ML082381084
| ML082381084 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/21/2008 |
| From: | - No Known Affiliation |
| To: | NRC/RGN-II |
| References | |
| 50-338/08-301, 50-339/08-301 | |
| Download: ML082381084 (508) | |
Text
Draft Submittal (Pink Paper)
NORTH ANNA JUNE 2008 EXAM 05000338/2008301 & 05000339/2008301
- 1. ADMINISTRATIVE TOPICS OUTLINE (ES-301-1)
- 2. CONTROL ROOM SYSTEMS & FACILITY WALK-THROUGH TEST OUTLINE (ES-301-2)
- 3. ADMINISTRATIVE JPMS
- 4. IN-PLANT JPMS
- 5. CONTROL ROOM JPMS (SIMULATOR JPMS)
ES*301 Administrative Topics Outline Form ES*301*1 Facility: North Anna Power Station Date of Examination:
Examination Level: Combined (See Below) Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (KA)
(see Note) Code*
M,R Determine the Quadrant Power Tilt Ratio by hand calculation Conduct of Operations (1-PT-23) and determine maximum allowable power level based on the calculation (ALL) (G2.1.7, RO 4.4/SRO 4.7)
N,R Determine minimum RHR flow based on time after shutdown Conduct of Operations and determine minimum RCS level to support that f10wrate (using 1-AP-11, Loss of RHR).
(ALL) (G2.1.25, RO 3.9 / SRO 4.2)
N,R Evaluate and apply Tech Specs and procedure requirements Equipment Control based on UNSAT condition from QTRLY PORV Block valve surveillance (1-PT-44.7). (1)
(ALL) (G2.2.40, RO 3.4/SRO 4.7)
M,R Determine dose considering task performance with and without Radiation Control the option of installing temporary shielding using a Survey map.
(ALARA)
(ALL) (G2.3.14, RO 3.4/SRO 3.8)
M,R Classify event and determine PAR (modified, activity was Emergency Plan performed on previous 2 exams)
(SRO ONLY) (G2.4.41, RO 2.9/SRO 4.6)
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (~3 for ROs; ~ 00 for SROs & RO retakes)
(N)ew or (M)odified from bank (~ 1)
(P)revious 2 exams (~ 1; randomly selected)
(1) - For SRO Candidates there is one additional element that is not included in the RO JPM.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: North Anna Date of Examination:
Exam Level: RO 181 SRO-IO SRO-UO Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System 1 JPM Title (KA) Safety Type Code" Function
- a. Emergency Borate for stuck rods following Reactor Trip (1-ES-0.1) (MOV-1350 will not open) A,M,S,E,L 1 024 - AA2.01, Ability to determine and interpret the following as they apply to the Emergency Boration: Whether boron flow andlor MOVs are malfunctioning, from plant conditions (3.814.1)
(CFR: 43.5/45.13)
- b. Respond to misaligned control rod, during realignment rod will drop causing negative rate trip; A,N,S,EN 2 (turbine fails to trip causing SI actuation, SI fails to automatically actuate but can be manually actuated).
013 - A4.03, Ability to manually operate andlor monitor in the control room: ESFAS initiation (4.5/4.7) (CFR: 41.7/45.5 to 45.8)
- c. Perform Quarterly PZR heater output determination (1-PT-44A); breaker trips on 2 nd set (entry A,N,S 3 conditions for TS 3.4.9.B. 72 hrs for one inoperable backup group) 010 - A4.02, Ability to manually operate andlor monitor in the control room: PZR heaters (3.6/3.4) (CFR: 41.7/45.5 to 45.8)
- d. Respond to Loss of RHR in Mode 4 (1-AP-11) 4 (Pri) 005 - A2.03, Ability to (a) predict the impacts of the following malfunctions or operations on the A, M, P, L, S RHRS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: RHR pumplmotor Malfunction (2.9/3.1)
(CFR: 41.5/43.5/45.3/45.5) 11I
- e. Respond to failure of non-controlling 1 Stage pressure transmitter (1-AP-3); after transfer to A,M,S,P 4 (SEC) pressure mode controller setpoint fails low steam dumps must be placed in off (1-AP-38) 041 - A4.04, Ability to manually operate andlor monitor in the control room: Pressure Mode (2.7/2.7) (CFR: 41.7/45.5 to 45.8)
- f. Add Nitrogen to PRT D,S,P 5 007 - A 1.02, Ability to predict andlor monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Maintaining quench tank pressure (2.7/2.9) (CFR: 41.5/45.5)
- g. Respond to Recirc Spray Heat Exchanger radiation monitor alarm (1-AP-5) D,C,E 7 073 - A4.01, Ability to manually operate andlor monitor in the control room: Effluent release (3.9/3.9) (CFR: 41.7/45.5 to 45.8)
- h. Respond to loss of 1 or more Circ Water Pumps (1-AP-13) N,S 8 075 - A2.02, Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of Circulating Water Pumps (2.5/2.7)
(CFR: 41.51 43.51 45/31 45/13)
In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
- i. Open the residual heat removal heat exchanger cooling water return valves using a jumper (1- D,E,R 8 AP-28, 0-FCA-1).
008-A2.05, Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Effect of loss of instrument and control air on the position of the CCW valves that are air operated (3.3/3.5) (CFR: 41.5/43.5/45/3/45/13)
- j. Re-energize a 120-volt vital bus from its inverter (1-MOP-26.6, Q-AP-10) D,E 6 057 - AA1.01, Ability to operate and 1 or monitor the following as they apply to the Loss of Vital AC Instrument Bus: Manual Inverter Swapping (3.7/3.7) (CFR 41.7/45.5/45.6)
- k. Trip the reactor locally by opening the Reactor Trip Breakers or the rod-drive motor generator A,D,E 7 breakers 029 - EA1.12, Ability to operate and monitor the following as they apply to a ATWS: MIG set power supply and reactor trip breakers (4.1/4.0) (CFR 41.71 45.5 1 45.6)
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
'Type Codes Criteria for RO 1SRO-II SRo-U (A~temate path 4-6 14-6 12-3 (C)ontrol room (D)irect from bank ~9/~8/~4 (E)mergency or abnormal in-plant ~1/~1/~1 (EN)gineered safety feature
- 1 - 1 ~ 1 (control room system)
(L)ow-Power I Shutdown 11 11 ~ 1 (N)ew or (M)odified from bank including 1(A) 21 21 ~ 1 (P)revious 2 exams (similar topic) 31 31 ~ 2 (randomly selected)
(R)CA 1 1 11 ~ 1 (Siimulator
ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2
"'~"h I\~~~ Date of Examination:
~Y"m I ..v..l: RO 0 SRO-ID SRO-U 181 Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System I JPM Title (KA) Safety Type Code* Function
- a. Not used for SRO-U candidates N/A N/A
- b. Respond to misaligned control rod, during realignment rod will drop causing negative rate trip; A,N,S,EN 2 (turbine fails to trip causing SI actuation, SI fails to automatically actuate but can be manually actuated).
013 - A4.03, Ability to manually operate and/or monitor in the control room: ESFAS initiation (4.5/4.7) (CFR: 41.7/45.5 to 45.8)
- c. Perform Quarterly PZR heater output determination (1-PT-44A); breaker trips on 2nd set (entry A,N,S 3 conditions for TS 3.4.9.B. 72 hrs for one inoperable backup group) 010 - A4.02, Ability to manually operate and/or monitor in the control room: PZR heaters (3.6/3.4) (CFR: 41.7/45.5 to 45.8)
- d. Respond to Loss of RHR in Mode 4 (1-AP-11) 4 (Pri) 005 - A2.03, Ability to (a) predict the impacts of the following malfunctions or operations on the A, M, P, L, S RHRS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: RHR pump/motor Malfunction (2.9/3.1)
(CFR: 41.51 43.5 I 45.3 I 45.5)
- e. Not used for SRO-U candidates N/A N/A
- f. Not used for SRO-U candidates N/A N/A
- g. Not used for SRO-U candidates N/A N/A
- h. Not used for SRO-U candidates N/A N/A In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
- i. Open the residual heat removal heat exchanger cooling water return valves using a jumper (1- D,E,R 8 AP-28,O-FCA-1).
008-A2.05, Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Effect of loss of instrument and control air on the position of the CCW valves that are air operated (3.3/3.5) (CFR: 41.5/43.51 45/3 I 45/13)
- j. Not used for SRO-U candidates N/A N/A
- k. Trip the reactor locally by opening the Reactor Trip Breakers or the rod-drive motor generator A,D,E 7 breakers 029 - EA1.12, Ability to operate and monitor the following as they apply to a ATWS: MIG set power supply and reactor trip breakers (4.1/4.0) (CFR 41.7 145.5/45.6)
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room .
- Type Codes Criteria for RO 1 SRO-II SRO-U (A)ltemate path 4-6 1 4-6 12-3 (C)ontrol room (D)irect from bank ~9/~8/~4 (E)mergency or abnormal in-plant ~ 11 ~ 11 ~ 1 (EN)gineered safety feature 1 ~ 1 (control room system)
(L)ow-Power 1Shutdown 1 1 11 ~ 1 (N)ew or (M)odified from bank including 1(A) 21 2/~ 1 (P)revious 2 exams (similar topic) 31 3 1~ 2 (randomly selected)
(R)CA 1 1 1/ ~ 1 iSiimulator
ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2
~~~a SRO-118I SRO-U 0 Date of Examination:
Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System I JPM Title (KA) Safety Type Code* Function
- a. Emergency Borate for stuck rods following Reactor Trip (1-ES-0.1) (MOV-1350 will not open) A,M,S,E,L 1 024 - AA2.01, Ability to detennine and interpret the following as they apply to the Emergency Boration: Whether boron flow andlor MOVs are malfunctioning, from plant conditions (3.814.1)
(CFR: 43.5 I 45.13)
- b. Respond to misaligned control rod, during realignment rod will drop causing negative rate trip; A, N, S, EN 2 (turbine fails to trip causing SI actuation, SI fails to automatically actuate but can be manually actuated).
013 - A4.03, Ability to manually operate andlor monitor in the control room: ESFAS initiation (4.5/4.7) (CFR: 41.7/45.5 to 45.8)
- c. Perfonn Quarterly PZR heater output detennination (1-PT-44A); breaker trips on 2 nd set (entry A,N,S 3 conditions for TS 3.4.9.B. 72 hrs for one inoperable backup group) 010 - A4.02, Ability to manually operate andlor monitor in the control room: PZR heaters (3.6/3.4) (CFR: 41.7/45.5 to 45.8)
- d. Respond to Loss of RHR in Mode 4 (1-AP-11) 4 (Pri) 005 - A2.03, Ability to (a) predict the impacts of the following malfunctions or operations on the A, M, P, L, S RHRS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: RHR pumplmotor Malfunction (2.9/3.1)
(CFR: 41.5 I 43.5 I 45.3 / 45.5) at
- e. Respond to failure of non-controlling 1 Stage pressure transmitter (1-AP-3); after transfer to A,M.S,P 4 (SEC) pressure mode controller setpoint fails low steam dumps must be placed in off (1-AP-38) 041 - A4.04, Ability to manually operate andlor monitor in the control room: Pressure Mode (2.7/2.7) (CFR: 41.7/45.5 to 45.8)
- f. Not used for SRO-I candidates N/A N/A
- g. Respond to Recirc Spray Heat Exchanger radiation monitor alann (1-AP-5) D,C,E 7 073 - A4.01, Ability to manually operate andlor monitor in the control room: Effluent release (3.9/3.9) (CFR: 41.7/45.5 to 45.8)
- h. Respond to loss of 1 or more Circ Water Pumps (1-AP-13) N.S 8 075 - A2.02, Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of Circulating Water Pumps (2.5/2.7)
(CFR: 41.51 43.51 45/3 I 45/13)
In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
- i. Open the residual heat removal heat exchanger cooling water return valves using a jumper (1- D,E,R 8 AP-28, 0-FCA-1).
008-A2.05, Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Effect of loss of instrument and control air on the position of the CCW valves that are air operated (3.3/3.5) (CFR: 41.5/43.5/45/3/45/13)
- j. Re-energize a 120-volt vital bus from its inverter (1-MOP-26.6, 0-AP-10) D,E 6 057 - AA1.01, Ability to operate and I or monitor the following as they apply to the Loss of Vital AC Instrument Bus: Manual Inverter Swapping (3.7/3.7) (CFR 41.7 145.5 145.6)
- k. Trip the reactor locally by opening the Reactor Trip Breakers or the rod-drive motor generator A,D,E 7 breakers 029 - EA1.12, Ability to operate and monitor the following as they apply to a ATWS: MIG set power supply and reactor trip breakers (4.1/4.0) (CFR 41.7 I 45.5/ 45.6)
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room .
- Type Codes Criteria for RO 1 SRO-II SRO-U (A)ltemate path 4-6 1 4-8 12-3 (C)ontrol room (D)irect from bank !,9/!,81!,4 (E)mergency or abnormal in-plant .1/.1/.1 (EN)gineered safety feature -= 1:- 1 ~ 1 (control room system)
(L)ow-Power 1 Shutdown 1 1 11 ~ 1 (N)ew or (M)odified from bank including 1(A) 21 2/* 1 (P)revious 2 exams (similar topic) 31 31 ~ 2 (randomly selected)
(R)CA 11 11 ~ 1 (Slimulator
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS Power reduction was in progress per 1-PT-34.3, Turbine valve Freedom Test when a dropped rod occurred.
The Unit has been stabilized at 85% power with Tavg and Tref matched.
Unit 1 PCS is unavailable.
All ex-core power-range channels are operable.
The Instrument Shop has obtained power range detector current readings.
A copy of 1-PT-22.4, Attachment 4, Unit 1 Power Range Calibration Data is available.
INITIATING CUE You are requested to:
- 1) Perform a Quadrant Power Tilt Ratio determination by hand calculation in accordance with 1-PT-23.
And
- 2) Based on your calculation provide the SRO with the maximum power level allowed by Technical Specification 3.2.4.
02/28/08 Page: 1 of 7
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM R709(M)
Determine the Quadrant Power Tilt Ratio by hand calculation (1-PT-23) and determine maximum allowed power level based on the calculation.
TASK STANDARDS Maximum QPTR correctly calculated, Quadrant of Max QPTR correctly identified, maximum power level allowed by TS 3.2.4 correctly determined.
KIA
REFERENCE:
G2.1.7 (4.4/4.7)
ALTERNATE PATH:
NIA TASK COMPLETION TIMES Validation Time = 24 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ ] SATISFACTORY [ ] UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature 1 Date EVALUATOR'S COMMENTS 02/28/08 Page: 2of7
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM R709 READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES The trainee has completed the applicable course knowledge training at the reactor operator level.
INITIAL CONDITIONS Power reduction was in progress per 1-PT-34.3, Turbine valve Freedom Test when a dropped rod occurred.
The Unit has been stabilized at 85% power with Tavg and Tref matched.
Unit 1 PCS is unavailable.
All ex-core power-range channels are operable.
The Instrument Shop has obtained detector current readings.
02/28/08 Page: 3 of 7
A copy of 1-PT-22.4, Attachment 4, Unit 1 Power Range Calibration Data is available.
INITIATING CUE You are requested to:
- 1) Perform a Quadrant Power Tilt Ratio determination by hand calculation in accordance with 1-PT-23.
And
- 2) Based on your calculation provide the SRO with the maximum power level allowed by Technical Specification 3.2.4.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT Calculator Copy of 1-PT-22.4, Attachment 4, Unit 1 Power Range Calibration Data.
PERFORMANCE STEPS START TIME 02/28/08 Page: 4 of 7
~ Record the expected 100% power current readings in attachment 2. IProcedure Step ISATl] UNSATl]
IStandards IData correctly transcribed from the copy of 1-PT-22.4 provided.
IN otes/Comments
~ Calculate the normalized detector currents and averages. IProcedure Step ISAT l] UNSAT l ]
I...;;;S;;;ta;;;n,;,;;;d;;;;;a,;,;rd;,;;;s~_ _ 1 Normalized detector currents and averages are calculated.
I NoteS/Comments
~ Calculate the guadrant power tilt ratios. I Procedure Step I_C_r_iti_c_al_S_t_e.L-P 1 SAT [] UNSAT l]
I Standards IUpper and lower guadrant power tilt ratios are calculated.
[Notes/Comments 02/28/08 Page: 5 of?
Record the value and location of the largest quadrant power tilt Procedure Step _ _
ratio.
I,-C_r_iti_c_a_1S_t_e-'-p 1 SAT [1 UNSAT [1 Standards -7 Maximum QPTR determined to be 1.03738 (acceptance criteria: +/- 0.001)
-7 Location determined to be N43L.
rates/com ments
~ Identifies QPTR exceeds maximum permissible value of 1.02. IProcedure Step ISAT [] UNSAT [ ]
IStandards IQPTR determined to exceed LeO 3.2.4 based on calculation.
INotes/comm ents 02/28/08 Page: 6 of 7
~ Determine maximum allowable power level per TS 3.2.4. IProcedure Step ICritical Step ISAT [1 UNSAT [ 1 Standards Maximum power level determined to be 88.79%
(from TS 3.2.4 power must be reduced ~ 3% from RTP for each 1% of QPTR> 1.00. Therefore 3.738% X 3 = 11.214).
For conservatism power may be rounded down to 88%; based on the acceptance criteria of step 5 the upper range is 89.086% (rounding to 89.1 is acceptable).
rates/com ment,
>>>>> END OF EVALUATION <<<<<
STOP TIME 02/28/08 Page: 7 of 7
Key Administraive JPM Perform QPTR DETECTOR 100% Current ACTUAL NORMALIZED TILT RATIO N41U 156.6 131 0.836526 1.00174 N41L 184.6 155 0.839653 1.00001 N42U 165.1 138 0.835857 1.00094 N42L 167.8 141 0.840286 1.00076 N43U 167.6 145 0.865155 1.03603 N43L 207.8 181 0.871030 1.03738 N44U 204.3 164 0.802741 0.96129 N44L 220.4 178 0.807623 0.96186 Maximum Power 88.79
'.U I 1-PT-22.4 Revision 13 Page 27 of 29
.. , "'= "" " . '= (Page ......'....,.,..,...
30f~4)~" ~~'-"-""'-=-o..~""'-""~~~,-,-,-,- ........~,-"=-=..........=o,,,,,,_=~, .. ' ",
Attachment 4 Unit 1 Power Range Calibration Data PCS POINTS Please enter the following PCS Computer Point values to reflect changes when the Power Range detectors, are calibrated AND initial the appropriate spaces. IF the PCS computer is inoperable, THEN mark the corresponding blanks N/A:
- 1. Expected Current Values: Updated K0821 N41 Upper 100% Expected Current: ' hi
~
K0822 N41 Lower 100% Expected Current:
K0823 N42 Upper 100% Expected Current: , ibs-. {
K0824 N42 Lower 100% Expected Current:
K0825 N43 Upper 100% Expected Current:
167.8 1fg
~
K0826 N43 Lower 100% Expected Current:
K0827 N44 Upper 100% Expected,Current: ~lf. 3 '.
K0828 N44 Lower 100% Expected Current:
- 2. AFD Calibration Constant Values: Updated K0855DF Calibration Constant for N41:
f8.C:.32o . l,)C.
K0856 DF Calibration Constant for N42: tl K0857 DF Calibration Constant for N43:
tI K0858 DF Calibration Constant for N44:
td(
pes
- 3. Incore Versus Excore Axial Offset Slope Values: Updated K0551 InJExcore Offset Ratio (Ch41, Q4): '(.>b'2't .
K0552 InJExcore Offset Ratio (Ch42, Q2):
K0553 InJExcore Offset Ratio (Ch43, Q I):
K0554 InJExcore Offset Ratio (Ch44,. Q3):
.~~d--- to-.g~D1_
Date Reviewed By to .:-r p ~C?7
, Date Calibration is complete and copy of this page has been placed in Reactor Data Book: .
Instrument Department
PROCEDURE NO:
'Dominion* 1-PT-23 REVISION NO:
NORTH ANNA POWER STA TION 29 PROCEDURE TYPE: UNIT NO:
OPERATIONS PERIODIC TEST 1 PROCEDURE TITLE:
QUADRANT POWER TILT RATIO DETERMINATION TEST FREQUENCY: UNIT CONDITIONS REQUIRING TEST:
- 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during steady state operation when the QPTR alarm is Inoperable SPECIAL CONDITIONS: None SURV REACT REQ MGT REVISION
SUMMARY
FrameMaker Template Rev. 030.
- Incorporated Plant Issue N-2005-4402-R5, Enhancements to Maintenance Steps in other Department Procedures, by adding PI to reference section step 2.4.1.
- Deleted "Fluke Model 8840 with" from Special Tools and Equipment table in section 5.
- Changed Shift Supervisor to SRO throughout procedure due to title change.
REASON FOR TEST (CHECK APPROPRIATE BOX):
o Surveillance o Post-Maintenance Work Order Number (Post-Maintenance Only): ______________
TEST PERFORMED BY (SIGNATURE): DATE STARTED: DATE COMPLETED:
TEST RESULT (CHECK APPROPRIATE BOX): WORK REQUEST NUMBERS AND DATE:
o Satisfactory o Unsatisfactory D Partial THE FOLLOWING PROBLEM(S) WERE ENCOUNTERED AND CORRECTIVE ACTIONS TAKEN:
______________________________________________ ~ ~__________ (Use back~raddffion~remarks.)
COGNIZANT SUPERVISOR or DESIGNEE: DATE:
ADDITIONAL REVIEWS: DATE:
Reactor Engineer:
CONTINUOUS USE
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 2 of 15 TABLE OF CONTENTS Section Page 1.0 PURPOSE 3
2.0 REFERENCES
4 3.0 INITIAL CONDITIONS 6 4.0 PRECAUTIONS AND LIMITATIONS 6 5.0 SPECIAL TOOLS AND EQUIPMENT 8 6.0 INSTRUCTIONS 9 7.0 FOLLOW-ON 9 ATTACHMENTS 1 Unit 1 PCS QPTR Calculation 11 2 QPTR Hand Calculation 13
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 3 of 15 1.0 PURPOSE To provide instructions for determining the core Quadrant Power Tilt Ratio (QPTR) by Computer Calculation or by hand calculation as required by Tech Spec SR 3.2.4.1.
The following Synopsis is designed as an aid to understanding the procedure, and is not intended to alter or take the place of the actual purpose, instructions, or text of the procedure itself.
The Quadrant Power Tilt Ratio (QPTR) is defined in the Technical Specifications as the ratio of the maximum upper excore detector output to the average upper excore detector outputs, or the maximum lower excore detector calibrated output to the average of the lower excore detector outputs, whichever is greater. With one excore detector inoperable and power below 75% RTP, the remaining three detectors are used for computing the average. With one excore detector inoperable and power above 75% RTP, QPTR is monitored using 1-PT-23.1 or l-PT-21.1. As per Tech Spec 3.2.4, the maximum QPTR is 1.02 when reactor power is above 50% of Rated Thermal Power. The purpose for this limit is as follows:
- To ensure that the radial power distribution satisfies design limits.
- To provide DNB protection with radial (x-y) power tilts.
- To provide linear heat generation rate protection for radial (x-y) power tilts.
- To ensure that FQ(z) and F~H remain within limits.
The QPTR is monitored to limit any gross changes in the radial power distribution between full core flux maps. The QPTR alarm setpoint is set as low as possible without causing spurious alarms.
DCP 95-161 installed test jacks in the front of the power range detector drawers.
Since the current meters on the front of the power range drawer are prone to meter drift and static charge effects, these jacks are the preferred way to perform the QPTR calculation by hand if the Unit 1 PCS is inoperable. These test jacks allow a DMM to be easily used to measure the voltage drop across the precision resistor which is in parallel to the face meter. A conversion of 1 millivolt = 1 )lamp is then used to obtain the detector current.
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 4 of 15
(
It should be noted that the test jacks may only be used if no Power Range Channel has been placed in trip. This satisfies the single failure criteria assumed in the safety analysis which permits use of the test jacks. If one Power Range Channel is in trip and the Unit 1 PCS is inoperable, then currents read from the meters on front of the power range drawers will have to be used in the QPTR calculations.
There are multiple methods of monitoring QPTR available to the OATC. These methods include the following:
- BOTH Annunciator Panel A-C7, NIS PR UP DET DEV-DEF <50% and Annunciator Panel A-C8, NIS PR LWR DET DEV-DEF <50%
- Annunciator Panel A-D4, CMPTR ALARM PR TILT
- PCS point #Ul170 With reactor power between 50% and 75% RTP, QPTR may be calculated with only three operable Power Range Channels, in accordance with Tech Spec SR 3.2.4.1, although as a good practice the first QPTR calculation should be compared with a symmetric or full incore flux map, to account for the loss of accuracy when one power range channel is removed from the average. With reactor power 75% RTP or greater, and only three Power Range Channels operable, then this procedure is not necessary, in accordance with Tech Spec SR 3.2.4.2, and the symmetric or full incore flux map is used to determine QPTR.
2.0 REFERENCES
2.1 Source Documents 2.1.1 UFSAR Section 4.3.2.2, Power Distribution 2.1.2 UFSAR Section 4.4.2.4, Flux Tilt Considerations 2.1.3 UFSAR Section 4.4.5.1, Incore Instrumentation 2.1.4 UFSAR Section 4.4.5.3, Instrumentation to Limit Maximum Power Output 2.1.5 UFSAR Section 7.7.1.3.1, Monitoring Functions Provided by the Nuclear Instrumentation System
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 5 of 15 2.2 Technical Specifications 2.2.1 Tech Spec 3.2.4 2.2.2 Tech Spec SR 3.2.4.1 2.2.3 Tech Spec SR 3.2.4.2 2.2.4 Tech Spec SR 3.3.1.1, Table 3.3.1-1, Items 2 and 3 2.3 Technical References 2.3.1 I-PT-23.1, Quadrant Power Tilt Ratio Determination Using the Incore Detector System 2.3.2 I-PT-21.1, Reactor Core Flux Mapping 2.3.3 Control Room Reactor Data Book (NASES 3.05) 2.3.4 Calculational Basis, approved 03-13-86 2.3.5 Westinghouse Position Statement on Core Tilt, VRA-92-034 (3/18/92) 2.3.6 DCP 95-161, Install NI Power Range Test Points for Unit 1 2.3.7 DCP 96-005, P-250 Upgrade 2.3.8 DCP 01-007, Phase 2 PCS Installation and P-250 Removal- Unit 1 2.4 Commitment Documents 2.4.1 Plant Issue N-2005-4402-R5, Enhancements to Maintenance Steps in other Department Procedures
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 6 of 15 Init Verif 3.0 INITIAL CONDITIONS NOTE: IF power level is 75% or greater AND any power range detector is inoperable, THEN do not perform this PT. I-PT-23.1, Quadrant Power Tilt Ratio Determination Using the Incore Detector System, must be performed.
3.1 IF power level is 75% or greater, THEN all 4 power range detectors are OPERABLE.
3.2 Verify at least three Excore Power Range Channels are operable.
3.3 Verify the Reactor is at a stable power level.
3.4 Notify SRO of this test.
NOTE: For the purposes of this procedure, the Unit 1 PCS is considered inoperable if any of the Power Range Channel inputs to the Unit 1 PCS are bad or unreliable.
3.5 IF the Unit 1 PCS is inoperable AND all four Excore Power Range Channels are operable, THEN notify the Instrument Department to send a technician with the equipment listed in Section 5.0 to the control room to measure voltages from the NI Power Range Detector drawer test jacks.
4.0 PRECAUTIONS AND LIMITATIONS 4.1 Comply with the following guidelines when marking steps N/A:
- IF the conditional requirements of a step do not require the action to be performed, THEN mark the step N/A.
- IF any other step is marked N/A, THEN have the SRO or designee approve the N/A AND submit a Procedure Action Request (PAR).
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 7 of 15 4.2 With input from one Power Range Channel inoperable and reactor power is less than or equal to 75% RTP, the remaining three power range channels can be used for calculating QPTR, in accordance with Tech Spec SR 3.2.4.1.
4.3 Observe the following when using the test jacks on the front of the NI Power Range Detector drawers:
- Only qualified individuals may use M&TE to measure the voltage from the test jacks.
- In order to ensure proper channel separation, connect a single DMM to only one NI Power Range Detector channel at a time. Both the Upper and Lower Detector test jacks of one Power Range Detector channel may be connected to the DMM at the same time.
- To further enhance channel isolation, do NOT connect multiple DMMs to the test jacks of two or more Power Range Detector channels at one time.
- To reduce the possibility of adversely impacting the NI Power Range Detector channels, minimize the amount oftime the DMM is connected to the test jacks.
- To avoid possible signal spikes due to static charge on the test leads, ground test leads to drawer case and short together prior to making connection to test jacks.
- The DMM shall NOT be left unattended while connected to the NI Power Range Detector drawer.
- ALL Power Range Detector channels shall be checked to ensure no Trip signal is locked in on ANY channel prior to connecting the DMM to ANY drawer.
- The DMM should not be connected to the test jacks of an inoperable NI Power Range Detector channel.
4.4 IF one Power Range Detector channel is in trip AND the Unit 1 PCS is inoperable, THEN QPTR MUST be determined by Hand Calculation using currents read from the analog current meters on the front of the operable Power Range Detector channels. It should be noted that due to possible static charge effects on the meters, this is the least accurate method to determine QPTR.
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 8 of 15 5.0 SPECIAL TOOLS AND EQUIPMENT IF the Unit 1 PCS is inoperable AND all four Power Range Channels are operable, THEN do the following:
- Obtain the equipment listed below and verify the equipment is calibrated to standards traceable to the National Institute of Standards and Technology M & TE Type Accuracy Requirement DMM At least 5 1/2 digits and ~ 1000 Megohm input resistance on 2 VDC range
- Obtain test leads to connect the DMM above to a double banana jack (0.75 inch centers).
- Record the following for the Digital Multimeter:
NQCNo.: Cal Due Date:
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 9 of 15 6.0 INSTRUCTIONS NOTE: Sections OR Attachments of this procedure NOT used may be discarded.
6.1 IF one Power Range Channel has become inoperable and Thermal Power is less than 75 percent RTP, THEN do ONE of the following:
6.1.1 The first perfonnance of this test shall be done concurrently with one of the following Periodic Test procedures. Mark procedure not perfonned N/A.
- 1-PT-23.1, Quadrant Power Tilt Ratio Detennination Using the Incore Detector System
- 1-PT-21.1, Reactor Core Flux Mapping 6.1.2 Obtain Reactor Engineer concurrence that an incore flux map is not required RXENG for comparison to the first QPTR calculation after the Power Range Channel has become inoperable.
NOTE: For the purposes of this procedure, the Unit 1 PCS is considered inoperable if the Power Range Channel inputs to the Unit 1 PCS are bad or unreliable.
6.2 Do at least one of the following to detennine QPTR: Mark methods not used N/A.
- IF desired to use the Unit 1 PCS to detennine QPTR, THEN complete Attachment 1, Unit 1 PCS QPTR Calculation.
- IF desired to perfonn a hand calculation to detennine QPTR, THEN complete Attachment 2, QPTR Hand Calculation.
7.0 FOLLOW-ON 7.1 Acceptance Criteria 7.1.1 A QPTR has been detennined from at least 3 Operable Excore Power Range detectors.
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 10 of 15 7.1.2 IF the Unit is at a power level above 50 percent, THEN the QPTR is 1.02 or less.
7.2 Follow-On Tasks NOTE: Sections OR Attachments of this procedure NOT used may be discarded.
7.2.1 IF any QPTR Report(s) were obtained, THEN attach them to this procedure.
7.2.2 IF the Unit is at a power level above 50 percent AND the QPTR is greater than 1.02, THEN do the following:
- Notify the SRO.
- Enter Action Statement of Tech Spec Action 3.2.4.A.
7.2.3 IF the QPTR is greater than 1.015, THEN notify the Reactor Engineer.
7.3 Completion Notification Notify the SRO that this test is complete.
Completed by: _ Date: _
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 11 of 15 (Page 1 of 2)
Attachment 1 Unit 1 PCS QPTR Calculation
- 1. Detennine the appropriate value for Unit I PCS constant K0829 below, (.I) check one:
o IF all Power Range Detector channels are operable, THEN K0829 should be O.
o IF N41 is inoperable, THEN K0829 should be 1.
o IF N42 is inoperable, THEN K0829 should be 2.
o IF N43 is inoperable, THEN K0829 should be 3.
o IF N44 is inoperable, THEN K0829 should be 4.
- 2. Display addressable constant K0829.
- 3. IF the present value ofK0829 is NOT equal to the value detennined in Step 1, THEN sv enter the appropriate value by selecting the UPDATE A CONSTANT button from the System Menu screen.
- 4. Request a QPTR Report by doing the following:
4.1 Select the NSSS AND BOP button from the Main Screen.
4.2 Select the QPTR button from the NSSS Menu Screen.
4.3 Select the F4=Report button from the NI: Quadrant Power Tilt Ratio screen.
4.4 IF using PCS, THEN select PRINT to print the QPTR Report.
- 5. WHEN the QPTR Report is complete, THEN remove the report from the printer and examine the report as follows:
5.1 IF any data is indicated as "NCAL" on the QPTR Report, THEN notify the Reactor Engineer for resolution.
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 12 of 15 (Page 2 of 2)
Attachment 1 Unit 1 PCS QPTR Calculation 5.2 IF any data indicated as "NCAL" cannot be resolved, THEN QPTR CANNOT be determined using the Unit 1 PCS. Mark the remainder of this attachment N/A AND complete Attachment 2, QPTR Hand Calculation.
5.3 IF no data is indicated as "NCAL" on the QPTR Report, THEN sign off the Current Verification section of the QPTR Report.
CAUTION The Expected lOO-percent currents from the Reactor Data Notebook should be used. IF new currents are being installed and a "mixture" of "OLD" and "NEW" Expected lOO-percent currents are used, THEN the QPTR will be inaccurate.
- 6. Compare the Expected lOO-percent currents from the Reactor Data Notebook to the NI Power Range Detector channel currents on the QPTR Report.
- 7. IF the Reactor Data Book Expected lOO-percent currents match the Expected lOO-percent currents on the QPTR Report, THEN sign off the QPTR Report for the expected 1OO-percent currents.
- 8. IF the Reactor Data Book Expected IOO-percent currents do NOT match the Expected IOO-percent currents on the QPTR Report, THEN do the following:
8.1 Notify the Reactor Engineer to resolve the discrepancy.
8.2 IF discrepancy cannot be resolved, THEN QPTR CANNOT be determined using the Unit I PCS. Complete Attachment 2, QPTR Hand Calculation.
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 13 of 15 (Page 1 of 3)
Attachment 2 QPTR Hand Calculation CAUTION To ensure adequate channel separation, only one NI Power Range Detector channel may be monitored at a time by a single DMM using the test jacks on the front of the drawers (See Precaution & Limitation Step 4.3).
- 1. IF no Power Range Detector Channel is in trip AND the Unit 1 SRO desires that the test jacks be used, THEN determine the NI Power Range Detector channel currents for each channel by doing the foHowing:
1.1 Notify the Instrument Department to send a technician with the equipment listed in Section 5.0 to the control room to measure voltages from the NI Power Range Detector drawer test jacks.
1.2 Have the Instrument Department technician place the DMM on the 2 VDC scale.
1.3 Have the Instrument Department technician ground the test leads to the drawer case and short together prior to making each connection to reduce the possibility of adversely impacting the Power Range Detector channel.
1.4 TM: Have the Instrument Department technician connect the DMM to the test jacks of each NI Power Range Detector channel one at a time and read the voltage indicated on the meter to the nearest millivolt.
1.5 Determine each NI Power Range Detector channel current by using the conversion 1 millivolt = 1 /lamp.
1.6 Record the measured currents on Page 3 of this attachment.
1.7 TM: Have the Instrument Department technician disconnect the DMM from the test jacks.
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 14 of 15 (Page 2 of 3)
Attachment 2 QPTR Hand Calculation
- 2. IF one Power Range Channel is in trip OR the Unit 1 SRO desires that the current meters on the Power Range drawers be used, THEN read the current meters on the front of each operable Power Range upper and lower Detector Channel AND record the measured currents on Page 3 of this attachment.
- 3. IF the Unit 1 PCS is operable AND the Unit 1 SRO desires to use the Unit 1 PCS to obtain Current Readings, THEN do the following:
3.1 Request a QPTR Report by doing the following on the Unit 1 PCS:
- c. Select the F4=Report button from the NI: Quadrant Power Tilt Ratio screen.
3.2 WHEN the QPTR Report is complete, THEN remove the report from the printer.
3.3 Using the QPTR Report, record the appropriate current readings in the Current Reading spaces provided on Page 3 of this attachment.
- 4. Complete the calculations on Page 3 of this attachment.
- 5. Have a qualified individual complete an independent verification of all calculations.
DOMINION 1-PT-23 North Anna Power Station Revision 29 Page 15 of 15 (Page 3 of 3)
Attachment 2 QPTR Hand Calculation NOTE: Current Readings may be obtained from the face meters of the Power Range "B" Drawer OR from the test jacks OR from the Unit 1 PCS, if operable.
Avg. of Upper N-41 N-42 N-43 N-44 Normalized Description Upper Upper Upper Upper Currents Current Reading (record to nearest Jla)
Expected Current (record as shown in Reactor Data Book or I-PT-22.4)
Normalized Current (Current Reading / Expected Current)
(4 decimal places)
QPTR (Normalized Current / Avg. of Norm. Currents)
(4 decimal places)
Avg. of Lower N-41 N-42 N-43 N-44 Normalized Description Lower Lower Lower Lower Currents Current Reading (record to nearest j.la)
Expected Current (record as shown in Reactor Data Book or I-PT-22.4)
Normalized Current (Current Reading / Expected Current)
(4 decimal places)
QPTR (Normalized Current / Avg. of Norm. Currents)
(4 decimal places)
Maximum QPTR: _ (Record Maximum Upper or Lower QPTR Value from above)
Quadrant of Max QPTR: _ (N41 Upper, N41 Lower N42 Upper, N42 Lower, etc.)
Completed by: Date:
Verified by: _
(Use this Time for recording Date/Time Verification Completed: _
when surveillance is completed)
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS The unit was shutdown 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> ago.
RCS draindown is in progress with the following equipment in operation:
- 1 RHR pump
- 2 RHR Heat exchangers
- 1 Charging pump The crew has implemented 1-AP-11, Loss of RHR and is at step 4b.
INITIATING CUE Using 1-AP-11, Loss of RHR, determine the following:
- 1. Minimum required RHR flow.
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM Determine minimum required RHR flow and RCS level.
TASK STANDARDS Given a copy of 1-AP-11, Loss of RHR, the examinee will determine:
- Minimum required RHR flow based on time after shutdown.
KIA
REFERENCE:
G2.1.25 (3.9/4.2)
ALTERNATE PATH:
NIA TASK COMPLETION TIMES Validation Time = 10 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ 1SATISFACTORY [ 1UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature 1 Date EVALUATOR'S COMMENTS Page 2 of6
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
INITIAL CONDITIONS The unit was shutdown 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> ago.
RCS draindown is in progress with the following equipment in operation:
- 1 RHR pump
- 2 RHR Heat exchangers
- 1 Charging pump The crew has implemented 1-AP-11, Loss of RHR and is at step 4b.
Page 3 of6
INITIATING CUE The task you are to perform is:
Using 1-AP-11, Loss of RHR, determine the following:
- Minimum required RHR flow.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT Calculator PERFORMANCE STEPS START TIME
)
cr=J Convert hours to days for use on att.3 I Procedure Step C~r.::....:it:....::.ic...::..::a_1S-'-t.:..-=-e...Lp I -: 1 SAT [1 UNSAT [1 Standards Operator determines time after shutdown for use on attachment 3 is 7days (168hrs/24hrs per day =7 days).
_ _ _-.!:I=N=ot=e=s/=c=o=m=m=en=t=s=:=========================::;1
1 Page 4 of6
~ Determine minimum required RHR flow IProcedure Step: Att.3 I_C_r_iti_c_a_15_t_e..p. . . 1 SAT [1 UNSAT [1 Standards Operator determines minimum flowrate of 2600 gpm.
Acceptance criteria of 2500-2700 gpm based on readability and figure minor increments of 200 gpm.
Notes/Comments: Operator will arrive at a value of 3800 gpm if the incorrect figure in Att. 3 is used.
Determine level that RHR pumps are required to be stopped based Procedure Step: Att.2 on minimum re uired RHR flow.
II-C_r_it_ic_a_15_t_e-'-p 1 SAT [1 UNSAT [1 Standards Operator determines minimum level of 7.25 in. above CL.
Acceptance criteria of 7.1 in. to 7.4 in. based on acceptance criteria for flow rate from previous step, and readability and figure minor increments of 0.2".
INotes/Comments: I
====================::::::;1
>>>>> END OF EVALUATION <<<<<
STOP TIME Page 5 of6
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
JOB PERFORMANCE MEASURE CHECKLIST Calculator Page 6 of6
NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR (WITH ELEVEN ATTACHMENTS) PAGE 1 of 23 PURPOSE To provide instructions for maintaining Core Cooling and protecting the Reactor Core in the event that RHR Cooling is lost.
ENTRY CONDITIONS This procedure is entered when RHR is required for Core Cooling and any of the following conditions exist:
- Air-binding of operating RHR pumps as indicated by:
- Flow oscillations, or
- Motor amps fluctuating, or
- Excessive pump noise.
- Annunciator "E" Panel A-6, RHR PP 1A AUTO TRIP, is LIT, or
- Annunciator "E" Panel A-7, RHR PP 1B AUTO TRIP, is LIT, or
- Annunciator "E" Panel A-8, RHR SYSTEM La FLOW, is LIT, or
- Loss of RHR pumps due to loss of power, or
- Loss of Service Water System with RHR System in service, or Loss of Component Cooling System with RHR System in service.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 2 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION:
- RCS make-up concentration MUST be greater than or equal to current Shutdown Margin and Boron concentration requirements of the COLR.
- Changes in RCS pressure due to boiling in the core can result in Reactor Vessel water level changes that may not show on RCS standpipe level indicator 1-RC-L1-1 03.
- 1. CHECK RCS LEVEL - DECREASING 0 GO TO Step 5.
0
- RCS standpipe level -
DECREASING OR 0
- RCS ultrasonic level indicator -
DECREASING OR 0
- PRZR level - DECREASING OR 0
- RCS makeup rate -
INCREASING OR 0
- PDn pumping frequency -
UNEXPLAINED INCREASE
- 2. INCREASE RCS MAKEUP FLOW
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 3 of 23 ACTIONI EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 3. ISOLATE RCS DRAIN PATHS:
a) Check the following Letdown o a) Manually close valves.
Isolation Valves - CLOSED:
0
- 1-CH-HCV-1200A 0
- 1-CH-HCV-1200B 0
- 1-CH-HCV-1200C 0
- 1-CH-LCV-1460A 0
- 1-CH-LCV-1460B o b) Check 1-CH-HCV-1142, RHR o b) Manually close valve.
System to Letdown Isolation Valve
- CLOSED c) Check loop drains - CLOSED: o c) Manually close valves.
o
- 1-RC-HCV-1557A o
- 1-RC-HCV-1557B o
- 1-RC-HCV-1557C d) While continuing with procedure, o d) Ensure valves are closed.
verify the following valves -
LOCKED CLOSED:
o
- 1-RH-36, Residual Heat Removal to RWST Isolation Valve (Containment) o
- 1-RH-34, Residual Heat Removal Supply to RP (Containment)
(STEP 3 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 4 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 3. ISOLATE RCS DRAIN PATHS:
(Continued)
D e) Close any known RCS drain paths D f) Initiate actions to stop level decreases due to maintenance covered by 0-GOP-13.3, ASSESSMENT OF MAINTENANCE ACTIVITIES FOR POTENTIAL LOSS OF REACTOR COOLANT INVENTORY CAUTION:
- Changes in RCS pressure can result in Reactor Vessel water level changes that may not show on RCS level indicator 1-RC-L1-1 03.
- 4. VERIFY ADEQUATE RCS MAKEUP FLOW:
D a) Check RCS level - STABLE OR a) Ensure the keylock switch for 1-RC-L1-105, INCREASING Independent RCS Level Indicator, is in ENABLE. GO TO appropriate procedure:
D
- 1-AP-52, LOSS OF REFUELING CAVITY LEVEL DURING REFUELING (STEP 4 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 5 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 4. VERIFY ADEQUATE Res MAKEUP FLOW: (Continued) b) Check RHR flow - LESS THAN OR 0 b) Reduce RHR flow to design flow rate of EQUAL TO DESIGN FLOW OF ATIACHMENT 3.
ATIACHMENT 3:
0
Page 1 of 2 0
THAN MINIMUM FOR INDICATED FLOW OF ATTACHMENT 2 0 1) Continue RCS makeup.
0 2) Stop RHR Pumps.
0 3) GO TO Step 11.
0 d) Check RCS level - AT LEAST 0 d) Increase RCS level to greater than +10 inches
+10 INCHES ABOVE above centerline.
CENTERLINE 0 .!E level cannot be increased to greater than
+10 inches above centerline, THEN GO TO 1-AP-17, SHUTDOWN LOCA.
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 6 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 5. VERIFY RHR ISOLATION VALVES -
OPEN:
a) RHR Inlet Isolation Valves - OPEN a) Do the following:
D
- 1-RH-MOV-1700 D 1) Stop RHR Pump(s).
D
- 1-RH-MOV-1701 D 2) Reduce RCS pressure as necessary.
D 3) WHEN RCS pressure is less than 418 psig, THEN open valves.
b) AT least one RHR Outlet Isolation D b) Open at least one RHR Outlet Isolation Valve.
Valve - OPEN D
- 1-RH-MOV-1720A D
- 1-RH-MOV-1720B CAUTION: RHR flow less than minimum requirements may cause RCS temperature to increase.
NOTE:
- Operating at low RHR system flow rates during reduced inventory operations greatly reduces the risk of air entrainment (vortexing).
- Indications of a pump sheared shaft are low flow and low motor amps. A degraded pump or a pump with a sheared shaft is to be considered as NOT running.
- 6. CHECK ONE RHR PUMP - RUNNING Do the following:
D a) !E the other RHR pump is available, THEN stop any degraded RHR pump.
(STEP 6 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 7 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 6. CHECK ONE RHR PUMP -
RUNNING (Continued) o b) IF a degraded RHR pump is running AND the other RHR pump is NOT available, THEN GO TO Step 7.
c) !E electrical power is available, THEN do the following:
- 1) Manually close the following RHR Control Valves:
o
- 1-RH-FCV-1605 o
- 1-RH-HCV-1758 o 2) !.E an RHR Pump was previously stopped due to air entrainment, THEN locally vent both RHR Pumps.
o 3) IF both RHR pumps are stopped, THEN start one RHR pump.
o
- 1-RH-HCV-1758 o
- 1-RH-FCV-1605 o 5) IF an RHR Pump has been started, THEN GO TO Step 7.
o !E no RHR Pump can be started, THEN GO TO Step 11.
(STEP 6 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 8 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 6. CHECK ONE RHR PUMP -
RUNNING (Continued) d) !E electrical power is NOT available, THEN do the following:
D 1) Initiate O-AP-10, LOSS OF ELECTRICAL POWER.
D 2) GO TO Step 11 .
- 7. VERIFY RHR SYSTEM - NORMAL: Do the following:
D
D
- RHR Motor amps - STABLE +10 inches above centerline by increasing charging flow.
D
- RCS temperature - STABLE
- 2) Check RHR flow - less than or equal to design flow of ATIACHMENT 3:
D
!E RHR flow is greater than the design flow rate of ATIACHMENT 3, THEN reduce flow to the design flowrate using:
D
- 1-RH-HCV-1758 D
- 1-RH-FCV-1605 (STEP 7 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 9 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7. VERIFY RHR SYSTEM - NORMAL:
(Continued)
D 3) Check RCS level - Greater than minimum for indicated flow of ATTACHMENT 2.
D IF RCS level is not greater than minimum for indicated flow of ATIACHMENT 2, THEN STOP the RHR Pumps and GO TO Step 11.
- 4) Send an Operator to locally check pump operation:
D
- RHR pump noise D
- RHR pump seals D
- RHR pump vibration D b) !E the running RHR pump is degraded AND the other RHR pump is available, THEN RETURN TO Step 6.
D c)!E RHR System cannot be stabilized, THEN stop running RHR Pump AND GO TO Step 11.
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 10 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 8. CHECK SERVICE WATER TO CC HEAT EXCHANGER - AVAILABLE:
0 a) Verify Service Water System - IN a) IF Service Water flow is NOT available, THEN SERVICE initiate the following while continuing with this procedure:
0
- 0-AP-12, LOSS OF SERVICE WATER 0
- 1-AP-15, LOSS OF COMPONENT COOLING 0 GO TO Step 11 .
b) Verify Service Water Supply Valves b) Open Service Water Supply Valves to CC to CC System - OPEN: System:
0
- 1-SW-MOV-108A 0
- 1-SW-MOV-108A 0
- 1-SW-MOV-108B 0
- 1-SW-MOV-108B 0 c) Locally check Service Water to CC 0 c) GO TO Step 11.
Heat Exchanger LlP - NORMAL
( NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 11 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
EXCHANGERS - NORMAL:
o
o
- 1-CC-FI-132B o
- 1-CC-TV-103A, A RHR Heat Exchanger Return Isolation o
- 1-CC-TV-103B, B RHR Heat Exchanger Return Isolation o
- 1-CC-MOV-100A, A CC Heat Exchanger Outlet Isolation o
- 1-CC-MOV-100B, B CC Heat Exchanger Outlet Isolation b) IF either 1-CC-TV-103A or 1-CC-TV-103B cannot be opened, THEN close the associated RHR CC MOV:
o
- 1-CC-MOV-1 OOA for 1-CC-TV-1 03A o
- 1-CC-MOV-100B for 1-CC-TV-103B o c) IF CC flow is restored, THEN GO TO Step 10.
o d) IF CC is NOT restored, THEN initiate 1-AP-15, LOSS OF COMPONENT COOLING, while continuing with this procedure.
o e) GOTO Step 11.
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 12 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 10. RETURN TO PROCEDURE AND STEP IN EFFECT CAUTION: If RCS boiling is determined to exist, then non-essential personnel should be evacuated from the Containment.
- 11. INITIATE PERSONNEL PROTECTIVE ACTIONS:
o a) Record most recent time to boiling estimate from 1-GOP-13.0, ALTERNATE CORE COOLING METHODASSESSMEN~
- Time (minutes): _
o b) Evaluate need to implement EPIP-1.01, EMERGENCY MANAGER CONTROLLING PROCEDURE c) Monitor Containment Radiation:
o
- 1-RM-RMS-159 o
- 1-RM-RMS-160 12._ INITIATE ATTACHMENT 11, CONTAINMENT CLOSURE, WHILE CONTINUING WITH THIS PROCEDURE 13._ VERIFY 1-RC-L1-105, INDEPENDENT o Place the keylock switch for 1-RC-L1-1 05 in RCS LEVEL INDICATOR - ENABLE.
ENERGIZED
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 13 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 14. START AVAILABLE CONTAINMENT AIR RECIRC FANS USING 1-0P-21.1, CONTAINMENT VENTILATION NOTE: If RCPs are stopped, then Attachment 10, NATURAL CIRCULATION should be used to establish and maintain natural circulation.
- 15. MAINTAIN CORE COOLING USING FORCED CIRCULATION:
0 a) Verify at least one RCP - 0 a) GO TO Step 16.
RUNNING b) Stabilize RCS temperature by dumping steam using either of the following:
0
- Condenser Steam Dumps OR 0
0
- Condensate o d) GO TO Step 18
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 14 of 23 ACTIONI EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 16. CHECK IF THE RCS SHOULD BE COOLED WITH SPENT FUEL POOL COOLING:
o a) Verify Reactor Cavity - FLOODED o a) GO TO Step 17.
o b) Verify Spent Fuel Pit level - o b) Initiate O-AP-27, MALFUNCTION OF SPENT NORMAL FUEL PIT SYSTEM, AND GO TO Step 17.
o c) Initiate ATTACHMENT 9, COOLING THE RCS WITH SFP COOLERS o d) GO TO Step 18
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 15 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION:
- Personnel working in Containment should be warned before the RCS is refilled to avoid contamination of personnel near any RCS openings.
- Only borated water should be added to the RCS to maintain adequate shutdown margin.
- Differences exist in RCS levels between active and inactive cold and hot legs during reduced inventory operations. At saturated conditions, the hot and cold leg levels can differ by several feet.
NOTE: The alternate cooling method priority is obtained from 1-GOP-13.0, ALTERNATE CORE COOLING METHOD ASSESSMENT.
- 17. DETERMINE APPROPRIATE ALTERNATE CORE COOLING METHOD:
o
- Natural Circulation - Initiate ATTACHMENT10,NATURAL CIRCULATION, while continuing with this procedure OR o
- Reflux Boiling - Initiate ATTACHMENT 8, REFLUX BOILING, while continuing with this procedure OR (STEP 17 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 16 of 23 ACTIONI EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 17. DETERMINE APPROPRIATE ALTERNATE CORE COOLING METHOD: (Continued) o
- Hot Leg Injection Forced Feed and Spill - Initiate ATTACHMENT 5, HOT LEG INJECTION FORCED FEED AND SPILL, while continuing with this procedure OR o
- Cold Leg Injection Forced Feed and Spill - Initiate ATTACHMENT 6, COLD LEG INJECTION FORCED FEED AND SPILL, while continuing with this procedure OR o
- Gravity Feed and Spill - Initiate ATTACHMENT 4, GRAVITY FEED AND SPILL, while continuing with this procedure
(
( NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 17 of 23 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION: If the Containment has been previously closed out using ATIACHMENT 11, CONTAINMENT CLOSURE, then personnel should not re-enter without first contacting Health Physics.
- 18. CONTINUE ATIEMPTS TO RESTORE RHR SYSTEM:
a) Vent RHR System as necessary:
0 1) Maintain RCS level while venting RHR by increasing makeup flow to RCS 0 2) Locally vent RHR System b) Establish conditions to start RHR Pumps:
0 1) Verify RHR Pumps- o 1) GO TO Step 19.
SECURED 0 2) Check RCS Level - AT o 2) Increase RCS level to greater than LEAST +10 INCHES +10 inches above centerline.
ABOVE CENTERLINE o IF level cannot be increased to greater than
+10 inches above centerline, THEN GO TO 1-AP-17, SHUTDOWN LOCA.
o 3) Check RHR Pump - o 3) Try to get an RHR Pump available.
AVAILABLE (STEP 18 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 18 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 18. CONTINUE ATTEMPTS TO RESTORE RHR SYSTEM:
(Continued)
- 4) Check RHR Inlet Isolation 0 4) Manually open valves.
Valves - OPEN:
0
- 1-RH-MOV-1700 0
- 1-RH-MOV-1701
- 5) Check RHR Outlet Isolation 0 5) Manually open desired valve.
Valves, Disch to Cold Legs -
OPEN:
0
- 1-RH-MOV-1720A
("B" Cold Leg)
OR 0
- 1-RH-MOV-1720B
("C" Cold Leg) 0 6) Check 1-RH-HCV-1758 - 0 6) Manually close valve.
CLOSED 0 7) Check 1-RH-FCV-1605- 0 7) Manually close valve.
CLOSED
0 a) Restore Service Water using O-AP-12, LOSS OF SERVICE WATER 0 b) Restore CC System using 1-AP-15, LOSS OF COMPONENT COOLING
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 19 of 23 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 20. CHECK SERVICE WATER TO CC HEAT EXCHANGER - AVAILABLE:
0 a) Verify Service Water System - IN a) IF Service Water flow is NOT available, THEN SERVICE continue attempts to restore Service Water using the following while continuing with this procedure:
0
- 0-AP-12, LOSS OF SERVICE WATER 0
- 1-AP-15, LOSS OF COMPONENT COOLING 0 RETURN TO Step 18.
b) Verify Service Water Supply Valves b) Open Service Water Supply Valves to CC to CC System - OPEN System:
0
- 1-SW-MOV-108A 0
- 1-SW-MOV-108A 0
- 1-SW-MOV-108B 0
- 1-SW-MOV-108B 0 c) Locally check Service Water to CC 0 c) RETURN TO Step 18.
Heat Exchanger ~P - NORMAL
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 20 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
o a) Verify CC System - IN SERVICE o a) IF CC flow is NOT available, THEN continue attempts to restore CC using 1-AP-15, LOSS OF COMPONENT COOLING, while continuing with this procedure.
o RETURN TO Step 18.
b) Check CC flow to RHR Heat b) Do the following:
Exchangers - NORMAL:
- 1-CC-FI-132A Exchanger:
o
- 1-CC-FI-132B o
- 1-CC-TV-103A, A RHR Heat Exchanger Return Isolation o
- 1-CC-TV-103B, B RHR Heat Exchanger Return Isolation o
- 1-CC-MOV-100A, A CC Heat Exchanger Outlet Isolation o
- 1-CC-MOV-100B, B CC Heat Exchanger Outlet Isolation
o
- 1-CC-MOV-1 OOA for 1-CC-TV-1 03A o
- 1-CC-MOV-100B for 1-CC-TV-103B o 3) IF CC flow is restored, THEN GO TO Step 22.
(STEP 21 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 21 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
(Continued) o 4) IF CC is NOT restored, THEN continue attempts to restore CC using 1-AP-15, LOSS OF COMPONENT COOLING, while continuing with this procedure.
o 5) RETURN TO Step 18.
CAUTION:
- During RHR flow restoration, flow must start at a lower rate to limit the initial sudden cooldown and to minimize level loss caused by collapsing voids.
- If the RHR System was not satisfactorily vented, then entrained air can be swept from the system by raising flow to 3300 gpm. This method could cause water hammer or pump damage.
- 22. RESTORE RHR FLOW:
a) Close the following valves:
0
- 1-RH-HCV-1758 0
- 1-RH-FCV-1605 0 b) Start one RHR Pump 0 b) RETURN TO Step 18.
0 c) Maintain RCS level within acceptable region of Attachment 2 d) Restore RHR flow by repositioning the following RHR Control Valves:
0
- 1-RH-HCV-1758 0
- 1-RH-FCV-1605
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 22 of 23 ACTIONI EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 23. VERIFY RHR SYSTEM - NORMAL: Do the fol/owing:
D
D
- RHR Motor amps - STABLE + 10 inches above centerline by increasing D
- RCS temperature - STABLE charging flow.
- 2) Check RHR flow - less than or equal to design flow of ATIACHMENT 3:
D
- 1 RHR HX in use - Page 2 of 2 IF RHR flow is greater than the design flow rate of ATIACHMENT 3, THEN reduce flow to the design flow rate using:
D
- 1-RH-HCV-1758 D
- 1-RH-FCV-1605 D 3) Check RCS level - Greater than minimum for indicated flow of ATTACHMENT 2.
D IF RCS level is not greater than minimum for indicated flow of ATIACHMENT 2, THEN STOP the RHR Pumps and RETURN TO Step 18.
- 4) Send an Operator to locally check pump operation:
D
- RHR pump noise D
- RHR pump seals D
- RHR pump vibration D b) !E RHR System cannot be stabilized, THEN stop running RHR Pump AND RETURN TO Step 18.
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 23 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 25. CHECK IF RCS MAKEUP SHOULD BE REDUCED:
0 a) RCS Temperature - LESS THAN 0 a) Continue cooldown with RHR.
200°F 0 b) RCS Level - STABLE OR 0 b) GO TO Step 26.
INCREASING c) Check Low Head SI Pump 0 c) GO TO Step 25e.
Suctions From Containment Sump
- CLOSED:
0
- 1-SI-MOV-1860A 0
- 1-SI-MOV-1860B 0 d) Stop any running Low Head SI Pump.
0 e) Control RCS level using makeup and letdown as required
THAN 140 OF 0 RETURN TO Step 24.
- 27. RETURN TO PROCEDURE AND STEP IN EFFECT
- END-
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 1 REFERENCES REVISION PAGE 25 1 of 3
- Generic Letter 87-12, Loss of Decay Heat Removal
- Generic Letter 88-17, Loss of Decay Heat Removal
- NUREG-1410, Loss of Vital AC Power and the Residual Heat Removal System During Mid-Loop Operations at Vogtle Unit 1 On March 20, 1990
- Background information for WOG Abnormal Response Guideline ARG-1, LOSS OF RHR WHILE OPERATING AT MID-LOOP CONDITIONS, Rev 0, March 15, 1990.
- NE Technical Report 825, EVALUATION AND DEVELOPMENT OF SETPOINTS FOR ABNORMAL RESPONSE GUIDELINE ARG-1 LOSS OF RHR WHILE OPERATING AT MIDLOOP CONDITIONS NORTH ANNA POWER STATION UNITS 1 AND 2, February 1991
- NE Technical Report 865, ENSURING ADEQUATE DECAY HEAT REMOVAL WITH RCS LOOP STOP VALVES CLOSED, December 6, 1991
- NE Technical Report 865, Revision 2, BACKGROUND AND GUIDANCE FOR ENSURING ADEQUATE DECAY HEAT REMOVAL FOLLOWING LOSS OF RHR SURRY AND NORTH ANNA POWER STATIONS, April 1995
- NSA-92180, NE TECHNICAL REPORT 865, REVISION 1, SUPPLEMENTAL INFORMATION, October 9,1992
- NAF ET-95030, Updated Time to Boiling Curves, 2/28/95
- 11715-FM-94A, RHR
- 11715-FM-88A, Fuel Pit Cooling
- 11715-FM-93A, Reactor Coolant
- 11715-FM-96A and B, Safety Injection
- TRM 3.7.8
- OP 95-1148, Incorporate TR 865 Rev 2 into procedures
- CTS 02-89-1750-003, 007, 036
- CTS 02-95-0001-004, Revise procedures to place ultrasonic level in service
- CTS 02-95-0001-012, Revise procedures for new time to boil curves
- 0-GOP-13.3, ASSESSMENT OF MAINTENANCE ACTIVITIES FOR POTENTIAL LOSS OF REACTOR COOLANT INVENTORY
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 1 REFERENCES REVISION PAGE 25 2 of 3
- 0-MCM-1204-3, EMERGENCY INSTALLATION OF THE EQUIPMENT DOOR AND ESCAPE LOCK
- 0-MCM-1204-5, EMERGENCY INSTALLATION OF EQUIPMENT DOOR AND TEMPORARY PENETRATION PLATE
- 0-AP-10, LOSS OF ELECTRICAL POWER
- 0-AP-12, LOSS OF SERVICE WATER
- 1-AP-15, LOSS OF COMPONENT COOLING
- 1-AP-17, SHUTDOWN LOCA
- 1-AP-52, LOSS OF REFUELING CAVITY LEVEL DURING REFUELING
- 1-0P-5.2, REACTOR COOLANT PUMP STARTUP AND SHUTDOWN
- 0-OP-16.1, SPENT FUEL PIT COOLING AND PURIFICATION SYSTEM
- 1-0P-21.1, CONTAINMENT VENTILATION
- 1-0P-21.5, OPERATION OF AUXILIARY BUILDING IODINE FILTERS
- 1-GOP-13.0, ALTERNATE CORE COOLING METHOD ASSESSMENT
- EPIP-1.01, EMERGENCY MANAGER CONTROLLING PROCEDURE
- The following EOP references this procedure:
1-FR-C.3, RESPONSE TO SATURATED CORE COOLING
- DCP 01-140, Boron Concentration Increase in RWST, CCT, SFP, SIA's/NAPS/Unit 1 & 2, associated with Tech Spec Change Request 375
- Tech Spec Change 385, Revised Containment Analysis
- DCP 05-010, LTOPS Setpoint Change and Cooldown Administrative Limit
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 1 REFERENCES REVISION PAGE 25 3 of 3
- CA005647/CA008770, Gravity feed and spill options in AP-11 for CR007311 (Attachment 4, Step 3, 4 & notes)
- Plant Issue N-2003-2005-R8, Return Seal Injection MOVs To Normal Service (Attachment 7, Step 2.h, Rev 24)
- DCP 06-015, NRC GSI-191, RWST Level ESFAS Function to Support Containment Sump Modifications/North Anna/Unit 1 (Rev 25)
Unit 1 & EOP setpoint Document Number 2007-002 (Rev 25)
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 2 MINIMUM RCS LEVEL FOR INDICATED FLOW REVISION ~ PAGE 25 1 of 1 11 10 9
1/
!.I
- J 0 Acceptable !/
Ql 8 Region
> 1/
0
.0 g
ctI
, I. ...
Qi Ql I rV, ')
....J (J) 7 1/
0 a:
c
/' II ~naccePtable
- E V Region II V
6 I
II 1.... /
II 5
4 o 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 RHR Flow Rate (gpm) F18-2006-08-01
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 3 DETERMINING ACCEPTABLE RHR FLOW REDUCTIONS REVISION (DESIGN FLOW CALCULATED TO REMOVE DECAY HEAT) PAGE 25 1 of 2 9000 8000 7000 Unacceptable Region 6000
~
I a::
I a:: 5000
~
E 0-S RHR Pump Flow Limit I
~ \
o \
u::: 4000 a::
I a:: :\
\ \ _ fIr-1§ o 1\ >vv I-3000 \
\
I Acceptable to I
'n-, I Reduce Flow I 2000 i'r-..
Design Flow 1000 o
o 5 10 15 20 25 30 35 40 45 Time After Shutdown (Days) L13-2006-08-01
( NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 3 DETERMINING ACCEPTABLE RHR FLOW REDUCTIONS (DESIGN REVISION PAGE FLOW CALCULATED TO REMOVE DECAY HEAT) 25 2 of 2 4500 I I Unacceptable Region 1\
4000 1\
1\
3500 \
\ I Acceptable to
\ I Reduce Flow 3000 \
\
X I
c::
~ 2500 E "-
Cl.
.9 ,....
~ 2000 ....
u:::
c:: ........ 1""- _ _
I c::
r- ~r-_
1500 Design Flow 1000 500 o
o 5 10 15 20 25 30 35 40 45 Time After Shutdown (Days) L14-2006-08-01
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 4 GRAVITY FEED AND SPILL REVISION PAGE 25 1 of 2 CAUTION:
- This mode of heat removal and inventory makeup cannot be used when RCS pressures exceed about 25 psig.
- Personnel working in Containment should be warned before the RCS is refilled to avoid inadvertent contamination of personnel near any RCS opening.
- Depending on equipment and RCS conditions, boiling in the core may lead to PRZR surge line flooding and cause RVLlS and RCS Standpipe level indications to read higher than actual.
NOTE: If there are no cold leg openings, then cold leg injection is preferable. If there are cold leg openings, then hot leg injection should be used.
- 1. IF desired to conserve Containment Sump inventory for RCS recirculation, THEN place the following Containment Sump Pumps in OFF:
- 1-DA-P-4A
- 1-DA-P-4B
- 2. Using available plant equipment, align RWST water to the RCS using one of the following flowpaths:
- 1-SI-P-1 A, A Low-Head SI Pump, hot leg injection flow path
- 1-SI-P-1 B, SLow-Head SI Pump, hot leg injection flow path
- 1-SI-P-1A, A Low-Head SI Pump, cold leg injection flow path
- 1-SI-P-1 S, SLow-Head SI Pump, cold leg injection flow path
- Charging Pump(s) hot leg injection flow path
- Charging Pump(s) cold leg injection flow path
- Charging Pump(s) normal charging flow path
- Charging Pump(s) alternate charging flow path
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 4 GRAVITY FEED AND SPILL REVISION PAGE 25 2 of 2 NOTE: Using the PRZR Safety flowpath method of core cooling should suppress boiling for at least one hour following initiation of Gravity Feed and Spill.
- 3. Verify at least one PRZR Safety Valve is removed.
NOTE: Using the PRZR PORV flowpath for core cooling is NOT an effective method to suppress boiling, but may provide some decay heat removal as a final option.
- 4. IF at least one PRZR Safety Valve is NOT removed, THEN do the following:
- Open both PRZR PORVs
- Open both PRZR PORV Block Valves NOTE: If forced feed capability is restored, then ATIACHMENT 5 OR ATIACHMENT 6, should be used for core cooling.
- 5. Continue attempts to restore forced cooling to RCS.
- 6. Return to 1-AP-11, LOSS OF RHR, step in effect.
- END-
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 5 HOT LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 1 of 10 CAUTION:
- If the RCS is vented to the PRT, then PRT pressure indication should be monitored as an indication of RCS pressure. Changes in RCS pressure can result in Reactor Vessel water level changes that may not show on RCS standpipe level indicator 1-RC-L1-103.
- Depending on equipment and RCS conditions, boiling in the core may lead to PRZR surge line flooding and cause RVLlS and RCS Standpipe level indications to read higher than actual.
- If RWST level decreases to 15%, then the SI System should be aligned for recirculation using ATIACHMENT 7, ALIGNING THE SI SYSTEM FOR RECIRC, to provide long-term cooling.
- Charging and Low-Head Pumps taking suction from the RWST must be stopped when RWST level decreases to 8%. An alternate water source will be necessary in order to prevent loss of pump suction.
NOTE: Hot leg injection using this Attachment is the preferred method of RCS makeup for forced feed and spill operations. If hot leg injection is not available, then ATIACHMENT 6, COLD LEG INJECTION FORCED FEED AND SPILL should be used.
- 1. IF desired to conserve Containment Sump inventory for RCS recirculation, THEN place the following Containment Sump Pumps in OFF:
- 1-DA-P-4A
- 1-DA-P-4B
- 2. Verify a Charging Pump is available AND is specified for RCS makeup by the Alternate Core Cooling Method Assessment. !E a Charging Pump is NOT available, THEN GO TO Step 5.
- 3. Verify a Charging Pump flow path to the RCS hot legs is available. !E a Charging Pump flow path is NOT available, THEN GO TO Step 5.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 5 HOT LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 2 of 10
- 4. Align a Charging Pump to make up to the RCS as follows:
- a. Open Charging Pump Suction from RWST Isolation Valves:
- 1-CH-MOV-1115B
- 1-CH-MOV-1115D
- b. Close Charging Pump Suction from VCT Isolation Valves:
- 1-CH-MOV-1115C
- 1-CH-MOV-1115E
- c. Open 1-CH-MOV-1373, Charging Pump Recirc Header Isolation Valve.
- d. Open the Charging Pump Recirc Valves:
- 1-CH-MOV-1275A for 1-CH-P-1 A
- 1-CH-MOV-1275B for 1-CH-P-1 B
- 1-CH-MOV-1275C for 1-CH-P-1 C
- e. Start one Charging Pump.
- f. Close the Normal Charging Isolation Valves:
- 1-CH-MOV-1289A
- 1-CH-MOV-1289B
- g. Align one of the following hot leg injection flow paths as desired:
- 1-SI-MOV-1869B OR
- 1-SI-MOV-1869A (STEP 4 CONTINUED ON NEXT PAGE)
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 5 HOT LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 3 of 10
- h. Close the Charging Pump Recirc Valves:
- 1-CH-MOV-1275A for 1-CH-P-1 A
- 1-CH-MOV-1275B for 1-CH-P-1 B
- 1-CH-MOV-1275C for 1-CH-P-1 C
- i. Check the following to determine if charging flow is adequate:
- RCS level is stable or increasing
- RCS temperature is stable or decreasing
- j. i.E charging flow is adequate, THEN GO TO Step 6. IF charging flow is NOT adequate, THEN GO TO Step 5 to align a Low-Head SI Pump.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 5 HOT LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 4 of 10
- 1-SI-MOV-1862A OR
- 1-SI-MOV-1862B
- b. Close both of the Low-Head SI Pump Discharge Isolation Valves to the Cold Legs:
- 1-SI-MOV-1890C
- 1-SI-MOV-1890D
- c. Close both of the Low-Head SI Pump Discharge Isolation Valves:
- 1-SI-MOV-1864A
- 1-SI-MOV-1864B
- d. Start the desired Low-Head SI Pump:
- 1-SI-P-1A OR
- 1-SI-P-1B
- e. Open the desired Low-Head SI Pump Discharge Isolation Hot Leg Injection Valve:
- 1-SI-MOV-1890A OR
- 1-SI-MOV-1890B
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 5 HOT LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 5 of 10
- 6. Establish RCS bleed path using one of the following methods:
- Verify at least one PRZR Safety Valve is removed OR
- Use PRZR PORVs:
- 1) Verify power is available or restore power to PRZR PORV Block Valves.
- 2) Open both PRZR PORV Block Valves.
- 3) Open both PRZR PORVs.
- 7. Maintain RCS makeup and heat removal:
- a. Maintain Charging or Low-Head SI flow.
- b. Maintain RCS bleed path.
- 8. WHEN RHR OR other means of decay heat removal is established, THEN consult TSC or Plant Staff to determine if SI flow can be stopped. WHEN SI flow can be stopped, THEN continue with Step 9.
- 9. IF both of the following Low-Head SI Containment Suction Valves are closed, THEN GO TO Step 11. IF either valve is open, THEN GO TO Step 10:
- 1-SI-MOV-1860A
- 1-SI-MOV-1860B
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 5 HOT LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 6 of 10 CAUTION: To provide adequate Charging Pump cooling, Charging flow must be maintained at least 60 gpm. During SI Recirculation Mode the Charging Pump recircs must remain closed to prevent lifting the Seal Water return relief valve.
- 10. IF a Low Head SI Pump is aligned to supply Charging Pump suction in the SI Recirculation Mode, THEN have TSC or plant staff ensure the following is the desired Recovery method.
!E NOT the desired Recovery method, THEN GO TO Step 14:
- a. Verify 1-CH-HCV-1311, Auxiliary Spray Valve is closed.
- b. Open Normal Charging Line Isolation Valves:
- 1-CH-HCV-1310
- 1-CH-MOV-1289A
- 1-CH-MOV-1289B
- c. Open 1-CH-FCV-1122 in Manual to establish 60 gpm Charging flow.
- d. Close the following hot leg injection valves:
- 1-SI-MOV-1869B
- 1-SI-MOV-1869A
- e. Establish and maintain greater than 60 gpm Charging flow using 1-CH-FCV-1122 in MANUAL.
- f. Have TSC or plant staff provide guidance on realigning systems for recovery.
- g. GO TO Step 14.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 5 HOT LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 7 of 10
- 11. ISOLATE HOT LEG INJECTION:
- a. Do the following:
- 1) Open 1-CH-MOV-1373, Charging Pump Recirc Header Isolation Valve.
- 2) Open Charging Pump Recirc Valves:
- 1-CH-MOV-1275A for 1-CH-P-1 A
- 1-CH-MOV-1275B for 1-CH-P-1 B
- 1-CH-MOV-1275C for 1-CH-P-1C
- b. Close the following hot leg injection valves:
- 1-SI-MOV-1869B
- 1-SI-MOV-1869A
- 12. Establish normal Charging and Letdown:
- a. Put controller for 1-CH-FCV-1122, Normal Charging Flow Control Valve, in MANUAL and close.
- b. Verify 1-CH-HCV-1311, Auxiliary Spray Valve, is closed.
- c. Open Normal Charging Line Isolation Valves:
- 1-CH-HCV-1310
- 1-CH-MOV-1289A
- 1-CH-MOV-1289B
- d. Open 1-CH-FCV-1122, Normal Charging Flow Control Valve, to establish desired flow.
(STEP 12 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 5 HOT LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 8 of 10
- e. Establish Letdown:
- 1) Verify at least one CC Pump is running. IF NOT, THEN at least one CC Pump using 1-0P-51.1, COMPONENT COOLING SYSTEM OR 1-AP-15, LOSS OF COMPONENT COOLING.
- 2) Put 1-CH-PCV-1145 in MANUAL and open to 100%.
- 3) Open the following:
- 1-CH-TV-1204A
- 1-CH-TV-1204B
- 4) Place desired Letdown path in service:
- Do the following to establish Letdown from RCS:
- a. Open the following:
- 1-CH-LCV-1460A
- 1-CH-LCV-1460B
- b. Open at least one of the following Letdown Orifice Valves:
- 1-CH-HCV-1200A
- 1-CH-HCV-1200B
- 1-CH-HCV-1200C
- 5) Adjust 1-CH-PCV-1145 in MANUAL or AUTO to establish desired letdown pressure.
(STEP 12 CONTINUED ON NEXT PAGE)
( NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 5 HOT LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 9 of 10
- f. Check VCT makeup control system, as follows:
- 1) Verify one Boric Acid Transfer Pump is aligned to Unit 1 blender. IF NOT, THEN align one Boric Acid Transfer Pump using the applicable 0-OP-8 series procedure.
- 2) Verify at least one PG Pump is running. IF NOT, THEN start one PG Pump.
- 3) Set makeup concentration at greater than 2600 ppm, as follows:
- a. Set Boric Acid Controller to 8.25 (16.5 gpm)
- b. Set PG Controller to 4.25 (65 gpm)
- 4) Place Blender control in AUTOMATIC.
- g. Align Charging Pump suction to VCT, as follows:
- 1) Verify VCT level is greater than 22%. IF NOT, THEN, WHEN VCT level is greater than 42%, THEN do Step 12.g.2 below:
- 2) Do the following:
- a. Open Charging Pump Suction From VCT Isolation Valves:
- 1-CH-MOV-1115C
- 1-CH-MOV-1115E
- b. Close Charging Pump Suction From RWST Isolation Valves:
- 1-CH-MOV-1115B
- 1-CH-MOV-1115D
- 13. SECURING LOW-HEAD SI PUMP:
- a. Close Low-Head SI Pump Discharge to Hot Leg Injection Valves:
- 1-SI-MOV-1890A
- 1-SI-MOV-1890B
- b. Stop Low-Head SI Pump.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 5 HOT LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 10 of 10
- 14. Do the following:
- a. Continue alignment of Charging and Low-Head SI Systems as directed by the Station Emergency Manager.
- b. RETURN TO 1-AP-11, LOSS OF RHR, step in effect.
- END-
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 6 COLD LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 1 of 10 CAUTION:
- If the RCS is vented to the PRT, then PRT pressure indication should be monitored as an indication of RCS pressure. Changes in RCS pressure can result in Reactor Vessel water level changes that may not show on RCS standpipe level indicator 1-RC-L1-1 03.
- Depending on equipment and RCS conditions, boiling in the core may lead to PRZR surge line flooding and cause RVLlS and RCS Standpipe level indications to read higher than actual.
- If RWST level decreases to 15%, then the SI System should be aligned for recirculation using ATTACHMENT 7, ALIGNING THE SI SYSTEM FOR RECIRC, to provide long-term cooling.
- Charging and Low-Head Pumps taking suction from the RWST must be stopped when RWST level decreases to 8%. An alternate water source will be necessary in order to prevent loss of pump suction.
NOTE: Hot leg injection using ATTACHMENT 5, HOT LEG INJECTION FORCED FEED AND SPILL is the preferred method of RCS makeup for forced feed and spill operations. If hot leg injection is not available, then this Attachment should be used.
- 1. !E desired to conserve Containment Sump inventory for RCS recirculation, THEN place the following Containment Sump Pumps in OFF:
- 1-DA-P-4A
- 1-DA-P-4B
- 2. Verify a Charging Pump is available AND is specified for RCS makeup by the Alternate Core Cooling Method Assessment. IF a Charging Pump is NOT available, THEN GO TO Step 5.
- 3. Verify a Charging Pump flow path to the RCS cold legs is available. IF a Charging Pump flow path is NOT available, THEN GO TO Step 5.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 6 COLD LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 2 of 10
- 4. Align a Charging Pump to make up to the RCS as follows:
- a. Open Charging Pump Suction from RWST Isolation Valves:
- 1-CH-MOV-1115B
- 1-CH-MOV-1115D
- b. Close Charging Pump Suction from VCT Isolation Valves:
- 1-CH-MOV-1115C
- 1-CH-MOV-1115E
- c. Open 1-CH-MOV-1373, Charging Pump Recirc Header Isolation Valve.
- d. Open the Charging Pump Recirc Valves:
- 1-CH-MOV-1275A for 1-CH-P-1 A
- 1-CH-MOV-1275B for 1-CH-P-1 B
- 1-CH-MOV-1275C for 1-CH-P-1 C
- e. Start one Charging Pump.
- 1. Close the Normal Charging Isolation Valves:
- 1-CH-MOV-1289A
- 1-CH-MOV-1289B (STEP 4 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 6 COLD LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 3 of 10
- g. Align one of the following cold leg injection flow paths as desired:
- BIT injection flow path:
- a. Close BIT Recirc Valves:
- 1-SI-TV-1884A
- 1-SI-TV-1884B
- 1-SI-TV-1884C
- b. Open BIT Outlet Valves:
- 1-SI-MOV-1867C
- 1-SI-MOV-1867D
- c. Open BIT Inlet Valves:
- 1-SI-MOV-1867A
- 1-SI-MOV-1867B OR
- Open 1-SI-MOV-1836, BIT Bypass Valve.
- h. Close the Charging Pump Recirc Valves:
- 1-CH-MOV-1275A for 1-CH-P-1 A
- 1-CH-MOV-1275B for 1-CH-P-1 B
- 1-CH-MOV-1275C for 1-CH-P-1C
- i. Check the following to determine if charging flow is adequate:
- RCS level is stable or increasing
- RCS temperature is stable or decreasing
- j. IF charging flow is adequate, THEN GO TO Step 6. !E charging flow is NOT adequate, THEN GO TO Step 5 to align a Low-Head SI Pump.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 6 COLD LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 4 of 10
- 1-SI-MOV-1862A OR
- 1-SI-MOV-1862B
- b. Close both of the Low-Head SI Pump Discharge Isolation Valves to the Hot Legs:
- 1-SI-MOV-1890A
- 1-SI-MOV-1890B
- c. Open the desired Low-Head SI Pump Discharge Isolation Valve:
- 1-SI-MOV-1864A OR
- 1-SI-MOV-1864B
- d. Start the desired Low-Head SI Pump:
- 1-SI-P-1A OR
- 1-SI-P-1 B
- e. Open the desired Low-Head SI Pump Discharge Isolation Valve to the Cold Legs:
- 1-SI-MOV-1890C OR
- 1-SI-MOV-1890D
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 6 COLD LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 5 of 10
- 6. Establish RCS bleed path using one of the following methods:
- Verify at least one PRZR Safety Valve is removed OR
- Use PRZR PORVs:
- 1) Verify power is available or restore power to PRZR PORV Block Valves.
- 2) Open both PRZR PORV Block Valves.
- 3) Open both PRZR PORVs.
- 7. Maintain RCS makeup and heat removal:
- a. Maintain Charging or Low-Head SI flow.
- b. Maintain RCS bleed path.
- 8. WHEN RHR OR other means of decay heat removal is established, THEN consult TSC or Plant Staff to determine if SI flow can be stopped. WHEN SI flow can be stopped, THEN continue with Step 9.
- 9. !E both of the following Low-Head SI Containment Suction Valves are closed, THEN GO TO Step 11. IF either valve is open, THEN GO TO Step 10:
- 1-SI-MOV-1860A
- 1-SI-MOV-1860B
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 6 COLD LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 6 of 10 CAUTION: To provide adequate Charging Pump cooling, Charging flow must be maintained at least 60 gpm. During SI Recirculation Mode the Charging Pump recircs must remain closed to prevent lifting the Seal Water return relief valve.
- 10. IF a Low Head SI Pump is aligned to supply Charging Pump suction in the SI Recirculation Mode, THEN have TSC or plant staff ensure the following is the desired Recovery method.
IE NOT the desired Recovery method, THEN GO TO Step 14:
- a. Verify 1-CH-HCV-1311, Auxiliary Spray Valve is closed.
- b. Open Normal Charging Line Isolation Valves:
- 1-CH-HCV-1310
- 1-CH-MOV-1289A
- 1-CH-MOV-1289B
- c. Open 1-CH-FCV-1122 in Manual to establish 60 gpm Charging flow.
- d. Close BIT Inlet Isolation Valves:
- 1-SI-MOV-1867A
- 1-SI-MOV-1867B
- e. Close BIT Outlet Isolation Valves:
- 1-SI-MOV-1867C
- 1-SI-MOV-1867D
- f. IF 1-SI-MOV-1836 is open, THEN place control power on AND close.
- g. Establish and maintain greater than 60 gpm Charging flow using 1-CH-FCV-1122 in MANUAL.
- h. Have TSC or plant staff provide guidance on realigning systems for recovery.
- i. GO TO Step 14.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 6 COLD LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 7 of 10
- 11. ISOLATE BIT:
- a. Do the following:
- 1) Open 1-CH-MOV-1373, Charging Pump Recirc Header Isolation Valve.
- 2) Open Charging Pump Recirc Valves:
- 1-CH-MOV-1275A for 1-CH-P-1 A
- 1-CH-MOV-1275B for 1-CH-P-1 B
- 1-CH-MOV-1275C for 1-CH-P-1C
- b. Close BIT Inlet Isolation Valves:
- 1-SI-MOV-1867A
- 1-SI-MOV-1867B
- c. Close BIT Outlet Isolation Valves:
- 1-SI-MOV-1867C
- 1-SI-MOV-1867D
- d. IF 1-SI-MOV-1836 is open, THEN place control power on AND close.
- 12. Establish normal Charging and Letdown:
- a. Put controller for 1-CH-FCV-1122, Normal Charging Flow Control Valve, in MANUAL and close.
- b. Verify 1-CH-HCV-1311, Auxiliary Spray Valve, is closed.
- c. Open Normal Charging Line Isolation Valves:
- 1-CH-HCV-1310
- 1-CH-MOV-1289A
- 1-CH-MOV-1289B
- d. Open 1-CH-FCV-1122, Normal Charging Flow Control Valve, to establish desired flow.
(STEP 12 CONTINUED ON NEXT PAGE)
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 6 COLD LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 8 of 10
- e. Establish Letdown:
- 1) Verify at least one CC Pump is running.!E NOT, THEN start at least one CC Pump using 1-0P-51.1, COMPONENT COOLING SYSTEM OR 1-AP-15, LOSS OF COMPONENT COOLING.
- 2) Put 1-CH-PCV-1145 in MANUAL and open to 100%.
- 3) Open the following:
- 1-CH-TV-1204A
- 1-CH-TV-1204B
- 4) Place desired Letdown path in service:
- Do the following to establish Letdown from RCS:
- a. Open the following:
- 1-CH-LCV-1460A
- 1-CH-LCV-1460B
- b. Open at least one of the following Letdown Orifice Valves:
- 1-CH-HCV-1200A
- 1-CH-HCV-1200B
- 1-CH-HCV-1200C
- 5) Adjust 1-CH-PCV-1145 in MANUAL or AUTO to establish desired letdown pressure.
(STEP 12 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 6 COLD LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 9 of 10
- f. Check VCT makeup control system, as follows:
- 1) Verify one Boric Acid Transfer Pump is aligned to Unit 1 blender. IF NOT, THEN align one Boric Acid Transfer Pump using the applicable 0-OP-8 series procedure.
- 2) Verify at least one PG Pump is running. IF NOT, THEN start one PG Pump.
- 3) Set makeup concentration at greater than 2600 ppm, as follows:
- a. Set Boric Acid Controller to 8.25 (16.5 gpm)
- b. Set PG Controller to 4.25 (65 gpm)
- 4) Place Blender control in AUTOMATIC.
- g. Align Charging Pump suction to VCT, as follows:
- 1) Verify VCT level is greater than 22%. IF NOT, THEN, WHEN VCT level is greater than 42%, THEN do Step 12.g.2 below:
- 2) Do the following:
- a. Open Charging Pump Suction From VCT Isolation Valves:
- 1-CH-MOV-1115C
- 1-CH-MOV-1115E
- b. Close Charging Pump Suction From RWST Isolation Valves:
- 1-CH-MOV-1115B
- 1-CH-MOV-1115D
- 13. SECURING LOW-HEAD SI PUMP:
- a. Close Low-Head SI Pump Discharge to Cold Legs Valves:
- 1-SI-MOV-1864A
- 1-SI-MOV-1864B
- b. Stop Low-Head SI Pump.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 6 COLD LEG INJECTION FORCED FEED AND SPILL REVISION PAGE 25 10 of 10
- 14. Do the following:
- a. Continue alignment of Charging and Low-Head SI Systems as directed by the Station Emergency Manager.
- b. RETURN TO 1-AP-11, LOSS OF RHR, step in effect.
- END-
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 7 ALIGNING THE SI SYSTEM FOR RECIRC REVISION PAGE 25 1 of 5 CAUTION: To prevent possible radioactive release from the RWST, VCT level should be maintained greater than 12%.
- 1. IF Containment Sump level is greater than 8 ft 0 in, THEN GO TO Step 3.
- 2. IF sump level is less than 8 ft 0 in, THEN have TSC or Plant Staff evaluate the following for guidance to provide alternate water source(s):
o
- Charging Pump Cross-Connect using 0-AP-48, CHARGING PUMP CROSS CONNECT OR o
- Casing Cooling Tank injection OR o
- Make up with Unit 1 Blender using 1-0P-7.7, REFUELING WATER STORAGE TANK SYSTEM OPERATION OR o
- Make up with Unit 2 Blender using 1-0P-7.7, REFUELING WATER STORAGE TANK SYSTEM OPERATION OR o
- Make up from Boron Recovery Tanks using 0-OP-16.1 0, Makeup to Unit 1 RWST from the Boron Recovery Tanks OR o
- Make up from the Spent Fuel Pool by cross-connecting piping using the RP System
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 7 ALIGNING THE SI SYSTEM FOR RECIRC REVISION PAGE 25 2 of 5 CAUTION: During SI Recirculation Mode the Charging Pump recircs must remain closed to prevent lifting the Seal Water return relief valve.
NOTE: If an SI signal is present, then the SI System will automatically align for recirculation at an RWST level of 16%.
- 3. WHEN Containment Sump level is greater than 8 ft 0 in, THEN do the following:
- a. Verify a Low-Head SI Pump is running. IF no Low-Head SI Pump is running, THEN start one Low-Head SI Pump on recirc using 1-0P-7.1, RECIRC OF RWST USING LOW HEAD SAFETY INJECTION PUMPS.
- b. Reset both trains of SI if necessary.
- c. Stop all but one Charging Pump and place in PTL.
- d. Open 1-CH-MOV-1373, Charging Pump Recirc Header Isolation Valve.
- e. Open the Charging Pump Recirc Valves:
- 1-CH-MOV-1275A for 1-CH-P-1 A
- 1-CH-MOV-1275B for 1-CH-P-1B
- 1-CH-MOV-1275C for 1-CH-P-1 C
- f. Close the Normal Charging Isolation Valves:
- 1-CH-MOV-1289A
- 1-CH-MOV-1289B (STEP 3 CONTINUED ON NEXT PAGE)
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 7 ALIGNING THE SI SYSTEM FOR RECIRC REVISION PAGE 25 3 of 5
- g. Ensure one of the following Charging Pump flow paths is aligned to the RCS. IF NOT, THEN manually align one desired flowpath:
- BIT injection flow path:
- a. BIT Recirc Valves - CLOSED:
- 1-SI-TV-1884A
- 1-SI-TV-1884B
- 1-SI-TV-1884C
- b. At least one BIT Outlet Valve - OPEN
- 1-SI-MOV-1867C
- 1-SI-MOV-1867D
- c. At least one BIT Inlet Valve - OPEN
- 1-SI-MOV-1867A
- 1-SI-MOV-1867B OR
- 1-SI-MOV-1836, BIT Bypass Valve - OPEN OR
- One of the following hot leg injection Valves - OPEN:
- 1-SI-MOV-1869B OR
- 1-SI-MOV-1869A (STEP 3 CONTINUED ON NEXT PAGE)
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 7 ALIGNING THE SI SYSTEM FOR RECIRC REVISION PAGE 25 4 of 5 h) Do the following:
- 1) Close 1-CH-MOV-1370, RCP Seal Water Injection Isolation Valve.
- 2) IF 1-CH-MOV-1370 cannot be closed, THEN do the following:
- a. Close 1-CH-HCV-1186, RCPs Seal Water Flow Control.
- b. Locally close 1-CH-MOV-1370, as time permits.
- i. Close the Charging Pump Recirc Valves:
- 1-CH-MOV-1275A for 1-CH-P-1 A
- 1-CH-MOV-1275B for 1-CH-P-1 B
- 1-CH-MOV-1275C for 1-CH-P-1 C
- j. Open Low-Head SI Discharge To Charging Pumps Valves:
- 1-SI-MOV-1863A
- 1-SI-MOV-1863B
- k. Close Low-Head SI Recirc Valves:
- 1-SI-MOV-1885A
- 1-SI-MOV-1885B
- 1-SI-MOV-1885C
- 1-SI-MOV-1885D I. Open Low-Head SI Containment Suction Valve:
- 1-SI-MOV-1860A
- 1-SI-MOV-1860B (STEP 3 CONTINUED ON NEXT PAGE)
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 7 ALIGNING THE SI SYSTEM FOR RECIRC REVISION PAGE 25 5 of 5
- m. Verify closed or close Low-Head SI Suction Valve from RWST:
- 1-SI-MOV-1862A
- 1-SI-MOV-1862B
- n. Close Charging Pump Suction From RWST Isolation Valves:
- 1-CH-MOV-1115B
- 1-CH-MOV-1115D
- o. Close Charging Pump Suction From VCT Isolation Valves:
- 1-CH-MOV-1115C
- 1-CH-MOV-1115E
- 4. IF Recirc Spray is available and required to provide a heat sink, THEN obtain Station Emergency Manager direction to initiate Recirc Spray and do the following:
- a. Evaluate Unit 2 Service Water System operability.
- b. Manually place one Recirc Spray Heat Exchanger in service.
- c. Manually start the associated Recirc Spray Pump.
- d. IF a CDA occurs on Unit 2, THEN terminate Service Water to Unit 1 CC Heat Exchangers.
- 5. Continue alignment of Charging and Low-Head SI Systems as directed by the Station Emergency Manager.
- 6. RETURN TO 1-AP-11, LOSS OF RHR, step in effect.
- END-
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 8 REFLUX BOILING REVISION PAGE 25 1 of 2 CAUTION: If the RCS is vented to the PRT, then PRT pressure indication should be monitored as an indication of RCS pressure. Changes in RCS pressure can result in Reactor Vessel water level changes that may not show on RCS standpipe level indicator 1-RC-L1-1 03.
NOTE:
- When RHR is restored, then the TSC or Plant Staff should be consulted to determine if Reflux Boiling should be terminated.
- Stable Reflux Boiling can be maintained when RCS level is kept above the core and below the top of the hot leg piping. Reflux Boiling can be effective regardless of the initial RCS level.
- RCS temperature will increase to saturation during establishment of Reflux Boiling. RCS pressure could increase to a positive pressure of as much as 20-50 psig. These are expected and necessary conditions during Reflux Boiling.
- 1. Maintain SG narrow range levels between 23% and 75% using any of the following:
- Main Feedwater
- Condensate
- 2. Dump steam using either of the following:
- Fully open two Condenser Steam Dump Valves OR
- 3. Verify effective Reflux Boiling by monitoring the following:
- Core Exit TCs - STABLE
- RCS pressure - STABLE AND ABOVE ATMOSPHERIC PRESSURE
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 8 REFLUX BOILING REVISION PAGE 25 2 of 2 NOTE: RVLlS may not be an accurate indication of actual RCS level during Reflux Boiling, but RVLlS may be used to trend RCS level.
- 5. Monitor Core Exit TCs - STABLE
- 6. Maintain stable plant conditions.
- 7. GO TO 1-AP-11, LOSS OF RHR, step in effect.
- END-
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 9 COOLING THE RCS WITH SFP COOLERS REVISION PAGE 25 1 of 4
- 1. As required, fill the Reactor Cavity to normal refueling level (water level at Reactor Cavity Skimmers):
a) Align one Low-Head SI Pump to fill the Reactor Cavity:
- 1) 1-SI-P-1A:
- Open 1-SI-MOV-1862A, Low-Head SI Pump A Suction.
- Open 1-SI-MOV-1864A, Low-Head SI Pump A Discharge.
- 2) 1-SI-P-1B:
- Open 1-SI-MOV-1862B, Low-Head SI Pump B Suction.
- Open 1-SI-MOV-1864B, Low-Head SI Pump B Discharge.
b) Open either of the following Low-Head SI to Cold Leg MOVs:
- 1-MOV-SI-1890C OR
- 1-MOV-SI-1890D c) Start the Low-Head SI Pump that was aligned:
- 1-SI-P-1A OR
- 1-SI-P-1B
- 2. WHEN the Reactor Cavity is full, THEN do the following:
_ a) Stop the Low-Head SI Pump.
b) Close Low-Head SI Pump Discharge Valve:
- 1-SI-MOV-1864A for 1-SI-P-1 A
- 1-SI-MOV-1864B for 1-SI-P-1 B c) Open the SFP gate valve.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-11 9 COOLING THE RCS WITH SFP COOLERS REVISION PAGE 25 2 of 4
_ a) Stop running RP Pumps.
b) Align the RP valves in Containment as follows:
- 1) Close 1-RP-1, Reactor Cavity Drain/R.P. Suction Isolation Valve
- 2) Close 1-RP-3, 1A Reactor Cavity Skimmer To RP Pumps Isol Valve
- 3) Open 1-RP-28, Refuel Prfcn Filter to Reactor Cavity Isol Valve c) Close the following RP valves in the Auxiliary Building basement:
- 1-RP-80, Refuel Prfcn Filters to Spent Fuel Pit Isol Vv d) Close the following RP valves in Unit 2 Penetration Area:
- 1-RP-134, Refuel Purification Fltrs To Unit 2 RWST Isol Valve
- 1-RP-84, Refuel Purification Fltrs To Reac Cavity Isol Vv e) Align the following RP valves in Unit 1 Penetration Area:
- 1) Close 1-RP-24, Refuel Purification Fltrs To Unit 1 RWST Isol Vv
- 2) Open 1-RP-26, Refuel Purification Fltrs To Reac Cavity Isol Vv f) Open the following RP valves in the Auxiliary Building basement:
- 1-RP-79, Refuel Purification Fltrs Outlet Hdr Xconn Isol Vv
- 1-RP-78, Refuel Purification Fltrs Outlet Hdr Xconn Isol Vv
- 1-RP-30, Spent Fuel Pit Coolers to RP Pps Suct Hdr Isol Vv
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 9 COOLING THE RCS WITH SFP COOLERS REVISION PAGE 25 4 of 4
- 6. Open the valves for the desired RP Pump(s):
a) 1-RP-P-1 A:
- 1-RP-12, 1A Refueling Purification Pump Suction Isol Valve
- 1-RP-16, 1A Refueling Purification Pump Disch Isol Valve b) 1-RP-P-1B:
- 1-RP-32, Refuel Purification Pps Suct Hdr Xconn Isol Valve
- 1-RP-34, 1B Refueling Purification Pump Suction Isol Valve
- 1-RP-38, 1B Refueling Purification Pump Disch Isol Valve c) 1-RP-P-1 C:
- 1-RP-32, Refuel Purification Pps Suct Hdr Xconn Isol Valve
- 1-RP-33, Refuel Purification Pps Suct Hdr Xconn Isol Valve
- 1-RP-55, 1C Refuel Purification Pump Suction Isol Valve
- 1-RP-119, 1C Refuel Purification Pump Disch Isol Valve
- 7. Place the SFP Cooling System in service using 0-OP-16.1, SPENT FUEL PIT COOLING AND PURIFICATION SYSTEM.
- 8. Start the RP Pump(s) that were aligned.
- 9. Throttle the following valves as necessary to maintain RP filter differential pressures less than or equal to 45 psid:
o
- 1-RP-23, 1A Refueling Purification Filter Outlet Isol Valve o
- 1-RP-65, 1B Refueling Purification Filter Outlet Isol Valve
- 10. RETURN TO 1-AP-11, LOSS OF RHR, step in effect.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 10 NATURAL CIRCULATION REVISION PAGE 25 1 of 2 CAUTION: If the RCS is vented to the PRT, then PRT pressure indication should be monitored as an indication of RCS pressure. Changes in RCS pressure can result in Reactor Vessel water level changes that may not show on RCS standpipe level indicator 1-RC-L1-1 03.
NOTE:
- 1. Stablize RCS temperature by dumping steam, using either of the following:
- Condenser Steam Dumps OR
- 2. Maintain SG narrow range levels between 23% and 75% using any of the following:
- Main Feedwater
- Condensate
- 3. Verify Natural Circulation by monitoring the following:
- 4. IF Natural Circulation was NOT verified, THEN increase dumping steam.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 11 CONTAINMENT CLOSURE REVISION PAGE 25 1 of 2
- 1. Evacuate all personnel from Containment not required for closure:
_ a) Notify Health Physics that Containment evacuation is required.
_ b) Sound the Containment Evacuation alarm for about 15 seconds and make the following announcement:
"ATIENTION UNIT ONE CONTAINMENT, CONTAINMENT CLOSURE IS REQUIRED. ALL PERSONNEL NOT REQUIRED TO CLOSE CONTAINMENT EXIT CONTAINMENT IMMEDIATELY."
_ c) Sound the Containment Evacuation alarm for about 15 seconds and make the following announcement:
"ATIENTION UNIT ONE CONTAINMENT, CONTAINMENT CLOSURE IS REQUIRED. ALL PERSONNEL NOT REQUIRED TO CLOSE CONTAINMENT EXIT CONTAINMENT IMMEDIATELY."
- 2. Verify or place Unit 1 Containment Purge Exhaust through the Iodine filters using 0-OP-21.5, OPERATION OF AUXILIARY BUILDING IODINE FILTERS.
- 3. IF RCS level is greater than 42 inches above centerline, THEN do the following:
_ a) Verify the Temporary Penetration Plate is installed. IF NOT, THEN install the Equipment Door and Temporary Penetration Plate using 0-MCM-1204-05, EMERGENCY INSTALLATION OF EQUIPMENT DOOR AND TEMPORARY PENETRATION PLATE.
_ b) Close at least one door on the Personnel Hatch.
_ c) GO TO Step 5.
NOTE: The trolley hoist must remain attached to the hatch without any change in load distribution on the hoist.
- 4. IF RCS level is 42 inches or less above centerline, THEN do the following:
_ a) Verify the Equipment Hatch is installed. IF NOT, THEN install the Equipment Hatch with at least 10 bolts torqued to 280 ft-Ib using 0-MCM-1204-03, EMERGENCY INSTALLATION OF THE EQUIPMENT DOOR AND ESCAPE LOCK.
b) Close at least one door on each hatch:
- Personnel Hatch
- Equipment Hatch
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-11 11 CONTAINMENT CLOSURE REVISION PAGE 25 2 of 2
- 5. IF any other penetration is open for maintenance OR Testing, THEN initiate the required contingency actions.
- 6. Place Containment ventilation and Containment cooling in service using 1-0P-21.1, CONTAINMENT VENTILATION.
- 7. Verify Containment evacuation is complete.
- END-
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS The Unit is in MODE 3 at normal operating temperature and pressure with preparations in progress for taking the Reactor Critical.
Pressurizer Pressure Controls are in automatic and all PORVs and Block valves are OPERABLE 1-PT-44.7, PORV Block Valves has been initiated.
The operator has stopped the procedure at Step 6.11 due to 1-RC-MOV-1535 indicating mid-position after taking the valve to close.
INITIATING CUE You are requested to determine the following:
- 1) Technical Specification action requirements if applicable.
- 2) Whether or not to continue procedure performance (test 1-RC-MOV-1536).
- 3) If Reactor may be taken critical (SRO ONLY).
02/25/08 Page: 1 of 7
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM NEW Apply Tech Specs.
TASK STANDARDS Applicable Tech Spec Actions are identified, including action that prohibits Mode change (SRO ONLY), and determination is made not to continue testing on opposite train equipment.
KIA
REFERENCE:
G-2.2.40 (3.4/4.7)
ALTERNATE PATH:
NIA TASK COMPLETION TIMES Validation Time = 20 minutes Start Time = - - -
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ ] SATISFACTORY [ ] UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature 1 Date EVALUATOR'S COMMENTS 02/25/08 Page: 2 of 7
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM NEW READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES The trainee has completed the applicable course knowledge training at the Reactor Operator level or Senior Reactor Operator level as applicable.
INITIAL CONDITIONS The Unit is in MODE 3 at normal operating temperature and pressure with preparations in progress for taking the Reactor Critical.
Pressurizer Pressure Controls are in automatic and all PORVs and Block valves are OPERABLE.
1-PT-44.7, PORV Block Valves has been initiated.
The operator has stopped the procedure at Step 6.11 due to 1-RC-MOV-1535 indicating mid-position after taking the valve to close.
02/25/08 Page: 3 of 7
INITIATING CUE You are requested to determine the following:
- 1) Technical Specification action requirements if applicable.
- 2) Whether or not to continue with procedure performance (test 1-RC-MOV-1536).
- 3) If Reactor may be taken critical (SRO ONLY).
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT Provide trainee blank copy of current revision of 1-PT-44.7 for reference.
PERFORMANCE STEPS START TIME 02/25/08 Page: 4 of 7
Recognizes condition is entry condition for Technical Procedure Step _ _
S ecifications.
ISAT [1 UNSAT [ 1 Standards Technical Specifications require BOTH PORV Block Valves operable in Modes 1 thru 3. Improper indication/operation renders 1-RC-MOV-1535 inoperable.
I Notes/Comments
~I Review Tech Specs and determine TS 3.4.11.0. applies. I Procedure Step I_C_r"..:.::.lti..:.::.c..:.;:,al:-S::-t:-e.L-p 1 SAT [1 UNSAT [ 1 Standards Determines the following actions apply:
- 1) Place associated PORV (1-RC-PCV-1456) in manual control -7 1 hr.
and
- 2) Restore Block Valve to OPERABLE status -7 72 hrs.
rates/comments 02/25/08 Page: 5 of 7
Determines procedure performance should NOT be continued on Procedure Step _ _
1-RC-MOV-1536.
ICritical Step ISAT [] UNSAT []
Standards Determination not to continue with testing on the opposite train based on procedure Step 4.5 "When one train of a redundantea ~elated system is inoperable, the valve on the remaining train ould t be exercised since its failure would cause a total loss of s nction."
Notes/Comments If the candidate is an RO candidate then the Evaluation ends at this point.
Determines Reactor should not be taken critical based on Procedure Step _ _
information rovided.
ICritical Step ISAT [] UNSAT [ ]
Standards Determines based on review of Tech Spec LCO 3.0.4 that Reactor should NOT be taken critical at this time since the mode change is prohibited by Tech Specs.
Notes/Comments This Step is applicable to, and thus ONLY critical for the ISRO or USRO candidates.
>>>>> END OF EVALUATION <<<<<
STOP TIME 02/25/08 Page: 6 of?
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
None 02/25/08 Page: 7 of 7
Procedure: 1-PT-44.7 Rev: 024 PAR: 0
Title:
PORV BLOCK VALVES Effective Date: 04/13/2006 Station: North Anna CONTINUOUS USE
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS Unit 1 is performing a scheduled Refueling Outage.
Two Engineering Department personnel will be working adjacent to the wall between 1B & 1C Recirc Spray Heat Exchangers.
The Engineering Department personnel estimate it will take them two (2) trips, each trip lasting 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to complete their required tasks.
HP estimates that 2 additional personnel, working in a 90 mrem/hr field, can install temporary shielding in 1.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. The shielding will reduce general area does rates where the Engineering Department personnel will be working by a factor of 2. Removal of the temporary shielding will take the 2 additional personnel 45 minutes, again working in a 90 mrem/hr field.
INITIATING CUE You are to review the radiological conditions for the area. Using the survey map and dose rate estimate provided, determine the following:
- 1. Total dose WITHOUT the use of temporary shielding.
- 2. Total dose WITH the use of temporary shielding.
- 3. Select which method should be used WITH temporary shielding or WITHOUT temporary shielding based on the Stations ALARA Program.
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM Use a survey map and dose rate estimate.
TASK STANDARDS Using a radiation survey map and dose rate estimate provided, the examinee will determine:
- Total dose WITHOUT the use of temporary shielding.
- Total dose WITH the use of temporary shielding.
- Which method should be used, WITH temporary shielding or WITHOUT temporary shielding, based on the Stations ALARA Program.
KIA
REFERENCE:
G2.3.12 (3.2/3.7)
G2.3.14 (3.4/3.8)
ALTERNATE PATH:
N/A TASK COMPLETION TIMES Validation Time = 20 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ 1SATISFACTORY [ 1UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature /
Date EVALUATOR'S COMMENTS Page 2 of7
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
INITIAL CONDITIONS Unit 1 is performing a scheduled Refueling Outage.
Two Engineering Department personnel will be working adjacent to the wall between 1B & 1C Recirc Spray Heat Exchangers.
The Engineering Department personnel estimate it will take them two (2) trips, each trip lasting 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to complete their required tasks.
HP estimates that 2 additional personnel, working in a 90 mrem/hr field, can install temporary shielding in 1.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. The shielding will reduce general area does rates where the Engineering Department personnel will be working by a factor of 2. Removal of the temporary shielding will take the 2 additional personnel 45 minutes, again working in a 90 mrem/hr field.
Page 3 of7
INITIATING CUE The task you are to perform is:
Review the radiological conditions for the area. Using the survey map and dose rate estimate provided, determine the following:
- 1) Total dose WITHOUT the use of temporary shielding.
- 2) Total dose WITH the use of temporary shielding.
- 3) Select which method should be used, WITH temporary shielding or WITHOUT temporary shielding based on the Stations ALARA Program.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT Survey map of work area Calculator PERFORMANCE STEPS START TIME
( Page 4 of7
Note: Candidate may perform calculations in any order.
o:==J Determine total dose WITHOUT the use of temporary shielding IProcedure Step I_C_r_iti_c_a_1S_t_e-'-p 1 SAT [] UNSAT []
Standards Operator determines total dose of 600mrem without the use of temporary shielding 2 trips X 3hrs/trip X 2 persons = 12 person-hrs 12 hrs X 50mr/Hr* = 600mrem total dose
- data obtained from survey map Notes/Comments: The dose rate on the survey map is used for estimating dose to ENG Department personnel.
~ Determine total dose WITH the use of temporary shielding I Procedure Step I_C_ri_ti_c_al_S_t_e......
p 1 SAT [] UNSAT []
Standards Operator determines total dose of 660mrem with the use of temporary shielding 2 persons X 2 hrs X 90mr/hr =360mr to install and remove temporary shielding The dose to engineering dept personnel now becomes:
~ 2 trips X 3hrs/trip X 2 persons = 12 person-hrs 12 hrs X 25mr/Hr** = 300mrem So total dose for the job now becomes 360mr + 300mr = 660mrem total dose
_ _ _----=1N=ot=e=s/=c=o=m=m=e=n=t=S============================;-1 I
Page 5 of7
3 Select which method should be used WITH temporary shielding or Procedure Step WITHOUT temporary shielding based on the Stations ALARA Program ICritical Step ISAT [] UNSAT [1 Standards Operator determines that work should be done WITHOUT temporary shielding since although the dose to the individuals doing the work (Eng. Dept. personnel) is substantially lower, the total dose to do the job with shielding is hiqher than without.
INotes/Comments: I
'--- 1
>>>>> END OF EVALUATION <<<<<
STOP TIME Page 6 of7
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
JOB PERFORMANCE MEASURE CHECKLIST Survey Map (May supply more than one map so that applicant has to choose correct one)
Calculator Page 7 of7
Unit 1 Containment, 241 Elevation 1 Pressurizer Relief Tank Cube LDWA - Low Dose Waiting Areas AMFrisking
- Continuous StationAir Monitor LHRA - Locked High Radiation Area HPA - Hot Particle Area CA - Contaminated Area ft;RCAB _Radiological Control Area Boundary HRA - High Radiation Area RA - Radiation Area RM Air~orn~ RaMdiOta'?l!~1 ARA- -Radioactive J a ena'\Sj Area NDCR _Neutron Dose Calculation Reauired o Gen. Area; 0 Contact; LCK Locked Gate; - Barrier All gamma and/or neutron readings in mRem/hr unless noted.
~~ Station
~,Dominion Administrative Procedure
Title:
Station ALARA Program Process I Program Owner: Manager Radiological Protection and Chemistry Procedure Number Revision Number Effective Date VPAP-2102 12 On File Revision Summary Revised to updated terminology throughout procedure to reflect implementation HIS-20 which replaced PREMS. Although changes are numerous, instructions did not change, only terminology was updated (including units (e.g., one man-rem became 1,000 person-mrem which is equivalent)).
Details:
- Updated terminology throughout procedure to reflect implementation of Canberra Health Physics Information System (HIS-20) which replaced Personnel Radiation Exposure Management System (PREMS). Examples include:
- Changed man-hours to person-hours.
- Changed man-rem to person-mrem.
- Changed estimate (or projected) to budget (or budgeted) (e.g., changed man-rem estimate to person-mrem budget), where appropriate.
- Deleted Personnel Radiation Exposure Management System and/or PREMS.
- Changed ALARA Hold to lock out.
- Revised attached forms, form instructions, and flow charts to reflect HIS-20 terminology:
- Revised Pre-Job and Post-Job ALARA Review Flow Chart (Attachment 1).
- Revised ALARA Evaluation Log (Attachment 3).
- Revised Multiple RWP Tracking (Attachment 4).
- Revised Pre-Job ALARA Worksheet (Attachment 5).
- Revised ALARA Action Plan (Attachment 6).
- Revised ALARA Action Plan Instructions (Attachment 7).
- Revised RWP/Task ALARA Re-Evaluation (Attachment 9).
- Revised Post-Job Review (Attachment 10).
- Revised ALARA Suggestion (Attachment 11).
Revision Summary continued on page 2 Approvals on File
DOMINION VPAP-2102 REVISION 12 PAGE 2 OF 66 Revision Summary continued
- Revised attached forms, form instructions, and flow charts to reflect HIS-20 terminology:
(continued)
- Revised ALARA Goals Flow Chart (Attachment 13).
- RevisedStation ALARA Goals (Attachment 14).
- Revised Department ALARA Goals (Attachment 15).
- Revised Department ALARA Goal Review (Attachment 16).
- Revised ALARA Goal Status Report (Attachment 17).
- Revised ALARA Goal Variance Report (Attachment 18).
- Revised Work in Progress ALARA Review (Attachment 20).
DOMINION VPAP-2102 REVISION 12 PAGE 3 OF 66 TABLE OF CONTENTS Section Page 1.0 PURPOSE 6 2.0 SCOPE 6
3.0 REFERENCES
/COMMITMENT DOCUMENTS 6 4.0 DEFINITIONS 7 5.0 RESPONSIBILITIES 8 6.0 INSTRUCTIONS 13 6.1 ALARA Organization and Administration 13 6.1.1 ALARA Coordinators 13 6.1.2 Station ALARA Committee (SAC) 14 6.2 Exposure Reduction Plans 15 6.3 Pre-Job Planning and Review 15 6.3.1 TEDE ALARA Evaluation 16 6.3.2 Pre-Job ALARA Review Initiation 18 6.3.3 Pre-Job ALARA Meeting 19 6.3.4 ALARA Action Plan 20 6.3.5 Less Than 5,000 Person-mrem 21 6.3.6 5,000 Person-mrem or More 21 6.4 Monitoring Job Performance 22 6.4.1 Exposure Tracking 22 6.4.2 Lock Out Notification 23 6.4.3 RWP/Task ALARA Re-evaluation Meetings 24 6.4.4 Work in Progress Reviews (WIPR) 25 6.5 Post-Job and Process Reviews 25 6.5.1 Post-Job Reviews 25 6.5.2 Job History Files 27
DOMINION VPAP-2102 REVISION 12 PAGE 4 OF 66 TABLE OF CONTENTS (continued)
Section Page 6.5.3 Process Review 27 6.6 Temporary Shielding 27 6.7 ALARA Suggestions 28 6.7.1 Submitting an ALARA Suggestion 28 6.7.2 Evaluating ALARA Suggestions 28 6.7.3 Review and Approval of ALARA Suggestions 29 6.7.4 ALARA Suggestion Awards 30 6.8 Design Change Package ALARA Review 31 6.9 Determining and Tracking ALARA Goals 31 6.9.1 ALARA Goals 31 6.9.2 Establishing Station ALARA Goals 31 6.9.3 Establishing Department ALARA Goals 33 6.9.4 Department/Organization Review of Proposed Person-mrem Goals 34 6.9.5 ALARA Goal Approval 35 6.9.6 Monitoring ALARA Goal Perfonnance 36 7.0 RECORDS 38
DOMINION VPAP-2102 REVISION 12 PAGE 5 OF 66 TABLE OF CONTENTS (continued)
Section Page ATTACHMENTS 1 Pre-Job and Post-Job ALARA Review Flow Chart 40 2 TEDE ALARA Evaluation - 720134(Jan 2001) 42 3 ALARA Evaluation Log - 727263(June 2006) 44 4 Multiple RWP Tracking - 729122(June 2006) 45 5 Pre-Job ALARA Worksheet -728680(June 2006) 46 6 ALARA Action Plan -725462(June 2006) 48 7 ALARA Action Plan Instructions 49 8 Station ALARA Committee Pre-Job ALARA Review -730285(June 2006) 50 9 RWP ALARA Re-Evaluation -727966(June 2006) 51 10 Post-Job Review - 728936(June 2006) 52 11 ALARA Suggestion -728518(June 2006) 55 12 ALARA Suggestion Log -728579(Jan 2001) 56 13 ALARA Goals Flow Chart 57 14 Station ALARA Goals - 727265(June 2006) 59 15 Department ALARA Goals - 727264(June 2006) 60 16 Department ALARA Goal Review -727266(June 2006) 61 17 ALARA Goal Status Report - 727268(June 2006) 62 18 ALARA Goal Variance Report -727761(June 2006) 63 19 Department ALARA Exposure Reduction Plan -722528(Jan 2001) 64 20 Work in Progress ALARA Review -722541(June 2006) 65
DOMINION VPAP-2102 REVISION 12 PAGE 6 OF 66 1.0 PURPOSE This procedure establishes the requirements and responsibilities for the ALARA Program. The objective of the ALARA program is to ensure that occupational radiation exposure, both individually and collectively, is maintained "As Low As Reasonably Achievable" (ALARA).
2.0 SCOPE The ALARA Program is applicable to all Station activities that involve exposure of individuals to ionizing radiation.
3.0 REFERENCES
/COMMITMENT DOCUMENTS 3.1 References 3.1.1 10 CFR 20, Standards for Protection Against Radiation 3.1.2 Regulatory Guide 8.8, Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable (Rev. 03, June 1978) 3.1.3 Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable (Rev. 01.-R, May 1977) 3.1.4 Regulatory Guide 8.15, Acceptable Programs for Respiratory Protection, Revision 1, October 1999 3.1.5 NUREG CR-0041 Revision 1, Manual of Respiratory Protection Against Airborne Radioactive Material 3.1.6 NUREG/CR-0446, Determining Effectiveness of ALARA Design and Operational Facilities 3.1.7 NUREG/CR-4254, Occupational Dose Reduction and ALARA at Nuclear Power Plants: Study on High-Dose Jobs, Radwaste Handling and ALARA Incentives 3.1.8 NUREG/CR-4373, Compendium of Cost Effectiveness Evaluation of Modifications for Dose Reduction at Nuclear Power Plants 3.1.9 INPO 91-014, Revision 1, Guidelines for Radiological Protection at Nuclear Power Stations 3.1.10 INPO Radiological Experience Notebook, ALARA Planning for Station Work, 82-001-0EN-08A 3.1.11 INPO Good Practice RP-601, Use of Goals in Achieving Reductions In Personnel Radiation Exposures, April1984 3.1.12 INPO Good Practice OA-103, Management Objective Program, September 1986 3.1.13 Surry and North Anna Technical Specifications 3.1.14 STD-GN-0001, Instructions for DCP Preparation
DOMINION VPAP-2l02 REVISION 12 PAGE 7 OF 66 3.1.15 STD-GN-0019, Engineering ALARA Design Guide 3.1.16 Atomic Industrial Forum, Compendium of Design Features to Reduce Occupational Radiation Exposure at Nuclear Plants, April 1981 3.1.17 VPAP-2l0l, Radiation Protection Program 3.1.18 VPAP-2l05, Temporary Shielding Program 3.1.19 C-HP-l042.2l0, Respiratory Hazards Evaluation And Respiratory Protection Selection 3.1.20 C-HP-l081.030, Radiation Work Permits: Extending, Revising, and Terminating 3.2 Commitment Documents 3.2.1 Plant Issues (Deviations) N-200l-3ll2-R4/S-200l-2920-R4, INPO SOER 01-1, Unplanned Radiation Exposures 4.0 DEFINITIONS 4.1 ALARA The operating principle of radiological protection that states that measures are to be implemented to keep doses and intakes "as low as reasonably achievable". Optimization is a key element of implementing ALARA.
4.2 Job/Activity A collection of sub-tasks that are required to be performed to complete a major task.
4.3 Person-Hour Estimate An estimate of the collective time spent performing the work (this procedure is concerned with the person-hours in an Radiological Control Area (RCA)).
4.4 Person-mrem Budget A calculated pre-work estimate of the collective exposure for performing a task in an RCA, based on job histories and realistic expectations.
4.5 Process A collection of similar tasks of a repetitive nature that, when taken as a whole, represent a significant person-mrem expenditure, but when examined individually, may not require extensive pre-job planning.
DOMINION VPAP-2102 REVISION 12 PAGE 8 OF 66 5.0 RESPONSIBILITIES 5.1 Site Vice President The Site Vice President is responsible for all aspects of Station operation, including the Radiation Protection Program. Site Vice President ALARA responsibilities include:
5.1.1 Ensuring the ALARA Program is implemented, that Station personnel support the ALARA Program, and that adequate resources are provided to achieve ALARA objectives and goals.
5.1.2 Reviewing and approving pre-job ALARA evaluations of jobs budgeted to expend 15,000 person-mrem or greater.
5.1.3 Reviewing and approving person-mrem goals approved by the Station ALARA Committee (SAC). Documenting approval or reasons for rejection of goals and suggestions.
5.1.4 Reviewing and approving Station and departmental ALARA goals.
5.2 Director Nuclear Station Operations and Maintenance The Director Nuclear Station Operations and Maintenance is responsible for acting as Chairman of SAC.
5.3 Station Supervision Station Supervision is responsible for:
5.3.1 Providing department personnel to serve as Department ALARA Coordinators as directed by this procedure.
5.3.2 Ensuring that department personnel comply with ALARA Program procedures and requirements, including performing pre-job ALARA planning and post-job ALARA reviews.
5.3.3 Integrating the ALARA concept into appropriate department procedures during preparation, routine review, and revision.
5.3.4 Routinely reviewing distributed reports to assess the exposure status of supervised personnel (including supplemental personnel).
5.3.5 Routinely reviewing exposure status of personnel within their department to ensure that exposures are distributed as evenly as practicable within different crafts.
DOMINION VPAP-2l02 REVISION 12 PAGE 9 OF 66 5.3.6 Routinely monitoring exposure accumulation of RWPs being supervised.
5.3.7 Contacting and meeting with Station ALARA Coordinator, as requested, when RWPs being used have or are expected to exceed 125 percent of budgeted dose.
5.3.8 Routinely reviewing progress toward meeting established ALARA goals and ensuring deviations from goals are investigated.
5.3.9 Developing and implementing exposure reduction plans if the annual budgeted dose for the department is greater than five percent of the total Station exposure, or as directed by the SAC Chairman.
5.4 Manager Radiological Protection and Chemistry The Manager Radiological Protection and Chemistry is responsible for:
5.4.1 Assigning an individual to serve as the Station ALARA Coordinator.
5.4.2 Supporting the ALARA Program as a part of the Radiation Protection Program.
5.4.3 Participating in design reviews for facilities and equipment that can potentially affect radiation exposures.
5.5 Manager Nuclear Training The Manager Nuclear Training is responsible for providing personnel training as necessary on ALARA Program requirements.
5.6 Manager Nuclear Site Engineering The Manager Nuclear Site Engineering is responsible for:
5.6.1 Performing engineering evaluations when required for installation of temporary shielding.
5.6.2 Providing technical support required to evaluate Design Change Packages (DCPs).
5.6.3 Providing engineering support for hot spot flushing or removal.
5.7 Manager Nuclear Outage and Planning The Manager Nuclear Outage and Planning is responsible for:
5.7.1 Ensuring jobs are planned with adequate time allotted to implement ALARA requirements.
DOMINION VPAP-2l02 REVISION 12 PAGE 10 OF 66 5.7.2 Including ALARA concerns in the planning stages of tasks.
5.7.3 Ensuring proper departmental interface and support from other disciplines are planned.
5.7.4 Providing a properly completed Radiation Work Permit (RWP) Request and component work history for ALARA job planning.
5.7.5 Providing the Station ALARA Committee and ALARA Coordinators with current outage planning schedules and job timetables.
5.7.6 Reviewing Station outage forecasts (e.g., five-year) and preparing dose/ALARA impact.
5.8 Station ALARA Coordinator NOTE: Where referred to in this and other administrative procedures, the "Station ALARA Staff' means the Station ALARA Coordinator and those personnel periodically assigned to the Station ALARA Coordinator for the purpose of performing evaluations, reviews, and other tasks required to ensure proper conduct of the Station ALARA Program.
The Station ALARA Coordinator is responsible for:
5.8.1 Administration of the Station ALARA Program and Station ALARA Staff.
- a. Assisting in preparing or reviewing pre-job ALARA packages ofjobs with exposure estimates greater than 1,000 person-mrem.
- b. Assisting in optimization of dose reduction techniques for planned work.
5.8.2 Serving as SAC secretary. As the SAC Secretary:
- a. As practical, preparing meeting agendas and material requiring SAC action prior to the meeting.
- b. Preparing meeting minutes for SAC proceedings.
- c. Distributing copies of SAC meeting minutes to Site Vice President and all SAC members.
- d. Ensuring SAC reviews and approves pre-job ALARA review packages for jobs budgeted to expend 5,000 person-mrem or more.
DOMINION VPAP-2102 REVISION 12 PAGE 11 OF 66
- e. Ensuring SAC reviews post-job reviews and critiques for jobs with collective doses 10,000 person-mrem or more.
- f. Evaluating and recommending, as appropriate, ALARA suggestions for review by the SAC.
- g. Coordinating preparation of Station ALARA goals.
- h. Presenting to SAC an annual Station ALARA Program status report and applicable outage reports.
5.9 Department ALARA Coordinators Department ALARA Coordinators are responsible for:
5.9.1 Serving as department ALARA representative and interfacing with the Station ALARA Coordinator, cognizant Job Supervisor/Foreman, RWP requestor, Health Physics, and other Department ALARA Coordinators in matters related to ALARA.
5.9.2 Serving as SAC member. Responsibilities as a SAC member include:
- a. Reviewing monthly department exposure reports that detail actual exposure versus department goals and summarizing department exposure by major jobs or tasks.
- b. Providing a report for SAC review whenever quarterly department cumulative exposure exceeds department goals by more than 15 percent and the total quarterly department exposure is 100 person-mrem or more.
- c. Evaluating applicable suggestions submitted as part of the ALARA Suggestion Program.
- d. Coordinating further evaluation and implementation of departmental assigned, approved ALARA suggestions.
- e. Reporting status ofdepartment-related ALARA suggestions, ifrequired by the SAC Chairman.
5.9.3 Assisting in development and tracking of:
- Department ALARA goals
- Exposure reduction plans, if required by the SAC Chairman 5.9.4 Distributing to department supervision exposure status reports, as required.
DOMINION VPAP-2102 REVISION 12 PAGE 12 OF 66 5.10 Station ALARA Committee The Station ALARA Committee is responsible for:
5.10.1 Evaluating pre-job ALARA reviews and ALARA Action Plans for jobs budgeted to expend a collective dose of5,000 person-mrem or more and other jobs as requested by the Station ALARA Coordinator.
5.10.2 Reviewing post-job evaluations ofjobs with collective doses of 10,000 person-mrem or more.
5.10.3 Reviewing other selected jobs, as part of the ALARA Program process review to measure its effectiveness.
5.10.4 Reviewing job exposure records and ALARA oriented audits as requested by the Station ALARA Committee Chairman.
5.10.5 Reviewing progress toward meeting specific ALARA Program goals, and as appropriate, recommending corrective actions and adjustments in goals.
5.10.6 Establishing and recommending to Site Vice President for approval, specific ALARA objectives and goals.
5.10.7 Assisting in the collection, analysis, and evaluation of data necessary to measure ALARA Program effectiveness.
5.10.8 Reviewing submitted ALARA suggestions and recommending applicable suggestions for approval and awards.
5.10.9 Reviewing ALARA cost benefit evaluations, as required.
5.10.10 Reviewing and approving departmental exposure reduction action plans through the Station ALARA Coordinator.
DOMINION VPAP-2102 REVISION 12 PAGE 13 OF 66 6.0 INSTRUCTIONS 6.1 ALARA Organization and Administration 6.1.1 ALARA Coordinators Specific individuals from the following departments and groups shall be assigned responsibility for, and authority to implement, the ALARA Program consistent with Dominion policy. These individuals shall be provided with staffing necessary to support the program.
- a. Station ALARA Coordinator A specific individual in Radiological Protection shall be assigned as Station ALARA Coordinator.
- b. Department ALARA Coordinators Department ALARA Coordinators shall be assigned from the following groups/departments with responsibility for supporting the ALARA Program:
- Engineering
- Training
(
- Operations
- Radiological Protection
- Maintenance
- Nuclear Site Services
- Others, as required by the Station ALARA Committee (SAC)
DOMINION VPAP-2102 REVISION 12 PAGE 14 OF 66 6.1.2 Station ALARA Committee (SAC)
A SAC shall be established with responsibility for overall coordination of the Station ALARA Program and for advising Site Vice President in matters relating to ALARA.
- a. Membership
- 1. SAC shall be composed of the following members:
- Chairman, Director Nuclear Station Operations and Maintenance
- Vice-Chairman, Manager Nuclear Maintenance
- Secretary, Station ALARA Coordinator
- Member, Manager Radiological Protection and Chemistry
- Member, Manager Nuclear Operations
- Member, Manager Nuclear Site Engineering
- Member, Manager Nuclear Site Services
- Member, Department ALARA Coordinators
- 2. SAC member alternates shall be appointed as follows:
- The Vice-Chairman, shall have authority to act as Chairman when Director Nuclear Station Operations and Maintenance is absent
- Department Managers and the Station ALARA Coordinator shall designate their own alternates
- Department ALARA Coordinator alternates shall be designated by responsible Department Managers
- b. Meeting Frequency
- 1. SAC shall meet on a regularly scheduled basis, or as convened by the Chairman.
- 2. SAC meetings shall be scheduled by the SAC Chairman or the Station ALARA Coordinator; however, meetings may be requested by any member.
- c. Meeting Quorum
- 1. A SAC quorum shall consist of:
- Director Nuclear Station Operations and Maintenance or Manager Nuclear Maintenance
- Manager Radiological Protection and Chemistry or alternate
- Station ALARA Coordinator or alternate
- One Department ALARA Coordinator or alternate
DOMINION VPAP-2102 REVISION 12 PAGE 15 OF 66
- 2. Committee decisions shall be reached by a vote of members present and determined by a majority rule.
- d. Procedural Rules
- 1. The Chairman shall conduct SAC meetings.
- 2. To be included on the SAC agenda, topics or items should be made known to the Station ALARA Coordinator at least three (3) business days before a meeting.
- 3. The Station ALARA Coordinator should prepare and distribute meeting agendas to SAC members before the meeting.
- 4. Individuals initiating an agenda item should make themselves available to the committee, as needed, for clarification or questions.
- 5. Sub-committees may be appointed as deemed necessary by SAC to research and report on matters pertinent to ALARA.
- 6. Departments and groups shall grant prompt and direct access to data, records, procedures, or other material pertinent to matters before the SAC.
6.2 Exposure Reduction Plans 6.2.1 Department ALARA Exposure Reduction Plan (Attachment 19) should be submitted by departments that are budgeted to expend greater than five (5) percent of the total station exposure, if required by the Station ALARA Committee Chairman.
6.2.2 Exposure Reduction Plans should include exposure reduction actions for both operational and outage periods.
6.2.3 Exposure Reduction Plans should be reviewed and approved by the Station ALARA Committee.
6.3 Pre-Job Planning and Review Pre-job planning should be initiated as far in advance of work initiation as possible to ensure jobs are planned with adequate time allotted to implement ALARA Actions. Exposure optimization and reduction techniques shall be considered and documented in all work planning. The ALARA Action Plan (Attachment 6) shall be used for this purpose.
DOMINION VPAP-2102 REVISION 12 PAGE 16 OF 66 6.3.1 TEDE ALARA Evaluation
- a. Radiation exposure control measures should be designed, selected, implemented, and maintained to ensure that anticipated and actual doses are ALARA.
- b. When application of process or engineering controls to keep radioactive material in air below values that constitute an Airborne Radioactivity Area are not practicable, then other controls shall be implemented to maintain the total effective dose equivalent (TEDE) ALARA. The use (or non-use) of a respiratory device with a lower protection factor (PF) than the peak concentration may be selected to be consistent with TEDE ALARA. A Respiratory Hazards evaluation should be a component of a TEDE ALARA Evaluation.
- c. C-HP-1 042.21 0, Respiratory Hazards Evaluation And Respiratory Protection Selection, provides several methods to evaluate respiratory hazards due to airborne radioactive material if historical conditions are not applicable or unavailable. If one of these methods are used in the TEDE ALARA Evaluation process, list methods used in part 2.1 ofTEDE ALARA Evaluation (Attachment 2) and attach worksheet from C-HP-1042.210 to the completed TEDE ALARA Evaluation.
- d. A TEDE ALARA Evaluation shall be done if any of the following conditions are met. Documented evaluations for use (or non-use) of individual respiratory protection shall be performed on TEDE ALARA Evaluation (Attachment 2) when activity exposure projections meet one or more of the following:
- Ratio ofDDE dose rate, mrem/hr, to DAC fraction is greater than 50 and without respirator use, individual DAC-Hour exposure would exceed 10 DAC-Hours
- Requested by Station ALARA Coordinator
- e. HP Operations should initiate TEDE ALARA Evaluation (Attachment 2) based on the above criteria, unless deemed appropriate to do so by the Station ALARA Coordinator.
DOMINION VPAP-2l02 REVISION 12 PAGE 17 OF 66
- f. When perfonning a TEDE ALARA evaluation, the following conditions should be considered as per NUREG CR-004l:
- Heat stress to workers
- Type and quantity of protective clothing needed for the job
- Environmental conditions and how they affect the use or non-use of respirators
- Skill and experience level of the workers
- Post work consequences of not using respirators, such as the need for personal decontamination, skin-dose assessments, responding to portal monitor alanns, psychological strain on workers who are contaminated.
- g. A respirator inefficiency factor up to 15% is acceptable for calculating TEDE ALARA. Inefficiency factors greater than 15% are acceptable with adequate justification. Justification could consist of factors such as:
- Prior field experience
- Professional judgement
- Time-motion studies
- Mock-up exercises
- Job planning interviews
- h. If initiated outside the Station ALARA Staff, forward TEDE ALARA Evaluation fonn to the Station ALARA Staff for completion.
- i. The Station ALARA Staff shall forward a copy of completed TEDE ALARA Evaluation fonn to HP Operations for concurrence and implementation. If discrepancies exist, the HPSS shall contact Station ALARA Staff to resolve discrepancies prior to implementation.
DOMINION VPAP-2102 REVISION 12 PAGE 18 OF 66 6.3.2 Pre-Job ALARA Review Initiation NOTE: A flowchart detailing the ALARA review process is shown on Pre-Job and Post-Job ALARA Review Flow Chart (Attachment 1).
The pre-job ALARA review process shall be initiated as follows:
- a. When a Radiation Work Pennit request is submitted in accordance with VPAP-21 01, Radiation Protection Program, or some other means or notification of radiological work (e.g., plan of the day, outage planning) is received, Radiological Protection shall perfonn a dose budget in accordance with approved RP procedures.
If the dose budget for the task is less than 1,000 person-mrem and a TEDE ALARA evaluation is not required, then the RWP may be issued following standard Health Physics RWP preparation.
- b. A refined dose budget shall be perfonned by Station ALARA Staff if:
- The initial dose budget by Health Physics is 1,000 person-mrem or greater.
- The RWP is part of a multiple RWP job/activity that has a cumulative dose budget of 1,000 person-mrem or greater.
- The job activity is detennined by the Station ALARA Coordinator to require further review and evaluation
- 1. If the refined dose budget is less than 1,000 person-mrem, the RWP package may be returned to Health Physics for RWP preparation and issuance.
- 2. If the refined dose budget is 1,000 person-mrem or greater, or part of a job or activity being reviewed by ALARA, and the job or activity requires a Specific RWP, perfonn the following:
NOTE: Although ALARA Reviews are not specifically required for General RWPs, reviews should be perfonned if the General RWP is budgeted to expend 10,000 person-mrem or more on an annual basis.
- Complete a pre-job ALARA meeting (Step 6.3.3)
DOMINION VPAP-2102 REVISION 12 PAGE 19 OF 66 6.3.3 Pre-Job ALARA Meeting A pre-job ALARA meeting shall be held for RWPs or jobs/activities with dose estimates of greater than 1,000 person-mrem.
- a. The cognizant Job Supervisor/Lead Person shall be contacted by Station ALARA Staff to review estimated person-hours and other applicable data used to perform the dose budget.
- b. If appropriate, Station ALARA Staffmay refine the dose budget based on cognizant Job Supervisor/Lead Person input.
- 1. Ifthe refined dose budget is less than 1,000 person-mrem, the RWP package may be returned to Health Physics for RWP preparation and issuance.
NOTE: Pre-Job Worksheets are not required for routine or repetitive tasks (e.g., reactor disassembly) unless deemed appropriate by the Station ALARA Coordinator.
- 2. If the dose budget remains 1,000 person-mrem or greater and the job or activity is a non-routine or non-repetitive task, then complete a Pre-Job ALARA Worksheet (Attachment 5).
- 3. Log the RWP on an ALARA Evaluation Log (Attachment 3) and assign an ALARA Evaluation (AE) number to package. For jobs/activities with multiple RWPs totaling more than 1,000 person-mrem, use a Multiple RWP Tracking (Attachment 4) to log RWP numbers/descriptions. Indicate whether a Multiple RWP Tracking Sheet is used by checking the applicable block on ALARA Evaluation Log.
- c. Station ALARA Staff shall record a job description on the Pre-Job ALARA Worksheet by completing Part 1 (Job Description).
- d. The cognizant Job Supervisor/Lead Person shall complete Part 2 of the Pre-Job ALARA Worksheet. Station ALARA Staff shall provide assistance or guidance as necessary.
- e. Station ALARA Staff shall indicate agreement with content of the worksheet by completing and signing Part 3 (Worksheet Review) of the Pre-Job ALARA Worksheet.
DOMINION VPAP-2102 REVISION 12 PAGE 20 OF 66
- f. The cognizant Job Supervisor/Lead Person should ensure contents of the Pre-Job ALARA Worksheet are discussed with the work crew, preferably at a departmental pre-job discussion.
6.3.4 ALARA Action Plan Additional actions, work evolutions, preparation requirements, or exposure control techniques not already planned or identified to maintain exposures ALARA should be specified on an ALARA Action Plan (Attachment 6) as one of the following:
- ALARA Design Considerations
- Good Practices to Use During Task
- Lessons Learned from Past Performance of Task (i.e., internal OEs)
[Commitment 3.2.1J
- Identify and incorporate external OEs related to the work being reviewed
[Commitment 3.2.1J
- Required Shielding for Task
- Contingency Plans for Potential Problem Areas
- Specific ALARA Actions to Reduce Exposure for Task
Foreman to ensure that actions are fully understood. This review may either occur during the pre-job ALARA meeting or at a future time acceptable with ALARA and the Job Supervisor/Foreman.
- 2. Sign and date the form.
- 3. Have RWP Requestor or cognizant Job Supervisor/Foreman acknowledge the requirements by signing and dating the form.
DOMINION VPAP-2102 REVISION 12 PAGE 21 OF 66
- 1. Ensure that any comments identified on the ALARA Action Plan for inclusion on the RWP are added to the RWP.
6.3.5 Less Than 5,000 Person-mrem The pre-job ALARA review process for the RWP or job/activity budgeted to expend less than 5,000 person-mrem may be concluded as follows:
- a. If applicable, the cognizant Job Supervisor/Foreman shall ensure that ALARA Actions are satisfied on the ALARA Action Plan.
- b. Station ALARA Staff shall forward the original RWP Request and other applicable ALARA evaluation documentation to Health Physics RWP writers.
6.3.6 5,000 Person-mrem or More
- a. If the budgeted dose for the RWP or job/activity is 5,000 person-mrem or more, the following actions shall be taken in addition to those specified in Step 6.3.5.
- 1. The responsible department head shall review all Pre-Job Worksheets and ALARA Action Plans for task. Acknowledge review by signing on ALARA Action Plan (Attachment 6).
- 2. Station ALARA Coordinator and, if applicable, the cognizant Job Supervisor/Foreman shall present the ALARA evaluation package to the Station ALARA Committee (SAC) for review and approval.
- 3. SAC shall review the proposed RWP or job/activity and associated ALARA evaluation documentation to ensure that sufficient planning and exposure control methods have been applied.
- 4. The SAC review and approval shall be documented on Station ALARA Committee Pre-Job ALARA Review (Attachment 8). Station ALARA Staff may complete applicable sections of the form before the SAC meeting.
DOMINION VPAP-2102 REVISION 12 PAGE 22 OF 66
- b. If the budgeted dose for the job/activity is less than 15,000 person-mrem, the ALARA evaluation package shall be forwarded to Health Physics for RWP preparation and issuance.
- c. Ifthe budgeted dose for the job/activity is 15,000 person-mrem or more, in addition to the actions specified in Step 6.3.6.a., the following review and approval shall be obtained before the RWP is issued:
- The Site Vice President shall review the proposed RWP or job/activity and associated ALARA evaluation. Approval and recommendations shall be documented on the Station ALARA Committee Pre-Job ALARA Review.
- d. Upon proper review and approval, Station ALARA Staff shall forward applicable ALARA evaluation documentation to Health Physics for RWP preparation and issuance.
6.4 Monitoring Job Performance 6.4.1 Exposure Tracking The Station ALARA Staff shall operate an exposure tracking system.
- a. A unique number shall be assigned to each RWP reviewed; however, a single number may be assigned for a group of tasks that make up a single job for purposes of tracking job exposure (e.g., overhaul of a reactor coolant pump). The system shall:
- Provide exposures by RWP number, work group, and individual integrated doses
- Provide comparison of actual dose to exposure estimates
- Provide appropriate reports to supervision and management to enable assessment of performance and trend analysis
- b. Station ALARA Staff should:
- 1. Monitor dose accumulation of active RWP tasks with actual or budgeted doses of 1,000 person-mrem or more.
- 2. Ensure that departments have access to exposure reports that list status ofRWPs issued to department including:
- Budgeted RWP dose
- Cumulative exposure to date
DOMINION VPAP-2102 REVISION 12 PAGE 23 OF 66
- 3. Notify cognizant job supervisors when RWPs budgeted to expend greater than 1,000 person-rnrem accumulate 100% of budgeted exposure.
NOTE: HIS-20 will automatically place an RWP or a task on lock out when the actual exposure exceeds the budgeted exposure by 125 percent.
- 4. Ensure that RWPs or tasks budgeted to expend 1,000 person-rnrem or more are placed on lock out when accumulated person-mrem exceeds 125 percent of budgeted person-mrem, and applicable job supervision has failed to participate in a RWP/task ALARA re-evaluation meeting as requested.
- c. Departments shall:
- 1. Periodically review exposure status of personnel and jobs being supervised.
- 2. Contact and meet with Station ALARA Staff when active RWP or tasks budgeted to exceed 1,000 person-rnrem or greater, or are expected to exceed 125 percent of budgeted dose.
6.4.2 Lock Out Notification To control and document reasons for exceeding budgeted RWP or task dose, the following actions shall be taken when RWPs or tasks accumulate 1,000 person-mrem or more, and reach 100 percent or more of budgeted person-rnrem:
- a. Department representative or cognizant Job Supervisor/Foreman shall:
- 1. Review current status ofjob.
- 2. Consider possible reasons for higher than budgeted person-rnrem accumulation.
- 3. Determine if job can be completed within 125 percent of budgeted person-rnrem.
- b. If it is determined that the job can be completed within 125 percent of budgeted person-rnrem, no further actions are required.
- c. Ifit is determined that the job cannot be completed within 125 percent of the budgeted person-mrem, the responsible job supervisor shall contact Station ALARA Staff and schedule an RWP/task ALARA re-evaluation meeting as soon as practicable.
DOMINION VPAP-2102 REVISION 12 PAGE 24 OF 66
- d. RWP or tasks that accumulate a dose of more than 1,000 person-mrem, or are budgeted to expend 1,000 or more person-mrem and exceed 125 percent of budgeted person-mrem, will be placed on lock out.
- e. Lock outs may be rescinded after either:
- An RWP/task ALARA re-evaluation meeting
- As directed by Manager Radiological Protection and Chemistry 6.4.3 RWP/Task ALARA Re-evaluation Meetings RWP/task ALARA re-evaluation meetings shall be conducted by Station ALARA Staff and attended by cognizant job personnel to:
- Review current status ofjob and any problems encountered
- Evaluate reasons for exceeding budget
- Calculate new budgeted person-mrem required to complete job
- Recommend actions to minimize dose
- a. RWP/task ALARA re-evaluation meetings shall be documented on a RWP/Task ALARA Re-Evaluation (Attachment 9) and include:
- Revision of RWP/task person-mrem budget, if applicable
- Identification of probable reasons for higher than expected person-mrem accumulation
- Evaluation of additional exposure reduction actions (e.g., modification of work practices or installation of temporary shielding)
- Identification of meeting attendees
- c. Station ALARA Staff shall ensure that a revised RWP/task person-mrem budget has been input into HIS-20, if applicable, and update ALARA tracking records.
DOMINION VPAP-2102 REVISION 12 PAGE 25 OF 66 6.4.4 Work in Progress Reviews (WIPR)
Work in Progress ALARA Review (Attachment 20) should be performed as directed by the Station ALARA Staff to:
- Monitor work progress
- Identify problem areas or good practices during task performance
- Implement corrective actions or stress good practices where applicable prior to completion of the task
- a. Work in Progress Reviews may be performed by any station employee. Individuals performing the WIPR should review the WIPR with the task foreman or lead individual and obtain applicable signatures.
- b. Areas identified as recommended for or requiring implementation of corrective actions should be discussed with the job supervisor, HP Shift Supervisor, and the ALARA Staff, as applicable, to ensure exposure reduction action items are addressed.
- c. WIPRs shall be routed to the Station ALARA Staff for review and then forwarded to HP for attachment to the appropriate RWP package.
- d. WIPR information should be utilized during ALARA post-job reviews as applicable.
6.5 Post-Job and Process Reviews 6.5.1 Post-Job Reviews
- a. If actual expended jOb/activity dose is less than 1,000 person-mrem, post-job reviews may be limited to that which is normally performed as a part of the job close-out process.
DOMINION VPAP-2102 REVISION 12 PAGE 26 OF 66 NOTE: Separate post-job reviews may be required for jobs and activities utilizing multiple RWPs, or RWP with multiple tasks.
- b. RWPs or jobs/activities that meet any of the following criteria shall require a post-job review that includes a meeting with the cognizant job supervision and workers (as required by Station ALARA Staff). Post-job reviews shall be documented on Post-Job Review (Attachment 10).
- Job/Activity expending 10,000 person-mrem or more
- Job/Activity expending more than 1,000 person-mrem and exceeded 125 percent of original budget
- Station ALARA Coordinator deems it prudent to document relative success/failure of special exposure reduction techniques or lessons learned
- c. During post-job ALARA review meetings, attendees shall:
- 1. Discuss job performance. If expended person-mrem was less than 10,000 person-mrem, completion of the Post-Job Review form shall serve as documentation ofjob performance discussion. If expended person-mrem was equal to or greater than 10,000 person-mrem, the Post-Job Review form shall be reviewed by the Station ALARA Committee.
- 2. Identify specific dose reduction techniques that were particularly effective/ineffective.
- 3. Discuss relative success or failure of any special work techniques.
- 4. Discuss problems encountered and lessons learned.
- 5. Identify possible reasons for significant differences between actual and budgeted person-mrem, if applicable.
DOMINION VPAP-2102 REVISION 12 PAGE 27 OF 66 NOTE: RWPs that cover multiple components or areas of the plant with multiple components (e.g., MOV maintenance, scaffold installation, valve work in loop rooms) may not require a post-job critique ifpre-determined by the Station ALARA Committee during the pre-job committee review of the RWP.
- 1. Assist preparation of a briefjob summary and post-job critique, by cognizant job supervision.
- 2. Schedule SAC review and approval of post-job review and critique report.
- e. SAC review ofjobs that expend 10,000 person-mrem or more shall be documented on Post-Job Review (Attachment 10) and include:
- Review of post-job summary
- Review of applicable ALARA evaluation documents
- Comments and recommendations as required
- Approval of SAC Chairman 6.5.2 Job History Files A work activity historic file should be instituted and maintained by the Station ALARA Staff that contains Post-Job Review forms, surveys, and other applicable documents. The job history file should be maintained as a resource for review guidance on similar future jobs.
6.5.3 Process Review The Station ALARA Staff may conduct selected reviews of processes that exceed 5,000 person-mrem per year. The results of this review shall be documented, and those findings that identify opportunities for reducing exposure should be submitted to the Station ALARA Committee for review and implementation. These opportunities may involve changes to methods, procedures, and equipment.
6.6 Temporary Shielding 6.6.1 Temporary shielding is shielding installed for a specific time period, (e.g., for duration ofjob) to reduce personnel exposure.
DOMINION VPAP-2102 REVISION 12 PAGE 28 OF 66 NOTE: The need for temporary shielding should be considered at the earliest possible stage in work planning to afford ALARA, Engineering, and Health Physics sufficient time to review, approve, and install required shielding.
6.6.2 Individuals or work group supervision who identify the need for temporary shielding shall submit a Temporary Shielding Request in accordance with VPAP-2105, Temporary Shielding Program.
6.7 ALARA Suggestions Employees (Dominion and supplemental personnel) are encouraged to submit ALARA Suggestions that will provide for continuous improvement of the ALARA Program.
6.7.1 Submitting an ALARA Suggestion Submit an ALARA suggestion as follows:
- a. Obtain a copy of form ALARA Suggestion (Attachment 11) from a suggestion forms box location, the ALARA office, or from Electronic Document Management System(s) (EDMS).
- b. Complete Part 1 of the ALARA Suggestion form.
- c. Sign and date the request. Submit completed form to Station ALARA Coordinator for review and evaluation.
6.7.2 Evaluating ALARA Suggestions
- 1. Review the ALARA Suggestion for applicability. Previously submitted suggestions that do not save personnel exposure or are obviously cost-prohibitive may be rejected without using an ASR number. Rejected suggestions should be returned to the requestor with an explanation for the rejection.
- 2. If the suggestion is applicable, assign an identifying number to the suggestion (e.g., For a 2002 ASR, 02S-ASR-00I at Surry or 02N-ASR-00I for North Anna).
- 3. Record applicable information on the ALARA Suggestion Log (Attachment 12).
DOMINION VPAP-2102 REVISION 12 PAGE 29 OF 66 NOTE: The Station ALARA Committee (SAC) does not approve and budget for the implementation ofREAs, plant modifications, and/or DCPs. The SAC should evaluate suggestions for ALARA benefit. Approved suggestions will be assigned to an appropriate department for further evaluation and implementation using existing plant processes.
- b. The suggestion may be evaluated by an individual(s) or by a SAC subcommittee as appropriate.
- 1. The suggestion should be evaluated for ALARA benefit and value added using the CostlBenefit Matrix.
- 2. The evaluator(s) shall recommend approval or rejection of the ALARA Suggestion. If rejected, provide explanation.
- 3. Return the suggestion to the Station ALARA Coordinator.
- 4. The Station ALARA Staff shall complete part 3 of the suggestion form approving or rejecting the suggestion. If rejected, provide explanation.
6.7.3 Review and Approval of ALARA Suggestions
- a. ALARA Suggestions recommended for approval by the evaluator(s) or the Station ALARA Staff shall be reviewed by the SAC.
- b. The Station ALARA Coordinator shall schedule the SAC presentation and document the SAC evaluation of ALARA Suggestions.
- 1. Document the reason for rejection.
- 2. Inform the originator of the evaluation results as soon as practicable.
- 3. Update ALARA Suggestion Log.
DOMINION VPAP-2102 REVISION 12 PAGE 30 OF 66 NOTE: Approval of ALARA Suggestions by the SAC does not provide a guarantee of suggestion implementation. It will be the responsibility of the assigned department to perform further review of the suggestion for applicability and initiate any required actions for budget, REA, or implementation support of the suggestion, as appropriate.
- d. If the SAC approves the suggestion:
- 1. The Station ALARA Coordinator or SAC Chairman shall document actions approved by the committee.
- 2. Designate a responsible department (part 5) to further evaluate and implement the suggestion as appropriate.
- 3. The SAC Chairman shall sign the suggestion form and return the form to the Station ALARA Staff for distribution to the responsible department.
- 4. The Station ALARA Staffshould obtain the signature and date ofthe individual from the responsible department who accepts the suggestion.
- 5. Ensure a copy of the completed suggestion is maintained for the ALARA file and provide a copy or original to the responsible department.
- e. Close out the ALARA Suggestion:
- 1. Notify the originator of evaluation results and assigned department point of contact.
- 2. Update the ALARA Suggestion Log.
- 3. File copy of suggestion for future reference. The Station ALARA Coordinator will determine the length of suggestion retention.
- 4. If applicable, the Station ALARA Staff should distribute approved suggestions to other Dominion nuclear units.
6.7.4 ALARA Suggestion Awards NOTE: Station ALARA Committee and the Station ALARA Staff are encouraged to submit ALARA Suggestions, but are not eligible for awards.
- a. ALARA Suggestion recognition and/or awards are encouraged for approved suggestions.
DOMINION VPAP-2102 REVISION 12 PAGE 31 OF 66
- b. The Station ALARA Committee and/or the Station ALARA Coordinator will determine the type of awards to be presented.
6.8 Design Change Package ALARA Review DCPs shall receive an ALARA Evaluation in accordance with General Nuclear Standards STD-GN-0019, Engineering ALARA Design Guide, and STD-GN-0001, Instructions for DCP Preparation.
6.8.1 As required by the General Nuclear Standards, Project Engineers shall forward copies of the draft DCP performed by Engineering to the Station RP Department.
NOTE: For design changes with little ALARA impact, comments may be limited to the ALARA Section of the Program Review.
6.8.2 RP shall review the DCP. If ALARA concerns exist that have not been addressed, contact or meet with the Project Engineer to discuss recommendations.
6.9 Determining and Tracking ALARA Goals 6.9.1 ALARA Goals
- a. ALARA goals are established to actively involve, guide, and monitor the performance of the Station and departments towards meeting ALARA objectives.
- b. Goals of the ALARA Program shall be established with Station ALARA Committee and Site Vice President concurrence. Goals should be realistic, specific, measurable, reviewed on a periodic basis, revised to ensure relevancy, and include:
- Collective radiation exposure (person-rnrem) per year
- Collective radiation exposure (person-rnrem) per department/organization
- Collective radiation exposure (person-rnrem) per outage
- Collective radiation exposure (person-rnrem) for major jobs 6.9.2 Establishing Station ALARA Goals NOTE: Department ALARA Goals (Attachment 15) provides an overview of the approval and tracking process for Station and department ALARA goals.
- a. The Station ALARA Coordinator should propose a Station person-rnrem goal for the upcoming year before the end of the current year.
DOMINION VPAP-2102 REVISION 12 PAGE 32 OF 66
- b. The Station ALARA Coordinator should compile, review, and evaluate:
- NBU Management long-term ALARA goals and proposed Station person-mrem goals and options
- Outage schedule for upcoming year, with emphasis on high person-mrem jobs
- Design changes to be implemented, with emphasis on high person-mrem jobs
- Cumulative Station person-mrem for current year
- Station exposure trends, eighteen (18) month averages and progress toward meeting long-term ALARA goals
- Applicable industry averages and performance of similarly designed Stations
- c. The Station ALARA Coordinator shall record applicable person-mrem data for the current year and previous two years on Part 1 of Station ALARA Goals (Attachment 14). Such data includes:
- Cumulative Station person-mrem
- Average person-mrem/day for scheduled outages
- Number of scheduled outage, unscheduled outage, and non-outage days
- d. Based upon evaluation of the items above, the Station ALARA Coordinator shall record proposed goals and scheduled outage days in Part 2 of Station ALARA Goals (Attachment 14) including:
- Station person-mrem goal for the upcoming year
- Number of scheduled outage days and non-outage days for the upcoming year
- Historical person-mrem/day from the scheduled outage and non-outage blocks on Part 1. The average person-mrem per day figure for scheduled outage should be used for both Unit 1 and Unit 2 as applicable
- Budgeted exposure of non-routine tasks (e.g., DCPs, major maintenance) for the upcoming year
- Compute expected person-mrem for upcoming year
- Propose a percent reduction factor for the person-mrem/day goals for scheduled outage days and non-outage days so the total budgeted person-mrem for the year equals the proposed Station person-mrem goal
- Review sign and date the form
DOMINION VPAP-2l02 REVISION 12 PAGE 33 OF 66
- e. The Station ALARA Coordinator should schedule an SAC meeting to review, discuss and agree on a proposed Station goal.
- f. At the meeting, the Station ALARA Coordinator shall distribute copies of the proposed Station goal and present an overview of ALARA goals, including:
- 1. Long-term ALARA goals established by NBU Management and progress toward meeting them.
- 2. Current and historical exposure trends for Station and departments.
- 3. Outage schedule and design changes to be implemented during upcoming year.
- 4. Selected high-dose outage-related jobs and routine work activities likely to contribute a significant portion of Station dose during the upcoming year.
- 5. Proposed Station ALARA goals for upcoming year.
- g. During the meeting, SAC members should discuss proposed ALARA goals, outage schedules, exposure reduction trends, and progress toward meeting long-term ALARA goals.
- h. The SAC shall determine applicable ALARA Program performance indicators to be tracked during the upcoming year.
- i. If not completed at initial goals review meeting, another SAC meeting may be scheduled to finalize proposed Station and department ALARA goals.
6.9.3 Establishing Department ALARA Goals
- a. Station ALARA Staff shall record department specific information and proposed department ALARA goals on Department ALARA Goals (Attachment 15) as follows:
- 1. Compile and record historical person-mrem data.
- 2. Calculate percent of Station dose each department/organization has historically contributed.
DOMINION VPAP-2102 REVISION 12 PAGE 34 OF 66 NOTE: Proposed department person-mrem goals should be based on probable department work load for upcoming year and may not necessarily correspond to percent of Station dose expended in the past.
- 3. Propose an annual person-mrem goal for each department or organization that have or are expected to expend 5 percent of more of the Station annual dose.
Departments with smaller exposure inputs << 5% of total station exposure) may be grouped under a common group called "Other Departments" so the sum of department goals equals the proposed station annual person-mrem goal.
- b. The Station ALARA Coordinator shall review, and sign and date the form and notify applicable directors and managers of proposed ALARA goals.
6.9.4 Department/Organization Review of Proposed Person-mrem Goals
- a. Department directors/managers shall review and evaluate proposed person-mrem goals.
- b. Goal evaluation should include an estimate of collective dose required to accomplish assigned departmental activities during the upcoming year. Specific items to be considered include:
- 1. Outage schedule and high person-mrem jobs planned for upcoming year.
- 2. Previous department work trends and historical person-mrem.
- 3. Maintenance and inspection schedules.
- 4. Design changes to be implemented.
- 5. Routine and repetitive work activities that expend significant dose annually.
- 6. Department person-mrem breakdown by craft.
- 7. Supplemental personnel support requirements.
- c. Doses to supplemental personnel assigned to a department shall be included in the department person-mrem goal.
- 1. Assist in evaluating proposed goals.
DOMINION VPAP-2102 REVISION 12 PAGE 35 OF 66
- 2. If applicable, aid in developing department exposure reduction action plans.
- e. If department evaluation concludes that the proposed person-mrem goal is reasonably achievable, applicable director/manager shall denote this on Department ALARA Goal Review (Attachment 16). Forward Attachment 16 to the Station ALARA Coordinator.
- f. If department evaluation concludes that the proposed person-mrem goal is not reasonably achievable:
- 1. The department shall document reasons on Department ALARA Goal Review (Attachment 16). Include estimated person-mrem required to perform department functions and list major jobs/activities involved.
- 2. The director/manager shall sign and date the form, forward to the Station ALARA Coordinator, and present arguments to change the goal at the SAC meeting to finalize proposed goals.
6.9.5 ALARA Goal Approval The SAC should finalize proposed ALARA goals. Department directors, managers, or ALARA Coordinators that do not agree with the proposed goals should attend to discuss any changes in department person-mrem goals.
- a. The Station ALARA Coordinator shall present a summary of ALARA goal status and, if applicable, recommend revisions to specific ALARA goals based on a more complete evaluation.
- b. If applicable, department directors/managers may present reasons to change proposed ALARA goals, as documented on Department ALARA Goal Review (Attachment 16).
- c. SAC shall review Station and department ALARA goals and requests to identify and change goals to ensure that:
- 1. Long-term goals set by NBU Management are being met.
- 2. Sum of department person-mrem goals equals Station person-mrem goal.
- 3. Goals are challenging, yet reasonably achievable.
DOMINION VPAP-2102 REVISION 12 PAGE 36 OF 66
- d. Final proposed goals shall be entered on Attachments 14 and 15 and approved by the SAC chainnan. Attachments 14 and 15 shall be submitted to Site Vice President for review and approval.
- e. The Site Vice President shall:
- 1. Review recommended goals.
- 2. If applicable, approve goals by signing and dating Attachments 14 and 15.
- 3. Infonn the Vice President Nuclear Operations ofproposed Station person-mrem goal.
- 4. Forward fonns to the Station ALARA Coordinator.
- f. Proposed Station person-mrem goals shall be subject to final approval by Vice President Nuclear Operations. Changes to Station person-mrem goals approved by the Vice President shall result in a corresponding adjustment to department person-mrem goals.
- g. Upon approval of Station ALARA goals, Station ALARA Staff shall distribute copies of approved Station and department goals to applicable Station supervision.
6.9.6 Monitoring ALARA Goal Performance
- a. Within 30 days after the end of each calendar quarter, the Station ALARA Coordinator shall prepare and distribute to applicable directors/managers a copy of the ALARA Goal Status Report (Attachment 17).
- b. Attachment 17 shall include:
- 1. Station ALARA goal status, including:
- Annual person-mrem goal versus cumulative person-mrem to date
- Quarterly person-mrem goal versus actual person-mrem expended in quarter
- Scheduled outage and non-outage person-mremlday goals versus actual perfonnance
- 2. Department ALARA goal status, including:
- Annual person-mrem goal for each department versus cumulative person-mrem
- Quarterly person-mrem goal for each department versus actual person-mrem expended in quarter
DOMINION VPAP-2102 REVISION 12 PAGE 37 OF 66
- c. Applicable directors/managers shall review the report and evaluate department performance relative to established goals.
- 1. Departments or organizations that exceed annual or quarterly person-mrem goals by 15 percent and the total quarterly department exposure is 100 person-mrem or more shall complete an ALARA Goal Variance Report (Attachment 18) by documenting:
- Probable reasons for variance
- Jobs and activities that contributed the largest portion of department person-mrem
- Additional exposure reduction action plans that can be implemented, if applicable
- 2. Completed ALARA Goal Variance Reports shall be approved by responsible director/manager and forwarded to applicable Department ALARA Coordinator or Station ALARA Coordinator for presentation to SAC.
- d. On a quarterly basis, SAC shall:
- 1. Compare Station cumulative and quarterly person-mrem to established goals.
- 2. Compare departments' cumulative and quarterly person-mrem to established goals.
- 3. Review department variance reports as documented on ALARA Goal Variance Report (Attachment 18).
- 4. Review implementation status of applicable department exposure reduction plans.
- 5. Review status of key ALARA program performance indicators.
- e. At the end of each calendar year, the Station ALARA Coordinator shall review the performance of Station and departments relative to established ALARA goals.
DOMINION VPAP-2l02 REVISION 12 PAGE 38 OF 66 7.0 RECORDS 7.1 The following individual and packaged documents and copies of any related correspondence completed as a result of implementing or performing this procedure are records. They shall be transmitted to Records Management in accordance with VPAP-170 1, Records Management.
Prior to transmittal, the sender shall assure that:
- Each record is packaged when applicable,
- QA program requirements have been fulfilled for Quality Assurance records,
- Each record is legible, completely filled out, and adequately identifiable to the item or activity involved,
- Each record is stamped, initialed, signed, or otherwise authenticated and dated, as required by this procedure.
7.1.1 Individual Records
- Station ALARA Goals
- Department ALARA Goals
- Department ALARA Goal Review
- ALARA Goal Status Report
- ALARA Goal Variance Report NOTE: ALARA Evaluation records to be transmitted with RWPs in accordance with C-HP-1081.030, Radiation Work Permits: Extending, Revising, and Terminating.
7.1.2 Record Packages ALARA Evaluations:
- Pre-Job ALARA Worksheet
- ALARA Action Plan
- RWP/Task ALARA Re-evaluation
- Post-Job Review
- Work In Progress Reviews
DOMINION VPAP-2102 REVISION 12 PAGE 39 OF 66 7.2 The following documents completed as a result of implementing this procedure are not Quality Assurance records and are not required to be transmitted to Records Management.
- ALARA Suggestion/Evaluation form
- ALARA Suggestion Log
- Multiple RWP Tracking Sheet
- ALARA Evaluation Log
DOMINION VPAP-2102 REVISION 12 PAGE 40 OF 66 ATTACHMENT 1 (Page 1 of2)
Pre-Job and Post-Job ALARA Review Flow Chart Planners
- Estimates total person-mrem required to I
RWP Requestor
- Completes RWP Request Form Health Physics
- Estimates person-mrem required perform job or activity I
- Prepares proposed RWP I
~
~
Station ALARA Starr Station ALARA Staff Budgeted
- Verifies RWP schedule date NO
- Returns RWP package to exposure
- Reviews job history files Health Physics Cognizant Job Supervisor
- Ensures ALARA Evaluation is - NO person-mrem YES ~
- Refines person-mrem budget Station ALARA Staff
- Evaluates appropriate ALARA Actions Budgeted
- Requires completion of Pre-Job ALARA Worksheet, if applicable exposure
- Advises Job Supervisor if Pre-Job Planning Meeting required
~ 5,000
, person-mrem Responsible Department Head Station ALARA Staff YES
- Prepares and issues RWP exposure
~ 15,000 person-mrem
-
- Reviews, provides recommendations, and approves Pre-Job ALARA Review YES I
Site Vice President I
- Review person-mrem budget and discuss high dose tasks
- Review requirements specified on ALARA Action Plan
- Discuss exposure and contamination control techniques
- Discuss individuals current and expected dose
- Discuss coordination with other work groups
( Continued on page 2 of 2 )
DOMINION VPAP-2102 REVISION 12 PAGE 41 OF 66 ATTACHMENT 1 (Page 2 of2)
Pre-Job and Post-Job ALARA Review Flow Chart Cognizant Job Supervision Station ALARA Staff
- Ensures ALARA Actions are complied with
- Monitors dose accumulation of active RWPs with actual or budgeted
- Contacts/meets with Station ALARA Staffifperson-mrem has I------t~ doses 1,000 person-mrem or greater or is expected to exceed budget by 25%
- Notifies department when RWPs reach 100% of budgeted person-mrem and specifies accumulated person-mrero at which a lock out will be applied if no RWP ALARA re-evaluation meeting is held
- Holds RWPffask ALARA Re-evaluation meetings with job supervision to Health Physics revise person-mrero budget and evaluate reasons for exceeding original budget
- Terminates RWP at job completion
- Places RWP or task on lock out if actual person-mrem exceeds budgeted dose fooIf------j by 25% and job supervision fails to participate in RWPffask ALARA Re-evaluation meeting Post-Job ALARA Review Meeting Cognizant Job Supervision
- Critiques job performance
- Discusses problems encountered, effectiveness of Actual Exposure YES , - - - - - - - - - - - - - - ,
Station ALARA Staff ALARA techniques, lessons learned, etc.
> 10,000
- Schedules Post-Job ALARA meeting
- Reviews Post-Job ALAR A Review form Actual exposure NO
> 10,000 person-mrero NO Actual exposure
> 10,000 person-mrero YES YES Actual exposure YES exceeds budget by Station ALARA Staff 25%
- Assist job summary report preparation NO Station ALARA Committee
- Reviews and approves Post-Job Review and Job summary report Job activity requires 2 or more RWPffask YES ALARA Re-Evaluations NO Station ALARA Staff 1-- ------_ _-1~1
- Updates ALARA Evaluation Log
- Maintains Job History Files
VPAP-2102 REVISION 12 PAGE 42 OF 66 TEDE ALARA Evaluation VPAP-21 02 - Attachment 2 Page 1 of 2 Task being performed Indicate (mark) the following conditions which exist or are expected to exist forthis request.
I Individual estimated DDE for an RWP job while using respirator will exceed 0.5 rem.
I Ratio of DDE dose rate (mrem/hr) to DAC fraction is greater than 50 and w~hout respirator use, an individual DAC-Hour exposure would exceed 10 DAC-Hours.
1Station ALARA Coordinator determination that TEDE ALARA Evaluation is appropriate.
[1 _
Note: Work description from VPAP-21 02 Pre-Job Worksheet (Form No. 728680) and historical data (if available) may be used for this evaluation.
concern.
Describe radiological conditions from survey data (or historical, or projected).
Record time anticipated in specific area of concern.
Record air concentration value (fraction or percent DAC) from estimated or historical.
Other Describe Radiological Engineering controls.
Describe historical conditions and resu~s.
Describe similar work in other or adjacent areas.
Other Key: RWP-Radlatlon Work Permit; TEDE ALARA-Total Effective Dose Equivalent As Low As Reasonably Achievable; DDE-Deep Dose Equivalent; mrem-millirem; DAC-Derived Air Concentration; PF-Protectlon Factor Form No. 720134{Jan 2001)
VPAP-2102 REVISION 12 PAGE 43 OF 66 TEDE ALARA Evaluation VPAp*2102
- Attachment 2 Page 2 of 2
[ ] Individual [ ] By Task Without Respirators Number of workers xtime hours x Dose rate mrem/hr
- mrem, external dose DAC Fraction x time hours DAC-Hours DAC-Hours x2.5 mrem/DAC-Hour a mrem, internal dose Estimated dose for implementing engineering or process controls = mrem Total. mrem With Respirators (Note: Hour estimate should reflect productivity increase or decrease.)
Numberofworkers x time hours x Dose rate mrem/hr
- _ mrem, external dose DAC Fraction x time hours
_ _ _ _ _ _ _ _ DAC Hours respirator PF _
DAC-Hours x 2.5 mremlDAC-Hour
- mrem, internal dose Estimated dose for implementing engineering or process controls a mrem Total _ _ _ _ _ _ _ _ mrem Remarks Work Job
[ ] With respirators [ ] Without respirators [ ] As indicated below
[ ] Type of respirator Justification or Comments Prepared By (Printed Name) Prepared By (Initials) Date ALARA Coordinator Approval (Signature) Date Key: DAC*Derived Air Concentratio n; m rem-m iIIirem; hr*hour; PF*Protectlon Factor; ALARA-As Low As Reasonably Ach levable Form No. 720134(Jan 2001)
ALARA Evaluation Log
'Dominion-VPAP-2102 - Attachment 3 Page 1 of 1 Page ALARA RWprrask ALARA Re-Evaluation
- RWP OrigInal Final Percent of Post-Job Review Evaluation Person-mrem Revised Revised RWPITask Budgeted Number Number Work Description and Location Budget Budget Date Budget Date Person-mrem Person-mrem Status
[ 1Not Required
[ 1Completed Mull. RWP
[ 1Yes [ ] No Date
[ 1Not Required
[ ] Completed Mutt.RWP Date
[ 1Yes [ 1No
[ 1Not Required
[ 1Completed Mull. RWP Date
[ ] Yes [ ] No
[ ] Not Required
[ ] Completed Mutt.RWP
[ 1Yes [ ] No Date
[ 1Not Required
[ 1Completed Mull. RWP Date
[ 1Yes [ 1No
[ ] Not Required
[ ] Completed Mull. RWP Date
[ 1Yes [ ] No
- If tracking more than one RWP under the same AE number, use a Multiple Tracking Sheet (Form No. 729122) for recording RWP numbers.
Key: RWP*Radiatlon Work Permit
~
oG; Fam No. 727263(JIrMlI20OS) m<<
-I:>-cn~
-1:>- ...... ......-
00';0
'"rjZ~
0\-0 O\NN
VPAP-2102 REVISION 12 PAGE 45 OF 66
~
rf/jp~ Dominion*
Multiple RWP Tracking VPAP-2102 - Attachment 4 Page 1 of 1 AE# I Job Description RWP# Job BUdgeted Rev Actual Trm Percent Description Expos Expos Budgeted (Person~m re m) (Person*mrem) Expos Total Form No. 729122(June2006)
VPAP-2102 REVISION 12 PAGE 46 OF 66
~ Pre-Job ALARA Worksheet 1IiJ**Dominion' VPAP-2102 - Attachment 5 Page 1 of 2 ALARA Evaluation Number RWP Number BUdgeted Person-mrem See Multiple RWP Tracking Sheet
[ I Yes [ I No Job Description Note: The following items should be considered by supervision during job planning. Answer each question, provide a brief explanation.
- a. What are the controlling job procedures and do they include Health Physics Ho d Points?
- b. Describe the work to be performed "under this ALARA Evaluation. (Attach work list if appropriate.)
- c. Describe exposure reduction measures planned for this job.
- d. What job history files and industry operating experience have been reviewed? [Com mltment 3.2.1]
- e. What lessons learned from previous similar work or industry operating experience have been incorporated? [Comm Itment 3.2.1]
- f. What coordination w~h other groups is planned?
- g. What are the work area access/ex~ points?
- h. Where are the designated low dose wa~ing and staging areas?
- i. What special communication devices will be used (e.g., radios, closed circuit television)?
- j. What additional services are required (lighting, air, electrical) and have they been requested?
- k. What prefabrication, disassembly, and assembly of components is to be performed outside of radiation areas?
I. What will be done to minimize post-job cleanup and radwaste?
- m. What changes in equipment status could result in significant interruption of job or changes in radiological conditions?
- n. What other work activ~ies could cause undue interruption of job or cause increased radiation levels or airborne radioactivity?
Key: ALARA-As Low As Reasonably Achievable; RWP-Radiatlon Work Permit Form No. 728680(June 2006)
VPAP-2102 REVISION 12 PAGE 47 OF 66 Pre-Job ALA RA Worksheet VPAP-2102 - Attachment 5 Page 2 of 2 ALARA Evaluation Number
- a. What tools, equipment, and materials are necessary to perform the task? Utilize contaminated tools if available. (Attach tool list if available.)
- b. What special tools/equipment will be used to minimize time and exposure?
- c. What methods of tool control and/or foreign material entry control will be used?
- d. How are tools/equipment to be tested for operability prior to entry? (Manual, Pneumatic, etc.)
a.
- b. What systems/components will be filled with water or flushed to reduce job area dose rates?
- c. Describe Engineering controls to control airborne activity. (H EPA filters, G love bags, etc.)
- d. Review RWP survey and note hot spots and general area (G/A) dose rates:
- b. What special training is required/planned (mock-ups, dry-runs)? If none, explain.
- c. What photos/video library material have been reviewed to familiarize work crews with job site, equipment, or work tasks?
- d. How will shift turnovers be conducted?
- e. What job procedures need to be reviewed by crew members prior to work execution?
Completed By Cognizant Job Supervisor (Signature) Date ALARA Actions Applicable
. : ~
[ I Yes [ I No Reviewed and Approved By Station ALARA Staff (Signature) Date Key: ALARA-As Low As Reasonably Achievable; RWP-Radlation Work Permit; HEPA-Hlgh Efficiency Particulate Air Form No. 728680(June 20(6)
VPAP-2102 REVISION 12 PAGE 48 OF 66 ALARA Action Plan
'Dominion-VPAP*2102 - Attachment 6 Page 1 of 1 Job ALARA Evaluation Number i RWP Number Person-mrem History for Task to include last two outages in same Unit or non-outage performances, best ever, and current outage / or non-outage performance goal.
Year: !Year: _ Year/Unit: [Year' Dose: mrem Dose: mrem Dose: -------m-r-e-m DOS~: mrem I
ALARA Design Considerations Good Practices to Use During Task Lessons Learned During Past Performance of Task Required Shielding for the Task Contingency Plan(s) for Potential problem Area(s)
Potential Problem(s) Methods to Avoid/Minimize Impact on Task Exposure Specific ALARA Actions to Reduce Exposure for Task Craft Comments Specific ALARA Requirements to be Added to RWP Signature by Date Submitted by ALARA (Signature) Date Approved by Job Supervisor (Signature) Date Approved by Department Head for Tasks> 5,000 Person-mrem (Signature) Date Key: RWP-Radiation Work Permit Form No. 725462(June 2006)
DOMINION VPAP-2102 REVISION 12 PAGE 49 OF 66 ATTACHMENT 7 (Page 1 of 1)
ALARA Action Plan Instructions Instructions 1.0 Record applicable ALARA Evaluation and RWP numbers.
Record the following person-mrem values for comparison by work crew:
- Last two outages or task for the affected unit or component 2.0
- Lowest exposure total for task, either unit
- Budgeted for current outage/task ALARA Design Considerations:
3.0 Record any design enhancements that reduce exposure for this task. Include tooling and permanent modifications.
Good Practices to use during task:
4.0 Record generic ALARA practices in this section.
Lessons learned during past performance of task:
- a. Record items that have the potential to recur and will impact exposure for the task 5.0 b. Include both Good Practices and areas of improvement.
- c. This area can also be used to prompt review of any identified Operation Experiences (OEs) related to the work being reviewed.
Required shielding for the task:
6.0 Record shielding packages scheduled for this task. Note if the shielding must be installed prior to certain phases of work.
Contingency plans for potential problem areas:
7.0 If any particular hazards can arise during performance of work, document possible occurrences and the potential methods to avoid them from happening or to reduce the exposure impact.
Specific ALARA Actions to reduce exposure for task:
8.0 Record task specific actions to reduce exposure. Items can be chronological if needed.
Crew comments:
9.0 During ALARA briefing, record any additional ALARA techniques addressed by the work crew.
Station ALARA Staff to sign ALARA Action Plan prior to seeking approval from Cognizant 10.0 Job Supervisor (Craft Foreman).
Craft Foreman to sign ALARA Action Plan to signify concurrence with all items. If any 11.0 item(s) can not be accomplished, annotate in crew comments section.
VPAP-2102 REVISION 12 PAGE 50 OF 66 Station ALARA Committee Pre-Job
'Dominion< ALARA Review VPAp*2102*Attachment8 Page1of1 ALAR A Evaluation Number RWP Number See Multiple RW P Tracking Sheet
[ I Yes [ I No Projected Start Date of Job Budgeted Person-mrem Estimated Person-hours Scheduled SAC Review Date Part 1 Station ALARA Committee Evaluation (To Be Documented By Station ALARA Coordinator)
- 1. Work Procedures - Comments/Recommendations
- 2. Pre-Job Planning, Scheduling, and Coordination - Comments/Recommendations
- 3. Tools and Equipment - Comments/Recommendations
- 4. Radiological Controls - Comments/Recommendations
- 5. Worker Preparation and Training - Comments/Recommendations
- 6. Exposure Review Points and In-Process ALARA Assessment Techniques - Comments/Recommendations
- 7. ALARA Actions - Comments/Recommendations
- 8. Contingency Plans or Operating Experiences (OEs)
- 9. Additional Comments, Recommendations, and Requirements Review and Approval By Chairman Station ALARA Committee (Signature) Date Site Vice President (Signature) Date Key: SAC-Station ALARA Committee; ALARA-As Low As Reasonably Achievable Form No. 730285(June 2006)
VPAP-2102 REVISION 12 PAGE 51 OF 66 R WP/Task A LARA Re-Evaluation
'DominiOn" VPAP*2102
- Attachment 9 Page 1 of 1 Part 1 Job Description ALARA Evaluation Number RWP Number EvaluatIOn Initiated By Date Job Description Part 2 RWP Person*m rem Re*Evaluation Current Person-mrem % of Revised Person-mrem % of Job ALARA Coordinator Date/Time Expended Current Person-mrem Completed Initials/Remarks Estimate to Date Estimate Estimate Part 3 RWP Person*mrem Re*Evaluation Completed by ALARA with information provided by cognizant job personnel
- 1. Check one or more of the items listed below which may have contributed to higher than expected RWP Person-mrem accumulation.
Provide an explanation for all items checked.
[ I a. Job scope changed/expanded.
[ I b. Job site radiation levels differenUchanged.
[ 1c. Encountered scheduling/coordination difficulties.
[ I d. Work extended due to tool/equipment failure.
[ I e. Work extended due to wrong or unavailable parts/tools/material.
[ 1f. Work extended due to unplanned job site preparation requirements.
[ I g. Work extended due to interruptions/interferences caused by other work activities.
[ I h. Inadequate compliance with radiological controls/requirements.
[ I i. Inadequate consideration of Pre-Job ALARA Worksheet items.
[ I j.1 nadequate shielding.
[ I k. Other.
- 2. Exposure Reduction Action to Be Implemented (if applicable)
Meeting Attendees (if applicable)
Reviewed By Station ALAR A Coordinator (Signature) I Date Form No. 727966(June 2006)
VPAP-2102 REVISION 12 PAGE 52 OF 66 Post-Job Review
'DominiOn" VPAP-2102 - Attachment 10 Page 1 of3 See M ulliple RWP Tracking Sheet
[ ] Yes [ ] No Job Description RWP Person-mrem Original Budget RWP Person-mrem Final Revised Budget Actual RWP Person-mrem Actual person-mrem required is % of original budget and % offinal budget.
Part 2 Post-Job ALARA Review Instructi ons To be completed by cognizant job supervision with assistance provided by ALARA staff.
Address/evaluate the following areas, comparing as applicable responses on Pre-Job ALARA Worksheet. Mark appropriate box and provide comments which may have either saved or contributed to the person-mrem expended for the job.
Sat Unsat N/A 1. Work Procedures
[1 [1 [ ] a. Procedure flow; i.e., logical, unnecessary steps deleted, conducive to efficient work practices.
[ ] [ ] [ ] b. Adequacy/Effectiveness of Health Physics hold points.
[1 [1 [ ] c. Adequacy/Effectiveness of proceduralized ALARA techniques including ALARA Action Plan.
[ ] [ ] [ ] d. Applicability of Radiological Work Practice.
Comments (identify by number and letter, if applicable, describe recommended changeslimprovements to work procedures)
Sat Unsat N/A 2. Pre-Job Planning, Scheduling and Coordination IT [1 [ ] a. Was adequate time available to plan work activities?
[ ] [ ] [ ] b. Completeness of RWP Request including job description, task breakdown, estimated person-hours and scope.
[ ] [1 [1 c. Were job history files useful in planning for this task?
[ ] [ ] [ ] d. How was coordination/planninglinterface with the other groups?
[ ] [1 [ ] e. Was the route used to enter/exit work area accessible and adequate?
[ ] [ ] [ ] f. Was the location and size of step-off -pads adequate?
[ ] [1 [ ] g. Were special communication devices used? If yes, describe below.
[ ] [ ] [ ] h. Was lighting, electrical, air, housekeeping, etc. adequate?
[ ] [ ] [ ] i. Were the locations of low dose rate staging and waiting areas identified by HP and used by all personnel?
[ ] [ ] [ ] j. Were the number of workers utilized in radiation areas minimized to save person-mrem?
[ ] [ ] [1 k. Were pre-job briefs/shift turnovers correct and useful?
[ ] [ ] [ ] I. Coordination of work such that interruptions due to changes in radiological conditions and interferences from other work in area was minimized.
Comments (Identify by number and letter, summarize lessons learned, successful techniques, shortcomings and recommendations for similar work in the future)
Sat Unsat N/A 3. Tools and Equipment
[ ] [ ] [1 a. Did the tool/equipment list include all necessary items? If not, comment below.
[ ] [1 [1 b. How effective were special tools and equipment? Comment below.
[ ] [ ] [ ] c. Did all tools function as designed?
[ ] [ ] [ ] d. Were all tools reliable?
[1 [ ] [1 e. Were proper methods used to verify operability and calibration of tools, equipment, and instruments?
[ ] [ ] [ ] f. Were tools available from the tool crib when requested?
[1 [ ] [1 g. Did tools remain neatly at job site for entire task?
Comments (Identify by n umber and letter, summarize lessons learned and successful/unsuccessful tech niques, tools, equipment for future reference)
Form No. 728936(Jufle 2006)
VPAP-2102 REVISION 12 PAGE 53 OF 66 Post-Job Review VPAP-21 02 - Attachment 10 Page 2 of 3 ALARA Evaluation Number RWP Number I See Multiple RWP Tracking Sheet
[ I Yes
[ ] No Sat Unsat N/A 4. Radiological Controls
[I [I -[-] a. Adequacy of RWP requirements (e.g., H P coverage, protective clothing, dosimetry, and respiratory protection).
[ I [I [I b. Use of temporary shielding.
[I [I [I c. Methods utilized to reduce job site dose rates; flushing or draining system, filling system with water, decontaminate system.
[ I [I [I d. Methods used to reduce need for respirators; wetting surfaces, supplemental ventilation, enclosures, etc.
[ I [I [ ] e. Adequacy/effectiveness of Pre-Job ALARA Worksheet in identifying ALARA planning consideration.
[I [I [ ] f. Adequacy/effectiveness of ALARA Actions.
[ I [ ] [I g. Pre-job survey data (adequacy/appropriate).
[ ] [I [ ] h. RWP dose budget.
[ I [I [I i. Actual dose comparison to previous similar jobs.
Comments (Identify by number and letter, if applicable, summarize lessons learned and successful/unsuccessful ALARA techniques, problems encountered and recommendations to be considered when planning radiological controls for similar work)
Sat Unsat N/A 5. Worker Preparation, Training and Job Performance
[ I [I [I a. Were personnel experienced with similar work utilized?
[ I [I [ ] b. Was special training, mockups, or dry-runs performed effectively?
[ I [ ] [I c. Were HP pre-job briefings on radiological conditions and requirements adequate?
[ I [I [I d. Did workers comply with radiological controls/requirements and exhibit good work practices?
[ ] [I [ ] e. Was VIMS or photos used to familiarize workers with the job site?
[ I [ ] [I f. Was any rework required on this task? Describe below.
[ I [ ] [I g. Adequacy of work crew supervision.
Comments (Identify by number and letter, if applicable, summarize lessons learned and successful techniques, and recommendations to be considered when planning similar work)
Prepared By Cognizant Job Supervisor (Signature) Date Reviewed By Station ALAR A Coordinator (Signature) Date Part 3 Station ALARA Committee Review Required for ALARA Evaluations for jobs expending 10,000 Person-mrem or greater Comments and Recommendations Chairman Station ALARA Committee (Signature) I Date Form No. 728936(June 2006)
VPAP-2102 REVISION 12 PAGE 54 OF 66 Post-Job Review (Continuation)
VPAP-2102 - Attachment 10 Page 3 of 3 ALARA Evaluation Number RWP Number I See Multiple RWP Tracking Sheet I U*III1*m*il~*iIl*l1*Ii'a-~.~.I'.II='lIi1Im' .******************[.].y.es* * *[.].N.O* * * * *
- 1. Work Procedures
- 2. Pre-Job Planning, Scheduling, and Coordination
- 3. Tools and Equipment
- 4. Radiological Controls
- 5. Workers Preparation, Training, and Job Performance Comments Form No, 728936(June 2006)
VPAP-2102 REVISION 12 PAGE 55 OF 66 ALARA Suggestion
'DominiOn" VPAP*21 02 - Attachment 11 Page 1 of 1 ALARA Suggestion Number (Provided by ALARA Coordinator after sUbmittal)
Originator (Printed Name)
I Department I Telephone Number Describe ALARA Concern and Location of Concern Recommended Solution (Attach any additional information such as sketches, prints, and supporting documentation)
Estimated Person-mrem Savings I Estimated Costto Implement Originator (Signature) (Return to Station ALARA Coordinator after signing) I Date Cost/Benefit Matrix - Estimate Cost/Benefit Matrix Value based on labeled diagram. (Place relative mark on blank diagram)
Low Cost High Cost High Benefit High Benefit Low Cost High Cost Low Benefit Low Benefit Evaluation Comments [ 1Approved [ 1Rejected Evaluator (Signature) (Return to Station ALARA Coordinator after signing) I Date
- . * -
- I : . ' ., .. . . .
Concur with Evaluation? [ 1Yes [ 1No Concur with Recommendation? [ lYes [ 1No Comments Stalion ALARA Coordinator (Signature) Date Part 4 - To Be Completed By SAC Chairman
[ 1Approved: To be forwarded to responsible Department for further evaluation [ 1 Rejected Comments SAC Chairman (Signature) Date Part 5* For ALARA Office Use Only Responsible Department Assigned I Accepted By (Signature) Date Fonn No. 728518(June 2006)
ALARA Suggestion Log
'Dominion-VPAP-2102
- Attachmenl12 Page 1 of 1 ALARA Date Originator's Suggestion Sugges~on Evaiuation Suggestion Status Suggestion Close Out Received Name and Department Numb ...
Name Assigned To Responsible Department o Approved 0 Rejected o Originator Notified Dept Dept IDate Date Point of Contact Date Name Assigned To Responsible Department o Approved 0 Rejected o Originator Notified Dept Dept IDate Date Point of Contact Date Name Assigned To Responsibie Department o Approved 0 Rejected o Originator Notified Dept Dept IDate Date Point of Contact Date Name Assigned To Responsibie Department o Approved 0 Rejected o Originator Notified Dept Dept IDate Date Point of Contact Date Name Assigned To Responsibie Department o Approved 0 Rejected o Originator Notified Dept Dept IDate Date Point of Contact Date Name Assigned To Responsible Department o Approved 0 Rejected o Originator Notified Dept Dept IDate Date Point of Contact Date Name Assigned To Responsible Department o Approved 0 Rejected o Originator Notified Dept Dept IDate Date Point of Contact Date
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DOMINION VPAP-2102 REVISION 12 PAGE 57 OF 66 ATTACHMENT 13 (Page 1 of2)
ALARA Goals Flow Chart Nuclear Operations Management
- Develop methodology for establishing long-tenn ALARA goals
- Review NBU Management long-tenn goals
- Review outage schedule and design changes to be implemented
- Estimate current year total Station person-mrem
- Compile Station department historical person-mrem data
- Propose Station ALARA goals for upcoming year
- Complete Station ALARA goals fonn
- Schedule SAC meeting to review proposed Station goals 11r Station ALARA Committee
- Meets before year end to discuss ALARA goals
- Committee adjusts ALARA goals as required
- Identifies key ALARA Program perfonnance indicators to be tracked 1
Station ALARA Coordinator
- Proposes department ALARA goals
- Distributes proposed ALARA goals to applicable departments and organizations
- Assists departments in proposed goals evaluation and provides historical person-mrem data
- Participates in department exposure reduction action plan development, if applicable
- Schedules SAC meeting to finalize proposed ALARA goals, if needed 11r Department Directors and Managers
- Review proposed ALARA goals
- Review outage schedule, previous exposure trends, maintenance and inspection schedules, design changes to be implemented, with emphasis on high dose jobs
- Complete department ALARA Goals Review fonn
- If applicable, identifY valid reasons for changing goal for presentation to SAC
DOMINION VPAP-2102 REVISION 12 PAGE 58 OF 66 ATTACHMENT 13 (Page 2 of2)
ALARA Goals Flow Chart Station ALARA Committee
- Reviews Station and department ALARA goals
- Ensures the goals are challenging, yet reasonably achievable
- Evaluates requests to change department ALARA goals
- Finalizes and approves Station and department ALARA goals
- Forwards proposed goals to Site Vice President for review and approval 1
Site Vice President
- Reviews and approves goals and informs Vice President Nuclear Operations of Station ALARAgoals Ir Vice President Nuclear Operations
- Distributes finalized Station and department ALARA goals to Station supervision
- Distributes ALARA Goal Status Report to SAC, and department directors and managers Department Directors and Managers
- Review quarterly status report and department and organization performance
- Complete ALARA Goal Variance Report ifdepartment variance from ALARA goal is :2: 15% and department expended> 100 person-mrem for the quarter 1
Station ALARA Committee
- Reviews Station and department ALARA goal performance
- Reviews department variance reports
- Reviews status of key ALARA Program performance indicators
VPAP-2102 REVISION 12 PAGE 59 OF 66 Station ALARA Goals
'DominiOn" VPAP-21 02 - Attachment 14 Page 1 of 1 Part 1 Station Person-mrem Data (Completed By Station ALARA Staff)
C urrent and Person-mrem/Day Number of Days Previous Two Years Total Scheduled Unscheduled Scheduled Unscheduled Person-mrem* Non-Outage Non-Outage (enter year) Outage Outage Outaqe Outaoe Davs Total Average
- Estimate Annual Dose for Current Year as Calculated Follows:
From To (Month, Day, Year) Total Year-To-Date Person-mrem for Current Year-to-Date Jan.1, Person-mrem Add Estimated Dose Based on Total Scheduled Outage Days Remaining days x actual Person-mrem/day Person-mrem Add Estimated Dose Based on Total Scheduled Non-O utage Days Remaining days x actual Person-mrem/day Person-mrem Estimated Annual Person-mrem for Current Year Part 2 Station ALARA Goals (Completed By Station ALARA Staff)
The Proposed Station ALARA Goal for the Year is Person-m rem Mode of Projected Historical
- Non-Routine Total % Reduction Revised P erso n-m rem/Day Person-mrem Person-mrem P erson-mrem Operation Days Person-mre m/Day Tasks Projection Applied Goal Total Non-Outage Unit 1 Outage Unil2 Outage Totals Unit #
- Description of DCP/Major Non-Routine Tasks Budgeted Person-mrem Station ALARA Coordinator (Signature) Date Part 3 Review and Approval The Approved Station ALARA Goal for the Year is Person-mrem Comments Chairman Station ALARA Committee Approval (Signature) Date Site Vice President Approval (Signature) Date Vice President Nuclear Operations Approval (Signature) Date Form No. 727265(June 20(6)
VPAP-2102 REVISION 12 PAGE 60 OF 66
,; Dominion~ Department ALARA Goals VPAP*2102 - Attachment 15 Page 1 of 1 Historical Three Year % of Station Proposed Goal Approved Goal Departments' Current Yr... Previous Two Years Average Dose *** (Person-mrem) (Person-mrem)
TOTALS
- Including Supplemental Support Estimate Cumulative Person-mrem for Current Year = Actual Person-mrem Expended + Estimated Person-mrem for Remainder of Year
... % of Station Dose = (Department 3 Year Average I Station 3 Year Average) x 100 Comments Station ALARA Coordinator Approval (Signature) Date Chainman Station ALARA Committee Approval (Signature) Date Site Vice President Approval (Signature) Date Form No. 727264 (June 2006)
VPAP-2102 REVISION 12 PAGE 61 OF 66 Department ALARA Goal Review
'Dominion. VPAp*21 02* Attachment 16 Page 1 of 1 Department Group (If Applicable)
The Proposed Department ALARA Goal for the Year is Person-m rem ALARA Goal Review (To Be Completed By Applicable Department Supervision)
The Proposed ALARA G oalls:
[ I Reasonably Achievable
[ I Not Reasonably Achievable If Not Achievable, Explain Reviewed by Department ALARA Coordinator (Signature)
Comments Department Director/Manager (Signature) Date Form No. 727266(June 2006)
VPAP-2102 REVISION 12 PAGE 62 OF 66 ALARA Goal Status Report
'DominiOn" VPAP-2102 - Attach ment 17 Page 1 of 1 Prepared By Name I Date Calendar Quarter Year
[ ] 1st [ ] 2nd [ ] 3rd [ ] 4th Exceedln g ALARA Annual Year to Date Quarterly Quarterly Person-mrem/Day Person-mrem/Day Goals By > 15%
Person-mrem Person-mrem Person-m rem Person-mrem Non-Outage
& total >100 m rem Sch ed uled Outage Goal Actual Goal Actual Yes No Goal Actual Goal Actual Annual Year to Date Quarterly Quarterly
- Exceeding Department Person-m rem Person-mrem Person-m rem Person-m rem Goals By > 15% Remarks Goal Actual Goal Actual Yes No TOTALS
- Departments exceeding established ALARA goals by 15% or more and the total quarterly department exposure is more than 100 person-mrem should complete an ALARA Goal Variance Report and forward the form to the Station ALARA Coordinator for presentation to the Station ALARA Committee.
Comments Form No. 727268(June 2006)
VPAP-2102 REVISION 12 PAGE 63 OF 66 ALARA Goal Variance Report VPAp*2102
- Attachment 18 Page 1 of 1 uepartment Group (If Applicable)
Prepared By (Printed Name) Calendar Quarter IYear
[ ]1 st [ ] 2nd [] 3rd [] 4th (A) Annual Department Person-mrem Goal (C) Quarterly Department Person-mrem Goal (B) Year to Date Department Person-mrem (D) Quarterly Department Person-mrem Actual Variance' = 100 x [(B-A)/A] % Variance' = 100 x [(D-C)/C] %
- Variance should be either negative or less than 15%
Part 2 Possible Reasons For Variance Check one or more possible causes for variance and provide explanation. If necessary, attach explanation.
[ ] Outage Extended
[ ] Special Work
[ ] Unscheduled Work
[ ] Other Part 3 Exposure Sig nificant Activities Review department exposure and list major jobs or ongoing activities which contribute the largest portion of department dose.
RWP Number Job Description Person-mrem Part 4 Exposure Reduction Action Plans Describe any exposure reduction action plans which will be implemented Part 5 Review Director/Manager Review (Signature) Date Station ALARA Committee Chairman (Signature) Date Form No. 727761(June 2006)
Department ALARA Exposure
'Dominion- Reduction Plan VPAP*2102
- Attachment 19 Page 1 of 1 Department Group (If Applicable)
Exposure Reduction Recommendations for Current Year:
[ J Routine , Activities:
[ 1Outage Activities:
[ 1Other:
Prepared By Name Date Department Management Review Comments Department DrectorlManager (Signature) I Date Comments Station ALARA Comm~tee Chairman (Signature) I Date
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VPAP-2102 REVISION 12 PAGE 65 OF 66 Work in Progress ALARA Review VPAp*2102
- Attachment 20 Page 1 of 2 Part I. Work Identification Job Description RWP# AE # _ Date Performed Part II. Current Work Status Current Exposure Budget Actual Exposure Percent of Budget Percent Job Complete Part III. Job Evaluation Place an 'X" in the appropriate block.
Explain "No" answers in the "Comments" Section Questions Yes No NIA
- 1. Is Job progressing as planned?
- 2. Will job complete WIThin the present budget?
- 3. Are the job procedures adequate?
- 4. Is the iob scope the same as that which was evaluated?
- 5. Was ore-iob olanninQ adequate?
- 7. Are Low Dose WaITing Areas being used?
- 8. Are tools used meeting the needs of the job?
- 9. Are workers familiar with the job site radiological conditions?
- 10. Is shielding adequate for the job?
- 11. Did the workers use VIMS prior to performing work?
- 12. Are radwaste reduction measures adequate for iob?
Comments: Use Page 2 of this form or attach additional sheets if necessary Follow up/Action Items discussed and/or performed with Job Supervisor Performed by: Date: _
Reviewed by: Date: _
Responsible Job Supervisor Reviewed by: Date: _
Station ALARA Coordinator Note: File form with Post-Job package. Form No. 722541 (June 2006)
VPAP-2102 REVISION 12 PAGE 66 OF 66 Work in Progress ALARA Review VPAP-2102 - Attachment 20 Page 2 of 2 Job Description RWP # _ AE #
Comments Form No. 722541(June 2006)
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS Unit shutdown was commenced due to primary to secondary leakage in the B SG.
At 80% power, operators initiated a manual trip due to inability to maintain Pressurizer level and pressure.
Prior to the trip, 1-CH-RI-128, Reactor Coolant Letdown Radiation Monitor, was reading 6 X104 mR/hr and had been steadily increasing over the last hour.
Operators isolated AFW flow to B SG due to level increasing uncontrollably following the trip.
- A Safety valve on the B SG has opened and failed to reseat.
Pressurizer level is off-scale low.
All safeguards equipment is functioning as designed and offsite power is available.
Emergency Response Organization response has been delayed due to icy roads and a severe winter storm.
INITIATING CUE You are requested to classify the emergency event and determine any Protective Action Recommendation if required.
This is a time critical JPM.
02/25/08 Page: 1 of 8
Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM NEW Classify event and determine PAR.
TASK STANDARDS Event classified as a General Emergency in accordance with EPIP-1.01 TAB B.2 PAR A (shelter in place 2 Mile radius and 5 Miles downwind) selected due to known impediments making evacuation dangerous.
KIA
REFERENCE:
GEN-2.4.41 (2.9/4.6)
GEN-2.4.44 (2.4/4.4)
ALTERNATE PATH:
N/A TASK COMPLETION TIMES Validation Time = 10 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ ] SATISFACTORY [ ] UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature 1 Date EVALUATOR'S COMMENTS 02/25/08 Page: 2 of 8
( Dominion North Anna Power Station ADMINISTRATIVE JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM NEW READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES The trainee has completed the applicable course knowledge training at the Senior Reactor Operator level.
INITIAL CONDITIONS Unit shutdown was commenced due to primary to secondary leakage in the B SG.
At 80% power, operators initiated a manual trip due to inability to maintain Pressurizer level and pressure.
Prior to the trip, 1-CH-RI-128, Reactor Coolant Letdown Radiation Monitor, was reading 6 X10 4 mR/hr and had been steadily increasing over the last hour.
02/25/08 Page: 3 of 8
Operators isolated AFW flow to B SG due to level increasing uncontrollably following the trip.
A Safety valve on the B SG has opened and failed to reseat.
Pressurizer level is off-scale low.
All safeguards equipment is functioning as designed and offsite power is available.
Emergency Response Organization response has been delayed due to icy roads and a severe winter storm.
INITIATING CUE You are requested to classify the emergency event and determine any Protective Action Recommendation if required.
This is a time critical JPM.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT None PERFORMANCE STEPS START TIME 02/25/08 Page: 4 of 8
(~' .*.* < <
Determine the event category using the emergency action level Procedure Step _ _
table index.
ICritical Step ISAT[] UNSAT[]
Standards Event is identified as a RCS EVENT, Tab B, Attachment 1 of EPIP-1.01.
IN otes/Comm ents Review the emergency action level tab associated with the event Procedure Step _ _
cate or .
I SAT [] UNSAT [ ]
IL.;;;S;,;;ta;;;;n,;,;;d;;;;a;;,,;rd;;,;;s~_ _1 Tab B reviewed to determine highest level of classification.
IN otes/Com ments Use available resources to obtain indications of emergency Procedure Step _ _
conditions.
ISAT [] UNSAT [ ]
IL.,;;S;,;;;ta;;;,;,n,;,;;d;,;;;a~rd;;,;;s~_ _1 Tab B indications compared to Initial Conditions provided in JPM.
rates/comments 02/25/08 Page: 5 of 8
~ Verify that an emergency action level has been exceeded. IProcedure Step I_C_r_iti_c_a_1S_t_e.L-P 1 SAT [] UNSAT []
IStandards IEvent is classified as a General Emergency lAW TAB B.2 INotes/Comments Identify Protective Action Recommendation (PAR) is applicable Procedure Step _ _
based on exceed in action level for General Emer enc .
ISAT [] UNSAT [ ]
Standards EPIP-1.06 Protective Action Recommendation Matrix compared to Initial Conditions rovided in JPM.
INotes/Comments
~ Determine Protective Action Recommendation. IProcedure Step ICritical Step ISAT [] UNSAT [ ]
Standards Determine Protective Action Recommendation from EPIP-1.06 is PAR A based on known im ediments make evacuation dan erous.
IN otes/Comments 02/25/08 Page: 6 of 8
>>>>> END OF EVALUATION <<<<<
STOP TIME 02/25/08 Page: 7 of 8
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required) 02/25/08 Page: 8 of 8
VIRGINIA POWER NORTH ANNA POWER STATION EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER PROCEDURE TITLE REVISION EPIP~l.Ol EMERGENCY MANAGER CONTROLLING PROCEDURE 43 (With 3 Attachments) PAGE 1 of 7 PURPOSE To assess potential emergency conditions and initiate corrective actions.
ENTRY CONDITIONS Any of the following:
- 1. Another station procedure directs initiation of this procedure.
c 2. A potential emergency condition is reported to the Shift Manager.
Approvals on File Effecti ve Date S /l7/A()O?
I."; ~:~ \: \-.,:.: '.:,~. ~ ':."" ~'. ,. "\\" " ,: ,.,; " ,", ";".": (";',"\, ,;-;.,,:.-.o::::;~.':~'.~;C~*.-** '~."': ,;r:~M '"_-;". 7.<: <:.;,',:-, .~r '~.". ':, '\~.'.; '". '", \'~.' 'r.' .'.', ;'.~. F_". "._, "'I ** :,,~, *** ~. "' .. "'~' ' *. ', ", -, ~,-.". ' * .' **:~~'."<" *.".-. '.-~'. -**-. ", -.: , -:' ,~.~ ', ';"~'. '.~' .. ",-', ";", '.~ .-:";"",-':'";"; '1,"""'-"" .;-,; ".";.7( *c ' ,'_-:(.r;*~~; ,; '" r.<.~. '.~'.', '", "., ,:1 ','>",,-. ';~.";-';' .""....*.. ,~;:
NUMBER PROCEDURE TITLE REVISION EPIP-I.OI EMERGENCY MANAGER CONTROLLING PROCEDURE 43 PAGE .
2 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION: Declaration of the highest emergency class for which an Emergency Action Level is exceeded shall be made.
- * * * * * * * * * ~ * * * * * * * * * * * * * * * * * * * * * * * * * * *
- NOTE: The PCS is potentially unreliable in the event of an earthquake.
Therefore. PCS parameters should be evaluated for accuracy should this situation occur.
I EVALUATE EMERGENCY ACTION LEVELS:
a) Determine event category using Attachment 1. EMERGENCY ACTION LEVEL TABLE INDEX b) Review EAL Tab associated with event category c) Use Control Room monitors. PCS.
and outside reports to get indications of emergency conditions listed in the EAL Table d) Verify EAL - CURRENTLY EXCEEDED d) IE basis for EAL no longer exists when discovered AND no other reasons exist for an emergency declaration. THEN do the following:
- RETURN TO procedure in effect .
- GO TO VPAP-2802.
NOTIFICATIONS AND REPORTS. to make one-hour. non-emergency reports for classification without declaration.
1£ EAL was NOT exceeded. THEN RETURN TO procedure in effect.
(STEP 1 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION EPIP-1.0I EMERGENCY MANAGER CONTROLLING PROCEDURE 43 PAGE 3 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1 EVALUATE EMERGENCY ACTION LEVELS: (Continued) e) Record procedure initiation:
- By:
Date:
Time:
f) Initiate a chronological log of events g) Declare position of Station Emergency Manager NOTE: Assembly. accountability and/or initiation of facility staffing may not be desired during certain situations (e.g .* security event.
severe weather, anticipated grid disturbance) or may have already been completed. These activities should be implemented as quickly as achievable given the specific situation.
_ _ 2 CHECK - CONDITIONS ALLOW FOR 1£ deviation from normal emergency NORMAL IMPLEMENTATION OF EMERGENCY response actions warranted. THEN RESPONSE ACTIONS do the following:
a) Refer to Attachment 3.
Considerations for Operations Response Under Abnormal Conditi ons.
b) Consider applicability of 50.54(x).
c) IE classification/assembly announcement deferred. THEN GO TO Step 4.
NUMBER PROCEDURE TITLE REVISION
( EPIP-l.OI EMERGENCY MANAGER CONTROLLING PROCEDURE 43 PAGE 4 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
_ _ 3 NOTIFY PLANT STAFF OF ALERT OR HIGHER
. CLASSIFICATION:
a) Check classification - ALERT OR a) GO TO Step 4.
HIGHER b) Check if emergency assembly and bY Do the following:
accountability - PREVIOUSLY CONDUCTED 1) Have Control Room sound EMERGENCY alarm and make announcement on station Gai-Tronics system as follows:
"(Emergency classification) has been declared as the result of (event)
"All Emergency Response personnel report to your assigned stations*
"All contractor personnel not responding to the emergency and all visitors report to the Security Building" "All other personnel report to your Emergency Assembly Areas*
- 2) Repeat RNO Step 3.b.I.
- 3) GO TO Step '4.
c) Have Control Room sound EMERGENCY alarm and make announcement on station Gai-Tronics system as follows:
"(Emergency classification) has been declared as the result of (event) d) Repeat Step 3.c
NUMBER PROCEDURE TITLE REVISION EPIP-I.OI EMERGENCY MANAGER CONTROLLING PROCEDURE 43 PAGE 5 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- * * * * * * * * * * * * * * * * * * * * * * * ~ * * * * * * * * * * * * *
- CAUTION: Continue through this and all further instructions unless otherwise directed to hold.
_ _ 4 INITIATE SUPPORTING PROCEDURES:
a) Direct Emergency Communicators to initiate the following procedures:
- 1) EPIP-Z.01. NOTIFICATION OF STATE AND LOCAL GOVERNMENTS
- 2) EPIP-2.02.,NOTIFICATION OF NRC b) Direct HP to initiate EPI P-4. 01, .RADIOLOGICAL ASSESSMENT DIRECTOR CONTROLLING PROCEDURE c) Establish communications with Security Team Leader:
- 1) Provide Security with current emergency classification
- 2) Notify Security which Operations Shift is designated for coverage
- 3) Direct Security to initiate EPIP-5.09. SECURITY TEAM LEADER CONTROLLING PROCEDURE
NUMBER PROCEDURE TITLE REVISION EPIP-l.OI EMERGENCY MANAGER CONTROLLING PROCEDURE 43 PAGE 6 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
__'_ 5 CHECK TSC - ACTIVATED If TSC NOT activated. THEN do the following:
a) Have STA report to the Control Room.
b) Notify Manager Nuclear Operations or Operations Manager On Call.
c) Consider having Radiological Assessment Director report to the Control Room.
d) WHEN relief SEM arrives, THEN perform turnover using EPIP-I.OI. Attachment 2, Turnover Checklist.
__ 6 IMPLEMENT [PIP FOR EMERGENCY CLASSIFICATION IN EFFECT:
- Notification of Unusual Event -
GO TO EPIP-I.02, RESPONSE TO NOTIFICATION OF UNUSUAL EVENT
- Alert -
GO TO EPIP-I.03. RESPONSE TO ALERT
- Site Area Emergency -
GO TO EPIP-I.04, RESPONSE TO SITE AREA EMERGENCY
- General Emergency -
GO TO EPIP-I.05. RESPONSE TO GENERAL EMERGENCY
NUMBER PROCEDURE TITlE REVISION
£PIP-l.Ol EMERGENCY MANAGER CONTROLLING PROCEDURE 43 PAGE 7 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
______ 7 NOTIFY OFFSITE AUTHORITIES OF EMERGENCY TERMINATION:
a) State and local governments (made by LEOF or CEOF when activated) b) NRC
_____ 8 NOTIFY STATION PERSONNEL ABOUT THE FOLLOWING:
- Emergency termination
- Facility de-activation
- Selective release of personnel
- Completion and collection of procedures
- Recovery
______ 9 TERMINATE EPIP-I.OI:
- Give completed EPIPs. forms and other applicable records to Nuclear Emergency Preparedness (TSC Emergency Procedures.
Coordinator if TSC activated)
- Completed By:
Date:
Time:
-END-
NUMBER ATTACHMENT TITLE REVISION EPIP-1.01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT PAGE INDEX 1
1 I"\T A?
CAUTION:
- Declaration of the highest emergency class for which an EAL is exceeded shall be made .
- Emergency Action Levels shall be conservatively classified based on actual or anticipated plant conditions.
EVENT CATEGORY:
- 1. Safety. Shutdown. or Assessment System Event A
- 2. Reactor Coolant System Event B
- 3. Fuel Failure or Fuel Handling Accident ...............*............ C
. 4. Conta i nment Event .....................*..*.*.. , D
- 5. Radioactivity Event ...........*.................*.......*......... E
- 6. DELETED i
- 7. Loss of Secondary Coolant ...............*.*........*.............. B
- 8. Electrical Failure H
- 9. Fire .....*.......... ~ I
- 10. Security Event J
- 11. Hazard to Station Operation ....................................*.. K
- 12. Natural Events ............*.......*....*....*..................... L
- 13. Miscellaneous Abnormal Events ..*.................................. M
NUMBER ATTACHMENT TITLE REVISION EPI P-1. 01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB A) PAGE SAFETY. SHUTDOWN. OR ASSESSMENT 1 SYSTEM EVENT 2 of 41' CONDITION/APPLICABILITY INDICATION CLASSI FICATION CAUTION: EAL C.2 is duplicated below for cross-reference/comparison to EAL A.l:
. C. 2
- Pro b a b1 e 1 a r g e Loss of Main Feedwater GENERAL radioactivity System, Condensate System EMERGENCY release initiated and Auxiliary Feedwater by loss of heat System sink leading to core degradation MODES 1. 2. 3 & 4
- 1. Loss of functi on
- Total loss of the SITE AREA needed for unit HSD Charging/SI System EMERGENCY conditi on OR MODES 1. 2, 3 & 4 Total loss of the Main Feedwater and Auxiliary Feedwater systems
- 2. Fail ure of the
- Reactor trip setpoint and SITE AREA Reactor Protection coincidences - EXCEEDED EMERGENCY System to initiate and complete a AND required trip while at power
- Automatic trip from RPS -
FAILED MODES 1 & 2 AND
- Manual triE from Control Room - FAI ED
NUMBER ATTACHMENT TITLE REVISION EPIP-l.01. EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB A) PAGE SAFETY, SHUTDOWN, OR ASSESSMENT 1 SYSTEM EVENT 3 of 42 CONDITION/APPLICABILITY INDICATION CLASS! FICATION
- 3. Inability to monitor
- Most (>75%) or all SITE AREA a significant annunciator alarms on EMERGENCY transient in panels "A" to "K" - NOT progress AVAILABLE MODES 1, 2. 3 & 4 AND
- All computer monitoring capability (e.g .* PCS)
- NOT AVAILABLE AND
- Significant transient - IN PROGRESS (e.g .. reactor trip, SI actuation, turbine runback >25% thermal reactor power. thermal power oscillations >10%).
AND
- Inability to directly monitor anyone of the following using Control Room indications:
- Subcriticality
- Core Cooling
- Heat Sink
- Vessel Integrity
- Containment Integrity
- 4. Evacuation of Main Evacuation of the Control Room SITE AREA Control Room with with local shutdown control not EMERGENCY control not established within 15 minutes establi~hed within 15 minutes ALL MODES
.:.. : ~ .* :.,.~~; ..:; ~ I:..':' :: :': ,'., .. ~'.' '." , "._~~ ,' : :-~"_"""'..l.. ~" ..' __
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB A) PAGE SAFETY, SHUTDOWN, OR ASSESSMENT 1 SYSTEM EVENT 4 of 47 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 5. Total loss of
- Secondary system cooling ALERT function needed for capabil i ty - UNAVAILABLE unit CSD condition AND MODES 5 & 6
- Loss of any of the following systems:
- Component Cooling
- RHR AND
- RCS temperature GREATER THAN 140 F 0
- 6. Fail ure of the
- Reactor trip setpoint and ALERT Reactor Protection coincidences - EXCEEDED Sy~tem to com~lete a trip which ta es the AND Reactor Subcritical
- Automatic trip from RPS -
MODES 1 & 2 FAILED AND
- Manual trip - REQUIRED AND
- Manual trip from Control Room - SUCCESSFUL
NUMBER ATTACHMENT TITLE REVISION EPIP-1.01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB A) PAGE SAFETY, SHUTDOWN, OR ASSESSMENT 1 SYSTEM EVENT 5 of 42 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 7. Unplanned loss of
- Unplanned loss of most ALERT safety system (>751) or all annunciator annunciators with alarms on panels "A" to nK" compensatory for GREATER THAN 15 minutes indicators unavailable or a AND transient in progress
- All computer monitoring capability (e.g., PCS)
MODES 1. 2, 3 & 4 - NOT AVAILABLE OR Significant transient -
INITIATED OR IN PROGRESS (e.g .* reactor trip, SI.
turbine runback > 25%
thermal reactor power.
thermal power oscillations
> 10%)
- 8. Evacuation of Main Evacuation of the Control Room ALERT Control Room with shutdown control required established within 15 minutes ALL MODES
- 9. Inability to reach
- Intentional reduction in . NOTI FICATION required mode within power, load or temperature OF UNUSUAL technical lAW T.S. Action Statement - EVENT specification limits HAS COMMENCED MODES 1. 2. 3 & 4 AND
- T.S. Action Statement time limit for mode change -
CANNOT BE MET
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE . 43 ATTACHMENT* (TAB A) PAGE SAFETY
- SHUTDOWN. OR ASSESSMENT 1 SYSTEM EVENT h of 42 CONDITION/APPLICABILITY INDICATION CLASSIFICATION
- 10. Fai 1ure of a safety
- ReS NOTI FICATION or relief valve to OF UNUSUAL close after pressure
- RCS pressure - LESS EVENT reduction. which may THAN 2000 psig affect the health and safety of the OR public NDT Protection System -
MODES 1. 2. 3. 4 & 5 IN SERVICE AND
- Any indication after lift or actuation that Pressurizer Safety or PORV - REMAINS OPEN AND
- Flow - UNISOLABLE
- Excessive Steam Generator Safety. PORV or Decay Heat Release flow as indicated by rapid RCS cool down rate AND
- Main Steam pressure greater than 100 psi below setpoint of affected valve
- 11. Unplanned loss of
- Unplanned loss of most NOTIFICATION most or all safety (>75%) or all annunciators OF UNUSUAL system annunciators on panels "A" to."K" for EVENT for greater than 15 GREATER THAN 15 minutes minutes MODES 1. 2, 3 & 4
NUMBER ATTACHMENT TITLE REVISION
( EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB A) PAGE SAFETY
- SHUTDOWN. OR ASSESSMENT 1 SYSTEM EVENT 7 of 42 CONDITION/APPLICABILITY INDICATION CLASSIFICATION
- 12. Loss of * . Station PBX phone system - NOTIFICATION communications FAILED OF UNUSUAL capabil ity EVENT ALL MODES
- Station Gai-tronics system - FAILED AND
- Station UHF radio system -
FAILED
NUMBER ATTACHMENT TITLE REVISION EPIP-1.0I EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB B) PAGE REACTOR COOLANT SYSTEM EVENT I
R of 42 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 1. Loss of 2 of 3 Any two of a). b) or c) exist GENERAL fission product and the third is imminent: EMERGENCY ba rri ers with potential loss of a) Fuel clad integrity failure 3rd barrier as indicated by any of the following:
ALL MODES
- RCS specifi~ activity greater than or equal to 300.0 ~Ci/gram dose equivalent 1-131 OR 5 or more core exit
. thermocouples greater than 1200 F0 OR Containment High Range Radiation Monitor RM-RMS-165. -166 or RM-RMS-265. -266 GREATER THAN 1.88xl0 2 R/hr b) Loss of ReS integrity as indicated by any of the following:
- RCS pressure greater than 2735 psig OR Loss of Reactor Coolant /
in progress c) Loss of containment integrity as indicated by any of the following:
- Containment pressure greater than 60 psia and not decreasing OR Rel ease path to /.
environment -EXISTS
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB B) PAGE REACTOR COOLANT SYSTEM EVENT 1
q nf 4?
CONDITION/APPLICABILITY INDICATION CLASSIFICATION
- 2. Fuel fail ure with Any two of a), b) or c) exist and the GENERAL steam generator third is imminent: EMERGENCY tube rupture
,a) Fuel cl ad integrity fail ure as ALL MODES indicated by any of the following:
- RCS specific activity greater than 300 ~Ci/gram dose equivalent 1-131 OR 5 or more core exit thermocouples GREATER THAN 1200 OF OR High Range Letdown radiation monitor 1-CH-RI-128 or 2-CH-RI-228 . //
GREATER THAN 5.9 x 10 4 mR/hr v b) Steam Generator tube rupture as indicated by both of the following:
- SI coincidence - SATISFIED AND
- Steam Generator tube rupture -IN/
PROGRESS c) Loss of secondary integrity associated with ruptured steam generator pathway as indicated by any of the following:
- Steam Generator PORV - OPEN ~
OR Main Steam Code Safety Valve - OPEN OR Loss of secondary coolant outside containment - IN PROGRESS
NUMBER ATTACHMENT TITLE REVISION EPIP-l.OI EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB B) pAGE REACTOR COOLANT SYSTEM EVENT 1
10 of 42 CONDITION/APPLICABILITY INDICATION CLASSI FI CATION
- 3. RCS leak rate exceeds
- Primary system leak (LOCA) SITE AREA makeup capacity - IN PROGRESS EMERGENCY MODES 1, 2. 3. & 4 AND
- Safety Injection - REQUIRED AND
- RCS subcooling based on Core Exit Thermocouples -
LESS THAN 30° F OR RCS inventory cannot be maintained based on pressurizer level or RVLIS indication
- 4. Gross primary to
- Steam Generator Tube SITE AREA secondary l~akage Rupture - IN PROGRESS EMERGENCY
( with loss of offsite .
power AND MODES 1. 2.3. & 4
- Safety Injection - REQUIRED AND
- Vent Vent A MGPI Monitor RM-VG-179 GREATER THAN 1.25 x 10 8 ~Ci/sec Steam Generator Slowdown monitor on affected pathway RM-SS-I22. -222 RM-SS-123. -223 RM-SS-124. -224 GREATER THAN lxl0 6 cpm AND
- A subsequent loss of ,-.
offsite power indicated by ;~~
zero volts on voltmeters for 4I60V buses D. E. & F
-':"' .*.. ,., .. '- ."." .'. ' .. ". ~;.-.\-.::":. ~.:. ~.~ ...
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (T AB B) PAGE REACTOR COOLANT SYSTEM EVENT 1
11 of 42 CONDITION/APPLICABILITY INDICATION CLASSIFICATION
- 5. RCS leak rate limit
- Pressurizer level cannot be ALERT
- EXCEEDED maintained greater than 20%
with one (1) Charging/SI MODES 1. 2. 3, & 4 pump in operation AND
- RCS inventory balance indicates l~akage - greater than 50 gpm
- 6. Gross primary to Steam Generator Tube Rupture ALERT secondary leakage I N PROGRESS MODES 1. 2. 3. & 4 AND Safety Injection - REQUIRED
- 7. Excessive primary to .
- Intentional reduction in ALERT secondary leakage power, load or tem~erature with loss of offsite lAW T.S. 3.4.13 prlmary-power to-secondary leakage LCO Action Statement MODES 1. 2. 3, & 4 AND
- Vent Vent A MGPI Monitor RM-VG-179 GREATER THAN 1.73 x 10 6 ~Ci/sec OR Steam Generator Blowdown
,monitor on affected pathway RM-SS-122,. -222 RM-SS-123. -223 RM-SS-124. -224 GREATER THAN lxl0 5 cpm AND
- A subsequent loss of offsite power indicated by zero volts on voltmeters for 4160V buses D. E. & F
,".'.'." '*.'.h". ~.". ,'. :~',._ , ..:." ..*' '. ;::'... ~
NUMBER ATTACHMENT TITLE REVISION EPIP-l.OI EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB B) PAGE REACTOR COOLANT SYSTEM EVENT 1
I? of 42 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 8. RCS operational Intentional reduction in power NOTI FICATION leakage requiring load or temperature lAW T.S. OF UNUSUAL plant shutdown 3.4.13 leakage limit action EVENT lAW 1.S.3.4.13 statement - HAS COMMENCED MODES 1. 2. 3. & 4
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB C) PAGE FUEL FAILURE OR FUEL HANDLING ACCIDENT 1
B of 4?
CONDITION/APPLICABILITY INDICATION CLASSIFICATION
- 1. Probable large
- Loss of reactor coolant in GENERAL radioactivity progress EMERGENCY release initiated by LOCA with ECCS AND failure leading to core degradation
- RCS specific activity -
greater than 300 ~Ci/gram ALL MODES dose equivalent 1-131 OR Containment High Range Radiation Monitor RM-RMS-165. -166 or RM-RMS-265. -266 GREATER THAN l.88xl0 2 R/hr AND
- High or low head ECCS flow not being delivered to the core (if expected by plant conditions)
CAUTION; EAL A.l is duplicated below for cross-reference/comparison to EAL C.2:
A.l. Loss of function
- Total loss of the SITE AREA needed for unit Charging/51 System EMERGENCY HSD conditi on OR MODES I, 2. 3 & 4 Total loss of the Main Feedwater and Auxiliary Feedwater systems
- 2. Probabl e 1arge Loss of Main Feedwater System. GENERAL radi oact i vity Condensate System and Auxiliary EMERGENCY release initiated by Feedwater System loss of heat sink leading to core degradation MODES 1. 2. 3 & 4
NUMBER ATTACHMENT TITLE REVISION EPIP-1.01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB C) PAGE FUEL FAILURE OR FUEL HANDLING ACCIDENT 1
14 of 42 CONDITION/APPLICABILITY INDICATION CLASSI FI CATION
- 3. Probable large
- Rx nuclear power after a GENERAL radioactivity trip - greater than 5% EMERGENCY release initiated by failure.of AND protection system to bring Rx subcritical
- RCS pressure greater than and causing core or equal to 2485 psig degradation OR ALL MODES Containment pressure and temperature rapidly increasing
- 4. Probabl e 1 arge
- Loss of all onsite and GENERAL radioactivity offsite AC power EMERGENCY release initiated by loss of AC power and AND all feedwater
( AND
- Restoration of either of the above not likely within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB C) PAGE FUEL FAILURE OR FUEL HANDLING ACCIDENT 1
15 of 42 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 5. Probable large
- Loss of reactor coolant in GENERAL radioactivity progress EMERGENCY release initiated by LOCA with loss of AND ECCS and containment cooling
- High or low head ECCS flow not being delivered to the ALL MODES core (if expected by plant conditions)
AND
- Containment RS sump temperature greater than 190 0 F and NOT decreasing OR All Quench Spray and Recirculation Spray systems
- NOT OPERABLE
NUMBER ATTACHMENT TITLE REVISION EPIP-L01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB C) PAGE FUEL FAILURE OR FUEL HANDLING ACCIDENT 1
Hi nf 42 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 6. Core damage with a) Fuel clad failure as SITE AREA possible loss of indicated by any of the EMERGENCY cool able geometry following:
MODES 1, 2, 3. & 4
- ReS Specific activity greater than 60 llCi/gram dose equivalent I-131 OR High Range Letdown radiation monitor l-CH-RI-128 or 2-CH-RI-228 GREATER THAN 1.2xl0 4 mR/hr AND b) Loss of cooling as indicated by any of the
( following:
- 5 confirmed core exit thermocouples greater than 1200 F 0 OR Core delta T - zero OR Core delta T ~ rapidly diverging
,,:~.' ;.. '. :". -' .... ; ...', , ,~ .. :.~,~ .. :.::.. \:;,.; ~., .. ';.,~ ;.'::,:, ~4~,. "~" _', .'~.' ';._ ..* ,.~' .
'_~.'\; ~.'- .._.. :.. -._:.. ~: .._.~ :.*.
NUMBER ATTACHMENT.TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB C) PAGE FUEL FAILURE OR FUEL HANDLING ACCIDENT 1
17 of 4/
CONDITION/APPLICABILITY INDICATION CLASSI FI CATI ON
- 7. Major fuel damage
- Water 1evel in Rx vessel SITE AREA accident with during refueling below the EMERGENCY radioactivity top of core release to containment or fuel OR buildings Water level in spent fuel ALL MODES pool below top of spent fuel AND
- Verified damage to i rradi ated fuel resul ti ng in readings on Vent Vent "B" MGPI monitor RM-VG-IBO GREATER THAN 2.69 x lOB ~Ci/sec I
- 8. Severe Fuel Clad
- High Range Letdown ALERT Damage radiation monitor MODES* 1. 2. 3. & 4 l-CH-RI-128 or 2-CH-RI-228 Increases to GREATER THAN Hi Hi Alarm setpoint (representing 1% fuel failure) within 30 minutes and remains for at least 15 minutes OR
- RCS specific activity -
greater than 300 ~Ci/gram dose equivalent 1-131
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB C) PAGE FUEL FAILURE OR FUEL HANDLING ACCIDENT 1
lR nf 4/
CONDITION/APPLICABILITY INDICATION CLASSIFICATION
- 9. Fuel damage accident
- Verified accident involving ALERT with rel ease of damage to irradiated fuel radioactivity to containment or fuel AND buildings
- Health Physics confirms fission ALL MODES product release from fuel OR Vent Vent "B" MGPI monitor RM-VG-180 GREATER THAN 1.99 x 10 6 ~Ci/sec I
- 10. Potential for fuel Continuing uncontrolled ALERT damage to occur decrease of water level in during refueling Reactor Refueling Cavity or Spent Fuel Pool MODE 6
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB C) PAGE FUEL FAILURE OR FUEL HANDLING ACCIDENT 1
lCJ of 4/
CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 11. Fuel clad damage
- Intentional reduction in NOTI FICATION indication power, load or temperature OF UNUSUAL lAW reactor coolant EVENT MODES 1, 2, 3, & 4 activity T.S. Action Statement - HAS COMMENCED OR High Range Letdown.
radiation monitor l-CH-RI-128 or 2-CH-RI-228 Increases to GREATER THAN Hi Alarm setpoint (representing 0.1%
fuel failure) within 30 minutes and remains for for at least 15 minutes
- 12. Independent Spent
- Verified Sealed Surface Non FICATION Fuel Storage Storage Cask (SSSC) seal OF UNUSUAL Installation. leakage EVENT (lSFSl) event OR ALL MODES Sealed Surface Storage Cask (SSSC) dropped or mishandled
NUMBER ATTACHMENT TITLE REVISION EPI P*-1. 01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB D) PAGE CONTAINMENT EVENT 1
/0 of 42 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 1. Extremely high
- Containment High Range GENERAL containment radiation monitor EMERGENCY radiation. pressure and temperature RM-RMS-165. -166 or RM-RMS-265. -266 MODES 1. 2. 3. & 4 GREATER THAN 3.76 x 10 2 R/hr AND
- Containment pressure greater than 45 psia and not decreasing OR Containment temperature greater than 2800F
- 2. High-high
- Containment High Range SITE AREA containment radiation monitor EMERGENCY
( radiation. pressure.
~, .. ,
and temperature RM- RMS-165. -166 or RM-RMS-265. -266 MODES 1. 2. 3. & 4 GREATER THAN
- 1. 88 x 10 2 R/hr AND
- Containment pressure -
greater than 27.75 psia and not decreasing OR Containment temperature greater than 200 OF
NUMBER ATTACHMENT TITLE REVISION EPIP-1.01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB D) PAGE
- CONTAINMENT EVENT 1
21 of 42 .
CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 3. High Containment
- Containment High Range ALERT radiation. pressure radiation monitor and temperature RM-RMS-165. -166 or MODES 1; 2. 3. & 4 RM-RMS-265. -266 GREATER THAN
- 81. 5 R/hr
- Containment pressure greater than 17 psia OR Containment temperature -
greater than 1500F
,_, '", '", ", ". \'.'. , ,.,. ':', ..,.~.~ ~.~._. "'_, _, ~ *** _, :-**,-.-.."' ,., "7. f",'".-:' """'.,." *""t"",,,.,,-.,. ~.~. ',' , ,.',_,',.,"..'- ,..,~._~. t'**" ._ *** ' , '.>** ',,'Y __: ******* ~._\ *** ~._ ~ ",. ,' ",'."". _.' '-' ** " **.* "'.* -.'.. ~',': ~ ':." v ,".' .\,~.:".,.,., or', ',"' , r. ~ ** < **** '<" . , , " ' , " ,. ',"":, .r'***** : . : ~ , , * ".' * *':". " . ' ' . ' ,,,,.,, :. , . , , " .
NUMBER ATTACHMENT TITLE REVISION EPI P-l. 01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB E) PAGE RADIOACTIVITY EVENT 1
22 of 42 CONDITION/APPLICABILITY INDICATION CLASSI FI CATION
- 1. Release imminent or
- HP assessment indicates GENERAL in progress and site actual or projected doses EMERGENCY boundary doses at or beyond site boundary projected to exceed greater than 1.0 Rem TEDE 1.0 Rem TEDE or 5.0 or 5.0 Rem Thyroid CDE Rem Thyroid CDE ALL MODES
- 2. Release imminent or
- HP assessment indicates SITE AREA in progress and site actual or projected dose at EMERGENCY boundary doses or beyond Site Boundary projected to exceed exceeds 0.1 Rem TEDE or 0.5 0.1 Rem TEDE or 0.5 Rem Thyroid CDE Rem Thyroid COE ALL MODES
- v.* " ~ ,.' , , _~'**'.'" .;' *. : .~.) ,. ,. , ,.'.,:..',"'" , ,,"~ *. " ~ *.,:_ *...*..*..*. ' .. ': '-. ,'. ~.: ,*r. ".' ".';"
NUMBER ATTACHMENT TITLE REVISION EPI P-l, 01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB E) PAGE RADIOACTIVITY EVENT 1
23 of 42 CONDITION/APPLICABILITY INDICATION CLASSIFICATION
- 3. Effluent release a) Any of the foll~wing ALERT greater than 10 monitors indicate valid times ODCM allowable readings above the 1 imit specified values for greater than 15 minutes ALL MODES
- Clarifier Effluent RM-LW-111 GREATER THAN 4.8 x 10 5 cpm .
- Discharge Canal RM-SW-130 or -230 GREATER THAN 5 x 10 4 cpm
- Vent Vent A MGPI RM-VG-179 GREATER THAN 1.73 x 10 6 ~Ci/sec
- Vent Vent B MGPI RM-VG-180 GREATER THAN 1.99 x 10 6 ~Ci/sec
- Process Vent MGPI RM-GW-178 GREATER THAN 1.35 x 10 7 ~Ci/sec OR b) HP assessment (sample results or dose projections) indicate greater than 10 times ODCM allowable limit
..* ' O'!:~.:. ':.' ..Ol:* ".....:: '.' ..... ' ';. '. - , , .... ,-',.:"'., * '. '- ..... " ._. , ., *'-. ..... ~, .,:.,",: \.~.. ~~.' 0~*O:':':":>:_-'_.:.l. .* -~';.:~.' **:.~..:;*.:.::: ::.. ;.:.. :.: ~;:;; :.;; ~:: ':-.:. .~ . ; .:-. ~ ~: .... .:'; .*
NUMBER ATTACHMENT TITLE REVISION EPIP-l.01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB E) PAGE RADIOACTIVITY EVENT 1
?4 nf 4?
CONDITION/APPLICABILITY INDICATION CLASSI FI CATION
- 4. High radiation or Valid readings on any of the ALERT airborne following monitors have contamination levels . increased by a factor of 1000 indicate a severe and remain for .at least 15 degradation in minutes:
control of radioactive material
- Ventilation Vent Multi-sample gaseous or ALL MODES *particulate monitor.
I RM-VG-106 or -105
- Control Room Area I RMS-157
- Aux. Bldg. Control Area I RMS-154
- Decon. Bldg. Area I RMS-151
- Fuel Pool Bridge Area I RMS-153
- New fuel storage Area I RMS-152
- Laboratory Area I RMS-158
- Sample Room Area I RMS-156
NUMBER ATTACHMENT TITLE REVISION EPIP-l.01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB E) PAGE RADIOACTIVITY EVENT 1
2!'i of 42 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 5. Effluent release a) Any of the following NOn FICATION greater than aDCM monitors indicate valid OF UNUSUAL allowable limit readings above the EVENT sgecified value for more ALL MODES t an 1 hour:
- Clarifier Effluent RM-LW-11l GREATER THAN 4.8 x 10 cpm
- Discharge Canal RM-SW-130 or -230 GREATER THAN 5 x 10 3 cpm
- Vent Vent A MGPI RM-VG-179 GREATER THAN 1.73 x 10 5 ~Ci/sec
- Vent Vent B MGPI RM-VG-180 GREATER THAN 1.99 x 105 ~Ci/sec
- Process Vent MGPI RM-GW-178 GREATER THAN 1.35 x 10 6 ~Ci/sec OR b) HP assessment (sample results or dose projections) indicates greater than ODCM allowable limit
NUMBER ATTACHMENT TITLE REVISION EPIP-1.01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB G) PAGE LOSS OF SECONDARY COOLANT 1
?f1 nf 4?
CONDITION/ APPLICABI LITY INDICATION CLASSIFICATION
- 1. Major secondary Conditions a) and b) exist with c): SITE AREA 1 i ne break with a) Uncontrolled loss of secondary EMERGENCY significant primary coolant - IN PROGRESS to secondary leakage and fuel AND damage indicated b) RCS specific activity exceeds MODES 1, 2. 3. & 4 limits of T.S. Figure 3.4.16-1 OR High Range Letdown radiation monitor l-CH-RI-128 or 2-CH-RI-228 GREATER THAN Hi Alarm setpoint AND c) Vent Vent A MGPI Monitor RM-VG-179 GREATER THAN 6.21 x 107 ~Ci/sec OR Affected pathway Steam Generator Blowdown monitor RM-SS-122. -123. -124.
-222. -223. -224 GREATER THAN 1 x 10 6 cpm OR Affected pathway Main Steam Line High Range monitor RM-MS-170, -171. -172,
-270, -271, -272 GREATER THAN 12.2 mR/hr
NUMBER ATTACHMENT TITLE REVISION EPIP-l.01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB G) PAGE LOSS OF SECONDARY COOLANT 1
27 of 42 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 2. Major secondary line
- Uncontrolled loss of ALERT break with secondary coolant - IN significant primary PROGRESS to secondary leakage AND MODES 1. 2. 3. & 4
- Vent Vent A MGPI Monitor RM-VG-179 GREATER THAN 1.76 x 10 6 ~Ci/sec OR Steam Generator Blowdown monitor on affected pathway RM-SS-122. -123. -124 RM-SS-222. -223. -224 GREATER THAN 1x10 5 cpm OR Main Steam Line High Range monitor on affected pathway RM-MS-170. -171, -172 RM-MS-270. -271. -272 GREATER THAN 0.14 mR/hr
- 3. Major secondary line Uncontrolled loss of secondary Non FICATION break coolant - IN PROGRESS OF UNUSUAL EVENT MODES 1. 2. 3. & 4
',,-,' ** " '~". ~ ~ _ '. " .** ~ ~-"~' *. " .. :": ~.; , -,.., ;., *. ,._- , . ~.<.::.:, .. :.(, ~ ':;., ..*..-.. ~.-' -.." :.' : ,..*_ - .,:-,._~.~ .
NUMBER ATTACHMENT TITLE REVISION EPIP-I.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB H) PAGE ELECTRICAL FAILURE 1
28 of 42 CONDITION/APPLICABilITY INDICATION CLASSI FICATION I. Loss of offsite and The following conditions exist SITE AREA onsite AC power for for greater than 15 minutes: EMERGENCY more than 15 minutes
- Ammeters for 4160V Reserve ALL MODES Station Service Buses D, E,.
& F all indicate - zero (0) amps AND
- Ammeters for 4160V Station Service Buses A, B, & Call indicate - zero (0) amps AND
- Ammeters for 4160V Emergency Buses H & J both indicate - zero (0) amps
- 2. Loss of all onsite The following conditions exist SITE AREA DC power for greater for greater than 15 minutes: EMERGENCY than 15 minutes
- All station battery ALL MODES voltmeters indicate zero CO) volts AND
- No light indication available to Reserve Station Service breakers 1501, 15El and 15Fl
. 'q,'.' .-.. , , " .. '~ , ..- .. ~ " ~ .. ~".~..,.* '.,,~ .. ; ',,; ...* ' ~.~ *.* ;~;"" *. :'.'.::':.:"'~
NUMBER ATTACHMENT TITLE REVISION EPIP-1.0I EMERGENCY ACTION LEVEL TABLE 43.
ATTACHMENT (TAB H) PAGE ELECTRICAL FAILURE 1
2g of 4?
CONDITION/APPLICABILITY INDICATION CLASSIFICATION CAUTION: EAL A.l is duplicated below for cross-reference/comparison to EAL H.3:
A.l. Loss of function
- Total loss of the SITE AREA needed for unit Charging/SI System EMERGENCY HSD condition OR MODES 1, 2, 3 & 4 Total loss of the Main Feedwater and Auxiliary Feedwater Systems
- 3. Loss of all offsite
- Ammeters for 4I60V Reserve ALERT and onsite AC power Station Service Buses D. E.
& F all indicate - zero (0)
ALL MODES amps AND
- Ammeters for 4I60V Station Service Buses A. B. & Call indicate - zero (0) amps
. AND
- Ammeters for 4160V Emergency Buses Hand J both indicate - zero (0) amps
- 4. Loss of all onsite
- All stati on battery ALERT DC power voltmeters indicate - zero (0) volts ALL MODES AND
- No light indication available to Reserve Station Service Breakers 1501. 15E1 and 15F1
NUMBER ATTACHMENT TITLE REVISION EPI P-l. 01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB H) PAGE ELECTRICAL FAILURE 1
30 of 42 CONDITION/APPLICABILITY INDICATION CLASSI FI CATI ON
- 5. Loss of offsite
- Unit main generator and NOTI FICATION power or both emergency diesel oF" UNUSUAL onsite AC power generators out of service EVENT capabil ity.
OR ALL MODES Loss of all 34.5 KV reserve station service buses
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB 1) PAGE FIRE 1
11 of 4?
CONDITION/APPLICABILITY INDICATION CLASS! FI CATI ON
- 1. Fire resulting in
- Fire which causes major SITE AREA degradation of degradation of a safety EMERGENCY safety systems system function required for protection of the MODES 1. 2, 3. & 4 public AND
- Affected systems are caused to be NOT operable as defined by Tech. Specs.
- 2. Fire potentially Fire which has potehtial for ALERT affecting station causing a safety system not to safety systems be operable as defined by Tech.
Specs.
MODES 1. 2. 3. & 4
- 3. Fire lasting greater Fire within the Protected Area NOTI FICATION than 10 minutes in or Service Water Pump/Valve* OF UNUSUAL Protected Area or House which is not under EVENT Service Water control within 10 minutes after Pump/Valve House Fire Brigade - DISPATCHED ALL MODES
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB J) PAGE SECURITY EVENT 1
- i? of 4?
CONDITION/APPLICABILITY INDICATION CLASS! FICATION
- l. Loss of physical A hostile force has taken GENERAL control of the control of plant equipment EMERGENCY facil i ty such that plant personnel are unable to operate equipment ALL MODES required to maintain safety functions
- 2. Imminent loss of
- A confirmed security event SITE AREA physi cal control within a plant Vital Area EMERGENCY of the plant OR ALL MODES A notification from the site security force that an armed attack. explosive attack. airliner impact.
or other.hostile action is occurring or has occurred within the Protected Area
- 3. Ongoing Security
- A confirmed security event ALERT compromise within the Protected Area ALL MODES OR A validated notification from NRC of an airliner attack threat less than 30 minutes away OR A notification from the site security force of an armed attack. explosive attack. airliner impact. or other hostile action within the Owner Controlled Area
NUMBER ATTACHMENT TITLE REVISION EPIP-1.0! EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB J) PAGE SECURITY EVENT 1
- n nf 4?
CONDITION/APPLICABILITY INDICATION CLASSIFICATION
- 4. Security threat.
- A credible site-specific NOlI FI CATI ON unauthorized security threat OF UNUSUAL attempted entry. or notification EVENT attempted sabotage OR ALL MODES A validated notification from NRC providing information of an aircraft threat OR A confirmed security event which indicates a potential degradation in the level of safety of the plant such as a violent civil disturbance or strike action. attempted sabotage, a hostage/extortion situation. or attempted intrusion in the Protected Area
NUMBER ATTACHMENT TITLE REVISION EPIP-l.OI EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB K) PAGE HAZARD TO STATION OPERATION 1
- i4 of 4?
CONDITION/APPLICABILITY INDICATION CLASSI FI CATI ON
- 1. Aircraft damage to Aircraft crash which affects SITE AREA vital pl ant systems vital structures by impact or EMERGENCY fire MODES 1, 2, 3. & 4
- 2. Severe explosive Explosion which results in SIn AREA damage severe degradation of any of EMERGENCY th following systems required MODES 1, 2, 3, & 4 for safe shutdown:
- 3. Entry of toxic or
- Uncontrolled release of SITE AREA flammable gases into toxic or flammable agents EMERGENCY plant vital areas greater than life other than the threatening or explosive Control Room limits in Vital Areas MODES 1, 2, 3, & 4 AND
- Evacuation of Vital Area other than Control Room -
REQUIRED OR Significant degradation of plant safety systems resulting in loss of*a safety system function required for protection of the public
....'. " ....*...,:.:...:.... ' .. .:.~~.: ......:.~.; *. ~ ...:..... ::.r.' .',' :.'.' : ., i*; ~' ..*. ,., . ' :." .:-: ,: ',': ;" .... '..:.: ::.. :.::....... :'. :.' ;....... :. " .. ... : ..*.. /.:.. ,.,., .*.... ".' ::., ':'. '.',.:...;'..<>,. f . .* :'.~ c **** ' .'.: .. , . '* * ' . , ., ****** ,'.'"
NUMBER ATTACHMENT TITLE REVISION EPIP-1.01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB K) PAGE HAZARD TO STATION OPERATION 1
111 r1f 4?
CONDITION/APPLICABILITY INDICATION CLASSI FI CATION
- 4. Severe missile Missile impact causing severe SITE AREA damage to safety degradation of safety systems EMERGENCY systems required for unit shutdown MODES 1. 2. 3. & 4
- 5. Aircraft crash on Aircraft crash within the ALERT the facil i ty Protected Area or Switchyard (other than ALL MODES impact from airliner attack - See TAB J)
- 6. Explosion damage to Unplanned explosion resulting ALERT facil ity in damage to plant structure or equipment that affects plant ALL MODES operations
- 7. Entry of toxic or Notification of uncontrolled ALERT flammable gases or release of toxic or flammable liquids into plant agent which causes:
facility
- Evacuation of personnel ALL MODES from* plant areas AND
- Safety related equipment is rendered inoperable*
- 8. Turbine failure or Failure of turbine/generator ALERT missile impact rotating equipment resulting in casing penetration MODES 1 & 2
.~.: .. ;. ;..'; .:: ,~ ...'.' ,;":.. \".";.:,(.~' . :..>:. ':..'--:':,;~ ';..;'~':.:';::,:'.:;.......:::..~.:.:;~:..::.:::\ ';.'.:.:~ ;'::'; ;';':'.' ~.,
NUMBER ATTACHMENT TITLE REVISION EPIP-l.01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB K) PAGE HAZARD TO STATION OPERATION 1
36 of 42 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 9. Missile damage to Notification of missile impact ALERT safety related causing damage to safety equipment or related equipment or structures structures MODES 1. 2. 3, & 4
- 10. Aircraft crash or unusual ai rcraft
.. Confirmed notification of an aircraft crash within NOTI FICATION OF UNUSUAL acti vity the site boundary (other EVENT than impact from airliner ALL MODES attack - See TAB J)
OR Unusual aircraft activity in the 'vicinity of the site as determined by the Operations Shift Manager/
Station Emergency Manager or the Security Shift Supervisor
- 11. Train derailment Confirmed report of train Non FICATION withi n Protected derailment within Protected OF UNUSUAL Area Area EVENT ALL MODES
- 12. Explosion within Confirmed report of unplanned NOTI FICATION Protect-ed Area explosion within Protected OF UNUSUAL Area EVENT ALL MODES
- 13. Onsite ornearsite Notification of unplanned NOTIFICATION release of toxic or release of toxic or flammable OF UNUSUAL flammable liquids or agents which may affect safety EVENT gases of station personnel or equipment ALL MODES
NUMBER ATTACHMENT TITLE REVISION EPIP-I. 01 EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB K) PAGE HAZARD TO STATION OPERATIO~
1 17 of 4?
CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- 14. Turbine rotating Failure of turbine/generator NOTIFICATION c9mponent f~ilure rotating equipment resulting in OF UNUSUAL wlth no caslng immediate unit shutdown EVENT penetration MODES 1 & 2
(
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB l) PAGE NATURAL EVENTS 1
- lR of 47 CONDITION/APPLICABILITY INDICATION CLASSI FICATION
- l. Earthquake greater
- Confirmed earthquake SITE AREA than or equal to DBE which activates the Event EMERGENCY levels Indicator on the Strong Motion Accelerograph MODES 1. 2. 3. & 4 AND
- Alarms on the Peak Shock Annunciator indicate a horizontal motion of greater than or equal to 0.12 9 or a vertical motion of greater than or equal to 0.08g
- 2. Sustained winds in Sustained winds 150 mph SITE AREA excess of design OR GREATER experienced EMERGENCY levels experienced or projected or projected MODES 1. 2. 3. & 4
- 3. NOT USED
NUMBER ATTACHMENT TITlE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB l) PAGE NATURAL EVENTS 1
39 of 42 CONDITION/APPLICABILITY INDICATION CLASS I FI CATION
- 4. Earthquake greater
- Confirmed earthquake which ALERT than or equal to OBE activates Event Indicator 1 evel s on the Strong Motion Accelerograph ALL MODES AND
- Alarms on the Peak Shock Annunciator indicate a horizontal motion of greater than or equal to 0.06 g or a vertical motion of greater than or equal to 0.04g
- 5. Tornado striking Tornado visually detected ALERT facil ity striking structures within the Protected Area or Switchyard ALL MODES
- 6. Hurricane winds Hurricane winds 120 mph ALERT near design basis OR GREATER experienced level experienced or projected .
or projected ALL MODES
- 7. Flood near design Flood in the Lake Anna ALERT levels Reservoir with indicated level -
greater than 263 feet MSL ALL MODES
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB L) PAGE NATURAL EVENTS 1
40 of 47 CONDITION/APPLICABILITY INDICATION CLASSI FI CATION
- 8. Earthquake detected Confirmed earthquake which NOTI FICATION activates the Event Indicator OF UNUSUAL ALL MODES on the Strong Motion EVENT Accelerograph
- 9. Tornado within Tornado visually detected NOTI FICATION Protected Area or within Protected Area or OF UNUSUAL Switchyard Switchyard EVENT ALL MODES
. 10. Hurricane force
- Confirmation by Weather NOTI FICATION winds projected Center that hurricane force OF UNUSUAL onsite within 12 winds (greater than 73 mph) EVENT hours ~rOjected onsite within 12 ours ALL MODES 1l. 50 year flood Flood in the Lake Anna NOTI FICATION Reservoir with indicated level - OF UNUSUAL ALL MODES greater than 254 feet MSL EVENT
NUMBER ATTACHMENT TITLE REVISION EPIP-1.0l EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB M) PAGE MISCELLANEOUS ABNORMAL EVENTS 1
41 of 4?
CONDITION/APPLICABILITY INDICATION CLASSIFICATION
- 1. Any major internal Shift Manager/Station Emergency GENERAL or external events Manager judgement EMERGENCY which singly or in combination cause massive damage to station facilities or may warrant evacuation of the public ALL MODES
- 2. Station conditions Shift Manager/Station Emergency SITE AREA which may warrant Manager judgement EMERGENCY notification of the public near the site ALL MODES
- 3. Station conditions Shift Manager/Station Emergency ALERT which have the Manager judgement potential to degrade or are actually degrading the level of safety of the station ALL MODES
.. ~ -: ;:. ...
- . :.\ ' ": ".' ::_' "'~'.:.:-:-'
NUMBER ATTACHMENT TITLE REVISION EPIP-l.O! EMERGENCY ACTION LEVEL TABLE 43 ATTACHMENT (TAB M) PASE MISCELLANEOUS ABNORMAL EVENTS 1
42 of 4?
CONDITION/APPLICABILITY INDICATION CLASSIFICATION
- 4. Station conditions Shift Manager/Station Emergency NOTI FICATION which warrant Manager judgement that any of OF UNUSUAL increased awareness the following exist: EVENT of state and/or local authorities
- Unit shutdown is other than ALL MODES a controlled shutdown OR Unit is in an uncontrolled condition during operation OR A condition exists which has the potential for escalation and therefore warrants notification
"'j
NUMBER ATTACHMENT TITLE REVISION EPIP-1.01 TURNOVER CHECKLIST 43 ATTACHMENT PAGE 2 1 of 1 Conduct a turnover between the onshift and relief SEM in accordance with the following checklist. Use placekeeping aid at left of item.>> -. to track completion.
- 1. Determine the status of primary responder notification.
- 2. Determine the status of -Report of Emergency to State and Local Governments." EPIP-2.01. Attachment 2. Get completed copies if available.
- 3. Determine status of the "Report of Radiological Conditions to the State.- EPIP-2.01. Attachment 3. Get completed copy if available.
- 4. Determine status of Emergency Notification System (ENS) communications and completion status of NRC Event Notification Worksheet (EPIP-2.02 Attachment 1).
- 5. Review classification and initial PAR status.
- 6. Review present plant conditions and status. Get copy of Critical Safety Functions form.
- 7. Review status of station firewatches and re-establish if conditions allow.
- 8. Determine readiness of TSC for activation.
- 9. After all information is obtained. transfer location to TSC.
1£ the TSC is functional. THEN the State and Local Communicator in the Control Room will relocate to TSC with the SEM.
1£ the TSC is NOT functional. THEN the responsibilities may be transferred to relief in another facility. e.g. LEOF/CEOF.
- 10. Call the Control Room and assess any changes that may have occurred during transition to the TSC.
- 11. When sufficient personnel are available. the relief SEM is to assume the following responsibilities from the onshift Station Emergency Manager:
- a. Reclassification.
- b. Protective Action Recommendations until LEOF activated.
- c. Notifications (i.e .. state. local. & NRC). Upon LEOF activation.
transfer notification responsibilities except for the NRC ENS.
- d. Site evacuation authorization.
- e. Emergency exposure authorization.
- f. Command/control of onsite response.
- 12. Formally relieve the Interim SEM and assume control in the TSC.
Announce name and facility activation status to facility.
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NUMBER ATTACHMENT TITLE REVISION EPIP-l.OI CONSIDERATIONS FOR OPERATIONS RESPONSE 43 ATTACHMENT UNDER ABNORMAL CONDITIONS PAGE 3 1 of 2 This attachment provides procedural guidance for controlling selected emergency response actions when their implementation would have adverse results. Station Emergency Manager (SEM) approval is required before any required action is postponed. suspended or modified. The guidance below is not all-inclusive.
UNANTICIPATED HAZARD EXISTS (e.g., security event. tornado or toxic release):
IE notifying off-duty augmentation could create a safety hazard for personnel coming to the station, THEN consider the following alternatives:
- Station Security (if available) can be directed to notify off-duty personnel to report to the remote mustering area (Louisa Fire Training Center).
- Corporate Security, at 804-273-3161. can be directed to notify off-duty personnel to report to the remote mustering area (Louisa Fire Training Center).
- Corporate Security, at 804-273-3161, can be directed to notify corporate emergency response organization only using CPIP-3.4, INNSBROOK SECURITY SUPPORT.
- Notifications can be deferred until hazardous conditions are resolved.
IE implementation of emergency response actions could compromise Security Plan response strategies. THEN consider postponing or suspending emergency response actions until threat has been resolved. e.g., on-site announcement directing assembly and emergency response facility activation, pager activation and call-out per EPIP-3.05, AUGMENTATION OF EMERGENCY RESPONSE ORGANIZATION. dispatch of Security Team members to the LEOF per EPIP-3.04, ACTIVATION OF LOCAL EMERGENCY OPERATIONS FACILITY, and staging of road blocks* per EPIP-5.04, ACCESS CONTROL.
IE assembling on-site personnel for accountability or activation of emergency response facilities could endanger plant personnel, THEN consider postponing emergency assembly until hazardous conditions are resolved. Corporate Security, at 804-273-3161, can be directed to notify corporate emergency response organization only using CPIP-3.4, INNSBROOK SECURITY SUPPORT. Personnel in unaffected areas on-site can be notified selectively.
1£ primary ingress/egress route is NOT available, THEN evaluate alternate route for use during site evacuation or off-duty augmentation (e.g., access via Dyke 1).
NUMBER ATTACHMENT TITLE REVISION EPIP-l.Ol CONSIDERATIONS FOR OPERATIONS RESPONSE 43 ATTACHMENT UNDER ABNORMAL CONDITIONS PAGE 3 2 of 2 ANTICIPATED SITUATION (e.g .. forecasted severe weather or grid disturbance):
IE all or part of the ERO has been staged in anticipation of a predicted event. THEN notify Security to omit performance of augmentation notification (as described in EPIP-3.05. AUGMENTATION OF EMERGENCY RESPONSE ORGANIZATION).
IE adequate controls have been established to continually account for personnel staged in anticipation of a predicted event. THEN notify Security to omit performance of initial accountability (as described in EPIP-5.03. PERSONNEL ACCOUNTABILITY).
II a decision has been made to staff the Central EOF in lieu of the LEOF. THEN notify Security that performance of EPIP-3.04. ACTIVATION OF LOCAL EMERGENCY OPERATIONS FACILITY. is not required.
IF environmental conditions are hazardous, THEN consult with Security Team Leader about suspending procedural requirements for staging road blocks (lAW EPIP-5.04. ACCESS CONTROL).
~~
riiiJJ~* Dominion NORTH ANNA POWER STATION EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER PROCEDURE TITLE REVISION 7
EPIP-1.06 PROTECTIVE ACTION RECOMMENDATIONS (WITH 5 ATTACHMENTS) PAGE 1 of 4 PURPOSE Give guidance to the Station Emergency Manager or Recovery Manager regarding determination of Protective Action Recommendations.
ENTRY CONDITIONS Anyone of the following:
- 1) Activation by EPIP-1.05, RESPONSE TO GENERAL EMERGENCY.
- 2) Activation by CPIP-3.1, CERC AND CEOF ACTIVATION.
- 3) Activation by CPIP-6.0, LEOF RECOVERY MANAGER GUIDANCE.
- 4) As directed by the Station Emergency Manager or Recovery Manager.
REFERENCE USE
NUMBER PROCEDURE TITLE REVISION 7
EPIP-1.06 PROTECTIVE ACTION RECOMMENDATIONS PAGE 2of4 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1 INITIATE PROCEDURE:
- By: _
Date: __
Time: _
2 USE ATTACHMENT 2, PROTECTIVE ACTION RECOMMENDATION MATRIX, TO DETERMINE INITIAL PAR 3 COMPLETE ATTACHMENT 3, PROTECTIVE ACTION RECOMMENDATION FORM:
o a) Fill in Item 1 Meteorological Data (obtained from EPIP-2.01, Notification of State and Local Governments, Attachment 2, page 1 of 3) o b) Determine Downwind Sectors using ATTACHMENT 4, DOWNWIND SECTORS TABLE c) Fill in Item 2 Protective Action Recommendation:
o 1) Mark appropriate PAR box(s) o 2) Record Mile radius and Miles downwind o 3) Record Downwind Sectors o d) Sign and date form
NUMBER PROCEDURE TITLE REVISION 7
EPIP-1.06 PROTECTIVE ACTION RECOMMENDATIONS PAGE 3 of 4 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED 4 DIRECT EMERGENCY COMMUNICATORS TO NOTIFY OFF-SITE AUTHORITIES OF PAR:
o . State Emergency Operations Center notified lAW EPIP-2.01, NOTIFICATION OF STATE AND LOCAL GOVERNMENTS o . NRC notified lAW EPIP-2.02, NOTIFICATION OF NRC (notification made from Control Room or TSC, when activated) o . NRC notified lAW EPIP-4.33, HEALTH PHYSICS NETWORK COMMUNICATIONS (notifications made from TSC or LEOF/CEOF only after NRC requests HPN be established) 5 HAVE RADIOLOGICAL ASSESSMENT DIRECTOR (RAD) IMPLEMENT EPIP-4.07, PROTECTIVE MEASURES [RADIOLOGICAL ASSESSMENT COORDINATOR (RAC) IF IN LEOF]
6 CHECK IF RADIOLOGICAL-BASED PAR o !E. PAR in effect - UNCHANGED, THEN RECOMMENDS EITHER OF THE FOLLOWING: GO TO Step 8.
o . Protective actions in any new area(s) o . Implementation of (Commonwealth of Virginia) Potassium Iodide (KI) strategies for the general public 7 RETURN TO STEP 3 8 CHECK EMERGENCY - TERMINATED o IF RAD/RAC recommends a PAR change, THEN RETURN TO Step 6.
NUMBER PROCEDURE TITLE REVISION 7
EPIP-1.06 PROTECTIVE ACTION RECOMMENDATIONS PAGE 40f4 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 9 TERMINATE EPIP-1.06:
o . Give completed EPIP-1.06, forms, and other applicable records to TSC Emergency Procedures Coordinator or LEOF Services Coordinator o . Completed by: _
Date: _
Time: _
-END-
NUMBER ATTACHMENT TITLE ATTACHMENT EPIP-1.06 1 SECTOR MAP REVISION PAGE 7 1 of 1 Graphics No. SV634A
NUMBER ATTACHMENT TITLE ATTACHMENT EPIP-1.06 2 PROTECTIVE ACTION RECOMMENDATION MATRIX REVISION PAGE NAPS 7 1 of 1
- 1. Start with the first statement at top, left side of table below (Refer to Attachment 5, Reference Information, for additional information as needed).
- 2. Go to the COL # that provides the response to this statement (YES or NO).
- 3. !.E no other responses are required in that column, THEN use the PAR at the bottom of the column.
IF another response is required in that column, THEN go to next statement at left side of table below AND RETURN TO Step 2 of this attachment.
Known impediments make evacuation dangerous Dose assessment results -
AVAILABLE Dose at or beyond 2 miles either:
- ~ 5 Rem Thyroid CDE Dose at or beyond Site Boundary ~ 5 Rem Thyroid CDE Dose at or beyond Site Boundary ~1 Rem TEDE Release controlled or terminated PAR A Shelter-in-place: 2 Mile radius and 5 Miles downwind.
PAR B Evacuate: 2 Mile radius and 5 Miles downwind.
PAR C Expanded Par: Derive from EPIP-4.07, PROTECTIVE MEASURES.
PAR D Evacuate: 2 Mile radius and 5 Miles downwind and recommend implementation of Potassium Iodide (KI) strategies for the general public. The projected dose at the site boundary is ~ 5 Rem Thyroid CDE.
NUMBER ATTACHMENT TITLE ATTACHMENT EPIP-1.06 3 PROTECTIVE ACTION RECOMMENDATION FORM REVISION PAGE 7 1 of 1 NOTE:
- Wind Direction and Wind Speed is obtained from EPIP-2.01. Notification of State and Local Governments, Attachment 2, Report of Emergency to State and Local Governments, page 1 of 3, transmitting this PAR.
- 1. METEOROLOGICAL DATA:
Wind Direction (degrees from) ; Wind Speed mph
- 2. PROTECTIVE ACTiON RECOMMENDATION:
[ ] SHELTER-iN-PLACE: Mile radius (360 0
) and Miles downwind in the following sectors: _
[ ] EVACUATE: Mile radius (360 0
) and __ Miles downwind in the following sectors:
[ ] BEYOND 10 MILE EPZ: (Data obtained from Radiological Assessment Coordinator (RAC)
[derived from EPIP-4.07, PROTECTIVE MEASURES])
[ ] Evacuate Area: Centerline in degrees; Distance in Miles; Width in feet
[ ] Shelter-in-place: Centerline in degrees; Distance in Miles; Width in feet
[ ] POTASSIUM IODIDE:
Recommend implementation of Potassium Iodide (KI) strategies for the general public. The projected dose at the site boundary is ~ 5 Rem Thyroid CDE.
- 3. APPROVED BY: 1 _
Station Emergency Manager or Date 1 Time Recovery Manager
NUMBER ATTACHMENT TITLE ATTACHMENT EPIP-1.06 4 DOWNWIND SECTOR TABLE REVISION PAGE 7 1 of 1 NOTE: Rounding shall be used when determining affected sectors using wind direction.
For example: Wind Direction (degrees from) 11.5 to 11.9 shall be rounded up to 12.0.
Wind Direction (degrees from) 11.1 to 11.4 shall be rounded down to 11.0.
AVERAGE WIND DIRECTION AFFECTED SECTORS (Degrees From) 349-11 H, J, K 12-33 J,K,L 34-56 K, L, M 57-78 L, M, N 79-101 M,N,P 102-123 N,P,O 124-146 P,O, R 147-168 O,R,A 169-191 R,A,B 192-213 A,B,C 214-236 B,C,D 237-258 C,D,E 259-281 D,E,F 282-303 E,F,G 304-326 F, G, H 327-348 G,H,J
NUMBER ATTACHMENT TITLE ATTACHMENT EPIP-1.06 5 REFERENCE INFORMATION REVISION PAGE (NAPS) 7 1 of 1 Refers to already known conditions, for example:
- Severe weather such as hurricanes, tornados, flooding, or blizzards.
- Traffic issues such as inadequate roads, major accident(s).
KNOWN EVACUATION
- An attack on an off-site infrastructure such as water and IMPEDIMENTS: power lines, transportation and communication systems, and public institutions including schools, post offices and prisons.
It is not expected that Protective Action Recommendation (PAR) development be delayed by attempting to obtain information from outside resources.
The initial PAR must be included with the initial notification of a General Emergency, which must be made to the State within 15 minutes follow-PAR NOTIFICATION ing declaration of the General Emergency.
( TIMES:
Notification of a revised PAR must be made to the State within 15 minutes of its development.
Downwind sectors may be determined from Attachment 4, Downwind SECTORS/ Sector Table.
WIND DIRECTION:
Wind direction is always given in degrees from.
There is no dose threshold for recommending sheltering-in-place.
Sheltering-in-place may be recommended as a result of controlled releases, evacuation impediments or other known conditions which SHELTERING make evacuation dangerous.
ill PLACE: A controlled release is a short-term release, which is a controlled evolution and the release duration can be accurately determined (such as containment venting).
Previously issued Protective Action Recommendations should not be PREVIOUSLY reduced until the threat is fully under control and after consulting with ISSUED PAR: Commonwealth of Virginia emergency response organization.
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS Instrument air has been lost INITIATING CUE You are requested to open the Residual Heat Removal Heat Exchanger Cooling Water Return Valves 1-CC-TV-103A and 1-CC-TV-1038 locally in accordance with 1-AP-28, Attachment 3.
03/04/08 Page: 1 of 8
(.
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM N925 Open the residual heat removal heat exchanger cooling water return valves using a jumper (1-AP-28,0-FCA-1).
TASK STANDARDS 1-CC-TV-103A and 1-CC-TC-103B have been jumpered open.
Work was performed in compliance with the Radiation Work Permit; exposure to surface and airborne contamination was minimized; and ALARA principles were applied.
KJA
REFERENCE:
008-A2.05 (3.3/3.5)
ALTERNATE PATH:
N/A TASK COMPLETION TIMES Critical Time =NO Validation Time = 20 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ 1SATISFACTORY [ 1UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature 1 Date EVALUATOR'S COMMENTS 03/04/08 Page: 2 of 8
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM N925 READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES Before being evaluated on the task, the trainee must have completed the reactor operator's course checkout during which the objectives listed below would have been addressed.
INITIAL CONDITIONS Instrument air has been lost INITIATING CUE You are requested to open the Residual Heat Removal Heat Exchanger Cooling Water Return Valves 1-CC-TV-103A and 1-CC-TV-103B locally in accordance with 1-AP-28, Attachment 3.
03/04/08 Page: 3 of 8
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT Adjustable wrench Nitrogen jumper rig Nitrogen or air bottle PERFORMANCE STEPS START TIME rr::=J Obtain a jumper rig, nitrogen or air bottle, and adjustable wrench. IProcedure Step ICritical Step ISAT [1 UNSAT [1 Standards Operator explains where to obtain jumper rig, nitrogen or air bottle, and ad'ustable wrench.
Verbal-Visual Show me where you would obtain the jumper rig.
Cues Explain to me how you could obtain a nitrogen or air bottle.
Tell me where you would go to obtain an adjustable wrench.
After the above have been done, give the following cue:
You may assume that you have the jumper rig, nitrogen bottle, and adiustable wrench.
INoles/Comm enls 03/04/08 Page: 4 of 8
(
[L] Isolate the instrument air supply to the trip valves. I Procedure Step I SAT [1 UNSAT [ 1 IL...,;;S;,;;ta;;;,;n,;,;;d;,;;;a;.;,;rd;;,;;s~_ _ 1 Local air supply valves for 1-CC-TV-103A and 1038 are closed.
Verbal-Visual Confirm actions once operator identifies components and describes Cues closing valves.
[Notes/Comments
~ Connect the regulator to the nitrogen bottle. IProcedure Step I SAT [1 UNSAT [ 1 IL...,;;S;,;;ta;;;,;n,;,;;d;,;;;a;.;,;rd;;,;;s~ _ _ Jumper rig is connected to the nitrogen bottle.
1 Verbal-Visual Have operator describe using attachment 3 and confirm that rig is Cues connected as described.
INotes/Comments 03/04/08 Page: 5 of 8
~I Connect the jumper rig to the trip valves. IProcedure Step I--=C~r::.::.iti::.::.c=a=-'S~t=eL..p 1SAT [] UNSAT []
I...,;;S~ta;;;,n,;,;;d;,;;;a,;",;rd;;;;;s~_ _ 1 Jumper rig is connected to 1-CC-TV-103A and 1-CC-TV-103B.
Verbal-Visual Have operator describe using attachment 3 and confirm that rig is Cues connected as described.
I Notes/Com ments
~ Align nitrogen or air bottle to open the trip valves. IProcedure Step IL-CL-r_itiL-cL-a_'S_t=-e-'-p 1 SAT [] UNSAT []
Standards The following actions are performed in sequence:
- Adjust regulator to full counterclockwise direction.
- Slowly adjust the regulator in the clockwise direction until the residual heat removal heat exchanger component cooling outlet isolation valves are open.
Verbal-Visual Have operator describe using Attachment 3 and confirm actions as Cues described.
rotes/com ments 03/04/08 Page: 6 of 8
Procedure Step _ _
ISAT [1 UNSAT [ 1 Standards Verbal-Visual The control room acknowledges that 1-CC-TV-103A and 1038 have Cues been jumpered open.
A jumper log entry will be made.
I Notes/Comments
>>>>> END OF EVALUATION <<<<<
( STOP TIME 03/04/08 Page: 7 of 8
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
(
03/04/08 Page: 8 of 8
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-28 3 LOCAL OPERATION OF 1-CC-TV-103A AND 1-CC-TV-103B REVISION PAGE 30 1 of 3 NOTE:
- A key is required to unlock the Appendix R storage cabinet.
- An adjustable wrench is required to connect the nitrogen jumper rig.
- 1. Obtain Nitrogen/Air Jumper Rig from the Appendix R storage cabinet (located in the TSC HVAC room).
- 2. Obtain Nitrogen/Air Bottle, containing less than 4000 psig, from the bottle storage area.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-28 3 LOCAL OPERATION OF 1-CC-TV-103A AND 1-CC-TV-103B REVISION PAGE 30 2 of 3
- 3. In the Unit 1 Auxiliary Building Penetration Area, open the RHR Heat Exchanger CC Outlet Isolation Valves as follows:
_ a) Close the local air supply isolations to 1-CC-TV-103A and 1-CC-TV-103B.
b) Connect the Nitrogen/Air Jumper Rig to the Nitrogen/Air Bottle.
c) Connect the Nitrogen/Air Jumper Rig to 1-CC-TV-103A actuator as follows:
- 1) Disconnect the Instrument Air copper tubing from the end of the actuator.
- 2) Remove the tube to pipe 90 0 fitting from the bushing on the actuator.
- 3) Connect the quick disconnect adapter from the Nitrogen/Air Bottle hose to the bushing at the end of the actuator.
- 4) Connect the hose from the Nitrogen/Air Jumper Rig to the quick disconnect adapter at the end of the actuator.
d) Connect the Nitrogen/Air Jumper Rig to 1-CC-TV-1 03B actuator as follows:
- 1) Disconnect the Instrument Air copper tubing from the end of the actuator.
- 2) Remove the tube to pipe 90 0 fitting from the bushing on the actuator.
- 3) Connect the quick disconnect adapter from the Nitrogen/Air Bottle hose to the bushing at the end of the actuator.
- 4) Connect the hose from the Nitrogen/Air Jumper Rig to the quick disconnect adapter at the end of the actuator.
_ e) Adjust the regulator to full counterclockwise direction.
_ f) Open Nitrogen/Air Bottle isolation valve.
_ g) Open both Nitrogen/Air Jumper Rig shut-off valves.
_ h) Slowly increase the regulator setpoint until the RHR Heat Exchanger CC Outlet Isolation Valves are open.
- 4. Notify the Control Room that the RHR Heat ExchangerCC Outlet Isolation Valves are open.
- 5. Make a Jumper Log entry.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-28 3 LOCAL OPERATION OF 1-CC-TV-103A AND 1-CC-TV-103B REVISION PAGE 30 3 of 3 Regulator Shut Off Valves Nitrogen/Air Bottle Valve Operating Cylinders For 1-CC-TV-1 03A&B Graphics No: WT1037B l Quick Disconnect Adapter Bushing NITROGEN JUMPER RIG CONNECTION DIAGRAM
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS The Unit is in Mode 1 and Power to 1-1 vital bus has been lost.
1-1 vital bus has been inspected and approved for use by Electrical Department.
The 1-1 inverter and regulating transformer were recently untagged and are available for use.
Equipment status has been reviewed and station configuration will support placing the inverter in service.
Loaded testing is not required.
Inverter Trouble annunciator was not defeated.
There are 3 circulating water pumps in operation.
INITIATING CUE You are requested to re-energize vital bus 1-1 from inverter 1-1 and to energize the bypass transformer in accordance with 1-MOP-26.60.
03/04/08 Page: 1 of 11
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM N30 Re-energize a 120-volt vital bus from its inverter (1-MOP-26.60, O-AP-10).
TASK STANDARDS Vital bus 1-1 has been energized from inverter 1-1 and its bypass transformer is energized.
KIA
REFERENCE:
057-AA1.01 (3.7/3.7)
ALTERNATE PATH:
N/A TASK COMPLETION TIMES Validation Time = 25 minutes Start Time = ---
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ ] SATISFACTORY [ ] UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature I Date EVALUATOR'S COMMENTS 03/04/08 Page: 2 of 11
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM N30 READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES Before being evaluated on the task, the trainee must have completed the reactor operator's course checkout during which the objectives listed below would have been addressed.
INITIAL CONDITIONS The Unit is in Mode 1 and power to 1-1 vital bus has been lost.
1-1 vital bus has been inspected and approved for use by Electrical Department.
The 1-1 inverter and regulating transformer were recently untagged and are available for use.
Equipment status has been reviewed and station configuration will support placing the inverter in service.
03/04/08 Page: 3 of 11
Loaded testing is not required.
Inverter Trouble annunciator was not defeated.
There are 3 circulating water pumps in operation.
INITIATING CUE You are requested to re-energize vital bus 1-1 from inverter 1-1 and to energize the bypass transformer in accordance with 1-MOP-26.60.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT None PERFORMANCE STEPS START TIME 03/04/08 Page: 4 of 11
Ir=J Energize and align the inverter for normal operation. IProcedure Step ICritical Step ISAT [1 UNSAT [ 1 Standards The following actions are performed in sequence
- Depress the PRECHARGE push-button until the precharge light is lit.
- Close the battery input breaker.
- Close the inverter output breaker.
Verbal-Visual Precharge light is lit.
Cues Output voltage is approximately 120 volts.
Frequency is approximately 60 Hz.
I Notes/Comments Close the vital bus panel's isolation breaker, the main feeder Procedure Step _ _
breaker from the transformer switch, and re-install the tam er seal.
ICritical Step ISAT [1 UNSAT [1 Standards 1-EP-CB-001 and 1-EP-CB-4A breaker 35 are taken to CLOSE osition.
Verbal-Visual You may assume that another operator has installed the tamper seal.
Cues I Notes/Comm ents 03/04/08 Page: 8 of 11
9 Transfer the vital bus load to the inverter by performing the Procedure Step _ _
following actions:
- Place the manual bypass switch in the INVERTER TO LOAD position.
- Depress the INVERTER to LOAD pushbutton.
I_C_r_iti_c_a_1S_t_e.L..P
_ 1 SAT [] UNSAT []
Standards Vital bus 1-1 transfer switch is rotated to the INVERTER TO LOAD position.
Standards Inverter to LOAD pushbutton depressed.
rotes/com ments Verify that the vital bus is energized and is indicating about 120 Procedure Step _ _
volts.
ISAT [] UNSAT [ ]
IStandards IOperator checks vital bus voltage is approximately 120 volts.
Verbal-Visual Vital bus voltage indicates approximately 120 volts.
Cues Verbal-Visual Vital bus voltage indicates approximately 120 volts.
Cues I Notes/Comments 03/04/08 Page: 9 of 11
11 If the vital bus bypass transformer was also de-energized, perform Procedure Step _ _
the following actions:
- Close the bypass transformer circuit breaker.
- Close the AC input to regulator breaker on the inverter cabinet.
- Allow the transformer to warm up until the desired output voltage is obtained.
- Close the bypass source AC input breaker.
ISAT [1 UNSAT [ 1 Standards Bypass transformer circuit breaker is closed.
AC input breaker is closed.
Bypass source AC input breaker is closed.
Verbal-Visual Transformer has warmed up and desired output voltage is indicated.
Cues
( Assume that another operator will complete the procedure.
IN otes/Comments
>>>>> END OF EVALUATION <<<<<
STOP TIME 03/04/08 Page: 10 of 11
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required) 03/04/08 Page: 11 of 11
PROCEDURE NO:
'Dominion- 1-MOP-26.60 REVISION NO:
NORTH ANNA POWER STA TION 27 PROCEDURE TYPE: UNIT NO:
MAINTENANCE OPERATING PROCEDURE 1 PROCEDURE TITLE:
VITAL BUS 1-1 (1-EP-CB-04A)
SUMMARY
NON-UPGRADED PROCEDURE Incorporated CR021347 and CA018409, Unit 1 Safeguards Ventilation Aligned to Iodine Filter Due to Loss of Vital Bus I-I, as follows:
- Added new Reference 2.4.8 ofCR021347 and CA018409.
- Added P&L 4.7 and Note for Steps 5.2.12, 5.4.16, 5.4.21, 5.5.13, 5.5.17, and 5.6.12 that When Vital Bus I-I is deenergized, the Unit 1 SFGDS ventilation dampers and the Aux Building Iodine Filter dampers will fail to the FILTER position. The only indication of this is PCS points. To restore dampers to BYPASS position, damper control switches must be cycled to FILTER and then back to BYPASS after power is restored.
- Deleted old Step 5.4.19.c.3 and Step 5.5.19.c.3 to return the Unit 1 SFGD Exhaust Ventilation using O-OP-21.5, Operation Of Auxiliary Building Iodine Filters.
- Added new Steps 5.2.12, 5.4.21, 5.5.17, and 5.6.12 to restore Unit 1 SFGD Exhaust ventilation as follows:
- Place BOTH control switches for Unit 1 SFGDS 1-HV-AOD-128-1,2,3,4, to the FILTER position.
- Place BOTH control switches for Unit 1 SFGDS 1-HV-AOD-128-1,2,3,4, to the BYPASS position.
- Verify that flows on Vent Stack A and B remain stable as recorded on 1-HV-FR-1212A and 1- HV-FR-1212B.
continued on the next page PROBLEMS ENCOUNTERED: D NO DYES Note: If YES, note problems in remarks.
REMARKS:
(Use back for additional remarks.)
SRO: DATE:
CONTINUOUS USE
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 2 of 77 REVISION
SUMMARY
(Continued):
- Added new Step 5.1.1 OU) to verify that flows on Vent Stack A and B remain stable as recorded on I-HV-FR-1212A and I-HV-FR-1212B.
- Added a Note for new Step 5.1.10U) that When Vital Bus I-I is deenergized, the Unit 1 Safeguards ventilation dampers and the Aux Building Iodine Filter dampers will fail to the FILTER position.
- Added new Steps 5.2.13, 5.4.22, 5.5.18, and 5.6.13 thatIF the Aux Building iodine filter bank was NOT aligned to FILTER prior to the event, THEN restore the Aux Building iodine filter bank to BYPASS as follows:
- Place BOTH control switches for Aux Bldg Iodine Filter I-HV-AOD-I07A-l,2,3,4, to the FILTER position.
- Place BOTH control switches for Aux Bldg Iodine Filter I-HV-AOD-I07A-l,2,3,4, to the BYPASS position.
- Verify that flows on Vent Stack A and B remain stable as recorded on I-HV-FR-1212A and 1-HV-FR-1212B.
- Added new Steps 5.2.14, 5.4.23, 5.5.19, and 5.6.14 to operate the Unit 1 SFGD Exhaust ventilation as desired using O-OP-21.5, Operation Of Auxiliary Building Iodine Filters.
- Added new Step 5.4.16 and new Step 5.5.13 that IF l-EP-CB-19A, Main Control Board 120VAC Vital Bus SOV Panel A, is energized, THEN do the following to restore ventilation to normal.
Followed each of these steps with the steps to restore dampers to correct positions:
- Restore Unit I SFGD Exhaust ventilation
- Restore the Aux Building iodine filter bank to BYPASS
- Verify that flows on Vent Stack A and B remain stable
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 3 of 77 TABLE OF CONTENTS Section Page 1.0 PURPOSE 4
2.0 REFERENCES
4 3.0 INITIAL CONDITIONS 9 4.0 PRECAUTIONS AND LIMITATIONS 9 5.0 INSTRUCTIONS 12 5.1 Actions to be Taken in the Event of Loss of Bus 12 5.2 Transferring Panels from Vital Bus I-I to Alternate Power Supply 19 5.3 Removing Vital Bus I-I (l-EP-CB-04A) from Service 23 5.4 Reenergizing Vital Bus I-I from Inverter I-I 27 5.5 Reenergizing Vital Bus I-I from Voltage Regulating Transformer 41 5.6 Transferring Panels from Alternate Power Supply to Vital Bus I-I 53 ATTACHMENTS 1 Returning Containment Trip Valves To Normal 57 2 Appendix R Diagram 66 3 Downpowering Vital Bus I-I (l-EP-CB-04A) 67 4 Vital Bus I-I Inverter 77
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 4 of??
1.0 PURPOSE 1.1 To provide instructions for reenergizing Vital Bus 1-1 after being lost.
1.2 To provide instructions for taking actions in the event ofloss of bus.
1.3 To provide instructions for transferring panels powered from Vital Bus 1-1 to an alternate supply.
2.0 REFERENCES
2.1 Source Documents 2.1.1 UFSAR 8.3.1.2 2.1.2 UFSAR Section 6.2.5 2.1.3 UFSAR Section 3.1.37 2.2 Technical Specifications 2.2.1 Tech Spec 3.3.1 2.2.2 Tech Spec 3.3.2 2.2.3 Tech Spec 3.3.3 2.2.4 Tech Spec 3.3.4 2.2.5 Tech Spec 3.4.12 2.2.6 Tech Spec 3.4.15 2.2.7 Tech Spec 3.7.8 2.2.8 Tech Spec 3.7.11 2.2.9 Tech Spec 3.7.12 (PREACS)
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 5 of 77 2.2.10 Tech Spec 3.7.13 2.2.11 Tech Spec 3.8.1 2.2.12 Tech Spec 3.8.2 2.2.13 Tech Spec 3.8.7 2.2.14 Tech Spec 3.8.8 2.2.15 Tech Spec 3.8.9 2.2.16 Tech Spec 3.8.10 2.2.17 Tech Spec 3.9.3 2.2.18 Tech Spec 3.9.4 2.2.19 TRM 3.3.7 2.2.20 TRM 3.3.8 2.2.21 TRM 3.4.5 2.2.22 TRM 3.7.11 2.2.23 TRM 3.9.5 2.2.24 TRM7.4 2.2.25 TRM7.5 2.3 Technical References 2.3.1 11715-FE-1AA, 120V AC One Line Diagram, Vital Bus I 2.3.2 11715-FE-3QA, Wiring Diagram, Remote Monitoring Panell-EI-CB-203 2.3.3 117l5-ESK-6PY, 480V Circuits Miscellaneous Solenoid Valves 2.3.4 11715-ESK-6PY-2, 480V Circuits Miscellaneous Solenoid Valves
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 6 of??
2.3.5 1-0P-46.3, Containment Instrument Air System 2.3.6 1-0P-32.1, Steam Generator Blowdown Utilizing Aux Bldg Tank 1-BD-TK-1 2.3.7 1-0P-33.2, Steam Generator Transfer System 2.3.8 1-0P-33.3, Placing A Steam Generator in Wet Layup 2.3.9 1-0P-33.4, Placing B Steam Generator in Wet Layup 2.3.10 1-0P-33.5, Placing C Steam Generator in Wet Layup 2.3.11 1-0P-36.1, Main Condenser Hogger Ejectors 2.3.12 I-OP-36.2, Main Condenser Air Ejectors 2.3.13 1-PT-213.15, Valve Inservice Inspection Misc 2.3.14 1-0P-5.2, Reactor Coolant Pump Startup and Shutdown 2.3.15 1-0P-26.5, 120-Volt Vital Bus Distribution 2.3.16 1-0P-21.2, Containment Purge 2.3.17 Solid State Controls, Inc. Technician Manual for Inverter 3SV12090 2.3.18 Technical Requirements Manual, TR 7.4 and 7.5 2.3.19 Admin Lock Reduction Program 2.3.20 0-AP-10, Loss of Electrical Power 2.3.21 1-E-0, Reactor Trip or Safety Injection 2.3.22 0-ECM-2501-01, Trouble-Shooting and Repair of Single Phase Static Inverters 2.3.23 EPIP-1.01, Emergency Manager Controlling Procedure 2.3.24 1-AP-10.13, Restoration of AC Vital Busses, deleted by this procedure
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 7 of 77 2.3.25 1-AP-15, Loss of Component Cooling 2.3.26 1-AP-6, Loss of Main Control Room Annunciators 2.3.27 1-AP-3, Loss of Vital Instrumentation 2.3.28 1-AP-4.l, Malfunction of Source Range Nuclear Instrumentation 2.3.29 1-AP-4.2, Malfunction of Nuclear Instrumentation (Intermediate Range) 2.3.30 l-AP-4.3, Malfunction of Nuclear Instrumentation (Power Range) 2.3.31 DCP 00-174, Installation of Chilled Water and Bearing Cooling Water Bypass Switches to Prevent Low Flow Trips of Mechanical Chiller During System ManipulationslNAPSlUnit 1 & 2.
2.3.32 Plant Issue N-200l-l045, Addressing Source Range High Voltage Energizing 2.3.33 DCP 00-169, Control Room Bottled Air System Modification 2.3.34 NA-DW-271C351, sh. 18 2.3.35 LOOP-HV-033 2.3.36 11715-ESK-6NX & 6NW 2.3.37 DCP 01-159, Replacement Of Vital Bus Inverters I-I And I-II 2.3.38 IPR to DCP 02-186, Mechanical Chiller Refrigerant Conversion / NAPS /
Unit 2 2.4 Commitment Documents 2.4.1 CTS Assignment 02-92-1803, Commitment 003, Tech Spec Amendment 155/137 2.4.2 CTS Assignment 02-92-1804, Commitment 003, Tech Spec Amendment 156/138 2.4.3 Plant Issue N-1999-2192, Cycling of2-BP-SW-1 interlock pin
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 8 of 77 2.4.4 Tech Spec Change Request N-034 and TRCR #084 2.4.5 License Amendment 242, Tech Spec change N-029, Heat Up and Cooldown operating limits, PORV lift setpoints, LTOPS enable temperature 2.4.6 Plant Issue N-2005-4355, Received SW system loss of control power alarm (l/2J-D5) during the performance of2-MOP-26.62 2.4.7 Plant Issue N-2006-1552, Noted Numerous Unit 1 Annunciator Panel "L" Lites Not Lit After Return Of Vita1Bus IV 2.4.8 CR021347, CA018409, Unit 1 Safeguards Ventilation Aligned to Iodine Filter Due to Loss of Vital Bus I-I
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 9 of 77 Init Verif 3.0 INITIAL CONDITIONS Review the equipment status to verify station configuration supports the performance of this procedure.
4.0 PRECAUTIONS AND LIMITATIONS 4.1 Comply with the following guidelines when marking steps N/A:
- IF the conditional requirements of a step do not require the action to be performed, THEN mark the step N/A.
- IF any other step is marked N/A, THEN have the SRO approve the N/A and justify the N/A on the Procedure Cover Sheet.
4.2 IF Reactor Power is greater than P-6 (l x 10 --10 amps), THEN N-32 Source Range high voltage will energize on loss of Vital Bus I-I due to loss of high voltage cutout circuit.
4.3 This procedure affects Appendix R equipment. IF a problem is found that cannot be immediately resolved, THEN inform the SRO to refer to the Technical Requirements Manual, TRM 7.4 and 7.5.
4.4 Vital Bus I-I provides power to the controls for the Aux Building Central Exhaust System filter and bypass dampers. The action of Tech Spec 3.7.12 applies.
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 10 of 77 4.5 The following Tech Specs apply:
- Tech Spec 3.3.1, Reactor Trip System Instrumentation
- Tech Spec 3.3.2, Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation
- Tech Spec 3.3.4, Bases Table B 3.3.4-1, Auxiliary Shutdown Panel Monitoring Instrumentation
- Tech Spec 3.4.12, Low Temperature Overpressure Protection System
- Tech Spec 3.4.15, RCS Leakage Detection Instrumentation
- Tech Spec 3.7.11, MCRIESGR ACS
- Tech Spec 3.7.12, PREACS
- Tech Spec 3.7.13, MCRIESGR Bottled Air System
- Tech Spec 3.8.7, Inverters - Operating
- Tech Spec 3.8.8, Inverters - Shutdown
- Tech Spec 3.8.9, Distribution Systems - Operating
- Tech Spec 3.8.10, Distribution Systems - Shutdown
- Tech Spec 3.9.3, Nuclear Instrumentation
- Tech Spec 3.9.4, Containment Building Penetrations Operability
- Tech Spec 3.3.3, Table 3.3.3-1, Accident Monitoring Instrumentation
- TRM 3.3.7, Table 3.3.7-1, Radiation Monitoring Instrumentation
- TRM 3.4.5, Primary-to-Secondary Leakage Detection Systems
- TRM 3.9.5, Containment Purge and Exhaust Isolation System Operability
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 11 of 77 4.6 IF 1-EP-CB-4A, Vital Bus I-I, will be removed from service AND 2-EP-CB-4A, Vital Bus 2-1, is deenergized, THEN Train III of Control Room Bottled Air will be inoperable and the Action of Tech Spec 3.7.13 may apply.
4.7 When Vital Bus I-I is deenergized, the Unit 1 Safeguards ventilation dampers and the Aux Building Iodine Filter dampers will fail to the FILTER position. The only indication of this is PCS points. To restore dampers to BYPASS position, damper control switches must be cycled to FILTER and then back to BYPASS after power is restored. (Reference 2.4.8)
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 27 of??
5.4 Reenergizing Vital Bus 1-1 from Inverter 1-1 5.4.1 Verify Initial Condition is satisfied.
5.4.2 Review Precautions and Limitations.
5.4.3 Ensure the following conditions are satisfied:
- a. All required repairs have been completed.
- b. The Electrical Department has inspected the Inverter and bus as required.
- c. IF reenergizing after maintenance, THEN the Electrical Department has given its approval AND applicable tag outs have been cleared.
- d. The Electrical Department has been informed of any circuit breakers found tripped.
WARNING Use caution when entering bay, due to exposed energized electrical circuits.
NOTE: Refer to Attachment 4, Vital Bus I-I Inverter, for location of Crest factor Board and Auto-Retransfer Switch.
- e. IF maintenance was performed on the inverter, THEN ensure the AUTO-RETRANSFER SWITCH is in OFF. (Small toggle switch located on the Crest Factor Board.)
- f. IF loaded testing is required, THEN perform Step 5.4.19.a AND then return to Step 5.4.4.
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 28 of 77 5.4.4 IF only one or two Circulating Water Pumps are running AND l-EP-CB-19A, Main Control Board 120VAC Vital Bus SOV Panel A, is deenergized, THEN place l-EP-CB-19A Breaker No. 19, (Condenser Waterbox Trip Valves) in OFF.
5.4.5 Ensure l-EP-CB-12A, DC Bus I-I, Breaker No. 13, Pwr Supply To l-VB-INV-Ol (Appendix R Power Supply) is in ON.
CAUTION
- To avoid damage to the cord and to the plug internals, the patch cord MUST be grasped by the plastic covering on the plug and NOT by the cord.
- The patch cord MUST be inserted into the designated socket very SLOWLY, while maintaining a 90 degree angle to the socket board. This wi11limit the possibility of arcing, which could seriously damage the power supply to the board.
5.4.6 IF Panel H-Al, INVERTER I-I TROUBLE, was defeated, THEN enable as follows:
- a. Unlock and open the door to l-EI-CB-2l, Hathaway Cabinet (located in Unit 1 Hathaway Room).
- b. Remove the Danger Tag from patch cord 1304 and insert the plug on the sv cord into socket number H-1.
- c. Verify Panel H-Al is LIT.
- d. Close and lock the door to l-EI-CB-21.
- e. Remove H-Al from the Disabled Annunciator log.
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 29 of 77 5.4.7 Clear Danger Tags as follows: (N/A components not tagged)
- BYPASS SOURCE AC INPUT Breaker at I-VB-INV-Ol sv
- I-VB-HS-l, Manual Bypass Switch Vital Bus 1-1 Power, at I-VB-INV-Ol sv in BYPASS TO LOAD position.
- AC INPUT TO REGULATOR breaker at I-VB-INV-Ol in OFF.
sv
- l-EP-CB-OOl, Vital Bus Panel 1-1 Isolation Breaker sv
- l-EE-BKR-IHI-4 C3L, Vital Bus 1-1 Bypass Transformer Circuit Bkr sv
- l-EP-CB-04A Breaker No. 35, Main Feeder From Transfer Switch sv I-VB-HS-l
- l-EP-CB-12A, DC Bus 1-1, Breaker No. 13, Pwr Supply To I-VB-INV-Ol sv 5.4.8 Do the following at I-VB-INV-Ol, Vital Bus Distribution Panel 1-1 Inverter:
- a. Press and hold the PRECHARGE button.
- b. WHEN the PRECHARGE light is LIT, THEN release the PRECHARGE button.
- c. Close the DC INPUT breaker.
- d. Close the INVERTER OUTPUT breaker.
- e. Allow the Inverter to warm-up until output voltage is 114 to 125 volts.
- f. Verify frequency is about 60 Hz.
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 30 of 77 CAUTION WHEN the Inverter is left running unattended with the output breaker open for warm-up or testing periods, THEN the output breaker MUST be Danger Tagged to prevent an inadvertent return to service.
(Reference 2.3.22) 5.4.9 Do the following:
- a. Close the following:
- l-EP-CB-OOl, Vital Bus Panel I-I Isolation Breaker sv
- l-EP-CB-04A, Breaker No. 35, Main Feeder From Transfer Switch sv I-VB-HS-l
- b. Re-install Tamper Seal on l-EP-CB-OOl, Vital Bus Panel I-I Isolation sv Breaker 5.4.10 At the I-VB-INV-Ol, place I-VB-HS-l, Manual Bypass Switch Vital sv Bus I-I Power, in INVERTER TO LOAD position.
5.4.11 At the I-VB-INV-Ol, Vital Bus Distribution Panel I-I Inverter, push sv INVERTER TO LOAD pushbutton.
5.4.12 Verify Vital Bus I-I is energized by a voltage indication of about 120 volts.
5.4.13 IF l-EE-TRAN-79A, 120V AC Vital Bus I-I Bypass Transformer, was deenergized, THEN do the following:
- a. Close l-EE-BKR-IHI-4 C3L, Vital Bus I-I Bypass Transformer Circuit sv Bkr.
- b. At I-VB-INV-Ol, ensure AC INPUT TO REGULATOR breaker is in sv ON.
- c. Allow the Voltage Regulating Transformer to warm-up until output voltage is 114 to 125 volts.
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 31 of 77
- d. At 1-VB-INV-01, ensure BYPASS SOURCE AC INPUT breaker is ON.
sv 5.4.14 Ensure Excore Flux Monitor power supply is normal as follows:
- a. Obtain Excore Flux Monitor Transfer Switch key from the Appendix R key locker (key number 73).
- b. At 1-EI-CB-202 in the Unit 1 Emergency Switchgear Room, place 2-XFR-SW-202, Excore Flux Monitor Power Supply Transfer, key switch in NORMAL.
- c. At 2-EI-CB-202 in the Unit 2 Emergency Switchgear Room, place 1-XFR-SW-202, Excore Flux Monitor Power Supply Transfer, key switch in NORMAL.
- d. Return Excore Flux Monitor Transfer Switch key to Appendix R key locker.
5.4.15 Clear the following TRM Action Statements:
- TR 7.4 for Unit 2 ChI Excore Neutron Flux Monitor
- TR 7.4 for Unit 2 Auxiliary Monitoring Panel
- TR 7.5 for 1-EI-CB-002
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 32 of??
NOTE: When Vital Bus I-I is deenergized, the Unit 1 Safeguards ventilation dampers and the Aux Building Iodine Filter dampers will fail to the FILTER position. The only indication ofthis is PCS points. To restore dampers to BYPASS position, damper control switches must be cycled to FILTER and then back to BYPASS after power is restored.
5.4.16 IF 1-EP-CB-19A, Main Control Board 120VAC Vital Bus SOY Panel A, is energized, THEN do the following to restore ventilation to normal:
- a. IF Unit 1 SFGD Exhaust ventilation was NOT aligned to the Aux Bldg iodine filter bank prior to the event, THEN restore Unit 1 SFGD Exhaust ventilation to BYPASS as follows: (Reference 2.4.8)
- 1. Place BOTH control switches for Unit 1 SFGDS l-HV-AOD-128-1,2,3,4, to the FILTER position.
- 2. Place BOTH control switches for Unit 1 SFGDS 1-HV-AOD-128-l,2,3,4, to the BYPASS position.
- 3. Verify that flows on Vent Stack A and B remain stable as recorded on l-HV-FR-1212A and 1-HV-FR-1212B.
- b. IF the Aux Building iodine filter bank was NOT aligned to FILTER prior to the event, THEN restore the Aux Building iodine filter bank to BYPASS as follows: (Reference 2.4.8)
- 1. Place BOTH control switches for Aux Bldg Iodine Filter 1-HV-AOD-107A-l,2,3,4, to the FILTER position.
- 2. Place BOTH control switches for Aux Bldg Iodine Filter 1-HV-AOD-107A-1,2,3,4, to the BYPASS position.
- 3. Verify that flows on Vent Stack A and B remain stable as recorded on 1-HV-FR-1212A and 1-HV-FR-12l2B.
- c. Operate the Unit 1 SFGD Exhaust ventilation as desired using O-OP-21.5, Operation Of Auxiliary Building Iodine Filters.
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 33 of 77 5.4.17 IF 1-EP-CB-19A, Main Control Board 120VAC Vital Bus SOY Panel A, 1-EP-CB-80A, 120VAC Instrumentation Distribution Panel I-I, and 1-EI-CB-21, MCR Annunciator/Lamp Panels, are energized from Vital Bus I-I, THEN mark N/A the remaining steps of this subsection and return Containment trip valves to normal using Attachment 1.
5.4.18 IF 1-EP-CB-19A, Main Control Board 120VAC Vital Bus SOY Panel A, 1-EP-CB-80A, 120VAC Instrumentation Distribution Panel I-I, and 1-EI-CB-21, MCR Annunciator/Lamp Panels, are energized from alternate power supply AND it is desired to reenergize from Vital Bus I-I, THEN mark N/A the remaining steps of this subsection and GO TO Subsection 5.6.
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 34 of 77 5.4.19 IF l-EP-CB-19A, Main Control Board 120VAC Vital Bus SOV Panel A, l-EP-CB-80A, l20VAC Instrumentation Distribution Panel I-I, OR l-EI-CB-2l, MCR Annunciator/Lamp Panels, are deenergized, THEN do the following:
- a. Close the following feeder breakers:
- l-EP-CB-04A, Breaker No.1, l-EI-CB-36, Control Power to NI Rack sv (N3l, N35, and N4l) (Protection)
- l-EP-CB-04A, Breaker No.2, l-EI-CB-36, Inst. Power to NI Rack sv (N3l, N35, and N4l) (Non-Protection)
- l-EP-CB-04A, Breaker No.3, Feeder to l-EI-CB-5l, Primary Plant sv Process Rack 1 (Protection)
- l-EP-CB-04A, Breaker No.4, l-EI-CB-55, Power Supply to Primary sv Process Rack 5
- l-EP-CB-04A, Breaker No.5, l-EP-CB-28H, SW Logic Cabinet sv (Train A)
- l-EP-CB-04A, Breaker No.6, l-EP-CB-28H, SW Logic Cabinet sv "Train A"
- l-EP-CB-04A, Breaker No.7, l-EI-CB-47A, SSPS Channell, sv Rack "A" Inputs
- l-EP-CB-04A, Breaker No.8, l-EI-CB-47B, SSPS Channell, sv Rack "B" Inputs
- l-EP-CB-04A, Breaker No.9, Feeder to l-EE-EG-lC, Emergency sv Generator "lH" Control Cabinet
DOMINION 1-MOP-26.60 North Anna Power Station Revision 2?
Page 35 of??
Close the following feeder breakers: (Continued)
- 1-EP-CB-04A, Breaker No. 10, AUK Relay Rack 1 (Train A) sv
- 1-EP-CB-04A, Breaker No. 11, Turb. Bldg. Flood Test Pnl (Train A) sv
- 1-EP-CB-04A, Breaker No. 13, Feeder to 1-EP-CB-19A, Main Control sv Board Vital Bus SOV Panel (Train A)
- 1-EP-CB-04A, Breaker No. 18, Feeder to 1-EI-CB-156A, Bottled Air sv Panel
- 1-EP-CB-04A, Breaker No. 22, Feeder to 1-EI-CB-23A, Secondary sv Plant Process Control Rack "A" Power Supplies
- 1-EP-CB-04A, Breaker No. 25, Feeder to 1-EI-CB-156A, CR Bottled sv Air Panel, Secondary Plant Process Rack CE
- 1-EP-CB-04A, Breaker No. 27, 1-IA-D-2A, Containment Instrument sv Air Dryer
- l-EP-CB-04A, Breaker No. 28, Feeder to 1-EP-CB-80E, sv Instrumentation Distribution Panel I-V
- 1-EP-CB-04A, Breaker No. 29, Feeder to 1-EI-CB-47E, SSPS Train sv "A" Outputs
- 1-EP-CB-04A, Breaker No. 30, SSPS AUK Relay Rack Train "A" sv
- l-EP-CB-04A, Breaker No. 32, Feeder to 1-EP-CB-80A, sv Instrumentation Distribution Panel I-I
- 1-EP-CB-04A, Breaker No. 33, Feeder to 1-EI-CB-49B, Radiation sv Monitoring System Cabinet 1-2
- 1-EP-CB-04A, Breaker No. 34, Feeder to l-EI-CB-21, Main sv Annunciators
- 1-EP-CB-04A, Breaker No. 35, 1-EP-CB-04A, Main Supply Breaker sv
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 36 of 77
- b. Clear the following Tech Spec Actions:
- Tech Spec 3.3.3 for Trip Valves
- TRM 3.3.8 on I-HC-H2A-I01, H2 Analyzer
- c. Return the following equipment to service as required:
- 1. One Containment Sump Pump, place control switch in AUTO: Mark control switch not used N/A.
- I-DA-P-4A, 4A Reactor Containment Sump Pump
- I-DA-P-4B, 4B Reactor Containment Sump Pump
- 2. l-RM-RMS-159 and l-RM-RMS-160, Containment Gaseous and Particulate Rad Monitors.
- 3. One Containment Vacuum Pump in service using 1-0P-19.2, Operation Of Containment Vacuum Pumps: Mark other pump N/A.
- l-CV-P-3A, 3A Containment Vacuum Pump
- l-CV-P-3B, 3B Containment Vacuum Pump
- 4. 2-CD-MR-l, Chilled Water Mechanical Chiller, using O-OP-51.5, Operation Of Chilled Water Systems: Mechanical Chiller.
- 5. Service Water to Containment Air Recirculation Fans using 1-0P-21.2, Containment Purge.
5.4.20 Return Containment trip valves to normal using Attachment 1.
DOMINION 1-MOP-26.60 North Anna Power Station Revision 2?
Page 3? of??
NOTE: When Vital Bus I-I is deenergized, the Unit 1 Safeguards ventilation dampers and the Aux Building Iodine Filter dampers will fail to the FILTER position. The only indication of this is PCS points. To restore dampers to BYPASS position, damper control switches must be cycled to FILTER and then back to BYPASS after power is restored.
5.4.21 IF Unit 1 SFGD Exhaust ventilation was NOT aligned to the Aux Bldg iodine filter bank prior to the event, THEN restore Unit 1 SFGD Exhaust ventilation to BYPASS as follows: (Reference 2.4.8)
- a. Place BOTH control switches for Unit 1 SFGDS I-HV-AOD-128-l,2,3,4, to the FILTER position.
- b. Place BOTH control switches for Unit 1 SFGDS I-HV-AOD-128-1,2,3,4, to the BYPASS position.
- c. Verify that flows on Vent Stack A and B remain stable as recorded on I-HV-FR-1212A and l-HV-FR-1212B.
5.4.22 IF the Aux Building iodine filter bank was NOT aligned to FILTER prior to the event, THEN restore the Aux Building iodine filter bank to BYPASS as follows: (Reference 2.4.8)
- a. Place BOTH control switches for Aux Bldg Iodine Filter I-HV-AOD-I07A-l,2,3,4, to the FILTER position.
- b. Place BOTH control switches for Aux Bldg Iodine Filter I-HV-AOD-I07A-l,2,3,4, to the BYPASS position.
- c. Verify that flows on Vent Stack A and B remain stable as recorded on l-HV-FR-1212A and l-HV-FR-1212B.
5.4.23 Operate the Unit 1 SFGD Exhaust ventilation as desired using O-OP-21.5, Operation Of Auxiliary Building Iodine Filters.
5.4.24 IF Action C of Tech Spec 3.3.2 was entered for this procedure, THEN clear Action C of Tech Spec 3.3.2.
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 38 of??
5.4.25 IF the 1H EDG was placed in MANUAL LOCAL solely for this procedure, THEN place the 1H EDG in AUTO REMOTE using 1-MOP-6.90, Emergency Diesel Generator 1-EE-EG-1H.
5.4.26 WHEN the load sequencing timer for 1-SW-P-1A is operable, THEN clear the Action of Tech Spec 3.7.8 or TRM 3.7.11 entered for this procedure.
5.4.27 Return Steam Generators to wet layup using the following procedures as applicable:
- 1-OP-33.3, Placing "A" Steam Generator In Wet Layup
- 1-OP-33.4, Placing "B" Steam Generator In Wet Layup
- 1-0P-33.5, Placing "C" Steam Generator In Wet Layup 5.4.28 IF Steam Generator transfer was in service previously, THEN start 1-WT-P-35, Steam Generator Transfer Pump 35, AND open 1-WT-431, 1-WT-P-35 Discharge.
5.4.29 IF required, THEN secure Containment Instrument Air using 1-0P-46.3, Containment Instrument Air System.
5.4.30 Have I&C install the 3A Fuse M13 in 1-EI-CB-48A, Aux Relay Rack 1 to I&C sv RTS the "High Flux At Shutdown Alarm" containment evacuation hom.
5.4.31 Clear the Power Range N-41 rate trip by placing the Rate Mode switch momentarily to RESET.
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 39 of77 5.4.32 Clear Tech Spec 3.3.1 Actions for the following:
- N-3l
- N-35
- N-41 5.4.33 Notify Instrument & Controls to verify the status of all cabinets powered from Vital Bus I-I appears normal. (Refer to the Station Load List for affected cabinets) (Reference 2.4.7) 5.4.34 IF required, THEN place Main Condenser Air Ejectors in service using 1-0P-36.2, Main Condenser Air Ejectors.
5.4.35 IF required, THEN secure Main Condenser Hoggers using 1-0P-36.1, Main Condenser Hogging Ejectors.
5.4.36 IF required, THEN restore Steam Generator Blowdown using on of the following:
- 1-0P-32.1, Low Capacity Steam Generator Blowdown System
- 1-0P-32.3, High Capacity Steam Generator Blowdown System Operation 5.4.37 Clear the ABNORMAL Status entry on l-EI-CB-48A, Aux. Relay Rack 1, 3A Fuse M13.
5.4.38 Do the following:
- a. Clear 7 Day Action on Tech Spec 3.7.12 (PREACS).
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 40 of??
- b. Clear Tech Spec 3.7.11 action on l-HV-F-42 fan (l-EP-CB-04A Bkr 18).
- c. Clear Tech Spec 3.7.13 action on Train 1 bottled air (l-HV-TV-1306A)
(l-EP-CB-04A Bkr 18).
Completed by: _ Date: - - - -
DOMINION 1-MOP-26.60 North Anna Power Station Revision 27 Page 77 of 77 (Page 1 of 1)
Attachment 4 Vital Bus 1-1 Inverter Crest Factor Board (Inside top right cabinet, 1,-,-1..l-J.--'-l.-L...L.J 1-1 on back wall,
,.,,"" ".""." woO ~~ Auto Retransfer Switch On = Up and Off = Down o --- ---
(i;0)
@X§) o <WD
@ @:§)
- o 0 0 0 000 o
- 0
- 0 I
o IPrechargel1 Precharge Light I Pushbutton I o 000
~ IBypass Sourcel
~I I
AClnput I
1 Graphics No. CS2831 VITAL BUS INVERTER
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS The OATe attempted to trip UNIT 2 and the reactor DID NOT trip.
INITIATING CUE You are requested to trip the UNIT 2 reactor locally in accordance with Attachment 4, Remote Reactor trip, of 2-FR-S.1, Response to Nuclear Power Generation/ATWS.
02/28/08 Page: 1 of 7
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM N1047 Trip the reactor by opening the reactor trip breakers or the rod-drive motor generator breakers locally (2-FR-S.1, 2-AP-20).
TASK STANDARDS The reactor was tripped by locally tripping MG set supply breakers in 307 Switchgear.
KIA
REFERENCE:
029-EA1.12 (4.1/4.0)
ALTERNATE PATH:
N/A TASK COMPLETION TIMES Validation Time = 10 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ ] SATISFACTORY [ 1UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature /
Date EVALUATOR'S COMMENTS 02/28/08 Page: 2 of 7
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM N1047 READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES Before being evaluated on the task, the trainee must have completed the Reactor Operator's course checkout during which the objectives listed below would have been addressed.
INITIAL CONDITIONS The OATC attempted to trip UNIT 2 and the reactor did not trip.
INITIATING CUE You are requested to trip the UNIT 2 reactor locally in accordance with Attachment 4, Remote Reactor trip, of 2-FR-S.1, Response to Nuclear Power Generation/ATWS.
02/28/08 Page: 3 of 7
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT None PERFORMANCE STEPS START TIME Note: The operator may perform actions 1-4, checking for a response after each individual action is performed, or may elect to perform all of the actions (1-4) then check the responses afterwards, either method is acceptable.
[r=J Locally trip the reactor from the rod drive room. IProcedure Step ISAT I] UNSAT I ]
Standards TRIP push-buttons for the Reactor Trip and Reactor Trip Bypass breakers are de ressed.
Verbal-Visual Inform the operator that breaker indication is as he/she sees them now.
Cues INotes/Comments 02/28/08 Page: 4 of 7
Put both M-G Set Generator Output Breakers control switches to Procedure Step _ _
TRIP.
ISAT [] UNSAT [ ]
IStandards IControl switches for both motor generator sets are placed in TRIP.
Verbal-Visual Inform the operator that breaker indication is as he/she sees them now.
Cues I
Notes/Comments Press the TRIP buttons for both M-G Set Generator Output Procedure Step _ _
Breakers.
ISAT [] UNSAT [ ]
Standards Depresses the TRIP push-button for each motor generator output breakers.
Verbal-Visual Inform the operator that breaker indication is as he/she sees them now.
Cues I
Notes/Comments 02/28/08 Page: 5 of 7
Put both M-G Set Motor Supply Breakers control switches to TRIP. Procedure Step _ _
I SAT [1 UNSAT [ 1 Standards Control switches for both motor generator set supply breakers are laced in TRIP.
Verbal-Visual Inform the operator that breaker indication is as he/she sees them now.
Cues r otes/Com ments 5 Procedure Step _ _
I_C_ri_ti_c_al_S--:t_e'-p 1 SAT [1 UNSAT [1 I...;;;S~ta;;,:.n~d~a;,,;;rd~s~ _ _ 1 Operator proceeds to Step 3 based on the Response Not Obtained.
Verbal-Visual Inform the operator that all breaker indications are as he/she sees them Cues now (cue may not be necessary if information was provided previously as the individual actions were performed).
INotes/Comments 02/28/08 Page: 6 of 7
cz==J Locally trip the reactor from the 307 Switchgear room. IProcedure Step ICritical Step ISAT[] UNSAT[]
Standards Local TRIP push-buttons for both M-G set motor supply breakers are depressed:
- 2-EP-BKR-24A1-3, 2-ED-MG-1A Rod Drive M-G Set Motor Supply Breaker.
- 2-EP-BKR-24C2-12, 1-ED-MG-1 B Rod Drive M-G Set Motor Supply Breaker.
Verbal-Visual When the operator describes the expected response for the actions Cues (breaker changes state as indicated by the green OPEN mechanical flaq) inform the operator that the breaker shows a green OPEN flaq.
[Notes/Comments Notify the control room operator of status of the rod power supply Procedure Step _ _
breakers.
ISAT [] UNSAT [ ]
Verbal-Visual Operator directs you to return to your normal duties.
Cues rotes/comments
>>>>> END OF EVALUATION <<<<<
STOP TIME 02/28/08 Page: 7 of 7
NUMBER ATTACHMENT TITLE ATTACHMENT 2-FR-S.1 4 REMOTE REACTOR TRIP REVISION PAGE 14 1 of 2 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: Because of M-G Set Flywheel coastdown, Reactor trip may be delayed for approximately one minute following opening of the M-G Set Motor Supply Breakers.
- 1. LOCALLY TRIP THE REACTOR FROM THE ROD DRIVE ROOM:
a) Do the following:
- Press the TRIP buttons for both Reactor Trip Breakers:
o . 2-EP-BKR-RTA o . 2-EP-BKR-RTB
- Press the TRIP buttons for both Bypass Breakers:
o . 2-EP-BKR-BYA o . 2-EP-BKR-BYB o . Put both M-G Set Generator Output Breakers control switches to TRIP o . Press the TRIP buttons for both M-G Set Generator Output Breakers o . Put both M-G Set Motor Supply Breakers control switches to TRIP (STEP 1 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 2-FR-S.1 4 REMOTE REACTOR TRIP REVISION PAGE 14 2 of 2 ACTIONI EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 1. LOCALLY TRIP THE REACTOR FROM THE ROD DRIVE ROOM: (Continued) b) Verify at least one of the following conditions - o b) GO TO Step 3.
SATISFIED 0
- All Reactor Trip and Bypass Breakers -
OPEN OR 0
- Both M-G Set Generator Output Breakers - OPEN OR 0
- Both M-G Set Motor Supply Breakers -
OPEN
- 2. NOTIFY THE CONTROL ROOM OF STATUS OF ROD POWER SUPPLY BREAKERS AND AWAIT FURTHER INSTRUCTIONS
- 3. LOCALLY TRIP THE REACTOR FROM 307 SWITCHGEAR:
a) Press the Mechanical Trip buttons for both M-G Set Motor Supply Breakers:
o . 2-EP-BKR-24A1-3, 2-ED-MG-1A Rod Drive M-G Set Motor Supply Breaker o . 2-EP-BKR-24C2-12, 2-ED-MG-1B Rod Drive M-G Set Motor Supply Breaker o b) Notify the Control Room of status of rod power supply breakers and await further instructions
-END-
Initial conditions are satisfied and precautions and limitations are Procedure Step _ _
reviewed.
ISAT[] UNSAT[]
Standards Initial conditions of JPM are checked and procedure precautions and limitations are reviewed.
IN otes/Com ments Verify that the Electrical Department has approved the bus to be re- Procedure Step _ _
ener ized.
ISAT[] UNSAT[]
IL..,;;S;;,;t;;;;:an:.;,;d;;;a;;;,r,;;;,ds~ llnitial conditions of JPM are checked to verify bus can be re-energized.
Verbal-Visual Shift manager reports that the Electrical Department has approved re-Cues ener izin the bus.
Verbal-Visual No circuit breakers have been found tripped.
Cues
[Notes/Comments 03/04/08 Page: 5 of 11
~ Ensure that the AUTO RE-TRANSFER switch is in OFF. IProcedure Step ISAT [1 UNSAT [ 1 NOTE TO THE Have operator explain where switch is located and how it would be EVALUATOR checked. DO NOT 0 en cabinet. Refer to Attachment 4 of rocedure Standards Operator explains that switch is located inside top, right cabinet and that the switch would be down if it was OFF.
Verbal-Visual Explain how you would perform this step.
Cues Assume another operator has verified that the auto re-transfer switch is OFF.
rotes/com ments 4 Open breaker #19 on SOV paneI1-EP-CB-19A if the following Procedure Step _ _
conditions exist.
- One or two circulating water pumps are running .
- Vital SOV paneI1-EP-CB-19A is de-energized .
ISAT [1 UNSAT [ 1 I~S;,;;ta;;;;,n;,;;d~a;",;;rd;;;;s~_ _ 1 Operator N/As step per initial condition.
rotes/Com ments 03/04/08 Page: 6 of 11
Ensure that the DC vital bus 1-1 (1-EP-CB-12A) breaker #13 is Procedure Step _ _
closed.
ISAT [1 UNSAT [ 1 I.,;;;S;;;;ta;,;,n;,;;;d,;;;,a;,;;rd;,;;;s~_ _11-EP-CB-12A, breaker #13 is verified to be closed.
INotes/Comments
~ Enable annunciators and clear associated danger tags, if required. IProcedure Step ISAT [1 UNSAT [ 1 Standards Operator verifies that annunciators are enabled and danger tags cleared er the Initial Conditions rovided.
INotes/Comm ents 03/04/08 Page: 7 of 11
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS A Reactor Trip occurred from 100% power.
Reactor Coolant System boron concentration prior to trip was 980 ppm.
Core burnup is approximately 9,000 MWO/MTU.
'A' BAST is the on-service BAST and is at 14,043 ppm.
The crew has transitioned from 1-E-0, Reactor Trip or Safety Injection to 1-ES-0.1, Reactor Trip Response with the plant stable.
INITIATING CUE You are requested to perform Step 7 of 1-ES-0.1, Reactor Trip Response.
02/25/08 Page: 1 of 11
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM R115 (modified)
Emergency Borate for stuck rods following a Reactor Trip (1-ES-0.1).
TASK STANDARDS Emergency Boration established using 1-CH-241 manual Emergency Borate Valve. Termination criteria determined to be 30% change in BAST level or SDM verified using 1-PT-10 series procedure.
KIA
REFERENCE:
024 - AA2.01, (3.8/4.1)
ALTERNATE PATH:
NIA TASK COMPLETION TIMES Validation Time = 15 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ 1SATISFACTORY [ 1UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature 1 Date EVALUATOR'S COMMENTS 02/25/08 Page: 2 of 11
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM R115 (modified)
READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES The trainee has completed the applicable course knowledge training at the reactor operator level.
INITIAL CONDITIONS A Reactor Trip occurred from 100% power.
Reactor Coolant System boron concentration prior to trip was 980 ppm.
Core burnup is approximately 9,000 MWD/MTU.
'A' BAST is the on-service BAST and is at 14,043 ppm.
02/25/08 Page: 3 of 11
The crew has transitioned from 1-E-O, Reactor Trip or Safety Injection to 1-ES-O.1, Reactor Trip Response with the plant stable.
INITIATING CUE You are requested to perform Step 7 of 1-ES-O.1, Reactor Trip Response.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT None PERFORMANCE STEPS START TIME 02/25/08 Page: 4 of 11
Note: all required cues provided by booth operator unless performed on dead simulator.
0==1 Verify AlIlRPls -10 Stepsor Less. I Procedure Step I....::.C~ri:....:.:ti...::...ca.:.:...I_S::.....:t:.-eL..p 1 SAT [] UNSAT []
Standards Observes IRPJs and determines 3 rods reading >10 steps (operator ma use IRPI console indications and/or PCS .
Verbal-visual When operator demonstrates how he would check IRPI indications Cues inform them that all are as expected except for the following:
Rod indicates steps Rod indicates steps Rod indicates steps Rod bottom IiQht for Rod is OFF.
I Notes/Com ments Place the control switch for the in-service boric acid transfer pump Procedure Step _ _
in FAST.
ICritical Step ISAT [] UNSAT []
1L.,;;~;;,;;ta;;;;n,;,;;d;,;;;a;;,,;rd;;;;s~_ _ 1 Control switch for the pump is placed in FAST.
Verbal-visual Once operator describes expected response, (red fast speed light is lit Cues all other lights associated with the pump are off), then confirm the indications.
INotes/Comments 02/25/08 Page: 5 of 11
~ Open emergency boration valve 1-CH-MOV-1350. IProcedure Step ICritical Step ISAT [] UNSAT [ ]
Standards Control switch for the emergency boration valve is placed in OPEN.
Verbal-visual Once operator describes expected response, tell the operator that red Cues and reen Ii hts for 1-CH-MOV-1350 are OFF.
Notes/Comments Valve will thermal out; this may be diagnosed by operator based on the loss of light indication; operator may elect to dispatch Aux bldg. operator to check valve and possibly another operator to look at breaker.
~ Verify Emergency Boration flow IProcedure Step ICritical Step ISAT [] UNSAT [ ]
IStandards IOperator observes FI-111 0 indicates < 35 gpm Verbal-visual Once operator identifies meter inform the operator that FI-111 0 Cues indicates as he sees it, (0 gpm).
If operator checks VCT level provide feedback that it reads % and is stable.
INotes/Comments 02/25/08 Page: 6 of 11
[C] Locally open 1-CH-MOV-1350 and verify flow. I Procedure Step ICritical Step
_~_----'----'L- _ ISAT [] UN SAT [ ]
IStandards IDispatches Aux Bldg. operator to locally open 1-CH-MOV-1350.
Verbal-visual As the Aux Bldg. operator inform operator that the handwheel will not Cues en a eon 1-CH-MOV-1350 and he is unable to 0 en the valve.
Notes/Comments The Aux Bldg. operator will call back to inform operator that the handwheel will not engage on 1-CH-MOV-1350 and he is unable to open the valve).
Operator may have dispatched the aux. bldg. operator earlier based on the 1-CH-MOV-1350 malfunction and may elect to have 1-CH-241 opened at this time based on the report that 1-CH-MOV-1350 can not be opened, this course of action is acceptable.
02/25/08 Page: 7 of 11
~I Establish boration via local emergency borate valve. *1 Procedure Step C_r_it_ic_a_1S_t_e.L...P
,-I 1 SAT [1 UNSAT [1 Standards
- Places blender mode switch in BORATE.
- Places blender control switch in START.
- Places 1-CH-FCV-1113A control switch in OPEN.
- Places 1-CH-FCV-1113B control switch in CLOSE.
- Dispatches Aux Bldg. operator to locally open 1-CH-241.
Verbal-visual As the Aux Bldg. operator inform operator that 1-CH-241 is open.
Cues Confirm indications for expected response when stated by operator:
- Red light for blender is ON, green light is OFF.
- Red light for 1-CH-FCV-1113A is ON, green light is OFF.
- Red light for 1-CH-FCV-1113B is OFF, green light is ON.
If operator checks VCT level provide feedback that it is % and increasing.
Notes/Comments Operator may initiate the action to open 1-CH-241 earlier based on the report that MOV-1350 cannot be opened locally, this course of action is acceptable.
02/25/08 Page: 8 of 11
[r=I Record information. I Procedure Step ISAT [l UNSAT [ 1 Standards The following information is recorded in the procedure:
- Start time.
- On-service (A) BAST level.
Verbal-visual When indicator for A BAST is identified by operator point to the Cues 90% mark on the indicator usin a en.
I Noles/Comments 02/25/08 Page: 9 of 11
~ Initiate Attachment 2 to determine stopping criteria. IProcedure Step ICritical Step Standards
- Determines 2 equivalent stuck rods.
-7 1 rod reading> 20 steps.
-7 2 rods reading between 11-20 steps.
- Determines stopping criteria is 30% change in BAST level (operator may identify that less acid may be required based on performance and results of a PT-10 series procedure to determine shutdown marqin).
Verbal-visual If operator does not identify clearly what the termination criteria is, then Cues as the SRO ask him what criteria must be met to terminate the Emergency Boration.
Tell the operator that an additional operator will complete the remainder of the procedure.
INotesiComm ents
>>>>> END OF EVALUATION <<<<<
STOP TIME 02/25/08 Page: 10 of 11
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
Emergency Borate for stuck rods following a Reactor Trip (1-ES-0.1).
CHECKLIST
_ _ RecalilC #171 (0% power, post-trip)
Booth Operator Actions:
If called to investigate 1-CH-MOV-1350 due to loss of lights tell the operator you don't see anything obviously wrong with the valve and if asked there is nobody working in the area.
If called to investigate the breaker for 1-CH-MOV-1350 due to loss of lights tell the operator you don't see anything obviously wrong with the breaker and if asked there is nobody working in the area.
When directed to locally open 1-CH-MOV-1350 inform the operator that the handwheel will not engage on 1-CH-MOV-1350 and you are unable to open the valve.
When called to open 1-CH-241 open the valve by actuating trigger 1 and report action when complete (valve is set to open on 30 second ramp).
02/25/08 Page: 11 of 11
NORTH ANNA POWER STATION EMERGENCY PROCEDURE NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE (WITH THREE ATIACHMENTS) 1 of 21 PURPOSE To provide instructions to stabilize and control the plant following a Reactor Trip without a Safety Injection.
ENTRY CONDITIONS This procedure is entered from:
- 1-E-O, REACTOR TRIP OR SAFETY INJECTION.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 2 of 21 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- STEAM DUMPS - CONTROLLING: 0 a) Stop dumping steam.
0
- STABLE AT 54rF 0 b) Verify SG Blowdown Trip Valves are closed.
OR 0 IF NOT, THEN manually close valves.
0
0
- STABLE AT 551°F OR d) !Lcooldown continues, THEN close the following:
0
- TRENDING TO 551°F 0
- MSTVs 0
- MSTV Bypass Valves (STEP 1 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 3 of 21 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 1. CHECK RCS AVERAGE TEMPERATURE:
(Continued)
IF temperature is greater than control value AND increasing, THEN do the following:
o
- Dump steam to the Condenser OR o
- Dump steam using Decay Heat Release Valve:
- 1) Locally open isolation valve(s) for NON-RUPTURED SG(s) to Decay Heat Release Valve:
o
- 1-MS-19, A Steam Line to 1-MS-HCV-104 Non-Return Valve o
- 1-MS-58, B Steam Line to 1-MS-HCV-104 Non-Return Valve o
- 1-MS-96, C Steam Line to 1-MS-HCV-104 Non-Return Valve o 2) Locally open 1-MS-20, Decay Heat Release Valve Upstream Isolation Valve.
o 3) Manually open 1-MS-HCV-104, Decay Heat Release Valve.
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 4 of 21 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 2. CHECK FEEDWATER STATUS:
0 a) Check RCS average temperature - 0 a) WHEN temperature is less than 554°F, LESS THAN 554°F THEN do Step 2b.
0 Continue with Step 2c.
0 b) Verify Main Feed Reg Valves - 0 b) Manually close valves.
CLOSED c) Verify AFW Pumps - RUNNING c) !E AFW Pumps are required, THEN do the following:
0
- Motor-Driven AFW Pumps -
RUNNING 0
- Manually start Motor-Driven Pumps 0
- Turbine-Driven AFW Pump -
- Manually open Turbine-Driven AFW RUNNING Pump Steam Supply Valves:
0
- 1-MS-TV-111A 0
d) Verify total feed flow to SGs: d) Establish feed flow to the SGs as necessary:
0
OR OR 0
- Main Feedwater on bypass 0
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 5 of 21 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 3. CHECK SG LEVELS:
D a) Check narrow range level in D a) !E any affected SG wide range level is ALL SGs - GREATER THAN 11 % NOT increasing, THEN increase feed flow to that SG.
D b) Control feed flow to maintain narrow range D b) !E narrow range level in any SG levels between 23% and 33% continues to increase, THEN stop feed to that SG.
- 4. VERIFY CHARGING - IN SERVICE Establish charging:
D a) Put controller for 1-CH-FCV-1122 in MANUAL and close.
D b) Close 1-CH-HCV-1311, Auxiliary Spray Valve.
c) Open Normal Charging Line Isolation Valves:
D
- 1-CH-MOV-1289A D
- 1-CH-MOV-1289B D
- 1-CH-HCV-1310 D d) Open 1-CH-FCV-1122 to establish at least 25 gpm flow.
D e) Adjust charging to restore PRZR level to greater than 15%.
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 6 of 21 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 5. CHECK PRZR LEVEL CONTROL:
o a) Level - GREATER THAN 15% a) Do the following:
o 1) Verify letdown isolation.
o !E letdown is NOT isolated, THEN manually isolate.
o 2) Verify PRZR Heaters are off.
o !E PRZR Heaters are NOT off, THEN put heaters in PTL.
o b) Verify CC system - IN SERVICE o b) Place CC system in service using 1-0P-51.1, Component Cooling System.
o WHEN CC system is in service, THEN perform Step 5c.
o Continue with Step 5d.
(STEP 5 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 7 of 21 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 5. CHECK PRZR LEVEL CONTROL:
(Continued) o c) Verify Letdown - IN SERVICE c) WHEN PRZR level is greater than 15%.
THEN manually put letdown in service:
o 1) Open 1-CH-FCV-1122 to establish at least 25 gpm flow.
o 2) Put 1-CH-PCV-1145 in MAN UAL and open to 100%.
- 3) Open the following Letdown Isolation Valves:
o
- 1-CH-TV-1204A o
- 1-CH-TV-1204B o
- 1-CH-LCV-1460A o
- 1-CH-LCV-1460B
- 4) Open one of the following Letdown Orifice Isolation Valves:
o
- 1-CH-HCV-1200A OR o
- 1-CH-HCV-1200B OR o
- 1-CH-HCV-1200C o 5) Adjust 1-CH-PCV-1145 to establish 300 psig letdown pressure and put in AUTO.
(STEP 5 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 8 of 21 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 5. CHECKPRZRLEVELCONTRO~
(Continued) 0 6) Reset PRZR Heaters GROUP 3 CONTROL.
- 1) IF RWST is aligned to Charging Pump Suction, THEN do the following:
- a. !E VCT level is less than 22%,
THEN do the following:
0 1. WHEN VCT level is greater than 42%, THEN do Step b to align charging pump suction to VCT.
0 2. Continue with Step 5d.
- b. Align charging pump suction to VCT:
- 1. Open Charging Pump Suction From VCT Isolation Valves:
0
- 1-CH-MOV-1115C 0
- 1-CH-MOV-1115E
- 2. Close Charging Pump Suction From RWST Isolation Valves:
0
- 1-CH-MOV-1115B 0
- 1-CH-MOV-1115D 0 d) PRZR level - BETWEEN 20% AND 29% 0 d) Control charging and letdown to maintain level between 20% and 29%.
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 9 of 21 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 6. CHECK PRZR PRESSURE CONTROL:
o a) Pressure - GREATER THAN 1780 PSIG o a) Verify SI Actuation.
o !E SI is NOT actuated, THEN manually actuate SI.
o GO TO 1-E-0, REACTOR TRIP OR SAFETY INJECTION, STEP 1.
o b) Pressure - STABLE AT OR TRENDING TO b) !E pressure is less than 2235 psig and 2235 PSIG decreasing, THEN:
o 1) Verify PRZR PORVs are closed.
o !E NOT, THEN manually close.
o !E any valve cannot be closed, THEN manually close its Block Valve.
o 2) Verify PRZR Spray Valves are closed.
o IF NOT, THEN manually close using controller.
o Verify PRZR spray valves - CLOSED.
!E NOT, THEN place failed valve remote close switch in CLOSE:
o
- 1-RC-SOV-1455A, 1-RC-PCV-1455A REMOTE CLOSE SOV o
- 1-RC-SOV-1455B, 1-RC-PCV-1455B REMOTE CLOSE SOV (STEP 6 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 10 of 21 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 6. CHECK PRZR PRESSURE CONTROL:
(Continued)
!E spray valves can NOT be closed, THEN do the following:
!E 1-RC-PCV-1455A failed open, THEN:
o
- Stop 1-RC-P-1 C.
o
- Stop 1-RC-P-1 A.
!E 1-RC-PCV-1455B failed open, THEN:
o
- Stop 1-RC-P-1 C.
o * !E 1-RC-P-1A is running, THEN stop 1-RC-P-1 B.
o 3) Verify PRZR Heaters are on.
o !E NOT, THEN manually energize heaters.
!E pressure is greater than 2235 psig and increasing, THEN:
o 1) Verify PRZR Heaters are off.
o !E NOT, THEN put heaters in PTL.
o 2) Control pressure using normal PRZR spray.
o !E normal PRZR spray is NOT available AND letdown is in service, THEN use Auxiliary Spray.
o IF normal PRZR spray is NOT available AND letdown is NOT in service, THEN use PRZR PORV(s).
(
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 11 of 21 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7. VERIFY ALL IRPls - 10 STEPS OR LESS IF TWO or more IRPls indicate greater than 10 STEPS, THEN emergency borate as follows:
D a) Place the in-service Boric Acid Transfer Pump in FAST.
D b) Open 1-CH-MOV-1350, Emergency Borate Valve.
c) Verify Emergency Boration flow:
D
- Emergency Boration Flow - 35 GPM OR GREATER D
- On-service Boric Acid Tank Level -
DECREASING D d) !E Emergency Boration flow NOT verified, THEN locally open 1-CH-MOV-1350 and verify flow.
e) IF Emergency Boration flow STILL NOT verified, THEN do the following:
D 1) Place Blender Mode switch in BORATE.
D 2) Place Blender Control switch in START.
D 3) Fully open 1-CH-FCV-1113A.
D 4) Close 1-CH-FCV-1113B.
D 5) Locally open 1-CH-241 , Manual Emergency Borate Valve.
(STEP 7 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1-ES-O.1 REACTOR TRIP RESPONSE PAGE 12 of 21 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7. VERIFY ALL IRPls - 10 STEPS OR LESS (Continued) o f) Record the following:
- Time Emergency Boration started: _
- Initial on-service BAST level: _
o g) Initiate ATTACHMENT 2, EMERGENCY BORATION FOR CONTROL RODS NOT FULLY INSERTED, to determine when emergency boration can be secured.
o h) Have the SRO refer to Tech Spec 3.1.1.
- 8. VERIFY ADEQUATE HP TURBINE GLAND o Throttle 1-MS-MOV-106, GLAND STEAM STEAM PRESSURE ON 1-MS-PI-131 DUMP BYPASS VALVE.
o !E gland steam pressure can NOT be increased, THEN throttle open 1-MS-198, Gland Steam Supply Header 1-MS-PCV-120 Bypass Valve, to control pressure on 1-MS-PI-118 between 1.5 and 5 psig.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-ES-0.1 2 EMERGENCY BORATION FOR CONTROL RODS NOT FULLY REVISION INSERTED PAGE 26 1 of 2 NOTE: If 1-CH-241 is used as the flow path, then the boration amount should be verified by the change in BAST level or by a 1-PT-1 0 series procedure.
- 1. Determine conditions to stop Emergency Boration:
a) Determine total Equivalent Stuck Rods using the following table:
Record IRPllDs Convert Actual Record ActualiRPI for IRPls IRPI to Equivalent Equivalent Indication indicating NOT Stuck Rods Stuck Rod fully inserted (EQSR): Subtotals:
Any Rod >20 steps 1 rod = 1 EQSR 1-5 rods = 1 EQSR Rods indicating 6-9 rods = 2 EQSR 11-20 (inclusive) 10- 16 rods = 3 EQSR steps withdrawn 17 - 32 rods = 4 EQSR 33 or more = 5 EQSR Total Equivalent Stuck Rods:
D b)!E ONLY ONE Total Equivalent Stuck Rod was recorded in Step 1a table, THEN GO TO ATTACHMENT 2, EMERGENCY BORATION FOR CONTROL RODS NOT FULLY INSERTED, Step 2 to stop Emergency Boration.
c) IF TWO or more Total Equivalent Stuck Rods were recorded in Step 1a table, THEN monitor for one of the following conditions to stop Emergency Boration:
~
D
- 25 minutes for each Equivalent Stuck Rod has elapsed.
OR D
- 15% BAST Level for each Equivalent Stuck Rod has been inserted.
OR D
- Adequate shutdown margin has been verified using a 1-PT-1 0 series procedure.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-ES-O.1 2 EMERGENCY BORATION FOR CONTROL RODS NOT FULLY REVISION PAGE INSERTED 26 2 of 2
- 2. WHEN Emergency Boration is no longer required, THEN stop Emergency Boration as follows:
0 a) Place Boric Acid Transfer Pump in AUTO.
b) Close valves that were opened:
0
- 1-CH-MOV-1350 0
- 1-CH-241 c) Ensure the following valves are in AUTO:
0
- 1-CH-FCV-1113A 0
- 1-CH-FCV-1113B 0 d) Place Blender in AUTO or OFF.
e) Record the following:
(~
0
- Time Emergency Boration stopped:
0
- Final on-service BAST level:
- 3. Return to procedure and step in effect.
- END-
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS Unit is stable at 100% power.
Quarterly control rod operability test was completed 20 minutes ago.
Control bank A control rod P-10 is at 217 steps, as indicated by individual rod position.
1-AP-1.3, Control Rod Out of Alignment, is complete up to the point of aligning the rod using Attachment 2, Realigning Control Rod--Rod"Low.
The SRO desires the affected rod be withdrawn in 3 step increments during realignment.
John Leake has been briefed with Attachment 2 and is standing by to perform any actions required outside the control room.
INITIATING CUE You are requested to realign rod P-10 by performing Attachment 2, Realigning Control Rod--Rod Low, of 1-AP-1.3.
02/25/08 Page: 1 of 12
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM R477 (modified)
Respond to a misaligned control rod (1-AP-1.3).
TASK STANDARDS The misaligned rod P-1 0 is withdrawn at least 1 rod pull (3 steps) per Attachment 2 of 1-AP-1.3; Upon observing 2 control rods dropped a manual reactor trip is initiated as required by 1-AP-1.2; During performance of 1-E-O, the failure of SI to automatically actuate is identified and SI is manually initiated.
KIA
REFERENCE:
013-A4.03 (4.5/4.7)
ALTERNATE PATH:
YES TASK COMPLETION TIMES Validation Time = 22 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ ] SATISFACTORY [ ] UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature /
Date EVALUATOR'S COMMENTS 02/25/08 Page: 2 of 12
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM R477 (modified)
READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES The trainee has completed the applicable course knowledge training at the Reactor Operator level.
INITIAL CONDITIONS Unit is stable at 100% power.
Quarterly control rod operability test was completed 20 minutes ago.
Control bank A control rod P-10 is at 217 steps, as indicated by individual rod position.
1-AP-1.3, Control Rod Out of Alignment, is complete up to the point of aligning the rod using Attachment 2, Realigning Control Rod--Rod Low.
02/25/08 Page: 3 of 12
The SRO desires the affected rod be withdrawn in 3 step increments during realignment.
John Leake has been briefed with Attachment 2 and is standing by to perform any actions required outside the control room.
INITIATING CUE You are requested to realign rod P-10 by performing Attachment 2, Realigning Control Rod--Rod Low, of 1-AP-1.3.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT Copy of 1-AP-1.3 signed off to the point of performing Attachment 2, Realigning Control Rod--Rod Low.
PERFORMANCE STEPS START TIME 02/25/08 Page: 4 of 12
Note: Step 1 of Att. 2 is N/A since rod is not in controlling bank.
[C] Record the affected bank position. IProcedure Step ISAT [] UNSAT [ ]
IL...Z;S;,;;ta;;;;n,;,;;d=a;;,;rd;;,;;s~_ _1 Records CSA 229 steps in procedure.
Notes/Comments Control bank A group step counters are reading 229 steps (value corresponds to the ARO position).
~ Record the misaligned rod position. IProcedure Step ISAT [] UNSAT [ ]
IStandards IRecords rod P-10 @ 217 steps.
rotes/comments
((::=J Record the number of steps that the rod is misaligned. I Procedure Step ISAT[] UNSAT[]
IStandards IRecords rod P-10 misaligned by 12 steps.
INotes/Comments 02/25/08 Page: 5 of 12
II::=:J Place the control rod bank selector switch in BANK SELECT. IProcedure Step ICritical Step ISAT [] UNSAT [ ]
IL,;;S;,;;ta;;;;n,;,;;d;,;;;a;;,,;rd;;,;;s~_ _1 Rod control selector switch positioned to CONTROL BANK A position.
INotes/Comments Manually adjust the group step counter to the actual position of the Procedure Step _ _
misali ned rod.
ICritical Step ISAT [] UNSAT []
Standards INotes/Comments Locally record the pulse-to-analog converter reading for control Procedure Step _ _
bank A.
ISAT [] UNSAT [ ]
I Standards ICalls operator (John Leake) to obtain reading per Step 7 of AU. 2 Notes/Comments Expected reading of 229 steps provided by booth operator.
02/25/08 Page: 6 of 12
~ Locally reset the pulse-to-analog converter for control bank A. IProcedure Step ISAT [1 UNSAT [ 1 Standards Verbal-visual If operator questions rod bank 10 limit alarm tell him as the SRO that it cues will be evaluated and to continue with the task.
Notes/Com ments The bank A rod bank 10 limit alarm will come in when the PIA converter is pulsed down. The operator should acknowledge this as an expected alarm. The recorder on the vertical panel may be checked, this recorder is calibrated in % NOT steps.
Open all lift coil disconnect switches for the affected bank, except Procedure Step _ _
for the switch for the misali ned rod.
ICritical Step ISAT [1 UNSAT [1 Standards All lift coil disconnect switches for control bank A are open EXCEPT for rod P-10.
Demonstration The balance-of-plant operator has the board while the operator Cues ositions lift coil disconnects.
Verbal-Visual The balance-of-plant operator has the board while the operator Cues ositions lift coil disconnects.
I Notes/Comments 02/25/08 Page: 7 of 12
[L] Have an extra operator independently verify lift coil disconnects. IProcedure Step ISAT[] UNSAT[]
IStandards IRequests extra operator (e.g. BOP) to perform step 10 of attachment 2. I Demonstration The balance-of-plant operator has completed step 10 of attachment 2.
Cues Verbal-Visual The balance-of-plant operator has completed step 10 of attachment 2.
Cues INotes/Comments
~ Manually withdraw the affected control rod. IProcedure Step ICritical Step ISAT [] UNSAT [ ]
IStandards IOperator begins withdrawal of rod P-10 in 3 step increments.
Notes/Comments Based on rod height and power level no actions should be necessary for maintaining temperature or Reactor power; Rod Control urgent failure alarm will annunciate as noted in the procedure.
( 02/25/08 Page: 8 of 12
11 Respond to dropped control rods (1-AP-1.2) Procedure Step _ _
Alternate path step, Step 1 of 1-AP-1.2 has the operator verify ONLY ONE CONTROL ROD DROPPED, the RNO is to go to 1-E-O (i.e. trip the reactor).
ICritical Step ISAT [] UNSAT []
Standards Operator identifies more than 1 RCCA dropped and initiates manual reactor tri er 1-AP-1.2, Oro ed Rod, Ste 1 RNO.
Notes/Comments Tripping the reactor manually before an automatic trip signal is received is not critical - the operator only needs to realize that one is required by the procedure if it occurs the actions to actuate the reactor trip switch, manually trip the turbine, etc. are identical.
~I Verify reactor tripped - YES. IProcedure Step ISAT[] UNSAT[]
Standards Operator verifies reactor is tripped (RTBs open, rod bottom lights on, flux decreasin .
I Notes/Comments 02/25/08 Page: 9 of 12
(
13 Verify turbine trip - NO .. Procedure Step _ _
Alternate path step, step 2 of 1-E-O has the operator manually trip turbine (automatic trip signal should have occurred also),
the RNO for this step has the operator manually runback the turbine.
ISAT [] UNSAT [ ]
Standards Attempts to manually trip turbine; based on failure of turbine to manually trip runs back turbine (verifies turbine tripped, stop valves closed) and resets Reheaters.
IN otes/Com ments
~ Verify both ac emergency busses energized - yes. I Procedure Step ISAT [] UNSAT [ ]
IStandards IVerifies 1Hand 1J busses both energized.
INotes/Comments 02/25/08 Page: 10 of 12
15 Check if SI is actuated . Procedure Step _ _
Alternate path step, although the operator remains in the left hand column of the procedure the logic for automatic safety injection was made up and the signal failed to actuate, thus the operator must diagnose and respond to the malfunction, and take manual action in order to effect SI actuation that otherwise would not be required.
I_C_r_iti_c_a_1S_t_e-'-p 1 SAT [1 UNSAT [1 Standards
- Checks low head pumps running- NO.
- Any first out annunciator lit - YES.
- Operator manually actuates SI.
Demonstration JPM is complete once operator states the E-O immediate operator Cues actions are com lete.
Verbal-Visual JPM is complete once operator states the E-O immediate operator Cues actions are com lete.
Notes/Comments Annunciator D-E/3 will be lit due to failure of turbine to trip unless acknowledged prior to performing this step (MSTVs close so condition clears after MSTVs shut).
>>>>> END OF EVALUATION <<<<<
STOP TIME 02/25/08 Page: 11 of 12
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
Respond to a misaligned control rod (1-AP-1.3).
CHECKLIST
_ _ RecalilC #168 (100% power).
_ _ Do Simspray.
Ensure rod banks and TavelTref recorder are correct.
Booth Operator Actions:
Respond as extra operator (John Leake) to perform the following:
Using extreme view report PIA converter reading (229 steps) when directed in Step 7 of Attachment 2.
Using extreme view reset PIA converter for CBA to 217 steps when directed in Step 8 of Attachment 2 and report completion of step to operator (note: adjust to 1 step increments for realism prior to pulsing down).
02/25/08 Page: 120f12
NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 11 1-AP-1.3 CONTROL ROD OUT OF ALIGNMENT (WITH FIVE ATTACHMENTS) PAGE 1 of 6 PURPOSE To provide instructions for recovering from the following misaligned rod events:
- Any Single Control Rod is out of alignment by any number of steps, or
- Multiple Control Rods are out of alignment provided:
- Misaligned Control Rods are all in the same group, and
- All of the misaligned Control Rods are at the same position, and
- All of the misaligned Control Rods are less than 12 steps out of alignment.
ENTRY CONDITIONS This procedure is entered when any of the following conditions exist:
- Any Control Rod is known to be out of alignment from group step counter, or
- Multiple rods in the same group are known to be out of alignment by the same amount from the group step counter demand position, or
- Any group is out of alignment by more than one step from the other group in that bank.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 11 1-AP-1.3 CONTROL ROD OUT OF ALIGNMENT PAGE 2 of 6 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 1. PLACE CONTROL ROD BANK SELECTOR SWITCH IN MANUAL
- 2. PERFORM NOTIFICATIONS:
0 a) Notify SRO 0 b) Notify Operations Manager On Call 0 c) Notify Reactor Engineer 0 d) Notify STA
- 3. VERIFY REACTOR - CRITICAL AND Do the following:
ABOVE THE POINT OF ADDING HEAT o a) !.E Reactor is critical, THEN GO TO 1-E-O, REACTOR TRIP OR SAFETY INJECTION.
b) !.E Reactor is NOT critical, THEN have the following personnel verify that the rod(s) can be withdrawn without achieving criticality using a Kef! calculation, as applicable:
o
- Reactor Engineer o
- NA&F - If Reactor Engineer requires NA&F assistance o
- STA o !.E rod(s) can be withdrawn, THEN obtain permission from the Operations Manager or Operations Manager On Call to withdraw rod(s)
AND GO TO Step 5.
o !.E NOT, THEN GO TO 1-E-O, REACTOR TRIP OR SAFETY INJECTION.
NUMBER PROCEDURE TITLE REVISION 11 1-AP-1.3 CONTROL ROD OUT OF ALIGNMENT PAGE 3 of 6 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: While a Control Rod is being aligned, Tave is controlled by the Main Turbine or Steam Dump System.
- 4. RECORD TIME ROD WAS MISALIGNED ON ATTACHMENT 4, MAXIMUM ROD WITHDRAWAL AND RAMP RATE
- 5. RECORD TIME ROD WAS MISALIGNED IN CRO LOG
- 6. CHECK AFFECTED ROD - MORE o GO TO Step 8.
THAN 12 STEPS FROM GROUP e
"*0 STEP COUNTER DEMAND POSITION
- 7. TAKE ACTIONS REQUIRED BY TECHNICAL SPECIFICATION 3.1.4
- 8. DETERMINE CAUSE OF MISALIGNED ROD
- 9. RODS BEING WITHDRAWN FOR o GO TO Step 11 .
REACTOR STARTUP
- 10. CHECK ROD STATUS:
o a) Rod(s) stuck on bottom at o a) GO TO Step 11 .
zero steps o b) System Engineering requires o b) GO TO Step 11.
alignment using 1-ICM-RCS-G-001, Rod Control System Troubleshooting and Maintenance for Rod Alignment o c) RETURN TO procedure and Step in effect with OMOC concurrence
NUMBER PROCEDURE TITLE REVISION 11 1-AP-1.3 CONTROL ROD OUT OF ALIGNMENT PAGE 4 of 6 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 11. VERIFY THE CAUSE OF D WHEN the cause of misaligned rod is repaired, MISALIGNED ROD - REPAIRED THEN GO TO Step 12.
- 12. RECORD CURRENT POWER LEVEL:
Power level: _
- 13. REACTOR - CRITICAL D GO TO Step 15.
- 14. DETERMINE MAXIMUM ROD WITHDRAWAL RATE AND POWER RAMP RATE USING ATTACHMENT 4, MAXIMUM ROD WITHDRAWAL AND RAMP RATE
- 15. VERIFY NO NUCLEAR D Reset rate trip signal.
INSTRUMENTATION POWER RANGE RATE TRIP SIGNAL - LIT NOTE: More than one rod in the same group may be aligned simultaneously.
- 16. ALIGN AFFECTED RODS USING ONE OF THE FOLLOWING:
D
- ATTACHMENT 3, REALIGNING CONTROL ROD - ROD HIGH
( NUMBER PROCEDURE TITLE REVISION 11 1-AP-1.3 CONTROL ROD OUT OF ALIGNMENT PAGE 5 of 6 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 17. VERIFY PULSE-TO-ANALOG D Notify Instrument Department to resolve CONVERTER READING IS THE discrepancy.
SAME AS THAT RECORDED IN:
D
- ATIACHMENT 2, REALIGNING CONTROL ROD-ROD LOW D
- ATIACHMENT 3, REALIGNING CONTROL ROD - ROD HIGH NOTE: The Bank Overlap Counters are located in the Rod Drive Logic Cabinet.
- 18. VERIFY PROPER READING OF BANK OVERLAP COUNTER:
D a) Perform ATIACHMENT 5, VERIFYING PROPER CONTROL ROD BANK OVERLAP D b) Verify reading - CORRECT b) Do the following:
D 1) Notify Instrument Department to resolve discrepancy.
- 2) Notify Reactor Engineer to evaluate the need to perform incore flux mapping as described by the following Technical Specifications:
D
- 3.1.4 D
- 3.2.1 D
- 3.2.2
NUMBER PROCEDURE TITLE REVISION 11 1-AP-1.3 CONTROL ROD OUT OF ALIGNMENT PAGE 6 of 6 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 19. PLACE CONTROL ROD BANK SELECTOR SWITCH IN MANUAL
- 20. VERIFY STEAM DUMPS - NOT o Reset Steam Dump arming signal.
ARMED
- 21. CHECK ALL INSTRUMENT o Notify Instrument Department to reset instrument SETPOINTS CHANGED BECAUSE setpoints.
OF TECHNICAL SPECIFICATION ACTION STATEMENTS - RETURNED TO NORMAL VALUES
- 22. CHECK TAVE - WITHIN 1.5 of OF Restore Tave to within 1.5 of of Tref by doing the TREF following:
- Manually adjust Turbine Load using one of the following procedures:
o
- 1-0P-2.2, Unit Power Operation from Mode 1 to Mode 2 o
- 1-0P-2.1, Unit Startup from Mode 2 to Mode 1 o
- Manually operate Control Rods.
o
- Manually borate/dilute the RCS.
- 23. DO ONE OF THE FOLLOWING:
o
- Place Control Rod Bank selector switch in AUTO OR o
- Leave Control Rod Bank selector in MANUAL
- END-
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-1.3 1 REFERENCES REVISION PAGE 11 1 of 1
- Technical Specifications:
3.1.1 3.2.1 3.2.2 3.2.4 3.1.4
- NE Technical Report No. 417, October 1984
- SOER 84-02, CONTROL ROD MISPOSITIONING
- Memo from J.O. Erb to C.G. Meyer, dated January 21, 1991
- CTS 02-89-2803-002
- ET No. NAF-970123, Rev 0, Control Rod Withdrawal Rate Limits For Recovery Of Misaligned Control Rods During Unit Startup (Rev 5)
- 1-0P-2.2, Unit Power Operation From Mode 1 To Mode 2
- 1-0P-2.1, Unit Startup From Mode 2 To Mode 1
- ET-NAF-2004-0012, Rev 0, Control Rod Withdrawal Guidelines from Framatome for Misaligned/Dropped Rods at North Anna (Rev 7)
- Plant Issue N-2005-1682 addressing rod withdrawal below 75% and greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for ET-NAF-2004-0012 (Rev 8)
- DCP 05-125, Replacement of Step Counters/NAPS/Units 1 & 2 (Attachment 2 & 3, Step 6 Note, Rev 9)
- CA009474, CR011496, Nonconservative Actions (Step 3, Rev 10)
- 1-ICM-RCS-G-001, Rod Control System Troubleshooting And Maintenance
- CA016898, CR01 9402, Unit 1 RFO Readiness Focused Assessment (07-30-N)
Recommendations (Step 9 & 10, Rev 11)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-1.3 2 REALIGNING CONTROL ROD REVISION - ROD LOW PAGE 11 1 of 4
- 1. !.E the misaligned rod(s) are in the controlling bank, THEN position rods to place both groups in that bank at the same reading.
- 2. Record affected bank position:
Gt I gt 11 ru~f1 Bank Steps
- 3. Record the misaligned rod(s) positions:
ROD(s) STEPS
- 4. Record the number of steps the rod(s) are misaligned:
ROD(s) STEPS MISALIGNED
- 5. Place the Rod Control Bank Selector switch in BANK SELECT for the bank containing the misaligned rod(s).
NOTE: A flashing Group Step Counter display indicates low battery.
- 6. Manually adjust the Group Step Counter for the affected group to the actual position of the misaligned rod(s) as recorded in Step 3, using the down (DN) or (UP) button, as required.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-1.3 2 REALIGNING CONTROL ROD - ROD LOW REVISION PAGE 11 2 of 4 NOTE: The shutdown banks do not have Pulse-to-Analog Converters.
- 7. Locally record the affected banks Pulse-to-Analog Converter readings (located in 1-EI-CB-41 Bin the Instrument Rack Room): steps NOTE: If the Pulse to Analog Converter is pulsed to zero, then the affected banks ROD BANK LO/
LO-LO LIMIT annunciator will alarm.
- 8. Locally reset the affected banks Pulse-to-Analog Converter by doing the following (located in the Instrument Rack Room):
_ a) Place the Manual/Automatic switch in MANUAL.
_ b) Place the Pulse-to-Analog Bank Selector switch to the affected bank.
_ c) Press the Down pulse button until the Pulse-to-Analog Converter reads the position of the misaligned rod(s).
( d) Place the Manual/Automatic switch in AUTOMATIC.
- 9. Open all Lift Coil Disconnect switches for the affected bank, except for the misaligned rod(s)
(located behind Main Control Room Vertical Board).
- 10. Have a second person independently verify that all Lift Coil Disconnect switches for the affected bank, except for the misaligned rod(s), are open.
- 11. Do the following during the Control Rod motion:
- Adjust Turbine load, as required at a rate consistent with the Tave change caused by rod motion AND ATTACHMENT 4, MAXIMUM ROD WITHDRAWAL AND RAMP RATE, as applicable.
AND
- Maintain Tave within 1.5 OF of Tref by adjusting Turbine load or Steam Dumps as necessary.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-1.3 2 REALIGNING CONTROL ROD - ROD LOW REVISION PAGE 11 3 of 4 CAUTION:
- Exceeding the maximum withdrawal rate calculated on ATIACHMENT 4 could cause fuel damage.
- If the Reactor is subcritical, then the Reactor must be monitored for inadvertent criticality during control rod withdrawal.
NOTE: When the affected rod(s) are withdrawn, then Annunciator Panel "A" 0-1, ROD CONTROL URGENT FAILURE, may annunciate, indicating the affected banks lift coils are de-energized.
- 12. Manually withdraw the affected rod(s) by placing the Rod Control switch in OUT:
a) Verify the OUT direction lamp is LIT.
b) Verify the affected Group Step Counter indicates outward motion.
c) IF affected rod(s) will NOT move, THEN have the System Engineer determine required actions AND obtain OMOC concurrence, before continuing with this procedure.
- 13. !E the misaligned rod is in Group 1 of the affected bank, THEN do the following:
a) Withdraw the Control Rod until it reaches a value of one step greater than the value recorded in Step 2 of this attachment.
b) Drive the Control Rod in one step to the value recorded in Step 2 of this attachment.
- 14. !E the misaligned rod is in Group 2 of the affected bank, THEN withdraw the Control Rod until the affected Group Step Counter reaches the value recorded in Step 2 of this attachment.
- 15. Record affected Group Step Counter steps: steps
- 16. Verify the following conditions are met:
- All rods in the affected bank are at the same height.
AND
- All Rod Bottom lights in the affected bank are NOT LIT.
- !E either condition is NOT satisfied, THEN have the System Engineer determine required actions AND obtain OMOC concurrence, before continuing with this procedure.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-1.3 2 REALIGNING CONTROL ROD - ROD LOW REVISION PAGE 11 4 of 4
- 17. Close all Lift Coil Disconnect switches.
- 18. Have a second qualified Operator verify that all lift Disconnect switches are closed.
- 19. Reset ROD CONTROL URGENT FAILURE alarm from the control board with the Alarm Reset pushbutton.
- 20. Step the affected banks Control Rods in one step and verify proper Group 2-Group 1 sequencing.
- 21. Step the affected banks Control Rods out one step and verify proper Group 1-Group 2 sequencing.
- 22. RETURN TO Step 17 of 1-AP-1.3, CONTROL ROD OUT OF ALIGNMENT.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-1.3 3 REALIGNING CONTROL ROD REVISION - ROD HIGH PAGE 11 1 of 4
- 1. !.E the misaligned rod(s) are in the controlling bank, THEN position rods to place both groups in that bank at the same reading.
- 2. Record affected bank position:
Bank Steps
- 3. Record the misaligned rod(s) positions:
ROD(s) STEPS
- 4. Record the number of steps the rod(s) are misaligned:
ROD(s) STEPS MISALIGNED
- 5. Place the Rod Control Bank Selector switch in BANK SELECT for the bank containing the misaligned rod(s).
NOTE: A flashing Group Step Counter display indicates low battery.
- 6. Manually adjust the Group Step Counter for the affected group to the actual position of the misaligned rod(s) as recorded in Step 3, using the (UP) or down (ON) button, as required.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-1.3 3 REALIGNING CONTROL ROD - ROD HIGH REVISION PAGE 11 2 of 4 NOTE: The shutdown banks do not have Pulse-to-Analog Converters.
- 7. Locally record the affected banks Pulse-to-Analog Converter readings (located in 1-EI-CB-41 Bin the Instrument Rack Room): steps
- 8. Locally reset the affected banks Pulse-to-Analog Converter by doing the following (located in the Instrument Rack Room):
_ a) Place the Manual/Automatic switch in MANUAL.
b) Place the Pulse-to-Analog Bank Selector switch to the affected bank.
c) Press the Up pulse button until the Pulse-to-Analog Convertor reads the position of the misaligned rod(s).
_ d) Place Manual/Automatic switch in AUTOMATIC.
- 9. Open all Lift Coil Disconnect switches for the affected bank, except for the misaligned rod(s)
(located behind Main Control Room Vertical Board).
- 10. Have a second person independently verify that all Lift Coil Disconnect switches for the affected bank, except for the misaligned rod(s), are open.
- 11. Do the following during the Control Rod motion:
- Adjust Turbine load, as required at a rate consistent with the Tave change caused by rod motion AND ATTACHMENT 4, MAXIMUM ROD WITHDRAWAL AND RAMP RATE, as applicable.
AND
- Maintain Tave within 1.5 of of Tref by adjusting Turbine load or Steam Dumps as necessary.
( NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-1.3 3 REALIGNING CONTROL ROD - ROD HIGH REVISION PAGE 11 3 of 4 NOTE: When the affected rod(s) are inserted, then Annunciator Panel "An 0-1, ROD CONTROL URGENT FAILURE, may annunciate, indicating the affected bank lift coils are de-energized.
- 12. Manually drive the affected rod(s) by placing the In/Hold/Out switch in IN:
_ a) Verify the IN direction lamp is LIT.
b) Verify the affected Group Step Counter indicates inward motion.
c) IE affected rod(s) will NOT move, THEN have the System Engineer determine required actions AND obtain OMOC concurrence, before continuing with this procedure.
- 13. Drive the rod two steps below the affected bank position.
- 14. IF the misaligned rod is in Group 1 of the affected bank, THEN do the following:
_ a) Withdraw the Control Rod until it reaches a value of one step greater than the value recorded
( in Step 2 of this attachment.
b) Drive the Control Rod in one step to the value recorded in Step 2 of this attachment.
- 15. IF the misaligned rod is in Group 2 of the affected bank, THEN withdraw the Control Rod until the affected Group Step Counter reaches the value recorded in Step 2 of this attachment.
- 16. Record affected Group Step Counter steps: steps
- 17. Verify the following conditions are met:
- All rods in the affected bank are at the same height.
AND
- All Rod Bottom lights in the affected bank are NOT LIT.
- IF either condition is NOT satisfied, THEN have the System Engineer determine required actions AND obtain OMOC concurrence, before continuing with this procedure.
- 18. Close all Lift Coil Disconnect switches.
- 19. Have a second qualified Operator independently verify that all Lift Coil Disconnect switches are closed.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-1.3 3 REALIGNING CONTROL ROD - ROD HIGH REVISION PAGE 11 4 of 4
- 20. Reset ROD CONTROL URGENT FAILURE alarm from the control board with the Alarm Reset pushbutton.
- 21. Step the affected banks Control Rods in one step and verify proper Group 2-Group 1 sequencing.
- 22. Step the affected banks Control Rods out one step and verify proper Group 1-Group 2 sequencing.
- 23. RETURN TO Step 17 of 1-AP-1.3, CONTROL ROD OUT OF ALIGNMENT.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-1.3 4 MAXIMUM ROD WITHDRAWAL AND RAMP RATE REVISION PAGE 11 1 of 3
- 1. Time rod misaligned: _
NOTE: Power may be reduced and maintained at ~ 75 percent to allow a less restrictive rod withdrawal rate.
- 2. Using the time recorded in Step 1, determine the maximum Rod Withdrawal Rate, from the table below:
Total Time from RCCA Amount of Maximum Power Level to Misaligned to RCCA Misaligned Rod Withdrawal Recover RCCA Recovery Completion Withdrawal Rate
< 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 0-100% unrestricted
~ 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
~ 75% unrestricted and ~ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Any Rod Misalignment 5 Step increments
> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ~ 75%
spaced every 30 minutes First 100 Steps > 75% op5 Steps/h r
~ 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Remainder of withdrawal beyond > 75% ~Steps/hr First 100 Steps
_ a) IF < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, THEN Rod Withdrawal Rate is unrestricted, unless 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is exceeded.
_ b) IF ~ 75% power AND ~ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> misaligned, THEN Rod Withdrawal Rate is unrestricted, unless 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is exceeded.
_ c) IF ~ 75% power AND> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> misaligned, THEN Rod Withdrawal Rate is limited to 5 Step increments spaced every 30 minutes apart.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-1.3 4 MAXIMUM ROD WITHDRAWAL AND RAMP RATE REVISION PAGE 11 2 of 3 NOTE: Steps per Hour = 0.5/P (for first 100 Steps)
Steps per Hour = 1/P (for remainder beyond first 100 Steps)
Where: P = fraction of rated power If the RCCA is misaligned by less than 100 Steps, then the withdrawal rate for the first 100 Steps apply for the entire withdrawal.
Whole Steps Only:
Example for first 100 Steps: 100% power = 0.5 Steps/hour which becomes 1 Step every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to maintain a whole step.
Rate is not cumulative, meaning that there is no benefit for postponing rod movement:
2 Steps following a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold is not allowed.
_ d) !E: > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND power is> 75%, THEN determine and record Rod Withdrawal Rates, as follows:
o 1) First -0.5 = -0.5 =
100 Steps: P _ _ _ _ _ Steps/hour Whole Step Withdrawal Rate: _
Steps per Unit Time o 2) Withdrawal Beyond First 100 Steps: 1 = 1 =
P _ _ _ _ _ Steps/hour Whole Step Withdrawal Rate: _
Steps per Unit Time
( NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-1.3 4 MAXIMUM ROD WITHDRAWAL AND RAMP RATE REVISION PAGE 11 3 of 3 NOTE: !E recovery completion time OR required power level changes during Rod recovery, THEN any new maximum Ramp Rate, shall be applied.
- 3. Using the time recorded in Step 1, determine the maximum Ramp Rate based on power level, from the table below:
Total Time from RCCA Maximum Misaligned to RCCA Power Range Core Power Recovery Ramp Rate
< 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 0-100% unrestricted 2:: 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and:S; 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0-100% :s; 30% / hr 0-50% :s; 30% / hr
> 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and:S; 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 50 - 90% :s; 15% / hr 90 - 100% :s; 5% / hr
> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0-100% :s; 3% / hr o Record the applicable Power Range(s): Percent o Record the maximum Ramp Rate(s): Percent/hour
- 4. IF Rod withdrawal OR Ramp Rate restrictions are imposed by 1-0P-2.1, Unit Startup from Mode 2 to Mode 1, THEN use the more restrictive values between procedures.
- END-
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-1.3 5 VERIFYING PROPER CONTROL ROD BANK OVERLAP REVISION PAGE 11 1 of 1
- 1. Record the Bank Overlap Counter value displayed in the Rod Control logic Cabinet
- 2. Record the step Counter value for the controlling Control Rod Bank:
_ _ _ _ _ Steps
- 3. Record the difference in the values recorded in Step 1 and Step 2:
- 4. Identify which Bank is controlling:
_ _ _ _ _ Bank
- 5. !E 0 Bank is the controlling bank, THEN verify that the value recorded in Step 3 is 384.
- 6. IF C Bank is the controlling bank, THEN verify that the value recorded in Step 3 is 256.
- 7. !E B Bank is the controlling bank, TH EN verify that the value recorded in Step 3 is 128.
- 8. IF A Bank is the controlling bank, THEN verify that the value recorded in Step 3 is O.
- END-
NUMBER PROCEDURE TITLE REVISION 13 1-AP-1.2 DROPPED ROD PAGE 2 of 7 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED 1 1_ VERIFY ONLY ONE CONTROL GO TO 1-E-0, REACTOR TRIP OR SAFETY ROD-DROPPED INJECTION.
[ 2 1_ PLACE CONTROL ROD BANK SELECTOR SWITCH IN MANUAL
- 3. PERFORM NOTIFICATIONS:
D a) Notify SRO D b) Notify Operations Manager On Call D c) Notify Reactor Engineer D d) Notify STA
- 4. VERIFY REACTOR - CRITICAL AND D GO TO 1-E-0, REACTOR TRIP OR SAFETY ABOVE THE POINT OF ADDING INJECTION.
HEAT
TEMPERATURE - 541 0 FOR GREATER D
- Increase each RCS Loop Tave to at least 541 0 F by diluting the RCS or reducing Turbine Loads using 1-0P-2.2, Unit Power Operation from Mode 1 to Mode 2.
NORTH ANNA POWER STATION EMERGENCY PROCEDURE NUMBER PROCEDURE TITLE REVISION 36 1-E-O REACTOR TRIP OR SAFETY INJECTION PAGE (WITH SEVEN ATTACHMENTS) 1 of 24 PURPOSE This procedure provides actions to verify proper response of the automatic protection systems following manual or automatic actuation of a Reactor trip or Safety Injection, to assess plant conditions, and to identify the appropriate recovery procedure.
ENTRY CONDITIONS
- 1) The following are symptoms that require a Reactor trip, if one has not occurred:
- A Reactor protection system setpoint has been exceeded
- A Turbine protection system setpoint with power greater than P-8 setpoint
- 2) The following are symptoms of a Reactor trip:
- Any Reactor trip first out Annunciator - LIT
- Reactor Trip and Bypass Breakers - OPEN
- Rod Bottom Lights - LIT
- Neutron flux - DECREASING
- 3) The following are symptoms that require a Reactor trip and Safety Injection, if one has not occurred:
- Low PRZR pressure
- High Containment pressure
- Steamline differential pressure
- High steamflow with 10-10 Tave
- High steamflow with low steam pressure
- 4) The following are symptoms of a Reactor trip and Safety Injection:
- Any SI first out Annunciator - LIT
- Any Low-Head SI Pumps - RUNNING
- 5) Transition from another plant procedure.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 36 1-E-O REACTOR TRIP OR SAFETY INJECTION PAGE 2 of 24 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1 ] _ VERIFY REACTOR TRIP: 0 !.E Reactor will NOT trip, THEN GO TO 1-FR-S.1, RESPONSE TO NUCLEAR 0 a) Manually Trip Reactor POWER GENERATION/ATWS, STEP 1.
b) Check the following:
0
- Reactor Trip and Bypass Breakers -
OPEN 0
- Rod Bottom Lights - LIT 0
- Neutron flux - DECREASING 2 ] _ VERIFY TURBINE TRIP:
0 a) Manually Trip Turbine 0 b) Verify all Turbine Stop Valves - CLOSED 0 b) Put both EHC Pumps in PTL.
0 !.E Turbine is still NOT tripped, THEN manually run back Turbine.
0 !.E Turbine cannot be run back, THEN close MSTVs and Bypass Valves.
0 c) Reset Reheaters 0 d) Verify Generator Output Breaker - OPEN 0 d) !.E Generator Output Breaker does NOT open after 30 seconds, THEN manually open G-12 AN D Exciter Field Breaker.
NUMBER PROCEDURE TITLE REVISION 36 1-E-O REACTOR TRIP OR SAFETY INJECTION PAGE 3 of 24 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED 3 ]_ VERIFY BOTH AC EMERGENCY Do the following:
BUSSES - ENERGIZED o a) IF no AC Emergency Bus is energized, THEN immediately restore power to at least one AC Emergency Bus.
o !.E power cannot be restored, THEN GO TO 1-ECA-O.O, LOSS OF ALL AC POWER, STEP 1.
o b) Try to restore power to de-energized AC Emergency Bus using O-AP-10, LOSS OF ELECTRICAL POWER, as time permits.
o Continue with Step 4.
NUMBER PROCEDURE TITLE REVISION 36 1-E-O REACTOR TRIP OR SAFETY INJECTION PAGE 4 of 24 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED 4 ]_ CHECK SI STATUS:
a) Check if SI is actuated: a) Check if SI is required as indicated by any of the following:
o
- Low-Head SI Pumps - RUNNING o
- Low PRZR pressure o
- Any SI First-Out Annunciator - LIT o
- High Containment pressure o
- Steamline differential pressure
- High steamflow with either:
- La-La Tave OR
- Low steam pressure
!.E SI required, THEN GO TO Step 4b.
!.E SI is NOT required, THEN GO TO 1-ES-O.1, REACTOR TRIP RESPONSE, STEP 1.
o b) Manually actuate SI
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS Unit 1 is at 100% power with no equipment out of service.
Unit 2 operator is available to supply pes data as needed.
INITIATING CUE You are requested to perform 1-PT-44A, Pressurizer Heater Output Determination. If any acceptance criteria are not met then applicable Technical Specification Action requirements are identified.
02/25/08 Page: 1 of 10
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM NEW Perform 1-PT-44A, Pressurizer Heater Output Determination. If any acceptance criteria are not met then applicable Technical Specification Action requirements are identified.
TASK STANDARDS Activity is conducted in accordance with the applicable procedure.
Tech Spec Action 3.4.9.8. (72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action is identified).
KIA
REFERENCE:
010 -A4.02, (3.6/3.4)
ALTERNATE PATH:
Yes TASK COMPLETION TIMES Validation Time = 17 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ 1SATISFACTORY [ 1UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature /
Date EVALUATOR'S COMMENTS 02/25/08 Page: 2 of 10
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM NEW READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES The trainee has completed the applicable course knowledge training at the reactor operator level.
INITIAL CONDITIONS The Unit is at 100% power with no equipment out of service.
Unit 2 operator is available to supply PCS data as needed.
INITIATING CUE 02/25/08 Page: 3 of 10
You are requested to perform 1-PT-44A, Pressurizer Heater Output Determination. If any acceptance criteria are not met then applicable Technical Specification Action requirements are identified.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT None PERFORMANCE STEPS START TIME Obtain controlled copy of procedure from Documentum and review Procedure Step _ _
recautions and limitations.
ISAT [1 UNSAT [ 1 IStandards I Precautions and limitations have been reviewed.
Notes/Comments Obtaining and verifying a controlled copy of a procedure is a generic conduct of operations task (2.1.21, 3.5/3.6). If desired by the evaluator a blank copy may be provided to the operator rather than having the operator obtain the copy himself.
02/25/08 Page: 4 of 10
Record the status (AUTO or ON) of the control switch for Procedure Step _ _
Pressurizer Heaters Grou 4-Backu .
ISAT[] UNSAT[]
Standards Heater status determined to be AUTO based on output and/or status Ii ht on vertical anel.
INotes/Comments Procedure Step _ _
ICritical Step ISAT [] UNSAT []
IStandards ISwitch taken to on to lock on applicable set of heaters.
IN otes/Com ments 02/25/08 Page: 5 of 10
~I Record data from PCS. IProcedure Step ISAT [1 UNSAT [ 1 I.,;;;S;;;ta;;,;n.:.;d;;,;;a;,;"rd;;;;s~ 1 Values recorded in procedure.
Demonstration *As the additional operator tell the operator the value for the Unit 2 PCS Cues display; use value displayed on Sim PCS if performed on simulator, if simulated in plant provide a value of 290 kW.
Verbal-Visual *As the additional operator tell the operator the value for the Unit 2 PCS Cues display; use value displayed on Sim PCS if performed on simulator, if simulated in plant provide a value of 290 kW.
Notes/Comments
- if operator pages Unit 2 the booth operator will provide the information.
Return control switch for Pressurizer Heaters Group 4-Backup to Procedure Step _ _
the status recorded in Ste 6.1.1.
ISAT [1 UNSAT [ 1 Standards Momentarily place control switch for Pressurizer Heaters Group 4-Backu in OFF.
Demonstration If operator requests input from SRO then respond as SRO that it is Cues desired to return heaters to the status recorded in Ste 6.1.1.
Verbal-Visual If operator requests input from SRO then respond as SRO that it is Cues desired to return heaters to the status recorded in Ste 6.1.1.
INotes/Comments 02/25/08 Page: 6 of 10
Record the status (AUTO or ON) of the control switch for Procedure Step _ _
Pressurizer Heaters Grou 1-Backu .
ISAT [1 UNSAT [ 1 Standards Heater status determined to be AUTO based on output and/or status Ii ht on vertical anel.
rotes/comm ents Procedure Step _ _
ICritical Step ISAT [1 UNSAT [1 IStandards ISwitch taken to on to lock on applicable set of heaters.
Notes/Comments Correct response not obtained since this action will result in overcurrent trip of the associated breaker. Operator may dispatch personnel to investigate - booth operator will respond accordingly.
02/25/08 Page: 7 of 10
8 Evaluate Pressurizer Heater Backup Group I status Procedure Step _ _
Alternate path step, operator is required to terminate the procedure based on the abnormal condition of the backup heaters. The successive step of the JPM evaluates the candidate's ability to apply Tech Specs for the abnormal condition.
ISAT [J UNSAT [ ]
Standards Recognizes procedure must be stopped based on response not obtained.
Dem onstration As the SRO acknowledge the operator and if the operator recommends Cues dis atchin ersonnel to investi ate, then confirm the recommendation.
Verbal-Visual Cues Notes/Comments Operator should state that he would inform the SRO of the condition and that the procedure cannot be satisfactorily completed based on the malfunction.
~ Reviews Tech Spec and determines TS 3.4.9.B applies. I Procedure Step ICritical Step ISAT [J UNSAT [ ]
Demonstration Cues Verbal-Visual Cues Standards Candidate identifies the following from TS 3.4.9.B:
Condition -7 One required group of Pressurizer heaters inoperable.
Action Required -7 Restore required group of Pressurizer heaters to OPERABLE status.
Completion Time -7 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> rOleS/COm ments 02/25/08 Page: 8 of 10
>>>>> END OF EVALUATION <<<<<
STOP TIME 02/25/08 Page: 9 of 10
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
Perform 1-PT-44A, Pressurizer Heater Output Determination. If any acceptance criteria are not met then applicable Technical Specification Action requirements are identified.
CHECKLIST
_ _ RecalilC #169 (100% power).
_ _ Do Simspray.
Ensure rod banks and TavelTref recorder are correct.
Booth Operator Actions:
If called as Unit 2 to provide HTR kW for bank 4 respond with 290 kW.
If called to investigate the breaker for Group I Backup heaters tell the operator you don't see any obvious damage but there is an overcurrent target; if asked there is nobody working in the area.
02/25/08 Page: 10 of 10
/ Dominion' PROCEDURE NO:
REVISION NO:
1-PT*44A NORTH ANNA POWER STATION 9
PROCEDURE TYPE: UNIT NO:
OPERATIONS PERIODIC TEST 1 PROCEDURE TITLE:
PRESSURIZER HEATER OUTPUT DETERMINATION TEST FREQUENCY: UNIT CONDITIONS REQUIRING TEST:
92 Days Modes 1, 2, and 3 SPECIAL CONDITIONS: None SURV REO PMT REVISION
SUMMARY
- FrameMaker Template Rev. 030.
- Administrative changes:
- Added Unit 1 water mark to cover sheet.
- Incorporated Concern OP 08-0022, Pressurizer Heater Output Determination as follows:
- Changed Steps 6.1.5 and 6.2.5 to give more detailed performance guidance in manipulating the control switches for the 1 and 4 pressurizer backup heaters.
REASON FOR TEST (CHECK APPROPRIATE BOX):
D Surveillance D Post-Maintenance Work Order Number (Post-Maintenance Only):
TEST PERFORMED BY (SIGNATURE): DATE STARTED: DATE COMPLETED:
TEST RESULT (CHECK APPROPRIATE BOX): WORK REQUEST NUMBERS AND DATE:
D Satisfactory D Unsatisfactory D Partial THE FOLLOWING PROBLEM(S) WERE ENCOUNTERED AND CORRECTIVE ACTIONS TAKEN:
(Use back for additional remarks.)
COGNIZANT SUPERVISOR or DESIGNEE: DATE:
ADDITIONAL REVIEWS: DATE:
CONTINUOUS USE
DOMINION 1-PT-44A North Anna Power Station Revision 9 Page 2 of 8 1.0 PURPOSE To provide instructions for verifying that both Emergency Bus powered Pressurizer Heater groups are operable with the capacity of each group at least 125 kw per Tech Spec SR 3.4.9.2.
2.0 REFERENCES
2.1 Source Documents 2.1.1 NUREG 0737 2.1.2 NUREG 0053, Supplement 11 2.1.3 NUREG 0452, Westinghouse Standard Technical Specifications 2.2 Technical Specifications 2.2.1 Tech Spec 3.4.9 2.3 Technical References 2.3.1 ncp 96-005, P-250 Upgrade 2.3.2 ncp 01-005, ERF Computer System Replacement 2.4 Commitment Documents 2.4.1 CTS Assignment 02-98-1809, Commitment 003, Pressurizer Heater Emergency Power Source Clarification Init Verif 3.0 INITIAL CONDITIONS 3.1 Verify the Unit is in a Mode with conditions to support Pressurizer Heater operation.
DOMINION 1-PT-44A North Anna Power Station Revision 9 Page 3 of 8 3.2 Verify the Pressurizer level is greater than 15 percent.
3.3 Verify Unit 1 and Unit 2 PCS are in service.
3.4 Notify the SRO of this test.
4.0 PRECAUTIONS AND LIMITATIONS 4.1 Comply with the following guidelines when marking steps N/A:
- IF the conditional requirements of a step do not require the action to be performed, THEN mark the step N/A and initial.
- IF this test is being performed as a Post-Maintenance Test, THEN mark inappropriate steps N/A.
- IF any other step is marked N/A, IHEN have the SRO or designee approve the N/A and submit a Procedure Action Request (PAR).
4.2 Use caution when manipulating Pressurizer Heaters with the Master Controller in MANUAL.
4.3 IF EDGs are connected to the Emergency Bus, THEN, when manipulating heaters, closely monitor EDG parameters.
4.4 IF the Pressurizer is solid AND Pressurizer Heaters must be turned on to obtain kw readings, THEN RCS pressure should be monitored and controlled to prevent undesired pressure increases.
5.0 SPECIAL TOOLS AND EQUIPMENT None
DOMINION 1-PT-44A North Anna Power Station Revision 9 Page 4 of 8 6.0 INSTRUCTIONS 6.1 IH Pressurizer Heater Bank Test CAUTION The Charging Pump control switches are next to the Pressurizer Heater control switches. Inadvertently operating the Charging Pump control switches could be defined as an ESF actuation.
6.1.1 Record the status (AUTO or ON) of the control switch for Pressurizer Heaters Group 4-Backup:
CAUTION IF the Pressurizer is solid AND Pressurizer Heater No.4 is NOT already energized upon entry into Step 6.1.2, THEN RCS pressure should be monitored and controlled to prevent undesired pressure increases.
6.1.2 Verify the PRZ HTRS ON white light for PNL NO.4 is LIT. IF NOT, THEN momentarily place control switch for Pressurizer Heaters Group 4-Backup in ON. This will lock on (red light, red flag) the heaters and illuminate the white light.
6.1.3 Display computer point QIRCOOIA or PVID330 on Unit 1 PCS. Record the IH Pressurizer Heater output below:
_ _ _ _ _ _ _ _ _ kw 6.1.4 Display computer point QIRCOOIA or PVID330 on Unit 2 PCS. Record the 1H Pressurizer Heater output below:
_ _ _ _ _ _ _ _ _ kw
DOMINION 1-PT-44A North Anna Power Station Revision 9 Page 5 of 8 6.1.5 Return control switch for Pressurizer Heaters Group 4-Backup to the status recorded in Step 6.1.1 OR as determined by the SRO as follows:
- a. IF placing the control switch in AUTO, THEN do the following:
- 1. Place the Control Switch for Przr Heaters Group 4-Backup in the STOP position, hold for approximately 5 seconds, return to AFTER-STOP.
- 2. Verify that the breaker indicating light for Przr Heaters Group 4-Backup, is red, and there is a green flag.
- 3. Verify that the white indicating light for PNL 4 is NOT illuminated.
- b. IF leaving the control switch in LOCK ON,:IllE.N do the following:
- 1. Verify that the breaker indicating light for PRZR HEATERS GROUP I-BACKUP, is red.
- 2. Verify that the white indicating light for PNL 1 is illuminated.
DOMINION 1-PT-44A North Anna Power Station Revision 9 Page 6 of 8 6.2 11 Pressurizer Heater Bank Test CAUTION The Charging Pump control switches are next to the Pressurizer Heater control switches. Inadvertently operating the Charging Pump control switches could be defined as an ESF actuation.
6.2.1 Record the status (AUTO or ON) of the control switch for Pressurizer Heaters Group I-Backup:
CAUTION IF the Pressurizer is solid AND Pressurizer Heater No.1 is NOT already energized upon entry into Step 6.2.2, THEN RCS pressure should be monitored and controlled to prevent undesired pressure increases.
6.2.2 Verify the PRZ HTRS ON white light for PNL NO.1 is LIT. IF NOT, THEN momentarily place control switch for Pressurizer Heaters Group I-Backup in ON. This will lock on (red light, red flag) the heaters and illuminate the white light.
6.2.3 Display computer point QlRC002A or PVID331 on Unit 1 PCS. Record the lJ Pressurizer Heater output below:
_ _ _ _ _ _ _ _ _ kw 6.2.4 Display computer point QlRC002A or PVID331 on Unit 2 PCS. Record the lJ Pressurizer Heater output below:
_ _ _ _ _ _ _ _ _ kw
DOMINION 1-PT-44A North Anna Power Station Revision 9 Page 7 of 8 6.2.5 Return control switch for Pressurizer Heaters Group I-Backup to the status recorded in Step 6.2.1 OR as determined by the SRO as follows:
- a. IF placing the control switch in AUTO, ~ do the following:
- 1. Place the Control Switch for Przr Heaters Group I-Backup in the STOP position, hold for approximately 5 seconds, return to AFTER-STOP.
- 2. Verify that the breaker indicating light for Przr Heaters Group I-Backup, is red, and there is a green flag.
- 3. Verify that the white indicating light for PNL 1 is NOT illuminated.
- b. IE leaving the control switch in LOCK ON, ~ do the following:
- 1. Verify that the breaker indicating light for PRZR HEATERS GROUP I-BACKUP, is red.
- 2. Verify that the white indicating light for PNL 1 is illuminated.
DOMINION 1-PT-44A North Anna Power Station Revision 9 Page 8 of 8 7.0 FOLLOW-ON 7.1 Acceptance Criteria 7.1.1 Each recorded output of Pressurizer Heater Group No. 1 is at least 125 kw.
7.1.2 Each recorded output of Pressurizer Heater Group No. 4 is at least 125 kw.
7.2 Follow-On Tasks IF Step 7.1.1 AND Step 7.1.2 CANNOT both be satisfied, THEN refer to the Action Statement of Tech Spec 3.4.9.
7.3 Completion Notification Notify the SRO that this test is complete.
Completed by: _ Date: _
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS Residual Heat Removal System is in service.
Reactor Coolant Pumps "A" and "C" are in operation.
Unit is in Mode 4 at approximately 301°F.
Charging pumps and LHSI pumps are aligned per 1-0P-3.3, Unit Shutdown from Mode 4 to Mode 5 in preparation for going below 280°F within the next hour.
RHR SYSTEM La FLOW, annunciator (E-A8) has just alarmed.
INITIATING CUE You are requested to respond in accordance with 1-AP-11, Loss of RHR.
02/21/08 Page: 1 of 11
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM R514 (Modified)
Respond to a Loss of RHR (1-AP-11).
TASK STANDARDS Attempts to restore RHR are made; Steam Dumps are controlled to control RCS temperature.
KIA
REFERENCE:
KIA 005A2.03 (2.9/3.1)
ALTERNATE PATH:
NIA TASK COMPLETION TIMES Validation Time = 20 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ ] SATISFACTORY [ ] UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature 1 Date EVALUATOR'S COMMENTS 02/21/08 Page: 2 of 11
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM R514 (Modified)
READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant eqUipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES The trainee has completed the applicable course knOWledge training at the Reactor Operator level.
INITIAL CONDITIONS Residual Heat Removal System is in service.
Reactor Coolant Pumps "A" and "C" are in operation.
Unit is in Mode 4 at approximately 301°F.
Charging pumps and LHSI pumps are aligned per 1-0P-3.3, Unit Shutdown from Mode 4 to Mode 5 in preparation for going below 280°F within the next hour.
02/21/08 Page: 3 of 11
RHR SYSTEM LO FLOW, annunciator (E-A8) has just alarmed.
INITIATING CUE You are requested to respond in accordance with 1-AP-11, Loss of RH R.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT None PERFORMANCE STEPS START TIME 0::=] Check RCS Level decreasing - NO. IProcedure Step ISAT [] UN SAT [ ]
Standards Pressurizer level, VCT level, RCS makeup rate, Containment Sump and/or PDTT level/ um in fre uenc are verified unchan ed.
Verbal-Visual Confirm no change is indicated for pressurizer level, VCT level, RCS Cues makeup rate, containment sump or PDTT level/pumping frequency when described by operator.
I Notes/Comments 02/21/08 Page: 4 of 11
Verify that the Residual Heat Removal System isolation valves are Procedure Step _ _
o en - YES.
ISAT [1 UNSAT [ 1 IL.,;;S~ta~n~d~a~rd;;,;;s~_ _1 RHR inlet valves 1-RH-MOV-1700 and 1701 are verified open.
Verbal-Visual Cues I Notes/Comments Verify that at least one Residual Heat Removal System outlet valve Procedure Step _ _
is 0 en -YES.
ISAT [1 UNSAT [ 1 IStandards IRHR outlet valve 1-RH-MOV-1720A is verified open.
Verbal-Visual 1-RH-MOV-1720A red light is lit and green light is NOT lit.
Cues I Notes/Comments 02/21/08 Page: 5 of 11
~ Check One RHR pump running - NO. I Procedure Step ISAT [] UNSAT [ ]
Standards
- Determines 1-RH-P-1A has a sheared shaft and the other RHR pump is available.
- Stops 1-RHR-P-1A.
Notes/Comments No cue is needed, the Initial Conditions given and control room indications provide adequate information to access System status.
5 If electrical power is available, then attempt to Jestore RHR flow. Procedure Step _ _
Alternate Path Step: Based on Initial Conditions provided candidate will attempt to restore flow per the RNO of the Step, and once the determination is made that NO RHR pumps are available the candidate will transition forward based on a loss of ALL RHR.
I_C_ri_ti_c_al_S_t_e-'-p 1 SAT [] UNSAT []
Standards
- Places 1-RH-FCV-1605 controller in manual and reduces demand to zero (0).
- Adjusts HIC for 1-RH-HCV-1758 demand to zero (0).
- Determines venting not required since there are no indications of air entrainment.
- Places Control switch for 1-RH-P-1 B to start.
- Determines RHR pump breaker failed to close.
Verbal-Visual When pump indications are identified by the operator, then inform him Cues that 1-RH-P-1 B GREEN light ON, AMBER light ON, RED light OFF, and amps are as they see them (zero).
Notes/Comments The only critical actions are:
- 1) Attempt to start 1-RHR-P-1B and
- 2) Determine RHR pump 1B breaker failed to close.
02/21/08 Page: 6 of 11
m::=:=:J Initiate personnel protective actions. IProcedure Step
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ISAT[] UNSAT[]
IStandards I Records time to boiling form 1-GOP-13.0 in procdure.
Notes/Comments Cue is not needed since this information is provided in the Initial Conditions.
Evaluate the need to implement EPIP-1.01, Emergency Manager Procedure Step _ _
Controllin Procedure.
ISAT [] UNSAT [ 1 Iw=S=ta;;;,;n,,;,;;d=a=rd=s~_ _ 1 Notifies Annex (WCCS) or other operator to evaluate EPIP-1.01.
Verbal-Visual Operations shift manager or unit supervisor will perform this step.
Cues INotes/Comments 02/21/08 Page: 7 of 11
~ Monitors Containment Radiation. I Procedure Step
/SAT[] UNSAT[]
IL..:S;;,;;ta;;;;,n,;,;;d=a;o;;rd;;,;;s~_ _1 Checks recorder and/or monitors 1-RM-RMS-159 and 160.
Verbal-Visual Cues INotes/Comments
~ Initiate Attachment 11 for Containment closure. I Procedure Step I SAT [] UNSAT [ ]
IStandards IIdentifies need to perform Attachment 11.
Verbal-Visual Inform the operator that an additional operator has been assigned to Cues perform Attachment 11.
INotes/Comments 02/21/08 Page: 8 of 11
[}[=] Verifies 1-RC-L1-105 energized. I Procedure Step ISAT [1 UNSAT [ 1 I-=S=ta;;;.n;,;;;d;,;;;a;,,;rd;,;;;s~ _ _ 1 Directs operator to energize 1-RC-L1-1 05.
Verbal-Visual Inform the operator that an additional operator has been assigned to Cues ener ize 1-RC-L1-1 05.
I Notes/Comments rrr=J Start Available Containment Air Recirc Fans. I Procedure Step ISAT [1 UNSAT [ 1 Iw;;S;;,;,ta;;;,;n,;,;d;;;a;;,,;rd;;;;s=-----_ _ 1 Checks all 3 CARFs are operating.
Verbal-Visual Once CARF are identified by candidate then confirm indications that all Cues three are operating (breaker RED light ON, GREEN light OFF and backdraft damper open).
INotes/Comm ents 02/21/08 Page: 9 of 11
~ Maintain core-cooling using forced circulation. I Procedure Step I-Critical Step ISAT[l UNSAT[l Standards
- Verifies RCP 1A and 1C running.
- Raises demand on 1-MS-PC-1464 to control RCS temperature.
- Increases feed flow to SGs if necessary to control level.
Verbal-Visual Once CETs are verified stable or decreasing assume that another Cues a erator will com lete the rocedure.
roteslCom ments
>>>>> END OF EVALUATION <<<<<
STOP TIME 02/21/08 Page: 10 of 11
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
SIMULATOR SETUP JOB PERFORMANCE MEASURE R514 (Modified)
Respond to a Loss of RHR (1-AP-11).
CHECKLIST
_ _ Recall the IC 180 (Mode 4 with temperature -301 F). 0
_ _ Do Simspray (remember startup 2, since this is a shutdown scenario).
Booth operator actions Respond as operator to check breaker for 1B RHR pump if requested; call back in 2 minutes and report that there is nothing obviously wrong with the breaker, but there is an acrid odor coming from it. If asked to locally close the breaker tell him you will get the necessary brief, procedure, and equipment and call him once you are ready. If requested you will get support from electrical maintenance also.
IF asked as ANNEX/SM respond that you will evaluate EPIP-1.01 (and VPAP-2802, and make any required notifications, etc.).
If asked as Annex/HP, respond that you will coordinate evacuation of CNTMT.
Respond as extra operator to energize 1-RC-L1-1 05.
02/21/08 Page: 11 of 11
~
?
Domlnlon-NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR (WITH ELEVEN ATTACHMENTS) PAGE 1 of 23 PURPOSE To provide instructions for maintaining Core Cooling and protecting the Reactor Core in the event that RHR Cooling is lost.
ENTRY CONDITIONS This procedure is entered when RHR is required for Core Cooling and any of the following conditions exist:
- Air-binding of operating RHR pumps as indicated by:
- Flow oscillations, or
- Motor amps fluctuating, or
- Excessive pump noise.
- Annunciator "E" Panel A-6, RHR PP 1A AUTO TRIP, is LIT, or
- Annunciator "E" Panel A-7, RHR PP 18 AUTO TRIP, is LIT, or
- Annunciator "E" Panel A-B, RHR SYSTEM La FLOW, is LIT, or
- Loss of RHR pumps due to loss of power, or Failure of RHR system to control HCStemperature due to loss of valve failures, or Loss of Service Water System withBHR System in service, or
- Loss of Component Cooling System with RHR System in service.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 2 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION:
- RCS make-up concentration MUST be greater than or equal to current Shutdown Margin and Boron concentration requirements of the COLR.
- Changes in RCS pressure due to boiling in the core can result in Reactor Vessel water level changes that may not show on RCS standpipe level indicator 1-RC-L1-1 03.
- 1. CHECK RCS LEVEL - DECREASING 0 GO TO Step 5.
0
- RCS standpipe level -
DECREASING OR 0
- RCS ultrasonic level indicator -
DECREASING OR 0
- PRZR level - DECREASING OR 0
- RCS makeup rate -
INCREASING OR 0
- PDn pumping frequency -
UNEXPLAINED INCREASE
- 2. INCREASE RCS MAKEUP FLOW
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 3 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 3. ISOLATE RCS DRAIN PATHS:
a) Check the following Letdown o a) Manually close valves.
Isolation Valves - CLOSED:
o
- 1-CH-HCV-1200A o
- 1-CH-HCV-1200B o
- 1-CH-HCV-1200C o
- 1-CH-LCV-1460A o
- 1-CH-LCV-1460B o b) Check 1-CH-HCV-1142, RHR o b) Manually close valve.
System to Letdown Isolation Valve
- CLOSED c) Check loop drains - CLOSED: o c) Manually close valves.
o
- 1-RC-HCV-1557A o
- 1-RC-HCV-1557B o
- 1-RC-HCV-1557C d) While continuing with procedure, o d) Ensure valves are closed.
verify the following valves -
LOCKED CLOSED:
o
- 1-RH-36, Residual Heat Removal to RWST Isolation Valve (Containment) o
- 1-RH-34, Residual Heat Removal Supply to RP (Containment)
(STEP 3 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 4 of 23 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 3. ISOLATE RCS DRAIN PATHS:
(Continued) o e) Close any known RCS drain paths o f) Initiate actions to stop level decreases due to maintenance covered by 0-GOP-13.3, ASSESSMENT OF MAINTENANCE ACTIVITIES FOR POTENTIAL LOSS OF REACTOR COOLANT INVENTORY CAUTION:
- Changes in RCS pressure can result in Reactor Vessel water level changes that may not show on RCS level indicator 1-RC-L1-1 03.
- 4. VERIFY ADEQUATE RCS MAKEUP FLOW:
o a) Check RCS level - STABLE OR a) Ensure the keylock switch for 1-RC-L1-1 OS, INCREASING Independent RCS Level Indicator, is in ENABLE. GO TO appropriate procedure:
o
- 1-AP-52, LOSS OF REFUELING CAVITY LEVEL DURING REFUELING (STEP 4 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 5 of 23 ACTIONI EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 4. VERIFY ADEQUATE RCS MAKEUP FLOW: (Continued) b) Check RHR flow - LESS THAN OR o b) Reduce RHR flow to design flow rate of EQUAL TO DESIGN FLOW OF ATTACHMENT 3.
ATTACHMENT 3:
o
Page 1 of 2 o
Page 2 of 2 o c) Check RCS level - GREATER c) Do the following:
THAN MINIMUM FOR INDICATED FLOW OF ATTACHMENT 2 0 1) Continue RCS makeup.
0 2) Stop RHR Pumps.
0 3) GO TO Step 11.
o d) Check RCS level - AT LEAST 0 d) Increase RCS level to greater than +10 inches
+10 INCHES ABOVE above centerline.
CENTERLINE o !.E level cannot be increased to greater than
+10 inches above centerline, THEN GO TO 1-AP-17, SHUTDOWN LOCA.
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 6 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 5. VERIFY RHR ISOLATION VALVES -
OPEN:
a) RHR Inlet Isolation Valves - OPEN a) Do the following:
o
- 1-RH-MOV-1700 o 1) Stop RHR Pump(s).
o
- 1-RH-MOV-1701 o 2) Reduce RCS pressure as necessary.
o 3) WHEN RCS pressure is less than 418 psig, THEN open valves.
b) AT least one RHR Outlet Isolation o b) Open at least one RHR Outlet Isolation Valve.
Valve - OPEN
( o
- 1-RH-MOV-1720A o
- 1-RH-MOV-1720B CAUTION: RHR flow less than minimum requirements may cause RCS temperature to increase.
NOTE:
- Operating at low RHR system flow rates during reduced inventory operations greatly reduces the risk of air entrainment (vortexing).
- Indications of a pump sheared shaft are low flow and low motor amps. A degraded pump or a pump with a sheared shaft is to be considered as NOT running.
- 6. CHECK ONE RHR PUMP - RUNNING Do the following:
o a) !.E the other RHR pump is available, THEN stop any degraded RHR pump.
(STEP 6 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 7 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 6. CHECK ONE RHR PUMP -
RUNNING (Continued) o b)!.E a degraded RHR pump is running AND the other RHR pump is NOT available, THEN GOTO Step 7.
c) !.E electrical power is available, THEN do the following:
- 1) Manually close the following RHR Control Valves:
o
- 1-RH-FCV-1605 o
- 1-RH-HCV-1758 o 2) !.E an RHR Pump was previously stopped due to air entrainment, THEN locally vent both RHR Pumps.
o 3) IF both RHR pumps are stopped, THEN start one RHR pump.
o
- 1-RH-HCV-1758 o
- 1-RH-FCV-1605 o 5) !.E an RHR Pump has been started, THEN GO TO Step 7.
o !.E no RHR Pump can be started, THEN GO TO Step 11 .
(STEP 6 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 8 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 6. CHECK ONE RHR PUMP -
RUNNING (Continued) d) IF electrical power is NOT available, THEN do the following:
o 1) Initiate 0-AP-1 0, LOSS OF ELECTRICAL POWER.
o 2) GO TO Step 11.
- 7. VERIFY RHR SYSTEM - NORMAL: Do the following:
o
o
- RHR Motor amps - STABLE +10 inches above centerline by increasing charging flow.
o
- RCS temperature - STABLE
- 2) Check RHR flow - less than or equal to design flow of ATTACHMENT 3:
o
- 1 RHR HX in use - Page 2 of 2 IE RHR flow is greater than the design flow rate of ATTACHMENT 3, THEN reduce flow to the design flow rate using:
o
- 1-RH-HCV-1758 o
- 1-RH-FCV-1605 (STEP 7 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 9 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7. VERIFY RHR SYSTEM - NORMAL:
(Continued)
D 3) Check RCS level - Greater than minimum for indicated flow of ATIACHMENT 2.
D !E RCS level is not greater than minimum for indicated flow of ATIACHMENT 2, THEN STOP the RHR Pumps and GO TO Step 11.
- 4) Send an Operator to locally check pump operation:
D
- RHR pump noise D
- RHR pump seals D
- RHR pump vibration D b) IF the running RHR pump is degraded AND the other RHR pump is available, THEN RETURN TO Step 6.
D c) IF RHR System cannot be stabilized, THEN stop running RHR Pump AND GO TO Step 11.
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 10 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 8. CHECK SERVICE WATER TO CC HEAT EXCHANGER - AVAILABLE:
0 a) Verify Service Water System - IN a) !E Service Water flow is NOT available, THEN SERVICE initiate the following while continuing with this procedure:
0
- 0-AP-12, LOSS OF SERVICE WATER 0
- 1-AP-15, LOSS OF COMPONENT COOLING 0 GO TO Step 11 .
b) Verify Service Water Supply Valves b) Open Service Water Supply Valves to CC to CC System - OPEN: System:
0
- 1-SW-MOV-108A 0
- 1-SW-MOV-108A 0
- 1-SW-MOV-108B 0
- 1-SW-MOV-108B 0 c) Locally check Service Water to CC 0 c) GO TO Step 11.
Heat Exchanger ~P - NORMAL
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 11 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
EXCHANGERS - NORMAL:
o
o
- 1-CC-FI-132B o
- 1-CC-TV-103A, A RHR Heat Exchanger Return Isolation o
- 1-CC-TV-103B, B RHR Heat Exchanger Return Isolation o
- 1-CC-MOV-1 OOA, A CC Heat Exchanger Outlet Isolation o
- 1-CC-MOV-100B, B CC Heat Exchanger Outlet Isolation b) i.E either 1-CC-TV-1 03A or 1-CC-TV-1 03B cannot be opened, THEN close the associated RHR CC MOV:
o
- 1-CC-MOV-1 OOA for 1-CC-TV-1 03A o
- 1-CC-MOV-100B for 1-CC-TV-103B o c) i.E CC flow is restored, THEN GO TO Step 10.
o d) i.E CC is NOT restored, THEN initiate 1-AP-15, LOSS OF COMPONENT COOLING, while continuing with this procedure.
o e) GO TO Step 11.
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 12 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 10. RETURN TO PROCEDURE AND STEP IN EFFECT CAUTION: If RCS boiling is determined to exist, then non-essential personnel should be evacuated from the Containment.
- 11. INITIATE PERSONNEL PROTECTIVE ACTIONS:
o a) Record most recent time to boiling estimate from 1-GOP-13.0, ALTERNATE CORE COOLING METHOD ASSESSMENT:
- Time (minutes): _
o b) Evaluate need to implement EPIP-1.01, EMERGENCY MANAGER CONTROLLING PROCEDURE c) Monitor Containment Radiation:
o
- 1-RM-RMS-159 o
- 1-RM-RMS-160 12._ INITIATE ATTACHMENT 11, CONTAINMENT CLOSURE, WHILE CONTINUING WITH THIS PROCEDURE 13._ VERIFY 1-RC-L1-105, INDEPENDENT o Place the keylock switch for 1-RC-L1-1 05 in RCS LEVEL INDICATOR - ENABLE.
ENERGIZED
( NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 13 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 14. START AVAILABLE CONTAINMENT AIR RECIRC FANS USING 1-0P-21.1, CONTAINMENT VENTILATION NOTE: If RCPs are stopped, then Attachment 10, NATURAL CIRCULATION should be used to establish and maintain natural circulation.
- 15. MAINTAIN CORE COOLING USING FORCED CIRCULATION:
o a) Verify at least one RCP - o a) GO TO Step 16.
RUNNING b) Stabilize RCS temperature by dumping steam using either of the following:
o
- Condenser Steam Dumps OR o
o
- Condensate o d) GO TO Step 18
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 14 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 16. CHECK IF THE RCS SHOULD BE COOLED WITH SPENT FUEL POOL COOLING:
D a) Verify Reactor Cavity - FLOODED D a) GO TO Step 17.
D b) Verify Spent Fuel Pit level - D b) Initiate O-AP-27, MALFUNCTION OF SPENT NORMAL FUEL PIT SYSTEM, AND GO TO Step 17.
D c) Initiate ATTACHMENT 9, COOLING THE RCS WITH SFP COOLERS D d) GO TO Step 18
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 15 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION:
- Personnel working in Containment should be warned before the RCS is refilled to avoid contamination of personnel near any RCS openings.
- Only borated water should be added to the RCS to maintain adequate shutdown margin.
- Differences exist in RCS levels between active and inactive cold and hot legs during reduced inventory operations. At saturated conditions, the hot and cold leg levels can differ by several feet.
NOTE: The alternate cooling method priority is obtained from 1-GOP-13.0, ALTERNATE CORE COOLING METHOD ASSESSMENT.
- 17. DETERMINE APPROPRIATE ALTERNATE CORE COOLING METHOD:
o
- Natural Circulation - Initiate ATTACHMENT10,NATURAL CIRCULATION, while continuing with this procedure OR o
- Reflux Boiling - Initiate ATTACHMENT 8, REFLUX BOILING, while continuing with this procedure OR (STEP 17 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 16 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 17. DETERMINE APPROPRIATE ALTERNATE CORE COOLING METHOD: (Continued) o
- Hot Leg Injection Forced Feed and Spill - Initiate ATTACHMENT5,HOTLEG INJECTION FORCED FEED AND SPILL, while continuing with this procedure OR o
- Cold Leg Injection Forced Feed and Spill - Initiate ATTACHMENT 6, COLD LEG INJECTION FORCED FEED AND SPILL, while continuing with this procedure OR o
- Gravity Feed and Spill - Initiate ATTACHMENT 4, GRAVITY FEED AND SPILL, while continuing with this procedure
( NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 17 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION: If the Containment has been previously closed out using ATTACHMENT 11, CONTAINMENT CLOSURE, then personnel should not re-enter without first contacting Health Physics.
- 18. CONTINUE ATTEMPTS TO RESTORE RHR SYSTEM:
a) Vent RHR System as necessary:
0 1) Maintain RCS level while venting RHR by increasing makeup flow to RCS 0 2) Locally vent RHR System b) Establish conditions to start RHR Pumps:
o 1) Verify RHR Pumps - o 1) GO TO Step 19.
SECURED o 2) Check RCS Level - AT o 2) Increase RCS level to greater than LEAST +10 INCHES +10 inches above centerline.
ABOVE CENTERLINE o !E level cannot be increased to greater than
+10 inches above centerline, THEN GO TO 1-AP-17, SHUTDOWN LOCA.
o 3) Check RHR Pump - o 3) Try to get an RHR Pump available.
AVAILABLE (STEP 18 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 18 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 18. CONTINUE ATTEMPTS TO RESTORE RHR SYSTEM:
(Continued)
- 4) Check RHR Inlet Isolation 0 4) Manually open valves.
Valves - OPEN:
0
- 1-RH-MOV-1700 0
- 1-RH-MOV-1701
- 5) Check RHR Outlet Isolation 0 5) Manually open desired valve.
Valves, Disch to Cold Legs -
OPEN:
0
- 1-RH-MOV-1720A
("B" Cold Leg)
OR 0
- 1-RH-MOV-1720B
("C" Cold Leg) 0 6) Check 1-RH-HCV-1758 - 0 6) Manually close valve.
CLOSED 0 7) Check 1-RH-FCV-1605 - 0 7) Manually close valve.
CLOSED
0 a) Restore Service Water using O-AP-12, LOSS OF SERVICE WATER 0 b) Restore CC System using 1-AP-15, LOSS OF COMPONENT COOLING
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 19 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 20. CHECK SERVICE WATER TO CC HEAT EXCHANGER - AVAILABLE:
0 a) Verify Service Water System - IN a) IF Service Water flow is NOT available, THEN SERVICE continue attempts to restore Service Water using the following while continuing with this procedure:
0
- 0-AP-12, LOSS OF SERVICE WATER 0
- 1-AP-15, LOSS OF COMPONENT COOLING 0 RETURN TO Step 18.
b) Verify Service Water Supply Valves b) Open Service Water Supply Valves to CC to CC System - OPEN System:
0
- 1-SW-MOV-108A 0
- 1-SW-MOV-1 08A 0
- 1-SW-MOV-108B 0
- 1-SW-MOV-108B 0 c) Locally check Service Water to CC 0 c) RETURN TO Step 18.
Heat Exchanger ~P - NORMAL
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 20 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
o a) Verify CC System - IN SERVICE o a) IF CC flow is NOT available, THEN continue attempts to restore CC using 1-AP-15, LOSS OF COMPONENT COOLING, while continuing with this procedure.
o RETURN TO Step 18.
b) Check CC flow to RHR Heat b) Do the following:
Exchangers - NORMAL:
- 1-CC-FI-132A Exchanger:
o
- 1-CC-FI-132B o
- 1-CC-TV-103A, A RHR Heat Exchanger Return Isolation o
- 1-CC-TV-103B, B RHR Heat Exchanger Return Isolation o
- 1-CC-MOV-100A, A CC Heat Exchanger Outlet Isolation o
- 1-CC-MOV-100B, B CC Heat Exchanger Outlet Isolation
o
- 1-CC-MOV-1 OOA for 1-CC-TV-1 03A o
- 1-CC-MOV-1 OOB for 1-CC-TV-1 03B o 3) IF CC flow is restored, THEN GO TO Step 22.
(STEP 21 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 21 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
(Continued) o 4) IF CC is NOT restored, THEN continue attempts to restore CC using 1-AP-15, LOSS OF COMPONENT COOLING, while continuing with this procedure.
o 5) RETURN TO Step 18.
CAUTION:
- During RHR flow restoration, flow must start at a lower rate to limit the initial sudden cooldown and to minimize level loss caused by collapsing voids.
- If the RHR System was not satisfactorily vented, then entrained air can be swept from the system by raising flow to 3300 gpm. This method could cause water hammer or pump damage.
- 22. RESTORE RHR FLOW:
a) Close the following valves:
0
- 1-RH-HCV-1758 0
- 1-RH-FCV-1605 0 b) Start one RHR Pump 0 b) RETURN TO Step 18.
0 c) Maintain RCS level within acceptable region of Attachment 2 d) Restore RHR flow by repositioning the following RHR Control Valves:
0
- 1-RH-HCV-1758 0
- 1-RH-FCV-1605
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 22 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 23. VERIFY RHR SYSTEM - NORMAL: Do the following:
D
D
- RHR Motor amps - STABLE
+10 inches above centerline by increasing D
- RCS temperature - STABLE charging flow.
- 2) Check RHR flow - less than or equal to design flow of ATIACHMENT 3:
D
- 1 RHR HX in use - Page 2 of 2 IF RHR flow is greater than the design flow rate of ATIACHMENT 3, THEN reduce flow to the design flow rate using:
D
- 1-RH-HCV-1758 D
- 1-RH-FCV-1605 D 3) Check RCS level - Greater than minimum for indicated flow of ATIACHMENT 2.
D !E RCS level is not greater than minimum for indicated flow of ATIACHMENT 2, THEN STOP the RHR Pumps and RETURN TO Step 18.
- 4) Send an Operator to locally check pump operation:
D
- RHR pump noise D
- RHR pump seals D
- RHR pump vibration D b) !E RHR System cannot be stabilized, THEN stop running RHR Pump AND RETURN TO Step 18.
NUMBER PROCEDURE TITLE REVISION 25 1-AP-11 LOSS OF RHR PAGE 23 of 23 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 25. CHECK IF RCS MAKEUP SHOULD BE REDUCED:
0 a) RCS Temperature - LESS THAN 0 a) Continue cooldown with RHR.
200°F 0 b) RCS Level - STABLE OR 0 b) GO TO Step 26.
INCREASING c) Check Low Head SI Pump 0 c) GO TO Step 25e.
Suctions From Containment Sump
- CLOSED:
0
- 1-SI-MOV-1860A 0
- 1-SI-MOV-1860B 0 d) Stop any running Low Head SI Pump.
0 e) Control RCS level using makeup and letdown as required
THAN 140 of 0 RETURN TO Step 24.
- 27. RETURN TO PROCEDURE AND STEP IN EFFECT
- END-
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS The Unit is operating at 100% power, steady state.
st Turbine 1 stage pressure channeI1-MS-PT-1447 has failed low.
Immediate actions of 1-AP-3, Loss of Vital Instrumentation have been performed.
INITIATING CUE You are requested to continue performance of 1-AP-3 up to and including completion of Step 10.
03/04/08 Page: 1 of 10
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM R664 (modified) st Respond to failure of non-controlling 1 Stage Pressure Channel.
TASK STANDARDS Steam dumps were transferred to steam-pressure mode lAW 1-AP-3; upon failure, steam dumps were taken to OFFIRESET as required by 1-AP-38.
KIA
REFERENCE:
041-A4.04 (2.7/2.7)
ALTERNATE PATH:
YES TASK COMPLETION TIMES Validation Time = 12 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ ] SATISFACTORY [ ] UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature 1 Date EVALUATOR'S COMMENTS 03/04/08 Page: 2 of 10
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM R664 (modified)
READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES The trainee has completed the applicable course knowledge training at the Reactor Operator level.
INITIAL CONDITIONS The Unit is operating at 100% power, steady state.
Turbine 1st stage pressure channeI1-MS-PT-1447 has failed low.
Immediate actions of 1-AP-3, Loss of Vital Instrumentation have been performed.
03/04/08 Page: 3 of 10
INITIATING CUE You are requested to continue performance of 1-AP-3 up to and including completion of Step 10.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT None PERFORMANCE STEPS START TIME 03/04/08 Page: 4 of 10
NOTE: Although not required based on the initial conditions provided, operator may review Steps 1-3 to verify all step performance requirements are met and ascertain pertinent information such as the positions of SGWLC transfer switches, etc.
0:==1 Verify systems affected by pressurizer level channels - NORMAL. IProcedure Step ISAT [1 UNSAT [ 1 Standards
- Verifies operable pressurizer level channels are selected (all channels 1459, 1460, 1461 indicate normal; switch may be checked in 461/460 position but doesn't matter since all channels are normal).
- Verifies letdown in service (flow indicated, 1200B I/S orifice).
- Verifies pressurizer level control in auto (1-RC-LC-1459G and 1-CH-FCV-1122 in auto).
- Verifies control group heaters not tripped (annunciator NOT lit and white liQht ON for control group heaters).
Verbal-visual Provide feedback to operator for expected conditions:
Cues
- All PRZR level channels indicated - %,
PRZR level control transfer switch is as they see it.
- 1-CH-HCV-1200B red light is ON, green Light OFF Letdown flow indicates - gpm.
- 1-RC-LC-1459G auto light lit demand is - %.
1-CH-FCV-1122 auto light lit demand is - %.
- White light for control group heaters is ON.
I Notes/Comments
( 03/04/08 Page: 5 of 10
Procedure Step _ _
ICritical Step ISAT [] UNSAT [ ]
Standards
- Place both STEAM DUMP INTLK switches to OFF/RESET.
- Place STEAM DUMP CONTROLLER to MANUAL.
- Place MODE SELECTOR switch to Steam PRESS.
- Ensure Steam Dump demand is ZERO (observes controller output and may also check meter on Vertical Panel).
- Return STEAM DUMP CONTROLLER to AUTO.
- Verify Steam Dump demand is ZERO (observes controller output and may also check meter on Vertical Panel).
- Place both STEAM DUMP INTLK switches to ON.
Verbal-visual Provide feedback to operator for expected conditions:
Cues
- Status light M-D4 for steam dumps extinguishes when placed in OFF/RESET.
- STEAM DUMP CONTROLLER MANUAL light is ON, AUTO light is OFF when manual PB depressed.
- STEAM DUMP CONTROLLER demand indicates 0%, demand meter on vertical panel indicates 0% (point ot zero on resoective indicators using a pen).
- STEAM DUMP CONTROLLER AUTO light is ON, MANUAL light is OFF when auto PB depressed.
- STEAM DUMP CONTROLLER demand indicates 0%, demand meter on vertical panel indicates 0% (point ot zero on resoective indicators using a pen).
- Status light for steam dumps ILLUMINATES when placed in ON.
Note: IF operator checks 1-MS-PT-1464, point to psig mark on indicator.
IN otes/Comments 03/04/08 Page: 6 of 10
~ Verify operable channels selected for SGWLC instruments. I Procedure Step ISAT [1 UNSAT [ 1 Standards Checks all of the following indicating normal (with the exception of 1-MS-PT-1447):
st
- Turbine 1 Stage Pressure (Blue Channel, 1-MS-PT-1446).
- All selected Steam Flow Channels (Blue channel).
- All selected Feed Flow can nels (Blue Channel).
Note: Since all channels were checked previously operator may just observe the vertical panel and note that there are no other instrument failures.
Verbal-visual Provide feedback to operator for expected conditions:
Cues
- 1-MS-PT-1446 indicates - 100%.
- All transfer switches for SGWLC are as they see them.
Notes/Comments The malfunction that causes 1-MS-TCV-1408A to open is on a set time delay; Operator will most likely not complete the above step since they may notice the failed steam dump and take actions of 1-AP-38.
03/04/08 Page: 7 of 10
4 Respond Excessive Load Increase (1-AP-38 immediate actions) Procedure Step _ _
Alternate path step, the malfunction requires the candidate to stop performance of the current procedure, implement IOAs of 1-AP-38 from memory, and to terminate the transient, take manual actions per the RNO of 1-AP-38.
I_C_ri_ti_c_al_S_t_e-'----p 1 SAT [] UNSAT []
Standards
- Verify ALL steam dumps closed - NO 1-MS-PCV-1408A is open.
- Place both STEAM DUMP INTLK switches to OFF/RESET.
- Verifies steam dumps (1-MS-TCV-1408A) go closed.
- Verify turbine load normal- YES/NO (see note).
- Verify Reactor Power.:': 100% and stable - YES/NO (see note).
Verbal-visual Provide the following information:
Cues Tavg is trending down and reactor power is trending up.
- When identified by the operator the red light for 1-MS-PCV-1408A is ON the Green light is off (all other valves have green light on and red light off).
- After operator places steam dumps in off confirm 1-MS-PCV-1408A red light is OFF and Green light is ON (all other valves have green light on and red light off).
- When operator checks reactor power it is the same as when they took the shift and Stable.
Verbal-visual JPM is complete once operator states the immediate operator actions Cues of 1-AP-38 are com lete.
Notes/Comments Depending on when the operator identifies the malfunction and completes the required actions power may be high enough to warrant ram ping. If Reactor Power is slightly greater than 100% because it has not returned to the pre-event value when checked the operator may elect to ramp slightly.
03/04/08 Page: 8 of 10
>>>>> END OF EVALUATION <<<<<
STOP TIME 03/04/08 Page: 9 of 10
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
SIMULATOR SETUP JOB PERFORMANCE MEASURE R664 (modified) st Respond to failure of non-controlling 1 Stage Pressure Channel CHECKLIST RecalilC 167.
_ _ Do Sims pray.
Ensure rod banks and Taveffref recorder are correct.
_ _ Provide copy of 1-AP-3 with immediate action steps signed off.
Note: The malfunction (1-MS-TCV-1408 goes open), is set to trigger on time delay automatically after the operator completes transferring steam dumps to steam pressure mode and will close once the switches are taken to OFF/RESET.
03/04/08 Page: 100f10
~~
~Dominion-NORTH ANNA POWER STATION
(
ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 22 1-AP-3 LOSS OF VITAL INSTRUMENTATION (WITH TWO ATTACHMENTS) PAGE 1 of 19 PURPOSE To provide instructions to follow in the event of a loss of vital instrumentation.
ENTRY CONDITIONS This procedure is entered when a faulty indication occurs on any of the following vital instrumentation channels:
- Reactor Coolant Flow, or
- Pressurizer Level, or
- Pressurizer Pressure Protection, or
- DELTA TITAVE Protection, or
- Containment Pressure Protection, or
- RWST Level, or
- Steam Generator Level, or
- Turbine Stop Valves Indication, or
- Turbine First Stage Impulse Pressure, or
- Turbine Auto Stop Oil Low Pressure Trip Signal, or
- Steam Flow, or
- Feed Flow,or
- Steam Pressure, or
- Station Service Bus Undervoltage,or
- Station Service Bus Underfrequency.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 22 1-AP-3 LOSS OF VITAL INSTRUMENTATION PAGE 2 of 19 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED 1 1_ VERIFY REDUNDANT o IF unable to determine Reactor is in a safe INSTRUMENT CHANNEL operating condition, THEN GO TO 1-E-O, INDICATION - NORMAL REACTOR TRIP OR SAFETY INJECTION.
21_ VERIFY STEAM GENERATOR Do the following:
LEVEL CONTROLLING CHANNELS - NORMAL: a) Place the associated valves in MANUAL:
o
- Main Feed Reg Valves o
- Main Feed Reg Bypass Valves o b) Control Steam Generator level.
o
- Steam Flow o
- Feed Flow o
- Steam Generator Level Ch III o
- Steam Pressure 3 1_ VERIFY TURBINE FIRST STAGE o IF the controlling channel failed, THEN place PRESSURE INDICATIONS - Control Rod Mode Selector switch in MANUAL.
NORMAL
NUMBER PROCEDURE TITLE REVISION 22 1-AP-3 LOSS OF VITAL INSTRUMENTATION PAGE 3 of 19 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 4. VERIFY SYSTEMS AFFECTED BY PRESSURIZER LEVEL CHANNELS -
NORMAL o a) Verify operable Pressurizer level a) Do the following:
channels - SELECTED o 1) Place 1-CH-FCV-1122, Charging Flow Control Valve in MANUAL and control level at program.
o 2) Select operable Pressurizer level channels for control.
- 3) Verify the following Annunciators are NOT LIT:
o
- Panel B-F8, PRZ LO LEVEL o
- Panel B-G6, PRZ HI LEVEL -
BU HTRS ON o
- Panel B-G7, PRZ LO LEV HTRS OFF -
LETDWN ISOL o
- Panel B-G8, PRZ HI LEVEL o b) Verify Letdown - IN SERVICE o b) Restore letdown using Attachment 2, LETDOWN RESTORATION.
(STEP 4 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 22 1-AP-3 LOSS OF VITAL INSTRUMENTATION PAGE 4 of 19 ACTIONI EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 4. VERIFY SYSTEMS AFFECTED BY PRESSURIZER LEVEL CHANNELS
- NORMAL (Continued) o c) Verify Pressurizer Level Control - c) Do the following:
IN AUTO o 1) Verify level restored to program.
o 2) Verify expected output of 1-RC-LCV-1459G, Pressurizer Level Control.
o 3) Place 1-CH-FCV-1122, Charging Flow Control Valve in AUTO.
o d) Verify Pressurizer Control Group o d) Reset Pressurizer Control Group Heaters by Heaters - NOT TRIPPED placing control switch to START position.
NUMBER PROCEDURE TITLE REVISION 22 1-AP-3 LOSS OF VITAL INSTRUMENTATION PAGE 5 of 19 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 5. VERIFY BOTH TURBINE FIRST IF Condenser Steam Dumps are available, THEN STAGE PRESSURE CHANNELS - transfer to Steam Pressure Mode by doing the NORMAL following:
o a) Place both STEAM DUMP INTLK switches to OFF/RESET.
o b) Place STEAM DUMP CONTROLLER to MANUAL.
o c) Place MODE SELECTOR switch to STEAM PRESS.
o d) Ensure Steam Dump demand is ZERO.
o e) Return STEAM DUMP CONTROLLER to AUTO.
o f) Verify Steam Dump demand is ZERO.
o g) Place both STEAM DUMP INTLK switches to ON.
NUMBER PROCEDURE TITLE REVISION 22 1-AP-3 LOSS OF VITAL INSTRUMENTATION PAGE 6 of 19 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 6. VERIFY OPERABLE CHANNELS Do one of the following as directed by the Unit 1 SELECTED FOR ALL OF THE SRO:
FOLLOWING SGWLC INSTRUMENTS:
0
- Turbine First Stage Pressure 0 * !.E desired to swap ONLY the failed channel, THEN GO TO Step 8.
0 * "A" SG Steam Flow OR 0 * "B" SG Steam Flow 0 * !.E desired to swap ALL SGWLC channels to the 0 * "C" SG Steam Flow same channel, THEN GO TO Step 9.
0 * "A" SG Feed Flow 0 * "B" SG Feed Flow 0 * "C" SG Feed Flow
- 7. GO TO STEP 10
NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 14 1-AP-38 EXCESSIVE LOAD INCREASE (WITH TWO ATTACHMENTS) PAGE 1 of 10 PURPOSE To provide instructions to follow in the event of an excessive load increase.
ENTRY CONDITIONS This procedure is entered when any of the following conditions exist:
- Rapid increase in Steam Flow, or
- Rapid increase in Main Generator MW Output due to increased Steam addition to the Main Turbine, or
- Rapid decrease in Main Generator MW Output due to Steam diversion from the Main Turbine, or
- Rapid increase in Reactor Power Level, or
- Control Rods stepping OUT in fast speed.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 14 1-AP-38 EXCESSIVE LOAD INCREASE PAGE 2 of 10 ACTIONI EXPECTED RESPONSE RESPONSE NOT OBTAINED 1 1_ VERIFY ALL STEAM DUMP o Place Both Steam Dump Interlock Switches to VALVES - CLOSED OFF/RESET.
o . 1-MS-TCV-1408A o !.E any valve does NOT close, THEN locally close using ATTACHMENT 2, STEAM DUMP o . 1-MS-TCV-1408B INSTRUMENT AIR ISOLATION.
o . 1-MS-TCV-1408C o . 1-MS-TCV-1408D o . 1-MS-TCV-1408E o . 1-MS-TCV-1408F o . 1-MS-TCV-1408G o . 1-MS-TCV-1408H 21_ VERIFY ALL SG PORVs- o Place any OPEN SG PORV controller in MAN and CLOSED shut affected SG PORV.
o . 1-MS-PCV-101A (A SG) o . 1-MS-PCV-101B (B SG) o . 1-MS-PCV-101C (C SG)
NUMBER PROCEDURE TITLE REVISION 14 1-AP-38 EXCESSNE LOAD INCREASE PAGE 3 of 10 ACTION! EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTiON: Ramp rates shall be limited to 5% per minute or less.
[ 3] _ VERIFY THE FOLLOWING: Use one of the following methods to reduce Reactor Power to the power level before the event started:
o . MAIN TURBINE LOAD - o . OPERATOR AUTO NORMAL OR o . REACTOR POWER - LESS o . TURBINE MANUAL THAN OR EQUAL TO 100%
AND STABLE IF Reactor Power cannot be reduced to less than or equal to 100% OR is NOT stable, THEN do one of the following:
o .!.E. Reactor power is 30% or greater OR Steam Dumps are NOT available, THEN GO TO 1-E-0, REACTOR TRIP OR SAFETY INJECTION.
OR o . IF Reactor power is less than 30%, THEN GO TO 1-AP-2.1, TURBINE TRIP WITHOUT REACTOR TRIP REQUIRED.
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS The Unit is at 100% power, steady state, with NO equipment out of service.
Pressurizer relief tank pressure is below the 8 psig minimum specified in Operator logs.
Nitrogen System is not supplying the power-operated relief valve nitrogen accumulators.
INITIATING CUE You are requested to add nitrogen to the Pressurizer Relief Tank in accordance with 1-0P-5.7, Operation of the Pressurizer Relief Tank (PRT), and raise pressure to 10 psig.
02/25/08 Page: 1 of 8
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM R640 Add Nitrogen to the PRT (1-0P-5.7).
TASK STANDARDS Task was performed as directed by the procedure referenced in the task statement.
KIA
REFERENCE:
007 -A1.02, (2.7/2.9)
ALTERNATE PATH:
NIA TASK COMPLETION TIMES Validation Time = Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ 1SATISFACTORY [ 1UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature 1 Date EVALUATOR'S COMMENTS 02/25/08 Page: 2 of 8
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM R640 READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES The trainee has completed the applicable course knowledge training at the Reactor Operator level.
INITIAL CONDITIONS The Unit is at 100% power, steady state, with NO equipment out of service.
Pressurizer Relief Tank pressure is below the 8 psig minimum specified in Operator logs.
Nitrogen System is not supplying the power-operated relief valve nitrogen accumulators.
02/25/08 Page: 3 of 8
INITIATING CUE You are requested to add nitrogen to the Pressurizer Relief Tank in accordance with 1-0P-5.7, Operation of the Pressurizer Relief Tank (PRT) and raise pressure to 10 psig.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT None PERFORMANCE STEPS START TIME Obtain controlled copy of procedure from Documentum and review Procedure Step _ _
recautions and limitations.
ISAT [1 UNSAT [ 1 IStandards I Precautions and limitations have been reviewed.
Notes/Comments Obtaining and verifying a controlled copy of a procedure is a generic conduct of operations task (2.1.21, 3.5/3.6). If desired by the evaluator a blank copy may be provided to the operator rather than having the operator obtain the copy himself.
02/25/08 Page: 4 of 8
~ Verify that nitrogen is aligned to the supply header. I Procedure Step ISAT [] UNSAT [ ]
rotes/comments Reduce the demand on Containment Nitrogen Supply Header Procedure Step _ _
1-SI-HIC-100 to 0%.
ISAT [] UNSAT [ ]
IStandards j1-SI-HIC-100 demand is reduced to 0%.
IN otes/Com ments
~I Open Nitrogen Supply To CNTMT 1-SI-TV-100. IProcedure Step ICritical Step ISAT [] UNSAT [ ]
Standards OPEN push-buttons for the solenoids 1-SI-TV-100A AND 100B are de ressed.
INotes/Com ments 02/25/08 Page: 5 of 8
~I Open Pressurizer Relief Tank Nitrogen Isol. 1-RC-HCV-1550. IProcedure Step I.....:.C....::....ri~ti-=-c..:.:..al=-S_t~e.L..p 1 SAT [] UNSAT []
IStandards !1-RC-HCV-1550 is open.
I Notes/Comments
~I Open N2 to PRZR Relief Tank Isol. 1-SI-HCV-1898. IProcedure Step ICritical Step ISAT[] UNSAT[]
I..,;;;;S,,;,;;ta=n=da_r,;;;,ds~_ _ 11-SI-HCV-1898 is open.
I Notes/Com ments 02/25/08 Page: 6 of 8
Stop pressurizing the Pressurizer Relief Tank at the desired Procedure Step _ _
ressure.
I_C_ri_ti_c_al_S_t_e'-p 1 SAT [] UNSAT []
Standards Desired pressure of 10 (+/- 1 psig) is reached in the PRT and the following valves are closed:
- 1-SI-HCV-1898 is closed.
- 1-RC-HCV-1550 is closed.
Notes/Comments It is only critical that one of the valves is closed, as closing either one isolates the flowpath.
~ Restore N2 header alignment. I Procedure Step ISAT [] UNSAT [ ]
Standards Alignment restored by performing the following:
- CLOSE push-buttons for the solenoids 1-SI-TV-100A and/or 1008 are depressed.
- 1-SI-HIC-100 demand is reduced to 0%.
IN otes/Com ments
>>>>> END OF EVALUATION <<<<<
STOP TIME 02/25/08 Page: 7 of 8
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
Add Nitrogen to the PRT (1-0P-5.7).
CHECKLIST RecalllC 170.
_ _ Do Simspray.
Ensure rod banks and Tave/Tref recorder are correct.
02/25/08 Page: 8 of 8
PROCEDURE NO:
- VIRGINIA POWER 1-OP-5.7 UNIT NO: REVISION NO:
NORTH ANNA POWER STATION 1 9-Pl PROCEDURE TYPE: EFFECTIVE DATE: EXPIRATION DATE:
OPERATIONS ON FILE N/A PROCEDURE TITLE:
OPERATION OF THE PRESSURIZER RELIEF TANK (PRT)
REVISION
SUMMARY
- Revised to incorporate OP 02-0378 to prevent RCS dilution in the DEGAS mode. Added new Step 5.1.3 to verify the Unit is not in the DEGAS mode. Added Step 5.1.3 Caution and Precaution and Limitation Step 4.4 to address that the Unit must not be in DEGAS mode during PRT draining to the PDTT because PDTT discharge will be aligned to the Gas Stripper in the DEGAS mode and will cause an RCS dilution.
{P I}
Added Section 5.5, Venting PRT to Process Vents via Sample Sink, per Ops request to make this procedure like Unit 2.
Writer: Julius Coppa IReviewer: Ben Spencer ELECTRONIC DISTRIBUTION - APPROVAL ON FILE PROBLEMS ENCOUNTERED: DYes o No NOTE: If yes, note problems in Remarks.
REMARKS:
I (use back for additional space)
SHIFT SUPERVISOR: DATE:
VIRGINIA POWER 1-0P-5.7 NORTH ANNA POWER STATION REVISION 9-P1 PAGE 2 OF 16 1.0 PURPOSE This procedure provides instructions for performing the following actions with the Pressurizer Relief Tank (PRT):
- Draining the PRT
- Filling the PRT
- Venting the PRT to the Gas Strippers
- Using "A" Safety Injection Accumulator to supply Nitrogen to the PRT during RCS draindown using a procedurally controlled Temporary Modification The following synopsis is designed as an aid to understanding the procedure, and is not intended to alter or take the place of the actual purpose, instructions or text of the procedure itself. This procedure provides instructions for performing operations on the PRT. If purging the PRT to the Process Vents via the Vent Pot is desired, the operator should refer to 1-0P-5.l, Filling and Venting the Reactor Coolant System.
In previous Unit 1 outages, an additional controlled source of nitrogen to blanket the PRT has been required due to flow restrictions existing in the installed N2 supply line.
A jumper was installed from the "A" Safety Injection Accumulator vent to a drain for the RHR relief valve discharge line. A section of this procedure will now provide instructions for installation, use, and removal of this jumper.
VIRGINIA POWER 1-0P-5.7 NORTH ANNA POWER STATION REVISION 9-Pl PAGE 4 OF 16 3.0 INITIAL CONDITIONS 3.1 Review the equipment status to verify station configuration supports the performance of this procedure.
4.0 PRECAUTIONS AND LIMITATIONS 4.1 Comply with the following guidelines when marking steps N/A:
- IF the conditional requirements of a step do not require the action to be performed, THEN mark the step N/A.
- IF any other step is marked N/A, THEN have the Shift Supervisor (or designee) approve and justify the N/A on the Procedure Cover Sheet.
4.2 PRT pressure should be maintained between 8 and 14 psig during normal operations.
4.3 PRT pressure should be maintained 2 3 psig while draining.
4.4 Unit 1 MUST NOT be in DEGAS mode during PRT draining to the PDTT. The PDTT discharge will be aligned to l-BR-EV-2A, Gas Stripper in the DEGAS mode and will cause an RCS dilution.
VIRGINIA POWER 1-0P-5.7 NORTH ANNA POWER STATION REVISION 9-Pl PAGE 8 OF 16 5.4 Adding Nitrogen to the PRT 5.4.1 Verify Initial Condition is satisfied.
5.4.2 Review Precautions and Limitations.
5.4.3 Verify nitrogen is aligned to the supply header.
5.4.4 Reduce demand on 1-SI-HIC-100, CONTAINMENT NITROGEN SUPPLY HEADER, to 0%.
NOTE: To open 1-SI-TV-100, the pushbuttons on both the H Safeguards Panel (1-SI-TV-100A) AND the J Safeguards Panel (l-SI-TV-100B) must be used.
5.4.5 Open 1-SI-TV-100, NITROGEN SUPPLY TO CNTMT.
5.4.6 Open 1-RC-HCV-1550, PRESSURIZER RELIEF TANK NITROGEN ISOL.
5.4.7 Open I-SI-HCV-1898, N2 TO PRZR RELIEF TANK ISOL.
VIRGINIA POWER I-OP-5.7 NORTH ANNA POWER STATION REVISION 9-Pl PAGE 9 OF 16 5.4.8 WHEN the desired pressure is obtained in the PRT, THEN do the following:
- a. Close I-SI-HCV-I898, N2 TO PRZR RELIEF TANK ISOL.
- b. Close I-RC-HCV-I550, PRESSURIZER RELIEF TANK NITROGEN ISOL.
NOTE: To close I-SI-TV-IOO, the pushbutton on either the H Safeguards Panel (I-SI-TV-IOOA) or the J Safeguards Panel (I-SI-TV-IOOB) may be used.
If desired, both pushbuttons may be used.
- c. IF the N2 System is NOT supplying the PORV N2 Accumulators, THEN do the following:
- 1. Close I-SI-TV-IOO, NITROGEN SUPPLY TO CNTMT.
- d. IF the N2 System is supplying the PORV N2 Accumulators, THEN lower output of I-SI-HIC-IOO, CNTMT NITROGEN SUPPLY HEADER, to
° percent.
Completed: Date:
Dominion North Anna Power Station CONTROL ROOM JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS Recirculation Spray System is in operation due to a valid Containment Depressurization Actuation (CDA) signal.
Increasing radiation is indicated on Recirculation Spray Heat Exchanger service water outlet radiation monitor 1-RM-SW-126.
Trend recorder 1-RM-RR-100 indicates increasing radiation for radiation monitor 1-RM-SW-126.
High and High-High alarms are illuminated on radiation monitor 1-RM-SW-126.
Control room annunciator 1K-D2, RAD MONITOR SYSTEM HI RAD LEVEL is illuminated.
Control room annunciator K-D4, RAD MONITOR SYSTEM HI-HI RAD LEVEL is illuminated.
High Volume Slowdown of Service Water Reservoir is NOT in service.
INITIATING CUE You are requested to respond to the high radiation indicated on Recirculation Spray Heat Exchanger service water outlet radiation monitor 1-RM-SW-126 in accordance with 1-AP-5.
02/25/08 Page: 1 of 12
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM R775 Respond to a Recirculation Spray Heat Exchanger service water radiation monitor alarm (1-AP-5).
TASK STANDARDS CDA was reset, 1-RS-P-28 was stopped, and SW was isolated to the "C" RSHX.
KIA
REFERENCE:
073-A4.01 (3.9/3.9)
ALTERNATE PATH:
N/A TASK COMPLETION TIMES Validation Time = Start Time = - - -
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ 1SATISFACTORY [ 1UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature I Date EVALUATOR'S COMMENTS 02/25/08 Page: 2 of 12
Dominion North Anna Power Station IN-PLANT JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM R775 READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I prOVided you.
PREREQUISITES The trainee has completed the applicable course knowledge training at the Reactor Operator level.
INITIAL CONDITIONS Recirculation Spray System is in operation due to a valid Containment Depressurization Actuation (CDA) signal.
Increasing radiation is indicated on Recirculation Spray Heat Exchanger service water outlet radiation monitor 1-RM-SW-126.
Trend recorder 1-RM-RR-1 00 indicates increasing radiation for radiation monitor 1-RM-SW-126.
High and high-high alarms are illuminated on radiation monitor 1-RM-SW-126.
02/25/08 Page: 3 of 12
Control room annunciator 1K-D2, RAD MONITOR SYSTEM HI RAD LEVEL is illuminated.
Control room annunciator 1K-D4, RAD MONITOR SYSTEM HI-HI RAD LEVEL is illuminated.
High Volume Blowdown of Service Water Reservoir is NOT in service.
INITIATING CUE You are requested to respond to the high radiation indicated on Recirculation Spray Heat Exchanger service water outlet radiation monitor 1-RM-SW-126 in accordance with 1-AP-5.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT None PERFORMANCE STEPS START TIME 02/25/08 Page: 4 of 12
Checks 1-RM-SW-126 to determine if the alarm is the result of an Procedure Step _ _
obvious radiation monitor malfunction.
ICritical Step ISAT [] UNSAT []
Standards Checks drawer light indication, meter reading not pegged, trend on recorder, etc. to determine Radiation Monitor is NOT failed.
Verbal-Visual When indications are described by the operator, (drawer light indication, Cues meter reading not pegged, trend on recorder, etc.), then confirm the operators observations that the Radiation Monitor is NOT failed.
I Notes/Comments Secure Service Water Reservoir High Volume Slowdown, if in Procedure Step _ _
service.
ISAT [] UNSAT [ ]
Standards Operator determines Service Water Reservoir High Volume Slowdown is NOT in service as iven in the Initial Conditions.
Notes/Comments No cue is needed since information is provided in Initial Conditions.
02/25/08 Page: 5 of 12
3 Request the Health Physics Department to sample the affected Procedure Step _ _
Recirculation Spray Heat Exchanger and to check radiation levels in the Quench spray basement area.
ISAT [1 UNSAT [ 1 I~S;,;;ta;;;,n;,;;d~a;,,;;rd;;;;s~_ _ 1 Operator requests HP to perform sampling and check radiation levels.
Verbal-Visual Health Physics reports that the sample is currently being analyzed and Cues that the radiation level in the Quench Spray basement has increased significantly.
INotes/Com ments Refer to Tech Spec 3.6.7 for the Recirc Spray System Procedure Step _ _
re uirements.
ISAT [1 UNSAT [ 1 Standards Operator identifies the need to refer to Tech Spec 3.6.7.
Verbal-Visual Inform the operator that the SRO will perform this action.
Cues INotes/Comments 02/25/08 Page: 6 of 12
Request the station management to determine if the affected heat Procedure Step _ _
exchan er should be isolated.
ISAT[) UNSAT[)
I..;;;S~ta~n;,;;;d:;;;,a;,;;rd;,;;;s~_ _ 1 Operator contacts Shift Manager for guidance.
Verbal-Visual Station management determines that the affected Recirculation Spray Cues Heat Exchan er should be isolated.
I Notes/Comments
~ Reset both trains of CDA. I Procedure Step I_C_r_iti_c_al_S_t_e-'-p 1 SAT [) UNSAT [)
Standards Both SPRAY ACTUATION RESET switches are placed in the RESET osition.
Verbal-Visual *If the operator identifies the expected response (1 K-H6, CDA Cues INITIATED annunciator CLEAR), then inform him that 1K-H6 appears as he sees it now.
Notes/Com ments *Checking of annunciator status following switch manipulation is not critical; Cue information is only provided in the event the operator states that he would expect the subject annunciator to clear.
02/25/08 Page: 7 of 12
cz.==Jldentify the Recirculation Spray Heat Exchanger to be isolated. IProcedure Step I_C_r_iti_c_al_S_t_e.......
p 1 SAT [1 UNSAT [1 Standards 1-RS-E-1 C is identified as the Recirculation Spray Heat Exchanger to be isolated.
Notes/Comments No cue is needed since information is provided in Initial Conditions.
8 Request the Safeguards operator to place the key-lock switches for Procedure Step _ _
the applicable Recirculation Spray Heat Exchanger's Service Water Isolation Valves in the DEFEAT position.
p I_C_ri_ti_c_al_S_t_e....... 1 SAT [1 UNSAT [1 Standards Safeguards operator is requested to place the key-lock switches for 1-SW-MOV-103C and 1-SW-MOV-104C in the DEFEAT osition.
Verbal-Visual Safeguards operator has placed the key-lock switches for Cues 1-SW-MOV-103C and 1-SW-MOV-104C in the DEFEAT position.
- If identified by the operator, then confirm receipt of the following annunciators:
- Identifying expected annunciator is not critical; Cue information is only provided in the event the operator states that he would expect the subject annunciators.
02/25/08 Page: 8 of 12
~ Stop the applicable Recirc Spray pump. I Procedure Step ICritical Step ISAT [] UNSAT [ ]
I.=S=ta==n=d=a;,;;;rd;,;;s~_ _ 1 Control switch for 1-RS-P-2B is placed in AUTO-AFTER-STOP.
Verbal-Visual Operator will expect pump to be running (RED light LIT, GREEN light Cues NOT LIT) based on initial conditions; if operator indicates he is checking light indication to determine pump status, then confirm RED light LIT, GREEN light NOT LIT, when controls for 1-RS-P-2B are identified.
When the operator identifies the expected change in pump status following the simulated control switch manipulation, then inform the operator that indication lights for the pump are as they see them now.
I Notes/Comments Stop the radiation monitor sample pump for the applicable Procedure Step _ _
Recirculation S ra Heat Exchan er.
I_C_ri_ti_ca_I_S_t_ep'------ 1 SAT [] UNSAT []
S~ta~n,,;,;;d~a~rd~s,,----_ _ Control switch for 1-SW-P-7 taken to STOP position.
t..,;;1 I I Notes/Comm ents 02/25/08 Page: 9 of 12
Close the applicable Recirculation Spray Heat Exchanger's Service Procedure Step _ _
Water Isolation Valves.
I_C_ri_ti_c_al_S_t_e.L-P 1 SAT [1 UNSAT [1 Standards CLOSE push-buttons for recirculation spray heat exchanger isolation valves 1-SW-MOV-103C and 1-SW-MOV-104C are de ressed.
Verbal-Visual Operator will expect valves to be OPEN (RED light LIT, GREEN light Cues NOT LIT) based on initial conditions; if operator indicates he is checking light indication to determine valve status, then confirm RED light LIT, GREEN light NOT LIT, when push-buttons for the MOVs are identified.
When the operator identifies the expected change in valve status following the simulated push-button manipulations, then inform the operator that indication liQhts for the valves are as they see them now.
r oles/Comm enls IJ::~=] Submit a Work Request to initiate repairs. IProcedure Step ISAT [1 UNSAT [ 1 Standards Operator identifies the need to generate a Work Request.
Verbal-Visual Inform the operator that the SRO will perform this action.
Cues INoles/Comm enls 02/25/08 Page: 10 of 12
(
I:JLI Return to the procedure in effect. I Procedure Step ISAT [1 UNSAT [ 1 Verbal-Visual Inform the operator that another operator will complete the procedure.
Cues rotes/com ments
>>>>> END OF EVALUATION <<<<<
STOP TIME 02/25/08 Page: 11 of 12
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
SIMULATOR SETUP JOB PERFORMANCE MEASURE R775 Respond to a recirculation spray heat exchanger service water radiation monitor alarm (1-AP-5).
CHECKLIST
_ _ RecalilC #1 (100% power)
Place the simulator in RUN
_ _ Enter the following malfunctions
- MRS0503, time delay = 0, ramp = 0, degradation = 100%
- MRC0303, time delay = 0
_ _ Perform the immediate and subsequent actions of 1-E-0 to the point of going to 1-E-1
_ _ Ensure that the reactor coolant pumps are secured and that the charging pump recirculation valves are closed Ensure that the Hi Hi alarm is received on 1-RM-SW-126 Place the simulator in FREEZE
_ _ To go to DEFEAT on Monitor, enter the following data
- V2KE3 W=ON
- SWMOV103_NORM(3) = F
- SWMOV104_NORM(3) = F or use Extreme View (includes annunciator for Logic Cabinet)
SW RSHX screen 02/25/08 Page: 12 of 12
~i
~Dominion-NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 26 1-AP-5 UNIT 1 RADIATION MONITORING SYSTEM (WITH THIRTEEN ATTACHMENTS) PAGE 1 of 5 PURPOSE To provide instructions to follow when an abnormal reading or alarm on the Unit 1 Radiation Monitoring System occurs.
ENTRY CONDITIONS This procedure is entered when an abnormal reading or alarm occurs on a Unit 1 Radiation Monitor instrument as indicated by annunciator:
- Panel "K" D-2, RAD MONITOR SYSTEM HI RAD LEVEL
- Panel "K" D-3, RAD MONITOR SYSTEM FAILURE TEST
- Panel "K" D-4, RAD MONITOR SYSTEM HI-HI RAD LEVEL
- Panel "K" G-6, N16 RAD DET
- Panel "K" B-8, N 16 SYS TRBL
- Panel "J" H-1, MAIN STEAM LOOPS 1A-B-C AUX STM LOOP HI RADIATION
- Panel "J" H-2, MAIN STEAM LOOPS 1A-B-C AUX STM LOOP RADIATION ALERT
- Unit 2, Panel "A" B-3, UNIT 1 CONT HI RANGE RADIATION TROUBLE
- Radiation monitor alarm response from 1-AR-32, 1-EI-CP-10, HIGH CAPACITY SG BLOWDOWN CONTROL PANEL CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 26 1-AP-5 UNIT 1 RADIATION MONITORING SYSTEM PAGE 2 of 5 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE:
- Only radiation monitors in alarm or with abnormal indications must be checked. Unaffected radiation monitors should be marked N/A.
- Any other radiation monitor(s) associated with alarming or abnormal radiation monitor(s) should be checked to verify the validity of an alarm.
- 1-RMS-RM-162 is expected to be de-energized in modes 1,2,3,4, and 5.
- 1. VERIFY THE FOLLOWING FOR THE o Initiate the appropriate attachment(s) listed in the AFFECTED RADIATION following table(s).
MONITOR(S):
o
- Indication - NORMAL o
- Trend Recorder - NORMAL o
- Switch Positions - NORMAL o
- Alarms -NOT LIT (STEP 1 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1-AP-5 UNIT 1 RADIATION MONITORING SYSTEM PAGE 3 of 5 ACTION/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 1. VERIFY THE FOLLOWING FOR THE AFFECTED RADIATION MONITOR(S): (Continued)
INIT RADIATION MONITOR RECORDER ATT. NO.
SG STEAM LINE N/A ATTACHMENT 11 1-MS-RM-170, 171, 172 TURBINE DRIVEN AFW PUMP EXHAUST N/A ATTACHMENT 11 1-MS-RM-176 PERSONNEL HATCH AREA 1-RM-RR-100 ATTACHMENT 6 1-RM-RMS-161 MANIPULATOR CRANE 1-RM-RR-100 ATTACHMENT 5 1-RM-RMS-162 CONTAINMENT 1-RM-RR-100 ATTACHMENT 7 1-RM-RMS-163 IN-CORE INST AREA 1-RM-RR-100 ATTACHMENT 7 1-RM-RMS-164 SG AND MAIN STEAM N-16 1-MS-RR-193 ATTACHMENT 2 1-MS-RI-190, 191, 192, 193 REACTOR COOLANT LETDOWN RADIATION 1-RM-RR-100 ATTACHMENT 8 MON RATEMETER, 1-CH-RI-128 DISCHARGE TUNNEL 1-RM-RR-100 ATTACHMENT 4 1-SW-RM-130 SG BLOWDOWN 1-RM-RR-100 ATTACHMENT 9 1-SS-RM-124, 122, 123 CONTAINMENT PARTICULATE 1-RM-RR-100 ATTACHMENT 5 1-RM-RMS-159 (STEP 1 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1-AP-5 UNIT 1 RADIATION MONITORING SYSTEM PAGE 4 of 5 ACTION I EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 1. VERIFY THE FOLLOWING FOR THE AFFECTED RADIATION MONITOR(S): (Continued)
INIT RADIATION MONITOR RECORDER ATT. NO.
CONTAINMENT GASEOUS 1-RM-RR-100 ATTACHMENT 5 1-RM-RMS-160 CONDENSER AIR EJECTOR 1-RM-RR-100 ATTACHMENT 3 1-SV-RM-121 RECIRC SPRAY HX SERVICE WATER OUTLET 1-RM-RR-100 ATTACHMENT 10 1-RM-SW-124, 125, 126, 127 CONTAINMENT HIGH RANGE 1-RM-RR-165 ATTACHMENT 11 1-RM-RMS-165,166 1-RM-RR-166 2._ LOCALLY VERIFY 1-SS-RM-125, D Initiate ATTACHMENT 9.
HIGH CAPACITY SG BLOWDOWN RAD MONITOR AS REQUIRED:
D
- Indication - NORMAL D
- History trends - NORMAL D
- Switch positions - NORMAL D
- Alarms - NOT LIT
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-5 10 RECIRC SPRAY HX SERVICE WATER OUTLET REVISION RADIATION MONITORS PAGE 26 1 of 4 CAUTION: Extremely high radiation levels could exist in the Quench Spray Basement area during accident conditions.
- 1. !E the abnormality of the Radiation Monitor was caused by an obvious malfunction, THEN do the following:
a) IF Radiation Monitor has failed high AND reset is desired by removing fuses, THEN do the following:
- 1) Remove Radiation Monitor fuses.
- 2) Install Radiation Monitor fuses.
NOTE: If the Radiation Monitor resets, then a Work Request is not required. The Condition Report will be assigned to the Radiation Monitor Engineer for trending and Work Request evaluation.
- 3) IF the Radiation Monitor resets, THEN do the following. !E NOT, THEN GO TO Step 1b:
- Submit a Condition Report.
- RETURN TO 1-AP-5, UNIT 1 RADIATION MONITORING SYSTEM, Step in effect.
_ b) Inform the HP Shift Supervisor of the date and time that the monitor was declared inoperable.
_ c) Submit a Work Request and Condition Report.
_ d) RETURN TO procedure and step in effect.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-5 10 RECIRC SPRAY HX SERVICE WATER OUTLET RADIATION REVISION PAGE MONITORS 26 2 of 4
- 2. !.E Recirc Spray is in service AND the Radiation Monitor has NOT malfunctioned, THEN do the following:
a) IF high volume blowdown of Service Water Reservoir is in service, THEN do the following:
- 1) Close at least one of the following valves:
- 1-SW-MOV-120B, SERVICE WATER DISCH TO CIRC WATER TUNNEL
- 2-SW-MOVc 220A, SERVICE WATER DISCH TO CIRC WATER TUNNEL
- 1-SW-1351, SW Blowdown Iso Valve
- 2) Secure high volume blowdown in accordance with 0-OP-49.7, High Volume Blowdown Of The Service Water Reservoir.
_ b) Have Health Physics obtain and analyze a Service Water sample from the affected Heat Exchanger AND check the area Radiation levels in Quench Spray Basement.
_ c) !.E a significant increase in sample radioactivity is detected, THEN have Station management determine if the affected Recirc Spray Heat Exchanger(s) should be isolated.
- 3. Refer to Tech Spec 3.6.7 for the Recirc Spray System requirements.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-5 10 RECIRC SPRAY HX SERVICE WATER OUTLET RADIATION REVISION PAGE MONITORS 26 3 of 4 NOTE: If one Recirc Spray Heat Exchanger's Service Water MOV fails to fully close, then the other MOV for that Heat Exchanger should be closed. When that MOV is fully closed, then the MOV that failed to fully close should be reopened, then closed.
- 4. lE Station management determined that the affected Recirc Spray Heat Exchanger(s) should be isolated, THEN do the following:
_ a) Ensure both trains of CDA are Reset using the Spray Actuation Reset Switches.
b) To isolate 1-RS-E-1A, do the following:
- 1) Place the keylock switches for 1-SW-MOV-1 03A and 1-SW-MOV-1 04A (located in Service Water logic cabinet 1-EP-CB-28H) to DEFEAT.
- 2) Stop 1-RS-P-1 A.
- 3) Stop 1-SW-P-5 (sample pump).
- 4) Close 1-SW-MOV-103A.
- 5) Close 1-SW-MOV-1 04A.
c) To isolate 1-RS-E-1 B, do the following:
- 1) Place the keylock switches for 1-SW-MOV-1 03B and 1-SW-MOV-1 04B (located in Service Water logic cabinet 1-EP-CB-28J) to DEFEAT.
- 2) Stop 1-RS-P-1 B.
- 3) Stop 1-SW-P-6 (sample pump).
- 4) Close 1-SW-MOV-103B.
- 5) Close 1-SW-MOV-104B.
d) To isolate 1-RS-E-1 C, do the following:
- 1) Place the keylock switches for 1-SW-MOV-1 03C and 1-SW-MOV-1 04C (located in Service Water logic cabinet 1-EP-CB-28J) to DEFEAT.
- 2) Stop 1-RS-P-2B.
- 3) Stop 1-SW-P-7 (sample pump).
- 4) Close 1-SW-MOV-103C.
- 5) Close 1-SW-MOV-104C.
NUMBER ATIACHMENT TITLE ATIACHMENT 1-AP-5 10 RECIRC SPRAY HX SERVICE WATER OUTLET RADIATION REVISION PAGE MONITORS 26 4 of 4 e) To isolate 1-RS-E-1 D, do the following:
- 1) Place the keylock switches for 1-SW-MOV-1 03D and 1-SW-MOV-104D (located in Service Water logic cabinet 1-EP-CB-28H) to DEFEAT.
- 2) Stop 1-RS-P-2A.
- 3) Stop 1-SW-P-8 (sample pump).
- 4) Close 1-SW-MOV-103D.
- 5) Close 1-SW-MOV-1 04D.
- 5. Submit necessary Work Requests to make repairs to any Recirc Spray Heat Exchanger(s) that were taken out of service.
- 6. RETURN TO procedure and step in effect.
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM INITIAL CONDITIONS Unit 1 is operating at 30% power holding for chemistry.
Unit 2 is in Mode 5 for scheduled refueling outage.
Circulating Water Pump 1-CW-P-1A is OOS for oil replacement.
CW PP 1A-1 B-1 C-1 0 AUTO TRIP annunciator (1 B-A5) has just been received.
INITIATING CUE You are requested to respond to plant conditions in accordance with 1-AP-13.
03/04/08 Page: 1 of 10
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE EVALUATION OPERATOR PROGRAM R517 (modified)
Respond to annunciator 1B-A5, CW PP 1A-1 B-1C-1 D AUTO TRIP.
TASK STANDARDS Immediate operator actions performed without reference to procedure. liquid waste releases secured in accordance with 1-AP-13.
KIA
REFERENCE:
075-A2.02 (2.5/2.7)
ALTERNATE PATH:
NIA TASK COMPLETION TIMES Validation Time = 15 minutes Start Time = _
Actual Time = minutes Stop Time = _
PERFORMANCE EVALUATION Rating [ ] SATISFACTORY [ ] UNSATISFACTORY Candidate (Print)
Evaluator (Print)
Evaluator's Signature 1 Date EVALUATOR'S COMMENTS 03/04/08 Page: 2 of 10
Dominion North Anna Power Station SIMULATOR JOB PERFORMANCE MEASURE (Evaluation)
OPERATOR PROGRAM R517 (Modified)
READ THE APPLICABLE INSTRUCTIONS TO THE CANDIDATE Instructions for Simulator JPMs I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
Instructions for In-Plant JPMs I will explain the initial conditions, and state the task to be performed. All steps, including any required communications, shall be simulated for this JPM. Under no circumstances are you to operate any plant equipment. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.
PREREQUISITES The trainee has completed the applicable course knowledge training at the reactor operator level.
INITIAL CONDITIONS Unit 1 is operating at 30% power holding for chemistry.
Unit 2 is in Mode 5 for scheduled refueling outage.
Circulating Water Pump 1-CW-P-1A is OOS for oil replacement.
CW PP 1A-1 B-1 C-1 D AUTO TRIP annunciator (1 B-A5) has just been received.
03/04/08 Page: 3 of 10
INITIATING CUE You are requested to respond to plant conditions in accordance with 1-AP-13.
EVALUATION METHOD Demonstration if conducted in the simulator or in a laboratory (use DEMONSTRATION cues)
Verbal-visual if conducted in the station or on a dead simulator (use VERBAL-VISUAL cues)
TOOLS AND EQUIPMENT None PERFORMANCE STEPS START TIME Note: All required cues are provided from the Booth operator
~ Verify that at least two circulating water pumps are running - YES. IProcedure Step ICritical Step ISAT [] UNSAT []
Standards Candidate checks panel and determines Two (2) Circulating Water Pump are running, (observes breaker indication, motor amps, discharge valve position, etc.).
INotes/Comm ents 03/04/08 Page: 4 of 10
~ Check Turbine Load Control. IProcedure Step ISAT [] UNSAT [ ]
Standards Checks EHC Panel and verifies the following:
- Turbine Valve Position Limiter Light - NOT LIT.
- Turbine Load Control, IMP-IN Light - LIT.
I Notes/Com ments
~I Verify Condenser Vacuum < 3.5" Hg and stable - YES. IProcedure Step ISAT [] UNSAT [ ]
Standards Verifies that Condenser vacuum stable at < 3.5 " Hg using recorder on vertical anel and/or PCS.
I Notes/Comments 03/04/08 Page: 5 of 10
~ Check operating CWP amps less than 340 amps - YES. IProcedure Step ISAT [] UNSAT [ ]
Standards Verifies operating CWP (1 Band 1D) amps less than 340 amps.
INotes/Com ments
~ Check bearing cooling system support. I Procedure Step ISAT [] UNSAT [ ]
Standards Verifies the following:
- Circ Water Intake Tunnel Full Light - LIT.
INotes/Comments 03/04/08 Page: 6 of 10
ILJ Determine if securing liquid waste releases required - YES. IProcedure Step ICritical Step ISAT[] UNSAT[]
Standards Contacts HP and informs them of CWP status and requests determination as to if releases need to be secured.
Notes/Comments HP will advise the operator that the Liquid Release Permit in effect requires a minimum of 3 CWPs running on Unit 1.
~ Verify SG Slowdown Trip Valves closed. IProcedure Step 1,--~_r_it_ic_a_1S.:. . . t.:. . . e-'-p 1 SAT [] UNSAT [1 Standards
- Closes SG Slowdown Trip Vavles on H Safeguards Panel (per RNO).
- Closes SG Slowdown Trip Vavles on J Safeguards Panel (per RNO).
- Verifies with Unit 2 OATC that Unit 2 Slowdown Trip Valves are closed.
Notes/Comments It is only critical to close 1 set of SG Slowdown Trip Valves (H panel or J panel) as closing either of the valves, which are in series isolates the flow path.
03/04/08 Page: 7 of 10
~ Locally close BO tank isolation valves. IProcedure Step ISAT [] UNSAT [ ]
Standards Operator dispatched to close the following valves:
- 1-BO-1005
- 2-BO-182 I Notes/Comm ents
~ Stop the following pumps to prevent clarifier overflow. IProcedure Step ICritical Step ISAT [] UNSAT [ ]
Standards
- 1-LW-P-6A and 1-LW-P-6B are stopped.
- 1-LW-P-1A and 1-LW-P-1 B are verified stopped.
- 1-BR-P-5A and 1-BR-P-5B are verified stopped.
Notes/Comments It is only critical to stop 1-LW-P-6A and 1-LW-P-6B as the other pumps are not running.
03/04/08 Page: 8 of 10
(
~I Locally secure Containment Matt Sump Pumps. I Procedure Step ISAT [] UNSAT [ J Standards Operator dispatched to perform the following:
- Open SOV breakers 1-EP-DB-2 bkr 6 & 7 AND 2-EP-DB-5 bkr 6 & 7.
- Verify Pumps are stopped.
INotes/Comm ents o.:r==l Verify Liquid waste release secured. IProcedure Step I_C_ri_ti_c_al_S_t_e"'_--p 1 SAT [] UNSAT []
Standards Operator places 1-LW-PCV-115 control switch in HAND and verifies valve closed (per RNO).
Notes/Comments At this time another operator will complete the rest of the procedure.
>>>>> END OF EVALUATION <<<<<
STOP TIME 03/04/08 Page: 9 of 10
SIMULATOR, LABORATORY, IN--PLANT SETUP (If Required)
SIMULATOR SETUP JOB PERFORMANCE MEASURE R517 (Modified)
You are requested to respond to plant conditions in accordance with 1-AP-13 CHECKLIST
_ _ Recall IC #172 (30% power)
_ _ Place clearance info stickers on 1-CW-P-1A control switch AND discharge MOV.
Booth Operator Actions:
If dispatched to investigate trip of 1-CW-P-1 C then wait approximately 3 minutes and respond that you see nothing obviously wrong with the pump but the breaker does have an over-current target dropped.
At Step 6 respond as HP that the Liquid Release Permit in effect requires a minimum of 3 CWPs running on Unit 1. If asked to confirm that releases need to be secured concur with the operator.
If called at Step 7 respond as Unit 2 and confirm that all SG Blowdown Trip Valves on Unit 2 are closed.
Respond as operator to perform Step 7b of the procedure to close valves 1-BO-1 005 and 2-BO-182 (no manipulations are necessary since these actions are invisible to the OATC) - call back in
- 5 minutes with actions complete.
Respond as operator to perform Step 7d of the procedure to secure Matt Sump Pumps by opening SOV breakers 1-EP-OB-2 BKR 6 & 7 AND 2-EP-OB-5 BKR 6 & 7 AND verifying Pumps are stopped (no manipulations are necessary) - call back in - 5 minutes with actions complete if JPM is still in progress.
03/04/08 Page: 10of10
VIRGINIA POWER NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 1- AP-13 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS 13 PAGE (WITH ONE ATTACHMENT) 1 of 7 PURPOSE To provide instructions to follow in the event that one or more Circulating Water Pumps are lost.
ENTRY CONDITIONS This procedure is entered when any of the following conditions exist:
- TRIP indication on breaker control switches on the Circulating Water Control Panel. or
- One or more Circulating Water Pump Motors read zero amps. or
- Decreasing Condenser vacuum. or
- Annunciator Panel "B" A-5. CW PP 1A-1B-1C-1D AUTO TRIP, is LIT.
RECOMMENDED APPROVAL: DATE EFFECTIVE RECOMMENDED APPROVAL - ON FILE DATE APPROVAL: DATE APPROVAL - ON FILE
NUMBER PROCEDURE TITLE REVISION 13 1- AP-13 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 2 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1] __ VERIFY AT LEAST TWO CIRCULATING GO TO 1-E-0. REACTOR TRIP OR WATER PUMPS- RUNNING SAFETY INJECTION while continuing with this procedure.
- 2. CHECK TURBINE LOAD CONTROL:
a) Verify Turbine valve position - a) Take Turbine off Valve Position OFF VALVE POSITION LIMITER Limiter.
b) Verify Turbine Load Control in b) Place Turbine Load Control in IMP-IN. IMP-IN by depressing the IMP-IN pushbutton.
- 3. VERIFY CONDENSER VACUUM - STABLE IF condenser vacuum cannot be AT 3.5 INCHES HG ABS OR LESS maintained. THEN initiate 1-AP-14.
LOW CONDENSER VACUUM. while continuing with this procedure.
- 4. CHECK OPERATING CIRCULATING WATER Reduce operating Circulating Water PUMP AMPS - LESS THAN 340 AMPS Pump amps by doing one of the following:
- Throttle discharge MOVs.
- Reduce number of water boxes in service.
NUMBER PROCEDURE TITLE REVISION 13 1-AP-13 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 3 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 5. VERIFY BEARING COOLING SYSTEM IF Bearing Cooling is aligned SUPPORT: Lake-to-Lake. THEN do the following:
- Circ Water Intake Tunnel Full -
LIGHT LIT
- Place l-WT-P-25 in service using 1-0P-48.4. OPERATION OF
- At least one Circulating Water AUXILIARY FLASH EVAPORATOR PUMP pump - RUNNING l-WT- P- 25.
- Align Bearing Cooling to Tower-to-Tower using 1-0P-50.2.
OPERATION OF THE BEARING COOLING WATER SYSTEM.
GO TO 1-AP-19. LOSS OF BEARING COOLING WATER. while continuing with this procedure.
- 6. __ DETERMINE IF SECURING LIQUID WASTE RELEASES REQUIRED:
a) Have HP Count Room determine if securing liquid waste releases is required based on Circulating Water system status b) Health Physics has determined b) GO TO Step 8.
securing liquid waste releases is required
NUMBER PROCEDURE TITLE REVISION 13 1-AP-13 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 4 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7. SECURE LIQUID WASTE RELEASES:
a) Verify SG Blowdown Trip Valves a) Do Sub step 1) OR 2) below as
- CLOSED: required:
- Unit 2 SG Blowdown Trip valves.
- 2) Locally close the following:
- 2-BD-182. SG BD FLASH TK TO SG BD FLASH TK DRN CLRS ISOL VV c) Stop the following pumps to prevent clarifier system overflow:
- Contaminated Drain Tank Pumps:
- 1-LW-P-6A
- 1-LW-P-6B
- Low Level Liquid Waste Tank Pumps:
- 1-LW-P-1A
- 1-LW-P-1B
- Boron Recovery Test Tank Pumps:
- 1-BR-P-SA
- 1-BR-P-SB (STEP 7 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 13 1- AP-13 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 5 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7. SECURE LIQUID WASTE RELEASES (Continued):
d) Send an operator to stop Containment Matt Sump Pumps:
- 1) Locally open SOV breakers:
- Unit 1 EP-DB-2.
Breakers 6 AND 7
- Unit 2 EP-DB-S.
Breakers 6 AND 7
- 2) Verify Pumps are stopped 2) Manually isolate Service Air to Pump(s).
e) Verify Liquid Waste release e) Place 1-LW-PCV-11S control in secured by verifying HAND and close.
1-LW-PCV-11S - IN HAND CONTROL AND CLOSED
- 8. VERIFY LP TURBINE RUPTURE DISCS Secure condenser air ejectors INTACT using 1-0P-36.2. MAIN CONDENSER AIR EJECTOR SYSTEM:
- WHEN condenser vacuum reaches zero. THEN secure Gland Steam System using 1-0P-39.1. GLAND SEAL STEAM SYSTEM .
- Submit Work Request to replace rupture discs.
- 9. CHECK HOTWELL TEMPERATURE - LESS Bypass the Powdex system using THAN 130°F 1-0P-30.2. POWDEX SYSTEM.
NUMBER PROCEDURE TITLE REVISION 13 1-AP-13 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 6 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 10. __ VERIFY LIQUID WASTE RELEASE Do the following:
PATHS - AVAILABLE:
a) Obtain Health Physics
- Verify liquid waste release concurrence on alternate Liquid forms in effect Waste Release paths.
- Re-establish liquid waste releases using 0-OP-22.11, b) Initiate liquid waste release RELEASING RADIOACTIVE LIQUID to Unit 2 discharge tunnel WASTE using 0-OP-22.11. RELEASING
- Re-establish Unit 1 High RADIOACTIVE LIQUID WASTE.
Capacity SG Blowdown using 1-0P-32.3, HIGH CAPACITY STEAM GENERATOR BLOWDOWN SYSTEM OPERATION or Unit 1 Low Capacity SG Blowdown using 1-0P-32.1, LOW CAPACITY STEAM GENERATOR BLOWDOWN SYSTEM (if desired)
- Re-establish Unit 2 High Capacity SG Blowdown using 2-0P-32.3. HIGH CAPACITY STEAM GENERATOR BLOWDOWN SYSTEM OPERATION or Unit 2 Low Capacity SG Blowdown using 2-0P-32.1, STEAM GENERATOR BLOWDOWN UTILIZING STEAM GENERATOR BLOWDOWN TANK 2-BD-TK-1 (if desired)
NUMBER PROCEDURE TITLE REVISION 13 l-AP-13 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 7 of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION: If the loss of Circulating Water Pumps was caused by SI/CDA load shed.
then O-AP-47. UNIT OPERATION DURING OPPOSITE UNIT EMERGENCY. must be used for instructions on resetting SI/CDA load shed. Circulating Water Pump Breaker control switches must be placed to TRIP position before load shed is reset to prevent a possible auto start of Circulating Water Pumps.
- 11. DETERMINE STATUS OF CIRCULATING WATER PUMP(S) :
a) Work Request(s) - SUBMITTED a) Submit Work Request(s) .
b) Affected Circulating Water b) Continue with other procedures Pump(S) - AVAILABLE in effect. WHEN affected pumps are available. THEN GO TO Step 12.
- 12. RESTORE CIRCULATING WATER SYSTEM USING 1-0P-48.2. OPERATION OF THE CIRCULATING WATER SYSTEM
- 13. RETURN TO PROCEDURE AND STEP IN EFFECT
- END -