Record of maintenance, surveillances, and calibrations of personnel dOSimetry dosimetry equipment and instrumentation 8.9.8 Occupational radiation exposure records may be shared nationwide via the Personnel Access Data System (P.A.D.S.). Any and all information transferred through this medium will conform to the requirements set forth in N.E.I documents 95-03 and 95-05.
(
MGR-0001 Rev 3
SNC PLANT E.I.E.!. HATCH I Pg 11 OF 13
(
DOCUMENT TITLE:
PERSONNEL DOSIMETRY PROGRAM II DOCUMENT NUMBER:
60AC-HPX-002-0 RevNerNo:
12.7 ATTACHMENT ..1 J. Att. Pg.
TITLE: DEFINITIONS 1 of 3 DEFINITIONS Airborne radioactive material means radioactive material dispersed in the air in the form of dusts, fumes, particulates, mists, vapors, or gases.
Annual limit on intake (ALI) means the derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a year. ALI is the smaller value of intake of a given radionuclide in a year by the reference man that would result in a committed effective dose equivalent of 5 rem (0.05 Sv) or a committed dose equivalent of 50 rem (0.5 Sv) to any individual organ or tissue.
Bioassay (radiobioassay) means the determination of kinds, quantities or concentrations, and, in some cases, the locations of radioactive material in the human body, whether by direct measurement (in vivo counting) or by analysis and evaluation of materials excreted or removed from the human body.
Controlled Area by definition in 10 CFR 20, means an area, outside of a restricted area but inside the site boundary, access to which can be limited by the licensee for any reason. At Plant Hatch, for radiological purposes, the controlled area means any area outside radiologically posted (restricted) areas up to the plant site boundary.
( equivalent(H d1 which applies to external whole-body exposure, is the dose equivalent at a Deep-dose eguivalent(H tissue depth of 1 cm (based on a tissue density of 1000 mg/cm 22 ).
Derived air concentration (DAC) means the concentration of a given radionuclide in air which, if breathed by the reference man for a working year of 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> under conditions of light work (inhalation rate 1.2 cubic meters of air per hour), results in an intake of one ALI.
Dose or radiation dose is a generic term that means absorbed dose, dose equivalent, effective dose equivalent, committed dose equivalent, committed effective dose equivalent, or total effective dose equivalent, as defined in 10 CFR 20.1003.
Escorted Radiation Worker is an individual authorized to perform work involving radioactive materials and may enter Contaminated, High Radiation or Airborne Radioactivity Areas. In general, these individuals have NOT completed the radiation worker training requirements of GET within the past year. The individual will be briefed and/or trained consistent with the work to be performed.
Escorted Visitor is an individual who is NEITHER a radiation worker NOR has received current GET radiation worker training. Visitors may tour the RCA, observe work activities and perform minor work tasks involving minimal use of radioactive materials (as determined by HP). Visitors are NOT allowed to enter High Radiation Areas, Very High Radiation Areas, Contaminated Areas or Airborne Radioactivity Areas. A visitor's dose will be restricted to 100 mrem/year or less, UNLESS assigned duties meet the definition outlined in "occupational dose", in which case the limits of 8.1.1 apply.
MGR-0001 Rev 3
SNC PLANT E.!.E.I. HATCH L I Pg120F13 DOCUMENT TITLE:
PERSONNEL DOSIMETRY PROGRAM II DOCUMENT NUMBER:
60AC-HPX-002-0 RevNer No:
12.7 ATTACHMENT .1 -.1 Att.
AU. Pg.
TITLE: DEFINITIONS 20f3 2 of 3 Exposure means being exposed to ionizing radiation or to radioactive material.
External dose means that portion of the dose equivalent received from radiation sources outside the body.
Extremity means hand, elbow, arm below the elbow, foot, knee, or leg below the knee.
Individual monitoring means--
(1) The assessment of dose equivalent by the use of devices designed to be worn by an individual and/or (2) The assessment of committed effective dose equivalent by bioassay or by determination of the time-weighted air concentrations to which an individual has been exposed (Le.
(i.e.
DAC-hours) and/or (3) The assessment of dose equivalent by the use of survey data.
Individual Monitoring Devices by definition in 10 CFR 20, means devices designed to be worn by a single individual for the assessment of dose equivalent such as film badges, thermoluminescent dosimeters (TLDs), and personal ("lapel") air sampling devices.
( Lens dose equivalent applies to the external exposure of the lens of the eye and is taken as the dose equivalent at a tissue depth of 0.3 cm (based on a tissue density of 300 mg/cm22 ).
Member of the Public means any individual except when that individual is receiving an occupational dose. Dose to a member of the public must not exceed 100 mrem/year.
Monitoring (radiation monitoring, radiation protection monitoring) means the measurement of radiation levels, concentrations, surface area concentrations, or quantities of radioactive material and the use of the results of these measurements to evaluate potential exposures and doses.
Non-monitored Occupationally Exposed Workers are individuals who, in the course of employment with the company, are NOT likely to receive greater than 100 mrem/yr.
Occupational Dose means the dose received by an individual in the course of employment in which the individual's assigned duties involve exposure to radiation and/or to radioactive material from licensed and unlicensed sources of radiation at Plant Hatch. It does not include dose received from background radiation, as a patient from medical practices, from voluntary participation in medical research programs, or as a member of the public. An individual with a picture security badge at Plant Hatch has assigned duties which meet the definition above and receives an occupational dose. The limits of 8.1.1 apply.
Rem is the special unit of any of the quantities expressed as dose equivalent. The dose equivalent in rem is equal to the absorbed dose in rad multiplied by the quality factor.
(
MGR-0001 Rev 3
SNC PLANT E.I. HATCH I PQ 13 OF 13 Pg130F13
(
DOCUMENT TITLE:
PERSONNEL DOSIMETRY PROGRAM IIDOCUMENT NUMBER:
60AC-HPX-002-0 RevNerNo:
RevNer No:
12.7 ATTACHMENT .1 -.1 AU. Pg.
Att.
TITLE: DEFINITIONS 3 of 3 30f3 Restricted Area by definition in 10 CFR 20, means an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. At Plant Hatch, for radiological purposes, the restricted area means any radiologically posted area.
Shallow-dose equivalent (HJ, (H§), which applies to the external exposure of the skin or an extremity, is taken as the dose equivalent at a tissue depth of 0.007 cm (based on a tissue density of 7 mg/cm22 )
averaged over the 10 square centimeters of skin receiving the highest exposure.
Unrestricted Area by definition in 10 CFR 20, means an area, access to which is neither limited nor controlled by the licensee. At Plant Hatch, for radiological purposes, the unrestricted area means an area to which access is not limited or controlled by the licensee, or any area within the site boundary used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
Whole body means, for purposes of external exposure, head, trunk (including male gonads), arms above the elbow, or legs above the knee.
(
MGR-0001 Rev 3
SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH ADMINISTRATIVE CONTROL PROCEDURE 1 OF 24 10F24
\. DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO:19.4 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MANAGER Jim Dixon DATE 11-10-06 DATE:
10/30/08 10/30108 N/A NIA NPGM/POAGM/PSAGM NPGMIPOAGMIPSAGM Dennis Madison DATE 11-10-06 1.0 OBJECTIVE This procedure establishes the requirements and responsibilities for monitoring and contrOlling controlling exposure to radiation and contamination. It includes criteria for Radiation Control Areas (RCAs),
the Radiation Work Permit (RWP) system, radiological surveys and sampling, and temporary shielding. This procedure, in part, satisfies the requirements ofTS 5.7.1 (a, b, c) and TS 5.7.2.
of TS 5.7.1(a, TABLE OF CONTENTS Section Page 8.1 RADIATION CONTROL AREA (RCA) REQUiREMENTS .................................................... 6 8.2 RADIATION WORK PERMITS (RWPS) ............................................................................ 11 8.3 CONTAMINATION CONTROLS ........................................................................................ 13 8.4 SURVEYS AND SAMPLING ............................................................................................. 18
( 8.5 TEMPORARY SHIELDING ............................................................................................... 20 8.6 RADIOACTIVE SPILL CONTROL. .................................................................................... 21 CONTROL .....................................................................................
8.7 ANNUAL RADIATION PROTECTION PROGRAM REViEW ............................................. 22 8.8 RECORD RETENTION ..................................................................................................... 22 ATTACHMENTS 1 CONTAMINATION PROTECTION GUiDELINES .............................................................. 23 2 DEFINITIONS ................................................................................................................... 24 2.0 APPLICABILITY 2.1 This procedure is applicable to all activities in which personnel may be exposed to radiation from by-product or licensed radioactive materials.
3.0 REFERENCES
3.1 DEVELOPMENTAL REFERENCES 3.1.1 Regulatory Guide 1.33, Appendix A, Quality Assurance Program Requirements
(
MGR-0002 Rev 8
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 20F24
( DOCUMENT NUMBER: REVISIONNERSION DOCUMENT TITLE:
RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 3.1.2 ANSI N13.2-1969, Guide for Administrative Practices in Radiation Monitoring, Sections 3 and4 and 4 3.1.3 Southern Nuclear Operating Company (SNC) Quality Assurance Topical Report (QATR) 3.2 IMPLEMENTING REFERENCES 3.2.1 10CFR20, Standards for Protection Against Radiation, Sections 20.1004, 20.1005, 1101(c), 1204, 1302, 1501-1906, and 2101-2103 3.2.2 FSAR, Unit 2, Chapter 12 3.2.3 TS 5.7.1 (a., b., c.) and TS 5.7.2 3.2.4 NMP-GM-002, Corrective Action Program 3.2.5 NMP-AD-008, Applicability Determination 3.2.6 62RP-RAD-012-0, Selection and Use of Temporary Shielding 3.2.7 62RP-RAD-017-0, Release Surveys
( 3.2.8 62 RP-RAD-044-0, Identification and Tracking of Hot Spots 62RP-RAD-044-0, 3.2.9 DI-RAD-03-1087, Survey/Inspection Survey/lnspection Frequency and Work Scheduling 3.3 FULL SIZE FORMS
- HPX-0010, Request for Radiation Work Permit / Survey Update Request
- HPX-0297, Temporary Shielding Request
- HPX-0583, Additional Work Scope Approval Form
- HPX-1034, Annual Radiation Protection Program Review 4.0 RESPONSIBILITIES 4.1 HEALTH PHYSICS (HP)
NOTE: Health Physics personnel have the responsibility and authority to stop or prevent initiation of any activity that, if continued, would result in a violation of HP policies or procedures, an unplanned radiation exposure, exposure to airborne radioactivity, the release of radioactive materials or the spread of contamination.
4.1.1 Reviews radiation safety incidents and approves corrective action.
(
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 30F24 3 OF 24 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 4.1.2 Establishes and maintains a radiation and contamination control program which includes appropriate procedures and instructions.
4.1.3 Establishes and maintains an RWP system to control and minimize exposure to radiation and contamination.
4.1.4 Performs and documents radiation and contamination surveys, air sampling and analyses.
4.1.5 Classifies, posts and barricades RCAs to control radiation exposure and spread of contamination.
4.1.6 Determines protective clothing requirements and stay times based on radiological conditions in work area(s).
4.1.7 Requests and supervises the installation of temporary shielding.
4.1.8 Supervises and monitors the decontamination of personnel, equipment and facilities.
4.1.9 Controls the use of contaminated materials and equipment.
4.1.10 Performs maintenance, calibrations and surveillances on HP instrumentation and equipment.
(
4.1.11 Performs trend analysis on radiation and contamination exposure controls. Recommends actions necessary to correct adverse trends.
4.1.12 Coordinates with Operations to perform flushing of Hot Spots.
4.2 ENGINEERING SUPPORT 4.2.1 Performs seismic and design loading evaluations upon systems and/or equipment on which temporary shielding will be placed.
4.2.2 Coordinates with HP to determine shielding specifications.
4.3 OPERATIONS 4.3.1 Operations, as directed by 30AC-OPS-003-0, Plant Operations, will ensure plant operations are conducted with the requirements of the operating license, Technical Specifications, plant operating procedures and all applicable state and federal regulations.
4.3.2 Operations will be responsible for determining if components containing Hot Spots can be flushed.
4.3.3 If flushing is possible, Operations will provide support in reducing radiation exposure by
( manipulating valves, reviewing plant drawings and layouts in response to identified deficiencies (Hot Spots).
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 40F24 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 4.4 MAINTENANCE Maintenance will provide the necessary support activities as directed by the Maintenance Manager for Hot Spot removal, when flushing cannot be performed or is unsuccessful.
4.5 PLANT SUPERVISORS AND FOREMEN 4.5.1 Ensure that personnel under their direction comply with rules, regulations and procedures associated with radiation safety and comply with RWPs.
4.5.2 Request RWPs to support work in RCAs.
4.6 PLANT PERSONNEL 4.6.1 Practice radiation safety and maintain their radiation exposure As Low As Reasonably Achievable (ALARA).
4.6.2 Read and comply with all radiation protection postings, rules, regulations, procedures and RWP requirements.
(
4.6.3 On each shift, prior to the start of any work in an RCA other than Operations or Chemistry routine surveillance or sampling, planned activities MUST be clearly communicated to HP.
Deviations from routine activities should be communicated as they occur.
4.6.4 Upon exit from a contaminated area, personnel should proceed directly to a personnel contamination monitor. Exceptions to this requirement can only be approved through Health Physics.
4.6.5 Notify HP promptly when radiological problems occur and interface with HP to resolve deficiencies.
5.0 REQUIREMENTS 5.1 PERSONNEL REQUIREMENTS HP personnel will receive training in radiation and contamination control requirements and procedures.
5.2 MATERIAL AND EQUIPMENT N/A - Not applicable to this procedure MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 50F24 5 OF 24 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 5.3 SPECIAL REQUIREMENTS The terms "safety evaluation" and "10 CFR 50.59 evaluation" are to be interpreted as the evaluation(s) required by the completion of an Applicability Determination in accordance with NMP-AD-OOB.
6.0 PRECAUTIONS/LIMITATIONS 6.1 PRECAUTIONS N/A - Not applicable to this procedure 6.2 LIMITATIONS 6.2.1 Orex protective clothing MUST NOT be used for work with a "spark producing" potential or work with energized equipment.
6.2.2 Clothing used to protect personnel from radioactive contamination MUST NOT be used for any other purpose.
( 6.2.3 Protective clothing (PCs) MUST NOT be removed from the Power Block without the approval of HP.
6.2.4 No Orex clothing is to be removed from the Protected Area without the approval of HP.
7.0 PREREQUISITES On each shift, prior to the start of any work in an RCA other than Operations or Chemistry routine surveillance or sampling, planned activities MUST be clearly communicated to HP. Deviations from routine activities should be communicated as they occur.
(
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 60F24 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 8.0 PROCEDURE 8.1 RADIATION CONTROL AREA (RCA) REQUIREMENTS This subsection defines radiological areas at Plant Hatch and establishes the criteria for their control.
Figure 1 8.1.1 Radiation Control Area
( 8.1.1.1 An RCA is any posted radiological area (Le., Radiation Area, High Radiation Area, Very High Radiation Area, Contaminated Area, Airborne Radioactivity Area, or area containing Radioactive Material).
8.1.1.2 HP SHALL conspicuously post Radioactive Material and entrances and perimeters of Radiation Areas, High Radiation Areas, Very High Radiation Areas or Airborne Radioactivity Areas with the standard radiation caution symbol and the type of RCA.
8.1.1.2.1 High Radiation Areas SHALL be barricaded.
8.1.1.2.2 High Radiation Areas SHALL be locked or continuously guarded when the dose rate is greater than or equal to 1000 mrem/hr at 30 cm from the radiation source.
8.1.1.2.3 Very High Radiation areas SHALL be locked.
8.1.1.2.4 IF, during the movement of radioactive material, it is found to be impractical to establish and maintain posted boundaries/barricades for Radiation or High Radiation areas, an HP technician may act as the boundary/barricade while the material is in movement. An HP Technician can act in this capacity for maximum of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after which the area must be posted and/or barricaded in accordance with applicable procedures and regulations.
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 70F24 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 Figure 2 8.1.2 Radiation Area 8.1.2.1 Any area accessible to personnel in which radiation levels could result in an individual receiving a dose equivalent in excess of 5 mrem in one hour at 30 cm from the radiation source OR from any surface that the radiation penetrates SHALL be conspicuously posted with the standard radiation caution symbol and the words "Caution Radiation Area" (See figure 2).
8.1.2.2 Any yard area, outside buildings or enclosures which contain RCAs accessible to personnel in which radiation levels could result in an individual receiving a dose
( equivalent in excess of 2 mrem in one hour at 30 cm from the source OR from any surface that the radiation penetrates SHALL be conspicuously posted with the 6.
standard radiation caution symbol and the words "Caution Radiation Area" (See figure 2).
- DANGER a* HIGH RADIATION AREA a*
Figure 3 8.1.3 High Radiation Area 8.1.3.1 High Radiation Area means an area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 0.1 Rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source OR 30 cm from any surface that the radiation penetrates.
8.1.3.2 High Radiation Areas SHALL be barricaded AND conspicuously posted with a sign unique to that type area, that has the standard radiation caution symbol and the words "Danger High Radiation Area" (See figure 3)
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 80F24 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 8.1.3.3 Entrance to High Radiation Areas SHALL be controlled by issuance of an RWP. A whole body TLD is required to enter a High Radiation Area. Any individual or group of individuals permitted to enter such areas SHALL be provided with OR accompanied by one or more of the following:
- A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset cumulative dose is received. Entry into such an area is permitted after the dose rates have been made known to personnel
- An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual SHALL be responsible for controlling activities within the area, keeping personnel exposure ALARA and performing radiological monitoring at the frequency specified in the RWP
(
Figure 4 8.1.4 Locked High Radiation Area NOTE: The Shift Supervisors maintain a high radiation door master key for emergency use only.
8.1.4.1 Each High Radiation Area in which the intensity of radiation is greater than or equal to 1000 mrem/hr at 30 cm SHALL be locked or continuously guarded to prevent unauthorized entry into such areas. The doors SHALL be locked in such a way that no individual will be prevented from leaving the area. The keys will be controlled in accordance with 62RP-RAD-016-0, Control of High Radiation Areas. These doors are normally painted red. (See figure 4)
(
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 90F24
( DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 Figure 5 8.1.5 Very High Radiation Area NOTE: For beta and gamma radiation, 1 Rad = = 1 Rem.
8.1.5.1 Very High Radiation Areas are areas which could be made accessible to an individual in which the potential radiological condition could result in the following situations:
8.1.5.1.1 An individual entering this area could receive from radiation sources external extemal to the body an absorbed dose in excess of 500 Rad in one hour at 1 meter from a radiation source OR 1 meter from any surface that the radiation penetrates.
( 8.1.5.1.2 An area where the radiological conditions could increase very rapidly resulting in an individual receiving an acute overexposure, exceeding Administrative and/or Federal limits.
8.1.5.2 Very High Radiation Areas SHALL be locked AND conspicuously posted with a sign unique to that type area that has the standard radiation caution symbol and the words "Grave Danger, Very High Radiation Area" (see figure 5). The doors to Very High Radiation Areas are normally painted orange.
8.1.5.3 Entrance to a Very High Radiation Area SHALL be controlled by issuance of an RWP.
Any individual or group of individuals permitted to enter such areas SHALL be provided with all of the following:
- A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset cumulative dose is received.
- A whole body TLD for each individual.
- An individual qualified in radiation protection procedures, equipped with a radiation dose rate monitoring device. This individual SHALL be responsible for controlling activities within the area, keeping personnel exposure ALARA and performing radiological monitoring at the frequency specified in the RWP
(
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 10 OF 24
{
\ DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 NOTE:
- Doors to Very High Radiation Areas SHALL have keys unique to the specific door. No master keys will be made for these doors.
- Keys to Very High Radiation Area doors have additional requirements for issue and are issued in accordance with 62RP-RAD-016-0, Control of High Radiation Areas.
8.1.5.3.4 Each Very High Radiation Area SHALL be locked to prevent unauthorized entry into such areas. The doors SHALL be locked in such a way that no individual will be prevented from leaving the area.
Figure 6
(
8.1.6 Contaminated Area 8.1.6.1 Any area where the surface contamination exceeds one of the following limits will be considered a Contaminated Area:
Beta-gamma 1000 dpm/100cm22 Alpha 20 dpm/100cm 22 8.1.6.2 Contaminated Areas MUST be conspicuously posted with the standard radiation caution symbol and the words "Caution Contaminated Area" or "Danger Contaminated Area".
8.1.6.3 HP will take measures to minimize the migration of high contamination to low or non-contaminated areas.
8.1.6.4 Areas where activities such as opening a contaminated system or grinding/machining on contaminated surfaces MUST be controlled as Contaminated Areas.
(
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 11 OF 24
( DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 8.1.7 Airborne Radioactivity Area 8.1.7.1 Any area in which airborne concentrations of radioactive materials exceed 30% of the limits in 10CFR20, Appendix B, Table I, Column 3 SHALL be considered an Airborne Radioactivity Area.
8.1.7.2 8.1 .7.2 All entrances to these areas SHALL be posted with the standard radiation symbol with the words "Caution Airborne Radioactivity Area".
8.1.7.3 Posting is also recommended for operations likely to cause airborne contamination such as initial opening of reactor systems, grinding, welding, burning or operating air-operated equipment in Contaminated Areas.
8.1.8 Radioactive Materials Area 8.1.8.1 Any area in which radioactive material is used or stored and which contains radioactive material in an amount exceeding 10 times the quantity of such material specified in 10CFR20 Appendix C SHALL be considered a radioactive materials area.
8.1.8.2 These areas SHALL be conspicuously posted with the standard radiation caution symbol and the words "Caution Radioactive Materials" or "Danger Radioactive Materials" .
(
8.2 RADIATION WORK PERMITS (RWPs)
NOTES:
- All personnel will normally access an RWP through the DAD (Digital Alarming Dosimeter) System.
- The RWP also provides a mechanism for evaluating person-Rem expenditures on each job.
8.2.1 All entrances into an RCA require the use of an RWP. On each shift, prior to starting any work in an RCA other than routine surveillance or sampling, planned activities should be clearly communicated to HP. Deviations from routine activities should be communicated as they occur.
8.2.2 HP SHALL establish and maintain an RWP system to control and minimize radiation exposure during operation and maintenance activities.
8.2.3 RWPs fall into one of two categories: General or Specific.
8.2.4 General Radiation Work Permits MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 12 OF 24 I
\ DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 8.2.4.1 General RWPs are used for inspection, plant tours and general area walkdowns in areas other than Highly Contaminated, High Radiation, Very High Radiation or Airborne Radioactivity Areas.
8.2.4.2 General RWPs specify basic radiation safety requirements such as special training requirements, protective clothing requirements, dosimetry, respiratory protection, access restrictions and the type and/or extent of work that is authorized.
8.2.5 Specific Radiation Work Permits Specific RWPs are used for controlling work that is not controlled by a General RWP and occurs in a High Radiation Area, High Contamination Area or areas with changing radiological conditions. The Specific RWP will be modified on the basis of surveys.
Radiological surveys SHALL be performed as required to support RWP authorized work.
8.2.5.1 IF multiple breaches are made within the same system in locations with similar radiological conditions, THEN a single RWP may be used.
8.2.5.2 When a job is in a High Radiation Area, HP may determine that performing pre-job RWP surveys will not be consistent with ALARA. RWPs will be created in accordance with 62RP-RAD-006-0, RWP Processing.
(
8.2.5.3 A Specific RWP will be considered if any of the following conditions exist:
- General area radiation levels ;:::100 mrem/hr
2
- Loose surface contamination levels >200,000 dpm/100 cm
- Breach of a highly contaminated system, OR in a High or Very High Radiation Area 8.2.6 Issuance of Radiation Work Permits 8.2.6.1 The group responsible for performing the job/surveillance or HP may initiate an RWP by submitting form HPX-0010, "Request for Radiation Work Permit / Survey Update Request".
8.2.6.2 HP will normally be notified at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> before the work is to begin to allow sufficient time to generate the RWP and perform an ALARA review, if necessary. Jobs requiring immediate attention may be exempted from this requirement.
8.2.6.3 HP will review the work to be performed in RCAs.
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 13 OF 24
( DOCUMENT NUMBER: REVISIONNERSION REVISIONIVERSION DOCUMENT TITLE:
RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 8.2.6.4 HP will determine the RWP requirements, type of HP coverage and whether or not an ALARA briefing is required.
8.2.6.5 If required, ALARA briefings will be performed in accordance with 60AC-HPX-009-0, ALARA Program.
8.2.6.6 HP will prepare the RWP.
8.2.6.7 RWP authorized work may commence after notification of HP.
8.2.7 The work assignment of personnel added to an existing RWP MUST be directly related to or associated with the job description on the RWP.
8.2.8 Addition of Work Scope to an Existing RWP 8.2.8.1 IF the work scope is to be increased on an RWP, THEN form HPX-0583, Additional Work Scope Approval Form, will normally be completed. This form could be submitted by Central Scheduling, the job supervisor or the group responsible for the additional work.
8.3 CONTAMINATION CONTROLS
( 8.3.1 Use of Protective Clothing
\
8.3.1.1 Personnel MUST don, wear and remove protective clothing consistent with the training received.
8.3.1.2 HP will ensure protective clothing is appropriate for the levels and state of contamination expected for the area entered and consistent with ALARA considerations of Total Effective Dose Equivalent (TEDE).
8.3.1.3 While outside the separate dressing areas provided, personnel subject to procedures for donning and removing protective clothing will be expected to remain modestly attired in the workplace.
8.3.1.4 Modesty Garments
- consist of either scrub suits, shorts and a t-shirt, or any other appropriate non-revealing clothing. Tight fitting, revealing undergarments are not acceptable
- may be provided by either the employee or the employee's department (at the discretion of Management).
- may be worn under protective clothing.
- . should be made of the appropriate material to meet the safety requirements for working with energized circuits and "spark producing work", when necessary.
- MUST NOT be used as an outer garment in Contaminated Areas.
- are considered personal clothing and subject to reimbursement guidelines.
MGR-0001 Rev 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 14 OF 24
( DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION RADIATION AND CONTAMINATION CONTROL 60AC-HPX-004-0 NO: 19.4 8.3.1.5 Paper coveralls will be provided to personnel who lose their personal clothing because of radioactive contamination or to those who are required to shower and wear a paper suit for a whole body count 8.3.1.6 Protective Clothing includes, but is not limited to, the following:
- Cotton PCs
- Orex PCs
- Rubber gloves
- Cotton hoods
- Orex hoods
- Rubber shoe covers
- Cotton booties
- Orex booties
- High-top rubber boots
- Cotton glove liners
- Orex lab coats
- Plastic suits 8.3.1.7 Entries into Contaminated Areas for inspections, surveillances, Step Off Pad maintenance or Light Work (with HP approval) may be made with plastic and/or rubber shoe covers, OREX or cotton booties, cotton gloves, rubber gloves and an OREX lab coat (provided removable contamination levels within the area average less than 10,000 dpm/100 cm 22 and the RWP allows such entries). Lab coats may be worn over personal clothing or scrubs. No sitting, kneeling, climbing or moderate to heavy work is allowed while wearing this level of dress. Shorts are not allowed when wearing minimum dress in posted contaminated areas.
( OREX PROTECTIVE CLOTHING MUST NOT BE USED FOR WORK WITH A CAUTION: "SPARK PRODUCING" POTENTIAL OR WORK WITH ENERGIZED EQUIPMENT.
8.3.1.8 OREX PCs will normally be the protective garment of choice for most Contaminated Area applications.
8.3.1.9 OREX PCs should not be used as the outer garment in wet conditions. Waterproof or water-resistant materials such as plastic suits or other suitable substitutes should be used in these environments.
8.3.1.10 Cotton PCs may be used based on radiological conditions or when working with energized circuits, when approved by HP.
8.3.1.11 Fire retardant (orange) PCs should be used when welding or performing any other activity which produces a spark potential.
8.3.1.12 Care should be taken when working with rotating equipment. See Attachment 1 for protective clothing guidance when working with this condition.
8.3.2 Contamination Monitoring 8.3.2.1 Personnel will properly survey themselves for contamination when leaving a posted Contaminated Area or any building, enclosure or outside posted area which contains a Radiologically Controlled Area.
(
MGR-0001 Rev 3
SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH SYSTEM OPERATING PROCEDURE 1 OF 284 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
RESIDUAL HEAT REMOVAL SYSTEM 34S0-E11-010-2 33.14 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR G. L. Johnson DATE 01/19/07 DATE:
N/A SSM SSM// PM DATE 5-29-09 N/A N/A 1.0 OBJECTIVE This procedure gives instructions for operation of the Residual Heat Removal (RHR) System.
TABLE OF CONTENTS Section Page 2.0 APPLICABILITY APPLiCABILITy .................................................................................................................. 3
3.0 REFERENCES
................................................................................................................... 3 4.0 REQUIREMENTS ............................................................................................................... 4 5.0 PRECAUTIONS/LIMITATIONS ........................................................................................... 5 5.1 PRECAUTIONS ............................................................................................................ 5 5.2 LIMITATIONS ............................................................................................................... 7 6.0 PREREQUISITES ............................................................................................................... 8
(
7.0 PROCEDURE ..................................................................................................................... 9 7.1 STANDBy ..................................................................................................................... 9 7.2 SYSTEM STARTUP AND OPERATION ..................................................................... 13 7.2.1 Automatic LPCI Initiation ...................................................................................... 13 LPCllnitiation 7.2.2 Manual LPCI Initiation .......................................................................................... 16 7.2.3 Shutdown Cooling Mode (Condition 3) ................................................................. 19 7.2.4 Shutdown Cooling Mode (Conditions 4 & & 5) ......................................................... 37 7.2.5 Suppression Pool Cooling Mode .......................................................................... 58 7.2.6 RHR Service Water Startup ................................................................................. 61 7.3 SHUTDOWN ............................................................................................................... 66 7.3.1 Shutdown From LPCI ........................................................................................... 66 7.3.2 Shutdown From The Shutdown Cooling Mode ..................................................... 67 7.3.3 Shutdown From The Suppression Pool Cooling Mode ......................................... 71 7.3.4 Shutdown Of RHR Service Water Water.........................................................................
....................................................................... 72 7.4 INFREQUENT OPERATION ....................................................................................... 73 7.4.1 Filling And Venting The RHR System .................................................................. 73 7.4.2 Shifting Shutdown Cooling Loops ........................................................................ 73 7.4.3 RHR Assisted Fuel Pool Cooling .......................................................................... 74 7.4.4 Draining The Reactor Vessel To Radwaste ......................................................... 75 7.4.5 Draining The Suppression Pool To Radwaste ...................................................... 76 7.4.6 Containment Spray Mode .................................................................................... 78 7.4.7 Swapping Division I RHR Service Water Strainers ............................................... 81 7.4.8 Swapping Division II RHR Service Water Strainers .............................................. 82 7.4.9 Manual LPCI Initiation While In Shutdown Cooling .............................................. 83 LPCllnitiation 7.4.10 Emergency Filling Of The RHR System ............................................................... 85 MGR-0002 Ver. 8.1
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 2 OF 284 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
RESIDUAL HEAT REMOVAL SYSTEM 34S0-E11-010-2 33.14 7.4.11 RPV/RPV Cavity Makeup/Fill ............................................................................... 86 7.4.12 RHR Loop System Pressure Control Due To Inleakage ....................................... 87 7.4.13 Cycling 2E11-F015A (2E11-F015B} ..................................................................... 92 7.4.14 RHR Pump Miscellaneous Manual Startup .......................................................... 93 7.4.15 Placing Second Shutdown Cooling Loop in Service with Unit in Hot Shutdown(Paraliel SDC SOC Loop Operation} ............................................................. 95 7.4.16 Placing Second Shutdown Cooling Loop in Service with Unit in Cold Shutdown or Refuel ............................................................................................................ 113 7.4.17 Shutdown of One Loop of RHR SDC SOC with Two Loops of RHR SDC SOC in Service .. 134 7.4.18 RHR Shutdown Cooling Removal from Service and Restoration for Short Out of Service Periods in Condition 4 or 5 ................................................................ 136 7.4.19 RHR Shutdown Cooling Removal from Service and Restoration for Short Out of Service Periods in MODE 3 ........................................................................... 143 7.4.20 Reducing RHR Loop Pressure After Running RHR System ............................... 153 7.4.21 Flushing RHR With Condensate Transfer And Placing SOC SDC In Service ............. 154 7.4.22 Reversing RHRSW Pump Motor Cooling Water Flow ........................................ 155 7.4.23 RHRSW System Fill ........................................................................................... 156 7.4.24 Flushing To Remove Hot Spots In The RHR System Piping .............................. 159 7.4.25 Draining an RHR Loop to the Torus for a System Outage ................................. 160 Attachments
( 1 RHR SYSTEM RESTORATION ...................................................................................... 161 2 RHR SYSTEM ELECTRICAL LINEUP ............................................................................ 165 3 RHR SYSTEM VALVE LINEUP ...................................................................................... 176 4 RHR SYSTEM INSTRUMENT VALVE LINEUP .............................................................. 198 5 CONTAINMENT SPRAY INITIATION PLACARD ............................................................ 208 6 RHR PUMP START PREREQUISITES FOR SDC SOC MODE .............................................. 209 7 NOT USED ..................................................................................................................... 211 8 NOT USED ..................................................................................................................... 212 9 RHR SYSTEM GENERAL INFORMATION ..................................................................... 213 10 RHRSW DIVISION PUMP INITIATION PLACARD ......................................................... 217 11 VENTING RHR SYSTEM PRESSURE DUE TO INLEAKAGE ........................................ 218 12 RHR LOOP DRAINING AND FILLING IN PREPARATION FOR SDC SOC ............................ 220 13 VENTING RHR LOOP A(B) PIPING ............................................................................... 233 14 USING CONDENSATE TRANSFER TO FLUSH FOR SHUTDOWN COOLING ............. 248 15 SUPPRESSION POOL COOLING INITIATION PLACARD ............................................. 278 16 FLUSHING TO REMOVE HOT SPOTS IN RHR SHUTDOWN COOLING SUCTION PiPiNG ............................................................................................................................ 279 17 LPCIINITIATION ............................................................................................................ 284
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE i PLANT E. I. HATCH 3 OF 284
\. DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
RESIDUAL HEAT REMOVAL SYSTEM 34S0-E11-010-2 34S0-E 11-0 10-2 33.14 2.0 APPLICABILITY This procedure applies to the Unit Two Residual Heat Removal System.
3.0 REFERENCES
3.1 Technical Specifications, Unit Two, 3.3.6.1, 3.3.5.1, 3.3.3.2, 3.5.1, 3.5.2, 3.6.2.1, 3.6.2.2, 3.3.6.1,3.3.5.1,3.3.3.2,3.5.1,3.5.2,3.6.2.1,3.6.2.2, 3.6.2.3, 3.6.2.4, 3.6.1.3, and 3.7.1.
3.2 FSAR, Unit Two, Chapter 7, Section 7.4.3 and 7.4.5, and Chapter 5, Section 5.5.
3.3 H-26014 and H26015, RHR System P&IDs.
3.4 H-21039, RHR Service Water P&ID.
3.5 H-27635 through H27657, RHR System Elementary Diagrams.
3.6 General Electric BWR Services Information Letter No. 203, dated 10-29-76.
3.7 General Electric BWR Services Information Letter No. 175, dated 6-15-76.
( 3.8 General Electric BWR Services Information Letter No. 357, dated June, 1981.
3.9 General Electric BWR Services Information Letter No. 406, dated February 24, 1984.
3.10 General Electric BWR Services Information Letter No. 69, dated March 29,1974.
3.11 General Electric BWR Services Information Letter No. 284, dated October 23, 1978.
3.12 SX-21155, Instruction Manual RHR Service Water Pump Motor.
3.13 S-25749, RHR Pump Motor Instruction Manual.
3.14 S-53288, RHR SW Pump Spare Motor Instruction Manual.
3.16 1ST Program and Basis Document.
3.17 S62064 GE-NE-86000018-00-02, Hatch Unit 2 Reactor Pressure Vessel Temperature Control During Noble Metal Addition.
3.18 GE Report: DRF P86 00018-00, Section 8, Simultaneous Operation of RHR Shutdown Cooling System and Recirc Pump at Minimum Pump Speed for Hatch Unit 1 and 2.
3.19 34GO-OPS-087-2, Suppression Chamber Fill and Drain.
( 3.20 34SV-SUV-019-2, Surveillance Checks.
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 4 OF 284 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
RESIDUAL HEAT REMOVAL SYSTEM 34S0-E11-010-2 33.14 3.21 34S0-T41-001-2, 34S0-T 41-001-2, Core Spray and RHR Rooms Ventilation System 3.22 34S0-B31-001-2, Reactor Recirculation System 3.23 34S0-G41-003-2, Fuel Pool Cooling and Cleanup System 3.24 90AC-OAM-002-0, Scheduling Maintenance 3.25 NMP-AD-003, Equipment Clearance and Tagging 3.26 Response to Request for Engineering Review 2000-025 (RHRSW Pump Flow Limits) 3.27 EOPI SAG Appendix "C" calculations 3.28 Edwin I. Hatch Nuclear Plant Unit 2 Valve and Pump Inservice Testing Program 4.0 REQUIREMENTS 4.1 PERSONNEL REQUIREMENTS
( The number and qualification level of Operations personnel performing the sections in this procedure will be determined by the Unit Shift Supervisor.
4.2 MATERIAL AND EQUIPMENT N/A Not applicable to this procedure 4.3 SPECIAL REQUIREMENTS Independent verification, as described in 1OAC-MGR-019-0, Procedure Use AND Adherence, is required WHEN returning the Residual Heat Removal System to service following maintenance, testing, OR operation.
WHEN required, independent verification will be documented by performing the STANDBY section of this procedure.
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 5 OF 284
(
\
DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
RESIDUAL HEAT REMOVAL SYSTEM 34S0-E11-010-2 33.14 5.0 PRECAUTIONS/LIMITATIONS 5.1 PRECAUTIONS 5.1.1 An RHR loop in Shutdown Cooling will NOT automatically align for injection upon receipt of a LPCI initiation signal.
5.1.2 IF a valve must be changed from normal position for the current mode of operation, a clearance will be taken on such valve to document the change in position. IF this change in valve position effects operability of the system, Technical Specifications shall be complied with as stated in the LCO. In case of emergency, the valves' position may be changed PRIOR to the clearance being issued.
5.1.3 Room Cooler Operation must be CONFIRMED any time the RHR System is operating.
IF both SEC room coolers in a diagonal are lost, the availability of the RHRlCS pumps in that diagonal can be assured for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, !E 1.E no more than one pump is operated at a time AND the RHRlCS pump is the only significant source of heat addition to the diagonal.
5.1.4 IF a temperature of 145 00 F is reached in the respective RHR or Core Spray pump room, declare the RHR/CS RHRlCS subsystems inoperable and initiate a condition report to request that
( an EQ evaluation be performed.
5.1.5 IF a fire occurs, enter 34AB-X43-001-2, Fire Procedure, concurrently with the desired RHR operation.
5.1.6 UNLESS otherwise stated, all switches AND indications are on panel 2H11-P601 with the exception of 2E11-F040, 2E11-F022, and 2E11-F009 which are located on panel 2H11-P602.
5.1.7 WHEN using 2E11-F003A (2E11-F003B) OR 2E11-F048A (2E11-F048B) as throttling valves, monitor for vibration locally. IF vibration is occurring, adjust flow to minimize vibration.
5.1.8 IF a LOCA signal is present AND an RHR pump is secured WHILE a logic power failure exists, the affected pump's reset pushbutton must be depressed momentarily to restart the pump (RESET PBs for the 2E11-C002A & & 2E11-C002D pumps are associated with 125 VDC Division I logic power AND RESET PBs for the 2E11-C002B & & 2E11-C002C pumps are associated with 125 VDC Division IllogicII logic power (refer to Attachment 9, RHR System General Information, concerning restart of RHR pump following abnormal conditions affected by anti-pump logic and DCR OCR 98-011).
5.1.9 As time permits following LPCI initiation, notify the Health Physics Supervisor OR Foreman so that appropriate actions may be taken to control the Diagonal AND Torus areas.
5.1.10 Observe proper radiation protection practices AND procedures to maintain personnel exposure ALARA AND to limit the spread of contamination. Be alert for changing
( conditions which may require additional radiation protection.
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 6 OF 284 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
RESIDUAL HEAT REMOVAL SYSTEM 34S0-E11-010-2 33.14 5.1.11 RHR system maintenance that will render Shutdown Cooling inoperable OR restrict its performance for extended periods of time will be done in accordance with 90AC-OAM-002-0, Scheduling Maintenance.
5.1.12 Opening 2E11-F079A (2E11-F079B) and 2E11-F080A (B), WHILE in Shutdown Cooling 0
with reactor coolant temperature at OR above 212 212°F,F, may cause steam flow at 2P33-P1 01, sample panel, located in the RCIC diagonal.
5.1.13 IF Drywell Spray AND Torus Spray are initiated in the same loop of RHR AND the RHR pumps 2E11-C002A (2E11-C002B) and 2E11-C002C (2E11-C002D) TRIP, the operator must immediately CLOSE 2E11-F016A (2E11-F016B), 2E11-F021A (2E11-F021 (2E11-F021B),
B),
2E11-F027A (2E11-F027B), and 2E11-F028A (2E11-F028B), Containment Spray Valves, in that loop (refer to Attachment 9, RHR System General Information, concerning concurrent operation of Drywell and Torus Spray).
5.1.14 During single loop RHR Shutdown Cooling operations, RHR flow is normally maintained at 7700 to 8200 OR Reactor water level is maintained> 53 inches. The upper end of the flow band AND the highest reasonable water level will normally be utilized concurrently to ensure adequate core circulation.
(
NOTE: Throughout the procedure, steps that are not numbered may have a "0" box provided for the user to check off the step as the step is performed.
5.1.15 "0" may be provided for the user to check off a step when the step is completed.
5.1.16 The presence of air or gas in the suction piping could cause pump binding or damage and in the discharge piping could cause water hammer, which could lead to ruptured piping, relief valve lifting, and/or broken or damaged piping supports.
(
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 7 OF 2B4 284
(
DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
RESIDUAL HEAT REMOVAL SYSTEM 34S0-E11-010-2 33.14 5.2 LIMITATIONS 5.2.1 Except in an emergency, observe proper limitations on pump motor starts. Two starts in succession from ambient temperature OR one start from rated temperature are allowed.
For subsequent starts, allow thirty minutes running time OR sixty minutes idle time.
5.2.2 Cooling water is NOT required to operate 2E11-C002A, 2E11-C002B, 2E11-C002C & &
2E11-C002D, RHR pumps, IF the fluid media it is pumping is < 160'F.
5.2.3 Simultaneous operation of RHRSDC and Recirc is prohibited when the Unit is in MODE 5.
5.2.4 Simultaneous operation of one or both Recirc pumps at minimum speed (#1 speed limiter) and one loop of RHR SDC with one RHR pump operating is acceptable in Condition 3 or 4.
More than one RHR pump discharging into the same operating Recirc loop could cause excessive Recirc pump vibration and hydrostatic bearing damage.
5.2.5 IF Core Spray OR RHR drains are used to lower torus level per 34GO-OPS-087-2, 34GO-OPS-OB7-2, the associated Core Spray OR RHR pump must be declared inoperable.
5.2.6 DC Limitorque motors (2E11-F008 (2E11-FOOB & & 2E11-F049) have a duty cycle of three starts in five minutes AND a 50 minute cooldown.
( 5.2.7 Per the Unit 2 TRM, T1.2 Definitions:
Operations with the Potential to Drain the Reactor Vessel (OPDRV):
This is a self-defined phrase, only applicable with fuel in the reactor vessel. The following activity is an example of an OPDRV.
2E11-FOOB and Failure to maintain RHR primary containment isolation valves 2E11-F008 2E11-F009 OPERABLE per Unit 2 Technical Specifications LCOs 3.3.6.1 and 3.6.1.3 while in MODE 4 or 5. **If Required Actions of Unit 2 Technical Specifications LeOs 3.3.6.1 and 3.6.1 are satisfied, this does not apply. **
apply.**
MGR-0001 Ver. 3
SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 8 OF 284 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:
RESIDUAL HEAT REMOVAL SYSTEM 34S0-E11-010-2 33.14 6.0 PREREQUISITES 6.1 AC Electrical Power Available.
6.2 DC Electrical Power Available.
6.3 RPS and ECCS Initiation Instrument Lineup completed per 34GO-OPS-004-2.
6.4 Instrument Air System in operation per 34S0-P51-002-2.
6.5 Plant Service Water System in operation per 34S0-P41-001-2.
6.6 120 VAC RPS Power Supply System in operation per 34S0-C71-001-2.
NOTE: System valve, electrical, and instrument valve lineups provided on the attachments are to be completed as required by 34GO-OPS-003-2, Startup System Status Checklist.
6.7 RHR System Electrical Lineup complete per Attachment 2 of this procedure.
(
6.8 RHR System Valve Lineup complete per Attachment 3 of this procedure.
6.9 RHR System Instrument Valve Lineup complete per Attachment 4 of this procedure.
6.10 IF RHR system operation is a preplanned non-emergency operation, notify Health Physics at least one hour PRIOR to placing RHR system in Shutdown Cooling Mode, OR WHEN changing loops of Shutdown Cooling, to post the Diagonal AND Torus as required.
6.11 The following annunciators will activate when starting RHR pumps:
- SEC SYSTEM AUTO INITIATION SIGNAL PRESENT.
6.12 Annunciator 601-222, RHR LOW FLOW, will activate when securing RHR pumps:
(
MGR-0001 Ver. 3
E. I. HATCH I SNC PLANT E.I. Pg 176 of 284
\.
DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT l2. Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 1 of 22 PERSON(S) PERFORMING OR VERIFYING LINEUP (PRINT NAME) INITIALS
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 177 of 284 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:
RESIDUAL HEAT REMOVAL SYSTEM II 34S0-E11-010-2 33.14 ATTACHMENT ~ Att.
AU. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 2of22 2 of 22 PERSON(S) PERFORMING OR VERIFYING LINEUP (PRINT NAME) INITIALS
(
TI ME _ _ _ _ I/ DATE _ _ __
LINEUP COMPLETED: TIME REVIEWED BY: SS _ _ _ _ _ _ _ _ _ _ _ _ Time _ _ _ _-.:/ -"1 Date _ _ __
COMMENTS: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
I
\,
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
E. I. HATCH I SNC PLANT E.I. Pg PQ 178 of 284
(
DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT ~ AU.
Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 3 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Reactor Building, NE Diag, 87' 2E11-F132C RHR Torus Suction Test Vlv CLOSED 2E11-F249C ROOT Vlv to 2E11-PIR002C OPEN 2E11-F3000C ROOT Vlv to 2E11-PXN100C CLOSED LOCKED 2E11-F071C RHR Pump 2C Suction Drain CLOSED LOCKED 2E11-F018C RHR Pump 2C Manual Min. Flow OPEN 2E11-F013C RHR Pump 2C Inboard Vent CLOSED 2E11-F102C RHR Pump 2C Outboard Vent CLOSED
( 2E11-F132A RHR Torus Suction Test Vlv CLOSED 2E11-F249A ROOT Vlv to 2E11-PIR002A OPEN 2E11-3000A ROOT Vlv to 2E11-PX-N100A CLOSED 2E11-F013A RHR Pump 2A Inboard Vent CLOSED 2E11-F102A RHR Pump 2A Outboard Vent CLOSED LOCKED 2E11-F071A RHR Pump 2A Suction Drain CLOSED LOCKED 2E11-F018A RHR Pump 2A Manual Min Flow OPEN 2E11-F069A RHR Loop A to RIB Floor Drain Isolation CLOSED 2E11-F107A RHR Hx A Outlet Side Drain Inboard Isolation CLOSED 2E11-F108A RHR Hx A Outlet Side Drain Outboard Isolation CLOSED
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
HATCH I SNC PLANT E. I. HATCHI Pg 179 of 284 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:
( RESIDUAL HEAT REMOVAL SYSTEM ATTACHMENT .l.. ~
IL 34S0-E11-010-2 33.14 Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 4of22 40f22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION 2E11-F109A RHR Hx A Inlet Side Drain, Inboard Isolation CLOSED 2E11-F110A RHR Hx A Inlet Side Drain, Outboard Isolation CLOSED 2E11-F112A RHR Hx A Discharge Inboard isolation CLOSED Reactor Building, NE Diag, 87' Cont.
2E11-F113A RHR Hx A Discharge Outboard Isolation CLOSED 2E11-F079A RHR Sample Line Valve (Local PaneI2H21-P018) CLOSED 2E11-F080A RHR Sample Line Valve (Local PaneI2H21-P018)
Panel 2H21-P018) CLOSED Reactor Building, NE Diag, 96'
( 2E11-F248C ROOT Vlv, to 2E11-PI-R003C OPEN LOCKED 2E11-F072C RHR Pump 2C Discharge Drain CLOSED LOCKED 2E11-F034C RHR Pump 2C Manual Discharge OPEN 2E11-F248A ROOT Vlv to 2E11-PI-R003A OPEN LOCKED 2E11-F072A RHR Pump 2A Discharge Drain CLOSED 2E11-F130A RHR Pumps 2A, 2C Min Flow Test CLOSED 2E11-FD022 RHR Pumps 2A, 2C Discharge Drain CLOSED 2E11-F131A Loop A Disch To Suppression Pool Test CLOSED LOCKED 2E11-F002A RHR Hx A Service Water Outlet Man. Isolation OPEN 2E11-F120A RHR SW System I Flushing Supply Inboard Isolation CLOSED 2E11-F121A RHR SW System I Flushing Supply Outboard Isolation CLOSED
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
E. I. HATCH I SNC PLANT E.I. Pg 180 of 284
(
DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIl DOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT -1. 2.. Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 5 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION 2E11-F3001A RHR Hx A Shell Side Drain Isolation CLOSED 2E11-F251A Root Vlv to 2E11-CE-N001A OPEN 2E11-F252A Root Vlv to 2E11-CE-N001A OPEN 2P52-F377 Air Supply to 2E11-F053A OPEN Reactor Building, NE Diag, 96' Cont.
2E11-F294A 2E11-C002A Upper Reservoir Oil Drain CLOSED 2E11-F293A RHR Pump Motor C002A Lower Reservoir Oil Drain CLOSED 2E11-F294C 2E11-C002C Upper Reservoir Oil Drain CLOSED 2E11-F293C RHR Pump Motor C002C Lower Reservoir Oil Drain CLOSED Reactor Building, NE Diag, 106' LOCKED 2E11-F034A RHR Pump 2A Manual Discharge OPEN 2E11-FV022 RHR Pumps 2A, 2C Discharge Vent CLOSED 2E11-F124A Jockey Pumps 2A, 3A discharge to RHR OPEN 2E11-F3002A ROOT Vlv to 2E11-PI-R008A OPEN 2E11-F245A ROOT Vlv to 2E11-DPIS-N003A OPEN LOCKED 2E11-F014A RHR Hx A Service Water Inlet Man. Isolation OPEN 2E11-F256A ROOT Vlv to 2E11-PS-N017 A,C 2E11-PS-N017A,C OPEN
(
" OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 181 of 284 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:
\. RESIDUAL HEAT REMOVAL SYSTEM II 34S0-E11-010-2 33.14 ATTACHMENT ~ AU. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 6 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION 2P52-F375 Spare CLOSED 2P52-F377-AS 1 Air Knob to 2E11-F053A OPEN ** **
Reactor Building, NE Diag, 118' 2E11-F244A ROOT Vlv to 2E11-DPIS-N003A OPEN ** **
2E11-F250A ROOT Vlv to 2E11-PI-N026A OPEN ** **
2E11-FV005A RHR A Hx High point Vent CLOSED Torus Room 118' 2E11-F257B 11 j ROOT Vlv to 2E11-FT-N007B (AZ. 11) OPEN
( 2E11-F258B 11 j ROOT Vlv to 2E11-FT-N007B (AZ. 11) OPEN 2E11-F240B 22j ROOT Vlv to 2E11-PS-N022B (AZ. 22) OPEN 2E11-F241 B 33j ROOT Vlv to 2E11-FT-N015B (AZ. 33) OPEN 2E11-F242B 33j ROOT Vlv to 2E11-FT-N015B (AZ. 33) OPEN 2E11-F076B Loop B Crosstie Test (AZ. 56j
- 56) CLOSED 2E11-F129B Loop B Suppression Pool Test (AZ. 56j
- 56) CLOSED 2E11-FV001 RHR Hx B Outlet Vent (AZ. 56j
- 56) CLOSED ** **
2E11-FD001 56j RHR Hx B Outlet Drain (AZ. 56) CLOSED 2E11-FV002 56j RHR Hx B Inlet Vent (AZ. 56) CLOSED ** **
2E11-FD002 RHR Hx B Outlet Drain (AZ. 56j
- 56) CLOSED ** **
2E11-FV101 348j RCIC Suction from RHR Vent (AZ. 348) CLOSED OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg PQ 182 of 284 DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E 34S0-E11-010-2 11-0 10-2 Version No:
33.14 ATTACHMENT ~ Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 7 of 22 7of22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Torus Room, 118' Cont.
2E11-F241A ROOT Vlv to 2E11-FT-N015A (AZ. 326j OPEN 2E11-F242A ROOT Vlv to 2E11-FT-N015A (AZ. 326j OPEN 2E11-F240A ROOT Vlv to 2E11-PS-N022A (AZ. 335j OPEN 2E11-F066 Shutdown Cooling Sample Line (AZ. 348j CLOSED 2E11-F257A ROOT Vlv to 2E11-FT-N007A (AZ. 348j OPEN 2E11-F258A ROOT Vlv to 2E11-FT-N007A (AZ. 348j OPEN 2E11-F076A Loop A Crosstie Test (AZ. 348j CLOSED 2E11-F129A Loop A Suppression Pool Test (AZ. 348j CLOSED
( 2E11-FD003 RHR Hx A Outlet Drain (AZ. 348j CLOSED 2E11-FV004 RHR Hx A Inlet Vent (AZ. 348j CLOSED 2E11-FD004 RHR Hx A Inlet Drain (AZ. 348j CLOSED 2E11-F255 ROOT Vlv to 2E11-PS-N018 (AZ. 348j OPEN LOCKED 2P52-F361 Air Supply to 2E11-AOV-F065A (87' AZ. 340j OPEN LOCKED 2P52-F367 Air Supply to 2E11-AOV-F065C (87' AZ. 340j OPEN LOCKED 2P52-F364 Air Supply to 2E11-AOV-F065B (87' AZ. 20°)
OPEN LOCKED 2P52-F370 Air Supply to 2E11-AOV-F065D (87' AZ. 20°)
OPEN LOCKED 2P52-F362 2E11-F065A Accumulator Drain CLOSED LOCKED 2P52-F368 2E11-F065C Accumulator Drain CLOSED
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 183 of 284 DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT ..l..
.l... Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 88of22 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Torus Room, 118' Cont.
LOCKED 2P52-F365 2E11-F065B Accumulator Drain CLOSED LOCKED 2P52-F371 2E11-F065D Accumulator Drain CLOSED Reactor Building 130RJR19 Loop A Condensate Flush Supply Outboard 2E11-F081A CLOSED Isolation Loop A Condensate Flush Supply Inboard 2E11-F082A CLOSED Isolation Shutdown Cooling Condensate Flush Supply 2E11-F083 CLOSED Inboard Isolation Shutdown Cooling Condensate Flush Supply 2E11-F084 CLOSED Outboard Isolation
( 2E11-F085 Cross header Loop AlB Condensate Flush CLOSED Supply Outboard Isolation Cross header Loop A Condensate Flush Supply 2E11-F086A CLOSED Inboard Isolation Reactor Building 150RER15 2E11-F043A ROOT Vlv to 2E11-PT-N094A OPEN 2E11-F142 Drywell Pressure Sensing Line Test CLOSED 2E11-F043C ROOT Vlv to 2E11-PT-N094C OPEN 2E11-F141 Drywell Pressure Sensing Line Test CLOSED 2P52-F590 Air Supply to 2E11-F041A OPEN 2P52-F592 Air Supply to 2E11-F041C OPEN OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 184 of 284
(
DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT l Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 9 of 22 90f22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Reactor Building, SE Diag, 87' 2E11-F132D RHR Torus Suction Test Vlv CLOSED 2E11-F249D ROOT Vlv to 2E 11-PI-R002D OPEN 2E11-F3000D ROOT Vlv to 2E 2E11-PX-N100D 11-PX-N 1000 CLOSED LOCKED 2E11-F071D 2E11-F071 D RHR Pump 20 Suction Drain CLOSED 2E11-F013D RHR Pump 20 Inboard Vent CLOSED 2E11-F102D RHR Pump 20 Outboard Vent CLOSED LOCKED 2E11-F018D RHR Pump 20 Manual Min Flow OPEN 2E11-F132B RHR Torus Suction Test Vlv CLOSED
(
2E11-F249B ROOT Vlv to 2E11-PI-R002B OPEN 2E11-F3000B ROOT Vlv to 2E 11-PX-N 1OOB 2E11-PX-N100B CLOSED 2E11-F013B RHR Pump 2B Inboard Vent CLOSED 2E11-F102B RHR Pump 2B Outboard Vent CLOSED 2E11-F070 Core Spray AND RHR Drain to Radwaste Isolation CLOSED 2E11-F069B RHR Loop B to Reactor Building Floor Drain Isolation CLOSED LOCKED 2E11-F071B 2E11-F071 B RHR Pump 2B Suction Drain CLOSED LOCKED 2E11-F018B RHR Pump 2B Manual Min Flow OPEN 2E11-F124B Jockey Pumps 2B, 3B Discharge to RHR OPEN 2E11-F3002B ROOT Vlv to 2E11-PI-R008B 2E 11-PI-R008B OPEN
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 185 of 284 DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT .2. ~ Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 10 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION 2E11-F107B RHR Hx B Outlet Side Drain, Inboard Isolation CLOSED 2E11-F108B RHR Hx B Outlet Side Drain, Outboard Isolation CLOSED 2E11-F109B RHR Hx B Inlet Side Drain, Inboard Isolation CLOSED 2E11-F079B RHR Sample Line Valve (Local Panel 2H21-P021) CLOSED 2E11-F080B RHR Sample Line Valve (Local Panel 2H21-P021) CLOSED 2E11-F110B RHR Hx B Inlet Side Drain, Outboard Isolation CLOSED 2E11-F112B RHR Hx B Discharge Inboard Isolation CLOSED 2E11-F113B RHR Hx B Discharge Outboard Isolation CLOSED Reactor Building, SE Diag, 96' 2E11-F248D ROOT Vlv to 2E11-PI-R003D OPEN LOCKED 2E11-F072D RHR Pump 2D Discharge Drain CLOSED LOCKED 2E11-F034D RHR Pump 2D Manual Discharge OPEN 2E11-F248B ROOT Vlv to 2E11-PI-R003B OPEN LOCKED 2E11-F072B RHR Pump 2B Discharge Drain CLOSED 2E11-F130B RHR Pumps 2B, 2D Min. Flow Test CLOSED 2E11-FD021 RHR Pumps 2B, 2D Discharge Drain CLOSED 2E11-F131B Loop B Disch to Suppression Pool Test CLOSED LOCKED 2E11-F002B RHR Hx B Service Water Outlet Man. Isolation OPEN
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 186 of 284
(
\
DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT ~ Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 11 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Reactor Building, SE Diag, 96', cont 2E11-F120B RHR SW System II Flushing Supply Inboard Isolation CLOSED 2E11-F121B 2E11-F121 B RHR Sw System II Flushing Supply Outboard Isolation CLOSED 2E11-F3001B 2E11-F3001 B RHR Hx B Shell Side Drain Isolation CLOSED 2E11-F251 B 2E11-F251B Vlv to 2E11-CE-N001B ROOT Vlvto 2E11-CE-N001 B OPEN 2E11-F252B ROOT Vlv to 2E11-CE-N001 B OPEN 2P52-F378 Air Supply to 2E11-F053B OPEN RHR 2B Motor Lower Reservoir Oil Drain (Inside pump 2E11-F293B CLOSED shroud next to pump coupling)
RHR 2D Motor Lower Reservoir Oil Drain (Inside pump 2E11-F293D CLOSED shroud next to pump coupling)
RHR 2B Motor Upper Reservoir Oil Drain
( 2E11-F294B (near pump motor)
CLOSED RHR 2D Motor Upper Reservoir Oil Drain (Inside pump 2E11-F294D CLOSED shroud next to pump coupling)
Reactor Building, SE Diag, 106' 2E11-F034B RHR Pump 2B Manual Discharge LOCKED nPFN OPEN 2E11-FV021 RHR Pumps 2B, 2D Discharge Vent CLOSED 2E11-F245B ROOT Vlv to 2E11-DPIS-N003B OPEN 2E11-F014B RHR Hx B Service Water Inlet Man. Isolation LOCKED OPFN 2E11-F256B ROOT Vlv to 2E11-PS-N017B,D OPEN 2P52-F376 Spare CLOSED 2P52-F378-AS 1 Air Knob to 2E11-F053B OPEN Reactor Building SE Diag, 118' 2E11-F244B ROOT Vlv to 2E11-DPIS-N003B OPEN 2E11-F250B ROOT Vlv to 2E11-PT-N026B OPEN 2E 11-FV005B 2E11-FV005B RHR B Hx High point Vent CLOSED OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 187 of 284 DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 34S0-E 11-0 10-2 Version No:
33.14 A ITACHMENT -1..
ATTACHMENT ~ Att. Pg.
AU.
TITLE: RHR SYSTEM VALVE LINEUP 12 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Reactor Building 150RHR21 2E11-F043B ROOT Vlv to 2E11-PT-N094B OPEN 2E11-F146 Drywell Pressure Sensing Line Test CLOSED 2E11-F043D ROOT Vlv to 2E11-PT-N094D OPEN 2E11-F145 Drywell Pressure Sensing Line Test CLOSED 2P52-F140A Inst. Air Isolation to 2E11-F041 B OPEN 2P52-F591 Air Supply to 2E 11-F041 B OPEN 2P52-F140B IInst.
nst. Air Isolation to 2E 2E11-F041 11-F041 D 0 OPEN 2P52-F593 Air Supply to 2E 11-F041 D 0 OPEN Reactor Building Drywell Access
(
2E11-F058A Loop A Test Inboard Isolation CLOSED 2E11-F059A Loop A Test Outboard Isolation CLOSED 2E11-F058B Loop B Test Inboard Isolation CLOSED 2E11-F059B Loop B Test Outboard Isolation CLOSED 2E11-F063 Shutdown Cooling Test Inboard Isolation CLOSED 2E11-F064 Shutdown Cooling Test Outboard Isolation CLOSED 2E11-F253 ROOT Vlv to 2E11-LS-N041 OPEN 2E11-F254 ROOT Vlv to 2E11-LS-N041 OPEN LOCKED 2E11-F286 Inboard Isolation to 2T45-D016 CLOSED LOCKED 2E11-F287 Outboard Isolation to 2T 45-0016 2T45-D016 CLOSED 2E11-F288 Drain to 2E11-LS-N041 CLOSED 2T45-F024 Flush/Test to T45-D016 CLOSED 2E11-F036A Loop A Air Test Connection CLOSED OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 188 of 284 DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT .l... Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 13 of 22 NORMAL NORMAL I NUMBER DESCRIPTION POSITION I CHECKED VERIFIED Reactor Building 130RHR20 Loop B Condensate Flush Supply Inboard 2E11-F082B CLOSED Isolation Loop B Condensate Flush Supply Outboard 2E11-F081B 2E11-F081 B CLOSED Isolation Cross header Loop B Condensate Flush 2E11-F086B CLOSED Supply Inboard Isolation Reactor Building 158RHR23 (RWCU Hx Room) 2E11-F238 Spare ROOT Vlv CLOSED 2E11-F239 Spare ROOT Vlv CLOSED 2E11-F036B Loop B Air Test Connection CLOSED 2E11-FD023 Loop B Containment Spray Line Drain CLOSED
(
2E11-FV023 Fuel Pool to RHR Line Vent CLOSED Reactor Building 185RHR19 2E11-F061 Head Spray Test Inboard Isolation CLOSED 2E11-F062 Head Spray Test Outboard Isolation CLOSED 2E11-F236 ROOT Vlv to 2E11-LSN040 OPEN 2E11-F237 2E 11-LSN040 ROOT Vlv to 2E11-LSN040 OPEN LOCKED 2E11-F284 Inboard Isolation to 2T45-D015 CLOSED LOCKED 2E11-F285 Outboard Isolation to 2T45-D015 2T45-DO 15 CLOSED 2E11-F289 Drain to 2E11-LSN040 CLOSED
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 189 of 284 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:
I
\ RESIDUAL HEAT REMOVAL SYSTEM II 34S0-E11-010-2 33.14 ATTACHMENT .2 ~ Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 14 of 22 NUMBER DESCRIPTION NORMAL CHECKED VERIFIED POSITION Drywell 193'6" 2E11-F147 (AZ. 330j Head Spray Drain Vlv (Inbd) (AZ CLOSED 2E11-F148 (AZ. 330j Head Spray Drain Vlv (Outbd) (AZ CLOSED Drywe1l128' 2E11-F060A (AZ. 285j RHR Injection Manual Isolation (AZ LOCKED OPEN 2E11-F127A By Pass 2E11-F122A Disch Outbd Test (AZ 295j CLOSED 2E11-F128A By Pass 2E11-F122A Disch Inbd Test (AZ 295j CLOSED 2E11-F155A 2E11-F060A Body Drain (Inbd) (AZ(AZ. 285j CLOSED 2E11-156A 2E11-F060A Body Drain (Outbd) (AZ (AZ. 285j CLOSED 2E11-F161A 2E11-F060A Vlv Stem Leakoff (AZ 285j OPEN
( Drywe1l128' 2E11-F162A 2E11-F050A Vlv Stem Leakoff (AZ(AZ. 295j OPEN 2E11-F060B RHR Injection Manual Isolation (AZ 75j LOCKED OPEN 2E11-F127B By Pass 2E11-F122B Disch. Inbd Test (AZ 65j CLOSED 2E11-F128B By Pass 2E11-F122B Disch. Outbd Test (AZ 65j CLOSED 2E11-F155B 2E11-F060B Body Drain (Inbd) (AZ(AZ. 75j CLOSED 2E11-F156B 2E11-F060B Body Drain (Outbd) CLOSED 2E11-F161B 2E11-F060B Vlv Stem Leakoff OPEN 2E11-F067 SID Cooling Suction Manual Isolation LOCKED OPEN 2E11-F143 SID Cooling Suction Inbd Test Conn. CLOSED 2E11-F144 SID Cooling Suction Outbd Test Conn. CLOSED 2E11-F3091 D/W Pene. 12 Thermal Press. Relief Isol. Valve LOCKED OPEN
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 190 of 284
(
DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIl DOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT .1.. 2- Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 15 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Service Water Intake Building 2E11-F206A Pump 2E11-C001A Prelube Water Shutoff LOCKED OPEN 2E11-F222A Pump 2E11-C001A Prelube Water Vent OPEN 2E11-F202A Pump 2E11-C001A Seal Water Shutoff LOCKED OPEN 2E11-F207A RHR SW Pump 2A Minimum Flow Valve SEE NOTE 1 2E11-F265A ROOT Valve to PCV 2E11-F207 2E11-F207A A OPEN 2E11-F264A ROOT Valve to 2E11-PI-ROO4A CLOSED 2E11-F012A Pump 2E11-C001A Discharge LOCKED OPEN 2E11-F263A ROOT Valve to 2E11-PT-N048A OPEN 2E11-FD005 Pump 2E11-C001A Discharge Drain CLOSED 2P52-F602H Instrument Air to 2E11-F207A LW OPEN
(
2E11-F296A Pump 2E11-C001A Air Release Valve Isol. LOCKED OPEN 2E11-F206C 2E11-C001C Pump 2E11-C001 C Prelube Water Shutoff LOCKED OPEN 2E11-F222C Pump 2E11-C001C Prelube Water Vent OPEN 2E11-F202C Pump 2E11-C001C 2E11-C001 C Seal Water Shutoff LOCKED OPEN 2E11-F207C RHR SW Pump 2C Minimum Flow Valve SEE NOTE 1 2E11-F265C ROOT Valve to PCV 2E11-F207C OPEN 2E11-F264C ROOT Valve to 2E11-PI-ROO4C CLOSED 2E11-F012C Pump 2E11-C001C Discharge LOCKED OPEN 2E11-FD006 Pump 2E11-C001C Discharge Drain CLOSED 2P52-F602G Instrument Air to 2E11-F207C LWOPEN LW OPEN 2E11-F296C Pump 2E11-C001 C Air Release Valve Isolation LOCKED OPEN 2E11-F296B Pump 2E11-C001 B Air Release Valve Isolation LOCKED OPEN NOTE 1 Normal valve position is CLOSED. Normal handwheel position is FULLY CCW.
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 191 of 284
(
DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT ~ Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 16 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Service Water Intake Building 2E11-F296D Pump 2E11-C001 D Air Release Valve Isolation LOCKED OPEN 2E11-F206B Pump 2E11-C001 B Prelube Water Shutoff LOCKED OPEN 2E11-F222B Pump 2E11-C001 B Prelube Water Vent OPEN 2E11-F202B 2E11-C001B Pump 2E11-C001 B Seal Water Shutoff LOCKED OPEN 2E11-F207B RHR SW Pump 2B Minimum Flow Valve SEE NOTE 1 2E11-F265B ROOT Valve to PCV 2E11-F207B OPEN 2E11-F264B ROOT Valve to 2E11-PI-R004B CLOSED 2E11-F012B Pump 2E11-C001 B Discharge LOCKED OPEN 2E11-FD007 Pump 2E11-C001 B Discharge Drain CLOSED
( 2P52-F602F Instrument Air to 2E11-F207B LW OPEN 2E11-F263B ROOT Valve to 2E11-PT-N048B OPEN 2E11-F206D Pump 2E11-C001 D Prelube Water Shutoff LOCKED OPEN 2E11-F222D Pump 2E11-C001 D Prelube Water Vent OPEN 2E11-F202D 2E11-C001D Pump 2E11-C001 D Seal Water Shutoff LOCKED OPEN 2E11-F207D RHR SW Pump 2D Minimum Flow Valve SEE NOTE 1 2E11-F265D ROOT Valve to PCV 2E11-F207D OPEN 2E11-F264D ROOT Vlv to 2E 11-PI-R004D CLOSED 2E11-F012D 2E11-C001D Pump 2E11-C001 D Discharge LOCKED OPEN 2E11-FD008 Pump 2E11-C001D 2E 11-C001 D Discharge Drain CLOSED 2P52-F602E Instrument Air to 2E11-F207D LW OPEN 2E11-F259A ROOT Vlv to 2E11-DPIS-N045A OPEN 2E11-F260A ROOT Vlv to 2E11-DPIS-N045A OPEN NOTE 1 Normal valve position is CLOSED. Normal handwheel position is FULLY CCW.
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 192 of 284 DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM ATTACHMENT l~
IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 17 of 22 NORMAL NUM8ER NUMBER DESCRIPTION CHECKED VERIFIED POSITION Service Water Intake Building, 8uilding, Cont STR 2E11-DOO2A LOCKED 2E11-F115A 2E11-D002A Inlet Isolation OPEN
- LOCKED 2E11-F114A 2E 11-DOO2A Outlet Isolation STR 2E11-D002A OPEN 2E11-FD011 STR 2E11-D002A Inlet Drain CLOSED 2E11-FD014 STR 2E11-D002A and 3A Outlet Drain CLOSED LOCKED 2E11-F117A STR 2E11-D003A Inlet Isolation OPEN STR 2E11-D003A Outlet Isolation *LOCKED 2E11-F116A CLOSED 2E11-FD012 STR 2E11-D003A Inlet Drain CLOSED 2E11-F259B ROOT Vlv to 2E 11-DPIS-N0458 2E11-DPIS-N045B OPEN 2E11-F260B ROOT Vlv to 2E11-DPIS-N0458 2E11-DPIS-N045B OPEN
( 2E11-F115B STR 2E11-D0028 2E11-D002B Inlet Isolation LOCKED OPEN
- LOCKED 2E11-F114B STR 2E 11-00028 11-D002B Outlet Isolation OPEN 2E11-FD010 STR 2E11-D0028 2E11-D002B Inlet Drain CLOSED 2E11-F261 B 2E11-F261B ROOT Vlv to 2E11-DPIS-N0468 2E11-DPIS-N046B OPEN 2E11-F262B ROOT Vlv to 2E11-DPIS-N0468 2E11-DPIS-N046B OPEN 2E11-FD013 STR's 2E11-D0028 2E11-D002B AND 38 3B Outlet Drain CLOSED LOCKED 2E11-F117B STR 2E11-D0038 2E11-D003B Inlet Isolation OPEN
- LOCKED 2E11-F116B STR 2E 11-00038 11-D003B Outlet Isolation CLOSED 2E11-FD009 STR 2E11-D0038 2E11-D003B Inlet Drain CLOSED 2E11-F261A ROOT Vlv to 2E11-DPIS-N046A OPEN 2E11-F262A ROOT Vlv to 2E11-DPIS-N046A OPEN
- It is not necessary to reposition this valve IF the corresponding valve in the other train is in the opposite position. IF positions are different than specified, note actual position AND initial.
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 193 of 284
(
DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT .l...~ Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 18 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Service Water Intake Building, Cont.
Cant.
Pump 2E11-C001A Motor Cooling Water Iso. From LOCKED 2P41-F923A PSW (Supply) OPEN 2E11-C001 B Motor Cooling Water Iso.
Pump 2E11-C001B From LOCKED 2P41-F923B PSW (Supply) OPEN Pump 2E11-C001C Motor Cooling Water Iso. From LOCKED 2P41-F923C PSW (Supply) OPEN Pump 2E11-C001D Motor Cooling Water Iso. From LOCKED 2P41-F923D PSW (Supply) OPEN SEE 2P41-F1600A RHRSW Motor "A" Cooling Water Four-way Valve NOTE 1 SEE 2P41-F1600B RHRSW Motor "B" Cooling Water Four-way Valve NOTE 1 SEE 2P41-F1600C RHRSW Motor "C" Cooling Water Four-way Valve NOTE 1 SEE
( 2P41-F1600D RHRSW Motor "0" "D" Cooling Water Four-way Valve Pump 2E11-C001A Motor Cooling Water Return NOTE 1 LOCKED 2P41-F924A Isolation to PSW OPEN Pump 2E11-C001 B Motor Cooling Water Return LOCKED 2P41-F924B Isolation to PSW OPEN Pump 2E11-C001C Motor Cooling Water Return LOCKED 2P41-F924C Isolation to PSW OPEN Pump 2E11-C001 D 0 Motor Cooling Water Return Iso. To LOCKED 2P41-F924D PSW OPEN RHR SW Pump 2E11-C001A Motor Upper Reservoir Oil 2E11-F292A CLOSED Drain RHR SW Pump 2E11-C001A Motor Lower Reservoir Oil 2E11-F291A CLOSED Drain RHR SW Pump 2E11-C001 B Motor Upper Reservoir Oil 2E11-F292B CLOSED Drain RHR SW Pump 2E11-C001 B Motor Lower Reservoir Oil 2E11-F291 B CLOSED Drain 2E11-C001C RHR SW Pump 2E11-C001 C Motor Upper Reservoir Oil 2E11-F292C CLOSED Drain RHR SW Pump 2E11-C001 C Motor Lower Reservoir Oil 2E11-F291C CLOSED Drain NOTE 1: Valve has two acceptable positions. The valve must be CONFIRMED to be against one of the stops which will CONFIRM that the valve is in one of the acceptable positions.
(
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E.I. HATCH I E. I. HATCH[ Pg 194 of 284
(
DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM II DOCUMENT NUMBER:
34S0-E 11-0 10-2 34S0-E11-010-2 Version No:
33.14 ATTACHMENT .l.. AU. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 19 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Service Water Intake Building, Cont.
RHR SW Pump 2E11-C001 D Motor Upper Reservoir Oil 2E11-F292D CLOSED Drain RHR SW Pump 2E11-C001 D Motor Lower Reservoir Oil 2E11-F291 D 2E11-F291D CLOSED Drain Control Room Pane12H11-P602 2E11-F009 SDC Suction Vlv CLOSED 2E11-F040 RHR RH R To Radwaste Vlv CLOSED 2E11-F022 Rx Head Spray Vlv CLOSED Control Room Pane12H11-P601 Loop A 2E11-F080A RHR Sample Line Valve (Outboard) CLOSED
( 2E11-F079A RHR Sample Line Valve (Inboard) CLOSED 2E11-F075A RHRSW To RHR Crosstie Vlv CLOSED 2E11-F060A RHR Injection Vlv OPEN 2E11-F017A RHR Outbd Inj Vlv OPEN 2E11-F047A Hx Inlet Vlv OPEN 2E11-F004A Torus Suction Vlv OPEN 2E11-F068A Hx A Disch Vlv CLOSED 2E11-F015A RHR Inbd Inj Vlv CLOSED 2E11-F003A Hx Outlet Vlv OPEN 2E11-F006A Shutdown Cooling Vlv CLOSED 2E11-F048A Hx Bypass Vlv OPEN 2E11-F004C Torus Suction Vlv OPEN
{
\
OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 195 of 284 DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIl DOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT ~ Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 20 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Control Room Pane12H11-P601 Loop A 2E11-F007A Min FlowVlv OPEN 2E11-F010 RHR Crosstie Vlv CLOSED 2E11-F006C Shutdown Cooling Vlv CLOSED 2E11-F065A Torus Suction Vlv OPEN 2E11-F016A Cnmt. Spray Outbd Vlv CLOSED 2E11-F028A Torus Spray OR Test Vlv CLOSED 2E11-F026A RHR Hx To RCIC Vlv
- CLOSED 2E11-F103A HxVent Vlv CLOSED 2E11-F065C Torus Suction Vlv OPEN
(
2E11-F021A Cnmt Spray Inbd Vlv CLOSED 2E11-F027A Torus Spray Vlv CLOSED 2E11-F011A RHR Hx To Torus Vlv CLOSED 2E11-F104A Hx Vent Vlv HxVent CLOSED 2E11-F024A Full Flow Test Line Vlv CLOSED 2E11-F073A RHRSW Crosstie Vlv CLOSED 2E11-F122A Check Vlv F050A Bypass Vlv CLOSED 2E11-F119A Serv Wtr Crosstie Vlv CLOSED 2E11-F049 RHR To Radwaste Vlv CLOSED 2E11-F023 Rx Head Spray Vlv
SNC PLANT E. I. HATCH I Pg 196 of 284 PQ JI DOCUMENT TITLE: DOCUMENT NUMBER: Version No:
RESIDUAL HEAT REMOVAL SYSTEM 34S0-E11-010-2 33.14 ATTACHMENT .l..
~ Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 21 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Pane12H11-P601 Control Room Panel 2H 11-P601 Loop B 2E11-F080B RHR Sample Line Vlv (Outbd) CLOSED 2E11-F079B (lnbd)
RHR Sample Line Vlv (Inbd) CLOSED 2E11-F075B RHRSW To RHR Crosstie Vlv CLOSED 2E11-F060B RHR Injection Vlv OPEN 2E11-F017B RHR Outbd Inj Vlv OPEN 2E11-F047B Inlet Vlv Hx InletVlv OPEN 2E11-F004B Torus Suction Vlv OPEN 2E11-F068B Hx B Disch Vlv CLOSED
( 2E11-F015B RHR Inbd Inj Vlv CLOSED 2E11-F003B Hx Outlet Vlv OPEN 2E11-F006B Shutdown Cooling Vlv CLOSED 2E11-F048B Hx Bypass Vlv OPEN 2E11-F004D Torus Suction Vlv OPEN 2E11-F007B Min FlowVlv OPEN 2E11-F006D Shutdown Cooling Vlv CLOSED 2E11-F065B Torus Suction Vlv OPEN 2E11-F016B Cnmt Spray Outbd Vlv CLOSED 2E11-F103B HxVent Hx Vent Vlv CLOSED 2E11-F065D Torus Suction Vlv OPEN OPS-0277 Ver. 20.3 G16.030 MGR-0009 Ver. 4
SNC PLANT E. I. HATCH I Pg 197 of 284
\
DOCUMENT TITLE:
RESIDUAL HEAT REMOVAL SYSTEM IIDOCUMENT NUMBER:
34S0-E11-010-2 Version No:
33.14 ATTACHMENT .l ~ Att. Pg.
TITLE: RHR SYSTEM VALVE LINEUP 22 of 22 NORMAL NUMBER DESCRIPTION CHECKED VERIFIED POSITION Control Room Pane12H11-P601 Loop B, cont 2E11-F028B Torus Spray or Test Vlv CLOSED 2E11-F026B RHR Hx To RCIC Vlv
- CLOSED 2E11-F021 2E11-F021BB Cnmt Spray Inbd Vlv CLOSED 2E11-F027B Torus Spray Vlv CLOSED 2E11-F011B RHR Hx To Torus Vlv CLOSED 2E11-F104B HxVent Vlv CLOSED 2E11-F024B Full Flow Test Line Vlv CLOSED 2E11-F073B RHRSW Crosstie Vlv CLOSED
.\ 2E11-F122B Check Vlv F050B Bypass Vlv CLOSED 2E11-F119B Serv Wtr Crosstie Vlv CLOSED Control Room Pane12H11-P657 Panel 2H 11-P657 2E11-F041A Drwl Press Switch Xmtr Line Isol OPEN 2E11-F041 2E11-F041C C Drwl Press Switch Xmtr Line Isol OPEN Control Room Pane12H11-P654 2E11-F041 B Drwl Press Switch Xmtr Line Isol OPEN 2E11-F041D Drwl Press Switch Xmtr Line Isol OPEN
DRAFT
(
Southern Nuclear E. I. Hatch Nuclear Plant Operations Training JPM Admin 6, RO Only TITLE DURING AN EMERGENCY, PERFORM A PROMPT OFFSITE DOSE ASSESSMENT CALCULATION AUTHOR MEDIA NUMBER TIME Kip Wainwright LR-JP-25306-1 0 30.0 Minutes
.)lEC()l\f:M:E~)jE])B¥:
RECOMMENDED BY * * * * * * **.**.*APPROVEDBY*.**
APPROVED BY DATE NIR D.H. GIDDENS 08-04-2005 sM Energy to Serve Your World sM
SOUTHERN NUCLEAR OPERATING COMPANY PLANT E. I. HATCH Page 1 of 1 FORM TITLE: TRAINING MATERIAL REVISION SHEET Program/Course Code: OPERATIONS TRAINING Media Number: LR-JP-25306
- .**. **.il*p~_*l\T~;,* ****h~t.,;) i*****. . . . ************f;.;,.t.iiX~_1~._ri< . . .*. . .. . *<i* ......../. !A~~h? ..'s.
Rev. No. . ** .*
Author's Supv's *.*.**.*
.'$JlPy'f*
Date Reason for Revision j * ......
. . j\."),,**.* I</r.. ~?.*. *. <>>.> .. . . . . . . .*. . . ....;.....*..*. . . **'**/.***. . . .***** ** . . *.*.i)}j....... .........j>i.* .iUlitialSInitials Initials lniti1J.ls\
00 05/02/94 Initial development RAB DHG 01 09/06/96 Format change, convert to use SPDS information RAB DHG 02 09/12/96 Editorial, correct MIDAS result numbers for peak RAB RSG TEDE and site boundaries 03 10/03/97 Revised based on 1997 annual exam comments. SCB DHG 04 09116/98 Revised due to procedure/MIDAS program revision. SCB DHG 05 01121/99 01121199 Revised per feedback from EP. SCB DHG 06 03/21100 Format modification, change time allowance based on RAB DHG running average, change MIDAS for due to revision 07 11106/00 Include objective number RAB DHG 08 03/25/02 Include initial operator statement RAB RAB
( 09 3/22/05 Changed RO to NPO, added statement ensuring the ELl/CRC ELJ/CRC DHG procedure is the current version, changed SOS to SM, addressed the use ofTRN-052 and TRN-0146 flowchart and deleted the "s" from the procedure numbers. Modified steps and results to incorporate the use of the new MIDAS Windows based software.
10 8/04/05 Corrected "Alarms Illuminated" information to match BKW DHG plant conditions given in attachments.
Revised Initial License statement for successful completion.
LR-JP-25306-10 LR-JP-25306-l0 ofl5 Page 1 of15
(
UNIT 1 (X) UNIT 2 (X)
TASK TITLE: DURING AN EMERGENCY, PERFORM A PROMPT OFFSITE DOSE ASSESSMENT CALCULATION JPMNUMBER: LR-JP-25306-10 TASK STANDARD: Offsite The task shall be complete when the Total Off site Dose Rate has been calculated per 73EP-EIP-018-0 and the SMIED has been informed of the correct radiological assessment.
TASK NUMBER: 200.060 OBJECTIVE NUMBER: 200.060.0 PLANT HATCH JTA IMPORTANCE RATING:
NPO 4.57 SRO 3.92 KIA CATALOG NUMBER: 2.4.39 KIA CATALOG JTA IMPORTANCE RATING:
NPO 3.9 SRO 3.8 OPERATOR APPLICABILITY: Nuclear Plant Operator (NPO)
IGENERAL
REFERENCES:
Unit 1 & 2 34AB-DII-001-2 (current version) 73EP-EIP-018-0 (current version)
TRN-0052 (current version)
TRN-0146 (current version)
IREQUIRED MATERIALS: Unit 1 & 2 73EP-EIP-018-0 (current version)
TRN-0052 (current version)
TRN-0146 (current version)
Computer with the MIDAS program APPROXIMATE COMPLETION TIME: 30.0 Minutes SIMULATOR SETUP: NIA
(,
UNITl&2 UNIT 1 &2 READ TO THE OPERATOR OPERATOR INITIAL CONDITIONS:
- 1. A pipe break has occurred causing a Primary System discharge outside Primary Containment.
- 2. OFFGAS VENT RADIATION HIGH-HIGH is alarming on Unit 1.
- 3. 34AB-D11-001-2, 34AB-D 11-00 1-2, "Radioactivity Release Control," is in progress.
- 4. SPDS is available for use.
- 5. The Shift Manager (SM) is acting in the Emergency Director's position.
( 6. (A TWS), all rods did not fully insert.
The reactor failed to scram (ATWS),
INITIATING CUES:
DETERMINE Total Offsite Dose Rate per 73 73EP-EIP-018-0, EP-EIP-O 18-0, then:
NOTIFY the SM/ED if a release is progress.
SM/ED if emergency classification should be addressed.
NOTIFY the SMIED
(
LR-JP-25306-l0 LR-JP-25306-10 of 15 Page 3 of15 PERFO~ANCESTEP STANDARD SAT/UNSAT COMMENTS For INITIAL Operator Programs:
For OJT/OJE; ALL PROCEDURE STEPS must be completed for Satisfactory Performance.
For License Examinations; ALL CRITICAL STEPS must be completed for Satisfactory Performance.
START TIME:_ _ __
- 1. Operator locates the correct Operator has LOCATED SAT I UNSAT procedure. 73EP-EIP-0 18-0.
- 2. Operator reviews procedure's Operator has REVIEWED the SAT I UNSAT Precautions and Limitations. Precautions and Limitations section of the procedure.
73 EP-EIP-O 18-0 NOTE: 73EP-EIP-0 18-0 directs the operator to perform procedure steps or use the Job Aid in TRN-0146, "Prompt Offsite Dose Assessment Flowchart".
Either method is acceptable.
(
PROMPT: IF addressed by the operator, PROVIDE the attached Annunciator Status sheet.
PROMPT: IF addressed by the operator, INDICATE that the following annunciators are NOT ILLUMINATED.
1Hll-P650-2 IHII-P650-2 MAIN STACK EFFLUENT ANY COLLECTOR RADN LEVEL MAXIMUM MAXTh1UM MAIN STACK EFFLUENT ALL COLLECTORS RADN LEVEL MAXIMUM MAXTh1UM RB VENT EFFL ANY COLLECTOR RADN LEVEL MAX RB VENT EFFL ALL COLLECTORS RADN LEVEL MAX 2Hll-P650-2 2HII-P650-2 RB VENT EFFL ANY COLLECTOR RADN LEVEL MAX RB VENT EFFL ALL COLLECTORS RADN LEVEL MAX PROMPT: INFO~ the IF addressed by the operator, as the Shift Supervisor, INFORM
(, operator that SPDS is available.
(*
- Indicates critical step)
(**
LR-JP-25306-l0 LR-JP-25306-10 of 15 Page 4 of15 PERFORMANCE STEP STANDARD SAT/UNSAT COMMENTS ..
NOTE: The operator may collect the required data from SPDS prior to starting the computer. Therefore, Step 5 may be completed prior to Step 3.
NOTE: The computer that may be used is in the plant simulator or those computers that may have been setup by the Emergency Preparedness group specifically for this JPM evaluation.
NOTE: The MIDAS software prints to the user profiles default printer location.
The user should change their printer default to the local printer or a nearby printer location supplied by the Evaluator.
- 3. Operator logs onto a business LAN At a computer workstation, the SAT I UNSAT computer that has the MIDAS operator logs in with their software installed. business LAN ID and selects the "MIDAS Accident Calcs" icon.
- 4. Operator selects the correct MIDAS The operator selects "Start New SAT I UNSAT SAT/UNSAT version and menu option to perform a Run" and enters or verifies that
( prompt dose projection. the correct selections are made to perform a prompt dose projection
[Plant Hatch, Plant, Manual, &
Quick Dose Projection (Menu A)].
PROMPT: WHEN the operator addresses SPDS, the MIDAS Information screen, PROVIDE the operator the MIDAS Information attachment (NOT the Dose Projection Screen).
PROMPT: WHEN the operator addresses the individual flow recorders, as each recorder is identified, PROVIDE the operator the correct recorder attachment.
- 5. Operator records the required Using TRN-0052, "MIDAS Data SAT I UNSAT SAT/UNSAT information for the Dose Assessment. Input Acquisition" Form (or similar) the operator GA THERS/RECORDS the GATHERSIRECORDS appropriate information to perform the calculation.
(** Indicates critical step)
LR-JP-25306-10 Page 5 of 15
(
PERFO~ANCESTEP STANDARD SJ\1.'If);NSAT SATIUNSAT
~..
.c. OOM.Ml!1~lS COMMENTS**
- 6. Operator determines the Offsite Dose Using the Midas Program, the SAT/UNSAT Rates and Dose Projections. Operator INPUTS the collected data and performs the dose projection.
PROMPT: INFO~ the operator that a reactor trip HAS NOT occurred.
IF asked, INFORM NOTE: Durations other than 15 minutes will vary the resulting projected doses.
NOTE: An evaluator copy of a completed TRN-0052 is attached. Do NOT hand to operator.
( **8. Operator determines that a release is The operator READS the peak SAT I UNSAT SAT/UNSAT in progress and notifies the SMIED. TEDE Dose Rate Value from the screen and DETERMINES that it is greater than 10 times higher than the daily average. The operator NOTIFIES the SMIED that a release is in progress.
PROMPT: WHEN the operator addresses informing the SM/ED, ACKNOWLEDGE INFO~ the operator as the SM/ED that you the operator's report and INFORM understand a release is in progress.
- 9. Operator determines that the peak The operator READS the peak SAT I UNSAT SAT/UNSAT TEDE Dose Rate is greater than TEDE Dose Rate Value from the 0.057 mR/hr and notifies the SM/ED. screen and DETERMINES that it exceeds 0.057 mR/hr mRlhr and occurs beyond the site boundary. The operator NOTIFIES the SM/ED that the Emergency Classification Procedure should be addressed for releases.
(
(*** Indicates critical step)
(*
LR-JP-25306-l0 LR-JP-25306-10 Page 6 of 15 PERFO~ANCESTEP STANDARD SA{f/PN$1\1 SAT/UNSAT .
- C,QMMEN'tS; COMMENTS .. ~.;
NOTE: Peak TEDE dose rate should calculate to be approximately: 2.6 E-l mR/hr.
To successfully complete Step 9, the calculated values must agree with the values on the attached MIDAS screen print and ENN Fonn printout.
PROMPT: WHEN the operator addresses infonning the SMIED, ACKNOWLEDGE INFO~ the operator as the SM/ED that you will the operator's report and INFORM declare the Emergency and detennine the Protective Action Recommendations (PARs).
PROMPT: WHEN addressed by the operator, as the SM/ED, RECEIVE the printed ENNFonn.
PROMPT: INFO~ the operator that WHEN addressed by the operator, as the SM, INFORM
( another operator will perfOlm additional dose projections.
NOTE: Compare the student results with the TEDE Rate and ENN attachments in this JPM.
END TIME: _ _ __
NOTE: The tenninating cue shall be given to the operator when:
- With no reasonable progress, the operator exceeds double the allotted time.
- Operator states the task is complete.
TE~INATING CUE: We will stop here.
TERMINATING
(*
- Indicates critical step)
(**
(
MAIN STACK FLOW 1Dll-R625 Dl1-R62S INST BUS lA oo 5 10 15 20 25
~laxJlllll
~1~1 I I I I I IIII IIIIIIIIIIIIII I IIII IIII I II IIIII IIIIII III I II ~III IIII IIII ~I I II IIIII IIIII IIIII II o 5 10 15 20 25 IX1~1
~1oooll I I I I I I II I J i l l l I I I I II I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I II I II I II I II I II I II I I'.; .;
JJ1 0 5 -10 f1::: 1'~11 ...... p.
~ I"C" L'll::
I~~
I-~
(~"<~
( ..
~
~
B MAIN STACK FLOW RED-(CHANNEL B) 0-25 0-2S CFM X 1000 BLK-(CHANNEL A) 0-25 0-2S CFM X 1000 1DDll-R62S ll-R625 60S0K60-8S8 6050K60-858 Page 7 of of15 15
( STACK VENT FLOW lT41-R621 INST BUS 1A o++++-~ ~+++-I-DIftI-+++++-+-+4 +-H-t-H-++t 4+++++++++ 5
(
RX BLDG VENT FLOW RED-(CHANNEL B) 0-5 X 10 5sSCFM SCFM BLK-(CHANNEL A) 0-5 X 10 5sSCFM SCFM lT41-R621 CTA6445200
(
ofI5 Page 8 of 15
( STACK VENT FLOW 2T41-R621 INST BUS 2B X1()()() CFM 1000
(
RB STACK MONITOR RED-RB STACK FLOW (CH-B) 0-400,000 SCFM BLK-RB STACK FLOW (CH-A) 0-400,000 SCFM 2T41-R621 CTA6445100
(
Page 9 of 15
MIDAS INFORMATION METEOROLOGICAL 1
10M \lVI ND SPD m,jJ VVIND \lVIND SPD 1001*.J1 VVIND 100f**J! 1O~il O~/I VvlND VI/I ND DI DIRR 1OOrvl VvlND 100M VI/IND DIR lY33~R601 lY33-R601 lY33~R603 1Y33-R603 1Y33~R601 1Y33-R601 1V33-R603 1Y33-R603 3.0 5.0 60 65 At',JlBIE~JT AMBIEf\JT TEMP DELTA T DELTA T RAI~JFALL RAlhJFALL (F) 10M 60~.. 1a 60 100 1OO~.. 10 15 MIN. AVG 54 1..11 2.4 .000 RADIOLOGICAL MAIt~
t..jI,.l\lt~ STACK Ul RX. BLDG. VENT U2 RX. BLDG. VENT NOR1v'lc.,L R,ll,NGE NORrvlAL R.a.NGE KAMAr~
KAMAN NORf'J1AL NORhJlAL RANGE R.ANGE KAMAN KA MAN NOR ~/1AL R/\NGE NORrv'1AL Ri-\NGE KAI~AN KAMAN D11-K600A 1Dl1-K600A D1 1-R631 1D11-R631 1D11-K619. ~
lD11-K619A lDll~R631 1D11-R631 2D11-R619P, 2Dll-R619A 2Dl1-R631 2D11-R631 1.30E-02 4.00E 01 4.QOE 01 4.00E 1 D11-K600B 1D11-K600B 1D11-K619B 2011-R6198 2Dl1-R619B
'i:i 3.9DE 01 3.90E 4.o0E 01 4,OOE
~......
o g, STABI LIlY ClASS CLASS
.j:>. F
j.O'",,~ =-=~~_ """'~""""""""""'<""" ~""<'f< ""-~~ ~_~"".~_" "" ,,?;' ""P'''<-' "_~'='~~~-'""""~?s_~~ =''''''''''''~-&?' ~""'"" ~.-,.;..".",,~ '~""-""'¥"'"~"""""" ,~~~""'~"'",.." = --" ~~~""--'-"'&t~._.,....." ..
...., PLAN"rHATtThl :n' 56' 3.11"1'1 B2." 20' 19.6"W Version;L.5.10.032305
( Site: PLANT HATCH Unit: HT Menu: A Ql
Title:
TOTAL EFFECTIVE DOSE EQUIVALENT (TEOE) RATE Model: Proje(
'Current Time:
Time: - ArO.25 Houl Projectiol'l Peak values
, Peak TEDE I Dir (to): WS\
, Peak THY CI i Dir[to): WS\
TEDE/EDE i
(
Dose Rate at Graphic
~LqUid Start Time _ _ _ _ Date _1_ _fi _ __ _Stop Time _ _ _ _ Date _1 _ f_ _1 f __
15, PROJECTION PARAMETERS:
- 15. Projection period: _ _ _----'Hours -'Hours Estimated Release Duration 4,0 4.0 Hours Projection performed: Time _____ Date _-.1 _----1_ _1__ Accident Type: _ _ 2 __
Site boundary 1.0E+00 1,OE+00 5.4 E-03 2 Miles 8,5E*01 8.5E-01 1.1 E+OO 5 Miles 4.2E-01 1.9 E+OO 10 Miles 1.3E-01 1,3E-01 8.4 E-01
- 17. APPROVED BY: Title _ _ _ _ _ _ _ __ Time _____ Date_-'_-'__
( NOTIFIED BY: _ _ _ _ _ _ _ _ _ __
RECEIVED BY: _ _ _ _ _ _ _ _ _ _ _ _ Time _____ Date_-' Date_.1_--" _ _I__
(To be com ISled leted by receiving organization)
Page 11 of 15
ANNUNCIATOR STATUS Panel Number Annunciator Title Alarm Status IHII-P601-4 OFFGAS VENT RADIATION HIGH-HIGH Illuminated IHII-P601-4 OFFGAS VENT RADIATION HIGH Illuminated IHII-P601-4 IHII-P6014 DNSC/INOP OFFGAS VENT RADIATION DNSCIINOP Illuminated Not llluminated IHII-P601-4 IHll-P601-4 OFFGAS VENT SAMPLE FLOW HIGH/LOW Illuminated Not llluminated IHII-P601-4 IHII-P6014 RX BLDG VENT SAMPLE FLOW HIGH/LOW Not llluminated Illuminated IHII-P601-4 REFUELING FLOOR VENT EXHAUST RAD HI-HI Illuminated Not llluminated IHII-P603-2 RX BLDG STACK RADN MON HIGH-HIGH Illuminated Not llluminated IHII-P603-2 RX BLDG STACK RADN MON HIGH Illuminated Not llluminated 2HII-P601-2 RX BLDG VENT EXHAUST RADIATION HI-HI Illuminated Not llluminated 2HII-P601-2 RX BLDG VENT EXHAUST RADIATION HIGH Illuminated Not llluminated 2HII-P601-4 RX BLDG VT MON HIGH/LOW FLOW Illuminated Not llluminated DOWNSCALE/INOP DOWNSCALEIINOP 2HII-P601-4 REFUELING FLOOR VENT EXHAUST RAD HI-HI Illuminated Not llluminated
(
of15 Page 12 of 15
SOUTHeRN NUCLEAR SOUTHI::::RN PLANT E.I.
E.!. HATCH PAGE 13 OF 3 FORM TITLE:
MIDAS INPUT DATA ACQUISITION EVALUATOR USE ONLY (NOT A HANDOUT)
METEOROLOGICAL OAT DATAA ENTRY MIDAS SCREEN 10M WIND 100M WIND 10M WIND 100M WIND AMBIENT TEMP 15 MIN AVG. 15 MIN AVG. RAINFALL LABELS SPEED SPEED DIRECTION*
DIRECTION' DIRECTION*
DIRECTION' (OF) (10M) DIFFERENTIAL DIFFERENTIAL (15 MIN. AVG.)
TEMPERATURE (.H)
(llT) TEMPERATURE (AT)
(llT) 60M-10M 100M-10M MPL# 1Y33-R601 1Y33-R603 1Y33-R601 1Y33-R603 1Y33-R607 1Y33-R606 1Y33-R606 1Y33-R606 DATE TIME 5.0 6.0 65 54 1.1 2.4 000 3.0 Readings may be taken from SPDS (Emergency Screens/MIDAS Reporting Info. or Miscellaneous Screens/Meteorological Data),
Panel 1H11-P690 (Primary Tower), OR 1H11-P689 (Backup Tower). In the Simulator Building, the Met MIDAS System can be use to obtain 15 min. average Meteorological Data for the Primary and Back-up Towers.
Stability Class (LlT),
(LlT), wind speed, wind direction and rainfall readings taken directly from the panel must be 15 minute averages.
!E IF the indicated instrument is unavailable use the following table (TRN-0052 page 2 of 3) to identify the appropriate alternate instrument.
- For wind direction greater than 360 degrees, subtract 360.
Page 13 ofl5 of 15
.,~N NUCLEAR SOU"I .,{N SOU-I PLANT E.!. HATCH PAGE 14 OF 3 FORM TITLE:
MIDAS INPUT DATA ACQUISITION EVALUATOR USE ONLY (not a handout)
RADIOLOGICAL DATA ENTRY NOTES:
- Entering data for the KAMAN monitors when the KAMAN system is NOT operating will result in inaccurate offsite dose estimates.
!E.the normal range instrumentation for any release path is offscale high and KAMAN is NOT operating properly.
- Units for radiological data entries should be checked to ensure the values are the same as those required by MIDAS.
RELEASE U2 RX BLDG MAIN STACK RXBLDG.
U1 RX BLDG.
PATH MNSTKNR MNSTKKAM STACK U1RXBGNR U1 RXBGKM U1 RXBG U2 RXBGNR U2 RXBGKM U2 RXBG MIDAS RELEASE RELEASE FLOW RELEASE RELEASE FLOW RELEASE RELEASE FLOW SCREEN LABELS (CPS) (uCi\cc) (CFM) (CPM) (uCi\cc) (CFM) (CPM) (uCi\cc) (CFM) 1 D11-K600 AlB 1D11-R631 1D11-R631 1D11-R625 1 D11-K619 AlB 1D11-K619 1D11-R631 1T41-R621 2D11-R619 AlB 2D11-R619AlB 2D11-R631 2T41-R621 PANEL MPL# 1H11-P604 1H11-P689 1H11-P645 1H11-P604 1H11-P689 1H11-P645 2H11-P645 2H11-P689 2H11-P645 DATE TIME 1.3E-02 18.0E03 4.0E01 1.9E05 4.0E01 200E03 17.5E03 3.9E01 2.0E05 4.0E01 190E03 Default flow values are as follows: UNISOLATED ISOLATED A release is underway IF effluent monitors exceed the Unit 1 Reactor Building Vent 288,905 CFM 193,870 CFM values below:
Unit 2 Reactor Building Vent 198,840 CFM 162,340 CFM
- Main Stack Normal Range;:::
Range;:: 500 cps Page 14 of 15
(
UNITl&2 UNIT 1 &2 READ TO THE OPERATOR INITIAL CONDITIONS:
- 1. A pipe break has occurred causing a Primary System discharge outside Primary Containment.
- 2. OFFGAS VENT RADIATION HIGH-HIGH is alarming on Unit 1.
- 3. 34AB-D 11-00 1-2, "Radioactivity Release Control," is in progress.
- 4. SPDS is available for use.
5.
S. The Shift Manager (SM) is acting in the Emergency Director's position.
( 6. The reactor failed to scram (ATWS), all rods did not fully insert.
INITIATING CUES:
DETERMINE Total Offsite Dose Rate per 73 EP-EIP-O 18-0, then:
NOTIFY the SMiED SM/ED if a release is progress.
NOTIFY the SMiED SM/ED if emergency classification should be addressed.
(
SOUTHERN NUCLEAR IDOCUMENT POCUMENT TYPE: PAGE E.I. HATCH PLANT E.!. II EMERGENCY PREPAREDNESS PROCEDURE 1 OF 11
{, DOCUMENT TITLE: DOCUMENT NUMBER: VERSION PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5 EXPIRATION ~PPROVALS:
APPROVALS: EFFECTIVE DATE: DEPARTMENT MANAGER J. C. Lewis DATE 3/23/05 DATE:
~~~~~-----
~~~~------
10/15/08 NIA NPGMIPOAGMIPSAGM C. R. Dedrickson DATE 3/24/05 1.0 OBJECTIVE This procedure provides the initial method used for prompt dose assessment to determine the dose rate at the site boundary for use in Emergency Classifications based on gaseous effluent and release determination. This procedure also provides the projection of offsite TEDE and CDE Thyroid doses for use in determination of Emergency Classifications based on dose projections and initial protective action recommendations (PARs).
TABLE OF CONTENTS Section Title Page 2.0 APPLICABI LlTY 1
3.0 REFERENCES
2 4.0 REQUIREMENTS 2 5.0 PRECAUTIONS/LIMITATIONS 3
( 6.0 PREREQUISITES 4 7.0 PROCEDURE 5 7.2 SYSTEM START-UP 5 7.3 DATA ACQUISITION 5 7.4 DETERMINATION OF OFFSITE DOSE RATES 7 AND DOSE PROJECTIONS 7.5 DOCUMENTATION AND RECORDS 11 2.0 APPLICABILITY This procedure is applicable to initial determinations of offsite dose and dose rates based upon estimated noble gas release from the Main Stack and Unit 1 and Unit 2 Reactor Building Vents.
This procedure is performed as required.
MGR-0001 Rev. 3
SOUTHERN NUCLEAR PLANT E.I.
E.!. HATCH I PAGE 2 OF 11 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION
( PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5
3.0 REFERENCES
3.1 10AC-MGR-006-0, Hatch Emergency Plan 3.2 20AC-ADM-002-0,Quality Assurance Records Administration 3.3 73EP-EIP-001-0, Emergency Classification and Initial Actions 3.4 73EP-EIP-005-0, On-Shift Operations Personnel Emergency Duties 3.5 NMP-EP-109, Protective Action Recommendations 3.6 31 EO-EOP-013-1/2, Primary Containment Control 3.7 73EP-EIP-015-0, Offsite Dose Assessment
3.8 Forms
- TRN-0052 - MIDAS Input Data Acquisition
- TRN-0146 - Prompt Offsite Dose Assessment Flowchart 4.0 REQUIREMENTS
(
4.1 PERSONNEL REQUIREMENTS 4.1.1 Emergency response personnel who have been trained in and are responsible for offsite dose assessment are required to perform this procedure.
4.1.2 The number and qualifications of personnel performing this procedure during an emergency will be determined by the SM/ED.
4.2 MATERIAL AND EQUIPMENT N/A - not applicable to this procedure.
4.2.1 The system requirements for the Meteorological Information and Dispersion Assessment System (MIDAS) dose assessment program are any standard Southern Nuclear Company (SNC) Business LAN computer with the MIDAS software installed. It is recommended that a local printer be provided for at least one computer in the Control Room (CR) and Technical Support Center (TSC).
(
MGR-0009 Rev. 4
SOUTHERN NUCLEAR PLANT E.!.
E.I. HATCH I PAGE 3 OF 11 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION
( PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5 5.0 PRECAUTIONS I LIMITATIONS 5.1 PRECAUTIONS 5.1.1 Significant radiation hazards may be encountered during entries to obtain local instrument readings. Maintain occupational exposure As Low As Reasonably Achievable (A LARA) in the (ALARA) performance of assigned duties.
5.2 LIMITATIONS 5.2.1 This procedure can NOT be used to downgrade the severity of an emergency classification.
5.2.2 This procedure is based upon using the Meteorological Information and Dispersion Assessment System (MIDAS) for calculating indications of offsite dose and dose rates. MIDAS utilizes data from both radiological and meteorological plant instrumentation. Readings may be obtained from SPDS, control room monitors/recorders, and/or locally at the instruments.
Information is required for all three release paths (Main Stack, U1 Reactor Building Vent, and U2 Reactor Building Vent).
NOTE: Values obtained from MIDAS MI DAS for the purpose of Protective Action Recommendations are based on the avoided dose concept.
(
5.2.3 For the purpose of Protective Action Recommendations (PARs), the TEDE and CDE values for Site Boundary, 2 miles, 5 miles, and 10 miles from the printed "ENN Form" Worksheet provided by MIDAS may be utilized. These values can also be obtained from the program on the Integrated TEDE and Integrated Thyroid CDE graphic or table screens.
5.2.4 The MIDAS user will log in with their normal business LAN 10. ID. The MIDAS software is designed to print to the user's default printer location; therefore the user should change their printer default to the local printer if available or to a nearby printer location.
(
MGR-0009 Rev. 4
SOUTHERN NUCLEAR PAGE PLANT E.I. HATCH DOCUMENT TITLE:
I DOCUMENT NUMBER:
4 OF 11 VERSION PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5 6.0 PREREQUISITES 6.1 Contact Health Physics prior to making entries to obtain local radiation readings.
6.2 The annunciators outlined in white in the Main Control Room are indications of a potential radiological condition. The following annunciators may indicate an offsite radioactive release:
6.2.1 Main Stack Annunciators Annunciator Panel Number OFF GAS VENT RADIATION HIGH-HIGH (601-412) 1H11-P601-4 OFF GAS VENT RADIATION HIGH (601-418) 1H11-P601-4 1H11-P601-4 OFF GAS VENT DNSC/INOP (601-424)
RADIATION DNSCIINOP 1H11-P601-4 OFF GAS VENT SAMPLE FLOW HIGH/LOW (601-430) 1H11-P601-4 1H11-P601-4 6.2.2 Unit I Reactor Building Vent Stack annunciators Annunciator Panel Number
, RX BLDG VENT SAMPLE FLOW HIGH/LOW (601-433) 1 H11-P601-4 1H11-P601-4
( REFUELING FLOOR VENT EXHAUST RADIATION HI-HI 1 H11-P601-4 1H11-P601-4 (601-403)
RX BLDG STACK RADN MON HIGH-HIGH (603-216) 1H 11-P603-2 1H11-P603-2 RX BLDG STACK RADN MON HI (603-225) 1HH11-P603-2 11-P603-2 6.2.3 Unit 2 Reactor Building Vent Stack Annunciators Annunciator Panel Number RX BLDG VENT EXHAUST RADIATION HI-HI (601-229) 2H11-P601-2 RX BLDG VENT EXHAUST RADIATION HIGH (601-223) 2H11-P601-2 RX BLDG VT MON HIGH/LOW FLOW 2H11-P601-4 DOWNSCALE/INOP (601-433)
REFUELING FLOOR VENT EXHAUST RADIATION HI-HI 2H11-P601-4 (601-403)
(
MGR-0009 Rev. 4
SOUTHERN NUCLEAR PLANT E.!.
E.1. HATCH I PAGE 5 OF 11 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5 I REFERENCE I 7.0 PROCEDURE 7.1 lE IF there are indication of a potential release condition as indicated by step 6.2, and activation of this procedure, THEN perform the following steps or utilize the job aid in TRN-0146, Prompt Offsite Dose Assessment Flowchart ELSE exit the procedure.
7.2 SYSTEM START-UP NOTE: The MIDAS software is installed on the STA's computer and other pre-designated business LAN PCs in the Control Room. If the computer you are using doesn't have the MIDAS Icon on the computer's desktop, contact the SM or SS to identify a PC that has MIDAS installed.
7.2.1 Log onto a business LAN computer on which the MIDAS software has been installed using your normal company business LAN ID username and password.
7.2.2 Change your default printer location to the local printer if available or a nearby business LAN printer.
7.3 DATAACQUISITION DATA ACQUISITION NOTE: MIDAS input data can be collected on form TRN-0052 before initiating the program or at any point prior to its entry being required by the program.
NOTE: Readings may be taken from SPDS (Emergency Screens/MIDAS Reporting Info. or Miscellaneous Screens/Meteorological Data), Panel 1H11-P690 (Primary Tower), OR 1H11-P689 (Backup Tower). In the Simulator Building, the Met MIDAS System can be use to obtain 15 min. average Meteorological Data for the Primary and Back-up Towers.
7.3.1 The MIDAS code requires the input of meteorological and radiological data. The data required is listed in TRN-0052, MIDAS Data Input Acquisition. Form TRN-0052 may be used to gather data for input. Data is to be gathered in accordance with the following steps.
7.3.1.1 Record readings from the appropriate meteorological instrumentation. Appropriate meteorological instrumentation is listed on form TRN-0052 - Page 1 of 3. Record only valid readings. !.E lE the primary meteorological instrumentation listed is unavailable, THEN use information on form TRN-0052 - page 2 of 3 to determine which alternate meteorological
( instrumentation can be utilized.
MGR-0009 Rev. 4
SOUTHERN NUCLEAR E.!. HATCH PLANT E.I.
I PAGE 6 OF 11 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5 NOTES
- IF there is a difference between the A and B channel readings for the Unit 1 Rx Bldg. Stack Vent monitor (1D11-R619 (1 D11-R619 NB),
AlB), Unit 2 Rx Bldg. Stack Vent Monitors (2D11-R619 NB) AlB) OR Main Stack monitor readings (1D11-R600 (1 D11-R600 NB),
AlB), THEN use the most conservative (higher) reading for input into MIDAS.
- Normal Kaman Operation is indicated on SPDS primary screen by vent stack window changing red with a value showing.
- !E IF the Control Room flow instruments are inoperable or unreliable; THEN the default flow values listed on TRN-0052, page 3 of 3 may be utilized for input.
- MIDAS will sum the release and significantly overestimate the dose rate and output !E both normal range AND accident range effluent values projected dose output!E are entered.
7.3.1.2 Record readings from the appropriate effluent monitors. Appropriate effluent monitors are listed on form TRN-0052 - Page 3 of 3. Record only valid readings as indicated below.
Record EITHER Normal Range OR Accident Range Effluent Monitor (KAMAN) value and the flow associated with that value. Do NOT record or enter data for both monitors for
( each release point.
7.3.1.3 IF it is determined that the normal range effluent monitor is offscale high or not functioning AND the accident range instrumentation (KAMAN) does not function for the Main Stack, Unit 1 Rx Bldg. or Unit 2 Rx Bldg. vent monitors, THEN turn over dose assessment responsibilities to the TSC.
7.3.1.3.1 !E IF the TSC is not activated, contact the on-shift HP foreman and instruct him to initiate dose assessment activities on a designated dose assessment computer in the TSC or HP/Chem offices and proceed to Section 7.5.
7.3.1.4 IF the effluent monitors exceed AND either: THEN:
any of the following values:
- Main Stack Normal Range
- An abnormal plant Notify the SM/ED that a
~
- 2: 500 cps condition exists Or radioactive release is in progress and
- Either Reactor Bldg.
- An emergency has immediately proceed with Stack Vent ~ 10,000 cpm Vent:2: been declared OR a dose projection.
- The sum of Unit 1 and Unit 2 Reactor Bldg Stack Vent is :2: ~ 10,000 cpm
(
MGR-0009 Rev. 4
SOUTHERN NUCLEAR PLANT E.!.
E.I. HATCH I PAGE 7 OF 11 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5 7.4 DETERMINATION OF OFFSITE DOSE RATES AND DOSE PROJECTIONS NOTE:
- The MIDAS Program files are loaded on the hard drive of the computer located STA's desk.
at the STA's
- The MIDAS software is set-up in the "All Users" group so the MIDAS icon will display for any user logged onto the computer.
7.4.1 Select the "MIDAS Accident Calculation" icon from the computer "desktop".
7.4.2 IE IF it is determined that the MIDAS computer or software cannot be initiated properly or a prompt offsite dose assessment cannot be performed for any reason, THEN turn over dose assessment responsibilities to the TSC.
7.4.2.1 IF the TSC is not activated, contact the on-shift HP foreman and instruct him to initiate dose assessment activities on a designated dose assessment computer in the TSC or HP/Chem offices and proceed to Section 7.5.
7.4.3 Select "START NEW RUN" on "MIDAS - Accident Dose Calculation - sheet # 1".
7.4.4 The "Plant Selection" field should default to "Plant Hatch". IE IF not, THEN select "Plant
( Hatch" from the drop down menu.
7.4.5 The "Version Selection" field should default to "Plant". IE IF not, THEN select "Plant" from the drop down menu.
7.4.6 Select "Manual" for the mode of operation.
7.4.7 The "Accident Run Menu Selection" field should default to "Quick Dose Projection (Menu A)". !E IE not, THEN select "Quick Dose Projection (Menu A)" from the drop down menu.
7.4.8 Select "OK" on "MIDAS- Accident Dose Calculation - sheet # 1". The "MIDAS - Accident Dose Calculation Menu: A Sheet # 2" will then be displayed.
st 7.4.9 Select the "Next" button in the 1st section of "MIDAS - Accident Dose Calculation Menu: A nd Sheet # 2". The 2nd section of this screen will then be displayed.
nd 7.4.10 Select the "NEXT" button in the 2nd section of "MIDAS - Accident Dose Calculation Menu: A Sheet # 2". "Spreadsheet Control Sheet # 3" will then be displayed.
(
MGR-0009 Rev. 4
SOUTHERN NUCLEAR PLANT E.I.
E.!. HATCH I PAGE 8 OF 11 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION
( PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5 NOTE: A "Warning Message" may be displayed stating "The last run was for a different unit. All spreadsheets should be initialized! Select [OK] to initialize spreadsheets or [Cancel] to exit run." This message will only appear on the first run of a new installation OR if the computer has been used to run other MIDAS models or menus. Selecting [OK] on the message will allow the user to proceed with initializing the spreadsheets. MIDAS will then display the meteorological spreadsheet as stated in step 7.4.13.
NOTE: * "Spreadsheet Control Sheet # 3" will display the date and time the last spreadsheets were built by the program with two menu options. "New" "New" should be selected for the first projection in a series of projections or for a single projection. "Edit Last" should be selected to update a previous projection.
- A "Warning Message" will be displayed stating ""You are about to destroy "New" and "OK" are selected on "Spreadsheet previously entered data" when "New" Control Sheet #3". This message is warning you that when you perform a new projection, the spreadsheet is reinitialized and all stored data is erased.
This is appropriate for an initial projection. Selecting "OK" on the message
( will allow the user to proceed with initializing the spreadsheets.
7.4.11 "New" on "Spreadsheet Control Sheet # 3" !E Select "New" LE this is an initial projection, ELSE selected "Edit Last".
7.4.12 Select "OK".
7.4.13 7.4.13 The "Meteorological Data" spreadsheet (Sheet #4) will now appear. Enter meteorological data for the current time from form TRN-0052 as required by the spreadsheet. The current time frame will be highlighted. If performing an update, data previously entered will be displayed.
7.4.14 Select "OK".
CAUTION: RELEASE POINT MONITOR DATA MAY BE DISPLAYED ON SPDS IN UNITS OTHER THAN THOSE REQUIRED BY MIDAS. IF USING SPDS FOR DATA* DATA ACQUISITION, CHECK ALL UNITS TO ENSURE THAT THE UNITS MATCH THOSE REQUIRED BY MIDAS. SPDS MAY DISPLAY THE SAME MONITOR ON DIFFERENT SCREENS IN THE APPROPRIATE UNIT.
7.4.15 The "Radiation Monitor and Flow" Flow" (RM/F) spreadsheet (Sheet #5) will now appear. Enter the radiological data for the current time from form TRN-0052 as required by the spreadsheet. The current time frame will be highlighted. If performing an update, data previously entered will be displayed.
MGR-0009 Rev. 4
SOUTHERN NUCLEAR PAGE PLANT E.!.
E.I. HATCH DOCUMENT TITLE:
I DOCUMENT NUMBER:
9 OF 11 VERSION PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5 CAUTION
THE GASEOUS VENT AND FLOW SPREADSHEET CONTAINS (10) DATA COLUMNS, OF WHICH (6) DATA COLUMNS ARE DISPLAYED. THE ADDITIONAL (4) DATA COLUMNS CAN BE DISPLAYED BY USING THE "RIGHT" ARROW KEY ONCE DATA HAS BEEN ENTERED IN THE FIRST (6)
DATA COLUMNS SHOWN ON THE INITIAL SCREEN. FAILURE TO ENTER DATA FOR EACH RELEASE POINT WILL RESULT IN A POTENTIAL UNDERESTIMATE OF THE OFFSITE DOSE RATE AND PROJECTED DOSE ESTIMATES.
7.4.16 Select "OK".
NOTE:
- IF the "MET Status" and RM/F Status" bars display in green, this indicates that all required data fields have an entry and the value is within the error checking range.
- IE either the "MET Status" or RM/F Status" bars displays in red, this indicates that there is a missing input or a value has failed error checking. Selecting the status bar will display the error message and return you to the selected spreadsheet to allow you to correct the error or enter missing data.
(
7.4.17 The 3rd section of "MIDAS - Accident Dose Calculation Menu: A Sheet # 2" will be displayed with the "MET Status" and RM/F Status" bars.
7.4.18 IE IF the status bars are displayed in red, THEN select the applicable status bar to display the error message and allow you to return to the selected spreadsheet to correct the error or enter missing data. IE IF the status bars are displayed in green, THEN select the "NEXT" button in the 3 rd section 3rd of MIDAS - Accident Dose Calculation Menu A Sheet # 2. The 4th section of this screen will then be displayed.
NOTE: MIDAS will default to "No Trip Date" option for the "Date and Time of Trip/Shutdown" for all projections.
7.4.19 The "Date, Time of Trip/Shutdown" should default to the "No Trip Date" option. IE IF the unit has not reached hot shutdown, THEN the "No Trip Date" should remain selected. JE J.E the reactor is tripped or in hot shutdown, THEN select the "Set Trip Date" option and enter the date and time. IF the reactor has tripped but the exact date/time is not known, THEN the "At Current Time" option should be selected 7.4.20 Next, select "Start Calc". MIDAS will now perform the required calculations and display a plot of the Total Effective Dose Equivalent (TEDE) rate based on a .25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> (15 minute) projection. The display gives an estimated peak TEDE Dose Rate value at or beyond the site boundary (see the
( right center of the screen). The display also includes the direction and distance for this value.
MGR-0009 Rev. 4
SOUTHERN NUCLEAR I PAGE PLANT E.I. HATCH 10 OF 11 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION
( PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5 NOTE: The current calculated daily average site dose rate is is>::!
R! E-03 mRlhr.
7.4.21 IF the peak TEDE dose rate (mRlhr) value is an order of magnitude (10 times) higher than the current calculated daily average AND an emergency has been declared, THEN notify the Emergency Director that a radioactive release is in progress.
7.4.22 IF the peak TEDE Dose Rate value exceeds 0.057 mR/hr (5.7 E-02 mRlhr), THEN notify the Emergency Director for possible emergency classification declaration or upgrade AND notify the affected Unit Shift Supervisor for possible EOP Actions.
NOTE: MIDAS will print a working copy of the ENN Form. The projected offsite dose values listed on the working copy of the ENN Form are utilized for the purpose of making an emergency declaration/upgrade or Protective Action Recommendations (PARs) to state and local authorities.
7.4.23 IF the values on the "ENN form" exceed 1.0 E + 3 TEDE or 5.0 E + 3 CDE Thyroid, THEN immediately notify the SM/ED of the results.
NOTE: Printouts of the working copy of the Emergency Notification Form can be obtained if
( the printer does not print properly or if additional copies are needed. Effort should be made to correct the problem with the printer, if appropriate.
7.4.24 To obtain printouts of the working copy of the Emergency Notification Form, select the "Special Reports" button on the selection bar at the bottom of the screen. The button below the "Special Reports" button will illuminate in yellow and then display the report option (default is "State SNC"). Additionally, the buttons below the "Thyroid CDE" and "EDE" buttons will illuminate in yellow and then display "Exposure 4hr" and "Table".
7.4.25 Select the flashing "Confirm" button. This will display the working copy of the Emergency Notification form to the screen. Select the printer icon to print a copy of the form.
7.4.26 Report the projected offsite dose values (TEDE and Thyroid CD E) values from the working copy of the Emergency Notification Form to the Emergency Director.
NOTE: You may elect to continue your evaluation of this projection,perform an update projection, or exit the MIDAS program.
7.4.27 Select the "End Run"" button from the selection bar at the bottom of the screen. A selection box with the exit options will then be displayed.
7.4.27.1 !E the SM/ED has NOT directed performance of continuous projections, THEN select the "Save Run and Exit" option and then select the "OK" button. MIDAS will exit and return to
( your computer "desktop". NEXT, proceed to step 7.5.
MGR-0009 Rev. 4
SOUTHERN NUCLEAR PLANT E.I. HATCH I PAGE 11 OF 11 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION
( PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 NO:
8.5 TSC or EOF activation is NOT required prior to turnover of dose assessment activities.
7.4.27.2 !E IF the dose assessment capabilities are functional from the TSC or EOF, THEN turnover Offsite Dose Assessment to either facility. Select the "Save Run and Exif' Exit" option and then select the "OK" button. MIDAS will exit and return to your computer "desktop". NEXT, proceed to step 7.5.
7.4.27.3 IF the dose assessment capabilities are NOT functional from the TSC or EOF, THEN continue to monitor the gaseous effluent release parameters until instructed to terminate dose assessment activities by the SM/ED.
The "Run Next Time Step" option of MIDAS may initially be highlighted in Red with a counter to the right. This counter will count down to the appropriate time for running NOTE:
the next (15 minute) time step. The highlighted color will then turn green and the counter will disappear indicating the next time step can be run.
7.4.27.4 Select "Run Next Time Step" to perform additional projections, then select the "OK" button.
nd The 2nd section of the "MIDAS Accident Dose Calculation Menu: A Sheet #2" will now be displayed on the screen. Return to step 7.4.9 to perform the next projections.
(
7.5 REVIEWS AND RECORDS Records generated during actual emergencies will be maintained in accordance with 20AC-ADM-002-0, Quality Assurance Records Administration.
I
\
MGR-0009 Rev. 4
SOUTHEKN NUCLEAR SOUTHERN E.!. HATCH PLANT E.I. PAGE 1 OF 3 FORM TITLE:
MIDAS INPUT DATA ACQUISITION METEOROLOGICAL DATA ENTRY MIDAS SCREEN 10M WIND 100M WIND 10M WIND 100M WIND AMBIENT TEMP 15 MIN AVG. 15 MIN AVG. RAINFALL LABELS SPEED SPEED DIRECTION' DIRECTION' (OF) (10M) eF) DIFFERENTIAL DIFFERENTIAL (15 MIN. AVG.)
(~T)
TEMPERATURE (I'.T) (~T)
TEMPERATURE (I'.T) 60M-10M 100M-10M MPL# 1Y33-R601 1Y33-R603 1Y33-R601 1Y33-R603 1Y33-R607 1Y33-R606 1Y33-R606 1Y33-R606 DATE TIME Readings may be taken from SPDS (Emergency Screens/MIDAS Reporting Info. or Miscellaneous Screens/Meteorological Data), Panel 1H11-P690 (Primary Tower), OR 1H11-P689 (Backup Tower). In the Simulator Building, the Met MIDAS System can be use to obtain 15 min. average Meteorological Data for the Primary and Back-up Towers.
Stability Class (~T),
(<1T), wind speed, wind direction and rainfall readings taken directly from the panel must be 15 minute averages.
IE IF the indicated instrument is unavailable use the following table (TRN-0052 page 2 of 3) to identify the appropriate alternate instrument.
- For wind direction greater than 360 degrees, subtract 360.
TRN-0052 Rev. 13.0 G16.70 73EP-EIP-015-0 73EP-EIP-018-0
SOUTHERN NUCLEAR PLANT E.I. E.!. HATCH PAGE 2 OF 3 PAGE20F FORM TITLE:
MIDAS INPUT DATA ACQUISITION Alternate Meteorological Instrument Designation for Dose Assessment Use Main Stack Elevated Release (Meteorological Data Entry Form - 100 Mdata) 100M Wind Speedo 100M Wind Direction° 100M-1 OM Differential 10M AmbientTemperature Rainfall 5
15 Min. Avg. MPH 15 Min. Avg. ° From 5 Temp. 15 Av . AT ° F5 15 Min. Avg. of 15 Min. Avg.
Value MPL# Value MPL# Value MPL# Value MPL# Value MPL#
Primary 100 M Y33-R603 100 M Y33-R603 100M -10M Y33-R606 10 M Y33-R606 10 M Y33-R606 1S{ Alternate 60 M Y33-R602 60 M Y33-R602 60M - 10M Y33-R606 10M' Y33-R604 Estimate~
na 2 Alternate 45M' Y33-R604 45 M Y33-R604 45M-10M Y33-R604 Estimate" N/A ra 4 3 Alternate 10 M Y33-R601 10 M Y33-R601 aS Y33-R603 N/A N/A 100M or 60M Y33-R602 t},~(}i;~!:Y(,~:i(:}:i;;wti: i';;~*i\,:,E;~;;!,f!:?{~V~M~!;~:~};:*)t**!
Reactor
- .*.i} ** "i ** Building Vent Ground~!i&'f)"I';.j; Level Release :;!;i,,';"'"
(Meteorological "'!"~'lData Entry
- .,I,
- *.!\:.i'*'!'l,.?:7;.'ii':':;:![:7. .~;.;*;*~!n~".,~,:r:!~!i!\![***;fj;;Ff;~) .\:
Form - 10M data) !;~~.. ;.*~;;iii';i~~~':W";*~':i~(r"),1;tF;\!2(;~;!F\!:\;)
10M 1 OM Wind Speed" Speedo Directiono 10M Wind Direction" 60M-10M Differential 10M Ambient Rainfall 15 Min. Avg. mph 55 15 Min. Avg. 0 ° From 55 Temp. 15 Avg. liT AT °0 F5 5
Temperature 15 Min. Avg.
15 Min. Avg. OF of Value MPL# Value MPL# Value MPL# Value MPL# Value MPL#
Primary 10 M Y33-R601 10 M Y33-R601 60M-10M Y33-R606 10 M Y33-R606 10 M Y33-R606 1S Alternate 45 M' 45M' Y33-R604 45M' 45M ' Y33-R604 45M-10M' Y33-R604 10 M Y33-R604 Estimate~
Estimate" na 2na Alternate 60M 60 M Y33-R602 60 M Y33-R602 100M-10M Y33-R606 Estimate" N/A ra 3 ro Alternate 100 M Y33-R603 100 M Y33-R603 a844 aS Y33-R601 N/A N/A 1
10M or45M Y33-R604 1
- 1. These readings are obtained from the Back-Up Meteorological Tower.
- 2. Since this value has minimal impact on the dispersion calculation an estimated ambient Temperature is acceptable.
- 3. Input these letters as estimates for rainfall based on a visual observation, L for Light, M for Medium Rain, or H for Heavy Rain. i.E IF no information is available use 0 in.l15 min. rainfall.
- 4. i.E IF the temperature values are unavailable for the Delta T readings, use the Sigma Theta (a8) (as) (variation in wind direction (in degrees)) for the Stability Class- Class -
=
22,500 = A, 22.4
~ 22.5 22.4°0 --17.5° =
17.50 B, 17.4° 17.40 -12.5° =
- 12.50 = C, 12.4° = 7.4 0 - 3.8
- 7.5 0 = D, 7.4° 12.40 -7.5° =
3.8°0 E, 3.7 3.7°0 - 2.1° 2.10 F, > 2.10 = =
2.1° = G. Input the Stability Class "Letter" into MIDAS.
- 5. i.E IF all instruments in this data field are inoperable, call the National Weather Service (see Emergency Call List Section 1- Offsite Agencies Phone List) and ask for the information from the nearest available source.
TRN-0052 Rev. 13.0 G16.70 73EP-EIP-015-0 73EP-EIP-018-0
soun.
('\ n ~.--------------------------~----------
soun. ,NUCLEAR I NUCLEAR PLANT E.!. HATCH PAGE 3 OF 3 FORM TITLE:
MIDAS INPUT DATA ACQUISITION RADIOLOGICAL OAT DATA A ENTRY NOTES:
- Entering data for the KAMAN monitors when the KAMAN system is NOT operating will result in inaccurate offsite dose estimates.
- Units for radiological data entries should be checked to ensure the values are the same as those required by MIDAS.
RELEASE U2 RXBLDG MAIN STACK U1 RXBLDG.
RX BLDG.
PATH MNSTKNR MNSTKKAM STACK U1RXBGNR U1 RXBGKM U1 RXBG U2 RXBGNR U2 RXBGKM U2 RXBG MIDAS SCREEN RELEASE RELEASE RELEASE FLOW RELEASE RELEASE RELEASE RELEASE FLOW RELEASE RELEASE FLOW LABELS (CPS) (uCilcc)
(uCilcc) (CFM) 1 D11-K600 NB AlB 1D11-R631 1D11-R625 1D11-K619 NB 1D11-K619AlB 1D11-R631 1T41-R621 2D11-R619 NB 2D11-R619AlB 2D11-R631 2T41-R621 PANEL MPL# 1H11-P604 1H11-P689 1H11-P645 1H11-P604 1H11-P689 1 H11-P645 2H11-P645 2H11-P689 2H11-P645 DATE TIME Default flow values are as follows: UNISOLATED ISOLATED A release is underway IF effluent monitors exceed the values below:
Unit 1 Reactor Building Vent 288,905 CFM 193,870 CFM
- Range;::
Main Stack Normal Range <:: 500 cps Unit 2 Reactor Building Vent 198,840 CFM 162,340 CFM
- Either Reactor Bldg. Stack Vent <:: 10,000 cpm or Vent;::
- Sum of Unit 1 and 2 Reactor Bldg Stack Vent Vent;::<:: 10,000 cpm.
Main Stack 20,000 CFM 20,000 CFM TRN-0052 Rev. 13.0 G16.70 73EP-EIP-015-0 73EP-EIP-0 18-0 73EP-EIP-018-0
SOUTHERN NUCLEAR PAGE 1 OF 4 PLANT E.I. HATCH FORM TITLE: PROMPT DOSE ASSESSMENT FLOWCHART The Annunciators Listed Below Are Indicators Of A Potential Release Condition. (6.2)
Main Stack Annunciator Panel Number OFF GAS VENT RADIATION HIGH-HIGH 1H11-P601-4 1 H11-P601-4 OFF GAS VENT RADIATION HIGH 1H11-P601-4 OFF GAS VENT RADIATION DNSCIINOP DNSCflNOP 1H11-P601-4 OFF GAS VENT SAMPLE FLOW HIGH/LOW 1H11-P601-4 1H11-P601-4 U1 Rx Bldg Vent Stack Annunciators Panel Number RX BLDG VENT SAMPLE FLOW HIGH/LOW 1 H11-P601-4 1H11-P601-4 REFUELING FLOOR VENT EXHAUST RADIATION HI-HI 1H11-P601-4 1H11-P601-4 RX BLDG STACK RADN MON HIGH-HIGH 1H 11-P603-2 H11-P603-2 RX BLDG STACK RADN MON HI 1H11-P603-2 1H11-P603-2 U2 Rx Bldg Vent Stack Annunciators Panel Number RX BLDG VENT EXHAUST RADIATION HI-HI 2H11-P601-2 RX BLDG VENT EXHAUST RADIATION HIGH 2H 2H11-P601-2 11-P601-2 RX BLDG VT MON HIGH/LOW FLOW DOWNSCALElINOP 2H11-P601-4 2H 11-P601-4 REFUELING FLOOR VENT EXHAUST RADIATION HI-HI 2H11-P601-4 Start
(
Yes Notes:
MIDAS input data can be collected on form TRN-0052 before initiating the program or at any point prior to its entry being required by the program.
The MIDAS software is installed on the STAs computer and other pre-designated business LAN PCs in the CR. If the computer you using doesn't have the MIDAS Icons on the desktop, contact the SM or SS to identify a PC with MIDAS installed.
Log onto the PC The MIDAS user should log in with their normal business 10. ID. The MIDAS software prints to and change the the user profiles default printer location. The user should change their printer default to the printer location local printer, if available, or a nearby printer location.
Select the In Plant Selection, In Version "MIDAS Accident / ~--
+ ......
\ ~--"""" select " Plant /---~
~--"""" Selection, select Cales" Icon Hatch" "Plant" OJ{" to Select" OK" In Menu selection, Select" Manual"
~-----\ 14---~
enter inputs and 1+----4,. Select" Quick 1 Select"Quick + - - - \ for the mode of 1+-----3.,.
exit Sheet # 1 ose Projection" operation.
(
TRN-0146 Ver. 3.0 G16.70 73EP-EIP-018-0
SOUTHERN NUCLEAR PAGE 2 OF 4 PLANT E.I. HATCH FORM TITLE: PROMPT DOSE ASSESSMENT FLOWCHART Continue Select "Next" on the first section of Sheet # 2 Select "Next" on the second section of Sheet # 2 Note: Note:
MIDAS input data can be collected on MIDAS returns to this point from page 4 form TRN-0052 before initiating the Gather input data when performing a "fifteen" minute program or at any point prior to its entry using TRN-0052 update. When "Next" "Nexf' is selected it will being required by the program. bypass Sheet 3 Spreadsheet Control and go directly to Sheet 4 Meteorological Edit.
Spreadsheet Contro I Sheet #3 Select Update "Edit Last"
(
Initial Note:
MIDAS gives an expected "warning message" (You are about to destroy previously entered data! ) when "New" and "OK" are selected, select "OK" on the message to proceed.
METEOROLOGICAL EDIT Sheet # 4
~
Readings may be taken from SPDS (Emergency Screens/MIDAS Reporting Info. or Miscellaneous Screens/Meteorological Data), Panel 1H11-P690 (Primary Tower), OR 1H11-P689 (Backup Tower). In the Simulator Building, the Met MIDAS System can be use to obtain 15 min.
average Meteorological Data for the Primary and Back-up Towers.
IE the normal instrument is unavailable use the table on TRN-0052 page 2 of 3 to select the appropriate instrument.
Select "OK" All meteorological readings must be 15 minute averages.
For wind direction greater than 360 degrees, subtract 360 TRN-0146 Ver. 3.0 G16.70 73EP-EIP-018-0
SOUTHERN NUCLEAR PAGE 3 OF 4 PLANT E.I. E.1. HATCH FORM TITLE: PROMPT DOSE ASSESSMENT FLOWCHART Continue RAD MONITOR & & FLOW DATA EDIT No Sheet # 5 Sheet#5 Notes:
Effluent monitor readings can be obtained from SPDS (Emergency Screens/MIDAS Reporting Info. or Diagnostic Screens/
Main Stack Effluent & Vent Effluent HNP Yes 1&2) OR the recorders listed on Form TRN-0052, MIDAS Data Input
'~~~~~-:-rm~'""'7"'"~~'1 Acquisition. IE IF an emergency has IH.E1::l been declared, I.I::I.EI':t Input "2" as the Accident Mix, if needed notify the SM/ED that Input only 1 monitor value for each a Release is release point (Normal Range or KAMAN). underway Select "OK" IE IF flow Instrument readings are NOT available use the default flows from TRN-
........:....;..,~..;.;..;..:."'-'-~........,;...;.;..;..:...;.;..;..:...;.;..;..:.--..J 0052, page 3 of 3.
Select "Next" The third section of Sheet # 2 displays a status "bar" for the "Met Status" & "RM/F (Radiation Monitor & Flow)
Status" spreadsheets. If the bar is Green then no errors were detected by the program. If it is Red then a
( value was not entered or failed error checking. Select the status bar to return to the spreadsheet to review data or correct an error.
Select No---I1o\
NO--I1Il.
"No Trip Date" SM/ED Notify the S M/E D that a release is underway &
the Peak TEDE Rate may exceed EAL & EOP action levels.
Yes Yes No Input date and time of shutdown and select "OK" Select OR IF the trip time is "Start Calc""
unknown, THEN enter the current time b selectin "OK" MIDAS will compute Review the MIDAS
( the dose projection TE DE Rate Screen TEDE TRN-0146 Ver. 3.0 G16.70 73EP-EIP-018-0 73EP-EI P-018-0
SOUTHERN NUCLEAR PAGE40F4 PAGE 4 OF 4 PLANT E.1. HATCH FORM TITLE: PROMPT DOSE ASSESSMENT FLOWCHART Continue Review the projected offsite Report the TEDE and dose values on the No CDE dose projection MIDAS "ENN Form" values to the SMIED 8M/ED printout Yes Notify the SM/ED 8M1ED that the dose projections exceed the value to declare a Select "End Run" General Emergency & &
issue PARs
(
No Yes No Continue to monitor gaseous effluent No release values and meteorological data Yes Turnover dose Note: assessment to the Dose Assessment activities will continue until directed to TSCorEOF stop by the SM/ED.
SMIED.
On-shift activities by the CR should be transferred to the TSC HP/C staff or EOF Dose Assessment Staff ASAP.
TRN-0146 Ver. 3.0 G16.70 73EP-EIP-018-0
DRAFT Southern Nuclear E. I. Hatch Nuclear Plant Operations Training JPM Admin 7, SRO Only TITLE
, EVALUATE THE NEED FOR/RECOMMEND OFFSITE PROTECTIVE
( ACTIONS AUTHOR MEDIA NUMBER TIME D. GIDDENS LR-JP-25202-11 13.0 Minutes RECOMMENDED BY APPROVED BY DATE sM Energy to Serve Your World SM
(
\
SOUTHERN NUCLEAR OPERATING COMPANY
(
\ PLANT E. I. HATCH Page 1 of 1 FORM TITLE: TRAINING MATERIAL REVISION SHEET Program/Course Code: OPERATIONS TRAINING Media Number: LR-JP-25202
- .* >.. .*.. . . . *. / '.' *****.RPD~o~JOr~;~
- i*".*' .. )\~t~OI"'s.
Author's '. ~til>,,~~
1':~e~.1'l0:****
Rev. No. <~})~t~>
1**Ij.",.;>r ..... ? .*....*...*....... ~i Date Reason for Revision Supv's
'.i.e} . . . _.. ........>- . . * ****......-:...*i--.Z-. Initials *.**Iniijals**
. Initiiils Initials 01 09124/92 09/24/92 General revision and format fonnat change WMM SMC 02 08/05/94 General revision, word processor change, incorporate RAB MMG change to MIDAS, adjust format fonnat 03 08/21196 08/21/96 Format F onnat change RAB DHG 04 07/03/97 Revised initiating cue and added MIDAS screen. SCB DHG 05 03/21100 Format modification, change time allowance based on Fonnat RAB DHG running average, change MIDAS fonn form due to revision 06 11/06/00 Include objective number RAB DHG 07 03/25/02 Include initial operator statement RAB RAB 08 03/17/04 03117104 Rev to 73EP-EIP-054-0 DNM DHG 09 06/27/05 Revised Initial License statement for successful RAB RAB
( completion Updated to include latest Midas and ENN fonn, form, 10 03/21/06 03121106 RAB RAB removed Response Cues 11 12/04/06 Updated for forNMP-EP-109.
NMP-EP-l 09. DHG DHG
(
LR-JP-25202-11 Page 1 of6 UNIT 1 (X) UNIT 2 (X)
TASK TITLE: EVALUATE THE NEED FORIRECOMMEND OFFSITE PROTECTIVE ACTIONS JPMNUMBER: LR-JP-25202-11 TASK STANDARD: The task shall be completed when the Protective Action Recommendation has been made per NMP-EP-l NMP-EP-109.09.
'i'ASKNUMBltR.:
TASK NUMBER: . .
201.105 (EP 001.088)
OBJlLCTIVENUMUF)E.:
OBJECTIVE NUMBER: 200.105.A PLANT HATCH JTA IMPORTANCE RATING:
( KIA CATALOG NUMBER: 2.4.9 KIA CATALOG JTA IMPORTANCE RATING:
RO N/A SRO 4.00 OPERATOR APPLICABILITY: Senior Reactor Operator (SRO)
I GENERAL
REFERENCES:
Unit 1 & 2 NMP-EP-109 (current version)
I REQUIRED MATERIALS: Unit 1& 2 NMP-EP-109 (current version)
APPROXIMATE COMPLETION TIME: 13.0 Minutes SIMULATOR SETUP: N/A
(
UNITl&2 READ TO THE OPERATOR INITIAL CONDITIONS:
- 1. The Prompt Offsite Dose Assessment has just been completed. The ENN form has been printed and is available. The MIDAS screen is available.
- 2. The Dose Assessment Staff is not available yet.
- 3. The Shift Manager has declared a General Emergency due to the release.
- 4. The release started 30 minutes ago and is ongoing.
- 4. The Shift Manager is performing the functions of the Emergency Director.
( 5. SPDS is not available.
- 6. No adverse weather conditions exist.
- 7. No manmade threats, i.e. terrorist threats, exist INITIATING CUES:
Determine the Protective Action Recommendations for the EPZ only, per NMP-EP-109.
NMP-EP-I09.
LR-1P-25202-11 LR-JP-25202-11 Page 3 of6 PERFORMANCE**STEP PERFO~ANCESTEP STANDARD *S1\.'flIJN~A.T SAT/UNSAT
"',:. __ ",,,>-,;-~_"_",,< :.-; __ ,,-.~-~,_,_ *
- _ ~'' __ __ ~_" _",,_,-,,_,",'_,,_;;..:">_:~_:.o:'_>::';.:,c >' :-_~;-:c~ .QOMMENl'S COMMENTS For INITIAL Operator Programs:
For OJT/OJE; OJT10JE; ALL PROCEDURE STEPS must be completed for Satisfactory Performance.
For License Examinations; ALL CRITICAL STEPS must be completed for Satisfactory Performance.
START TIME: ___ __ ___
PROMPT: AT this time, GIVE the operator the Emergency Notification Form, and the MIDAS Screen.
- 1. Identifies the procedure needed to Operator has identified the correct SAT/UNSAT perform the task. procedure as NMP-EP-I09.
NMP-EP-109.
- 2. Review the procedure's precautions Operator has reviewed the SAT/UNSAT and limitations. precautions and limitations.
- 3. Utilize Attachment 1. Operator has identified the SAT/UNSAT SAT I UNSAT Attachment 1 as the correct
( procedure section.
- 4. Determine is a General Emergency The Operator determines a SAT/UNSAT SAT I UNSAT has been declared. General Emergency has been declared and answers YES on the flowchart.
- 5. Determine is a PUFF release is in The Operator determines a PUFF SAT I UNSAT SAT/UNSAT progress or been terminated. release is not in progress or been terminated and answers NO on the flowchart.
NOTE: The definition section ofNMP-EP-109 ofNMP-EP-1 09 defines a PUFF release as "less than an hour". The EN form has the estimated release at 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
- 6. Detennine if a known Site or Plant The Operator detennines determines from the SAT I UNSAT Event is Underway making initial conditions that evacuation Evacuation Dangerous. is not dangerous and answers NO on the flowchart.
- 7. Determine if doses at or beyond the The Operator determines that site SAT/UNSAT SAT IUNSAT site boundary are projected to exceed boundary doses will exceed the
( PAGs. PAGs P AGs and answers YES on the flowchart.
(*** Indicates critical step)
(*
LR-JP-25202-ll LR-JP-25202-11 Page 4 of6 PERFORMANCE STEP STANDARD SATIUNSAT COMMENTS
- 8 Refer to Table 1 The Operator refers to the SAT/UNSAT SAT I UNSAT
.. HATCH specific Table 1. 1.
NOTE: Operator may use the optional Attachment 5, PAR Worksheet to record the infonnation.
- 9 Detennine the wind direction. The Operator detennines that the SAT/UNSAT SAT I UNSAT wind direction is 25 degrees and selects Table 1 Row labeled "NNE, > 11 - 34".
(
PROMPT: IF the student ask about supplemental PARs infonn them that they are not desired at this time.
PROMPT: IF the operator addresses notifications, as the Shift Manager, INFORM the operator that another operator will make the State and Local notifications.
END TIME:_ _ __
NOTE: The tenninating cue shall be given to the operator when:
- With no reasonable progress, the operator exceeds double the allotted time.
- Operator states the task is complete.
TERMINATING CUE: We will stop here.
(
(**
(*
- Indicates critical step)
Southern Nuclear Emergency Notification I.~DRlLL l.~DRlLL ~ ACTIJAL AC11JAL EVENT MESSAGE #
2.~INTTIAL 2.~INTI1AL ~FOLLOW-UP NOTIF1CATION:
NOTIFICATION: TIME_____ DATE__..!_-1 AUTIlENTICATION # _~_
..!_-.l___ AUTHENTICATION
- 3. SITE: HATCH NUCLEAR PLANT HATCHNUCLEARPLANT Con:firmation Confirmation Phone # CR or TSC - (912) 366-2000 ext. _ _ _ _
EOF - 205 992-6586
- 4. EMERGENCY CLASSIE1CATION CLASSIFlCATION ~UNUSUAL
~ UNUSUAL EVENT I!!IALERT 19 SITE AREA EMERGENCY IQI GENERAL EMERGENCY BASED ON EAL EAL# # _ _ _ _ __ EAL DESCRIPTION: _ _ __ __ __ __ __ __ __ __ __ __ _ _ __ _ _ ___ __ __ __ __ __ ___
- 5. PROTECTIVE ACTION RECOMMENDATIONS: 0
~ NONE
~EVACUATE
~EVACUATE-_----------------------
~SmLTER-----------------------------------------
§S~LTER _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
IQI Advise Ad vise Remainder ofEPZ of EPZ to monitor local radiofTV stations/Tone Alert Radios for additional information and CONSIDER THE USE OF KI radiofIV stationsITone (POTASSIUM IODIDE) IN ACCORDANCE!
ACCORDANCE WITH STATE PLANS AND POllCY.
~m~R--------------------------------
~mmR--------------------------------
- 6. EMERGENCY RELEASE: ~None !ill IillIs Is Occurring 19 Has Occurred
- 7. RELEASE SIGNIFICANCE: ~ Not applicable ~ Within normal operating §lSI Above normal operating [QI 1!21 Under evaluation limits limits
- 8. EVENT PROGNOSIS: ~Improving
~ Improving ~Stable
~ Stable 19 Degrading
- 9. METEOROLOGICAL DATA: Wind Direction from ~ degrees Wind Speed ~mph Precipitation No Rain PreCipitation Stability Class Q .Q
~ DECLARATION IO.It;iDECLARATION
- 10. Illi llijTERMINATION TER1vITNATION Time ________ Date_-1 Date_-.l_ ____ _I' _ __
II. AFFECTED UNIT(S):
- 11. II] ~jg] ~
( 12. UNIT STATUS:
(Unaffected Unites) Status Not Required for Initial lei U1 ___
Ie:I _ _% Power Shutdown at Time _ _ _ _ _ Date _1 _ '___
_ _1__
Notifications) ~ U2 ___
_ _% Power Shutdown at Time Date _1
_ _1__
13.REMARKS: _ _ __ __ __ __ __ __ __ __ __ __ _ _ __ _ __ ____ __ __ __ __ __ __ _ _ __ __ __ __ __ __ __ __ __ _ __ _ __ __
FOLLOW-UP INFORMATION (Lines 14 through 16 Not Required for Initial Notifications)
EMERGENCY RELEASE DATA_ NOT REQUIRED IF LINE 6 A IS SELECTED.
- 14. RELEASE CHARACTERIZATION: lei Elevated ~Mixed TYPE:!ejElevated TYPE: ~ Mixed 19 Ground UNITS:~Ci UN ITS: ~ Ci I§I Ci/sec 19 Ci,secl9f1Cilsec
[.LCi/sec MAGNITUDE: Noble Gases: 9.7E+01 lodines: 1.2E+OO 1.2E+00 Particulates: O.OE+OO Other: _ _ _ __
FORM: 161Ie!I Airborne Start Time _____ Date _ _1 '_ _I _ _Stop Time _ _ _ _ Date _1_ _1_ __ _
~Liquid Start Time _____ Date _1_ _I _ _Stop Time Date _1_ _1_ __ _
- 15. PROJECTION PARAMETERS: Projection period: _____ _ _-'Hours.Hours Estimated Release Duration 4.0 Hours Projection performed: Time _____ Date _-.1_ _1__ Accident Type: --
--1--
Site boundary 1.1 E+03 2.0 E+03 2 Miles 1.0 E+03 6.7 E+03 5 Miles 5Miles 3.9 E+02 3.4 E+03 10 Miles 8.6 E+01 9.1 E+02
- 17. APPROVED BY: _ _ _ _ _ _ _ _ _ _ _ Title _ _ _ _ _ _ _ __ Time _____ Date_..!_ _I__
( NOTIFIED RECEIVED BY: _ _ _ _ _ _ _ _ _ __ BY: _ _ _ _ _ _ _ _ _ _ _ _ Time _____ Date_..!_ _C_ I __
(To be com leted by receiving organization)
Page 5 of6
Site: ~Hll-lAl'I'.K PLANT HATCH Unit:.* tiT
_._ ... ' ". ._ ._ . . . . Unit HT
Title:
TOTAL EFFECTIVE DOSE EQQIVJr.LENrlrE~~JFb\lE lQT,AJ;;£fFEcri\iEllBs£ EQUIVALENT ITEDEJ RA.TE '. Pl~e.::ted PUme Seslmer.1 lire: AliO:.25"_ptQt~Cl~
A.t 0.25 HCQ P.-oiEld:ion . CiArEfl111me: 1 0126109 09: 05 Run riTe: 1 0126109 09: 04
- 1~*hI~fE~Jltl.~tf)i!\1I"****
Mer.!.JaIErtrJl r1 MetData ~"7---'-'-'--'----;
OJlrert. Met \'v'S (llphJ:5.D IND (110m I: 25' 51: D Erd Date of 15 Mirl'J1e Rate CamplJtalkln: 1012610909:08 SI.;v-1 01 Rebase. 1 0126109 os: 53 Erd d ReI,"~e: 10126109 09:08 ManuatErlrJl r1 Monlol dcta .<<
OJrrert. ReleazeR<lle l.a/sec]: . ~UE t01 1otai Ct NG 8iIE +iN I: 1 1E+03 P: 0.0: +00 <
Peak yakJes: <
Pea.; lEDE (rrrem}ht Dil {tol SSW Dist (m~sl: 1.0 Pea., lHV CDElnu$mlhr): ~.:£+D2 Dh(oo:t ssw Dis!: (rrii:ls): 11 lEDElEDE .slPeak T£OE; HE tOO Dose Rate (rrrmV~fi 1 t!:ffl2 UE+02+of-2 5~.f\Jl*,
5JE t{Jl
- 1 i:E"~
t ((+02
- 3. 1.((+01 ~*5.QE l.tE:tD1 5.'OE...Ut 01
>4 <!lEI 5(£+00 *1.0E...4-Ut 50(["<<1./1.£1£ (O 5 i .t.I:E.~OO UE+OJ *-..5.:OE fiOE ...OO (O
6 1;(111;;(11 tOE*1]1 i : 1'(E tOO 1.((::i{lfj 7 f.cE~02*!J;llE1Jl-*
l.CE*02
- 1.0£.1.11 8 '(r:e;Il3~t:OE,l;I~
HE*(r3* [OE,02 g tcE:04~*1,0E.o3 1.0E*(14 1;/lE:03 10 ,.ilH15' ~10E4J4 UI:*(I5*UJE4J4 11 'H'E;(6.J:OEtt15 1.0:*00 *1.0E4)5 .
12-. 1(E*07 ~-l.lJE~J6 1JE*(t'* tllE.(]S
'"d
~
(1) 0\
a, 0\
UNITl&2 READ TO THE OPERATOR INITIAL CONDITIONS:
- 1. The Prompt Offsite Dose Assessment has just been completed. The ENN fonn has been printed and is available. The MIDAS screen is available.
form
- 2. The Dose Assessment Staff is not available yet.
- 3. The Shift Manager has declared a General Emergency due to the release.
- 4. The release started 30 minutes ago and is ongoing.
- 4. The Shift Manager is perfonning performing the functions of the Emergency Director.
( 5. SPDS is not available.
- 6. No adverse weather conditions exist.
- 7. No manmade threats, i.e. terrorist threats, exist INITIATING CUES:
Determine the Protective Action Recommendations for the EPZ only, per Detennine NMP-EP-l09.
NMP-EP-109.
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 EN~"DJ()Slt~yqNr'l,'(f_rJJ' E"~roIO&~Y_/"W_,IJ' Procedure Page 1 of 19 Procedure Owner: _ _----'W:...:..=:a=te=r-'H...:.;.:.....:L=e=e;....:./...:E=m""""7.er':'"g~e"'"'"n:::_cy'-'-P=la:=::n::_:_n:.:.;in":,-g-:::=S:..::u,:",p",,,e.;..;rv:...:,;is=-=o:.,:..r..:...1 Walter H. Lee 1 Emergency Planning Supervisor 1..;::C;;.;:o"-' Corporate rp""'o;.;..r=at=e'--__
(Print: Name / Title / Site)
Approved By: ___________Original O~rig~i~n~al~s~ig~n~e~d~b~y~W~a~lt~er~H~.L signed by Walter H. Lee ~e~e~0~n~0~5/~0~1/~2~0~0~8~---------
on 05/01/2008 (Peer Team Champion/Procedure Owner's Signature / Date)
Effective Dates: 05/02/2008 05/02/2008 05/02/2008 05/02/2008 Corporate FNP HNP VEGP The individuals listed below are the members of the Peer Team responsible for writing and maintaining this procedure.
Corporate Charles K. Brown Chris E. Boone
( Clint S. Hartfield Plant Farley Robert J. Vanderbye Plant Hatch Rachelie G. Reddick Rachelle Plant Vogtle Lawrence E. Mayo PROCEDURE USAGE REQUIREMENTS SECTIONS Procedure must be open and readily available Continuous at the work location. Follow procedure step by Use: step unless otherwise directed by the procedure.
Procedure or applicable section(s) available at Reference Use: the work location for ready reference by person ALL performing steps.
Information Available on site for reference as needed.
Use:
(
Printed: 9/7/2009
Southern Nuclear Operating Company Emergency NMP-EP-109
( SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 EtmvJlI~Y_,.ww-J.r EntD I l/&1rt!Y6JU>f1iIrU' Procedure Page 2 of 19 Revision Description Version Number Revision Description 1.0 Implements a common fleet procedure for developing offsite Protective Action Recommendations (PARs). The procedure incorporates the revised guidance from RIS 2004-13, RIS 2004-13, Supplement 1 and RIS 2005-08.
2.0 This procedure change adds separate definitions for "uncontrolled release" and "controlled release", incorporates procedural recommendations for completing the emergency notification forms, and adds human factorin to the PAR Worksheet 1, based on user feedback.
(
(
Printed: 91712009 9/7/2009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOOTHERNA SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 E~~"DIfJ'~rlf(l ~ltrWWJ,r Ene"l"lJ()'!knfC}'#":"W",U' Procedure Page 3 of 19 TABLE OF CONTENTS 1.0 PURPOSE ............................................................................................................ 4 2.0 APPLICABILITY ................................................................................................... 4
3.0 REFERENCES
..................................................................................................... 4 4.0 DEFINITIONS ....................................................................................................... 5 5.0 RESPONSIBILITIES ............................................................................................ 6 6.0 PRECAUTIONS AND LIMITATIONS ................................................................... 6 7.0 PROCESS DESCRIPTION .................................................................................. 8 8.0 RECORDS ............................................................................................................ 8
( 9.0 COMMITMENTS .................................................................................................. 8 ATTACHMENT 1 - ACTION CHECKLIST FOR PAR DEVELOPMENT .......................... 9 ATTACHMENT 2 - PLANT FARLEY AFFECTED ZONES FOR PARS ........................ 12 ATTACHMENT 3 - PLANT HATCH AFFECTED ZONES FOR PARS ......................... 14 ATTACHMENT 4 - PLANT VOGTLE AFFECTED ZONES FOR PARS ....................... 16 ATTACHMENT 5 - PAR WORKSHEET ....................................................................... 18 Printed: 91712009 9/7/2009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 EtI~rr.YJP&rpt' EtlUf.l ~Nr~,.M*
/oSkrl1t Y~Nr ~,.J.i. Procedure Page 4 of 19 1.0 PURPOSE This procedure provides guidelines for determining Protective Action Recommendations (PARs) which will be communicated to offsite authorities during a General Emergency.
PARs are provided as an input to the protective action decision (PAD) making process for the development of protective action orders. Protective action orders are communicated to the general public by offsite authorities to avoid or reduce the exposure incurred from an accident condition that results in a significant radiological effluent release or has the potential for a release based on degraded plant conditions.
2.0 APPLICABILITY Protective actions are recommended to offsite authorities to avoid or reduce the radiological exposure that may be incurred by the public from an accident condition that results in a significant radiological effluent release or has the potential for a release based on degraded plant conditions.
This procedure is performed, as required, during drills, exercises, and declared emergencies following declaration of a General Emergency. Attachments 2, 3, and 4 are site-specific and non-applicable site attachments may be removed and discarded to ensure usage of the correct site-specific attachment.
(
3.0 REFERENCES
3.1 NRC IN 83-28, Protective Actions Based on Plant Conditions 3.2 EPA-400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, October, 1991 3.3 NRC IN 91-72, "Issuance of a Revision to the EPA Manual of Protective Action Guides and Protective Actions for Nuclear Incidents" 3.4 NRC IN 92-08, "Revised Protective Action Guidance for Nuclear Incidents" 3.5 NRC RIS 2003-12, "Clarification of NRC Guidance for Modifying Protective Actions" 3.6 NUREG-0654/FEMA REP 1, Supplement 3 3.7 NRC RIS 2004-13, "Consideration of Sheltering in Licensee's Range of Protective Action Recommendations", August 2,2004 3.8 NRC RIS 2004-13, Supplement 1, "Consideration of Sheltering in Licensee's Range of Protective Action Recommendations, Dated Aug. 2004", March 10,2005 3.9 NRC RIS 2005-08, Endorsement of NEI Guidance "Range of Protective Actions for Nuclear Power Plant Incidents", June 6, 2005
(
Printed: 91712009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 E,,~ff.Y /uSn-1II: XoHr ~rld~
EIIr!ftLY/j)&rl'tx.u,.~rJ.i* Procedure Page 5 of 19 4.0 DEFINITIONS 4.1 EPA PROTECTIVE ACTION GUIDELINE (PAG) - exposure levels determined by the Environmental Protection Agency for the evacuation of the offsite public following a release of radioactive materials. These levels have been established at one (1) Rem TEDE or five (5) Rem CDE Thyroid. (VCMT# 1985304906) 4.2 PROTECTIVE ACTION RECOMMENDATIONS (PARs) - shelter, evacuation, monitor, and/or KI recommendations made by SNC to appropriate state agencies.
PARs are made by SNC personnel based on the Attachment 1 Flowchart whenever a General Emergency is declared. Additionally, if in the opinion of the ED, conditions warrant the issuance of PARs, a General Emergency will be declared (SNC will not issue PARs for any accident classified below a General Emergency).
4.3 UNCONTROLLED RELEASE - is a radiological effluent release that cannot be immediately stopped via positive control action (Example: Vent stack release from a known or unknown Containment leakage pathway which is not under the control of the shift and requires time to terminate.)
4.4 CONTROLLED RELEASE - is a planned radiological effluent release that can be immediately terminated by the licensee (Example: closure of the Post LOCA CTMT vent valves that were manually opened to lower Containment pressure.).
( 4.5 PUFF RELEASE - A controlled release that is projected to exceed the PAGs and will be terminated in less than an hour or an uncontrolled release that was projected to exceed the PAGs and has been terminated.
4.6 TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE) - The sum of the deep dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).
4.7 COMMITTED DOSE EQUIVALENT (CDE) - The dose equivalent to organs or tissues of reference that will be received from an intake of radioactive material by an individual during the 50-year period following the intake.
4.8 TONE ALERT RADIO (TAR) - Radio used to provide emergency information to the public living in the 10 mile emergency planning zone around the sites.
(
\
Printed; 91712009 Printed: 917/2009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 EII~rrtJ()San EIlt!f1!.Y IP&rH Y.,,,.r WorJJ~
Y#Nr WrnlJ" Procedure Page 6 of 19 5.0 RESPONSIBILITIES 5.1 The Emergency Director (ED) has the non-delegable responsibility for approving PARs (VCMT#1985304893).
5.1.1 The EOF Manager may sign approval for the ED after receiving verbal approval from the ED.
5.2 Once the TSC is operational, the TSC has responsibility for developing and communicating offsite PARs until relieved of that responsibility by the EOF.
5.3 Approved PARs may be communicated to applicable offsite authorities by the Control Room, TSC or EOF staffs as directed by the ED.
6.0 PRECAUTIONS AND LlMITIATIONS 6.1 Evacuation and Shelter Recommendations 6.1.1 PARs are only applicable when entering a General Emergency.
6.1.2 Evacuation is the preferred action unless conditions impose a greater risk
( from the evacuation than from the dose received.
6.1.3 Shelter is a preferred action when a 'Puff' type release has occurred.
6.1.4 A plant condition based PAR to shelter a 2-mile radius and 5 miles downwind may be issued when a Puff Release has occurred.
6.1.5 If onsite plant events are underway which would make evacuation dangerous (such as known hostile action) then sheltering should be considered over evacuation recommendations.
6.1.6 When prior knowledge of offsite impediments to evacuation exist (such as flooding, bridge/road closings, or other travel restrictions), then sheltering should be considered over evacuation recommendations.
6.1.7 A recommendation to evacuate or shelter a partial zone is not allowed.
6.1.8 Once an evacuation recommendation for an area has been given, it should not be reduced to a shelter recommendation.
9/7/2009 Printed: 917/2009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUfHERNA SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 EU~~11()&rUt X,Nr EU#fyI9&rpt' ~iUrW~rJJ'
~,.JJ' Procedure Page 7 of 19 6.2 ED Judgment 6.2.1 The ED may elect to modify PARs based on judgment, if conditions warrant.
6.2.2 The ED shall upgrade to a General Emergency if PARs are determined to be needed and not already in a General Emergency.
6.2.3 Protective action guidelines shall not imply an acceptable dose.
6.2.4 PARs are inherently conservative such that expanding the evacuation zone as an added precaution would result in a greater risk from the evacuation than from the radiological consequences of a release. It also would dilute the effectiveness of the offsite resources used to accommodate the evacuation.
6.3 Recommendations Beyond the 10 mile EPZ 6.3.1 Many assumptions exist in dose assessment calculations, involving both source term and meteorological factors, which make computer predictions over long distances less reliable. The ED should use the recommendation of the dose assessment staff when making recommendations beyond 10 miles 6.3.2 While evaluating the need to develop PAR 4 recommendations, issuance
( of appropriate PAR 1,2, or 3 recommendations should not be delayed.
6.4 Ingestion Pathway and Relocation Responsibilities 6.4.1 Protective actions taken in areas affected by plume deposition following the release are determined and controlled by offsite governmental agencies. SNC is not expected to develop offsite recommendations involving ingestion or relocation issues following plume passage.
6.4.2 SNC may be requested to provide resources to support the determination of post plume protective actions.
(
Printed: 91712009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUTHERN'\.
SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 E""~"II~&rplI YllurW"rlJ" l::.'1l<,rt.YI9&rp"X,."W"rlJ" Procedure Page 8 of 19 6.5 Continuing Assessment 6.5.1 Weather should not normally influence SNC protective action recommendations for the public except for changes in plume trajectory.
The States and Counties are the most knowledgeable concerning current weather conditions and weather forecast information. The States and Counties may incorporate existing or forecast weather in their decisions regarding implementation of recommended protective actions.
6.5.2 Only the MUTUALY AGREED UPON protective action recommendations specified in Attachment 1 should be recommended unless there are obvious relevant factors (e.g., severe natural phenomena like hurricanes) that probably were not anticipated when the PARs were developed and that would make the standard PAR recommendations impractical or obviously non-conservative. In such events, the ED should use judgment as appropriate.
6.5.3 Actual field readings from Field Monitoring Teams should be compared to dose assessment results and used as a dose projection method to validate calculated PARs and to determine whether the plant or dose based protective actions are adequate. (VCMT# 1986309134) 6.5.4 When available, actual sample data from monitored or unmonitored
( release points should be utilized in conjunction with other dose assessment and projection methods to validate calculated PARs and to determine whether the plant based protective actions are adequate.
6.5.5 VEGP and FNP off-site dose rates may be significantly higher (up to 10 times) due to volatilization of iodine if a steam generator (SG) water level falls below the break point during a SG tube rupture 7.0 PROCESS DESCRIPTION Guidance is provided in the form of attachments. Attachment 1, Action Checklist for Off-Site PAR Development", Attachment 2, "Farley Site Specific Data Sheets",
Attachment 3, "Hatch Site Specific Data Sheets", Attachment 4 "Vogtle Site Specific Data Sheets", and Attachment 5 "PAR Worksheet" direct the initial and supplemental actions.
8.0 RECORDS Records generated during actual emergencies will be maintained as QA records in accordance with applicable administrative procedure.
9.0 COMMITMENTS Farley - None Hatch - 1989301429, 1990303261, 1990303410 Vogtle - 1985304693, 1985304906, 1986309134
(
Printed: 917/2009 91712009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUTHERN COMPANY A
A. Implementing Protective Action Recommendations Version 2.0 En~f'f'lI~&rN En~~y I()!krpt Y4lNr WOrld'"
Y"'lo'rw"rl.r Procedure Page 9 of 19 Attachment 1
- Continuing Activity (Page 1 of 3)
Action Checklist for PAR Development NOTE: ONLY THE MUTUALY AGREED UPON PROTECTIVE ACTIONS SPECIFIED BELOW SHOULD BE RECOMMENDED UNLESS THERE ARE OBVIOUS RELEVANT FACTORS (E.G., SEVERE NATURAL PHENOMENA LIKE HURICANES) THAT PROBABLY WERE NOT ANTICIPATED WHEN THE PARS WERE DEVELOPED AND THAT WOULD MAKE THE STANDARD PAR RECOMMENDATIONS IMPRACTICAL OR OBVIOUSLY NON-CONSERVATIVE. IN SUCH EVENTS, THE ED SHOULD USE JUDGMENT AS APPROPRIATE.
A. INITIAL ACTIONS Please Check
- 1.
- Precautions and Limitations are applicable in development of Protective Action Recommendations (PARs) in subsequent steps. Attachment 5, Figure D
1, "PAR WORKSHEET', may be used to record affected zones or sectors.
- 2.
- Determine General Emergency PARs using the Attachment 1 Flowchart.
- PAR 1 - Shelter to 2 miles and 5 mile downwind zones D
- PAR 2 - Evacuate to 2 miles and 5 mile downwind zones
- PAR 3 - Evacuate to 5 miles and 10 mile downwind zones
(
CAUTION - PAR Revisions must include previous PARs
- 3. For PAR 1,2,1, 2, and 3, determine the affected zones using Site specific Table 1. D An electronic program may also be used.
NOTE: Once conditions requiring a PAR change are available, PARs should be developed as soon as possible. (The expectation for development is 15 minutes after the chan e in conditions.
D NOTE: Once PARs are developed they should be communicated to appropriate agencies as soon as possible. (The expectation for communication is 15 minutes after development, as directed by position s ecific instructions.
information is selected:
- Select block 5.B and record the "Evacuate" zones OR select block 5.C and record the "Shelter" zones
- Select block 5.D
Printed: 91712009 9/7/2009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 Etmul()$awy. ... rW&rJ.r Entl'nl()SawY.foIrW""U' Procedure Page 10 of 19
- Continuing Activity Attachment 1 (Page 2 of 3) of3)
Action Checklist for PAR Development (Cont)
B. SUPPLEMENTAL ACTIONS Please Check
- 1.
- Continue assessment actions applying applicable Precautions & & limitations.
o D
- 2.
- IE a release is in progress THEN it is appropriate to dispatch Field Monitoring Teams (FMT) to downwind and adjacent areas as soon as o
D possible. FMT data should be used to validate calculated exposure rates by comparison with actual field exposure rates to ensure issued PARs remain conservative.
- 3.
D
- Use any available FMT readings, IF available, to validate accuracy of the projection model prior to issuing PARs beyond 10 miles.
- IF dose assessment calculations indicate the need to recommend actions beyond 10 miles, THEN consult with affected State agency(s) to compare/
( validate model assumptions prior to issuing PARs beyond 10 miles.
- 4.
- IE conditions requiring PAR 1 entry are eliminated or dose projections 0D change such that additional PARs are required THEN return to the Initial Actions section. Once conditions requiring PAR change are available, PARs should be developed as soon as possible. (The expectation for development is 15 minutes after the change in conditions.) Once PARs are developed they should be communicated to appropriate agencies as soon as possible.
(The expectation for communication is 15 minutes after development, as directed by position specific instructions.)
- 5.
- Apply dose projection results in continuing assessment activities. Dose D 0
assessment results should be used to refine (but not reduce) protective action recommendations after adequate data becomes available.
- 6. Utilize real time meteorological and effluent radiation monitor readings in 0D continuing assessment activities. IF radiation monitor readings provide sufficient data for assessment, THEN, it is NOT appropriate to wait for field monitoring data to become available to confirm or expand a PAR within the 10-mile EPZ.
release is in progress. It is expected that a dose projection will be performed as soon as practicable at a General Emergency with a release in progress to determine if PAR change is needed.
Printed: 91712009 9/7/2009
Emergency NMP-EP-109 SOUTHERN'\ Implementing Protective Action Recommendations Version 2.0 COMPANY Procedure Page 11 of 19 Attachment 1, Flowchart Is a (Page 3 of 3)
Has a UFFRELEAS eneral Emergenc In Progress or Been PAR!
PARt Been Declared YES Terminated That Is YES ~ SHELTER rojected to Excee o - 2 mile radius PAGs* And downwind to 5 miles NO NO Advise Remainder of EPZ to Monitor Local CONTINUE s a Known Radio/TV Stations/
RadiolTV ASSESSMENT Site or Plant Event TARs for Additional Return to START Underway Making -YES.
-YES"- Information Iniomlation Evacuation Consider the use ofKI ofK1 PAR 2 in accordance with EVACUATE State plans and policy NO o - 2 mile radius Refer to site specific And downwind to 5 miles Table 1 or computer Do Known program for affected Advise Remainder of Offsite Conditions Make -YES zones EPZ to Monitor Local Evacuation CONTINUE
( Radio/TV Stations/
TARs for Additional Dangerous ASSESSMENT Information Return to START NO Consider the use ofKI ofK!
in accordance with PAR 3 State plans and policy Have Doses EVACUATE Refer to site specific at or Beyond the SB o - 5 mile radius NO- Been Projected to YES. And downwind to Table 1 or computer Exceed 10 miles program for affected PAGs*
zones Advise Remainder of EPZ to Monitor Local RadiolTV Radio/TV Stations/
TARs for Additional Information Have Doses Consider the use ofKI CONTINUE Beyond 10 Miles in accordance with ASSESSMENT .NO- Been Projected to State plans and policy Return to START Exceed PAGs* Refer to site specific Table 1 or computer program for affected YES zones PAR 4
( Evaluate the Need for PARs
\
- PAGs
'. Beyond 10 Miles - Refer to
Southern Nuclear Ql!eratinj!
Operating Company Emergency NMP-EP-109 i
\ SOUIHERNA SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 EN"'t'£rI(J~rptY.lI"rWf1TI;I" Ellern J(J SIr" ¥.IMr \%r/;I" Procedure Page 12 of 19 Attachment 2 Table 1 PLANT FARLEY AFFECTED ZONES FOR PROTECTIVE ACTION RECOMMENDATIONS DIRECTI~~
PAR 1 and 2 PAR 3 WIND DIRECTION AFFECTED AFFECTED FROM (degrees) (d~grees) '.'~.. ZONES ZONES N,~ A, N, > 349 - 11 A;, B5, C5, J5, K5 A, B5, C5, 05, E5, F5, 15, J5, K5, B10, C10, K10
.,~:N'NE,('>11 NNE,>11 - 34 A:'3~";rC5,;
A, B5,C5, 05,K5 05, K5 A, B5, C5, 05, E5, F5, 15, J5, K5, B10, Ci0, C10, 010
..* ;1 l~i];~NE >~2,,;!56
.;).:f
,;;$";3 ."
NE, >34 -,- 56 A, B5,C5,05 B5;'C5, 05 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, B10, 810, Ci0, C10, 010 EN;i;,
ENE,>56-79 ~~ 79 - :i! A, A,C5,05,E5 C5, 05, E5 A, B5, 85,C5, C5, 05, E5, F5, 15,J5, 15, J5, K5, Ci0, C10, 010,E10 010, E10
- 1i;i;'~* ./'/
E, >79 > '{ l:j"'f):l.'!j:r
.. 101 A, 05, E5, F5 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, Ci0, C10, 010, E10 ESE,~101-ESE, >101 - 124 A, 05, E5, F5 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, 010, E10, Fi0 F10
~ I.J SE, >124-146 A, E5, F5 A, B5, 85, C5, D5, 05, E5, F5, 15, J5, K5, E10, Fi0 F10
.}:i'JE;t~,~;t
"<~;;'E" SSE, >146 -169 - 169 A, E5, F5, 15 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, E10, Fi0, F10, Gi0 G10 S, >169 - 191 A, E5, F5, 15 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, Fi0, F10, Gi0, G10, Hi0 H10 SSW, >191 - 214 A, F5, 15 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, Fi0, F10, Gi0, G10, H10, HiD, 110 SW, >214-236 A, F5, 15, J5 A, B5, C5, 05, E5, F5, 15, J5, K5, Fi0, F10, G10, Hi0, H10, 110, Ji0 J10 WSW, >236-259 A, 15, J5 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, Gi0, G10, Hi0, H10, 110, Ji0 J10 W, >259 - 281 A, 15, J5 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, Hi0, H10, 110, Ji0, J10, Ki0 K10 WNW, >281 - 304 A, 15, J5, K5 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, 110, Ji0, J10, KiD K10 NW, >304 - 326 A, B5, 85, J5, K5 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, B10, 810, Ji0, J10, Ki0 K10 NNW, >326 - 349 A, B5, 85, C5, J5, K5 A, B5, 85, C5, 05, E5, F5, 15, J5, K5, B10, 810, Ki0 K10
(
91712009 Printed: 917/2009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUIHERNA SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 l:N..,," 1_S,rre HI.,.WWJ.r
&#YD/o&rn r#Nr w"yl,r Procedure Page 13 of 19 Attachment 2 Table 2 PLANT FARLEY GUIDANCE FOR PARS BEYOND THE 10 MILE EPZ
- 1. Calculate the Ev~.cuation v~uaLl'VI Distance by determining the maximum Projected Distance where MIDAS projectiQn~;exceed PAGs and adding 5 miles to the projected distance.
dose nrniA('1tir
_ _ _ _ 'Pi"(>>)~cted Distance (miles) + 5 miles = = Evacuation Distance (miles)
- 2. Determjo~ the affe~~ed ",ff.>'Bt~.rl sectors for the current 15 minute average (From) wind direction
~;*:;::f,*D. ';Y. _ _ _ _ _ _ _ _ _ _ _ _ _ Affected Affected Sectors Sectors 3.gebQmmend*~~~2d~ft&n from 10 miles to the Evacuation Distance (calculated in step 1) for the
" .* ' Affected Sectors (deterlJ'tiQ.edjn step 2).
-~~:j =- ' -;;-;>,.~:':~:j -~;:
it:~,.Check~j'le 5, Item E - otR~~r on the Emergency Notification Form and record the recommended
- ';:~§.est9rsarid dista,nce
-.0-.< _.~ -. ,
nce range in miles for Evacuation. (Note: Refer to 50 mile IPZ map as necessary)
,":\
PAR 4 WIND DIRECTION FROM (degrees) AFFECTED SECTORS N, > 349 -11 H, J, K NNE, >11-34 >11 - 34 J, K, L NE, >34- 56 K,L, K, L, M ENE, >56-79 L, M, N E, >79-101 M,N,P ESE, >101 -124 N,P,Q SE, >124-146 P,Q,R SSE, >146 --169 169 Q,R,A S, >169 -191 - 191 R,A,B SSW, >191 - 214 A,B,C SW, >214-236 B,C,D WSW, >236-259 C,D,E W, >259-281 >259 -281 D,E,F WNW, >281 - 304 E,F,G NW, >304 - 326 F,G,H NNW, >326 - 349 G, H,J H, J
(
Printed: 91712009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUIHERNA SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 EH~rcrJ(lSlr" Y4t",rWtJrlJ" F.nnnJlJ$ullt ~wr WorM" Procedure Page 14 of 19 Attachment 3 Table 1 PLANT HATCH AFFECTED ZONES FOR PROTECTIVE ACTION RECOMMENDATIONS
~AR1and2 PAR 1 and 2 PAR 3 WIND DlREC DIRECTION AFFECTED AFFECTED FROM (degrees) ZONES ZONES N~:-11 N, > 349 - 11 ~5 A, 85,C5 A,B5,C5,05,E5,C10,010,E10 A,85,C5,05,E5,C10, 010, E10
.~Jlit*lI::
NNi::, > >1111;~~~~~f,';
~ 34 A, B5, C5':t:~Ef A,85,C5 B5,C5, 05, E5, 010, E10, A, 85,C5, E10,F10Fi0
{ .,
.?~/
NE, >34 - 56 ENE, >56 >56-7~
56~;t?5*,
-79 A,85,C5 1:..:'
A, C5 C5 C5 A,B5,C5,05,E5, A, 85, C5, 05, E5, E10,F10,G10 E10, F10,G10 A,B5,C5,05,E5,E10,F10,G10 A, 85, C5, 05, E5, E10,F10,G10
>79-1 Q/fc1' E, >79-101 A,C5,05 A, C5, 05 A,B5,C5,05,E5, A, 85,C5, 05, E5, F10,G10,H10 ESE',';;,'i!'01 ESE, >101 - 124 A, A,C5,05 C5, 05 B5, C5, 05, E5, G10, H10, A, 85, Hi 0, 110
(
- ~~J
>124-146 A,C5,05, A, C5, 05, E5 A, B5, 85, C5, 05, E5, G10, Gi0, H10, 110
?
ji};*SSE, SSE, >146 - 169 A, A,C5,05, C5, 05, E5 B5, C5, 05, E5, H10, 110, J10 A, 85, Ji0 S, >169 - 191 A,05,E5 A, 85, B5, C5, 05, E5, 110, J10 Ji0 SSW, >191 - 214 A,05,E5 B5, C5, 05, E5, 110, J10 A, 85, Ji0 SW, >214-236 A,E5 A, E5 B5,C5, 05, E5,J10, K10, A, 85,C5, Kia, L10 wsw, >236-259 A,85,E5 WSW, A, B5,E5 A,B5,C5, A, 85,C5, 05, E5,J10, K10, Kia, L10 W, >259 -281 - 281 A, B5,E5 A,85,E5 B5, C5, 05, E5, 810,K10, A, 85,C5, B10, Kia, L10 WNW, >281 - 304 A,85,E5 A, B5,E5 A, 85, B5, C5, 05, E5, 810,C10, B10,C10, 010, K10, Kia, L10 NW, >304 - 326 A, B5 A,85 A,B5,C5, A, 85,C5, 05, E5, 810,C10, B10,C10, 010 NNW, >326 - 349 A,85,C5 A,B5,C5 A,B5,C5, 05, E5, B10,C10, 010, E10 A,85,C5,05,E5,810,C10,010,E10
(
9/7/2009 Printed: 91712009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 EN~t'C1Jo&rlft F:NO'rn Y.urW9rlJ" JOMrw Y.lo'r m,rt.r Procedure Page 15 of 19 Attachment 3 Table 2 PLANT HATCH GUIDANCE FOR PARS BEYOND THE 10 MILE EPZ
- 1. Calculate the Evacuation cuation Distance by determining the maximum Projected Distance where MIDAS dose projectiQj'\$;~xceed xceed PAGs and adding 5 miles to the projected distance.
____ .. ~;iiJ~9ted ed Distance (miles) + 5 miles = Evacuation Distance (miles)
~, "'-;','" ',,':;
- 2. DetermIne the affect~~~~qtors tors for the current 15 minute average (From) wind direction
.- .~".'_::.. -, ';;:*~;:.{*~~;oJ _ _ _ _ _ _ _ _ _ _ _ Affected Sectors
~R~9pmm'~N&;$,,§cuation froril!~1;oi~iles miles to the Evacuation Distance (calculated in step 1) for the Affeoted Secti:5r$(determinel In in step 2).
- 4. Check Line 5, Ite IterTl;:'~.;;.$')ther ther on the Emergency Notification Form and record the recomme nded recommended sectors and dista distanc.~;range nge in miles for Evacuation. (Note: Refer to 50 mile IPZ map as necessary) ne cessary)
'I'~'
," PAR 4
.;1'.;;& WIND DIRECTION FROM (degrees) AFFECTED SECTORS fC'*;
N, >349-11
> 349 -11 H, J, K
( NNE, >11 - 34 J, K, L NE, >34-56 K, L, M ENE, >56 -79 L, M, N E, >79-101 M,N,P ESE, >101 -124 N,P,Q SE, >124-146 P,Q,R SSE, >146 -169 Q,R,A S, >169 --191 191 R,A,B R,A,8 SSW, >191 - 214 A,B,C A,8,C SW, >214-236 B,C,D 8,C,D WSW, >236-259 C,D,E W, >259-281 D,E,F WNW, >281 - 304 E,F,G NW, >304 - 326 F, G, H NNW, >326 - 349 G, H, J
( ..
91712009 Printed: 917/2009
Operating Company Southern Nuclear Operatina Emergency NMP-EP-109 SOUTHERN'\
SOUTHERN"\. Implementing Protective Action Recommendations Version 2.0 COMPANY Enel'nJuSUf'r ElItrn 10 Sun XlUT Y4turWbr4r Worur Procedure Page 16 of 19 Attachment 4 Table 1 PLANT VOGTLE AFFECTED ZONES FOR PROTECTIVE ACTION RECOMMENDATIONS
~;'" PAR 1 and 2 PAR 3 WIND DIRECTION DIRECTl~N I"*, AFFECTED AFFECTED FROM (degrees) (deg~es) "., ZONES ZONES N, > 349 --'1!;1:.~;,;, 11 .bi~D~5~C5, A,85, C5, SRS Sgs to 2 Miles A, B5, 85, C5,05, E5, F5, B10, C10, 010, SRS to 5 Miles 810, Ci0,
":~'NE, NNE, >11-34 >11 - 34 ';~i~a5 A, 85, C5,C5,.$!~S
$RS to 2 Miles A, 85, C5, 05, E5, F5, Ci0, C10, 010, SRS to 5 Miles c.>. ~J~;
NE,>34 ,.,, - 56 A, 85, R~!li~~
C5, 05, SRS to 2 Miles A, B5, C10, 010, E10, SRS to 5 Miles 85, C5, 05, E5, F5, Ci0, ENE,':J~~,~
ENE, >56~ 79 1~~~\C5, A, C5, 05, E5, SRS to 2 Miles A, 85, C5, 05, E5, F5, 010, E10, F10,SRS to 5 Miles Mile~
- .i
" E, >79-101 >79-101'W~? A, C5, 05, D5, E5, F5, SRS to 2 Miles A, B5, 85, C5, 05, E5, F5, 010, E10, F10, SRS to 5 Miles
-"1--,._,'-.. ,',',_>,
ES\3;';':.r;r.1D1 ESE, >101 - 124 A, A. 05, E5, F5, SRS to 2 Miles A, 85, C5, 05, E5, F5, E10, Fi0, F10, G10,SRS to 5 Mile~
(
~t;"';."*
.,t;j;.~::~,;,
, >146 - 169 A, 05, E5, F5, SRS to 2 Miles A, E5, F5, SRS to 5 Miles A, B5, F10, G10,SRS to 10 Miles 85, C5, 05, E5, F5, E10, Fi0, A, 85, C5, 05, E5, F5, F10, G10, SRS to 10 Miles
<;**,*.tii:;i;{~"'***
S, >169 - 191 A, F5, SRS to 5 Miles A, 85, C5, 05, E5, F5, Fi0, F10, Gi0, G10, SRS to 10 Miles SSW, >191 - 214 A, A. F5, SRS to 5 Miles A, 85, C5, 05, E5, F5, Gi0, G10, SRS to 10 Miles SW, >214-236 A, SRS to 5 Miles A, 85, C5, 05, E5, F5, SRS to 10 Miles WSW, >236-259 A, SRS to 5 Miles A, 85, C5, 05, E5, F5, Hi0, H10, SRS to 10 Miles W, >259-281 >259 -281 A, 85, SRS to 5 Miles A, 85, C5, 05, E5, F5, 810, Hi0, H10, SRS to 10 Miles WNW, >281 - 304 A, 85, SRS to 5 Miles A, B5, 85, C5, 05, E5, F5, BiD, 810, Ci0, HiD, SRS to 10 Miles C10, H10, NW, >304 - 326 A, 85, SRS to 5 Miles A, B5, 85, C5, 05, E5, F5, BiD, 810, Ci0, HiD, SRS to 10 Miles C10, H10, NNW, >326 - 349 A, 85, SRS to 2 Miles A, B5, 810, C10, 010, SRS to 5 Miles 85, C5, 05, E5, F5, BiD, Printed: 91712009 9/7/2009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUTHERN COMPANY AJ!. Implementing Protective Action Recommendations Version 2.0 EN~fC'I /(1 EHefU ~!',e Y4Illr JoSu,e W'6rM" Y41MT W9"J.r Procedure Page 17 of 19 Attachment 4 Table 2 PLANT VOGTLE GUIDANCE FOR PARS BEYOND THE 10 MILE EPZ
- 1. Calculate the EYp(,;uation EX~Guation Distance by determining the maximum Projected Distance where MIDAS dose projectidhsEixceed projestlir'isgx<ceed PAGs and adding 5 miles to the projected distance.
_ _~ .-..;/'_:'Project~d
______ "t¥~\projeC~~,d Distance (miles) + 5 miles =
f.,;-?- -'I*,
= Evacuation Distance (miles)
- 2. Deternir~ethec;affected'~eqtors !:i":WaGtorsfor for the current 15 minute average (From) wind direction
"~'i)~,.:;' . ',;, Affected Sectors
~S?Recommend
'Recommend Eva Evacd~tlbn fo ';0'miles to the Evacuation Distance (calculated in step 1) for the from*tdmiles
,,-:". '.'< .' .ffected Sectors (de
'Affected (deteij~Jg,~(j' In step 2). 2) .
~:"~;~}.,~ ..:
+-
ch'eplyLine C Line 5, Item Item~"- ~ther on the Emergency Notification Form and record the recomme
'6";?'; - Other nded recommended sectors'and sec nd distance dista ne cessary) range in miles for Evacuation .. (Note: Refer to 50 mile IPZ map as necessary)
.0.;:-_-
- .', ,.".
- :}c~"c' PAR 4
','<~:'.:7-'-.
d' * ~
PAR 4
.*. WIND DIRECTION FROM (degrees) AFFECTED SECTORS N, > 349 -11 H, J, K NNE, >11 - 34 J, K, L NE, >34-
>34-56 56 K, L, M ENE, >56 -79 L, M, N E, >79-101 M,N,P ESE, >101 -124 N,P,Q SE, >124-146 P,Q,R SSE, >146 -169 Q,R,A 5,
S, >169 -191
- 191 R,A,8 R,A,B SSW, >191 - 214 A,8,C A,B,C SW, >214-236 8,C,D B,C,D WSW, >236-259 C,D,E W, >259 -281- 281 D,E,F WNW, >281 - 304 E, F, G NW, >304 - 326 F, G, H NNW, >326 - 349 G, H, J
(
Printed: 91712009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUIHERNA SOUTH£RN COMPANY J1 Implementing Protective Action Recommendations Version 2.0 EUlI!rnJtl~r't' Enj!ffY r.",r 'I<<1rlJO 1D ~T't' Y.",r W'tJrlJ" Procedure Page 18 of 19 Attachment 5 Figure 1 PAR WORKSHEET INSTRUCTIONS:
- 1. Check the box for the applicable PAR (1,2,3, or 4).
- 2. Record the 15 minute average "wind direction from" for the selected PAR.
Use met instrumentation corresponding to primary release point(s) (BWR) OR ground level release (PWR).
- 3. Use the applicable "Site Specific" PAR table (Table 1 or 2) to determine the affected zones.
!!!!!!!!!!!!!'!!!!!!!!!!!!!!!!!!!!\;I II CAUTION: II PAR Revisions must include previous PARs. II On the ENN Form for the selected PAR:
- Select block 5.B 5.8 and record the "Evacuate" zones OR select block 5.C and record the "Shelter" zones"
- Select block 5.0 5.D
Shelter Zones ENN Line 5 [0]
Advise remainder of EPZ to Monitor Local RadiolTV RadiollV Stations
{fone Alert Radios. Consider the use of KI (Potassium Iodide) in
!Tone accordance with State Plans and Policy
(
Wind direction from D
Evacuate Zones ENN Line 5 [0]
Advise remainder of EPZ to Monitor Local Radio/TV RadiolTV Stations
{fone Alert Radios. Consider the use of KI (Potassium Iodide) in
!Tone accordance with State Plans and Policy Wind direction from D
Evacuate Zones ENN Line 5 [0]
Advise remainder of EPZ to Monitor Local RadiolTV RadiollV Stations
{fone Alert Radios. Consider the use of KI (Potassium Iodide) in
!Tone accordance with State Plans and Policy Wind direction from ENN Line 5 [8]
D PAR 4 Evacuate Zones ENN Line 5 [0]
Advise remainder of EPZ to Monitor Local Radio/TV accordance with State Plans and Policy RadiolTV Stations/
Stationsl Tone Alert Radios. Consider the use of KI (Potassium Iodide) in ENN Line 5 [E] Evacuate Affected Sectors to OTHER miles
(
Approval:
~/-------
Printed: 9/7/2009
Southern Nuclear Operating Company Emergency NMP-EP-109 SOUTHERN COMPANY A Implementing Protective Action Recommendations Version 2.0 EneI'CY EneFfYJD$r,e l':.1I1> ~rnr UJ$UH Ye/llTW1'Irltl' Procedure Page 19 of 19 Emergency Director DatelTime
(
(
Printed: 91712009 9/7/2009