ML20147B751
| ML20147B751 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 01/24/1997 |
| From: | Michael B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9702030149 | |
| Download: ML20147B751 (160) | |
Text
{{#Wiki_filter:. . . - __ . _ . _ _ . . _ _ _ _ . _ . . . _ . . _ _ _ _ . - . . . . _ . . _ . - _ . _ . ._. . I l UNITED STATES l crog%, NUCLEAR REGULATORY COMMISSION g- a
" REGloN 11 101 MARIETTA STREET, N.W., SUITE 2900
, a j ATLANTA, GEORGIA 303234199 oss** , I January 24, 1997 NOTE T0: NRC Document Control Desk j Mail Stop 0-5-D-24 ! FROM: Beverly Michael, Licensing Assistant
- Operating Licensing and Human Performance Br ch, Region II j
SUBJECT:
OPERATOR LICENSING RETAKE EXAMINATION ADMINISTERED ON DECEMBER 18, 1996, AT NORTH ANNA POWER STATION - DOCKET NOS. 50-338 AND 50-339 (EXAMINATION REPORT 50-338/96-301) ] 1 l 1 a On December 18, 1996, Operator Licensing Examinations were administered at l i the referenced facility. Attached, you will find the following ! information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR: i Item #1 - a) Facility submitted outline and initial exam submittal, 3 designated for distribution under RIDS Code A070. 4 b) As given operating examination, designated for
- distribution under RIDS Code A070.
l Item #2 - Examination Report already submitted. Written exam - N/A. i f l l 1 l 9702030149 970124 PDR ADOCK 05000338 V PDR l I
______.__.m_ . . ._ 2 i { ! ES-301 Individual Walk-thmga Ta Outline - Set #1
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Form ES-301-2 i i i Exanunation Level (Circle One): @ SRO(1) / SRO(U) .Ibc/(, j i i Facility: North Anna Week of Exammation: Jam,-29,1996 l Exanuner's Name (print): i j System / JPM Safety Planned Follow-up Questions-l Function K/A/O // Importance // Description j i t l 1. Control Rod Drive - I-001 a. 003AA2.03 (3.6/3.8) i Retrieve a dropped rod SIM Effects of dropped rod on major plant parameters. I (R.. 476)
- b. 001K4.03 (3.5/3.8)
Given a set of conditions, deternme effect on rod control. 4 l 2. Chemical and Volume II-004 a. GEN-2.1.25 (2.8/3.1) j Control - Place excess MCR Given a set of plant conditions, use graphs to determine j { letdown in service (R.. 337j blended flow. I re-write for unit-2 { b, GEN-2.1.32 (3.4/3.8) j Describe the reasons for the procedure precautions. j 3. Emergency Core Cooling - III-006 s. 011EK3.13 (3.8/4.2) ' l Transfer the Safety Injection SIM Describe the reasons for swapping from hot-leg back to System from hot-leg to cold- cold-leg injection. 2 leg injection (R.. 736) ALT. ! ^ 'I I ) PATH / ESF / NEW l Actions required following a spurious SI 1 i 4. Residual Heat Removal - IV-005 a. GEN-2.1.20 (4.3/4.2) l Restore RHR Cooling ALT. SIM Given a set of conditions, determine required actions. ! PATH / SHUTDOWN / NEW (LOSS OF CC TO b. 005A2.04 (2.9/2.9) RHR HXs, R.. 514) Affect of loss of instrument air on RCS temperature. i l ,~i v w i 1 i a 1O , e A% eo E i 4
___l 1
- 5. Reactor Coolant Pump - IV-003 a. GEN-2.1.32 (3.2/3.3)
Start a rektor coolant pump SIM Given a set of conditions, rh RCP start limitations.
~
(SEAL DELTA-P IS LOST,
,6)
R.164) Given a set of plant conditions, determine the effect on calorimetric.
- 6. Contamment Spray - Align V-028 s. 028K5.01 (3.4/3.9):
contammmt spray systems SIM Given a set of plant conditions, determine hazards. (R..216)
- b. GEN-2.1.28 (3.2/.3.3)
Purpose of sample line heat tracing.
- 7. Pressurizer Pressure 111-010 a. 010A4.03 (4.0/3.8)
Control - Place NDT SIM Given a set of conditions describe response of PRZR protection in service during a PORV block valve. natural circulation cooldown OAl.M (3 A/3@ (R. 577) SHUTDOWN Given a set of conditions, determine tail pipe temperature
- 8. Emergency Diesel VI-064 a. 064A3.06 (3.3/3.4)
Generating - Unload and IN-PLANT Given a set of conditions, determine control of EDG. I shutdown an EDG in the control room emergency mode b 064K4.06 (2.8/3.2) (N.. 466) AP ACIION Describe when speed droop is in effect.
- 9. Auxiliary Feedwater - IV-061 a. 061 A2.06 (2.7/3.0)
Collapse steam voids in a IN-PLANT Affect of check-valve back-leakage. j steam-bound AFW pump
- b. 061K4.02 (4.5/4.6) l (N.. 935)
Given a set of conditions, apply interlocks to control of AFW pump.
- 10. Component Cooling VIII-008 a. 008K4.01 (3.1/3.3)
Water System - Drain the CC IN-PLANT Given a set of plant conditions, determine response of CC system to a LW test tank pumps. (N.. 230) RCA
- b. 008A1.01 (2.8/2.9)
Affects on CC flow from shifting common loads. Examiner: Chief Exanuner: - . _ . . _ _ _ _ _ _ _ _ _ _ - .. _~ _ - - - , - _.
1 i i. 1 i i 1 1 l 4 1 I 4 i. NORTII ANNA POWER STATION I i ADMINISTRATIVE WALK-TIIROUGII i I $ SET 1 .O i i o 1 1 O
ES-301 Administrative Topics Outline - SET # 1 Form ES-301-1 p Examination Level (Circle One): RO / SRO Facility: North Anna Week of Examination: Dec. 16, 1996 Examiner's Name (print): Administrative Describe method of evaluation: Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Status Control JPM - Enter a component into abnormal status. I l A.1 Shift Turnover Given a set of conditions, determine turnover requirements. Given a set of conditions, determine qualification of relief. l A.2 Maintenance JPM - Given a set of plant conditions, prioritize methods of plant cooling. A.3 Radiation Work JPM - Review a Radiation Work Permit and obtain a DAD. 'l Practices NOTE: This JPM will be incorporated into the "RCA" task, N.. 230 (JPM # 10). A.4 Emergency Given a set of conditions, determine required actions. Facilities List those facilities with protection from radiation / airborne. t Examiner: Chief Examiner: L
a 1 J s' 2 i, J ) 4 1 ,1 i. 1 1 k i I I i i 4 s l f ' ABNORMAL STATUS JPM I i e 4 i ,I l l
i . j- Virginia Power North Anna Power Station J
, REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Administrative i j Job Performance Measure i ,
4 i
. INITIAL CONDITIONS ,
- It has been determined during the unit 1 turbine building watchstander's rounds, that the main steam l to aux, steam PCV,1-AS-PCV-105, drifts open with its M/A station in automatic.
- The instrument department has identified the problem, but repair parts will not arrive on sight until 3 January 15th,1997.
i l Until it can be repaired, the US has directed that the M/A station be placed in manual with zero
- demand.
! l { The US has directed that t.nis action be documented in abnormal status.
- i j INITIATING CUE You are to enter the above into abnormal status, i
i a } } O
Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Administrative 4 { Job Performance Measure 1 i Candidate Evaluator Evaluation Date j Performance Evaluation Satisfactory Unsatisfactory i 4 TASK !, I
; Enter a component into Abnormal Status i
NOTE TO THE TRAINER AND THE EVALUATOR 4 Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure. !O INITIAL CONDITIONS t It has been determined during the unit 1 turbine building watchstander's rounds, that the main steam
; to aux. steam PCV,1-AS-PCV-105, drifts open with its M/A station in automatic.
e
; The instrument department has identified the problem, but repair parts will not arrive on sight until . January 15th,1997.
i Until it can be repaired, the US has directed that the M/A station be placed in manual with zero demand. 1
! The US has directed that this action be documented in abnormal status.
INITIATING CUE You are to enter the above into abnormal status. O
STANDARDS Self-checking practices were used throughout task performance Verbal communication related to any of the following modes was conducted in accordance with VPAP-1407 : e Emergency communication e Face-to-face communication e Giving and acknewledging orders e Phonetic alphabet e Telephone communication systems l TOOLS AND EOUIPMENT Simulator Ops Network Computer l EVALUATION METHOD l Demonstrate 1 PERFORMANCE STEPS ' O V1. Select VPPEQS from menu. ; i SAT [] UNSAT [ ] NOTE I 1 1
.. 1
- 2. Enter badge number and password.
SAT [ ] UNSAT [] NOTE
- 3. Select "B. Find by Type".
SAT [ ] UNSAT [ ] NOTE
- 4. Select "C. Abnormal Status".
SAT [ ] UNSAT [ ] NOTE
- 5. Select "A. Unit 1 and Common".
SAT [] UNSAT [ ] NOTE
e i .
- 6. Press " Insert".
i
- SAT [ ] UNSAT [] NOTE
!O 7. , Enter equipment mark number 1-AS-PCV-105 and required information. SAT [] UNSAT [ ] NOTE
> > > > > END OF EVALUATION < < < < <
O O
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l 4 4 l ES-301 Administrative Topics Outline - SET #1 Form ES-301-1 1 ! 'vTOPIC: Shift Turnover ! QUESTION #1 i
, gp < yJgggg ; gjpglf1;g g flgp ;ggg j g g g - 949% gg gg j jjgg j Given the following conditions: If a Control Room Operator with unit duty 2
expects to be away from the assigned station j e You are the unit-2 OATC. for situations other than meal and restroom j e The SS has just informed you that you breaks, then a complete, formal relief shall
- must report to " Fitness For Duty" for a be performed.
j random test. l REF. Vision Objective: 13598, OPAP-0005 j Describe the turnover requirements. section 6.2.3 1 j QUESTION #2 4+4-m..-
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[S/U$b In the above situation, could Mike No, the individual does not hold an active Burnette relieve you? license. Determined by referencing Virginia j Power Personnel Qualification System ? (VPPQS) data base. l
- REF. Vision Objective
- 13597, OPAP-0005 j sections 5.1.5 & 6.1.14 t
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a 4 e tO h l = 4 CORE COOLING ASSESSMENT JPM O O
f Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS
- Administrative Job Performance Measure INITIAL CONDITIONS e Unit I has been shutdown for 155 hours, e RCS temperature is 123*F.
e RCS is depressurized to atmospheric. e "A" RCS loop stops are closed. e "B" and "C" loop stops are open but de-energized, e Both PRZR PORVs are blocked open.
- One PRZR safety valve is removed.
e PRZR level is 20%. e One train of SI (HHSI and LHSI) is operable with both hot and cold leg injection paths. e The RSWT is at normal operating level. e The main condenser is open for maintenance, e All SG levels are at 33%. e The AFW system is tagged out for maintenance, e All other systems are operable and in the expected state for this mode of operation. INITIATING CUE You are to determine the Alternate Core Cooling Assessment IAW 1-GOP-13.0,1-GOP-13.1 for normal surveillance. You are not t.' determine time to boiling. O
Virginia Power North Anna Power Station l l O V REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Administrative Job Performance Measure Candidate Evaluator Evaluation Date Performance Evaluation Satisfactory Unsatisfactory TASK Determine the Alternate Core Cooling Method (s). NOTE TO TIIE TRAINER AND THE EVALUATOR Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component l or parameter is in the condition specified by the procedure. 1 INITIAL CONDITIONS a n i U e Unit I has been shutdown for 155 hours. e RCS temperature is 123*F. e RCS is depressurized to atmospheric. 1 e "A" RCS loop stops are closed. j e "B" and "C" loop stops are open but de-energized.
- e Both PRZR PORVs are blocked open.
e One PRZR safety valve is removed. e PRZR level is 20%. l
; e One train of SI (HHSI and LHSI) is operable with both hot and cold leg injection paths.
- The RSWT is at normal operating level.
e The main condenser is open for maintenance.
;
- All SG levels are at 33%.
e The AFW system is tagged out for maintenance. e All other systems are operable and in the expected state for this mode of operation. INITIATING CUE You are to determine the Alternate Core Cooling Assessment IAW 1-GOP-13.0,1-GOP-13.1 for normal surveillance. You are not to determine time to boiling. i
STANDARDS Task was performed as directed by the procedure referenced in the task statement within parentheses (one of the underlined procedures if several are cited) Self-checking practices were used throughout task performance Verbal communication related to any of the following modes was conducted in accordance with VPAP-1407 e Emergency communication i e Face-to-face communication e Giving and acknowledging orders e Phonetic alphabet e Telephone communication systems TOOLS AND EOUTPMENT None EVALUATION METHOD Verbal-visual ERFORMANCE STEPS
- 1. Enter reason for assessment STANDARD Records " surveillance".
SAT [ ] UNSAT [ ] NOTE
- 2. Verify Natural Circulation available.
STANDARD Determines that Nat. Circ. is not available due to AFW and main condenser are tagged out. MT t ,1 UNSAT [ ] NOTE v
- 3. Verify Reflux Boiling available.
! STANDARD Determines that Reflux Boiling is not available due to AFW and main condenser are tagged out.
SAT [ ] UNSAT [ ] NOTE i
- 4. Verify Forced Feed and Spill available.
j STANDARD Determines that: a.1 PRZR safety valve is removed,1 LHSI pump is available, time after shutdown is I hour, and RWST levelis 50%. AND/OR
- b. 2 PRZR PORVs are blocked open,1 LHSI pump is available, time after shutdown is 26 hours, and
. RWST levelis 50%.
l CRITICAL STANDARD 4 Denotes Forced Feed and Spill as Priority 1. i SAT [ ] UNSAT [] NOTE I
- 5. Verify Gravity Feed and Spill available.
] I i STANDARD i j Determines that: ! 1 PRZR safety valve i.; removed, time after shutdown is 58 hours,1 train of SI operable, RWST level is approx. 82%. CRITICAL STANDARD j Denotes Gravity Feed and Spill as Priority 2. SAT [ ] UNSAT [ ] NOTE , > > > > > END OF EVALUATION < < < < < 4 i i i !O 9 b
4 t 4 i a 1 i, 4 i i
- I l
t l 4 1 4 t 1 i 1 i -k ? ) 1 RWP REVIEW JPM l 4 (INTEGRATED INTO JPM #10) 1 i I I A l i. i l i i i 5 w I i i 4 1 4 i d e 1 I i l' 2 i. J 1
- - _ . - . - -- =. . .
1
- Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Administrative i
Job Performance Measure 5 ! INITIAL CONDITIONS l l l
- You have been assigned a task in the auxiliary building. '
INITIATING CUE You are to review the applicable RWP and obtain a DAD. O O
Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Administrative Job Performance Measure Candidate Evaluator _ Evaluation Date Performance Evaluation Satisfactory Unsatisfactory TASK Review an RWP and obtain a DAD. NOTE TO THE TRAINER AND THE EVALUATOR Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure. You have been assigned a task in the auxiliary building. INITIATING CUE You are to review the applicable RWP and obtain a DAD. O
STANDARDS Task wa performed as directed by the procedure referenced in the task statement within parentheses (one of the underlined procedures if several are cited) Self-checking practices were used throughout task performance Verbal communication related to any of the following modes was conducted in accordance with VPAP-1407 e Emergency communication e Face-to-face communication I e Giving and acknowledging orders l e Phonetic alphabet e Telephone communication systems l TOOLS AND EOUIPMENT None EVALUATION METIIOD Perform in-plant ERFORMANCE STEPS
- 1. Select applicable RWP on RWP review computer.
SAT [] UNSAT [] NOTE
- 2. Review RWP.
SAT [] UNSAT [ ] NOTE
- 3. Obtain DAD from shelf and ensure it reads " PAUSE".
SAT [ ] UNSAT [ ] NOTE O
- 4. Insert DAD into reader.
SAT [ ] UNSAT [] NOTE i tO 5. Enter TLD/ badge number. , SAT [ ] UNSAT [ ] NOTE
- 6. Enter RWP number, i
~ i SAT [ ] UNSAT [] NOTE ) , 1 l
- 7. Scan TLD.
i SAT [ ] UNSAT [ ] NOTE l f
- 8. Acknowledge review of RWP.
- SAT [ ] UNSAT [ ] NOTE i
Q9. V Review personal and RWP information. SAT [ ] UNSAT [ ] NOTE
- 10. Remove DAD from reader and ensure it is reading "0.000 rem".
SAT [] UNSAT [] NOTE 1
> > > > > END OF EVALUATION < < < < <
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4 4 l .I f W f 4 i i i d 4 EPIP QUESTIONS i 4 k n f a f l E i i.
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1 .. l - ES-301 Administrative Topics Outline - SET-1 l Form ES-301-1 p 0
- TOPIC
- Emergency Facilities
, QUESTION #1
~ j " ;,l';
- ,f<'/7 QU$5FION'f' c 's
, >f Nk'JE'A #$h2M'7 ',l ,J /:l C'If '[$$')
i i You are escorting a mechanical chiller The " visitor" must be taken to security
! tech. rep. in the turbine building when an building for accountability.
5 Alert is declared. What should you do
- with the tech. rep.? REF. Emergency Plan, page 6.8 i
! QUESTION #2 i I gy sw = c m,, w y ,,.s.
"TQUESTIONr
- 5.9 1 ma gi', $M < f4 ' i. ._ ,.,,, .,4NSWER> , @M. ,,.R J[S/U.a, ,.;,M.
Which emergency facilities were designed The following emergency facilities were to provide some form of protection from designed to provide some form of protection whole body radiation and airborne from whole body radiation and airborne contamination? contamination: O Main Control Room (MCR), Technical Support Center (TSC), and the Iocal Emergency Response Facility (LEOF). NOTE: Candidate may mention the Alternate Operation Support Center (OSC), which shares the MCR pressure envelop. REF. Emergency Plan, pages 7.3, 7.4, & 7.5. O
I Virginia Power i North Anna Power Station
- REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS l
i Job Performance Measure R..476 INITIAL CONDITIONS Unit was at 100% steady-state operation prior to the event Control bank A control rod P-10 is at 0 steps, as indicated by individual rod position 4 1-AP-1.2, " Dropped Rod," has been signed off up to the point of determining the l maximum withdrawal rate, and it has been determined that no rate applies. INITIATING CUE You are requested to complete the " Dropped Rod Retrieval" attachment 2 in 1-AP-1.2. 1 4 i e
r . Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure R..476 Candidate Evaluator ! Evaluation Date Performance Evaluation Satisfactory Unsatisfactory l TASK Retrieve a dropped rod (1-AP-1.2). NOTE TO TIIE TRAINER AND TIIE EVALUATOR You must supply key variable information needed for training or evaluating this task e BEFORE the session, fill in (or check the blank following the correct information in) all ILRACKETED blanks [ ] e DURINO the session, fill in (or check the blank following the correct information in) all NON-BRACKETED blanks Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure. INITIAL CONDITIONS Unit was at 100% steady-state operation prior to the event Control bank A control rod P-10 is at 0 steps, as indicated by individual rod position 1-AP-1.2, "Dropp.d Rod," has been signed off up to the point of determining the maximum withdrawal rate, and it has been determined that no rate applies.
j INITIATING CUE You are requested to complete the " Dropped Rod Retrieval" attachment 2 in 1-AP-1.2. l l STANDARDS 1 Task was performed as directed by the procedure referenced in the task statement withi.1 ; parentheses (one of the 'Jnderlined procedures if several are cited) ' f Self-checking practices were used througnout task performance i Verbal communication related to any of the following modes was conducted in accordance with VPAP-1407 l
- Emergency communication
- Face-to-face communication
- Givi.,g and acknowledging orders
- Phonetic alphabet l
- Telephone communication systems TOOLS AND EOUIPMENT 1
Copy of 1-AP-1.2 signed off to the point of completing the " Dropped Rod Retrieval"
- attachment 3
PREFERRED EVALUATION NIETIIOD Simulator
- PERFORNIANCE STEPS
- 1. Place the control rod bank selector switch in BANK SELECT.
CRITICAL STANDARDS Candidate places rod control selector switch in the CONTROL BANK A position t SAT [] UNSAT [ ] NOTE _ i 4
- 2. Record the affected bank's group step counter reading.
VERBAL-VISUAL CUES
- Control bank A group 2 step counter reading is 225.
4 STANDARD i Candidate records 225 for control bank A group 2 step counter reading. SAT [] UNSAT [ ] NOTE i ] 3. Manually reset the group step counter. 1
- CRITICAL STANDARDS
- Candidate resets thumbwheels for control bank A group 2 step counters to zero.
) SAT [ ] UNSAT [ ] NOTE 4 f
- 5. Record the affected bank pulse-to-analog converter reading.
! STANDARD
- . Candidate requests an extra operator to obtain the pulse-to-analog converter
!, reading for control bank A. 4 DEMONSTRATION CUES i Control bank A pulse-to-analog converter reading is 225.
- STANDARD i
Candidate records 225 for Control bank A pulse-to-analog converter reading. ?i
- SAT [] UNSAT[] NOTE I 4
i i 5 i
I a 1 . 6. Request an extra operator to reset the pulse-to-analog converter for control bank A. l STANDARD i Candidate requests an extra operator to reset the pulse-to-analog ;onverter for 4 control bank A.
- SAT [ ] UNSAT [ ] NOTE l
! 7. Open all lift coil disconnect switches for the affected bank, except the switch for the dropped rod, i CRITICAL STANDARDS Candidate opens all lift coil disconnect switches for control bank A except for rod P-10. SAT [ ] UNSAT [ ] NOTE
- 8. independently verify that all lift coil disconnect switches for the affected bank, except the switch for the dropped rod, are open.
DEMONSTRATION CUES Assume that another operator has performed this step. SAT [] UNSAT [ ] NOTE
- 9. Manually withdraw the affected control rod.
CRITICAL STANDARDS Candidate commences withdrawaling control rod P-10. DEMONSTRATION CUES Reactor Coolant System temperature control will be accomplished by the balance-of-plant operator. SAT [ ] UNSAT [ ] NOTE
, n
- 10. Continue to withdraw control rod P-10 to 225 steps.
CRITICAL STANDARDS 4 Candidate withdrawals control rod P-10 to 225 steps. STANDARD Candidate records 225 steps. 4 SAT [ ] UNSAT [ ] NOTE i l 11. Verify that all rods in the affected bank are at the same height and that no rod bottom j light is lit. 3
- STANDARD 1
Candidate verifies all rods in the affected bank are at the same height and that no rod bottom light is lit. 1 SAT [] UNSAT [ ] NOTE
- 12. Close all lift coil disconnect switches, i
CRITICAL STANDARDS ! Candidate closes all lift coil disconnect switches for control bank A. i SAT [ ] UNSAT [ ] NOTE
- 13. Reset the ROD CONTROL URGENT FAILURE alarm.
i
- STANDARD Candidate resets the ROD CONTROL URGENT FAILURE alarm.
i SAT [ ] UNSAT [ ] NOTE 4 4 d 9
- 14. Step the affected bank control rods in one step, and verify that group 2, and then group-1, are sequencing properly.
9EMONSTRATION CUES Assume another operator will coraplete this procedure. SAT [ ] UNSAT [ ] NOTE
> > > > > END OF EVALUATION < < < < <
l f
JPM QUESTION SilEET R..476 TOPIC: 1-001 SYSTEM: Rod Control K/A: 003AA2.03 (3.6/3.8) QUESTION #1 (QUESTION 7ASUVER7 ISli
-U1 Explain how each of the following plant
- Indicated reactor power will decrease by parameters change in response to a dropped varying amounts as indicated by the PRNIS control rod that does not cause a reactor trip due to rod shadowing and the relationship while the unit is at power. between the dropped rod and detector location.
o Indicated reactor power o Actual reactor power
- Actual reactor power should remain constant o Reactor Coolant System AT/T,,, if steam demand does not change unless the o Pressurizer level and pressure corresponding decrease in T,, causes steam o Axial flux difference pressure to decrease low enough to reduce steam demand.
- Reactor Coolant System AT will remain constant if actual power does not change. If power does decrease (for the reasons noted above), AT will correspondingly decrease.
e T,,, will decrease to provide positive reactivity (in response to the negative reactivity inserted by the dropped rod).
- PRZR level will decrease in response to the reduced T ,, due to " shrinkage" of the RCS as
, well as the decrease in the programir.ed pressurizer level that corresponds to the I reduced T.,,. e Pressurizer pressure will decrease in response to the reduced T,,, until the heaters can restore the system to normal operating pressure.
- Axial Hux will shift upward due to the greater amount of positive reactivity inserted in the top of the core due to the T.,, decrease (greater fractional change in density at the top of the core than at the bottom of the core).
REF. Vision Objective #: 11025
TOPIC: 1-001 SYSTEM: Rod Control K/A: 1001K4.03 (3.5/3.8) I QUESTION #2 l LQUESTION? VANSWERG LS/UL Given the following 40 steps per minute (SPM) conditions: e The unit is holding at 30% [72 SPM (max. speed) - 8 SPM (min speed) = power for chemistry. 2* mis-match between 3 F and 5 F e Rod control is in manual. e Tave is 557 F 32 SPM/ F + 8 SPM (min speed) = 40 SPM e Tref unit fails to +4 F above Tave. REF. Vision Objective #: 6512, PLS DOCUMENT What would indicated rod speed be if rod control was placed in automatic? i
1 l 11025 Explain how each of the following plant parameters change in response to a dropped control rod that does not cause a reactor trip while the unit is at power. I l
- Indicated reactor power l Indicated reactor power will decrease by varying amounts as indicated by the PRNIS due to rod shadowing and the relationship between the dropped rod and detector location. A noticeable tilt / imbalance may be indicated due to the proximity of the dropped rod to the detectors.
- Actual reactor power 1
Actual reactor power should remain constant if steam demand does not change unless the ' corresponding decrease in T,,, causes steam pressure to decrease low enough to reduce steam demand. (This should not be the case with minimal SG tube plugging, adequate initial SG l pressure and " room" for the governor valves to open in response to the decrease in steam pressure.) Reactor Coolant System AT/T,,, Reactor Coolant System AT will remain constant if actual power does not change. If power does decrease (for the reasons noted above), AT will correspondingly decrease. i T,,, will decrease to provide positive reactivity (in response to the negative reactivity inserted l by the dropped rod) to maintain the reactor critical. l l
- Pressurizer level and pressure Pressurizer level will decrease in response to the reduced T,,, due to " shrinkage" of the RCS as well as the decrease in the programmed pressurizer level that corresponds to the reduced T,,.
Pressurizer pressure will decrease in response to the reduced T,,, until the hea.ers can restore the system to normal operating pressure.
- Axial flux difference Axial Gux will shift upward due to the greater amount of positive reactivity inserted in the top of the core due to the T,., decrease (greater fractional change in density at the top of the core than at the bottom of the core).
~
6512 Explain the following concepts associated with the output signal generated by the Rod Control System's automatic temperature control unit. Differential temperature at which rod motion is initiated at minimum speed 1.5 *F I Differential temperature range at which rod speed is linearly increased to its maximum value 3.0 to 5.0 F Differential temperature range known as the " control rod lockup region"
-1.0 to -1.5 F and + 1.0 to + 1.5*F Why the lockup region is provided A lockup region is included in the program to prevent bistable chattering, when the error signal is near the end of the deadband.
l Differential temperature range known as the " control rod deadband region" 3.0 F (-1.5 to + 1.5*F) Why the deadband region is provided I l A deadband is provided in the rod control program to prevent continuous stepping of the control rods. I
s Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure R..333 INITIAI, CONDITIONS Maintenance on the Normal Letdown System is required Unit 1 is at 100% power INITIATING CUE i You are requested to shift from normal to excess letdown flowing to the volume control i tank. l 1 i i 1
- . . _ _ . . . - . - _ _ .- . .~.-- -. ___ _ . - - Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure R. 333 Candidate Evaluator Evaluation Date - Performance Evaluation Satisfaaory Unsatisfactory TASK Shift from normal letdown to excess letdown (1-OP-8,5). NOTE TO TIIE TRAINER AND THE EVALUATOR 1 Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure. 1 l l INITIAL CONDITIONS Maintenance on the Normal Letdown System is required i Unit is in mode 1 INITIATING CUE You are requested to shift from normal to excess letdown flowing to the volume control tank. 1 l l j
SJ'ANDARDS Task was performed as directed by the procedure referenced in the task statement within l parentheses (one of the underlined procedures if several are cited) Self-checking practices were used throughout task performance Verbal communication related to any of the following modes was conducted in 1 accordance with VPAP-1407 e Emergency communication e Face-to-face communication e Giving and acknowledging orders e Phonetic alphabet e Telephone communication systems I TOOLS AND EOUIPMENT None PREFERRED EVALUATION METTIOD Simulator PERFORMANCE STEPS
- 1. Review initial conditions, precautions, and limitations.
STANDARD Candidate reviews initial conditions, precautions, and limitations. SAT () UNSAT [ ] NOTE l 1 1 l I
s d i 2. Verify that component cooling water is being supplied to the excess letdown heat exchanger. J STANDARD l Candidate observes that annunciator G-E2, "EXC LTDN HX CC OUT LO FLOW" is not lit. ; SAT [] UNSAT [ ] NOTE i
- 3. Close excess letdown pressure control valve 1-CH-HCV-1137.
- STANDARD 1
Candidate closes excess letdown pressure control valve 1-CH-HCV-1137. 4 + . SAT [] UNSAT [ ] NOTE i j 1
- 4. Close the breaker fer the loop drain header isolation valves.
t i CRITICAL STANDARDS Candidate requests that breaker 1-EP-CB-26B-22 for loop drain header isolation valves 1-CH-HCV-1557A,1557B, and 1557C be closed. SAT [] UNSAT [ ] NOTE
- 5. Place the selector switch for excess letdown flow divert valve 1-CH-HCV-1389 in VCT. f i '
s ! STANDARD Candidate verifies selector switch for excess letdown flow divert valve j 1-CH-HCV-1389 is in VCT. SAT [] UNSAT [ ] NOTE l i i 4 l 4 2
- 6. Close the letdown orifice isolation valves.
CRITICAL STANDARDS Candidate closes letdown orifice valves 1-CH-HCV-1200A,1200B, and 1200C. SAT [ ] UNSAT [ ] NOTE
- 7. Place normal charging flow control valve 1-CH-FCV-1122 in MANUAL, and close the valve.
CRITICAL STANDARDS ! Candidate place normal charging flow control valve 1-CH-FCV-1122 in l MANUAL, and closes the valve. SAT [] UNSAT [ ] NOTE I 8. Open at least one Reactor Coolant System drain valve. I CRITICAL STANDARDS Candidate opens at least one Reactor Coolant System drain valve (1-CH-HCV-1557A,1557B, or 1557C). j l SAT [] UNSAT [ ] NOTE
- 9. Open the excess letdown heat exchanger isolation valve.
CRITICAL STANDARDS Candidate opens excess letdown heat exchanger isolation valve 1-CH-HCV-1201. SAT [] UNSAT [ ] NOTE l 1 l l
- 10. Open the excess letdown pressure control valve.
CRITICAL STANDARDS Candidate opens excess letdown pressure control valve 1-CH-HCV-1137. SAT [ ] UNSAT [ ] NOTE
- 11. Maintain the pressurizer level stable.
SAT [] UNSAT [ ] NOTE
> > > > > END OF EVALUATION < < < < <
4 I i 1 1 l i
JPM QUESTION SIIEET R. 333 4 TOPIC: 11-004 SYSTEM: Chemical and Volume Control K/A: GEN-2.1.25 (2.8/3.1) QUESTION #1 1
- QUESTION iANSWER S/U'-
1 Given the following plant Use 1-SC-2.1 to determine that boric acid flow should be 4
! conditions: gpm to achieve a blended flow of 500 ppm.
o RCS boron concentration is 500 (Note: the following equation can ONLY be used if the . ppm. range of indication starts at ZERO) o Desired PG flow rate wi.1 be 100 gpm. 4 oBAST concentration is 12,950 PU* ' Potentiorneter Setting = Desired Row hte gn g Total Rangeof Indication,
- Determine the potentiometer
- settings to achieve a blended flow equal to that of RCS boron
] concentration. 6.66 = ' 100gpin' 10 2.0 = '4 ###" 10 r 150gpm, r 20 gpm, i
- REF. Vision Objective #
- 3966, STATION CURVE BOOK J
4 1 ) i l 4 4
i TOPIC: 11004 SYSTEM: Chemical and Volume Control K/A: GEN-2.1.32 (3.4/3.8) QUESTION #2 1
' QUESTIONK LANSWER 2 iS/U1 ,
l Explain the consequences of the following
- This could present an unnecessary challenge l actions when placing excess letdown in to the excess letdown heat exchanger relief I service. valve (150 psig setpoint). Excessive pressure in !
the excess letdown line (when aligned to the
- Exceeding 130 psig excess letdown l VCT) exerts a higher than normal back-pressure pressure on the #1 RCP seals. I
- Exceeding 75 psig volume control
- This would present an unnecessary challenge :
tank pressure to the VCT relief valve (75 psig relieves to l HLLWT's) if the VCT were to be overfilled since there is no divert capability via LCV-1115A while on excess letdown. REF. Vision Objective #: 245,11715-FM-95 SERIES
j . [ 3966 Given the desired concentration of a blended makeup, determine the correct potentiometer settings for I 1 j the boric acid flow controller and primary grade water flow controller, j . Determining the potentiometer settings to achieve a given boron concentration. l h f
- l l
l I i i j I k i i j ! I l i a 1 i a I
1
. . n 245 Explain the consequences of the following actions when placing excess letdown in service.
Exceeding 130 psig excess letdown pressure This could present an unnecessary challenge to the excess letdown heat exchanger relief valve (150 psig setpoint), i Excessive pressure in the excess letdown line (when aligned to the VCT) exerts a higher than i normal back-pressure on the #1 RCP seals. Exceeding 75 psig volume control tank pres.sure This would present an unnecessary challenge to the VCT reli-f valve (75 psig relieves to j HLLWT's) if the VCT were to be overfilled since there is no divut capability via LCV-1115A while on excess letdown. e Rapidly placing excess letdown in service This would cause undue pressure / temperature stress on the excess letdown heat exchanger and l could cause tube leakage, i l l /
k Virginia Power } North Anna Power Station i REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure 2 R. 736 s 1
- INITIAL CONDITIONS 1-E-1, " Loss of Reactor or Secondary Coolant," has directed the transition to 1-ES-1.5, i
" Transfer From Hot Leg Recirculation to Cold Leg Recirculation" i
l Safety Injection System is in the hot-leg recirculation mode i 1-CH-P-1B was lined up to flow through the boron injection tank header for hot-leg recirculation, but has tripped on over-current and cannot be restarted. 4 1-CH-P-IC was tagged out at the start of the accident for coupling replacement and will not be available for 18 hours. ] l-CH-P-1 A is flowing through the alternate header for hot-leg recirculation j INITIATING CUE You are requested to transfer from hot-leg recirculation to cold-leg recircu'.ation in accordance with 1-ES-1.5, " Transfer From Hot Leg Recirculation to Cold Leg Recirculation."
-. _ _= - = _ _
Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure R..736 l Candidate Evaluator i Evaluation Date > Performance Evaluation Satisfactory Unsatisfactory TASK Transfer the Safety injection System from hot-leg to cold-leg recirculation (1-ES-1.5). ALTERNATE - PATII TOPIC 1-CII-P-1 A is the only available charging pump. NOTE TO TIIE TRAINER AND THE EVALUATOR Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure. INITIAL CONDITIONS 1-E-1, " Loss of Reactor or Secondary Coolant," has directed the transition to 1-ES-1.5,
" Transfer From Hot Leg Recirculation to Cold Leg Recirculation" Safety Injection System is in the hot-leg recirculation mode 1-CII-P-1B was lined up to flow through the boron injection tank header for hot-leg recirculation, but has tripped on over-current and cannot be restarted.
1-CII-P-lC was tagged out at the start of the accident for coupling replacement and will not be available for 18 hours.
1-CII-P-1A is flowing through the alternate header for hot-1eg reci rculation 6 4 s 4
INITIATING CUE You are requested to transfer from hot-leg recirculation to cold-leg recirculation in accordance with 1-ES-1.5, " Transfer From IIot Leg Recirculation to Cold Leg , Recirculation." STANDARDS i Task was perfonned as directed by the procedure referenced in the task statement within parentheses (one of the underlined procedures if several are cited) Self-checking practices were used throughout task performance Verbal communication related to any of the following modes was conducted in accordacce with VPAP-1407
- Emergency communication
- Face-to-face communication e Giving and acknowledging orders
- Phonetic alphabet
- Telephone communication systems TOOLS AND EOUIPMENT None PREFERRED EVALUATION METHOD Simulator PERFORMANCE STEPS
- 1. Close the following low-head safety injection pump hot-leg injection valves.
- 1-SI-MOV-1890A
- 1-SI-MOV-1890B CRITICAL STANDARDS !
i l Candidate p' laces key switches for 1-SI-MOV-1890A and 1-SI-MOV-1890B in j CLOSE. SAT [] UNSAT[] NOTE j l i
J ) l
- 2. Open the low-head safety injection pump discharge valves. ;
- CRITICAL STANDARDS '
, Candidate - depresses OPEN push-button for 1-SI-MOV-1864A or 4 1-SI-MOV-1864B. I, SAT [] UNSAT [ ] NOTE l 3. Open the low-head safety injection pump cold leg injection valves.
- CRITICAL STANDARDS i
.! Candidate places control switch for 1-SI-MOV-1890C or 1890D in OPEN. I i SAT [ ] UNSAT [ ] NOTE
- 4. Verify two charging pumps in service.
STANDARD Candidate determines that only one charging pump is available. SAT [ ] UNSAT[] NOTE
- 5. Establish charging pump recire for the running pump.
CRITICAL STANDARDS Candidate places control switch for 1-CH-MOV-1275A in OPEN. SAT [ ] UNSAT [ ] NOTE
4
- 6. Isolate hot-leg injection.
I I CRITICAL STANDARDS 4 Candidate depresses CLOSE push-buttons for 1-SI-MOV-1869B and 1869A. SAT [] UNSAT[] NOTE I j ] 7. Establish cold-leg injection flow path. 4
; CRITICAL STANDARDS i
j Candidate depresses OPEN push-buttons fo.1-SI-MOV-1867C and/or 1867D, I then depresses OPEN push-buttons for 1-SI-MOV-1867A and/or 1867B. SAT [] UNSAT [ ] NOTE ?
- 8. Verify that normal header cold-leg injection flow exists.
STANDARD l Candidate determines that cold-leg injection flow does not exist. 'Y SAT [] UNSAT [ ] NOTE l
- 9. Align 1-CII-P-1 A to the normal header.
- CRITICAL STANDARDS Candidate places control switch for 1-CH-MOV-1286A in OPEN.
SAT [] UNSAT [ ] NOTE l
- 10. Clo. ., ..arging pump recirc valve for 1-CH-P-1 A.
CRITICAL STANDARDS {. Candidate places control switch for 1-CH-MOV-1275A in CLOSE 4 l' SAT [ ] UNSAT [ ] NOTE i > > > > > END OF EVALUATION < < < < < W
JPM QUESTION SHEET R. 736 TOPIC: 111-006 SYSTEM: Emergency Core Cooling K/A: 011EK3.13 (3.8/4.2) QUESTION #1
- QOESTIONO ?ANSW5RI 55/UI Explain the reason Safety Injection Based on the NAPS ECCS design, the System is transferred from the hot-leg SI flow path should be alternated from recirculation mode back to the cold-leg the hot legs to the cold legs. Th!s recirculation mode in 1-ES-1.5, change in flow path is needed to
" Transfer From Hot Leg Recirculation address a concern relating to boron to Cold Leg Recirculation." precipitation in the reactor vessel core region. If a relatively large pipe break or leakage occurs in the RCS hot leg and SI is injected into the RCS hot legs, then the boron concentration in the reactor core region may increase and approach the precipitation limit. In order to alleviate this situation, the ir.jection flow path should be changed so the low-head SI pump discharge flows back to the cold legs and the charging pump primary flow path is via the bit with the alternate pump and flow path to the cold leg injection path.
REF. Vision Objective #: 13438
_ _- .~ _ . _ _ _ - _ - . _ _ _ . _ _ . - - - . _ . _ _ _ . _ . . _ _ _ _ - . _ t TOPIC: 111-006 SYSTEM: Emergency Core Cooling K/A: 006A2.13 (3.9/4.2) QUESTION #2 QUESTION:? TANSWER/ S/U[ Assuming that a spurious safety Reseting a Safety Injection signal using injecdon occurs with the unit at full the main benchboard's reset power, explain the potential long-term pushbuttons DOES NOT RESET the consequences of failing to reset the safety injection input signal for l recirculation mode signal using the SI automatic swapover of the LHSI pump I RECIRC MODE RESET push-buttons. suction to the containment sump. If this signal is not reset, it could result in the LHSI pumps shifting into the ! Recire Mode automatically when RWST level decreases below the 10-10 level setpoint, such as during a refueling, and may destroy the LHSI Pumps. To prevent this from occurring, the i operator MUST depress the SI RECIRC MODE RESET pushbuttons l after the safety injection signal has been reset. This action is required to reset the safety injection input required to actuate this interlock circuit. REF. Vision Objective #: 3432 1
_ . _ . _ . ~ . - . _ _ . . _ . _ _ _ _ . _ . . _ _ . _ . _ . _ - _ _ _ . . _ _ . . _ _ _ . SIMULATOR SETUP TASK R..736 Transfer the Safety Injection System from hot-leg to cold-leg recirculation (1-ES-1.5). CIIECKLIST Recall IC #1 (100% power) Tag out 1-Ch-P-1C - non-rotaional Enter malfunction MRC0101, time delay = 10, ramp = 60, start = 0, stop = 100 Go to RUN, and perform 1-E-0 to 1-E-1 When the Safety Injection System swaps to cold-leg recirculation, then perform the steps up through establishing redundant cold-leg injection flow paths (1-CH-P-1B flowing the boron injection tank and 1-CH-P-1A flowing the alternate header) Swap to hot-leg recirculation in accordance with 1-ES-1.4 Enter the following malfunctions: MCH1602, TD = 0 SEC, TRGR = N/A Place the simulator in FREEZE Place the keys in SI-MOV-1890A,1890B,1869A, and 1869B
i 13438 Explain why the Safety Injection System is transferred from the hot-leg recirculation mode back to the cold-leg recirculation mode in 1-ES-1.5, ' Transfer From Hot Leg Recirculation to Cold Leg Recirculation." " Based on the NAPS ECCS design, the SI Dow path should be alternated from the hot legs to the cold legs. This change in flow path is needed to address a concern relating to boron precipitation in the reactor vessel core region. If a relatively large pipe break or leakage occurs in the RCS hot leg and SI is injected into the RCS hot legs, then the boron concentration in the reactor core region may increase and approach the precipitation limit. In order to alleviate this situation, the injection flow path should be changed so the low-head SI pump discharge flows back to the cold legs and the charging pump primary flow path is via the bit with the alternate pump and flow path to the cold 4 leg injection path, e 3 J
. 1 R
3432 Assuming that a spurious safety injection occurs with the unit at full power, explain the potential long-term consequences of failing to reset the recirculation mode signal using ) the SI RECIRC MODE RESET push-buttons. 4
' Reseting a Safety Injection signal using the main benchboard's reset pushbuttons DOES NOT RESET the safety injection input signal for automatic swapover of the LIISI pump l
) 1 suction to the containment sump. l l If this signal is not reset, it could result in the LHSI pumps shifting into the Recire Mode ! } automatically when RWST level decreases below the 10-10 level setpoint, such as during i a refueling, and may destroy the LHSI Pumps. j To prevent this from occurring, the operator MUST depress the SI RECIRC MODE RESET pushbuttons after the safety injection signal has been reset. j This action is required to reset the safety injection input required to actuate this 4 interlock circuit. l l ) i j 4 1 + t i. d ) i i .h i l 1 T
.o 3
i A l Refueling water storage tank level is currently 81 %, and a large break loss of coolant accident i is in progress. Calculate how long it will be before manual swapover to cold leg recirculation
- is required. Assume the following pu nps are running at the specified flow rates
i "A" and "B" low head safety injection pumps (2500 gpm each)
"B" and "C" high head safety injection pumps (200 gpm each)
\ l -
"A" and "B" quench spray pumps (1900 gpm each) i Disregard the volume of the chemical addition tank.
A 1.00 30 minutes A 32 minutes-A 60 minutes A 63 minutes AAAAAAA Vision objective number-3381 1-SC-5.12 i
Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS JOB PERFORMANCE MEASURE R..514 INITIAL CONDITIONS Unit-1 is in mode 4. It is day 5 of a 30-day refueling outage 1-RH-P-1B is in service with 2 RHR heat exchangers in service INITIATING CUE You are the OATC on watch, i i
. e i
Virginia Power i North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS . JOB PERFORMANCE MEASURE R..514 Operator Evaluator l Observer Evaluation Date ] 1 Performance Evaluation [ 1 Satisfactory [ l Unsatisfactory TASK i l l Restore residual heat removal flow (1-AP-11). I j ALTERNATE - PATH TOPIC Running RHR pump has a sheared shaft. NOTE TO THE TRAINER AND THE EVALUATOR l 1 ', Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure. INITIAL CONDITIONS Unit-1 is in mode 4. It is day 5 of a 30-day refueling outage 1-RH-P-1B is in service with 2 RHR heat exchangers in service INITIATING CUE You are the OATC on watch.
STANDARDS Task was performed as directed by the procedure referenced in the task statement within parentheses (one of the underlined procedures if several are cited) Self-checking practices were used throughout task performance Verbal communication related to any of the following modes was conducted in accordance with VPAP-1407
- Emergency communication
- Face-to-face communication
- Giving and acknowledging orders
- Phonetic alphabet
- Telephone communication systems Residual Heat Removal System flow is restored to normal TOOLS AND EOUIPMENT None EVALUATION METHOD Simulator PERFORMANCE STEPS
- 1. Check if Reactor Coolant System level is decreasing.
STANDARD Candidate determines that a loss of inventory is not in progress. SAT [ ] UNSAT [ ] NOTE
- 2. Verify that the Residual Heat Removal System inlet isolation valves are open.
STANDARD Candidate determines that 1-RH-MOV-1700 and 1701 are open. SAT [] UNSAT [ ] NOTE
1
- I l
i 3
- 3. Verify that the Residual Heat Removal System outlet isolation valves are open.
J STANDARD Candidate detennines that 1-RH-MOV-1720A and 1720B are open. SAT [ ] UNSAT [ ] NOTE l
- 4. Check that at least one residual heat removal pump is running.
i STANDARD , Candidate determines that 1-RH-P-1B has a sheared shaft and is not to be considered as running. j SAT [ ] UNSAT [ ] NOTE
- 5. Manually close 1-RH-FCV-1605 and 1-RH-HCV-1758.
STANDARD I Candidate closes 1-RH-FCV-1605 and 1-RH-HCV-1758. SAT [] UNSAT [ ] NOTE
- 6. Start one RHR pump.
CRITICAL STANDARD Candidate starts 1-RH-P-1 A. SAT [ ] UNSAT [ ] NOTE
__ _ .__ _ _ ___.. .._ _ __. _ . - _ . . _ _ _ _ . m . _ _ . _ _ . _- . . _._ . l
- 7. Verify RHR system parameters normal.
4 STANDARD 1 i 1 Candidate determines that RHR system is normal. l j SAT [ ] UNSAT[] NOTE i 1 1' l
- 8. Check service water to CC heat exchangers delta-P normal.
), a 1 DEMONSTRATION CUES t i The auxiliary building watchstander reports that SW to CC heat exchanger delta P is 12 PSID. i STANDARD 1 1 1 ! Candidate detennines that the delta-p is normal. SAT [ ] UNSAT[] NOTE I ;
- 9. Check CC flow to RHR heat exchangers normal.
I STANDARD ' . Candidate determines that CC flow to RHR heat exchangers is normal. SAT [ ] UNSAT [ ] NOTE i,
- 10. Return to the procedure in effect.
i SAT [ ] UNSAT [ ] NOTE :
> > > > > END OF EVALUATION < < < < <
1 1 1 J
l l JPM QUESTION SHEET R..514 TOPIC: IV-005 SYSTEM: Residual Heat Removal K/A: GEN-2.1.20 (4.3/4.2) QUESTION #1 h fDUESTIONE V1NSWEM TS/U? Assume the following conditions existed during Since RCS level is +7 inches above the performance of the just completed JPM: centerline, current RHR tiow is in the o Unit 1 is in moc.e 5. " unacceptable region". The following actions l 0 It is day 5 of a 30-day refueling outage. should be taken: o RIIR pump amps are observed to be oscillating. 1. Continue RCS makeup. o RCS level is decreasing rapidly. 2. Stop RHR pumps. o The operating crew entered 1-AP-11 3. Go to step 11 o RCS inventory has been stabilized at +7 inches above centerline. REF.1-AP-11 o RHR flow has been reduced to equal " design flow" for time after shutdov,n. What are the next required actions that should be taken?
__ _ _ _ _ . __._..~.__ _ _ _ _ _ _ . . . _ _ _ . . _ . _ _ _ _ _ . . _ . . . . _ - . - _ . . _ _ . _ _ _ , _ - _ - _ - i 1 1 l TOPIC: IV-005 SYSTEM: Residual Heat Removal K/A: 005A2.04 (2.9/2.9) i ! QUESTION #2 I fQUESTION 'ANSWERL 'S/U Given the following conditions: RCS temperature will decrease as 1-RII-HCV-1758 fails open and 1-RH-FCV-i
- Unit 1 is in mode 5 for a refueling outage. 1605 fails closed.
- A VCT float has been established.
- Core off-load has not yet begun. REF.11715-FM-94
)
- An instrument air leak has developed inside containment.
- The contaiment instrument air TV has I been closed and outside instrument air j pressure is recovering.
l As containment instrument air decreases to j zero, what will happen to RCS temperature and why?
)
Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure l R..164 INITIAL CONDITIONS 1 Unit startup from mode 4 to 3 is in progress "A" and "C" RCPs are in service Conditions have been established for starting the "B" RCP ; The reactor coolant filter and a mixed bed ion exchanger are in service i INITIATING CUE You are requested to start the "B" RCP.
.-_ . . . - . .= ._ . - . - . _ - . . - . . . . - ..
i Virginia Power 4 North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS 4 Jeb Performance Measure i i R.. 164 1 Candidate Evai ator Evaluation Date Performance Evaluation Satisfactory Unsatisfactory TASK Start a reactor coolant pump. NOTE TO TIIE TRAINER AND THE EVALUATOR Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure. INITIAL CONDITIONS Unit startup from mode 4 to 3 is in progress "A" and "C" RCPs are in service Conditions have been established for starting the "B" RCP The reactor coolant filter and a mixed bed ion exchanger are in service INITIATING CUE You are requested to start the "B" RCP.
i I 4 l STANDARDS 1 Task was performed as directed by the procedure referenced in the task statement within j parentheses (one of the underlined procedures if several are cited) Self-checking practices were used throughout task performance Verbal communication related to any of the following modes was conducted in accordance with VPAP-1407 e Emergency communication e Face-to-face communication e Giving and acknowledging orders e Phonetic alphabet e Telephone communication systems l Failed channel is placed in TEST within one hour l TOOLS AND EOUIPMENT i None PREFERRED EVALUATION METIIOD Simulator PERFORMANCE STEPS
- 1. Review initial conditions, precautions, and limitations.
STANDARD Candidate reviews initial conditions, precautions, and limitations. SAT [] UNSAT [ ] NOTE
- 2. Ensure seal return MOVs are open.
STANDARD Candidate verifies 1-CII-MOV-1380 and 1381 are open. SAT [] UNSAT[] NOTE
. _ _ . . . _ _ _ _ _ _ . _ . . . _ _ . _ . . _ _ _ _ _ _ _ _ _ _ . . _ . _ _ . _ . _ _ _ . ~ _ _ _ _ . _ _ _ _ _ . . _ _ _ ,
1 1 1 4 . l 3. Verify "B" #1 seal leak off flow is within acceptable limits. 1 j STANDARD i [ Candidate verifies acceptable flow IAW attachment 1. l j SAT [ ] UNSAT [ ] NOTE ( 4. Verify "B" #1 seal delta-P is greater than 200 PSID. i i STANDARD Candidate verifies delta-P is >200 PSID. 4 { SAT [ ] UNSAT [ ] NOTE i
- 5. Start the "B" RCP bearing lift pump.
I CRITICAL STANDARDS i ) Candidate places control switch for 1-RC-P-1B1 in START. J l 1 SAT [ ] UNSAT [ ] NOTE i j 6. Verify "B" RCP oil pressure start permissive indicating light is on. 1 STANDARD i
- j. Candidate observes that light is lit.
i SAT [ ] UNSAT[] NOTE
- 7. Verify "B" RCP annunciators are not lit.
i STANDARD 1 } Candidate observes that annunciators are not lit. i i SAT [ ] UNSAT [ ] NOTE ]
i
- 8. Ensure CVCS parameters are within spec.
j STANDAP.D Candidate determines that CVCS parameters.are within spec. e SAT [ ] UNSAT [ ] NOTE
- 9. Ensure CC parameters are within spec.
1 , STANDARD i Candidate determines that CC parameters are within spec. !! SAT [ ] UNSAT [ ] NOTE 1
- 10. Ensure RCS pressure is above required value.
STANDARD Candidate verifies RCS pressure is > 280 PSIG. SAT [ ] UNSAT [ ] NOTE
- 11. Verify that bearing lift pump has been running at least 2 minutes.
CRITICAL STANDARDS Pump has been running at least 2 minutes. SAT [] UNSAT [ ] NOTE
- 12. Start the "B" RCP.
CRITICAL STANDARDS Candidate places the "B" RCP control switch to START, then to AUTO-AFTER-START. SAT [ ] UNSAT [ ] NOTE
.- - . . . - . . . - - . . . - _ . - . - ~ . - . . . . - . . . , - - .. __ . -
- 13. Verify RCS flow is increasing.
STANDARD Candidate observes RCS flow increasing. SAT [ ] UNSAT[] NOTE NOTE TO EVALUATOR: AT THIS POINT, #1 SEAL DELTA-P WILL BEGIN TO DECREASE.
- 14. Determine that #1 seal delta-P is decreasing to/is going below 2(X) PSID.
CRITICAL STANDARDS Candidate stops "B" RCP. i SAT [ ] UNSAT [ ] NOTE i
> > > > > END OF EVALUATION < < < < <
l
q
)
i JPM QUESTION SHEET 1 R. . 57 TOPIC: VII-015 SYSTEM: Nuclear Instrumentation K/A: GEN-2.1.28 (3.2/3.3) QUESTION #1
- .QUESTIONi fANSWER3 l:S/Uj Given the following conditions: The next start attempt may be made at 1715.
- Unit 1 is in mode 3
- The "A" RCP tripped at 1500 due REF. Vision Objective #: 10499,1-to a faulty relay OP-5.2
- The relay has been replaced
- At 1615, the OATC started the "A" RCP, but inadvertently stopped '
it as it was reaching full speed -
- Another start was attempted at 1645, but it too was unsuccessful When may the next start attempt be made?
l
__ _ _ _ . _ . _ _ _ _ . _ .._._._ _ ~ . _ _ _ _ _ . . . . _ . _ . _ .. _ _ _ _ _ _ . . l TOPIC: VII-015 SYSTEM: Nuclear Instrumentation K/A: 015A1.01 (3.5/3.8) QUESTION #2
- QUESTIONE VANSWERj *S/U)
Given the following conditions: By not considering the heat added by the RCPs, actual reactor power will less than calculated. e The OATC is performing a computer calorimetric.
- The OATC inadvertently omits . . , l the heat input from the RCPs. On
- Gsta
- Gsta mondown ~ ace's ~ ru neatus l
Based on the above, what will the relationship be between actual REF. Vision Objective #: 7804 power and calculated power?
- . ... - .-. . . . . . - . . - . - . - . - . . . ~ . . - . . - . . - - - . ~ _ . ..
4
- h i
} 10499 List the following requirements associted with starting a reacer coolant pump, i I Time required for the motor ta be idle before any start attempt Allow motor to idle at lean 30 minutes before any restart. Time required for the motor to be idle between the third and fourth start attempt Allow motor to idle at least I hour before fourth attempt. Number of start attempts allowed in a two-hour period Attempt only three starts in 2 hours. Reason for limiting the number of motor starts within a given time period To prevent damage to the stator due to overheating
7804 Explain the following concepts concerning the adjustment of the power-range nuclear instruments. Qualifications required of personnel who adjust the power-range instruments The channel is adjusted by a licensed operator under the direct supervision of a senior reactor operator. Parameters used to calculate the secondary plant calorimetric Rt SIG
- SIG Blowdown ~ RCP's ~ Pzt Heaters Maximum difference between the secondary plant calorimetric and the power-range nuclear instrument reading Each power range channel at the North Anna Power Station is required to be within two percent of the calorimetric pov'er level.
4 i
i k i Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure R..216 INITIAL CONDITIONS Unit 1 tripped from 100% power due to a main generator fault While stabilizing the unit in 1-ES-0.1, a large break LOCA occurred The STA has identified an orange path on Containment INITIATING CUE You are requested to respond in accordance with 1-FR-Z.1.
Virginia Power , North Anna Power Station } REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure R..216 d Candidate Evaluator Evaluation Date i Performance Evaluation Satisfactory Unsatisfactory TASK Align the containment spray systems in esponse to high containment pressure (1-FR- ! Z.1). NOTE TO THE TRAINER AND THE EVALUATOR Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure. ) INITIAL CONDITIONS Unit 1 tripped from 100% power due to a main generator fault While stabilizing the unit in 1-ES-0.1, a large break LOCA occurred The STA has identified an orange path on Containment INITIATING CUE You are requested to respond in accordance with 1-FR-Z.1.
STANDARDS a L was performed as directed by the procedure referenced in the task statement within
, .ntheses (one of the underlined procedures if several are cited) delf-checking practices were used throughout task performance Verbal communication related to any of the following modes was conducted in accordance with VPAP-1407 e Emergency communication
- e Face-to-face communication e Giving and acknowledging orders e Phonetic alphabet e Telephone communication systems TOOLS AND EOUIPMENT None PREFERRED EVALUATION METHOD Simulator PERFORMANCE STEPS
- 1. Verify Phase A isolation valves closed.
STANDARD Candidate observes Phase A trip valves are closed. SAT [ ] UNSAT [ ] NOTE
- 2. Check if CDA is required.
STANDARD Candidate determines that containment pressure has exceeded 28 PSIA. SAT [] UNSAT [ ] NOTE __
l 4
- 3. Manually actuate CDA.
) CRITICAL STANDARDS , Candidate places both of the CDA switches to INITIATE. i SAT [ ] UNSAT [ ] NOTE .
! 4. Verify CC pumps tripped.
a CRITICAL STANDARDS i
- Candidate stops 1-CC-P-1 A.
i . 4 SAT [ ] UNSAT [ ] NOTE i j 5. Stop all RCPs. t ! CRITICAL STANDARDS 1
- Candidate stops all RCPs.
l SAT [] UNSAT [ ] NOTE i ! 6. Verify Phase B isolation valves closed. l
- CRITICAL, STANDARDS I
- Candidate closes 1-CC-TV-105A, B, C,1-CC-TV-102A, C, E,1-CC-TV-101 A, i 103A, and 1-IA-TV-102A.
4 STANDARD 5
- Candidate observes all other Phase B isolation valves are closed.
4 SAT [] UNSAT [ ] NOTE i )
- 7. Verify at least two service water pumps running.
1 4 STANDARD j Candidate observes both SW pumps running. I
SAT [] UNSAT [ ] NOTE 4
? f
- 8. Verify SW supply to CC HXs closed.
CRITICAL STANDARDS Candidate closes 1-SW-MOV-108A. SAT [ ] UNSAT [ ] NOTE _
- 9. Verify SW supply to recirc spray HXs open.
CRITICAL STANDARDS l Canilidate opens 1-SW-MOV-103A,101A,103D, and 101C. SAT [ ] UNSAT [ ] NOTE l
- 10. Verify SW return from recirc spray HXs open.
! CRITICAL STANDARDS 1 Candidate opens 1-SW-MOV-104A,105A,104D, and 105C. l SAT [] UNSAT [ ] NOTE l l
- 11. Verify proper operation of the "H" train containment spray systems. I CRITICAL STANDARDS Candidate opens 1-QS-MOV-101A andl-RS-MOV-100A.
Candidate starts 1-RS-P-1 A, 2A, 3A, and 1-QS-P-1 A. SAT [ ] UNSAT [ ] NOTE
- 12. Verify proper operation of the "J" train containment spray systems.
STANDARD Candidate observes all "J" equipment operating. SAT [ ] UNSAT [ ] NOTE 6.
-v- --7 += -
? .- aw -
--7
_ . _ _ . . _ . . . __ - . _ _ . . _ - . _ _ . _ _ . - _ . .____.__m__ . . - _ _ . 3 1-1'
- 13. Verify MSTVs and bypass valves closed.
c STANDARD ' Candidate observes valves are closed. SAT [ ] UNSAT [ ] NOTE
- 14. Check if feed flow should be isolated to any SG.
STANDARD Candidate determines that SGs are not faulted. SAT [] UNSAT [ ] NOTE
- 15. Check containment hydrogen concentration.
EVALUATOR'S CUE Assume another operator will complete this procedure. SAT [ ] UNSAT [ ] NOTE
> > > > > END OF EVALUATION < < < < <
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j . , JPM QUESTION SHEET R..765 ] TOPIC: V-028 SYSTEM: Hydrogen Analyzer K/A: 028K5.01 (3.4/3.9): 3 QUESTION #1 i
!.. QUESTION'.) iANSWERi qS/Uj l Given the following conditions: Since containment hydrogen
- concentration is 17%, any spark l e Unit I has suffered a LOCA could cause an explosion, which j
- The crew is cooling down the plant would cause a spike in containment
- using 1-ES-1.2, " POST-LOCA pressure.
4 COOLDOWN AND j DEPRESSURIZATION". REF. Vision Objective #: 5450,
- e Containment sump level is 2.5' 12654, 13008 e Containment hydrogen concentration is i 17 %.
j e They are at the point of isolating j accumulators
- Unknown to the crew, the wiring to 1-SI-MOV-1865B has been damaged.
1 j Given the above conditions, what might I l occur if a spark developed during the ' energizing of 1-SI-MOV-1865B, and what indications would the control room see? l l l l i l l I
3 1 I TOPIC: V-028 SYSTEM: Hydrogen Analyzer K/A: GEN-2.1.28 (3.2/.3.3) !
- QUESTION #2 f.QUESTIONji FANSWERT RSIUM j Why does the procedure require you This maintains sample line 1 I
to place the hydrogen analyzer heat temperature at approximately 285*F tracing in service? to prevent condensation and
- equipment malfunction damage. ;
j This heat tracing extends from i j outside Containment to the inlet to the hydrogen analyzer. REF, Vision Objective #: 5433 I f i I i 4 s b a j 4 2 i i i t i l 4 i I i j i e
SIMULATOR SETUP TASK R. 765 Place a containinent hydrogen analyzer in operation (1-OP-63.2). GIECKLIST Reca!! IC #1 (100% power) l
' 1 1 ,- l
,' 5450 List the following information associated with hydrogen in containment.
- 1 Minimum hydrogen concentration required for hydrogen to burn if ignited '
Minimum hydrogen concentration required for hydrogen to detonate if ignited i Hydrogen characteristics in dry air:
.% hydrogen in air by volume Characteristics 4% to 15% Hydrogen will burn j 15% to 60% Hydrogen will detonate / explode 1
60% to 74% Hydrogen will burn s i 4 i 4 i i i i k i e i a i k 1 i 4 1
12654 Explain why an extreme challenge to the containment barrier exists if containment pressure is greater than the design pressure (1-F-0). The extreme challenge to the containment boundary does not necessarily come from the pressure alone, but rather from the potential additional pressure spike which would accompany a hydrogen ignition inside containment. The total pressure could then potentially exceed the strength of containment. Also, above containment design pressure, leakage may exceed design basis limits. i
__ ___ _ _ . _ . - _ _ _ _ _ . . _ . _ - _ . - . . _ . _ - . _ . . . _ . . . _ . - _ _ _ _ ___...._-_.m._.-. . . . - _ . . . _ . . 7 1 1 b ) 13008 Explain the following concepts associated with 1-FR-Z.1, " Response to High Containment Pressure." 2
. Why excessive hydrogen concentration in containment is a concern with respect to containment integrity 1
' A hydrogen burn / explosion could cause a pressure spike which may challenge the containments integrity.
4 I i i i 4 ) 4 4 k r ) i 1 i l. t I i E i e ; 1 1 4 } i 2 1 )
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. =
i I i
- I l
, 5433 Explain the following concepts associated with hydrogen analyzer H2 A-IIC-101. . Why the containment sample line is heat traced This maintains sample line temperature at approximately 285*F to prevent condensation and equipment malfunction damage. This heat tracing extends from 1 outside Containment to the inlet to the hydrogen analyzer. IIow a low temperature in the heated sample compartment affects detector operability Inlet gas is heated to 275 to 300 'F by electric heaters. This elevated temperature is necessary for the operation of the analyzer cell within its designed accuracy. If the temperature drops to 250 degree F, an alarm lamp is energized on the local analyzer panel. 1
. _ _ _ . _ . - . _ . . . _ _ - _ - _ _ . ~ _ . _ . _ _ _ _ - . . _ _ _. ._ . _
. Virginia Power 4 North Anna Power Station J REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure R..577 l INITIAL CONDITIONS i Residual Heat Removal System is in service j Reactor was manually tripped due to a loss of all reactor coolant pump seal cooling i j Unit-1 "H" emergency diesel generator is tagged out for maintenance
- Shift supervisor desires to maintain "J" bus powered equipment operable i
l 1-ES-0.2A, " Natural Circulation Cooldown with CRDM Fans," has been completed to j the point of determining if systems should be aligned for NDT protection i INITIATING CUE 4 4 You are requested to align each pressurizer PORV for NDT protection in accordance with step 22 of 1-ES-0.2A. e 4 4 1
1 1 i Virginia Power I , North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS , Job Performance Measure t R..577 i l Candidate Evaluator Evaluation Date 4 i Performance Evaluation Satisfactory Unsatisfactory 1 TASK Place NDT protection in service during a natural circulation cooldown (1-ES-0.2A, 1-ES-0.2B). NOTE TO Tile TRAINER AND THE EVALUATOR ' i Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure. INITIAL CONDITIONS Residual Heat Removal System is in service Reactor was manually tripped due to a loss of all reactor coolant pump seal cooling Unit-1 "H" emergency diesel generator is tagged out for maintenance Shift supervisor desires to maintain "J" bus powered equipment operable 1-ES-0.2A, " Natural Circulation Cooldown with CRDM Fans," has been completed to the point of determining if systems should be aligned for NDT protection
1 ) ) INITIATING CUE You are requested to align each pressurizer PORV for NDT protection in accordance I ) with step 22 of 1-ES-0.2A. 1 STANDARDS i . Task was performed as directed by the procedure referenced in the task statement within l parentheses (one of the underlined procedures if several are cited) Self-checking practices were used throughout task performance
- Verbal communication related to any of the following modes was conducted in j accordance with VPAP-1407 l j
- Emergency communication
,-
- Face-to-face communication
- Giving and acknowledging orders
- Phonetic alphabet
- Telephone communication systems TOOLS AND EOUIPMENT Pressurizer power-operated relief valve NDT protection keys Copy of 1-ES-0.2A signed off to the point of determining if systems should be aligned for NDT protection PREFERRED EVALUATION METHOD Simult tor PERFORMANCE STEPS
- 1. Verify that Reactor Coolant System cold-leg temperatures will be less than the required value within one hour.
SAT [ ] UNSAT [ ] NOTE
k l
- 2. Place one low-head safety injection pump in PULL-TO-LOCK.
CRITICAL STANDARDS Candidate places control switch for 1-SI-P-1A in PULL-TO-LOCK. ! i l SAT [ ] UNSAT [ ] NOTE
- 3. Place all but one charging pump in PULL-TO-LOCK.
CRITICAL STANDARDS i ( Candidate places control switches for 1-CH-P-1A and IC in PULL-TO-LOCK. SAT [ ] UNSAT [ ] NOTE
- 4. Verify that RCS pressure is less than the required value.
STANDARD i Candidate checks RCS pressure. ' SAT [ ] UNSAT [ ] NOTE
- 5. Close pressurizer power-operated relief block valve 1-RC-MOV-1535.
CRITICAL STANDARDS Candidate closes pressurizer power-operated relief block valve 1-RC-MOV-1535. SAT [ ] UNSAT [ ] NOTE
- 6. Open pressurizer power-operated relief valve 1-RC-PCV-1456.
CRITICAL STANDARDS Candidate opens pressurizer power-operated relief valve 1-RC-PCV-1456. SAT [ ] UNSAT [ ] NOTE
4 1 1 2
- 7. Close pressurizer power-operated relief valve 1-RC-PCV-1456.
j CRITICAL STANDARDS i Candidate closes pressurizer power operated relief valve 1-RC-PCV-1456. l 1 SAT [] UNSAT { ] NOTE-1
- 8. Open pressurizer power-operatad relief block valve 1-RC-MOV-1535.
1. j CRITICAL STANDARDS i I Candidate opens pressurizer power-operated relief block valve 1-RC-MOV 1535. l SAT [] UNSAT [ ] NOTE 4 i , 9. Close pressurizer power-operated relief block valve 1-RC-MOV-1536.
- CRITICAL STANDARDS Candidate closes pressurizer power-operated relief block valve 1-RC-MOV-1536.
4 SAT [] UNSAT [ ] NOTE ', 1 i ! 10. Open pressurizer power-operated relief valve 1-RC-PCV-1455C. CRITICAL STANDARDS i Candidate opens pressurizer power-operated relief valve 1-RC-PCV-1455C. SAT [] UNSAT [ ] NOTE ! 11. Close pressurizer power-operated relief valve 1-RC-PCV-1455C. f CRITICAL STANDARDS 1 l Candidate closes pressurizer power-operated relief valve 1-RC-PCV-1455C. 4 l 4 SAT [ ] UNSAT [ ] NOTE 1 i
1 4 . . i s
- 12. Open pressurizer power-operated relief block valve 1-RC-MOV-1536.
] CRITICAL STANDARDS i i Candidate opens pressurizer power-operated relief block valve 1-RC-MOV-1536. s i SAT [ ] UNSAT [ ] NOTE [ 13. Place the pressurizer power-operated relief valve key switches in AUTO. C CRITICAL STANDARDS t
- Candidate places PRZR power-operated relief valve 1-RC-PCV-1455C and 1456 NDT protection key switches in AUTO.
i ] SAT [] UNSAT [ ] NOTE i j 14. Depressurize safety injection accumulators. VERBAL-VISUAL CUES I j Assume that another operator will complete this procedure ! SAT [] UNSAT [ ] NOTE
> > > > > END OF EVALUATION < < < < <
JPM QUESTION SHEET R..577 TOPIC: 111-010 SYSTEM: Pressurizer Pressure Control K/A: 010A4.03 (4.0/3.8) QUESTION #1 IDU8STI N$ i;[ASSWER *(S/O) Given the following conditions: Once valve travel is initiated by placing the handswitch to the
- Unit is operating at 100% power. "Open" or "Close" position, the
- 1-RC-PCV-1456 fails open. handswitch can be shifted to and
- Actions to close it IAW 1-AP-44 maintained in opposite position, but have failed. valve travel will continue until either
- In an attempt to isolate the open the " full-open" or " full-closed" PORV, the operator mistakenly takes position is reached. At that time j the control switch for 1-RC-MOV-1535 the valve will reverse direction and 1 to CLOSE. continue stroking until the valve l
- Realizing his error, he immediately position matches the handswitch l
takes the switch back to OPEN. position. Describe the valve's response to these REF. Vision bjective #: 3499 action's. l l l P
1 TOPIC: 111-010 SYSTEM: Pressurizer Pressure Control K/A: 010A1.09 (3.4/3.7) QUESTION #2
- n. . . _ . . . . . .. . ..
; QUESTION) '
fANSWERS !!S/UK; Given the following conditions: 281 F e The unit is operating at 100% power when a PRZR safety valve REF. Steam Tables fails partially open. e PRZR pressure is 1910 PSIG e PRZR vapor temperature is 630 F e RCS Tave is 550 F e PRT pressure is 35 PSIG. What is the PRZR safety valve tail-pipe temperature?
1 l 3499 Explain the purpose of the following pressurizer components.
- Power-operated relief valve block valves i
j Each PORV has a blocking valve (MOV-1535 and 1536) located upstream of it. The motor-operated valves are normally open, but they are used to isolate a i PORV which does not fully reseat after opening or is experiencing leakage. These blocking valves are controlled from the MCR. Once valve travel is initiated by placing the handswitch to the "Open" or "Close" position, the i j handswitch can be shifted to and maintained in opposite position, but valve travel will continue until either the " full-open" or " full-closed" position is reached. Al ;
- that time the valve will reverse direction and continue stroking until the valve '
! position matches the handswitch position. This is to prevent thermalling out the i j breaker with the valve in mid-position. I 1 ? i T I i = 1 s
l a 1 Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure 4 N.. 466 INITIAL CONDITIONS Control room is uniahabitable The 1H emergency diesel generator is carrying the 1H emergency i bus in the CRE mode I The 1H EDG started on a UV signal j Normal emergency bus power supply is available Breaker 15F3 is closed INITIATING CUE You are requested to unload and shut down an emergency diesel generator in the control room emergency mode. 4 4 ) 1 i [ l s i I
1
. . l Virginia Power .
i North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance Measure N.. 466 Candidate Evaluator Evaluation Date Performance Evaluation Satisfactory Unsatisfactory TASK Unload and shut down an emergency diesel generator in the control room emergency mode (1-OP-6.5). l NOTE TO THE TRAINER AND THE EVALUATOR t Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure. INITIAL CONDITIONS Control room is uninhabitable l The lH emergency diesel generator is carrying the lH emergency bus in the CRE mode The lH EDG started on a UV signal Normal emergency bus power supply is available Breaker 15F3 is closed
INITIATING CUE You are requested to unload and shut down an emergency diesel generator in the control room emergency mode. STANDARDS Task was performed as directed by the procedure referenced in the task statement within parentheses (one of the underlined procedures if several are cited) Self-checking practices were used throughout task performance Verbal communication related to any of the following modes was conducted in accordance with VPAP-1407 l e Emergency communication e Face-to-face communication i e Giving and acknowledging orders e Phonetic alphabet 4 e Telephone communication systems i TOOLS AND EQUIPMENT Administrative key Sync key i 1H and 2H diesel swap-over remote panel isolation switch key 1 (#67) 4 PREFERRED EVALUATION METHOD 4 Verbal-visual i 4 1 4
1 I l l PERFORMANCE STEPS l
- 1. Verify that the emergency nur normal feeder breaker,15H11, is open.
STANDARD i i Candidate verifies that the emergency bus normal feeder breaker, 15H11, is open. $ VERBAL-VISUAL CUES The green indicating light for 15H11 is lit. SAT [] UNSAT [] NOTE
- 2. Ensure that breaker, 15F3, is closed.
j STANDARD candidate determines that 15F3, is closed from intial conditions. . SAT [] UNSAT [] NOTE i
- 3. Turn on the 15H11 synch switch.
l l CRITICAL STANDARDS Candidate turns on synch switch. ] SAT [ ] UNSAT [ ] NOTE i
- 4. Turn the droop /irochronous switch to droop.
CRITICAL STANDARDS candidate places the droop /isochronous switch in DROOP.
- SAT [ ] UNSAT [ ] NOTE 4
l 1
( . . _ - - _ . . .. _-. . _ . . _ . - - . _ . . - ..- _ - .- . . _ . . .- - _ _ - _ 1 l 4 5. Adjust 1H EDG volhge, as required. j STANDARD Candidate adjusts 1H EDG voltage (incoming) to 1 to'2 volts higher than "C" RSST (running). l VERBAL-VISUAL' CUES 1H EDG voltage (incoming) is 120, "C" RSST (running) is 118. SAT [ ] UNSAT [] NOTE
- 6. Adjust 1H EDG speed, as required. .
1 QITICAL STANDARDS Candidate adjusts 1H EDG speed until synch scope is moving very slowly (1 rotation in 20 seconds or longer) in the fast direction. SAT [ ] UNSAT [ ] NOTE
- 7. Verify that no diesel normal shutdown alarms are locked in.
STANDARD Candidate verifies that no diesel normal shutdown alarms are locked in. VERBAL-VISUAL CUES No diesel normal shutdown alarms are locked in. SAT [ ] UNSAT [ ] NOTE s v n n
- ._ .~ . . . - , _ . - - . _ _ . . . _ - . _ _ . - - - . _ . . .- - - . . - . .
5 t
- 8. Close the emergency bus normal power supply, 15H11.
. CRITICAL STANDARDS Candidate closes 15H11 when synch scope reaches 1-minute to 12 o' clock. 3 VERBAL-VISUAL CUES l The red indicating light for 15H11 is lit. SAT (] UNSAT [ ]' NOTE
- 9. Adjust power factor and turn off the 15H11 synch switch.
I I STANDARD Candidate adjusts power factor to 0 KVAR and turns off l 15H11 synch switch. 4 { VERBAL-VISUAL CUES i 4 Vars are at O. SAT [] UNSAT ( ) NOTE
- 10. Adjust the speed and voltage, as required, to unload the diesel.
CRITICAL STANDARDS Candidate adjusts lotd to below 100KW over a 2-3 minute period. VERBAL-VISUAL CUES Diesel load is 100 KW. SAT [ ] UNSAT ( ) NOTE
.- .._.~. . _ _ . _ _ . . _ _ _ _ ___ - _ _ _ __ _....__. - _ _ _ . _ _ . - - _ _ . _ _ _ .
- 11. Open 15H2.
1 CRITICAL STANDARDS Candidate opens 1SH2. VERBAL-VISUAL CUES The green indicating light for 15H2 is lit SAT [ ]~ UNSAT [ ] NOTE i
- 12. Turn the droop /isochronous switch to droop.
STANDARD Candidate verifies the droop /isochronous switch is in DROOP. SAT [] UNSAT [ ] NOTE
- 13. Adjust 1H EDG speed to 900 RPM. l STANDARD Candidate adjusts 1H EDG speed to 900 RPM.
VERBAL-VISUAL CUES Diesel speed is 900 RPM. SAT [] UNSAT [] NOTE
- 14. Turn the droop /isochronous switch to off.
CE TICAL STANDARDS Candidate places the droop /isochronous switch in OFF. SAT [ ] UNSAT [ ] NOTE
b l 5
- 15. Turn the 15H2 synch switch to on.
STANDARD Candidate places the 15H2 synch switch to on. 4 SAT [] UNSAT [] NOTE - I
- 16. Adjust 1H EDG voltage.
- j. STANDARD
! Candidate adjusts 1H EDG voltage (incoming) to 119-120 volts. VERBAL-VISUAL CUES 1H EDG voltage (incoming) is 120 volts. SAT [] UNSAT [] NOTE
- 17. Turn the 15H2 synch switch to off.
STANDARD Candidate places the 15H2 synch switch to off. SAT [] UNSAT [] NOTE
- 18. Cool down the diesel.
STANDARD Candidate allows the EDG to run for 5 minutes un-loaded. VERBAL-VISUAL CUES Assume 5 minutes has ellapsed. SAT [] UNSAT [] NOTE
_ ._ _ _ _ _ _ .. _ . _ . _ . _ _ . . . _ _ _ _ . . . _ . _ . . , __._____....m...___._ __ .. _ _ . _ l l l'
- 19. Ensure 15H2 is in auto-after-trip.
STANDARD Candidate ensures 15H2 is in auto-after-trip. 1 VERBAL-VISUAL CUES 15H2 has a green " flag". SAT [] UNSAT [] NOTE
- 20. Shutdown the diesel.
CRITICAL STANDARDS Candidate pushes both ENGINE STOP push buttons. i l VERBAL-VISUAL CUES The diesel is coasting to a stop. SAT [] UNSAT [] NOTE
- 21. Check if 1H EDG door was blocked open.
1 j CRITICAL STANDARDS l Candidate checks status of door. VERBAL-VISUAL CUES The door was as you found it. SAT [-] UNSAT [] NOTE i l l
- 22. Verify the control room is habitable. l VERBAL-VISUAL CUES Assume another operator will complete this procedure.
SAT [] UNSAT [] NOTE l
>>>>> END OF EVALUATION <<<<< I l
l l
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4 1 JPM QUESTION SHEET j N.. 466 I TOPIC: VI-064 SYSTEM: Emergency Diesel GtptsecrM3.06 (3.3/3.4) I ! QUESTION #1 l l i
- < QUESTION : ' ANSWERl:l
.: S /U' ; With the emergency diesel can be started using the !
i generator CRE switch in blue start pushbutton on I j EMERGENCY, how can the CRE panel. diesel be started and stopped? Can be stopped using either:
- 2/2 emergency stop PB's on 4
CRE panel. t , Or j i 2/2 emergency stop PB's on MCR diesel panel. REF. Vision Objective #:
- 6299, 11715-ESK-11C a
t 4 a j i r i 1
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TOPIC: VI-064 SYSTEM: Emergency Diesel GtpfterW4.06 (2.8/3.2) QUESTION #2 j jQUESTION..; 7 ANSWERS LS/US Concerning the emergency Speed droop is diesel generators, when is automatically in service speed droop in service? whenever: i the EDG is supplying power l to the 4 Kv emergency bus in parallel with another power source, or 1 1 the EDG output breaker is synchronized (i.e. synch key or.) to close onto an emergency bus that is already powered by another power source, or another power source breaker is synchronized (i.e. synch key on) to close onto an emergency bus that is already powered by the EDG. REF. VISION OBJ. 6305, ll715-ESK-11C i 3 a
* ~-
--. _ . . _ - . = . . - . -
i 6299 Identify the push-buttons that will start and stop a diesel
- generator when each of the following switch alignments exist.
Mode selector switch in AUTO REMOTE 2/2 emergency stop PB's on CRE panel 2/2 emergency stop PB's on MCR diesel panel Mode selector switch in MAN REMOTE j Start pushbutton on diesel panel in the MCR (Auto prelube i occurs prior to EDG start) 4 2/2 emergency stop PB's on CRE panel ) 2/2 emergency stop PB's on MCR diesel panel Mode selector switch in MAN LOCAL Start pushbutton on EDG skid gageboard , All automatic starts are blocked 2/2 emergency stop PB's on CRE panel 2/2 emergency stop PB's on MCR diesel panel i CRE switch in EMERG Blue start pushbutton on CRE panel (All odd # CRE contacts close with CRE switch in i emergency) 2/2 emergency stop PB's on CRE panel 2/2 emergency stop PB's on MCR diesel panel i a 1 5 l i i l
6305 Explain the following concepts concerning speed droop on the 1 diesel generators. l
- Purpose l Speed droop stabilizes the load sharing between the generator and a larger (infinite) power source, such as the Virginia Power grid.
Dif ferences between the two types of speed droop provided Speed droop is used during parallel operations. Without the droop characteristics (referred to as isochronous operation), the generator does not have a load sharing characteristic. When speed droop is automatically in service circuits have droop characteristics inserted whenever:
- 1. the EDG is supplying power to the 4 Kv emergency bus in parallel with another power source, or
- 2. the EDG output breaker is synchronized to close onto an emergency bus that is already powered by another power source, or
- 3. another power source breaker is synchronized to close onto an emergency bus that is already powered by the EDG, How the diesel generator would respond to load changes with speed droop in service while carrying the emergency bus alone (non-parallel operation)
With speed droop control present and the EDG as the sole power supplier to a load, as the generator load increases, the speed set of the mechanical governor is lowered, causing the engine to operate at lower steady-state speeds. How the diesel generator would respond to load changes with speed droop out o_ f service while carrying the emergency bus with another source (parallel operation) Without the droop characteristics (referred to as isochronous operation), the generator does not have a load sharing characteristic. If the flat isochronous EDG characteristic were used, the EDG would pick up no load, all the load, or the load would shift between the two in an unstable manner.
Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS Job Performance . Measure N..935 INITIAL CONDITIONS Auxiliary feedwater pump,1-FW-P-3A, discharge piping upstream of the discharge check valve is hot to the touch 1-FW-P-3A has been tagged out electrically INITIATING CUE You are requested to collapse steam voids in 1-FW-P-3A in accordance with 1-OP-31.09. You are not required to clear tags or perform periodic testing.
i
\
l
- l l Virginia Power North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS 1 I Job Performance Measure N..935 a 1 Candidate Evaluator 2
Evaluation Date i Performance Evaluation Satisfactory Unsatisfactory l TASK l Collapse steam voids in a steam-bound auxiliary feedwater pump (1-OP-31.09). NOTE TO TIIE TRAINER AND THE EVALUATOR l i i Unless a specific evaluator's cue is provided, you should provide a cue indicating that ] the component or parameter is in the condition specified by the procedure. I INITIAL CONDITIONS j } Auxiliaty feedwater pump,1-FW-P-3A discharge piping upstream of the discharge check
- valve is hot to the touch j 1-FW-P-3A has been tagged out electrically a
i INITIATING CUE 4 } You are requested to collapse steam voids in 1-FW-P-3A in accordance with 1-OP-31.09. 4 You are not required to clear tags or perform periodic testing. i 1 8
?
. . i STANDARDS Task was performed as directed by the procedure referenced in the task statement within parentheses (one of the u_nderlined procedures if several are cited) Self-checking practices were used throughout task performance Verbal communication related to any of the following modes was conducted in accordance with VPAP-1407
- Emergency communication
- Face-to-face communication
- Giving and acknowledging orders
- Phonetic alphabet
- Telephone communication systems Tagging was performed in accordance with approved practices TOOLS AND EOUIPMENT Administrative key - CANDIDATE MUST SIMULATE OBTAINING ADMIN. KEY PRIOR TO "UN-LOCKING" ANY ADMIN. LOCK.
Pipe wrench - CANDIDATE MUST SIMULATE OBTAINING PIPE WRENCII PRIOR TO " REMOVING" ANY PIPE CAP. PREFERRED EVALUATION METIIOD Verbal-visual PERFORMANCE STEPS
- 1. Tag out the pump's prime mover.
STANDARD Candidate notes initial conditions stated that 1-FW-P-3A breaker has been danger-tagged. SAT [ ] UNSAT[] NOTE l
. . I
- 2. Isolate the pump's suction and discharge paths.
CRITICAL STANDARDS Candidate unlocks and closes 1-FW-172 and 160. j STANDARD Candidate checks 1-FW-166 and 162 closed. SAT [] UNSAT [ ] NOTE-
- 3. Vent the pump.
l CRITICAL STANDARDS Candidate removes pipe caps from casing vents, and 1-FW-252 and 532 are l cracked open. l VERBAL-VISUAL CUES i Assume that several minutes have passed and that 1-FW-PI-156B reads 0 PSIG. 1 SAT [ ] UNSAT [ ] NOTE
- 4. Throttle open the pump's suction from the emergency condensate storage tank in order 1 to cool the pump's casing.
CRITICAL STANDARDS Candidate cracks open 1-FW-160. VERBAL-VISUAL CUES Assume that several minutes have passed and the pump casing is now cool SAT [ ] UNSAT [ ] NOTE
- 5. When the pump's casing has been cooled, return the pump to its normal alignment.
CRITICAL STANDARDS Candidate closes 1-FW-252 and 532. Candidate opens 1-FW-160 and 172. ' STANDARD Candidate checks 1-FW-166 and 162 closed. SAT [ ] UNSAT [ ] NOTE
- 6. Clear the tag-out on the pump's prime mover.
VERBAL-VISUAL CUES Assume that another operator will complete this procedure SAT [ ] UNSAT [ ] NOTE
> > > > > END OF EVALUATION < < < < <
JPM QUESTION SHEET N. 935 TOPIC: IV-061 SYSTEM: Aux. Feed K/A: 061 A2.06 (2.7/3.0) QUESTION #1 - [QUESTIDN3 ihNSW$Rt IS/.U} What are the adverse effects of If an AFW pump becomes steam having a steam-bound AFW pump? bound, the possibility exists that upon starting NPSII will not be available, and the pump will not deliver flow to the steam generator. REF. Vision Objective #: 5977 TOPIC: IV-061 SYSTEM: Aux. Feed K/A: 061K4.02 (4.5/4.6) QUESTION #2 IQUESTIUNi iANSWEIO , i!!S/UI Describe the operator actions If a safety injection signal is present, required for manually stopping a regardless of which signal actually l motor-driven auxiliary feedwater started the pump, then SI must be pump from the auxiliary shutdown reset before the pump can be panel following an automatic start. manually stopped. Assume control has already been transfered to the auxiliary shutdown If no SI signal is present, then the panel, pump can be manually stopped without clearing the AUTO-START signals, however if the control switch is returned to AUTO, the pump will re-start. REF. Vision Objective #:10173, 11715-ESK-5AA/5AB l
1 5977 Explain the following concepts associated with steam binding of the auxiliary feedwater pumps. l
- Probable cause l {
] The most probable cause of steam binding of an auxiliary feedwater pump is ! back-leakage through its associated discharge check valves. j
- Possible consequences if a pump becomes steam-bound 1
j If an AFW pump becomes steam bound, the possibility exists that upon starting j NPSH will not be available, and the pump will not deliver flow to the steam l l
- generator.
i Methods used to detect when a pump may be steam-bound The AFW pumps are monitored for possible steam binding by the safeguards operator. Indication of steam binding is a pump casing that is hot to the touch. i 4 1 b l i 1 j l l 4 i l i i I ) 4 i 1 e 4 4
l l l 10173 List the operator actions required for manually stopping a motor-driven auxiliary feedwater pamp from the auxiliary shutdown panel following an automatic start. If a safety injection signal is present, regardless of which signal actually started the pump, then SI must be reset before the pump can be manually stopped. {If the control room is inaccessible, the pump's control switch could be placed in PULL-TO-LOCK, then the breaker tripped open locally at the breaker cabinet. The breaker will not automatically re-close with the switch in PTL } If no SI signal is present, then the pump can be manually stopped without clearing the AUTO-START signals, however if the control switch is returned to AUTO, the pump will re-start. l I
! i
- I Virginia Power j North Anna Power Station REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS
JOB PERFORMANCE MEASURE
- N..230 i
1 INITIAL CONDITIONS ] Component cooling surge tank level is 70%. s INITIATING CUE { You are requested to drain 10% from the component cooling water surge tank to both
- 1-LW-TK-5A and 1-LW-TK-5B IAW 0-OP-51.3 1
1 ) h 4 l'
j . . 1 i-Virginia Power
.: North Anna Power Station 1
) REACTOR OPERATOR / SENIOR REACTOR OPERATOR LICENSE CLASS JOB PERFORMANCE MEASURE l N..230 4 1 Operator Evaluator J Observer Evaluation Date l, \ I Performance Evaluation [ l Satisfactory [ 1 Unsatisfactory f i 1 l' TASK ' 4 ) Drain the Component Cooling Water System to a liquid waste test tank on the auxiliary j building watchtation (0-OP-51.3). l
; TASK COMPLETION TIMES l 1
Approximate = 10 minutes Actual = minutes
Virginia Power North Anna Power Station l LICENSED OPERATOR REQUALIFICATION PROGRAM i ' JOB PERFORMANCE MEASURE l l N..230 1 TASK Drain the Component Cooling Water System to a liquid waste test tank on the auxiliary building watchtath (0-OP-51.3). f
- NOTE TO THE TRAINER AND TIIE EVALUATOR Unless a specific evaluator's cue is provided, you should provide a cue indicating that the component or parameter is in the condition specified by the procedure.
? l INITIAL CONDITIONS i Component cooling surge tank level is 70%. INITIATING CUE You are requested to drain 10% from the component cooling water surge tank to 1-LW- , TK-5A and 1-LW-TK-5B IAW 0-OP-51.3 STANDARDS
- Task was performed as directed by the procedure referenced in the task statement within parentheses (one of the underlined procedures if several are cited)
Self-checking practices were used throughout task performance 2 Verbal communication related to any of the following modes was conducted in accordance with VPAP-1407
- Emergency communication
,
- Face-to-face communication
- Giving and acknowledging orders
] j
- Phonetic alphabet 1
- Telephone communication systems f
. _ . ~ . ___ .- ~ _ . _ . _ _ . _ _ . .
TOOLS AND EOUIPMENT None , j EVAI,UATION METHOD 1 I Verbal-visual 1 ! PERFORMANCE STEPS l I i
- 1. ' Verify that the liquid waste test tank pumps are isolated.
STANDARD 1 Candidate verifies the following valves closed: 1-LW-76,1-LW-87,1-LW-1003 SAT [] UNSAT [ ] NOTE
- 2. Align the desired test tank (s).
CRITICAL STANDARDS Candidate opens 1-LW-86,1-LW-75,1-LW-74 and 1-LW-1002. SAT [ ] UNSAT [ ] NOTE
- 3. Request the control room operator to monitor component cooling surge tank level.
STANDARD Candidate informs OATC that he will be draining the CC head tank. SAT [ ] UNSAT [ ] NOTE
- 4. Begin draining the surge tank to the test tank (s).
j CRITICAL STANDARDS l Candidate throttles open 1-CC-1012, i SAT [ ] UNSAT [ ] NOTE
i l 5. When notified by the control room operator that the desired surge tank level is reached, ) secure the draining. I \ s 1
- VERBAL-VISUAL CUES j The control room reports that CC surge tank level is 60%.
CRITICAL STANDARDS Candidate closes 1-CC-1012. SAT [ ] UNSAT [ ] NOTE
- 6. Secure the test tank (s) alignment.
STANDARD Candidate closes 1-LW-86,1-LW-75,1-LW-74 and 1-LW-1002. SAT [ ] UNSAT [ ] NOTE
> > > > > END OF EVALUATION < < < < <
) .
JPM QUESTION SIIEET N. 230 i TOPIC: VIII-008 SYSTEM: Component Cooling K/A: 008K4.01 (3.1/3.3) f QUESTION #1
- NUESTION] EANSW55 - IS/Us
! 1-CC-P-1 A is running, normal feeder to 13 Both CC pumps will be running. bus trips open, IJ EDG starts and powers up the bus. What is the status of CC pumps? REF. Vision Objective #: 3656,11715-ESK-( 005P/Q i 1 a v 4 1 4 t 4 i i i 3
TOPIC: Vill-008 SYSTEM: Component Cooling K/A: 008A1.01 (2.8/2.9) QUESTION #2 IQUESTIONS LANSWERi l S/th Explain how shifting the Component Cooling Since the component cooling water system is Water System common loads supply from one normally operated cross-connected between unit to the other will affect each unit's reactor units, shifting common loads has no effect on coolant pump component cooling water flows. the system. If the CC system is being operated with the units split, then one unit or the other will supply common CC loads. When the Component Cooling Water System common loads supply is shifted from one unit to the other, the affect on each unit's reactor coolant pump component cooling water flows will be as follows: RCP CC flows will be lowered on the unit supplying common loads because the unit supply header pressure will continue to drop after the recirculation valve (PCV-110) has fully closed. REF. Vision Objective #: 3686, 11715-FM-79 SERIES
3656 List the following information associated with the component cooling water pumps. Power supply to each pump The component cooling water pumps are powered from the 4160 volt stub-busses: Interlocks associated with manually starting a pump i To manually start CC pump 1-CC-P-1A, the following conditions must exist: No ground or phase overcurrent condition exists Nomial voltage on the supply bus for at least 20 seconds No CDA signal present Interlocks associated with automatically starting a pump A CC pump will automatically start, provided all of the following conditions exist: Selected control switch in AUTO as appropriate (local or remote), No CDA signal No ground or phase overcurrent condition I Either: Auto-trip signal on opposite pump or undervoltage condition on the opposite bus, assuming no UV/DV signal on supply bus for at least 20 seconds Following a UV/DV on supply bus: power restored for at least 15 seconds, but no more than 20 seconds. Interlocks associated with automatically tripping a pump The following conditions will automatically trip a CC pump: Undervoltage condition on the supply bus CDA signal Ground or phase overcurrent condition
4 i ! 1 3686 ' Explain how shifting the Component Cooling Water System common loads supply from one unit to the l i other will affect each unit's reactor coolant pump component cooling water flows. { Since the component cooling water system is normally operated cross-connected between units, shifting j common loads has no effect on the system. If the CC system is being operated with the units split, then 1 one unit or the other will supply common CC loads. When the Component Cooling Water System l common loads supply is shifted from one unit to the other will affect each unit's reactor coolant pump . component cooling water flows as follows: RCP CC flows will be lowered on the unit supplying ! common loads because the unit supply header pressure will continue to drop after the recirculation valve ! (PCV-110) has fully closed, t i 4 l 1
l
- 1 l
ES-301 Scenario Events Form ES-301-3 Simulation Facility: North Anna Scenario No.: 1 Examiners: v 5b[au- Applicants: k- d,kug:Ro Initial Conditions: 100% power. Turnover: The scenario will begin with both units at 100 % power. The following unit 1 equipment is unavailable: 1-FW-P-3B is tagged out for scheduled lube oil maintenance (expected to take 18 more hours),1-FW-P-1B is tagged out for bearing cooling line clean out (expected to take 4 more days). Shift orders are to maintain current plant conditions. Event Malf. Event Type
- Event No. No. Dexdpdon 0.a. C:ALL Failure of single train SI.
0.b C:RO Running HHSI pump trips, standby doesn't auto start. 1.0 I:SRO/RO Continuous Uncontrolled Rod Motion 2.0 I:SRO/RO PRZR level transmitter fails low. 3.0 C:ALL Ietdown line develops > 10 GPM leak. 4.0 N:ALL Ramp unit off-line. ' g' R:RO
/
5.0 MT:ALL SBLOCA. / C:SRO/ BOP Failure of single train SI. (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor i Facility Author: ktiT" h. LINC ~ Chief Examiner: 4v Giav / , O7(J l S- ww ; r e o p1 6 s % 0[ ' A/N'r
l i s 1 . i i 4 i l i VIRGINIA POWER NORTH ANNA POWER STisTION I l INITIAL LICENSE CLASS NRC SIMULATOR EXAMINATION SCENARIO I-1 PICKERING AS RO SCENARIO 1 2 Revision 0
4 NRC SIMULATOR EXAMINATION SCENARIO i
- EVENT DESCRIPTION 1
- 1. Continuous Uncontrolled Rod Motion j 2. PRZR Level Failure
- 3. I2tdown Leak
- 4. Ramp Unit Off-Line 2
- 5. SBLOCA i
SCENARIO DURATION i 4 60 Minutes 1 1 1 1 1 4 i 4 l I l 1 1 1 SCENARIO 1 3 Revision 0 l
l I i SCENARIO
SUMMARY
) The scenario will begin with both units at 100 % power. Unit 1 is at 8000 MWD /MTU, Cb = 784 PPM. Aux. steam is being supplied from unit l's second point extraction steam. i The following unit 1 equipment is unavailable: 1-FW-P-3B is tagged out for scheduled lube oil maintenance (expected to take 18 more hours),1-FW-P-1B is tagged out for bearing cooling line clean out (expected to take 4 more days). Shift orders are to maintain current plant conditions.
)
The first event will be a failure in the automatic rod control circuitry, resulting in the ! control rods stepping in at 72 steps per minute. The crew will be expected to respond IAW 1-AP-1.1, " CONTINUOUS UNCONTROLLED ROD MOTION". The RO will place rod control in manual and verify rod motion stopped. The RO will restore Tave and PRZR pressure as required. After the US has requested assistance from the instrument department, and the plant
- has been stabilized, the next event will occur.
The selected PRZR level channel will fail low resulting in a loss of letdown and the opening of the charging flow control valve. The RO will be expected to respond IAW 1-AP-3,
" LOSS OF VITAL INSTRUMENTATION" and take manual control of PRZR level. After the
, crew has restored letdown and places the failed channel in trip, the next event will occur. As a result of the loss and restoration ofletdown, an un-isolable leak will develop in the letdown system. The crew will be expected to respond IAW 1-AP-5, " UNIT 1 RADIATION MONITORING S'YSTEM" and 1-AP-16, " INCREASING PRIMARY PLANT LEAKAGE". An RCS flow balance will indicate an approx.15 GPM leak that will not be isolated when letdown is secured. The US will direct the crew to commence ramping the unit off-line IAW TS 3.4.6.2. After the unit has been ramped a sufficient amount, the last event will occur. The leak in the letdown system will increase, resulting in a small break (<2") LOCA. The crew will be expected to respond IAW 1-E-0, " REACTOR TRIP OR SAFETY INJECTION". A single train SI will occur, with the "B" HHSI pump tripping and the standby pump not start automatically. The RO must manually start the "A" HHSI pump. The crew will i proceed through 1-E-0 and transition out to 1-E-1, " LOSS OF REACTOR OR SECONDARY I l COOLANT". The scenario may be terminated after the crew has checked QS pump status, or as directed by the lead examiner. l 4 1 1 i l 4 SCENARIO 1 4 Revision 0
l l EVENT 1: Continuous Uncontrolled Rod Motion i
~ TIME
- 1. RO identil'es control rods stepping inward.
- 2. RO determines that rod movement is not called for.
- 3. RO places control rod bank selector switch in manual IAW 1-AP-1.1,
" CONTINUOUS UNCONTROLLED ROD MOTION." l
- 4. RO verifies rod motion stopped.
l
- 5. RO checks RCS Tave within 1.5'F of Tref. l
- 6. Crew maintains stable plant conditions.
- 7. US notifies instrument department of problem.
NOTE: STEPS 8 THRU 13 REQUIRED IF Tave/ Tref WERE NOT WITHIN 1.5*F.
- 8. RO checks Tave less than Tref.
- 9. RO verifies Tave greater than 541 F.
_ RO restores Tave as directed by US,
- 10. RO checks PRZR pressure stable or trending to 2235 PSIG.
RO energizes heaters as required. Crew reduces turbine load if required.
- 11. RO checks PRZR level stable.
RO takes manual control of 1-CH-FCV-1122. COMMFNTS: SCENARIO 1 5 Revision 0
EVENT 1: Continuous Uncontrolled Rod Motion
~ TIME
- 12. RO verifies insertion limits not exceeded.
__ RO restores rods above insertion limit as directed by the US.
- 13. RO restores Tave to program as directed by the US.
NOTE: AFTER THE CREW HAS STABILIZED THE PLANT, THE NFXT EVENT WILL OCCUR. COMMENTS:
)
i SCENARIO 1 6 Revision 0
)
EVENT 2: PRZR Level Failure TIME
- 1. RO identifies various PRZR level alarms
- 2. RO identifies PRZR level channel III failing.
- 3. US directs crew to perform actions of 1-AP-3, " LOSS OF VITAL INSTRUMENTATION".
- 4. RO and BOP verify redundant instrument channel indication normal.
- 5. BOP verifies SG level parameters normal.
- 6. RO verifies PRZR level indications normal. (NO)
RO performs RNO step and places 1-CH-FCV-1122 in manual and controls PRZR level.
- 7. BOP verifles turbine 1st stage pressure indication normal.
- 8. RO verifies systems affected by PRZR level channels normal. (NO)
RO performs RNO step and selects operable channel RO restores charging and letdown as directed by the US. RO resets PRZR heaters.
- 9. Crew verifies remaining vital instrumentation normal.
- 10. US refers to TS 3.3.1.1 for operability.
- 11. TT5 notifies instrument dept. of failure.
- 12. Crew determines that channel should be placed in trip IAW 1-MOP-55.72,
" PRESSURIZER LEVEL INSTRUMENTATION".
COMMENTS: l SCENARIO 1 7 Revision 0
4 4 EVENT 3: Letdown Leak Inside Containment
~ TIME 3
j 1. Crew identifies annunciator K-D2, " RAD MONITOR SYSTEM HI RAD 4 LEVEL".
- 2. Crew identifies 1-RMS-RM-160, containment gaseous, in alarm.
j i
- 3. US directs BOP to perform actions of 1-AP-5, " UNIT 1 RADIATION
, MONITORING SYSTEM".
- 4. US directs RO to perform actions of 1-AP-16, " INCREASING PRIMARY l PLANT LEAKAGE".
- 5. RO verifies PRZR level and RCS subcooling under his control.
- 6. RO verifies 1-CH-LCV-1115A not diverted.
l 7. Crew identifies containment sump pumping rate has increased. 1
- 8. RO performs RCS flow balance and identifies approx. 24 GPM leak.
- 9. US refers to TS 3.4.6.2 for operability.
NOTE: THE OMOC WILL DIRECT A UNIT SHUTDOWN WHEN INFORMED OF i RCS LEAK. )'
- 10. US directs crew to ramp unit off-line IAW TS.
COMMENTS: SCENARIO 1 8 Revision 0
1 L . EVENT 4: Ramp Unit Off-Line 1
- TIME
- 1. US briefs crew on ramp.
i -{
!. 2. RO calculates amount of reactivity change due to ramp.
- 3. RO commences lowering Tave using boration/ control rods.
- 4. BOP commences lowering main turbine load.
Verifies load rate at .3%/ min. (may select higher rate until off
- limiter).
4 Lowers reference setter. ! Pushes GO button. j .
- 5. BOP directs turbine building watchstander to place LP Heater Drain j Pumps on recire and shutdown when power is approx. 94%.
l 4 t l COMMENTS: i l 4 i e 4 1 4 i l 1 1
- SCENARIO 1 9 Revision 0 1
EVENT 5: SBLOCA TIME
- 1. RO identifies rapidly decreasing pressurizer level and pressure.
, 2. RO informs the US of RCS conditions. 4
- 3. US directs crew to manually trip the reactor and perform actions of 1-E-0.
! 4. Crew trips reactor. i 5. BOP trips turbine.
- 6. RO verifies AC emergency busses energized.
j 7. Crew checks if safety injection has actuated.
- 8. Crew checks if safety injection is required. (YES)
- 9. US directs crew to manually initiate SI.
2
- 10. RO and BOP initiate SI.
i s 11. RO verifies both emergency diesels running. (NO)
- BOP starts 1J EDG as directed by the US.
1
- 12. Crew verifies SI pumps running. (NO)
RO performs RNO step and starts 1-CH-P-1A. BOP performs RNO step and starts 1-SI-P-1B. 1
- 13. BOP verifies main feedwater isolation.
COMMENTS: h e i l
, l I
j SCENARIO 1 10 Revision 0
EVENT 5: SBLOCA TIME
- 14. BOP verifies AFW Pumps are running.
- 15. Crew verifies Phase "A" isolation.
- 16. BOP verifies service water pumps running.
BOP starts 1-SW-P-1B, '
- 17. Crew checks if MS line should be isolated.
- 18. Crew checks if CDA is required. (NO)
- 19. Crew checks if QS is required. (NO)
- 20. BOP verifies SI flow indicated.
- 21. BOP ensures MSR vents aligned to main condenser.
- 22. BOP adjusts gland steam as required.
- 23. BOP verifies AFW flow.
- 24. US directs unit 2 to initiate 0-AP-47, " UNIT OPERATION WITH OTHER UNIT EMERGENCY".
- 25. RO checks RCS average temperature.
- 26. RO checks RCP trip and charging pump recirc criteria.
COMMENTS: SCENARIO 1 11 Revision 0
EVENT 5: SBLOCA
' TIME i
- 27. US notifies STA, initiates EPIPs.
- 28. RO checks PRZR PORVs, spray valves, and safety valves closed. j l
- 29. Crew checks SGs not faulted.
- 30. Crew checks SGs not ruptured.
- 31. Crew checks RCS intact. (NO)
- 32. US directs crew to transition to 1-E-1.
I
- 33. STA commences monitoring CSFs.
- 34. RO checks RCP trip and charging pump recirc criteria.
- 35. BOP checks SGs not faulted.
- 36. BOP checks intact SG levels.
- 37. Crew checks secondary radiation levels.
- 38. RO checks PRZR PORVs.
- 39. Crew checks if SI can be terminated.
COMMENTS: SCENARIO 1 12 Revision 0
l 3 I
! EVENT 5: SBLOCA i
' EME
} 40. RO/ BOP resets CDA. 1
- 41. BOP checks QS pump status.
J i NOTE: The scenario may be termi:iated after the crew has checked QS pump status, or j as directed by the lead examiner. COMMENTS: 1 1 i 1 1 4 i 3 i 4 f i 1 1 1 l, e 1 i a l i i j i 4 i I j SCENARIO 1 13 Revision 0 i . - _
SIMULATOR OPERATOR'S COMPUTER PROGRAM SCENARIO 1 14 Revision 0
1 l l SIMULATOR OPERATOR'S COMPUTER PROGRAM SCENARIO II-1 Initial conditions
- 1. RecallIC # (100% Power)
- 2. Ensure Tave, Tref, PD7T level, and VCT level are selected on trend recorders.
' 3. ENSURE PRZR LEVEL CONTROL IS SELECTED TO TIIE "461/460 " POSITION. 4 ] PREQ)][IBTHEROJR(ifiGjpg1pRj@]TfRTiOQCENXRi1O]
- 1. MSIO702, TD = 0 (FAILURE OF "J" TRAIN SI)
, 2. TAGOUT 1-FW-P-3B j 3. TAGOUT 1-FW-P-1B 2
- 4. MCH1602, TD = 0 SEC, TRGR = SIl ("B" HHSI PUMP TRIP ON SI)
EVENT MALFUNCTION / OVERRIDE / COMMUNICATIONS
- 1. Continuous Uncontrolled MALFUNCTION: MRD07 Rod Motion TD = 15 SEC TRGR = N/A
- 2. PRZR Level Failure MALFUNCTION: MRC0803 TD = 30 SEC RAMP = 5 START = 50 STOP = 0
- 3. Letdown Leak Inside MALFUNCTION: MRC04 Containment TD = 30 SEC RAMP = 60 START = 0 STOP = 3.5 NOTE: THE OMOC WILL DIRECT A UNIT SHUTDOWN WHEN INFORMED OF RCS LEAK.
SCENARIO 1 15 Revision 0
4 EVENT MALFUNCTION / OVERRIDE /CO.MMUNICATIONS
- 4. Ramp Unit Off-Line PROVIDE FEEDBACK AS REQUIRED
- 5. SBLOCA UPDATE MALFUNCTION: MRC04 i E=0 RAMP = 30 START = 3.5 i STOP = 75 4
i e l l SCENARIO 1 16 Revision 0
. . . - - - - - - _ _ _ = - - - - . - - . - . . - - - .. - - . - - .. _ . . - -
1 1 4
- ES-301 Scenario Events Form ES-301-3 4
Simulation Facility: North Anna Scenario No.: 2 1 Examiners: 9cd bbe Applicants: 2 'bker.ys 30P i i Initial Conditions: l 35% power. i j Turnover: l The scenario begins with unit I at 35 % chemistry hold. The unit had been i placed on line the previous shift following a unit trip. The trip was caused when ; ! starting 1-FW-P-1B un-coupled following motor bearing replacement. A fault l occurred in the breaker for 1-FW-P-1B2 which did not trip and resulted in a loss of l l the "B" station service bus, this in turn caused a reactor trip due to the loss of 1-RC- I l P-1B. Unit 2 is at 100 % power with no limiting actions. The following unit 1 l equipment is unavailable: 1-FW-P-3B is tagged for scheduled lube oil maintenance, ! i expected to be RTS in 18 hours,1-FW-P-1B is tagged for breaker repair and is :
, expected to be RTS in 16 hours. Shift orders are to start a second main feedwater i pump and ramp the unit to 90% power. ;
1 j Event Malf. Event Type
- Event No. No. Description j' O.a C: BOP Failure of automatic turbine trip.
- 1.0 N
- SRO/ BOP Start second main feedwater pump.
! 2.0 I:SRO/ BOP "C" SG level fails low. 4 3.0 C:ALL Loss of 1H emergency bus. , 4.0 MT:ALL Loss of all main feedwater pumps, terry-turbine C:SRO/ BOP trips on start. j (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Facility Author: kt aw tA LiML Chief Examiner: 'Pu\Q b f l i l
4.42_.m4.A .---4S-_,._msu,M -*mMa --wa. 4- Ar 4 4,As -&aAh,_- A4 g A4 J 3rw.AM-_he 4_m.4.A.As._JhJ_49 4.AA8. h_Ae,--,.eJ.2_--,A-arJ+*%+- m3eAmi4 4.4AA-A A A i , .1 h. m_14.,__,L4, Aa ..4m 4 s 4 i 1 i 4 i i' i VIRGINIA POWER NORTH ANNA POWER STATION i i I 1 i i INITIAL LICENSE CLASS NRC SIMULATOR EXAMINATION SCENARIO I-2 PICKERING AS BOP
4 NRC SIMULATOR EXAMINATION SCENARIO ) EVENT DESCRIPTION
- 1. Start Second Main Feedwater Pump
- 2. Ramp Unit To 90% Power
- 3. "C" Steam Generator Level Transmitter Fails Low l
,; 4. Loss Of 1H Emergency Bus j 5. Loss Of Heat Sink 4 SCENARIO DURATION 60 Minutes l l l l I SCENARIO 2 3 Revision 0 I t
SCENARIO
SUMMARY
- The scenario will begin with unit I at 35 % power, 8000 MWD /MTU, Cb = 1070 PPM, following a forced outage to repair a steam leak in an extraction steam line. Unit 2 is at 100%
} power, supplying aux. steam from its second point extraction. The following unit 1 equipment ! is unavailable: 1-FW-P-3B is tagged out for scheduled lube oil maintenance (expected to take 18 more hours),1-FW-P-1B is tagged out for breaker repair and is expected to be RTS in 16 hours. Shift orders are to start a second main feedwater pump and ramp the unit to 90% power. 5 The first event will be the starting of a second main feedwater pump by the BOP. After
- the pump has been started, the next event will occur.
i 4 The "C" SG level channel III will fail high requiring the BOP to respond IAW 1-AP-3 l and take manual control of "C" SG level. After the crew has placed the failed channel in trip, the next event will occur.
- A loss of the IH emergency bus will occur, with the IH emergency diesel wiU failing i to start, leaving the bus de-energized. In addition, the BOP willidentify that the "B" component ;
i cooling water pump did not auto start and manually start it. After the RO completes the ! l diagnostic portion of 1-AP-10, and reports to the US, the last event will occur.
)
l i A second "C" SG level channel will fail high, resulting in a loss of all main feedwater l pumps on P-14 (HI-HI SG LEVEL). The automatic turbine trip will not occur and the terry , j turbine AFW pump will trip on over-speed upon starting. This, along with 1-FW-P-3B being I
- tagged out for maintenance and the loss of the 1H emergency bus, will result in the total loss !
of feedwater to the SGs. The crew will be expected to respond IAW 1-E-0, " REACTOR TRIP l OR SAFETY INJECTION", where the turbine will trip manually. Eventually, the crew will transition to 1-FR-H.1, " RESPONSE TO LOSS OF SECONDARY HEAT SINK." The scenario may be terminated anytime after condensate flow is established to at least 1 SG or as directed by the lead examiner. l l l l SCENARIO 2 4 Revision 0 l 1
- .._. - - . . . -_. .. . _ - . = .- - .. . ..
e 1 EVENT 1: Start a Second Main Feedwater Pump I
~ TIME
- 1. BOP revie. initial conditions, precautions, and limitations.
l
- 2. BOP request the turbine building operator to verify the support conditions
] for the "C" feedwater pump. 1 1
- 3. BOP verifies that the FW PP 1C LUBE OIL LO PRESS annunciator (IE-H7) is not lit.
j 4. BOP places the control switch for the standby condensate pump in PULL-TO-LOCK.
- 5. BOP opens feedwater pump recirculation valve 1-FW-FCV-150A,150B, or 150C.
, 6. BOP notes the control switches for the standby feedwater pump are already in PULL-TO-LOCK for its tagout. I 7. BOP places the control switch for the "C1" feedwater pump motor in i PULL-TO-LOCK. , 8. BOP closes feedwater pump discharge motor-operated valve , 1-FW-MOV-150C. I
- 9. BC P places the control switch for the "C2" feedwater pump motor in START.
- 10. BOP verifies that the feedwater pump motor current lowers to a normal value.
COMMENTS: 1 SCENARIO 2 5 Revision 0
l i I EVENT 1: Start a Second Main Feedwater Pump TIME
- 11. BOP places the control switch for the "C1" feedwater pump motor in START.
- 12. BOP verifies that the feedwater pump motor current lowers to a normal value.
, 13. BOP opens feedwater pump discharge motor-operated valve i 1-FW-MOV-150C. i
- 14. BOP requests that the turbine building operator to place the control switch for the auxiliary lube oil pump in AUTO.
- 13. BOP p' aces the coairol switses for the 'C" feedwater pump inotors in AUTO-AFTER-STOP.
- 16. BOP verifies that feedwater pump discharge motor-operated valve 1-FW-MOV-150C is open.
- 17. BOP places the feedwater pump recirculation valve that was previously opened in AUTO.
- 18. BOP places the control switch for the standby condensate pump in AUTO-AFTER-STOP.
COMMENTS: SCENARIO 2 6 Revision 0
i j EVENT 2: "C" Steam Generator Level Transmitter Fails Low I TIME i 1. BOP identifies annunciator F-F2, "SG1B LEVEL ERROR". l I l
- 2. BOP identifies "C" SG level channel III failing.
- 3. US directs crew to perform actions of 1-AP-3, " LOSS OF VITAL INSTRUMENTATION".
- 4. RO and BOP verify redundant instrument channel indication normal.
- 5. BOP verifies SG level parameters normal. (NO)
BOP performs RNO step and places "C" MFRV control in manual. 1
- 6. RO verifies PRZR level indications normal.
4
- 7. BOP verifies turbine 1st stage pressure indication normal.
i
- 8. RO verifies systems affected by PRZR level channels normal.
l 9. Crew verifies remaining vital instrumentation normal. ! , 10. US refers to TS 3.3.1.1 for operability.
- 11. Crew places channel in trip IAW 1-MOP-55.76, " STEAM GENERATOR
- LEVEL INSTRUMENTATION".
I NOTE: AFTER THE CREW 'HAS PLACED THE CHANNEL IN TRIP, THE NEXT EVENT WILL OCCUR. COMMENTS: 9 SCENARIO 2 7 Revision 0
- ^ " ^ ^ --
, EVENT 3: Loss of 1H Emergency Bus. NOTE: PRE-BRIEF UNIT SRO THAT YOU WANT THE BOP TO REMAIN ON THE BOARD FOR THIS EVENT. TIME
- 1. Crew recognizes loss of IH Bus.
, 2. US directs crew to enter 0-AP-10. ! 3. RO commences 0-AP-10 diagnostic. j
- 4. Crew recognizes that 1H diesel did not re-energize the bus.
4
- 5. BOP identifies 1-CC-P-1B did not auto start.
___,__ BOP starts 1-CC-P-1.?
- 6. US evaluates T.S. 3.7.3.2, 3.8.1.1 and 3.8.2.1 for equipment lost.
J
- 7. US directs electrical department to investigate loss of 1H emergency bus and 1H EDG.
2 NOTE: AFTER THE RO/ BOP BRIEFS US ON EXTENT OF POWER LOSS, THE 1
- NEXT EVENT WILL OCCUR. l j COMMENTS
i 3 I SCENARIO ? 8 Revision 0
n -
- ,._..2 - -- .a _. - - .ns.- >
EVENT 4: Loss Of Heat Sink 1 TIME
- 1. BOP identifies failure of "C" SG level channel II.
- 2. BOP identifies loss of all main feedwater pumps.
- 3. BOP identifies that turbine did not automatically trip.
i
- 4. US directs crew to manually trip the reactor and turbine and perform actions of 1-E-0.
- 5. Crew trips reactor.
l 6. BOP trips turbine.
- 7. RO verifies AC emergency busses energized.
- 8. Crew checks if safety injection has actuated. (NO)
- 9. Crew checks if safety injection is required. (NO)
- 10. US directs crew to transition to 1-ES-0.1, " REACTOR TRIP RESPONSE."
l
- 11. Crew identifies red-path on heat sink.
- 12. US directs entry into 1-FR-H.1, " RESPONSE TO LOSS OF SECONDARY HEAT SINK."
COMMENTS: 1 SCENARIO 2 9 Revision 0
EVENT 4: Loss Of Heat Sink. TIME
- 13. Crew checks if secondary heat sink is required. (YES)
- 14. BOP attempts to establish AFW flow to at least one SG.
- 15. RO stops reactor coolant pumps.
- 16. Crew attempts to establish feed flow to at least one SG. I i
- 17. BOP verifies at least one condensate pump running. l
- 18. BOP establishes line-up for feed flow from the condensate system. l BOP opens at least one main feedwater pump discharge MOV. !
DOP opens at least one main feedwater reg, bypass valve.
- 19. Crew depressurizes RCS.
- 20. RO maintains RCS pressure less than 1950 PSIG.
- 21. RO blocks low PZR pressure SI.
- 22. US verifies condensate feed path aligned.
- 23. Crew injects BIT.
- 24. Crew depressurizes all SGs to between 610 and 120 PSIG.
- 25. RO blocks high steam flow SI. l COMMENTS:
SCENARIO 2 10 Revision 0
4 EVENT 4: Loss of Heat Sink.
' TIME
- 26. BOP verifies condensate flow to at least SG.
BOP verifies wide range SG level (s) increasing. RO verifies core exit TCs decreasing. NOTE: THE SCENARIO MAY BE TERMINATED ANYTIME AFTER l ESTABLISHING CONDENSATE FLOW TO AT LEAST 1 SG OR AS i DIRECTED BY THE LEAD EXAMINER. COMMENTS: l i j J i ) 1 l
- SCENARIO 2 11 Revision 0
j SIMULATOR OPERATOR'S COMPUTER PROGRAM SCENARIO 2 12 Revision 0
1 l 1 l SIMULATOR OPERATOR'S COMPUTER PROGRAM SCENARIO 2 Initial conditions
- 1. Recall IC # (30% Power)
- 2. Ensure Tave, Tref, PDTT level, and VCT level are selected on trend recorders.
l
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- 1. TAGOUT 1-FW-P-3B
- 2. TAGOUT 1-FW-P-1B
- 3. SWITCH OVER-RIDE: EG1H_ AUTO _ REMOTE, TD = 0, OVRD = OFF, TRGR =
N/A
- 4. 15H2 NOT close: SIMLOCH: ED15H2_LO = T
- 5. SIMLOCH: MTU03, FAILURE OF AUTO TURBINE TRIP.
- 6. ON SIMLOCH: CCP1_ AUTO _ DEFEAT (2) = T (BLOCK AUTO-START OF 1-CC-P-1B)
EVENT MALFUNCTION / OVERRIDE / COMMUNICATIONS
- 1. Start Second PROVIDE FEEDBACK AS REQUIRED Main Feedwater Pump
- 2. "C" Steam Generator MALFUNCTION: MFW01 Ievel Transmitter Fails Low TD = 15 SEC RAMP = 45 SEC START = 50 STOP = 0 POS = Y TRGR = N/A SCENA.RIO 2 13 Revision 0
EVENT MALFUNCTION / OVERRIDE / COMMUNICATIONS i
- 3. Loss Of IH Emergency Bus MALFUNCTION: MELO301 I
TD = 30 SEC TRGR = N/A
-or-SIMLOCH EL15H11_OC = T
- 4. Loss Of All Feedwater MALFUNCTION: MFW01 RAMP = 60 SEC START = 50 STOP = 0 i
POS = Y TRGR = N/A 5 l i l s 4 4 1 I 4 i SCENARIO 2 14 Revision 0
I l ES-301 Individual Walk-through Test Outline - Set #1 Form ES-3012 Examination Level (Circle One): / SRO(1) / SRO(U) Facility: North Anna Week of Exammation: Jan. 29, 1996 Exa==r's Name (print): G.uk S,Lu System / JPM Safety Planned Follow-up Questions: Function K/A/G // Importance // Description
- 1. Control Rod Drive - I-001 s. 003AA2.03 (3.6/3.8) l Retrieve a dropped rod SIM Effects of dropped rod on major plant parameters. l (R.. 476)
- b. 001K4.03 (3.5/3.8) l Given a set of conditions, determine effect on rod control.
- 2. Chemical and Volume II-004 a. GEN-2.1.25 (2.8/3.1)
Control - Place excess SIM Given a set of plant conditions, use graphs to determine l letdown in service (R.. 333) blended flow,
- b. GF.N-2.1.32 (3.4/3.8)
Describe the reasons for the procedure precautions.
- 3. Emergency Core Cooling - III-006 a. 011EK3.13 (3.8/4.2)
Transfer the Safety Injection SIM Describe the reasons for swapping from hot-leg back to System from hot-leg to cold- cold-leg injection. leg recirculation (R.. 736) ALT. PATH / ESF / NEW b. 006A2.13 (3.9/4.2) Actions required following a spurious SI
- 4. Residual Heat Removal - IT855 a. GEN-2.1.20 (4.3/4.2)
Restore RHR Cooling ALT. ISIM Given a set of conditions, determine required actions. PATH / SHUTDOWN / NEW (SHEARED SHAFT, b. 005A2.04 (2.9/2.9) R.. 514) Affect of loss of instrument air on RCS temperature. ; l
l 0
- 5. Reactor Coolant Pump - IV403 a. GEN-2.1.32 (3.2/3.3)
Start a reactor coolant pump SIM Given a set of conditions, determine RCP start limitations. (SEAL DELTA-P IS LOST, R.164) ( ) Given a set of plant conditions, determine the effect on calorimetric.
- 6. Containment Spray - Align 31028 a. 028K5.01 (3.4/3.9):
containment spray systems SIM Given a set of plant conditions, determine hazards. (R.. 216)
- b. GEN-2.1.28 (3.2/.3.3)
Purpose of sample line heat tracing.
- 7. Pressurizer Pressure III-010 a. 010A4.03 (4.0/3.8)
Control - Place NDT SIM Given a set of conditions describe response of PRZR protection in service during a PORV block valve. natural circulation cooldown (R. 577) SHUTDOWN b. 010A1.09 (3.4/3.7) Given a set of conditions, determine tail pipe temperature.
- 8. Emergency Diesel IVf 064 a. 064A3.06 (3.3/3.4)
Generating - Unload and IN-PLANT Given a set of conditions, determine control of EDG. shutdown an EDG in the control room emergency mode b. 064K4.06 (2.8/3.2) (N.. 466) AP ACI' ION Describe when speed droop is in effect.
- 9. Auxiliary Feedwater - 5id1 a 061A2.06 (2.7/3.0)
Collapse steam voids in a IN-PLANT Affect of check-valve back-leakage. steam-bound AFW pump (N.. 935) b. 061K4.02 (4.5/4.6) Given a set of conditions, apply interlocks to control of AFW pump. )
- 10. Component Cooling #11$008 a. 008K4.01 (3.1/3.3)
Water System - Drain the CC IN-PLANT Given a set of plant conditions, deterraine response of CC l system to a LW test tank H pumps. l
- b. 008A1.01 (2.8/2.9)
Affects on CC flow from shifting common loads. Facility Author: N rrd L A Chief Examiner: as b 1 l l l l l
ES-301 Administrative Topics Outline - SET # 1 Form ES-301-1 Examination Level (Circle One): hSRO Facility: North Anna n Week of Examination: Dec. 16,1996 Examiner's Name (print): t'a gl Sba . Administrative Describe method of evaluation:
; Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions I
A.1 Status Control JPM - Enter a component into abnormal status. A.1 Shift Turnover Given a set of conditions, detennine turnover requirements. Given a set of conditions, determine qualification of relief. A.2 Maintenance JPM - Given a set of plant coaditions, priori 3ze methods of plant cooling. A.3 Radiation Work JPM - Review a Radiation Work Permit and obtain a DAD. Practices NOTE: This JPM will be incorporated into the "RCA" task, N.. 230 (JPM # 10). A.4 Emergency Given a set of conditions, determine required actions. Facilities List those facilities with protection from radiation / airborne. Facility Author: V/vnwWiln k Chief Examiner: QU NCE<c}}