ML041620450
| ML041620450 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/02/2004 |
| From: | Conicella N Public Service Enterprise Group |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 50-272/04-301, 50-311/04-301 | |
| Download: ML041620450 (145) | |
Text
ES-401 Site-Specific RO Written Examination Form ES-401-87 Cover Sheet 31 of 34 NUREG-1021, Draft Revision 9
Given the following conditions:
- Salem Unit 2 is at 90% power, performing a 5% per hour load ascension.
- OHA E-24, "ROD DEV OR SEQ" is received,
- Control Rod 2D5 indicates 200 steps withdrawn.
- - The Group Demand Counter indicates 213 steps.
1-The power ascension is put on hold, and NO action is taken for one hour.
IAW Technical Specifications, which of the following describes the consequences, if any, of continuing to operate in this configuration with NO operator action?
!Axial power distribution limits may be exceeded, but adequate SHUTDOWN MARGIN will be
'maintained as long as the deviation remains less than 18 steps.
Radial power distribution limits may be exceeded, and adequate SHUTDOWN MARGIN may be lost.
There will be NO adverse consequences as long as the rod deviation does NOT grow larger. 1
\\The Accident Analysis for a mis-aligned rod will no longer be valid, and peaking factor limits
/will be maintained.
maintained, 3) limit the potential effects of rod misalignment on associated accident analyses, will all stay true as long as the control rods are positioned within +/- 12 steps >85% power. The correct answer, B, has 2 of these conditions. Distractor A has one correct and one incorrect part, as SDM will NOT always be maintained with a rod misalignment > Tech Specs. Distractor C is incorrect because the
!tech Spec is exceeded, and the bases spells out what is maintained by keeping rod deviation within ilimits. Distractor D has one correct part and one incorrect part because peaking factors may be exceeded i
l___
Indication Systems, including:
a)
The Limiting Condition(s) for Operation b)
The Bases for the LCO(s) d)
The LCO Action Statements(s)
I I
I Tuesday, Aph 20,20048 30.22 AM-
]
I Page I O f 8 9 I
'Given the following conditions for Unit 2:
- Reactor power - 14%.
- PZR Pressure control in AUTO.
- RCS Tavg - 551°F.
- Main feed in operation using BF40s in AUTO.
- SG levels stable at 33%.
- Steam Dumps are in the MN STM PRESS CONTROL mode when 21TB30 fails full open.
- Operators successfully isolate steam flow by turning Steam Dumps OFF.
4 jFollowing isolation of steam flow, and with NO further operator action, key parameters are:
- PZR pressure lowered to 1825 psig and is beginning to rise slowly.
- RCS Tavg 542°F and rising slowly.
- SG levels peaked at 48% and are currently 40% and lowering.
- Steam pressure - 870 psig and rising slowly.
Which of the following describes the status of feedwater to the SG's?.
- Main Feedwater is isolated, Aux
.- Feedwater pumps are stopped.
,Main Feedwater is isolated, Aux Feedwater pumps are running.
'Main
..................... Feedwater is supplying
.... feedwater through
... all BF40's, Aux
. Feedwater pumps are
..... stopped..
IMain Feedwater is sumlvina feedwater through all BF40's. Aux Feedwater Dumm are runnina. 1 I
'steam flow and is now dropping, but is still above AFP auto start setpoint. Distractors c and d are incorrect because the FW INTERLOCK closed all the BF40's. and no feed is beina sumlied to SG's.
IReactor Trip Response, Basis Document a) b)
c) d)
Condenser Hotwell Makeup and Rejection Condensate Pump Low Flow Recirculation Steam Generator Feedwater Pump Recirculation (N/A NEO)
Feedwater Isolation and Feedwater Interlock (N/A NEO)
I 1
1
[_::
Facility Exam Bank IS!gfl!fi~ n! y Modified esday, April 20, 2004 8:30:24 AM---
Tuesday, April 20, 2004 8:30:24 AM 1
I Page3of89 I
Given the following conditions:
!- Salem Unit 2 is operating at 90% power.
1-Pressurizer (PZR) is 2235 psig.
- PZR Power Operated Relief Valve (PORV) 2PR1 is leaking.
- Pressurizer Relief Tank (PRT) pressure is 5 psig.
- PORV discharge temperature has stabilized at 230 deg. F.
~
i
.Which one of the following will DIRECTLY cause the indicated PORV discharge temperature to I rise?
- The PRT rupture disk develops a leak.
- PRT pressure is allowed to rise to 10 psig.
Pressurizer t........
Spray... is removed from service..
0611 112004 enough to raise PRT pressure. C. is incorrect, Pressurizer Spray has no effect on PORV leakrate or on PRT messure.
LOSS of Coolant Accident I
I Control Room location of Pressurizer and Pressurizer Relief Tank control bezels and indications e effect each Pressurizer and Pressurizer Relief Tank control has upon Pressurizer and Pressurizer Relief Tank April 21,2004 7:40:14AM 1
I Page4of89 1
IThe following plant conditions exist:
.- A small break LOCA has occurred.
- The condenser is NOT available to receive steam.
i-All S/Gs have been determined to be intact.
I-PZR level indicates zero and RVLIS indicates a bubble in the reactor vessel.
1-Reactor vessel level is decreasing slowly.
'- RCS pressure is greater than all S/G pressures.
1-ECCS is running.
- - RCPs are secured.
'- Natural Circulation cooling has stopped due to steam void in the S/G U-tubes.
j
- Which...............
of the following describes the primary paths of removing core heat?
/Boiling removing almost all heat from core, condensation of steam in U-tubes, SI flow and break flow removing heat from primary.
Natural convection cooling removing all heat from core, break flow removing all heat from Boiling removing heat from core, break flow and SI removing heat from core, S/Gs do NOT primary, S/Gs do NOT contribute due to steam void in U-tubes.
contribute due
___- to - steam
- void
__ in
___ U-tubes.
,b - SlGs db coniribute. natural circulation is not primary heat removal, c - SlGs do contribute, Loss of Coolant Accident I
I A.
B.
C.
D.
E.
Decay heal removal by natural circulation with pressurizer pressure controlling Decay heat removal by natural circulation with reactor vessel pressure controlling Decay heat removal during transition from natural circulation to core boiling mode Decay heat removal by core boiling Decay heat removal during transition from core boiling to natural circulation mode
~.................
j
~.................
... 111.......
j 1
llNPO Exam Bank I
Exam 5/30/03 Tuesday, April 20, 2004 8:30:25 AM
]
E-:
I Page5of89 I
Tuesday, April 20, 2004 8:30:27 AM 1
1 Pase6of89 1
Given the following conditions:
'- Salem Unit 2 is operating at 100% power when a catastrophic failure of RCS loop 21 cold leg piping occurs.
- RCS pressure is 35 psig.
- Initial RWST level was 41.O feet.
/Given the RWST tank curve from S2.OP-TM.ZZ-0002 TANK CAPACITY DATA, which of the
'following choices identifies the time available until the swap to Cold Leg recirc will be required?
,I3 minutes.
I!?
m!!?tes.
28minites.
iEK2.02 IPumps j/2.6*//2.7"1 IQuestion stem describes the design basis LOCA, but with power. With the RCS at 35 psig, all ECCS pumps will be injecting at their maximum rate. The flow rates used are: Charging pumps 2x550=
i1 1 OOgpm; SI 2x650= 1300gpm; RHR 2x4600= 9200; and Containment Spray pump flow of
/2x2600=5200 So, 1100+1300+9200+5200= 16,800gpm total. With the initial RWST level of 41.I'
'equating to 370,000 gallons, and 15.2' level of 150,000, you need to pump in 220,000 gallons. That's 13.09 minutes. Distracter D is the time it would take to pump in the entire RWST volume. Distracter b is the time if CS pump flow is not included.
I I
LOCAOI EO01 1
I I Loss of Coolant Accident
\\Tank Capacity Data r
Given a plot of RCS pressure and other necessary plant parameterslconditions vs. time during a LOCA, perform the following in accordance with the handout:
A.
Describe the response of the ECCS and subsystems I
RWST Tank curve, page 28 of S2.OP-TMZ-0002 1-Tuesday, April 20, 2004 8:30 27 AM
]
I Page7of89 I
iSalem Unit 2 has experienced a rupture of a RCS cold leg which has resulted in containment pressure peaking at 18 psig.
i LOCAOIE007
!With all systems actuating as expected, which of the following choices identifies the containment il
.ALL feedwater to containment to Dreclude an excessive RCS cooldown event.
Identify possible radioactivity release paths for a Loss of Coolant Accident, and describe how the actions of 2-EOP-LOCA-1 minimize the potential for a release Phase B to isolate potential injection paths to containment; Containment Ventilation to ensure
,non-essential containment ventilation penetrations are isolated.
!Main Steamline to minimize potential primary-to-secondary leakage; Feedwater to prevent uncontrolled filling of any SG.
(Phase A Isolation to ensure non-essential containment penetrations are isolated; Phase B to
/isolate additional Dotential release Daths from containment.
EK3. ] Knowledge of the reasons for the following responses as they apply to Large Break LOCA:
jEK3.061 IActuation of Phase A and B during LOCA initiation 114.3"114.3*1
'Distractor a Feedwater Isolation only isolates Main Feedwater, it does not isolate ALL feedwater. AFW
'is still available for injection to SG's. Distractor b is incorrect because Phase B isolates leakage paths, not injection paths. Distractor c is incorrect because Main Stearnline Isolation is designed to minimize and/or terminate the mass and enerav releases associated with a hiah enerav secondarv line break.
1
- - ~ -
Tuesday, April 20, 2004 8:30:27 AM
[I...--
Which of the following choices describes a combination of conditions which would result in the
'most severe damage to an operating Reactor Coolant Pump at NOP / NOT if NO operator action
'is taken?
'Total Thermal Barrier CC return flow 170 gpm and rising, any RCP seal leak-off flow = 3.8 gpm 1 and steady.
RCP motor bearing temp 180 deg F and rising, any RCP seal injection flow 3 gpm and
.lowering.
ITotal Thermal Barrier CC return flow = 155 gpm and steady, any RCP seal injection flow = 14 1
/RCP seal leak-off flow = I
.3 gpm and steady, any RCP motor winding temp 260 deg F and
/steady.
I I
/gpm and lowering.
~
jAK3. I Knowledge of the reasons for the following responses as they apply to Reactor Coolant Pump Malfunctions:
AK3.-a-j [Potential
- damage
--- -- from high winding and/or bearing - temperatures I
11 2.5][
3.1) peal injection is high, but lowering, and will not affect RCP Derformance. D is incorrect because leak-off
/is within normal range, as is RCP motor winding temp..
IREACTOR COOLANT PUMP ABNORMALITY i I 1
1
~
-~
RCPUMPEOOB L -.....................
Describe the operation of the following system as applied to S2.OP-AB.RCP-0001:
a)
Basic RCP Construction b)
Seal Injection and Seal Water Configuration a) b)
c) d)
function e)
The Control Room location of Reactor Coolant Pump control bezels and indications (N/A NEO)
The function of each Reactor Coolant Pump Control Room control and indication (N/A NEO)
The effect each Reactor Coolant Pump control has upon Reactor Coolant Pump components and operation (N/A NEO)
The plant conditions or permissives required for Reactor Coolant Pump Control Room controls to perform their intended The setpoints associated with the Reactor Coolant Pump control room alarms I
Tuesday, April 20, 2004 8:30:28 AM I
I Page9of89 1
Given the following conditions:
- Salem Unit 2 is operating at 100% power, steady state, with all controls in AUTOMATIC.
I-Auctioneered high Tave (21 loop) is 570.1 degrees.
1-22-24 loop Tave's are all 569.5 I
I !The NCO is performing a lineup to remove loop 21 Tave instrument from service for calibration.
- When the 21 Loop Tave Deviation Defeat pushbutton is depressed, charging flow will
, and PZR level will until a new steady state PZR level is achieved.
I I
1 1
rise, rise
- rise,
.......... lower........ __....................................................
lower, lower Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup:
,AK1.03 I Relationship between charging flow and PZR level lmm Auctioneered high Tave is the input for PZR level setpoint. When 21 loop is removed from this circuit, the next highest input will be used automatically, which is 0.6 degrees lower. This will cause level setpoint to lower, and the Charging Master Flow Controller to reduce output, lowering charging flow, 1
i I I b) c)
The function of each Pressurizer Pressure and Level Control system Control Room control and indication The effect each control has upon Pressurizer Pressure and Level Control components and operation i
c:::-.
i Tuesday, April 20, 2004 8:30:29 AM i
1 Pageloof89 I
Given the fo I I owi n g conditions :
1 ABCCOl E005
'..- ~~~~~~~~~~~~~~~~~~~~~~~
i
- Unit 2 is operating at 88% power.
- The crew is attempting to isolate a Component Cooling Water leak in accordance with S2.OP-1- ~
~
Z
~
~
~
i p
is operating.
i-21 and 23 Component Cooling pumps are operating.
i-22 Component Cooling pump and heat exchanger are isolated.
I - Component Cooling Surge tank level indication, LI-628A, is 37% and lowering.
1-Component Cooling Surge Tank Makeup is isolated.
'- OHA Alarm C-2 CNTMT SUMP PMP START has actuated.
I AB.CC-0001, COMPONENT COOLING I
Given a set of initial plant conditions: a) Determine the appropriate abnormal procedure.
b) c)
Describe the plant response to actions taken in the abnormal procedure.
Describe the final plant condition that is established by the abnormal procedure.
Which of the following identifies the location of the leak?
- 21 Component Cooling header.
22 Component Cooling header.
Non-safeguards header.
L _____ ____
~ - - ___ -... - _____ - - -
I J I 2. 9 j j 3. 5 1 I
L AA2.01-ILocation of a leak in the CCWS
,the leak is on either of the CC headers and non-safeguards header, then checks if any RCP OHA's alarming. In this situation, the CNTMT SUMP PP START OHA is due to the leak being on the RCP supply or return header inside containment..
Tuesday, April 20, 2004 8:30:30 AM 1
I I
Paaellof89 I
Given the following conditions:
- Salem Unit 2 is in Mode 6, with fuel movement in progress.
- Refueling cavity level is 24' over the flange.
- A large Component Cooling Water (CCW) leak has resulted in ALL CCW pumps being secured
~
due to all pumps starting to cavitate.
IAW S2.OP-AB.CC-0001 COMPONENT COOLING ABNORMALITY, which of the following actions will be performed first?
2
,Immediately suspend core alterations i ABRHRIE005 I
Send
- i.....
an operator to
... adjust Service
...... Water flow._
. controllers for
. in.
service
_____ CCW
. HX's..
i Given a set of initial plant conditions: a) Determine the appropriate abnormal procedure.
b)
Describe the plant response to actions taken in the abnormal procedure.
c)
Describe the final plant condition that is established by the abnormal procedure.
- 2.1
]:Conduct
~
Of Operations j2.1.2 1 IKnowledge of operator responsibilities during all modes of plant operation.
0004 only applies in Modes 3 and 4. Distractor d is incorrect because it will only be performed when loss of SW is the cause of entry into AB.CC. C is the correct answer because step 4.0 of the CAS states to initiate AB-RHR-1 if RHR is aligned for shutdown cooling, which it will be if fuel is in the vessel in Mode 6.
komnonent Coolina Abnormalitv I
1 L.................
.~ ~-
Tuesday, April 20, 2004 8:30:31 AM
]
I Pane 12 of 89 1
!Given the following conditions:
- Salem Unit 2 is operating at 40% power.
- All control systems are in AUTOMATIC.
- The feedback linkage on 2PS3, PZR SPRAY VALVE, fails and causes 2PS3 to fail full open.
- All attempts to close 2PS3 have failed.
IAW S2.OP-AB.PZR-0001, PRESSURIZER PRESSURE MALFUNCTION, which of the following choices identifies the actions required to be taken?
i.-
(Energize ALL PZR heaters, commence a rapid power reduction in anticipation of tripping the I
Tuesdav. ADril 20. 2004 8:30:32 AM I
Page13of89 I
iGiven the following two sets of conditions:
I
/-
i i-i I-
/-
i i
Salem Unit 1 is in Mode 3, HOT STANDBY, @ NOP, NOT.
A cooldown caused by a Steam Dump malfunction caused pressurizer level to drop to 12%.
Pressurizer pressure fell to 2185 psig before the Steam Dumps were isolated.
Pressurizer level was quickly recovered to 22%, and pressurizer heaters were returned to AUTO.
Salem Unit 1 is in Mode 3, HOT STANDBY, @ NOP, NOT.
A depressurization caused by a PORV malfunction caused pressurizer level to rise to 25%.
Pressurizer pressure fell to 21 85 psig before the PORV was isolated.
Pressurizer level was quickly recovered to 22% and pressurizer heaters remained in AUTO.
IWhich malfunction will take a longer time for pressure recovery from 21 85 psig and why?
~ pressu re.
c3 ~
@ /The PORV because the pressure reduction will be faster for the failed oDen PORV.
I 'The Steam Dump because the pressurizer heaters are less effective since they had tripped and cooled off on low PZR level.
The Steam Dump because the subcooled water insurge during refill reduced the Pressurizer liquid space temperature.
Malfunction:
1 IS inconsequen
- because pressure reduction rate does not affect recovery time. D is the correct answer because after Ithe outsurge due to lowering pressure, the insurge will be of cooler water, and will require a greater time
- to reach saturation and cause a pressure rise.
i
-1 I!?
ressurizer Pressure Malfunction I
1 I
Wednesday, April 21, 2004 7:43:42 AM 1.....
J I
Page 14of89 1
Tuesday, April 20, 2004 8:30:33 AM I
I Paae15of89 I
Given the following conditions:
1 ABNISlE001 j-Salem Unit 1 is in Mode 5.
1-Rx power is 150 cps.
1-RCS pressure is 300 psig.
1-RHR is in service, RHR HX inlet temperature is 140 degrees.
'Which of the following choices describes a condition which would cause OHA E-5, "SR DET VOLT j
iTRBL" to alarm?
~~~~~
/Loss of power to either Source Range channel "C" or "D" Gamma Metrics.
/channel 1, to alternate power supply.
~~
~
IDeenergizing 1 B 23OVAC bus without manually transferring 1 N31, Source Range detector IRemoving 1A Vital Instrument Bus Inverter from service with 1A VIB alternate source AC
'disconnected.
i
~~~~
~
I
~
~~~
Describe the operation of the following as applied to S2.OP-AB.NIS-O001(Q), in accordance with this lesson plan:
a) source range instrumentation b) intermediate range instrumentation
___ c) power range instrumentation
!Deenergizing the I B 125VDC battery bus for maintenance.
~~~~~
1 I2004 01 13 pA2. Ability to determine and interpret the following as they apply to Loss of Source Range Nuclear Instrumentation:
not have an alternate power supply. C IS correct because the power supply to N3lwill be lost if the VIB inverter and alternate AC source are not available. D is incorrect because the 1B VIB inverter will still be sutmlvinq Dower to 1 B VIB.
2.9*1 Overhead Annunciators Window E Nuclear Instrumentation System Malfunction
~~~
~ ~~~.~
er supply to the following Excore Nuclear Instrumentation System components.
L................................
Tuesday, April 20, 2004 8:30:34 AM 1
J I
Paae16of89 I
Given the following conditions:
A SGTR has occurred on 21 S/G.
1-Operators are preparing to initiate a RCS cooldown IAW step 15 of EOP-SGTR-I, STEAM
'GENERATOR TUBE RUPTURE.
- Ruptured SG pressure is 605 psig.
- Containment pressure is 0.1 psig and steady.
- answer D is the only choice less than 445 deg, and is the required CET temperature required when a
'cooldown is initiated from 600-650 psig IAW Table D of SGTR-1. Containment pressure is normal so
'would NOT use adverse containment numbers. Distractor A is the choice for 1000 psig or greater, which is where the maioritv of SGTRs will fall. Distractors B & C were picked from Table D of SGTR-1
!as realistic numbers bbt for higher SG pressures.
I L...
~-.
~
I 3, 15, 15.3, 16.3,
~
1 17.2,21, 25, and 27 -
.....-r........
J
~~~
Tuesday, April 20, 2004 8:30:34 AM^--'---.]
I Paae17of89 I
Salem Unit I has been operating at 100% steady state power for 5 days. During a board
.walkdown, the RO observes the following conditions:
SGTROlEOOl SGTROl E006 I
- - Rx power is 100.0% and stable
- I I S/G steam flow is 5% higher than feed flow.
!- Containment pressure is 0.1 psig and stable.
- - Charging flow is 96 gpm and rising slowly.
- - Pzr level is 47% and dropping.
i-Tavg is stable at 570 deg F.
iWhich 1..
of the following conditions is causing these indications?
1 List 7 symptoms of a steam generator tube rupture. For each symptom:
A.
B.
Determine if it is unique to a steam generator tube rupture.
Determine if it can be used to positively identify a ruptured steam generator Determine the indications that are monitored to ensure proper systemlcomponent operation for each step in 2-EOP-SGTR-1
]Main
~~
feed line break on 11 SG.
~
~
~
~
!Steam 1
. line
.__ rupture of 11 SG :
!Tube ruDture on I 1 SG.
i000040A203 15
[&&2,1 Ability to determine and interpret the following as they apply..........
to Steam Line Rupture:
~.
11::.
STEAM GENERATOR TUBE RUPTURE I
I Tuesday, Apri 20, 2004 8:30'35 AM --- -1 1
Page18of89 I
iGiven the following plant conditions:
LOPAOOE007 1 ABCAOIEOOI I - Salem Unit 2 is operating in MODE 4 during a mid-cycle shutdown.
- 21 RHR loop is providing Shutdown Cooling at 3,000 gpm.
- Letdown flow from RHR is 40 gpm.
- RCS pressure is 300 psig.
- 21 RHR HX inlet temperature is 260 degrees.
- All Station Air Compressors trip.
/Which of the following choices identifies the indication which would be NO oDerator action?
Unit 2 ECAC fails to start.
I Describe the EOP mitigation strategy for a loss of all AC power.
Describe the operation of the following as applied to S2.OP-AB.CA-O001(Q):
a)
Station Air Compressors b)
Emergency control Air Compressors c)
Station Air Headers d)
Control Air Headers e)
~
~~
present after 5 minutes with
~
- I - -
~
~~~
~-
~
lAll MSI 0 Atmospheric Relief Valves are failed open.
~~
~
~~~
/Letdown t __.--
flow indicates 0 gprn
~...............
121 RHR pump flow is 2,000 gpm.
121 RHR HX inlet temperature rising slowly.
~~
~
~
header, and FC on loss of air. C is incorrect because the RH20 and RH18 both are supplied only from 2A header, and BOTH fail open, so RHR pump flow would rise. Distracter D is incorrect because RHR flow has risen, and CCW flow has remained the same (CC16 is MOV)
~.
lccw Tuesday, April 20,2004 8:30:36 AM 1
1 Page 19of89 I
F -....................................
/Given the following conditions:
1-Salem Unit 2 is operating at 100% power when the entire 500KV electrical switchyard suddenly land unexpectedly becomes de-energized.
I-The reactor trips automatically.
- All EDGs start and load in Mode II.
I-SI is NOT required.
1-All other expected automatic actions occur.
!- The control room crew is recovering in 2-EOP-TRIP-2, REACTOR TRIP RESPONSE.
1 i
5 minutes after the reactor trip, which of the following choices identifies the status of the Unit 2
'Control Area Ventilation System (CAV)?
- The CAV system.....
dent P kd will be operatina in the NORMAL Mode.
'will
- - have
- ~-.
... been MANUALLY initiated
. in Accident Pressurized Mode.
will need manual alignment due to the loss of power.
-~
[B 1
06/11/2004 17 AAI.-] Ability to operate and / or monitor the following as they apply to Loss of Off-Site Power:
L__
because an initiation signal will not be generated. Distractor c is incorrect because there is no procedure step (nor reason) to place CAV in AP mode manually. Distractor d is incorrect because CAV will not be re-aliqned, nor will it need to for a loss of off-site Dower.
~
Control Area Ventilation Operation a)
Control Area Ventilation System Control Room control and indication c) operation d) intended function e)
The Control Room location of Control Area Ventilation System control bezels and indications b)
The effect each Control Area Ventilation System control has upon Control Area Ventilation System components and The plant conditions or permissives required for Control Area Ventilation System Control Room controls to perform their The setpoints associated with the Control Area Ventilation System control room alarms The function of each April 20, 2004 8:30:37
................. AM
.-........... ]
1 Page20of89 1
/Given the following conditions:
I
'- Salem Unit 2 is operating at 90% power when a loss of 2D Vital Instrument Bus (VIB) occurs.
- The Control Room crew enters S2.0P-AB.115-0004, LOSS OF 2D 11 5V VITAL INSTRUMENT BUS, and performs actions to stabilize the plant.
- Subsequently, the cause of the 2D 11 5V bus failure is identified and corrected.
I
,Which of the following choices identifies the actions that are required to be performed PRIOR to re-ienergizing 2D 11 5V VIB IAW S2.0P-AB.115-0004, LOSS OF 2D 115V VITAL INSTRUMENT
~REAC COOL PZR LEVEL PROT SIGNAL COMP.
~
~~
L /Place
-.......... 21MS10 and
............ 24MS10,
___ ATMOSPHERIC RELIEF VALVES, in MANUAL.
~
~ - -
- $ iSecure Steam Generator Blowdown IAW S2.OP-SO.GBD-0002, STEAM GENERATOR
- B LOWDO W N 0 P ERATlO N.
Place 2CA2015,CONTROL AIR SUPPLY TO CV55 BYPASS VALVE, in BYPASS.
061 7G 112004 23 18
'manipulation. Distractor c is incorrect because blowdown vlvs to main condenser fail as is, and SGDB is not required to be isolated. Distractor d is incorrect because the CV55 fails open, and manual control of CV56 allows charaina control without maniwlatina 2CA2015.
LOSS of 2D 115V Vital Instrument Bus I
r r -
I AB1 151E003 Describe in aeneral terms the actions taken in SllS2 OP-AB.115-001(Q), S1/S2.OP-AB.115-0002(Q). S1/S2 OP-AB115
. a)
A, B, c, and "ita, Bus b)
Essential Control Power System c)
Essential Lighting Power System d)
Unit 2 RMS Power System e)
SPDS Power System Tuesday, April 20, 2004 8:30:38 AM i._
I 1
Page21 of89 1
.An accidental gaseous radwaste release could cause a nuclear plant worker to exceed which of the
'following 1 OCFR20 radiation exposure dose limits?
~
The
.... dose
... to a worker inside the SITE BOUNDARY may exceed...
lthe Total Effective Dose Eauivalent (TEDE) limit of 5 rem in one vear.
.____c_.__._
ithe Total Effective Dose Eauivalent (TEDE) limit of 3000 mrem in one vear.
- a lens dose
.- equivalent to
._ the.. eye of 50 rem in one year.
~ ______.
k!d !a shallow-dose equivalent of 15 rem in one year to the skin of the whole body.
-J
~000060K102 19 I
IAKl.
- c.
I/Knowledge of the operational implications of the following concepts as they apply to Accidental Gaseous
- Radwaste Release
' 1 2.5 '3.1*1 units used for radiation intensity measurements or
-- radiation exposure levels la is correct because 5 Rem/year is the IOCFR TEDE limit. B is incorrect because it is one of the
!administrative limits at Salem. C is incorrect because the limit is 15 rem to the lens of the eye. D is incorrect because the limit is 50 rem.
/Radiation Protection Program
/Salem Offsite Dose Calculation Manual s for Category 1 and 2 Workers, as well as extension limits and requirements nistrative dose control levels for Declared Pregnant Women its for members of the general public and minors RMSOOOEOOI I State the purpose of the Radiation Monitoring System I
................... I
~
,Additional
References:
Salem Tech Specs 6.8.4.9
- RADIOLOGICAL EFFLUENT
....... RELEASES NRC POSITION PAPER HPPOS-254 PDR-9303020117
-AP.ZZ-8011, UNPLANNED 1- -Wednesday, April 21,2004 I Page22of89 1
/Given the following conditions:
I L................................
1-Salem Unit 2 is operating at 100% power steady state.
1-Service water (SW) pressure on both 21 and 22 headers drops from 105 psig to 95 psig.
'- The SW pump selected to AUTO starts and header pressures stabilize at 120 psig.
- A field operator reports a SW leak in 2C EDG room, just upstream of 23SW39, C DIESEL CLG sw VLV.
a) b)
c) operation d) intended function e)
The Control Room location of Service Water - Nuclear Header control bezels and indications The function of each Service Water - Nuclear Header Control Room control and indication The effect each Service Water - Nuclear Header control has upon Service Water - Nuclear Header components and The plant conditions or permissives required for Service Water - Nuclear Header Control Room controls to perform their The setpoints associated with the Service Water - Nuclear Header control room alarms
- Which of the following choices describes the actions that will be taken IAW S2.OP-AB.SW-0001,
- LOSS OF SERVICE WATER HEADER PRESSURE, in response to this leak?
Shut 21SW37, C DIESEL CLG SW INLET VALVE, if leak is not isolated shut 22SW37, C
- DIESEL i
CLG
-. SW INLET VALVES, declare 2C EDG
-... INOPERABLE.
Declare 2C EDG INOPERABLE, shut 21SW21 OR 22SW21, DIESEL CLG SW INLET
'VALVES, I -
to isolate the leak : ___....
'Lock out 2C EDG, isolate the leak by shutting 21SW37 AND 22SW37, C DIESEL CLG SW
!INLET VALVES.
~.......
20 Per step 3.150 of the AB, lock out the EDG(s) that will be affected and isolate the leak. The only way to isolate the leak is to isolate BOTH supplies from both SW headers bv closina BOTH SW37's. lsolatina "
just one header with a SW37 will allow.the other header to continue supplyi@ the leak. Isolating the SW2l's will not stop the leak as the other header will continue supplying. EDG must be locked out to
- prevent starting with no SW available.
Service Water Nuclear dwg Tuesday, April 20, 2004 8:30:40 AM I
J 1
Page23of89 1
'Given the following conditions:
- Salem Unit 2 has received a FIRE alarm for Zone 69, Fuel Handling Building (FHB).
- An operator in the area reports a fire in a bin of Protective Clothing in the FHB truck bay Which of the following choices identifies the actions required IAW S2.OP-AB.FIRE-0001, CONTROL ROOM FIRE RESPONSE?
Place Control Room Ventilation in FIRE OUTSIDE CONTROL AREA on Unit 2 ONLY, and secure Unit 2 FHB supply fans ONLY.
Place Control Room Ventilation in FIRE INSIDE CONTROL AREA on both Unit 2 and Unit 1, Place Control Room Ventilation in F k E OUTSIDE CONTROL AREA on both Unit 2 and Unit and
- ONLY
- secure ALL FHB supply
- - - fans.
I, and secure
_ _ ALL Unit 2 FHB supply - and
___ ___ exhaust fans - ___ ONLY.
k!!
Place ControlRoom Ventilation in FIRE INSIDE CONTROL AREA on Unit 1 ONLY, and 06/11/200 000067A10.5 2
[&ll,_ljAbility I
to operate and /._____
or..
monitor
... the
..... following as. they apply
....... to Plant
.... Fire on Site:
retted to have for a fire in the FHB, Zone 69, the
'ouerator is directed to secure ALL FHB su~ulv and exhaust fans.
IC ontrol Room Fire Response
~
FIRpROEOi2 Identify and describe the Control Room control ed with the Fire Protection System, includinq A:
The Control Room location of Fire Protection System control bezels and indications (N/A NEO)
B. The function of each Fire Protection System Control Room control and indication (NIA NEO)
C.
The effect each Fire Protection System control has upon Fire Protection System components and operation (N/A D. The plant conditions or permissives required for Fire Protection System Control Room controls to perform their NEO) intended function precautions associated with each operating procedure which are required to be considered by either Licensed or Non-Licensed Operators Tuesday, April 20, 2004 8:30:41 AM I --...
~ - _ _
J 1
Paae24of89 1
000068K201 22 J
.zI/Knowledge of the interrelations between Control Room Evacuation and the following:
3.9; 4 o AFW pumps are not controlled from HSD panel, nor do Boric Acid Storage Tank levels have indication at HSD panel. The correct answer a contains 3 items available on the hot shutdown panel. Distractor b ABCROI EO01 I...............
'is wrong because it contains BAST level. Distractor c is incorrect because it contains AFW pump
'control switches. Distractor d is incorrect because it contains AFW pump control switches, lis CC flow out of 22 RHR HX, and feeds F1601 D which id on HSD Panel.
FE601 B Describe the operation of the following as applied to S2.OP-AB.CR-0001 (a):
A.
Hot Shutdown Panel ot-Shutdown Station Panel 213-1 drawin 1
I I
Page25of89 I
Given the following conditions:
i-Salem Unit 1 is operating at 100% power.
i-All systems are aligned normally for 100% power.
- Which of the following choices, when taken by itself, identifies a condition which will NOT
- automatically close the 1 CV3, Letdown 45 GPM orifice Isolation Valve?
[Loss i
____ of
____ 28VDC
... control voltage.
~
I I
~
06/11/2004 23
, A D. ]:Knowledge of the interrelations between Control Room Evacuation and the following:
ILogic Drawing
~..
Hot Shutbown Panel
]
Tuesday, April 20, 2004 8:30:42 AM I
Page26of89 I
iGiven the following conditions:
1-Salem Unit 2 has experienced a SBLOCA.
/- Operators have transitioned from 2-EOP-TRIP-1 REACTOR TRIP RESPONSE, to 2-EOP-LOCA-16 LOCA OUTSIDE CONTAINMENT.
' The 21SJ49 COLD LEG ISOLATION VALVE has just been closed in an attempt to isolate the I
I-jleak.
I loutside Containment).
Distractor a is incorrect because after the leak is isolated by closing the 21SJ49, the procedure leaves it closed. Also stable RCS pressure is not the required RCS pressure indicating leak is isolated. B is
!correct because rising RCS pressure indicates leak isolation, and the next step is to stop 21 RHR pump
'because the RHR system is split and there is no discharge path. Distractor c is incorrect because the i21RH19 would already have been closed at step 2. Distractor d is incorrect because both stable lPE3!.!?a?d stePP~nstheoPPos~te-!~oP-
_R_t.iRPu-mP-a~e-n~-correc~~
1..
Tuesday. Auril 20, 2004 8:30:43 AM 1
I Page27of89 I
I,Given the following conditions:
FRHSOOE004
- - A loss of heat sink has occurred.
'- The operating crew is establishing RCS Bleed and Feed in accordance with EOP-FRHS-1, Loss
'Of Secondary Heat Sink.
~
The RO opens one PORV. He reports that the other PORV will NOT open.
i /Which L-..
one of the foll i_
any consequences of the PORV failure?
CS must be terminated and sec depressurization to
!inject condensate pump flow must be immediately initiated.
Explain the effectivenesslineffectiveness of Safety Injection flow in mitigating the consequences of a loss of heat sink event, with no other operator action
~
'ALL L
.... SGs will
.. require
.__ depressurization to inject the alternate source of feedwater The RCS will rapidly re-pressurize when the SGs empty, resulting in a violation of the RCS
.Safety L............
Limit.
[The RCS may not depressurize quickly enough to ensure sufficient SI flow to provide RCS
!heat removal, and other RCS openings may have to be established.
i
[Salem I 2 06/11/2004
-1 00WE05K101 ibecause SI flow may NOT be adequate at the PORV setpoint. Distractor a is incorrect because action
/to align condensate pumps is already taken, and not as a contingency to Bleed and Feed. D is correct ibecause FRHS Basis document describes on page 33 the consequences of not having both PORV's open, and it is D.
Response to Loss of Secondary Heat Sink E
y Modified Beaver Valley NRC Exam 12/1/2002
[::--_Tuesday,
. April 20, 2004 8:30:44 AM I
Page28of89 j
~.___
- Given the following conditions:
1-A LOCA has occurred during a cooldown while in MODE 3.
1-i Operators have transitioned out of 2-EOP-TRIP-1.
- RCS pressure is 125 psig.
I-RCS Core Exit TCs read 380 deg F.
- RCS Cold Leg temperatures are 250 deg F.
- RCS has cooled down 125 degrees in the last 30 minutes.
- 22 RHR pump failed to start.
- 21 RHR Pump is running providing I 1 50 gpm flow.
/What is the required action taken in response to the above conditions?
Entry into 2-EOP-FRTS-1 RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS is...
I
~ _ _ _
~~
[made and a RCS temperature soak for a ONE hour period will be completed.
made but NO actions are imDlemented before returning to procedure in effect.
~
~..
@ !NOT required since RCS pressure is below 350 psig.
NOT required since S2.0P-AB.LOCA-0001 SHUTDOWN LOCA will address any Thermal ishock concerns.
1
/E.K3._] Knowledge of the reasons for the following responses as they apply to Pressurized Thermal Shock:
of temperature, pressure, and reactivity changes and operating limitations and reasons for perating characteristics.
ing characteristics during transient conditions, including coolant chemistry and the
' 3.4, 3 9'
~
B is the correct answer because entry is made to FRTS due to PURPLE PATH on FRTS from >IO0 deg cooldown and Tc's 280. Upon entry to FRTS-1, steps 1 and 1.1 establish that RCS pressure IS < 300 psig and RHR flow is > 300 gpm on either RHR train, and directs operator back to procedure in effect
/The stem conditions tell the operator that the EOP's are still in effect, so the distracter d is incorrect IDistractor c is incorrect because entry to FRTS will be made, THEN RCS pressure is evaluated
[Distractor a is incorrect because no soak is required prior to returning t procedure in effect.
IResponse to Imminent Pressurized Thermal Shock Conditions pritical Safety Function Status Trees L
Modified I
Page29of89 I
!Which of the following choices identifies the maximum cooldown rate allowed in EOP-TRIP-5,
!NATURAL CIRCULATION RAPID COOLDOWN WITHOUT RVLIS, after the initial cooldown to 500
'degrees is performed?
k00 degrees / hr.
1100 degrees / hr.
06/11 /2004
[OOWEI 06222 27
[Emergency and Abnormal Plant Evolutions ml Ell
/E101 !Natural Circulation with Steam Void in Vessel withhithout RVLIS i2.2 I Equipment Control
,2.2.22 ' /Knowledge of limiting conditions for operations and safety limits.
1 3 4 4 1 The Tech Spec LCO for RCS cooldown is I00 degrees per hour (3.4.10.1). This is reflected in the
/maximum cooldown rate in TRIP-5. Distracter a is the limit for PZR c/d. Distracter c is the limit for initial 1
c/d to 500 degrees : Distracter d is applicable in non-EOP procedures I
TRP004E005 Determine the basis for each step, caution. note or continuous action steD relative to a Natural Circulation Cooldown Tuesday, April 20, 2004 8:30:46 AM I
. I Page30of89 I
!Given the following conditions:
A Main Generator trip causes an automatic Main Turbine trip. With power >P-9 (49%), a turbine trip causes a Rx trip. Since the stem identifies that the Rx did not trip, an A M
has occurred Distractor a is incorrect because while the first part of answer is correct, the stabilization of power during an ATWT is incorrect. Distracter b is incorrect because the generator trip will cause a Turbine trip.
Distractor d is incorrect because rods will insert in AUTO at 72 spm due to the loss of load. C is correct because
- Salem Unit 1 is operating at 100% power, EOL, with all control systems in AUTO.
- A 500KV breaker failure in the switchyard causes the Main Generator to trip.
- The reactor does NOT automatically trip.
- All other AUTOMATIC actions occur as designed.
i FRSMOOE004
/
FRSMOOTOOI Which of the following choices identifies the plant response, and the required response by the control room crew?
/A Main Generator trip does NOT cause an automatic Rx trip. The crew will stabilize power <
149-% and determine why the main turbine failed to trip.
/An-OT delta T turbine runback will occur, and will continue until the Main Turbine is off line
/The crew will manually insert control rods when rod speed falls below 48 spm.
/Control rods insert in AUTO at 72 steps per minute, the crew manually attempts to trip the 1 reactor.
ntrol rods must be inserted MANUALLY, and the Main Turbine must be manually tripped.
~
IComprehension 1
Salem 1 & 2 06/11/2004 E
I Determine the indications that are monitored to ensure proper systemlcomponent operation for each step in:
A.
B.
C.
Shutdown Margin Status Tree RESPONSE TO NUCLEAR POWER GENERATION.
RESPONSE TO A LOSS OF CORE SHUTDOWN.
Given a set of plant conditions, perform actions for a Response To Nuclear Power Generation in accordance with 2-EOP-FRSM-1 28 I
2004 7:02:48 AM 1
[
Page31 of89 I
Given the following conditions:
- Salem Unit 2 is in MODE 6.
- The Reactor vessel head is removed and on its storage stand.
- The Refueling Cavity is being filled from the RWST IAW S2.OP-S0.SF-0003, FILLING THE REF U ELI N G CAVITY.
- Refueling cavity water level is 1 I O ' and rising.
Which of the following choices identifies the indications which will be present on the Control Room Console?
PZR Cold Calibrated level is 10%: RWST level is 27'.
PZR Cold Calibrated level is off-scale low; RWST level is 40'.
~--.
PZR Cold Calibrated level is off-scale high RWST level is IO".
' T l, A b i l i t y to predict and/or monitor changes in parameters associated with operating the Reactor Coolant System controls including:
.AI L...............I I J i Relative level indications in the RWST, the refueling cavity, the PZR and the reactor vessel during
~
2.7 3.2 IDreDaration for refuelina I
al operating in RWST will
- be 253,000 gallons, which will be -27.5'. The corresponding.PZR cold cal level will be > the 0% level at
,the 108' 11" elevation in containment. This question does not require memorization of tank levels, but rather an understanding of the physical connections and relative elevations of the systems I
]Draininn the Reactor Coolant Svstem
. \\
- A 9 '
. I *
--i---i RCSOOOE008 PZRP&LE008 Identify and describe the Control Room controls, indications, and alarms associated with the Reactor Coolant System including:
a) b)
c) d)
function (N/A NEO)
Identify and describe the Control Room controls and indications associa system, including:
a) b)
c) d)
uerform their intended function The Control Room location of Reactor Coolant System control bezels and indications (N/A NEO)
The function of each Reactor Coolant System Control Room control and indication (N/A NEO)
The effect each Reactor Coolant System control has upon Reactor Coolant System components and operation (N/A NEO)
The plant conditions or permissives required for Reactor Coolant System Control Room controls to perform their intended
... associated with the Reactor Coolant System Control Room alarms Pressure and Level Control The Control Room location of Pressurizer Pressure and Level Control system bezels and indications The function of each Pressurizer Pressure and Level Control system Control Room control and indication The effect each control has upon Pressurizer Pressure and Level Control components and operation The plant conditions or permissives required for Pressurizer Pressure and Level Control system Control Room controls to e)
The setpoints associated with the Pressurizer Pressure and Level Control system control room alarms
.... E I
Page32of89 I
[---- Tuesday, April 20, 2004 8:30:48 AM I
I Paae33of89 I
IGiven the following conditions:
- Salem Unit 2 is operating at 22% power.
- The Main Generator is supplying 130 Mwe.
- Main Steam Dumps are in Tavg control.
- 24 RCP shaft sheared completely.
- 24 RCP breaker did NOT trip.
,With NO operator action, which of the following choices describes the plant condition one minute after the event?
The reactor did NOT automatically trip because power is c P-8, and 24 SG pressure will be lower than the remaining SG's.
The reactor did NOT automatically trip because power is c P-8, and 24 SG pressure will be higher than the remaining SG's.
The reactor tripped automatically on 2/4 Low RCS Flow on 1/4 RCS loops > P-IO, and 24 SG pressure will
____. be higher than the remaining
.- SG's.
The reactor tripped automatically on 2/3 Low RCS Flow on 1/4 RCS loops > P-IO, and 24 SG pressure will be lower than the remaining SG's.
06/11/2004 002000K607 30 llowing will have on the Reactor Coolant 1 2 5 2 8 '
or 1/4 RCP bkrs open will cause a Rx trip >P-IO ressure will lower, as its D/T lowers. 24 loop will rise. Tc's n the 3 remaining loops will lower,
[Steam Dump System Operation RCPUMPEOOI I State the numose of the Reactor Coolant PumDs
[Dieit From Source 1.....................................................
Tuesday, April 20, 2004 8:30:48 AM 1
_J I
Paae34of89 I
I
~ \\Given the following conditions:
- Salem Unit 1 is operating at 100% power steady state.
- All control systems are in AUTOMATIC.
- 12 SGFP trips on low lube oil pressure.
- The Main turbine runs back to 65%.
- With NO operator action, which of the following choices describes the RCP Seal Injection flow one
'minute after the turbine runback is completed?
[Seal injection flow will rise due to lCV55, CENT CHG PMP FLOW CONT VALVE, modulating
/in I the open direction.
~
!Seal injection flow will lower due to 1CV71, CHG HDR PCV, modulating in the open direction.
ISeal injection flow will rise
__ ___ due to 1CV71, CHG HDR PCV, modulating in the closed direction.
Seal injection flow will lower due to 1CV55, CENT CHG PMP FLOW CONT VALVE, modulating in the closed direction.
31
/A3.:3_.]:Ability to monitor automatic operations of the Reactor Coolant Pump System including:
back is complete, programmed PZR level will be less than actual. This in the closed direction to lower PZR level. With the CV71 having no automatic feature to modulate, the closing of the CV55 will lower the charging header pressure, and cause less flow to be directed to the RCP seals. Distractor a is incorrect because seal injection flow will not rise. Distractors b and c are incorrect because CV71 will not move without operator action
[Charging, Letdown, and Seal Injection I-i..............
.. 1 I
I Identify and describe the Control Room controls, indications, and alarms associated with the Reactor Coolant Pump, including:
a) b)
c) d)
function The Control Room location of Reactor Coolant Pump control bezels and indications (N/A NEO)
The function of each Reactor Coolant Pump Control Room control and indication (NIA NEO)
The effect each Reactor Coolant Pump control has upon Reactor Coolant Pump components and operation (NIA NEO)
The plant conditions or permissives required for Reactor Coolant Pump Control Room controls to perform their intended rCoo!~nt..Pum~~~ontrol_roo Tuesday, April 20, 2004 8:30:49 AM I
J I
Page35of89 I
IGiven the following conditions:
i-Salem Unit 2 is in MODE 5.
- - 500 KV switchyard is aligned for normal operation with Unit 2 Main Generator Drops removed.
i-RHR HX inlet temperature is 150 deg F.
i-All RCS Tc are - 130 deg F.
j-All SGs are filed to 80% WR level.
j-RCS pressure is 290 psig.
All RCP's are secured.
- POPS is in service.
[
Which of the following choices describes a condition where it is NOT permissible to start 23 RCP ilAW L.
S2.OP-SO.RCP-001 REACTOR COOLANT...............
PUMP OPERATION, and why?
iH and E 4KV group busses are powered from # I Station Power Transformer (SPT), and 2A i4KV vital bus is powered from #24 SPT, to prevent 2A vital bus second level under voltage
/protection i
actuation.
.for available mass release required for 2 RCPs in operation.
'PZR level of 100% with SG secondary side temperatures at 140 deg F, to prevent exceeding
,RCS I
heat-up rate IAW LCO 3.4.10.1.a PZR level of 20%, and SG secondary side temperature of 80 deg. F, to prevent a RCS
,~ressure transient and PZR heater isolation.
003000K402 c31 7
32
- K4.
_______ //Knowledge of Reactor Coolant Pump System design feature(s) and or interlock(s) which provide for the lfollowing:
K4-. o2 -_-, p _-----__-
revention of cold water accidents or transients I 2 5 27*1 D is the correct answer because P&L 3.2.9 of procedure directs PZR level to be as high as possible when SG secondary side temperature is lower than RCS temp to minimize the RCS pressure transient caused by the secondary system heat sink and letdown isolation for heater protection. PZR level of 20% is just above heater cutout at 17%. Distractor a is wrong because the E group bus and any 4KV vital or CW bus are not allowed to be powered from the same 500/13kv transformer. The distracter has
'H,E powered from ISPT, and 2A 4kv vital powered from 24 SPT which receives it's power from 3 SPT under normal conditions, which were set forth in the stem. Reason is correct. Distractor b IS incorrect because there is a 2 RCP operation limit below 200 deg, with correct reason. Distractor c is incorrect because with RCS Tc < 312 either PZR level has to be less than 92% OR ALL SG secondary side temps have to be 4 0 degrees above RCS Tc. In this distracter, all SG secondary side temps are 10 I
I. Reactor Coolant Pump
_. Operation.-
prerequisites and precautions associated with each operating procedure which are required to be considered by e>her Licensed or Non-Licensed Ooerators
~. ~..
Tuesday, April 20, 2004 8:30:50 AM 1
1 Page36of89 I
r I
I Tuesday, April : 20,2004 8:30:51
.- AM
~
....... 1 I
Page37of89 1
- An electrical fault causes Auxiliary Building MCC I C West Valves and Misc. 230V Vital Control
'Center Motor Control Center (MCC-1 CY2AX) to become deenergized.
I iWith NO operator action, which of the following choices identifies a concern with this MCC de-energized while
_.. at 100% power?
~
BF22, MAIN FEEDWATER STOP CHECK VLV, excessive valve disc erosion due to ireleasina the valve into the feed flow.
Rundown of the 1A 28VDC batterv due to loss of 1A2 28V batterv charuer.
,Having to declare 12 Boric Acid Storage Tank INOPERABLE at 63 deg. F due to the loss of itank heaters.
'Lowering air pressure in the 1 C EDG Air Storage Reservoirs due to the loss of both 1 C EDG
/air system air compressors.
I 0611 1/2004 00K204 33 2.6 2.7 ers. C is correct I 'of
. time would result in lowering temp. of boric acid in the tank :
oss of 1 C 460/230 Vital Bus IC...........................
Describe the function and operating characteristics for the following Chemical and Volume Control System components:
a)
LetdownlCharging i) Letdown lsolaiton Valves, CV2, CV277 ii) Regenerative Heat Exchanger iii) Letdown Orifices iv) Letdown Orifice Isolation Valves, CV3, CV4, CV5 v)
Letdown Releif Valve, CV6 vi) Letdown Line Containment Isolation Valve, CV7 vii) RHR Flow Control Valve, CV8 viii) Letdown Heat Exchanger ix) Low Pressure Letdown Control Valve, CVl8 x) Temperature Control Valve, CV21 xi) Demineralizers (Mixed Bed, Cation, and Deborating xii) Inlet Valve to Deborating Demin, CV27 xiii) Reactor Coolant Filter xiv) Diversion Valve, CV35 xv) CVCS Holdup Tanks xvi) Volume Control Tank xvii) VCT Isolation Valves, CV40, CV41 xviii)
Chemical Mixing Tank xix) Charging Pumps (Centrifugal and PD) xx) Miniflow Recirc. Valves, CV139, CV140 xxi) Seal pressure Control Valve, CV71 xxii) Chg. Line Containment Isol. Valves, CV68, CV69 xxiii) xxiv) PZR Auxiliary Spray Valve, CV75 xxv) CCP Flow Control Valve, CV55 i) Seal Water Injection Filters ii) Seal Bypass Flow Valve, CV114 iii) Seal Water Return Isolation Valve, CV104 Charging to Loop 3 Valve, CV77, Loop 4 Valve, CV79 b)
RCP Seal Water r p a g e 38 Of 89 I
1
iv) Seal Water Return Relief Valve, CV115 v) Seal Return Cont. Isol. Valves, CV116, CV284 vi) Seal Return Filter vii) Seal Water Heat Exchanger Excess letdown i) ii) Excess Letdown Heat Exchanger iii) Excess letdown Flow Cotrol Valve, CV132 iv) Excess Letdown Diversion Valve, CVI 34 Makeup i)
Primary Water Storage Tank ii) Primary Water Makeup Pumps iii) Boric Acid Batch Tank iv) Boric Acid Tanks v)
Boric Acid Transfer Pumps vi) Boric Acid Filter vii) Boric Acid Blender viii) Primary Water Flow Control Valve, CV179 ix)
Boric Acid Flow Control Valve, CV172 x)
Charging Pump Suction Valve, CV185 xi) VCT Makeup Isolation Valve, CV181 Excess Letdown Isolation Valves, CV278, CV131 xii) Rapid Borate Stop Valve, CV175 I
Page39of89 I
/Given the following conditions :
I-Unit 1 is operating at 100% power.
i-11 charging pump is in service.
j - An AUTO make-up to the VCT occurs at 14%, and terminates at 24%.
i PUMPSOEOO3
' PUMPSOE007 I..-.
With NO operator action, which of the following choices describes the CVCS system response?
iVCT pressure will I
~~
, and the charging pump will operate cavitation.
I I
lower, closer to.
06/11/2004 34 b) pump head terms c) pump lift terms d)
Describe cavitation, including symptoms, effects on centrifugal pump operation and methods of prevention.
Describe the operation of centrifugal pumps, including requirements for:
a) b)
c) d)
minimum flow net positive suction head (NPSH) net positive suction head (NPSH) starting and venting a centrifugal pump operation with a centrifugal pump dead headed K5.
1 Knowledge of the operational implications of the following concepts as they apply to the Chemical and olume Control System' I
from cavitation.
1 Page40of89 1
/Given the following conditions:
L Unit 2 has experienced a Large Break LOCA.
!- 2-EOP-LOCA-3, "TRANSFER TO COLD LEG RECIRCULATION" is complete with NO
- abnormalities encountered.
i-Operators are currently at step 26, "Preparation for Hot Leg Recirc", of 2-EOP-LOCA-1, "LOSS iOF REACTOR COOLANT".
1-Off-site power is supplying all 4KV Vital busses.
I
~
h: 121 and 22 SI pumps would begin to cavitate.
122 Containment Spray Pump would
___ lose NPSH.
!Containment Spray flow would drop to zero.
LCA3U2TOOl LOCAOI E008 100 500 0 K305 L----
35 Ell 7-71 1
icI/Knowledge of the effect that a loss or malfunction of the Residual Heat Removal System will have~on the Given a set of plant conditions, perform actions for a Transfer to Cold Leg Recirculation in accordance with 2-EOP-LOCA-3 Determine the indications that are monitored to ensure proper systemlcomponent operation for each step in 2-EOP-LOCA-1 ifollowing:
K3.05.~
IECCS 137* 38*1 Distracter c is incorrect because both CS pps would be secured by the time LOCA-3 was completed.
D is the correct answer. LOCA-3 explicitly states that if BOTH RHR pumps are operating, then 22CS36 is opened to supply containment spray from 22 RHR pp discharge.
Distracter a is incorrect because the 21 and 22RH19's are closed in LOCA-3 to prevent runout conditions on the operating pump if the other RHR pp were to trip will in cold leg recirc. Distracter b is incorrect because NPSH would still be suoolied bv 21 RHR DO r r - - - :.--r.=.:
m Source Wednesday, April 21,2004 7:50:37 AM 1 I.__-.__ :
I Page41 of89 1
~
,Given the following conditions:
I ECCSOOE006 1
j-Salem Unit 2 has initiated a Rx trip and Safety Injection from 100% power due to a LOCA.
- RWST level has lowered to 15.2'.
Describe the interlocks associated with the following Emergency Core Cooling System components:
a) b)
c) d)
e) 9 g)
VCT Outlet Valves - CV40/41 SI Pump Mini-Flow Isolation Valves - SJ67168 RHR Hot Leg Suction Valves - RH112 RHR Pump Suction Isolation Valves - RH4s Containment Sump Suction Valves - SJ44s RHR Discharge to ChargingISI Pumps - SJ45s RHR Discharge to Containment Spray - CS36s
/With NO operator action, which of the following choices contains ONLY automatic actions that will loccur i__
with the current RWST level?
C16 RH T opens, 22SJ44 CONT opens.
/21SJ44 opens, 21 RH4 RHR PUMP SUCT VALVE shuts.
I22CC16 opens, 22SJ113 SI CHG PUMP X-OVER VALVE opens.
i21 SJI 13 SI CHG PUMP X-OVER VALVE opens when 21 RH4 is fully shut.
in LOCA-3. Since the stem of the question says that the crew is transitioning to LOCA-3, they will not yet have ARMED the SJ44's. The 21/22RH4's are interlocked to operate off of the SJ44's being full
,open, before they will start to close. The only answer which contains both a CC16 and a SJ113 is c, all 1
Tuesday, April 20, 2004 I
Page42of89 I
During a normal Unit I plant heatup and pressurization from Mode 5, the following conditions exists:
RCS Tcold is 150 F.
RCS pressure is 320 psig.
POPS is in service RCS heatup rate is 40 F per hour.
All reactor coolant loops are operable, but only one RCP is running.
The RHR system is aligned for shutdown cooling with 1 I RHR pump 12 SI pump is OPERABLE.
11 charging pump is in service running.
i 'The conditions described are IMPROPER because:
The number of ECCS pumps available to provide injection is inadequate.
'Running
....... -. one
.. RCP and
- one. RHR.._______
pump produces non-uniform core._
cooling..-.
- The heatup rate is too high for the RCS temperature and pressure.
Ilf
&___ the SI pump were
.......... to start, it might overpressurize the RCS.
00K505 37 Knowledge of the operational implications of the following concepts as they apply to the Emergency Core Cooling System:
3 4 38'
/OPERATION. Distractor c is incorrect because the HU rate is less than 100 deg / hour allowed in ITSAS 3.4.10.1.a. D is correct because only 1 charging/SI pp is allowed to be OPERABLE with loop Tc L-.-
I I
ergency Core Cooling System, including:
a) b)
c) d)
The Limiting Condition@) for Operation (N/A NEO)
The applicability of the LCO(s)
The LCO Action Statement@) (N/A NEO)
The Bases for the LCO(s)
I Modified Tuesday, April 20, 2004 8:30:55 AM I
_1
\\
Page43of89 1
- Given the following conditions
- Salem Unit 2 is operating at 100% power.
- 21 and 22 CCW pumps are running in MANUAL.
- 23 CCW pump is stopped and in LOCAL-MANUAL at the Hot Shutdown Panel.
- A breaker failure relay causes a loss of off-site power.
- All EDG's start and load correctly.
- Which one of the following choices describes the status of the Component Cooling Water pumps 1
!minute after the loss of off-site power?
~~
!21
_____ __ and
__._ 22 CCW
. pumps
_____ are running
,23
....... CCW. pump is stopped.
~NO CCW pumps are running, all CCW pumps swap to MANUAL.
'All
........... CCW __ pumps are running, all CCW pumps swap to MANUAL.
a 21 and 22 CCW pps swap to AUTO, all CCW pumps are running.
06/11/2004 OOOA401 38 I
w1 Ability to manually operate and/or monitor in the control room:
lA4.01 1 hCW indications and controls 3 3 3 1
,selected to local manual at the HSD panel, it will still receive a start signal and manual control lockout Logic Drawings 1 -
Tuesday, April 20, 2004 8:30:56 AM
]
I::.: __ -.-
I PageMof89 I
- SSPS testing and troubleshooting are in progress.
- A Phase B Containment Isolation signal is generated and all related valves closed.
Which of the following describes the required operator actions IAW S2.OP-AB.RCP-001 REACTOR COOLANT PUMP ABNORMALITY?
/Restore CCW to the thermal barrier within five minutes or initiate a MANUAL reactor trip and istop all RCP's.
Initiate a MANUAL reactor trip and stop all RCP's. 3-5 minutes later close all CV104 SEAL LEAKOFF valves.
- Trip the Main Turbine, insert control rods to achieve 5% power, open the Rx trip bkrs, stop
.ALL RCP's.
'Initiate a MANUAL reactor triD and stoD all RCP's ONLY.
06/11/2004 008000K302 39 K3, ___I Knowledge of the effect that a loss or malfunction of the Component Cooling Water System will have on the following:
procedure does not allow 5 minutes for restoration. Distracter b is incorrect because closing CV104's is not required for loss of CCW. Distracter c is incorrect because the procedure directs a reactor trip, since the RCP's must be stomed.
/Reactor Coolant Pump Abnormality I
Tuesday. ADril20. 2004 8:30:56 AM I
I Paae45of89 I
$Given the following conditions for Unit 2:
1-Pzr pressure - 2235 psig.
1-RCS temperature - 547 deg F.
I-Channel Ill (PT-457) has been selected as controlling channel.
j-Master Pressure Controller is in AUTO.
- Which one of the following correctly describes Pressurizer pressure response if PT-457 fails LOW
- and NO operator action is taken?
I i
j
,Pressurizer pressure will rise until...
IONE Pressurizer Code Safety lifts.
IONE Pressurizer PORV opens.
/BOTH Pressurizer Spray
........... valves open.
I 06/11/2004 inputs from channels 1,3 and 2,4 respectively for 2PR1 and 2PR2. Since channel 3 has failed low, and the coincidence for the PORV opening is 2/2, only 2PR2 will see 2 channels of pressure at 2335 psig, and open to control any further pressure rise. 2PR1 will remain closed because one of its inputs, 40 1
PZRP&LE007 1..................
i PZRP&LEOOS
/
Identify and describe the local controls and indications associated with the Pressurizer Pressure and Level Control system, including:
a) b)
c) their intended function d)
State the setpoints for automatic actuations associated with the Pressurizer Pressure and Level Control system The location of Pressurizer Pressure and Level Control system local controls and indications The function of Pressurizer Pressure and Level Control system local controls and indications The plant conditions or permissives required for Pressurizer Pressure and Level Control system local controls to perform The setpoints associated with the Pressurizer Pressure and Level Control system local alarm-Not applicable to this lesson 1PR1,PR2 PZR Power Relief Valves
[RPS PZR Pressure
. and Level Control I
I
.... Tuesday, April 20, 2004 8:30:57 AM
]
I Page46of89 1
jG iven the following conditions :
- Salem Unit 2 is operating at 100% steady state power, with all systems in AUTOMATIC.
- The output of the Pressurizer Master Pressure Controller to fail to 0%.
i IWhat effect will this have on PZR heaters and spray valves?
B/U heaters energize, and spray valves open.
IB/U heaters remain off, and spray valves open.
i
- B/U heaters energize, and spray valves remain closed.
iB/U heaters remain off, and spray valves remain closed.
06/11 /2004 010000K603 41
~-~
- Pressure L _____
Control System:
'K6.
m m
3.21 The MPC output varies from 0-1 OO%, with 0% demand acting as if PZR pressure is too low, and will energize B/U heaters and send a close demand air signal to the spray valves. C is the correct answe because it contains the correct actions that will take place. All the distracters are wrong combinations I sDrav and heaters.
/Knowledge of the of the effect of a loss or malfunction on the following will have on the Pressurizer 3.6' r
of I
..a.....................
I_. Pressurizer Pressure Control System Operation IPZR pressure and level control PZRP&LEOO8 I Identify and describe the Control Room controls and indications associated with the Pressurizer Pressure and Level Control system, including:
7 a)
The Control Room location of Pressurizer Pressure and Level Control svstem bezels and indications bj c) d)
Derform their intended function The function of each Pressurizer Pressure and Level Control system Cdntrol Room control and indication The effect each control has upon Pressurizer Pressure and Level Control components and operation The plant conditions or permissives required for Pressurizer Pressure and Level Control system Control Room controls to Tuesdav. Aoril20. 2004 8:30:58 AM 1
I Page47of89 I
~
i
- During a total loss of off-site and on-site AC power, how are safeguards valves prevented from
'automatically repositioning upon. restoration
.....__ of AC power?...............
- I25 VDC Vital batteries supply the Vital Instrument Busses, which allow SI circuitry actuation
.and reset capabilities from SSPS.
from repositioning if a different (SI, Phase A, Phase B, CVI) signal calls for movement when power IS restored. Distracter c is incorrect because the S signal is always locked in following a SI. It would 28 VDC Vital batteries maintain control power to valve control circuits, which keep the valves from moving without a control board manipulation.
restored.
'The Reset "S" signal is locked in to prevent valves from moving when 230 VAC power is
~D repressing the CLOSE PB for all Phase A isolation valves on the main control board.
of the physical connections and/or cause-effect relationships between Reactor Protection 3.4 ~ 3.7 All AC Po to reset SI. This actuation and reset prevents valves from ted to initiate or verify
~~
~
~ _ _
~LOSS of all AC Power i.........................
i
[--. Tuesday, April 20, I
Pase48of89 I
'What would be the effect on the Reactor Protection System if the 2B Vital Instrument Bus were to
- become deenergized with the unit at 100% power?
12RP4 bistable lights flashing for all channel II indications due to train disagreement between
'SSPS Trains A and B.
'OHA A-34 SSPS TRN A TRBL in alarm due to loss of 1 of 2 45VDC power supplies to Train A logic cabinet.
'SSPS Train B slave relays would not actuate on a Safety Injection signal.
'Logic coincidence for Containment Spray actuation would go from 2/4 to 1/3 due to channel II ibistable tripped.
43
[Plant
._ Systems El 1 'Reactor j..
Protection System
~ 3.3 3.7 ut bays from 2B
/both the same. Distracter b is incorrect because SSPS train A will not be affected by the loss of 28 VI6
]except as noted above for distracter a, and that will not cause an OHA A-34. Distracter d is incorrect ibecause the Containment Spray coincidence is energize to actuate, so the deenergization of 1 channel
/will cause the remaining 3 channels to still need 2 channels to actuate, so the logic will go from 2/4 to 1213.
C is correct because 2B VIB supplies power to the slave relays on SSPS Train B, so none of the 5 VIB. Therefore they will not disagree on contact status because the loss of power will affect them
/slave relays will actuate.
loverhead Annunciators Window A
[Solid State Reactor Prot Train A AC Power Distribution I
- ? *'-- ;* ;'<,,,e; 1*.
_I --_-
RXPROTEOll RXPROTEOZO State the power supply to the SSPS Identify and describe the Control Room controls, indications, and alarms associated with the Reactor Protection System, including:
a) b)
c) d)
intended function The Control Room location of Reactor Protection System control bezels and indications The function of each Reactor Protection System Control Room control and indication The effect each Reactor Protection System control has upon Reactor Protection System components and operation The plant conditions or permissives required for Reactor Protection System Control Room controls to perform their e)
The setpoints associated with the Reactor Protection System control room alarms Tuesday. ADril 20, 2004 8:31:00 AM 1
I Page49of89 t
Given the following conditions:
going to LOSC-2. Maintaining >1 E4 lbmlhr to each SIG keeps tubes from drying out, among other I
,- Salem Unit 2 has experienced a MSLB at the Main Turbine steam piping bifurcation point.
!- All attempts at Main Steamline Isolation have failed.
!- Operators have transitioned out of EOP-TRIP-1 with RCS pressure at 1400 psig.
I - RCS pressure is currently 1300 psig and dropping.
- Charging system SI flow meter indicates 290 gpm.
'- The RCS cooldown is NOT being controlled.
Which of the following choices identifies an action that must be performed IAW 2-EOP-LOSC-2, RCS cooldown rate is 120 deg F/hr.
I
.MULTIPLE STEAM GENERATOR DEPRESSURIZATION?
/Trip all RCP's.
Stop BOTH RHR pumps.
'Send
............... operators
.._ to close all BFIS's, BF~O'S, and BF22's.
lM.............................................
ultiple Steam Generator Depressurization
-..J I
..I 1 6. Determine an appropriate transition out of the EOP r
I I - - -
/Direct From Source I
Tuesday, April 20 2004 8 31:Ol AM 1
I Paae50of89 t
Which of the followina choices identifies the purpose of the Containment SDrav Svstem?
I
'Maintain containment pressure less than the design pressure of 47 psig following a Loss of I Coolant Accident (LOCA).
iMaintain containment pressure less than the test pressure of 54 psig following a Main Steam lLine Break (MSLB) inside containment.
I
-Fject a mixture of borated water and Sodium Chloride (NACI) into the containment atmosphere followinn a LOCA to minimize exDosure to the public followina a LOCA.
- Inject a mixture of borated water and Sodium Hydroxide (NAOH) into the containment atmosphere followinn a MSLB with failed fuel. to minimize exDosure to the public.
I 12.1 l;c
- 026 i Containment Spray System J t
~
2.8 inment Spray System, Design Basis, Purpose, "...is to spray cool e in the event of a loss-of-coolant accident (LOCA) and thereby I
j
- ensure that containment uressure does not exceed the desian Dressure of 47 Dsia....'I.
i.........
I
~ ~~~~~~-
7 IConcept Used Wednesday, April 21, 2004 7:51:51 AM i
1 Page51 of89 I
Given the following conditions:
- Salem Unit 2 has experienced a Large Break Loss of Coolant Accident (LOCA).
- Containment H2 concentration has risen to 2%.
- 21 H2 Recombiner has been placed in service with containment pressure at 4.1 psig.
- 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later, containment H2 concentration has remained at 2%, and containment pressure has
/risen to 5.1 psig.
IlAW S2.OP-SO.CAN-001 HYDROGEN RECOMBINER OPERATION, which of the following ia ion of the H2 recombiners?
ering cont
!Lower the Recombiner Power Adjust Potentiometer setting by 4KW below the previous setting. I
/Secure 21 Recombiner due to containment pressure rising above 5 psig.
......... ~-..........
-1
~~
d; Recalculate the Recombiner Power Setting due to the rise in containment pressure.
!because if H2 concentration has remained stable or dropping, with a rise in containment pressure,
!recalculation of the power setting is required.
\\Hydrogen Recombiner Operation Describe the function and operating characteristics of the components associated with the following Containment And Containment Support Systems subsystems:
a)
Containment Fan Cooler System b)
Containment Iodine Removal System c)
Rod Drive Ventilation System d) e)
Reactor Shield Ventilation System 9
g)
Hydrogen Recombiner System Identify and describe the Control Room controls, alarms, and indications associated with the Containment And Containment Support Systems, including:
a) b)
c)
Support Systems components and operation d) to perform their intended function e)
Reactor Nozzle Support Ventilation System Containment Pressure - Vacuum Relief System
~
The Control Room location of Containment And Containment Support Systems control bezels and indications The function of each Containment And Containment Support Systems Control Room control and indication The effect each Containment And Containment Support Systems control has upon Containment And Containment The plant conditions or permissives required for Containment And Containment Support Systems Control Room controls The setpoints associated with the Containment and Support Systems Control Room alarms
- Tuesday, April 20, 2004 8.31 02 AM
]
Page52of89 I
r:..::.:: Tuesday, April 20, 2004 8:31:04 AM I
Pase53of89 1
Given the following conditions:
- Salem Unit 2 is 7 days into a refueling outage.
i-The core is partially offloaded with 7 bundles remaining in the Rx.
1-Fuel movement is in progress, and S2.OP-I0.ZZ-0010 SPENT FUEL POOL MANIP1 iin effect.
- SFP temperature is 120 deg. F.
8-21 SFP becomes air bound, trips on motor OL, and can NOT be restarted.
!- 22 SFP pump will NOT start.
- SFP hi level alarm is in alarm.
- SFP heatup rate is 12 deg F/ hr.
I LATIONS is llf SFP cooling can NOT be restored, which of the following choices describes an adverse
- diverted to the charcoal filter.
.Increased radiation levels at the FHB charcoal filter due to Spent Fuel off-gassing at temps >
,150 deg. F.
~
il
@ ' nability to place a raised Spent Fuel bundle into any location in the pool due to rising radiation 1
,level on 2R32 Fuel Handling Crane Area Monitor.
a El 47 following:
filter. B is correct because rising radiation will be seen as fuel off-gassing is expected to occur as temp increase to 150 deg. Distractor c is incorrect because any overflow will go out the ventilation openings in the SFP.
Distractor d is incorrect because it is always possible to lower a SF bundle.
LOSS of SDent Fuel Coolina I
SFPOOOE007
' ABSFOlE002 Identify and describe the local controls, indications, and alarms associated with the Spent Fuel Pool Cooling System, including:
a) b)
c) function. (N/A STA)
Discuss the remonse to a loss of sDent fuel coolina.
The location of Spent Fuel Pool Cooling System local controls and indications. (N/A STA)
The function of Spent Fuel Pool Cooling System local controls and indications. (N/A STA)
The plant conditions or permissives required for Spent Fuel Pool Cooling System local controls to perform their intended I
Tuesday, April 20, 2004 8:31:04 AM I
........... J I
Page54of89 I
I Tuesday, April 20, 2004 8.31.05 AM 1
1 Page55of89 1
As Unit 2 Main Turbine power is raised from 20% to loo%, Steam Generator Narrow Range Water Level will.....
~ __
'rise from 33% to 44%.
... I
~
remain constant at 33%.
- remain I. __..-
.. -. constant
..... __ at 44%
....... 1._
~ _ _
imodulate between 33%-44%.
I Station.... Drawing
.I I
1.
Tuesday. April 20, 2004 1
Page56of89 1
,Unit 2 is at 7% power during a plant startup. The NCO is using Steam Dumps to slowly raise power
'when Steam Dump demand suddenly rises in an uncontrolled manner. Reactor power starts to rise
- more rapidly and all attempts to close the Steam Dump valves fail. IAW S2.OP-ABSTM-0001, the
/control room crew will.....
- initiate a Main Steam Line Isolation (MSLI) ONLY.
/initiate a Safetv Iniection (SI) ONLY.
I
'trip the reactor, confirm the trip and initiate a MSLI.
b) Describe the plant response to actions taken in the abnormal procedure.
c) Describe the final plant condition that is established by the abnormal procedure.
I
~
i'&
- trip the reactor, initiate a SI.
,A2. \\;Ability to (a) predict the impacts of the following on the Main and Reheat Steam System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
iA2.04 ]Malfunctioning steam dump
/ I 3. 4 j l ]
AB-STM Continuous action summary states that at any time, if Reactor power rises uncontrollably, the control room will trip the reactor, confirm the trip, (if trip cannot be confirmed GO TO EOP-TRIP-I) initiate MSLI. If the source of the steam leak is NOT isolated, then initiate SI and GO TO TRIP-1. The
'distracters a and b are incorrect because they contain only 1 of the 3 required actions. Distracter d is incorrect because the oDerator attemDts to isolate the steam leak Drior to initiatina SI.
1 I
i I
l?
xcessive
.. Steam Flow
~
i:.::
Tuesday, April 20, 2004 8:31:05 AM I
Paae57of89 1
~
Given the following conditions:
'- Salem Unit 2 is operating at 12% power.
1-Control Bank D rods are at 189 steps withdrawn in MANUAL.
I-Main turbine is rolling up to normal speed.
!- Main steam dumps are set for 950 psig, in MS PRESS CONTROL-AUTO.
/If Main steam dump AUTO setpoint is adjusted to 940 psig, what effect will this have on Tave and i
1 1
ems........................
K5. I Knowledge of the operational implications of the following concepts as they apply to the Main and Reheat Steam System:
m mjJD3.61 When the steam dump pressure setpoint is adjusted downward, the control system will automatically attempt to control steam header pressure at the new setpoint by opening the steam dump valves further to reduce pressure. Tc will lower to the new saturation temp for the new lower steam pressure. Tave will lower, and the Dositive reactivitv added will cause Rx Dower to rise.
I
[Steam
..... Dump System Operation C I I
_._~-
I I STMGENE008 Describe the purpose and operation of the following Steam Dump Controllers, a)
Steam Pressure Controller b)
Load Rejection Controller c)
Plant Trip Controllers Identify and describe the Control Room controls, indications, and alarms associated with the Steam Generator, SG Blowdown and Drain Systems, including:
a)
NEO) b)
c) and Drain Systems components and operation (N/A NEO) d)
controls to perform their intended function e)
The Control Room location of Steam Generator, SG Blowdown and Drain Systems control bezels and indications (N/A The function of each Steam Generator, SG Blowdown and Drain Systems Control Room control and indication (N/A NEO)
~
The effect each Steam Generator, SG Blowdown and Drain Systems control has upon Steam Generator, SG Blowdown The plant conditions or permissives required for Steam Generator, SG Blowdown and Drain Systems Control Room The setpoints associated with the Steam Generator, SG Blowdown and Drain Systems control room alarms i
Tuesday, April 20, 2004 8:31:06 AM------..]
I Page58of89 I
Salem Unit 2 is operating at 80% power with the following conditions:
- 2.4.49
- 21 condensate pump CTT.
- 23 Heater drain pump 01s.
- PT-506, Steamline Inlet Pressure Channel is 01s.
- Main condenser backpressure is 2.1"Hg.
- 21 and 22 SGFPs are in service.
Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
/I
'14.01
- 22 SGFP speed starts acting erratically, and quickly degrades to the point where the crew manually trips 22 SGFP.
Given the above scenario, what action, if any, is IMMEDIATELY required of the crew?
I I
kd Tric, the Reactor. and GO TO 2-EOP-TRIP-1.
,Verifv Automatic Turbine Runback has or is occurring.
I
~
- E nsure Rod Bank Selector Switch is in AUTO.
I No other IMMEDIATE ACTION is required.
IApplication Salem 1 & 2 04 1 steam Dump System and Turbine Bypass Control 1041 0006449 I
I ABCNOI E002 I State the immediateactions of S2.OP-AB.CN-0001.
I Tuesday, April 20, 2004 8 I
Page59of89 I
~.
RXPROTEOl2 Given the following co nd it ions :
List all Reactor Trips and Safety Injections, including a) b)
c)
Name of the trip or safety injection Setpoint and Coincidence (N/A NEO)
Any related permissives or block signals (N/A NEO)
- - Salem Unit I is operating at 80% power.
- A large quantity of river grass starts building up on the Circ water traveling screens and
~ condenser waterboxes.
i-A rapid power reduction is initiated IAW S I.OP-AB.LOAD-0001 RAPID LOAD REDUCTIONl to
- maintain condenser backpressure.
- During the power reduction, the NCO places rod control in MANUAL and continues to drive rods in.
- The turbine is put on hold at 20%, with condenser backpressure at 4.8" Hg and stable.
- Reactor power and temperature continue to lower due to an excess amount of negative reactivity inserted with control rods and boration, and reactor power reaches 7% before stabilizing.
- The NCO starts to withdraw control rods in manual to restore RCS Tave which has dropped to 545 deg. F.
As the NCO continues to withdraw control rods continuously, which of the following actions will act first to prevent the RCS from exceeding the required DNBR?
HIHigh power reactor trip (low range) at 25% on 2/4 PR Nl's.
High power (low range) reactor trip at 20% power equivalent IR current on 2/2 IR Nl's.
Hbh power reactor trip at 109% on 2/4 PR Nl's.
Overpower rod block at 103% on 1/4 PR Nl's.
i ___ - -.. - - -
I I
1 1045000K518 045 -
J iMain-Tu&ine G ~ ~ r a t o r S y S t e ~
J K5.1 Knowledge of the operational implications of the following concepts as they apply to the Main Turbine Generator System:
m mmi3T coincidence are correct and it would actuate before any high power trips. Distracter b is incorrect because the coincidence is 1/2 not 2/2 IR Nl's. Distractor c is incorrect because the low power reactor trip would cause the trip to protect against DNB first. Distractor d is incorrect because it would happen later in event. if needed.
~.
I
[------Tuesday, April 20, 2004 8:31:09 AM
]
I Page61 of89 I
Given the following conditions:
- Salem unit 2 is operating at 100% power.
- All control systems are in AUTOMATIC.
1-OHA G-7, ADFCS SWITCH TO MANUAL alarms.
1-24BF19 valve demand is steady and indicates 5% lower than the 21-23BFl9 valve demand.
1-24 SG NR level is 41% and dropping slowly.
IAW S2.OP-AB.CN-0001. which of the followina describes the reauired actions?
I I
!Remove an adjacent bezel and insert it into the slot for 24BF19 and attempt to establish
'MANUAL control for 24BF19.
-Initiate a 10% load reduction to reduce feed flow to approximately the value required to maintain 24 SG level constant.
I Ensure 24BF1 9-and 24BF40 are in MANUAL and depress the open pushbutton on 24BF19 to
/Place the SGFP Master Speed Controller in MANUAL and adjust SGFP speed to raise 24 SG level.
/raise 24 SG level.
I
& 2 06/11/2004
-1
/059000A408 3
I
[l.9*1
'Step 3.27 of AB.CN-0001 states. "ESTABLISH manual control of any controller that has transferred AND STABILIZE affected SG level (s). Distracter a is incorrect since it is not IAW the AB. Actual
'performance of this act may work, but it is not proceduralized. Distracter b is incorrect because a load reduction is not specified in the procedure, although it might work.
Distracter d is incorrect because it
/is not in the procedure to attempt this, even though this action has been performed at Salem.
IMain FeedwaterIMain Condensate Abnormality I I
--_I-Tuesday, April 20, 2004 8:31:09 AM 1
- 1.
I Paae62of89 1
Given the following conditions:
- Salem Unit 2 has experienced a reactor trip from 100% power.
- 22 AFP Pressure Override Protection circuit has malfunctioned, causing the AF2ls (Auxiliary Feedwater Isolation Valve) supplied from this pump to close.
1
/With NO operator action, which of the following choices describes the indications which would be
/present 2 minutes after the reactor trip?
~
'AFW flow indication reading 0 gpm for 21 and 22 SGs.
123 and 24 SG wide range (WR) levels rising slower than 21 and 22 SG WR levels.
~~
~
.AFW flow indication reading 0 gpm for._____
23 and 24 SGs.
~
~
~~~~
.2land
... __ 22 SG wide range
. (WR).. levels rising slower than 23 and 24 SG WR levels.
c -
r l
- Auxiliary / Emergency Feedwater System
[Application Salem 1 & 2 I061 OOOK601 304 54
'K6.
Knowledge of the of the effect of a loss or malfunction on the following will have on the Auxiliary /
Emergency Feedwater System:
i K 6. 0 u IControllers and Positioners 1
?.8*
With the 22 AFP Pressure override protection controlling 21 and 22 AF2ls, d would be the correct answer because 23 AFP would still be supplying AFW to 21 and 22 SGs through the AFI Is.
Distracters a and b are incorrect because TOTAL AFW flow (from MDAFW pps and TDAFW pps combined) is indicated on 2CC2. Distracter c is incorrect because 23 and 24 WR levels would be rising faster than 21 and 22 SG WR levels because of the combination of MD and TD AFW pps supplying feed to those aenerators.
-~
Reactor Trip or Safety Injection a) b)
c)
Auxilialy Feedwater Pump Automatic Start Motor-Driven AFW Pump Recirculation Flow Control Valves Motor-Driven AFW Pump Discharge Flow Control Valves r---- Tuesday, April 20,2004 8:31:10 AM I
I Page63of89 I
~
- Given the following conditions
-- - -~
i-2C EDG is operating in parallel with the 500KV grid for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> endurance run IAW S2.0P-
- While operating at 2525 KW three hours into the test, the operator mistakenly adjusts 2C EDG speed control resulting in MW loading increasing to 2790 KW.
'ST.DG-O014,2C DIESEL GENERATOR ENDURANCE RUN.
Licensed Operators.
Which of the following choices describes the consequences, if any, of continued EDG operation at ithis KW load?
Operation for the remainina 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> of the test...
I
~
~~
!will result in exceeding the 30 minute load limitation for 2C EDG.
,will result in exceeding
........... the 2000
. hour load..-
limitation
.........._ for.-. 2C
... EDG.
iwill not have anv adverse effect on 2CEDG.
06/11 /2004
[Memory
'Salem 1 & 2 55 lthe EDG will alwavs exceed the limit of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for oPeration between 2750-286OW.
2C Diesel Generator Endurance Run I _-
orereauisites I
I I
I PI' Tuesday, April 20, 2004 8:31 :I 1 AM I
-. I.....
I Paae64of89 I
.-~.... --- -.. -.....
/Given the following conditions:
I 1-Salem Unit 2 is performing a reactor startup.
Reactor power is stable at IXIOE-8 amps.
1-OHA B-3 2A VTL INSTR BUS INVRT FAIL alarms, accompanied by Aux Alarm Typewriter Point i-1147 2A VITAL INSTR BUS INV TROUBLE.
Using Attachments I and 2 of S2.OP-SO.ll5-OOl1 2A VITAL INSTRUMENT BUS UPS INVERTER OPERATION, determine the status of the 2A Vital Bus Inverter and its effect on the Rx startup.
The 2A Vital Bus Inverter...
e detector momentarily lost power.
,has undergone-a latched transfer, Rx startup may continue with no restrictions.
has undergone
__ an
- - unlatched
- transfer,
___ Rx startup
_____ - may ___ continue
..- with no restrictions.
has undergone an unlatched transfer, the reactor has tripped when I N35 Intermediate range
-~
detector momentarily lost power.
~~
10 56
/predictions, use procedures to correct, control, or mitigate the consequences of those abnormal ON INVERTER lamp will be lit, indicating the static switch momentarily swapped to the alternate source land automatically switched back. An unlatched transfer is corrected by depressing the alarm contact reset PB and the alarms will clear. An unlatched transfer has no effect on the OPERABILITY of the inverter, and will not affect the reactor S/U.
2A Vital Instrument Bus UPS System Operation I
115VACE007 ldentifv and describe the local controls and indications associated with the 115 Vac Electncal Svstems. includina lAnateria1 Required for Examination. *
' I s2.0P-S0.115-0011 Attachments I & 2 Tuesday, April 20, 2004 8:41:27 AM
]
c_
~
I Page65of89 I
~.
/Which of the following choices contains ONLY components supplied power from the Unit 2 i.. I 125VDC system?
'Control Room Emergency Lighting, 13KV switchgear,
............... SCADA system.
Pzr heater bus switchgear, EDG control power, 2RP4 status lights.
Alternate Shutdown System - (ASDS),
- IRPI,
__ 460V
- __ - switchgear.
kk! Vital Instrument Bus Inverters, 4KV switchgear, SGFP Emergency Bearing Oil Pump.
receives 125VDC power from bkr 2CDC31. All the aforementioned breakers are found on drawing 223720.
The2RP4 status panel receives 28VDC power from bkrs 2A11 DE and 2B12E, shown on drawing 21 1357.
The SCADA system ALSO receives 28VDC power.
also shown on 223720. IRPl is Dowered from 23OVAC and 28VDC. SGFP EBOP is 250VDC load ASDS supply IS 104 57 I*
IUnit 2 125 V.D.C One Line 1 Unit 2 28 Volt D.C. One Line
~ _ -
I I
DCELECEOOJ Draw one-line diagrams of the DC Electrical Systems similar to the TPs listed, which indicate the following:
A.
250 Vdc Distribution (TP-1)
B.
C.
125 Vdc Vital Distribution (TP-4) 125 VDC CW Distribution (TP-7)
~
[_---Tuesday,
....... April 20, 2004 8:31:12 AM
]
I Page66of89 I
Given the following conditions:
- Salem Unit 2 is operating at 30% power.
- 125 VDC breaker 2BDClAX12,2G 4KV Bus Control Power Supply (REG) trips due to a breaker
,malfunction.
/With NO operator action, which of the following choices identifies the status of the plant following I
r can on1 PB on the breaker locally.
The reactor is tripped, and 24 RCP breaker can only be tripped by depressing the STOP PB from the Control Room.
~
~~
.The L
reactor is tripped, and 24 RCP breaker is tripped.
/
!The reactor is at 30% power, and 24 RCP breaker is tripped.
~~~
58
-- -1
' T I Knowledge of the effect that a loss or malfunction of the D.C. Electrical Distribution will have on the--
following:
,incorrect because when control power is lost to the pump breaker, the Trip Coil cannot be energized to trip the breaker, and there is no trip signal present.
idemand for a reactor trip or breaker trip on a loss of control power to the pump breaker. Distracter d IS Distracter c is incorrect because there is no
'incorrect because the reactor is not tripped.
[RCS Reactor coolant pumps and lift oil pumps RCS Reactor coolant pump
~
~~~
~
I
~ - - - -
I DCELECEOO7 4KVACOE007 Identify and describe the local controls, indications, and alarms associated with the DC Electrical Systems, including:
A.
The location of DC Electrical Systems local controls and indications B. The function of DC Electrical Systems local controls and indications C.
The plant conditions or permissives required for DC Electrical Systems local controls to perform their intended function D.
The location of 4160 Electrical System local controls and indications. (N/A STA)
The function of 4160 Electrical System local controls and indications. (N/A STA)
The plant conditions or permissives required for 41 60 Electrical System local controls to perform their intended function The setpoints associated with the DC Electrical Systems local alarms
~~
Identify and describe the local controls, indications, and alarms associated with the 4160 Electrical System, including:
a) b)
c)
(N/A STA)
L __......
Tuesday, April 20, 2004 8.31 13 AM
]
I I
Page67of89 I
Given the f o I I ow i n g co n d it i o ns :
A fire has occurred in Unit 2 Relay Room.
I-
,- There has been substantial damage to equipment, wiring and relays.
'- As a result of the damage, all Emergency Diesel Generators (EDG) are supplying their associated vital bus and numerous individual valve and pump controls have been locally selected to EMERGENCY, including 21 Diesel Fuel Oil Transfer Pump (DFOTP).
- 22 DFOTP is tagged 00s.
Which one of the following correctly describes how the DFO Day Tank levels will be maintained?
- In EMERGENCY, 21 DFOTP operates off of the automatic controls from 22 DFOTP. Unless 122 DFOTP controls are damaged, automatic pump cycling and overflow protection is
- maintained.
In EMERGENCY, 21 DFOTP will run continuously until the switch is returned to NORMAL.
Overfilling/overflow protection is provided by the DFO Day Tank high level overflow connection 21 DFOTP starts automatically but continues to run until stopped-manually.
Overfilling/overflow protection is provided by the DFO Day Tank high level overflow connection 21 DFOTP must be manually started but the high level trip will automatically stop the pump to back to the DFOST.
back to
___ the DFOST.
provide protection against overfilling/overflow of the DFO Day Tanks.
-~
~
064000G128 59 2.1.28 /Knowledge of the purpose and function of major system components and controls.
3 2 ~, 3 3 efer to drawing 223825 during this explanation. When the 21 FOST local EMERGENCYINORMAL 13 close allowing continuous energization of the "6" coil, which is inline with breaker feeding the FO Xfer pp. With no OL condition present (49), the pump will start. The 5 and 7 contacts of FTP-3 add a
/redundant fuse to the control power circuits around the normal control power fuses to allow an added
!degree of certainty that they pump will run as required. As the 6 relay is energized, the 6X relay will
!energize through the 3/2 contact of the 6 relay. With the 6X relay energized AND the FRT-3 switch in
- EMERGENCY, the red EMERGENCY light will light giving indication that the pump should be running When the FTP-3 switch was taken from NORMAL to EMERGENCY, it removed the FO Day tank level
'control circuit from affecting the 6 coil by opening FTP-3 contacts 2 and 4.
oDens to remove the REGULAR and BACKUP function from the FO Xfer DumD.
itch (FTP-3) is taken to EMERGENCY, contacts 1,3,5,and 7 close, and 2,4,6,8 open. Contacts 1 and The FTP-3 switch 8 contact FUEOILEOIO Oil System, a) b)
c) d)
The Limiting Condition(s) for Operation The Bases for the LCO(s)
The applicability of the LCO(s) (N/A NEO)
The LCO Action Statement(s) (N/A NEO)
Tuesday, April 20,2004 8:31:14 AM 1 Page68of89 I E:.
~
]
[:-.:- Tuesday, April 20, 2004 8:31:15AM
]
~
I Page69of89 I
IWhich of the following choices identifies the initiating signal and the response to that signal which
/will minimize the radiological consequences of a Fuel Handling Accident in the Unit 2 Fuel
[Handling Building (FHB)?
BOTH 2R5 SFP AREA RADIATION MONITOR AND 2R9 FHB NEW FUEL STORAGE AREA
!RADIATION MONITOR, must be in alarm to cause FHB ventilation exhaust fans to start and
- exhaust
.......... through
- a
-. charcoal filter.
2R32A FUEL HANDLING CRANE AREA MONITOR in warning will cause all Fuel Handling ICrane functions to be locked out.
I
~
FHVENTEOOG
[EITHER 2R5, SFP AREA RADIATION-MONITOR~-OR 2R9 FHB-NEW-FUEL STORAGE AREA RADIATION MONITOR, in alarm causes FHB ventilation exhaust fans to start and exhaust throuah a charcoal filter.
Describe the interlocks associated with the following Fuel Handling Area Ventilation System components:
a)
Supply Fan Unit b)
Exhaust Fan Controls c)
Exhaust Filter Units 2R32A-, Fuel Handling Crane Area Monitor in warning willcause downward Fuel Handling Crane movement to be locked out.
06/11 /2004 60 K4.
I Knowledge of ARM system design feature(s) and or interlock(s) which provide for the following:
,K4.02
/Fuel building isolation 3 2* 3 4*
Answer c is correct because either of the R5 or R9 in alarm (a7.0 mremlhr) will cause the Fuel Handling Building exhaust fans to start if not in service, and to re-align their exhaust flow to the FHB
'Charcoal filter.
'Distracters I....................................
b and d are incorrect because the R32A only locks out OUTWARD rod motion Distracter a is incorrect because only 1 monitor is needed to be in alarm, not BOTH
?.
t RMSOOOE006
...~ ___._
Tuesday, April 20, 2004 8:31:15 AM--'--'.]
I Page70of89 I
I Tuesdav. Aoril20. 2004 8:31:16 AM 1
1 Page71 of89 I
!Given the following conditions:
- Salem Unit 1 is operating at 100% power.
- Control room operators are preparing to perform a Containment Pressure Relief IAW S I.OP-
- Containment radiation levels are NORMAL for 100% power operation with no failed fuel.
SO. CAN -0 002, C 0 N TA I N M E NT PRESS U R E-VAC U U M RE LI E F SYSTEM 0 P E RAT I 0 N.
1 ;After opening the 1VC5 and 1VC6 to initiate the pressure relief, which of the following choices
/describes how the respective radiation monitors indication will respond?
1 R12A-Containment Gas Effluent 1 R41 B-Plant Vent Noble Gas Intermediate Range 1 R41 D-Plant Vent Noble Gas Release Rate 1 R12A rises; 1 R41 B constant; 1 R41 D constant.
-1 R12A constant; I R41 B rises; I R41 D constant.
11 R12A rises; 1 R41 B rises; 1 R41 D rises.
11 R12A constant: 1 R41 B constant-1 R41 D rises.
06/11/2004 61
,Monitoring System controls including:
3 2 1 3 5
,flow will start when the lo range 1 R41A monitor nears its high end of monitoring range. It's indication
/will not change during a pressure relief with NORMAL containment radiation levels. The R41D
'provides the gaseous effluent release rate (uCi/sec) by combining (product of) the on-range R41A through R41C with plant vent flow (cdsec). It will rise when the pressure relief is initiated, and also
'provides
--............. -.-. automatic termination
.. of release on hi gaseous effluent.
Abnormal Radiation I....... -_
~._
including:
A.
The Control Room location of Radiation Monitoring System control bezels and indications (N/A NEO)
B. The function of each Radiation Monitoring System Control Room control and indication (NIA NEO)
C. The effect each Radiation Monitoring System control has upon Radiation Monitoring System components and D.
The plant conditions or permissives required for Radiation Monitoring System Control Room controls to perform their operation (N/A NEO) intended function Tuesday, April 20, 2004 8:31:16 AM L
1 Page72of89 I
Tuesday, April 20, 2004 8:31:17 AM 1
.-- 1 I
Paae73of89 1
,Which one of the choices describes a difference between Unit 1 and Unit 2 SG Blowdown RMS L..-- lchannel response?
!On Unit 1, a WARNING on an R-I 9 closes the respective GB4, Blowdown Isolation Valve. On Unit 2. it does not.
/On Unit I, an ALARM on an R-I 9 closes only the respective GB4. On Unit 2, the same signal
'closes all interlocked blowdown valves.
On Unit 2, an ALARM on an R-19 channel closes just the respective GB4. On Unit 1, the same sianal closes all interlocked blowdown valves.
~ ;o d;. ~
n Unit 2, a WARNING on an R-19 closes the respective GB4, Blowdown Isolation Valve. On
~
i Unit 1. it does not.
06/11 /ZOO4 LO73
' Process Radiation Monitoring System Modified Wednesday, April 21, 2004 75456 AM
]
1 Page74of89 I
Given the following conditions:
i-Salem Unit 1 has initiated a MANUAL Safety Injection (SI) following a Loss of the 500kv
I-4kv Vital bus "AI has been de-energized due to a Bus Differential relay actuation.
i
/Which of the following choices identifies the Service Water pumps that will be running 2 minutes 11 __ and - 14.
12 and 16.
075000K203 63
[K2,Q..d.Knowledge of bus power supplies to the following:
2.6* 2.7*
- t. Distracter a is and the B SE will
.will not start. C is correct because the C bus SEC will start the C bus lead pump (1 start whichever pump is selected to "LEAD" (13 or 14). The "A" bus pumps (15 and 16) cannot start because their rewective 4kv vital bus is deenerclized
[Service Water Pump Operation I
a)
Mechanical Trash Rake b)
Traveling Screens c)
Service Water Pumps Ventilation Fans & Heaters i- --..
I
~.______.
Tuesday, April 20, 2004 8:31:18 AM 1
- Given the following conditions:
'- Salem Unit I is operating at 100% power.
1-Unit 1 ECAC is shutdown and aligned for normal AUTOMATIC operation.
'- 11 Chilled Water pump is in operation.
1 12 Chilled Water pump is in standby.
1-
- A momentary dip of Station Air header pressure causes Unit 1 ECAC to automatically start.
'Which of the following describes the operation of the Chilled Water Pumps?
I 12 chilled -
water pump starts
__ and
- 11
- - chilled
-_ water
- pump stops after a 2 second time delay 12 chilled water pump remains in standby and I 1 chilled waterpump remains running.
12 chilled water pump starts
___ - - and
- - 11 chilled
___ ___ __ water
- __ - - pump
. - remains
- running.
~
d-! 12 chilled water pump is locked out from starting in automatic while 11 chilled water pump is running.
06/11/2004 078000G132 64 2.1 I - Conduct
__ Of - Operations 2.1.32 [Ability to explain and apply all system limits and precautions.
3 4 3 8 AC sends an auto start signal to 12 CH pump If signal to 11 CH pump. It does not trip 11 pump if 1-running.
- So with the conditions in the stem,
- 11 will
- __ remain running and 12 will
- start.
1 Control Air System Operation Tuesday, April 20, 2004 8:31 :I 9 AM
.. ]
I Page76of89 I
Given the following conditions:
1-All 3 Salem Station Air Compressors have become unavailable.
i-The NORMAL cooling water supply to the Unit 1 Emergency Control Air Compressor (ECAC) has
'been lost.
- CONAIRE007
~~~~~~~~~
1 CONAIREOIP Operation of the Unit 1 ECAC....
!can continue since cooling water will automatically swap to Demineralized Water through a bj Control Air Riceivers c) d)
Emergency Control Air Dryers e) 9 Station Blackout Compressor g)
CA Containment Isolation Valves h)
PORV Control Air Accumulators i)
Redundant Air Supply Panels Identify and describe the local controls and indications associated with the Control Air System, including:
a) b)
c)
Describe the procedures which govern the operation of the Control Air System including, significant prerequisites and precautions associated with each operating procedure which are required to be considered by either Licensed Operators or Non-Licensed Ooerators.
Emergency Control Air Compressors (ECACs)
Excess Flow Check Valves (EFCVs)
The location of Control Air System local controls and indications The function of Control Air System local controls and indications The plant conditions or permissives required for Control Air System local controls to perform their intended function
~..
must be discontinued until cooling water can be manually aligned through a spool piece from can continue since cooling water will automatically swap to Service Water through a check valve.
!must be discontinued until cooling water can be manually aligned through a spool piece from Service Water.
078000K104 65 2 6 2 9 a supply and a return spool piece.
(Service - Water-Nuclear Drawing Control Air System Operation CONAIRE004 Describe the function and operating characteristics for the following Control Air System components:
4 a)
Control Air Drvers
.- -~.__-
Tue~ay~ApriI20,2004 8.31.I9 AM 1
Page77of89 1
Fromthe choices below, identify the ONE set of conditions that constitutes a VIOLATION of conditions
___ and limitations in Technical
- Specifications for Salem
__ - Unit
2.
'Operation at 40% power for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> with:
- - QPTR of 1.017.
- AFD of +11.0.
'Operation at 100% power for 36hours with:
- RCS net unidentified leakage = 0.08 gpm.
- - 24 _ _ - -
SG primary to secondary
. leakage
-- - =
- 0.4
- gpm.
- QPTR = 1.01 as indicated by - BEACON.
Operation at 80% power for I O hours with:
- Power Range Channel N-41 inoperable and in tripped condition.
Operation at 35% power for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> with:
- Reactor coolant specific activity of 1 OE-2 uCi/gram DOSE EQUIVALENT 1-1 31.
- Reactor coolant specific activity of 50/E bar uCi/gram.
r Salem1 & 2 J
06/11/2004
- b- ]
[e- - 1 bompreh.ension_
66 TI Conduct of Operations 2 7 3 9 E SG which is hours or be in HSB within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the LCO to become active, then 10 more hours mean that HSB must be achieved within 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />. Distracter a is incorrect because the Tech spec for AFD and QPTR is only applicable > 50% power. The AFD condition will prevent raising power above 50% due to 100 penalty minutes having been accumulated.
instrument is allowed to be inoperable as long as it is placed in tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
,QPTR is monitored Id12 hours by in-core OR BEACON if above 75% power. Distracter D is incorrect
!because the DEI limit of 1.O must be exceeded 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, and the plant not in HSB below 500 degrees Distracter C is incorrect because 1 power range if o be
__ in __ violation.
__ Also
__ the E-bar
- - limit
__ ___ is
___ lOO/E-bar.
1 Salem
__ Technical
_ ' - Specifications
[Core Operating Limits Report TECHSPE010 I Define the term Lirnitina Condition for Operation as it amlies to the Technical Specifications
[Material Required for Examination I NFS-0231, Core Operating Limits Report, Unit 2, Cycle 14, Figure 2, "Axial Flux Difference Limits as a Function of Rated Thermal Power tly Modified
- TMI I, 5/19/2003 NRC Exam, modified to fit Salem Tech Specs I.-.. - -.....-......-.
I Page78of89 I
/Which of the followhg -choices contains ONLY items that are described in 10 CFR 50.46 as Acceptance Criteria for Emergency Core Cooling systems for light-water nuclear power reactors?
I I.
The cladding thickness shall nowhere be lower than 17% of the total cladding thickness prior to
'oxidation.
II. The calculated maximum fuel element cladding temperature shall not exceed 2200 deg F.
!Ill. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
IV. The calculated total amount of hydrogen generated from the zirc-water reaction shall not exceed 0.01 times the amount that would be generated if all the cladding surrounding the fuel were to react.
- V. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total
,cladding thickness before oxidation.
VI. After any successful initial operation of the ECCS system, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long lived radioactivity remaining in the core.
VII. The zirc-water reaction postulated to occur at 1800 deg. F will not become self sustaining
/under any circumstance.
I, Ill, v, VII.
i,
_ _ 711,
__ IV,
- - VI.
,lV, VI VI, VII.
[I, _ _ -
II, Ill, IV.
r Salem 1 & 2 061
~
194001 G 112004 28 67 2.1 IjConduct of Operations 2.1.28-IKnowledge of the purpose and function of major system components and controls.
3 2 3 3
/10CFR46 paragraph b delineate the Acceptance Criteria for a LOCA. Also Salem FSAR contains the same criteria coPied from 10CFR46 in section 15.4.1.I.
The only answer which contains ONLY correct is answer b. All'the distracters contain either numeral I or VII, neither of which are correct
/Code of Federal Regulations I
associated reference material.
I Tuesday, April 20, 2004 8 31:21 AM 1
I Paae79of89 I
~..........................................................
.-~.
- Which of the following choices describes what SHALL occur PRIOR to performing action steps
- contained in Salem
............ Operations Department Implementing Procedure,
... and
_._~
why?.-
ISpecific approval from the Shift Manager SHALL be obtained prior to manipulating any
- component that will have an effect on Control Room indication, to preclude an NCO from
.responding
-. -........ incorrectly
~-.............. to an expected condition.
- All Precautions and Limitations for the procedure SHALL be reviewed and compared against current plant conditions, to minimize the possibility that a system operating limit will not be
,exceeded
. inadvertently during procedure
___ __ performance.
iAn Infrequently Performed Test or Evolution (IPTE) brief SHALL be held if the procedure has hot j
been performed within the past 6 months to familiarize operators with the procedure prior to use.
All Prerequisites SHALL-be completed unless approval is obtained from the unit Reactor Operator to N/A a particular prerequisite to allow expedited completion of procedure.
0611 1/2004
' T I ; C o n d u c t of Operations Distracter a is incorrect because SM approval is not required.
IPTE is not required based ONLY on the periodicity of the procedure.
Distracter d is incorrect because expedited performance is not a reason to N/A procedure steps, and the unit CRS must
,approve any such N/Aing.
Distracter c is incorrect because an Answer b is correct because P&L are required to be reviewed and signed
- k L rior to reaching any
__ performance step, and they guard against inadvertently exceeding system limits.
[Use of Procedures tate the requirements associated with the use of the following types of procedures in accordance with NC NA-AP ZZ-0001, uclear Procedure System:
- a.
Category I
- b.
Category It Tuesday, April 20, 2004 8:31:22 AM I
I Page80of89 I
/Given the following conditions :
Salem Unit 2 is in Mode 1, exiting a refueling outage.
~
Rx power is 16%.
- The Main Turbine is rolling unloaded at 1800 rpm.
I
/Which of the following choices identifies the actions required to raise power to 18% IAW S2.0P-jSO.MS-0002 STEAM DUMP SYSTEM OPERATION?
I URE the STM DUMP PRESSURE SETPOINT DECREASE pushbutton, then withdraw Control Rods in MANUAL or dilute the RCS
- to establish
-.-. 18%
___. - power.
-Raise the Main Steam Dump PRESSURE MODE-AUTO setpoint by depressing the STM DUMP PRESSURE SETPOINT INCREASE pushbutton, then withdraw Control Rods in OPEN the Main Steam Dump control valves in PRESSURE MODE-MANUAL by depressing the OPEN VLV pushbutton, then ensure Control Rods move in AUTOMATIC to maintain Tavg
,CLOSE the Main Steam Dump control valves in PRESSURE MODE-MANUAL by depressing the CLOSE VLV pushbutton, then ensure Control Rods move in AUTOMATIC to maintain I ITavg on programmed value.
MANUAL or borate the
_ - RCS to establish - 18%
- __ power.
on programmed value.
06/11/2004 1940016202 AUTOMATIC Mode. Rod Control will be in MANUAL. Turbine inlet pressure will be very low, and the requirement for placing rods in AUTOMATIC (PT505 > 15%, P-2) will not have been met since the Main Turbine has not been synchronized. IOP-3 step 5.4.20 initiates a power ascension to 10-20 % using Main Steam Dumps IAW SO.MS-002, Steam Dump Operation. Section 5.4 of SO.MS-0002 identifies the actions to be performed to raise Rx power (Step 5.4.1.8) of adjusting the pressure setpoint DOWN (to raise steam flow-->raise Rx power) then to withdraw control rods or dilute the RCS until predetermined power level is achieved. Distracter B has the wrong direction of adjustment for the steam dumps. Distracters c and d both have the wrona control mode of MANUAL.
L
. MSTEAME008 Y
ISTEAM DUMP SYSTEM OPERATION a) b)
c) 1.0~10-8 IRAmps Identify and describe the Control Room controls, indications, and alarms associated with the Main Steam System. including:
a) b)
c)
SR Permissive light is energized (P-6).
2/4 PRs are greater that 10% (P-10).
The Control Room location of Main Steam System control bezels and indications The function of each Main Steam System Control Room control and indication The effect each Main Steam System control has upon Main Steam System components and operation IOP003E004 Recosnize the actions which are reauired or may be taken when power reaches various levels listed below-
! RXOPERE021 Tuesday, April 20, 2004 8:31:24 AM I
~
~
J d) function e)
Explain the relationship between steam flow and reactor power given specific conditions.
The plant conditions or permissives required for Main Steam System Control Room controls to perform their intended The setpoints associated with the Main Steam System control room alarms I
Page82of89 I
~-
- On November 21st at 1300, with Unit 2 in Mode 1, it is discovered that the monthly surveillance to verify that ECCS piping is full of water has NOT been performed since October 1 Ith at 1300.
TECHSPE014
~
/Which of the following choices identifies the LATEST time for satisfactorily performing the required
- I300 on December 12th.
/I 300 on December 21 st.
11300 on December 22nd.
Describe the general requirements associated with Specifications 4.0.1 through 4.0.5 relating to implementation of the Technical Specification Surveillance Requirements 1G212 70 3.4 i f the surveillance to prevent ektering the LCO and taking the required actions therein. The periodicity for 3.5.2 to ensure the ECCS piping is full of water is 31 days per Surv. Req. 4.5.2.b.2.(tn stem) A
,"month" in Tech Specs is always 31 days.
31 days from the time of discovery (1 1/21 @ 1300) is Salem Technical Specifications r--
Tuesday, April 20, 2004 9:37:28 AM I.... ~...............
J I
Page83of89 I
........ --- - - ~
~..
- Which of the following parameter limits is established to ensure that radiation releases will remain
-- he limits of 10CFR20?
Secondary system
.. activity.
_ _ _ ~. ___.
~-.-.~-
. ~ -
~
- ~ -
auid Waste discharae activitv.
1 WASLIQE002 El Primarv svstem activitv.
Describe the design bases of the Radioactive Liquid Waste System 06/11/2003 1
194001 G301
- 1 71 2 6 3 0 to unrestricted r secondary activity, primary activity and primary to secondary leakage limits ensure conformance to 1 OCFRl 00 limits.
rn Source
'5/5/03 Salem NRC Exam Tuesday, April 20, 2004 8:31:25 AM 1
....... 1 1
Page 84 of 8 9 - 1
While performing a review of 21 GDT release papetwork IAW S2.OP-SO.WG-0008, it is noticed that the calculated maximum release rate is 28 scfm.
1 WASGASE012
~~~~~~~~~~~~~
Which of the following choices identifies the actions that this release rate will cause?
@! The GDT release cannot be initiated because flow rates c 32 scfm cannot guarantee that the dose received by a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY will not exceed 500 mremlyr whole body.
'The GDT release cannot be initiated because flow rates dose received by Site Staff inside the Protected Area will not exceed 3000 mrem/yr whole 32 scfm cannot guarantee that the Describe the procedures which govern the operation of the Radioactive Waste Gas System, including significant prerequisites body.
The GDT release can be initiated if a double, independent verification of the release rate has been performed.
The GDT release can be initiated if the Auxiliary Building Ventilation Exhaust flow is verified to be above 125,000 scfm.
aiem I
& 2
-J 06/11 /2003 k l IB I
l M i 6 6 k _.
!2.3 1 Radiation Control 2 7 3 2 gamma dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the whole body and 3000 mrem/year to the skin The release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem/year."
The waste gas
,release procedure specifies in the Precautions and Limitation section that..." Tanks with a calculated
,maximum allowable release rate of less than or equal to 32 scfm shall not be released."
I 121 Waste Gas. Decay Tank
__ Preparation
____ and
- Release I
tly Modified I
Tuesday, April 20,2004 8 31.26 AM I
Page85of89 1
- The crew is performing EOP-FRCC-I, RESPONSE TO INADEQUATE CORE COOLING. Both
/channels of Reactor Vessel Level Indication System (RVLIS) are INOPERABLE. Preparations are
!being made to start RCPs. Which one of the following indications provides the status of RCS
!inventory under these conditions?
/Core
....... Exit Thermocouples.
j Pressurizer level.
iSafetv lniection flow.
~
~.--
'Core delta temperature (Thot-Tcold).
06/11/2004 01 G403 73 3 5 3 8 During inadequate core cooling events, there will most likely be no pressurizer level indication, and even if there is, it's probably because water was displaced from the outlet plenum. Therefore, in FRCC-1, pressurizer level is not used as an indicator. SI flow can be an indicator of core cooling capability, but not an actual inventory indication. And in this case, since the crew is preparing to start RCPs, SI has not been restored yet. Core Delta T gives a reliable indication of natural circulation heat removal during subcooled conditions, but for the conditions given, can be any value due to loop stagnation and loss of reflux, most likely a low Delta T since core cooling has been lost. FRCC-1 uses RVLIS and LET'S Wednesday, April 21, 2004 7:56:32 AM I
I Page86of89 I
Given the following:
- Salem Units 1 and 2 are operating at 100% power.
- 1A EDG is C/T for maintenance.
- A loss of off-site power occurs, and both units trip.
'- 1B EDG fails to start.
- I C EDG starts and trips on low lube oil pressure.
Which of the following choices describes how the control room response will differ if the crew directly enters 1-EOP-LOPA-1 LOSS OF ALL AC POWER, instead of I-EOP-TRIP-1 REACTOR TRIP OR SAFETY INJECTION?
afety Injection will always be initiated by the operators prior to power restoration to
/Functional Restoration Procedures (FRPs) will NOT be implemented until after a transition out 1-EOP-LOPA-1 verifies AFW flow to be > 44E4 Ibm/hr vs. >22E4 Ibm/hr in 1-EOP-TRIP-1 ensure correct component position upon 4KV Vital bus restoration.
'of LOPA-I is performed.
due
- - to the
__ ___ increased decay
- heat
. - load
- which might
- - _--- be
- present if the
_ _ reactor did not trip 1-EOP-LOPA-I does NOT confirm the reactor trip because a transition to I-FRSM-1 RESPONSE TO NUCLEAR POWER GENERATION with all AC busses deenergized would 06/11/2004 194001 G406 74 3 1 4 0 us IS energized so
/because LOPA-I verifies 22E4 lbmlhr to verify that the TDAFW pump is supplying the minimum safeguards AFW flow for heat removal. Distractor c is incorrect because FRP's are implemented
'PRIOR to leavina LOPA-1 at step 20 when prompt restoration of a vital bus has occurred
/Loss of All AC Power Basis Document
~
Tuesday, April 20,2004 8:31:28 AM 1
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Page87of89 I
iGiven the following:
- Salem Unit 1 has been manually tripped using the reactor trip switch from 80% power
- 3 of 4 Turbine Stop Valves indicate closed on 1 RP4.
- 3/3 Auto Stop Oil Pressure Low Bistables are lit on 1 RP4.
1-IR NI SUR is negative.
L Rx power is 1.2% and lowering.
- IR NI indicate lowering flux.
- No other action has been taken.
I Which one of the following choices describes the status of the plant, and any required IMMEDIATE
/actions in response to these indications?
~ _
- Rx trip IS confirmed, Turbine trip is NOT confirmed. Operate the turbine trip switch on 2CC3.
iRx trip is NOT confirmed, Turbine trip IS confirmed. GO TO FRSM-1 RESPONSE TO
/Rx
-.. trip
.............. IS
........... confirmed,
................. Turbine
............. trip IS confirmed. Verify any 4KV....... Vital bus is energized.
INUCLEAR POWER GENERATION.
Rx trip is NOT confirmed, Turbine trip is NOT confirmed. Open Reactor Trip Breakers.
06/11 /ZOO4 194001 G407 75 3 1 3 8 of reactor tripped SUR indication negative. The reactor is tripped. The Turbine Trip confirmation is defined as ALL turbine stop valves closed. The turbine is not tripped, and as no other action had been taken except tripping the reactor, the next step is to trip the turbine using the turbine trip switch.
Distractor b IS
,incorrect because the Rx trip is confirmed, and the turbine trip is not confirmed. The action IS correct for an ATWT when no other action opens the reactor trip breakers..
Distractor c is incorrect because the Iturbine trip is not confirmed. The action is correct for the status in c. Distractor d is incorrect because ithe Rx triD is confirmed. The action is correct for an ATWT (first alternate action to triD the reactor).
I -
Use of Procedures A.-Immediate Actions B.-Continuous Action Summaries C.-Communications D.-Log Keeping E.-Application of Notes and Cautions F.-Transitions 1
G.-Adverse Containment I
Page88of89 I
[---Wednesday, April 21, 2004 7 5 8 5 4 AM 1 1
Page89of89 I
- 1. b 2. a
- 3. b 4. a 5. a 6. d 7. b 8. c 9. c
- 10. c
- 11. d
- 12. d
- 13. c
- 14. d
- 15. d
- 16. b
- 17. b
- 18. a
- 19. a
- 20. d
- 21. c
- 22. a
- 23. b
- 24. b
- 25. d Reactor Operator Answer Key
- 26. b
- 27. b
- 28. c
- 29. a
- 30. a
- 31. d
- 32. d
- 33. c
- 34. b
- 35. d
- 36. c
- 37. d
- 38. c
- 39. d
- 40. b
- 41. c
- 42. a
- 43. c
- 44. b
- 45. a
- 46. d
- 47. b
- 48. c
- 49. c
- 50. b Page 1
Reactor Operator Answer Key
- 51. a
- 52. a
- 53. c
- 54. d
- 55. a
- 56. c
- 57. a
- 58. a
- 59. b
- 60. c
- 61. d
- 62. c
- 63. c
- 64. c
- 65. d
- 66. b
- 67. b
- 68. b
- 69. a
- 70. d
- 71. b
- 72. a
- 73. a
- 74. d
- 75. a Page 2
.*______I_______
--x--m-3.=-
- x-.=*---
-e." -e----
vs-Material Required for Examination Administration ExamLevel RA MaterialRequiredforExamination Exam section B
00001 1 K202 RWST Tank curve, page 28 of S2.OP-TM.22-0002 1
000037K101 Steam Tables 1
OOWE08K301 EOP-CFST-1 Figure 4A Thermal Shock Limit Curve 1
062000A210 S2.OP-S0.115-0011 Attachments 18 2 2
194001G110 NFS-0231, Core Operating Limits Report, Unit 2, Cycle 14, Figure 2, "Axial Flux Difference Limits as a Function of Rated Thermal Power 3
Tuesday, April 20, 2004 Page I of I
RO Answer Distribution Answer Number of Questions a
19 b
19 C
19 d
18 Tuesday, April 20, 2004 8:32:02 AM
RO Cognitive Level Cognifive Level Number of Questions Application Comprehension Memory 27 15 33 Tuesday, April 20, 2004 8:32:09 AM
Question Source-RO Question Source Modification Method RO Number Facility Exam Bank Direct From Source 5
Facility Exam Bank Editorially Modified 10 Facility Exam Bank Significantly Modified INPO Exam Bank Concept Used INPO Exam Bank Editorially Modified INPO Exam Bank Significantly Modified New New Direct From Source Previous 2 NRC Exams Direct From Source 3
1 4
2 43 2
5
ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet US. Nuclear Regulatory Commission Use the answer sheets provided of the answer sheets. To pass t this cover sheet on top final grade of at least require an 80.00 percent to ination, and three hours if you cant Certification Points Applicant s Grade Percent NUREG-1021, Draft Revision 9 32 of 34
Unit 1 is at 100% power when the following annunciators are received in the control room:
- OHA E-31, PR OVRPWR ROD STOP.
- OHA E-24, ROD DEV OR SEQ.
- OHA E-48, ROD BOTTOM.
- OHA E-39, PR CH DEV.
The following conditions exist:
- PR-N43, N44 indicate I01 % power.
- PR N41 indicates 103% power.
- PR-N42 indicates 97% power.
- All Shutdown Bank and Control Bank group demand counters are at 225 steps.
- IRPI indicates all rods fully withdrawn with the exception of one Shutdown Bank C rod indicating 8 steps.
Which of the following choices identifies the procedure which will be used to address this condition?
iAB-ROD-0002. DROPPED ROD.
.AB-RO D-0 004, RO D P 0 S IT I 0 N IN Dl CAT10 N-FAIL U RE.
- AB-NIS-OOOI, NI SYSTEM MALFUNCTIONS.
06/11/2004 03G404 1
!2.4 Emergency Procedures / Plan conditions for emergency and abnormal operating procedures 55.43(5)
B is correct because a dropped rod in the vicinity of a power range detector will skew the power range indication. The Nls are behaving as expected. There is no IRPl failure based on the indications given. The dropped rod is misaligned, but the stucWmisaligned procedure is NOT correct ifor a droooed rod L Dropped Rod
~-.
_.._.... - ~.
. __~_.~_.
A.
B.
C.
Determine the appropriate abnormal procedure.
Describe the plant response to actions taken in the abnormal procedure Describe the final plant condition that is established by the abnormal procedure I
[ Editorkiy Modified I
Tuesday, April 20, 2004 9:23:12 AM
.J I
Paael of34 I
Unit-2 has experienced a Rx trip and SI during a SBLOCA. These conditions are present after checking PORV's shut in 2-EOP-TRIP-I:
All S/G pressures are 810 psig and stable.
Total AFW flow is 24E4 Ibm/hr.
RCS pressure is stable at 1250 psig.
22 RCP flange vibration is 15 mils.
Containment pressure is 4.6 psig.
RCS Tave's are all - 554 and rising slowly.
All charging pumps have tripped.
21 and 22 SI pumps are running but are NOT injecting to the RCS.
2R2 is in ALARM, 2R7 reads O.OOE+O mr/hr.
Which of the following choices describes the correct action(s), if any, required to be taken, and p roced u re transition points?
- Stop ALL RCPS, transition to ~-EOP-LOSC-I at appropriate step.
Leave ALL RCPs runnina. transition to 2-EOP-LOSC-1 at appropriate step.
Stop ALL RCPs, transition to 2-EOP-LOCA-1 at appropriate step.
Stop 22 RCP ONLY, transition.
to 2-EOP-LOCA-1 at appropriate step. __.......
06/11/2004 00001 56406 2
2.4 IiEmergency Procedures I Plan
/Knowledge symptom based EOP mitigation strategies.
3 1 4 0 is required to be tripped due to its flange vibration exceeding the limit of 5 mils per AB RCP The other 3 RCP's will remain running because ECCS flow cannot be verified. Transition to LOCA-1 at Step 28.1 due to Containment pressure > 4 psig.
In the basis document for EOP-TRIP-1 page 44-45 ifor step 25, discusses what "ECCS flow" is. The setpoint of 100 gpm indicated on the SI pumps
/individual flowmeter is "the minimum SI flow (per the SI pump flow meter) which indicates iniection into
/the RCS". LOSC entry is NOT indicated.
]Reactor Trip or Safetv lniection (Basis)
L........
1 ABRCPl E005 For the following analyzed transientslaccidents.
a) b)
Partial Loss of Forced Reactor Coolant Flow Single Reactor Coolant Pump Locked Rotor and Reactor Coolant Pump Shaft Break
- 1) Determine the expected alarms and indications
- 2) Describe the analysis assumptions
- 3) Describe the protective features that mitigate the event.
- 4) Describe the expected plant response r-.
. Tuesday, April 20, 2004 9:23:13 AM
khanged stem conditions to add RCP malfunction, and changed answers for RCP malfunction
~
~
..- -~
Tuesday, April 20, 2004 9:23:14AM
]
I
~ a a e 3 o f 3 4 I
Given the following conditions :
,- Salem Unit 2 is operating at 100% power when 21 SGFP trips, and the Main Turbine runs back to 160% as expected.
- Operators receive OHA E-16, ROD INSERT LMT LO-LO.
- Steps to initiate Rapid Boration are initiated IAW S2.OP-S0.CVC-0008, RAPID BORATION.
Which of the following choices describes a condition in which Rapid Boration is NOT successfully established IAW S2.OP-S0.CVC-0008, RAPID BORATION?
Charging flow 90 gpm on 2FI-128B 2 CVC CHG SYS FLOW INDICATING BEZEL, 2CV175
,RAPID BORATE STOP VALVE open, boric acid flow 32 gpm on 2FI-113A 2 CVC RAPID
[BORATE FL INDICATING BEZEL Charging flow 85 gpm on 2FI-128BI 2CV174 BLENDER BYP VALVE open, 2CV172 BA FLOW CONTROL TO BLENDER open, boric acid flow 37 gpm on 2FI-I 1 OA 2 CVC BORIC ACID FLOW INDICATING BZL
'Charging flow 80 gpm on 2FI-128BI 2CV175 shut, 2CV172 and 2CV185 M/U FROM BLENDER TO CHG PUMP SUCTION LINE in MANUAL and open, boric acid flow 39 gpm on 2FI-110A.
B'Charging flow 95 gpm on 2FI-I28B, 2CV175 shut, 2SJ1 RWST TO CHG PUMPS STOP MOV
/open, 2CV40 VCT OUTLET STOP shut.
I Igpm. Distractor b,c and d are incorrect because all the conditions are met IAW step 5.2, 5.3, and 5.4
!rewectivelv. of the procedure.
,..Tuesday,.~
a)
LetdownlCharging i)
Letdown lsolaiton Valves, CV2, CV277 ii) Regenerative Heat Exchanger iii) Letdown Orifices iv) Letdown Orifice Isolation Valves, CV3, CV4, CV5 v)
Letdown Releif Valve, CV6 vi) Letdown Line Containment Isolation Valve, CV7 vii) RHR Flow Control Valve, CV8 viii) Letdown Heat Exchanger ix) Low Pressure Letdown Control Valve, CV18 x)
Temperature Control Valve, CV21 xi) Demineralizers (Mixed Bed, Cation, and Deborating xii) Inlet Valve to Deborating Demin, CV27 xiii) Reactor Coolant Filter xiv) Diversion Valve, CV35 xv) CVCS Holdup Tanks xvi) Volume Control Tank I
~ a g e 4 o f 3 4 1
1 CVCSOOE012 xvii) VCT Isolation Valves, CV40, CV41 xviii)
Chemical Mixing Tank xix) Charging Pumps (Centrifugal and PD) xx) Miniflow Recirc. Valves, CV139, CV140 mi) Seal pressure Control Valve, CV71 xxii) Chg. Line Containment Isol. Valves, CV68, CV69 xxiii) xxiv) PZR Auxiliary Spray Valve, CV75 xxv) CCP Flow Control Valve, CV55 RCP Seal Water i) Seal Water Injection Filters ii) Seal Bypass Flow Valve, CV114 iii) Seal Water Return Isolation Valve, CV104 iv) Seal Water Return Relief Valve, CV115 v)
Seal Return Cont. Isol. Valves, CV116, CV284 vi) Seal Return Filter vii) Seal Water Heat Exchanger Excess letdown i)
Excess Letdown Isolation Valves, CV278, CV131 ii) Excess Letdown Heat Exchanger iii) Excess letdown Flow Cotrol Valve, CV132 iv) Excess Letdown Diversion Valve, CV134 Makeup i) Primary Water Storage Tank ii) Primary Water Makeup Pumps iii) Boric Acid Batch Tank iv) Boric Acid Tanks v)
Boric Acid Transfer Pumps vi) Boric Acid Filter vii) Boric Acid Blender viii) Primary Water Flow Control Valve, CV179 ix)
Boric Acid Flow Control Valve, CV172 x)
Charging Pump Suction Valve, CV185 xi) VCT Makeup Isolation Valve, CV181 Charging to Loop 3 Valve, CV77, Loop 4 Valve, CV79 xii) Rapid Borate Stop Valve, CV175 Describe the procedures which govern the operation of the Chemical and Volume Control System. includina sianificant prerequisites and precautions associated with each operating procedure which are required io be considere; b; either Licensed or Non-Licensed
............... Operators...........
LT1...
AM 1
Tuesday, April 20, 2004 9:23:15 I
~ a g e 5 o f 3 4 1
, :Given the following conditions:
I iThe reactor and turbine are tripped which will always ensure operation within design basis iparameters.
i._
/The turbine is tripped, the reactor is NOT tripped, AFW pumps started when both SGFPs
/tripped, ATWS design basis parameters will be maintained even without a Rx trip.
"D f.........................
/The reactor is tripped, the turbine is NOT tripped, thermal shock design basis will be exceeded lif cooldown exceeds I00 deg F/hr.
!The reactor AND turbine are NOT tripped, AFW pumps have started at 9% NR levels in 2/4 SGs, heat sink design basis will be maintained if feed and bleed is established.
~...
- Salem Unit 2 is operating at 100% power.
- BOTH SGFPs trip simultaneously.
- The Reactor Protection System does NOT initiate a Rx trip at ANY time.
- ALL other automatic actions actuate as designed.
- AMSAC tripped the turbine. Distracter a is incorrect because AMSAC does not trip the reactor, and the
- stem states that RPS does NOT trip the rx. Distracter c is incorrect because the reactor is not tripped.
- B is the correct answer because the design criteria of maintaining the RCS below 3200 psig will be met
~
~
J I
I I
I i
I lNew I
1 i Tuesday, April 20, 2004 9:23:16 AM i
J I
Paae6of34 I
.Which of the following choices describes the bases of requiring at least ONE Vital Instrument Bus
,to be energized from its respective inverter connected to its respective DC Bus Train during
- MODES 5, 6 and the movement of irradiated fuel?
'Insure adequate power is available for operation of the Fuel Handling Building exhaust fans.
'Ensure the facility can be maintained in the shutdown or refueling condition for short time periods.
Ensure sufficient instrumentation and control capability is available for monitoring and maintainina the unit status.
Insure capability for communications within the facility and off-site agencies remain available.
1
,3/4.OBases. Distracter a is incorrect because the VIB doesn't supply the FHB exhaust fans, the 230V
- bus does. The stem asks for the bases for the VIB. Distracter b is incorrect because the bases states ithat the requirement is to be able to keep the unit in S/D or refueling condition for EXTENDED periods
!of time. Distracter d is incorrect because it's the bases for the 3.8.1.1 for MODES 1-4.
ISalem Tech Specs
~
I Systems, including:
a) b)
c)
The Limiting Condition@) for Operation The Bases for the LCO(s) (N/A NEO)
The applicability of the LCO(s)
..... d)
The...........
LCO Action Statement(s) (N/A NEO)
/New I
QuestionSourct ~
Question Source Comments: I ~-
-~
x I
Tuesday, April 20, 2004 9:23:17 AM i
J I
Paae7of34 I
.Given the following conditions:
1-Salem Unit 2 is in MODE 3, NOP, NOT.
1-The control room receives OHA B-18 2C 125VDC CNTRL BUS VOLT LO 1-Upon further investigation, the NCO reports that 2C 125VDC bus voltage is at 126 volts, and no
- current is indicated on 2RP9.
,Which of the following choices describes the condition which is present, and the actions required to j be taken?
- IC 125VDC bus is
...... DETECTION.
- above the Tech Spec minimum setpoint, ONLY continued monitoring for any indication of
,further voltage degradation is required.
ibelow the Tech Spec alarm setpoint, secure the operating battery charger and place the (standby battery charger in service.
~
IAA2 --., iAb-
/AA2.021 ll25V dc bus voltage, low/critical low, alarm ility to determine and interpret the following
___ as they apply to Loss of DC Power:
..................................................... p33.61
~
55.43(5) A is the correct answer because the control band as specified in the NCOs logs is 125-139.8V. Voltage is in the normal band, and the AR states to have maint adjust the float voltage.
i I
1 i
!Distractor b is incorrect because there is no indication of a ground. Distracter c is incorrect because
'action IS reauired IAW ARP. Distracter d is incorrect because voltaae is above the TS limit.
..~
IControl
... room Logs
. MODES 1-4 (Overhead Annunciators Window C A.
Batteries B.
Battery Chargers A. The Control Room location of DC Electrical Systems control bezels and indications B. The function of each DC Electrical Systems Control Room control and indication C.
The effect each DC Electrical Systems control has upon DC Electrical Systems components and operation D. The plant conditions or permissives required for DC Electrical Systems Control Room controls to perform their intended function
- i........
I Tuesday, April 20, 2004 9:23:17 AM I
1 Pane8of34 1
- Which of the following conditions, if left uncorrected, will require the Salem Operations Department Salem ECG
,to make an Off-Site notification?
'Unplanned loss of DID and CENTREX phone systems.
~
~A...
iji i single Unit 2 Hydrogen Recombiner declared INOPERABLE in MODE 3.
The location of Control Air System local controls and indications
[--
- Tuesday, April
.. ___ 20, 2004 I
Pagegof34 I
Given the following conditions:
j-Salem Unit I is in MODE 4.
I-21 RHR loop is in service supplying shutdown cooling.
i /Which of the following actions, if not corrected IMMEDIATELY, will lead to a loss of RCS inventory,
/steam binding of ECCS suction piping, a potential failure of all ECCS while shutdown, and potential 1
n MANUAL close.
Opening of 2RH21 RHR TO RWST STOP VALVE.
Opening 2SJ69 RHR SUCTION FROM RWST with BOTH 2RH1 and 2RH2 RHR COMMON SUCTION VALVES open.
Continued operation of the 21 RHR loop in shutdown cooling when RCS WR Thots rise above I 350 deg F.
T l I E m e r g e n c y Procedures / Plan 2.4.49
[Ability to perform without reference to procedures those actions that require immediate operation I [-4:a r4:0]
of system components and controls L.....................................
__............................................. __I
!55.43 (4) 55.43(5) S2.OP-SO.RHR-0001 P&L 3.23 states that 2RH21 shall not be open in MODE 4
!with RHR in shutdown cooling, for the aforementioned reasons found in the stem. Distracters a and d
!would have the effect of rising temperature, but occur over a period of time and do not require immediate mitigating action. Distracter c is incorrect because we routinely operate in this condition.
I NRC Generic letter 98-02 address this loss of coolant outside containment.
i
/Initiating RHR I
RHROOOEOI 3 i For plant or industry events associated with the Residual Heat Removal System a)
Summarize each event b) c)
d)
Station (N/A NEO)
Identify the root cause@) of the event (NIA NEO)
Describe the events likelihood of occurrence at Salem Nuclear Generating Station (N/A NEO)
Describe established or alternative actions which might prevent the events (re)occurrence at Salem Nuclear Generating I
Tuesday, April 20, _ _
20049.23 I
Given the following conditions:
- Salem Unit 1 was operating at 100% power when a Loss of Feedwater occurred.
- The crew is performing actions of EOP-FRHS-I, LOSS OF HEAT SINK.
- Following trip of the RCPs, RCS pressure is slowly rising. The pressure rise is terminated by PZR PORV operation.
- Steam Generator Wide Range levels are all currently 37% and lowering at approximately 1 %
every 5 minutes.
- Attempts to restore feed have been unsuccessful.
Which of the following parameters would indicate the need to immediately establish Bleed and Feed cooling?
/A rapid drop in any SG pressure.
'A I rapid drop in RCS pressure :
..................... I A high core Delta T.
A low core Delta T.
~
,EA2.2 /Adherence to appropriate procedures and operation within the limitations in the facility's license 1 14.3 and amendments.
55.43(5) to loss of heat sink. A high delta T and increasing RCS pressure are due to the normal characteristics of natural circulation setup. A drop in RCS pressure is not cause for bleed and feed, and could potentially result in transition to LOCA-1. A MSLB on one SG will only cause one SG level to go below the threshold for bleed and feed. 43.5 because this requires the SRO to make a judgment call regarding the implementation of the bleed and feed portion of FRHS-1. The SRO must understand what constitutes a rise in RCS pressure due to loss of heat sink versus a rise in pressure caused by Nat Circ setting up.
A low core delta T indicates that heat is not being removed, and that the pressure rise is due ILoss of Heat Sink I
~
IDirect From Source I
Page11 of34 I
While Unit 2 was operating at 100% power, a LOCA occurred. The crew is now in EOP-LOCA-1.
The following conditions exist:
- All rods are fully inserted.
- No RCP's are operating.
- 8 CETs are reading between 720-790 degrees.
- Containment pressure is 31 psig.
- Containment sump level is 55%.
- RWST level is 17 ft.
!- RVLIS indicates 37%.
- RCS pressure is 265 psig.
- Only 2A 4KV Vital Bus is energized.
i FRCCOOEOOI
- Which of the
......... following
.............. procedures must
..... be implemented?
@!!l EOP-FRCE-1,
RESPONSE
TO EXCESSIVE CONTAINMENT PRESSURE.
'EO IEOP-FRCC-1,
RESPONSE
TO INADEQUATE CORE COOLING.
I
[EOP-FRCC-2. RESPONSE TO DEGRADED CORE COOLING.
State the Red paths for the core cooling status tree k c i
/Emergency and Abnormal Plant Evolutions E07 i /Saturated Core Cooling
- EA2]'Ability
~.
to determine
......... and interpret
.- the
.... following as they
.. apply. to... Saturated Core Cooling:
,EM.
1
/Facility conditions and selection of appropriate procedures during abnormal and emergency 55.43(5) Using the Basis Docum (NO-Stem)), RCS subcooling > 0 (NO, RCS pressure of 265 psig corresponds to sat temp of -400 deg),
Is any RCP running (NO-stem), 5 or more CETs > 700 deg (YES-stem) RVLIS > 39% (NO-stem), GO
/TO FRCC-1 due to the RED PATH. All the distracters contain the wrong procedural transition.
I Critical Safety Function Status Trees
~.-
c 111.:.........
1 Tuesday, April 20,2004 9:23:21 AM 1
Page12of34 1
- Given the following conditions on Unit 2:
rEA2.2 '
- A LBLOCA has occurred.
- Operators are performing 2-EOP-LOCA-5 LOSS OF EMERGENCY RECIRCULATION.
,- Containment pressure is 15.1 psig and is rising slowly.
'Which one of the following correctly describes how the Containment Spray system will be operated, and why?
/The Containment Spray System is operated as directed in...
i Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.
(/13.
,LOCA-5 because it establishes minimum required containment spray flow and conserves iRWST inventorv.
LOCA-5 since FRPs are NOT implemented durina the performance of LOCA-5.
~
ci 2-EOP-FRCE-1 "RESPONSE TO EXCESSIVE CONTAINMENT PRESSURE" since irestoration of the critical safetv function takes precedence.
,2-EOP-FRCE-I
............... because actions concerning Containment Spray operation are more restrictive. 1
[:.:-__
]
Wednesday, April 21, 2004 7:28:25 AM
Given the following conditions:
,2.4.4
- Salem Unit 2 is in Mode 6.
- - A bowed fuel assembly is being raised from its location in the reactor vessel.
- Gas bubbles start rising to the surface of the refueling cavity directly above the fuel assembly ibeing raised.
1-Radiation in the Containment Building area starts rising steadily.
1-The fuel assembly being raised can NOT be inserted back into the core.
Ability to recognize abnormal indications for system operating parameters which are entry-level j4.0)131 conditions for emergency and abnormal operating procedures.
'Which of the following choices identifies the procedure which will be entered FIRST in response to
,these
__. conditions?
i S2.0 P-AB. F U E L-0 00 I F U EL H ANDLING-INC I DENT.
coolant, AB.RC-2 is only applicable in MODES 1-5.
/Fuel Handlina Incident I
I
~
~.
- ~-
(Material Required fo-p--2~'-A*-- ' '
I Tuesday, April 20, 2004 9:23:23 AM I..........
J I
Page 140f 34 I
Given the following conditions for Unit 2:
- SI has actuated due to a large break LOCA
- RWST LEVEL LOW alarm has actuated
- Once armed, 22SJ44 RHR Pump Suction Valve to Containment Sump did NOT open.
4AW I
2-EOP-LOCA-3 TRANSFER TO COLD LEG RECIRCULATION, which of the following
/identifies an operation that requires stopping 22 RHR pump in order to complete the switchover to I
I
/Cold Leg Recirculation?
'Closing 22SJ45 RHR to Charging/SI Pump Suction Valve.
Closing22CS36
__ RHR System to CS System Isolation
__ Valve.
Closing 22RH4 RWST to RHR Pump Suction Valve.
1
'VI Ability to (a) predict the impacts of the following on the Residual Heat Removal System and (b) based on Ithose predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
Iclosing RH4 and opening SJ44. Distracter d is incorrect because closing the RH1 9 will not require I
.... J
'stopping the RHR pump.
I..
Transfer
.............. to Cold Leg
... Recirculation I
I
... ~-..... -.....
[:I:..
Tuesday, April
..... 20, 2004 92324 AM
]
I Paae 15of34 I
Given the following conditions:
I
/Which of the following choices identifies how the ECCS system is affected by the deenergized DC
/bus, and what action(s), if any, is/are required to mitigate this effect?
2A SEC will initiate in Mode 1, but some ECCS equipment will NOT start. After at least I
.minute has elapsed since SI initiation, reset SI, reset 2A SEC, start ECCS pumps from the Control Room that did NOT start automatically.
- The loss of 2A 125VDC bus will affect indication only. All ECCS equipment will start upon the SEC Mode 1 signal due to the redundant nature of having 2 separate logic trains.
'All ECCS equipment will start as required except the Containment Spray pumps. If containment pressure rises to > 15 psig, CS pumps will NOT automatically start due to the energize-to-actuate feature, and must be manually started.
Some ECCS pumps will NOT start, and operators will be unable to start those pumps from the control room. Local operation will be required to start those pumps.
and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
E3.7114.21 55.43(5)
SSPS logic cabinets is provided from 115VAC vital instrument power. As such the actuation of SSPS The loss of the 2A 125VDC bus will not affect the SSPS cabinets. The dc power used in-the belays will not be affected. In tandem, the 2A SEC cabinet will not be affected by the loss of dc either. It will perform its function and send output signals required for a MODE 1 actuation. The dc power loss is felt at the actual equipment breakers, which need dc power to close and open (energize close and trip
/coils) The stem of the question specifies that NO action has been taken since the loss of the dc bus, so
,the control power for the busses supplying these breakers HAS NOT BEEN SWAPPED to b/u source.
- Without control power, remote operation of the breakers is unavailable, whether form the control room
/or the SEC. Local manual oDeration is the onlv wav to operate these bkrs.
Reactor Trip or Safety Injection 1
ECCSOOE008 1 Identify and describe the Control Room controls, indications, and alarms associated with the Emergency Core Cooling System, including:
a) b)
c) operation (N/A NEO) d)
intended function e)
The Control Room location of Emergency Core Cooling System control bezels and indications (N/A NEO)
The function of each Emergency Core Cooling System Control Room control and indication (N/A NEO)
The effect each Emergency Core Cooling System control has upon Emergency Core Cooling System components and The plant conditions or permissives required for Emergency Core Cooling System Control Room controls to perform their The setpoints associated with the Emergency Core Cooling System control room alarms
[:---.Tuesday, April 20, 2004 9:23:26 AM I
Page 17of34 I
Given the following conditions:
- Salem Unit 2 is operating at 100% power.
- 23 & 25 CFCUs have been C/T for emergent corrective maintenance for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- A crew of 5 people are inside containment investigating an increase in the RCS leakrate, with a
,Heat Stress stay time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
/- 22 CFCU breaker trips.
1 130 minutes after the crew entered containment, the on shift Radiation Protection Technician reports
- that 2R12A CONTAINMENT GAS EFFLUENT is reading double what it was when the crew entered containment, and wants the control room staff to have all personnel exit containment.
I I____.__
!ANY increase in radiation levels in containment while it is occupied requires dispatching a
- Radiation
.. Protection technician into containment to evacuate containment.
Contact the crew in containment by flashing the containment lights, and direct them to exit the contain men t.
'Since the 2R12A is expected to rise with an RCS leak, the crew may remain in containment luntil their Heat Stress stay time is complete.
- Which of the following choices identifies how control room personnel will respond IAW SC.SA-
.ST.ZZ-0001 SALEM CONTAINMENT ENTRIES IN MODES 1 THROUGH 4?
/rise in radiation level. If radiation levels on 2R12A increase by a factor of 4 from original level,
/use i __
the.- page system to direct
..... personnel in containment to exit.
rise in RMS data, shall prohibit any subsequent entries to containment and DIRECT the control room to contact any work parties and have them exit containment. IAW SC.SA-ST.ZZ-0001, 3.2.1...." The containment lighting, when flashed, is the preferred method the Control Room uses for requesting Tuesday, April 20, 2004 9:23:26 AM
]
l:.
I Paae18of34 1
~. _
[:..... Tuesday, April 20, 2004 9:23:27 AM
]
I
~ a a e 1 9 o f 3 4 I
Given the following conditions:
- Salem Unit 2 is in MODE 6.
- The Spent Fuel Pool (SFP) Gate Valve is open.
- A spent fuel assembly is being raised from the reactor vessel.
- The fuel transfer cart is at the SFP.
- Operators in the control room receive OHA C-35 SFP LVL LO.
- The NE0 at the SFP reports that level is just below the low level alarm point and dropping slowly.
- Which of the following choices describes the impact if SFP level drops 8 feet from normal, and
- what action will be performed in S2.0P-AB.FUEL-0002 LOSS OF REFUELING CAVITY OR SPENT FUEL POOL LEVEL to mitigate the consequences?
~.................
/Reduction in the amount of Iodine scrubbed by the water prior to bubbles reaching the surface, linitiate makeup from WHUT to SFP.
/RHR pump suction vortexing leading to possible loss of RHR cooling, close the SFP Gate
'Valve.
L 1
~~~~~~
~
~~
~
I
- Loss
-.. of Spent
__ Fuel Pool cooling from uncovering suction piping,............ stop
.. ALL SFP pumps.
Rising radiation on FHB area radiation monitors 2R5 and 2R9, initiate S2.OP-AB.RHR-0002 LOSS OF RHR AT REDUCED INVENTORY.
A2.1 Ability to (a) predict the impacts of the following on the Spent Fuel Pool Cooling System and (b) based on
~
~
those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal loperation:
-1
[3.1]1 1
55.43(7), 55.43(5) C is correct because the SFP suction piping is located - 4' below normal water level to prevent gravity draining pool, and the action of stopping SFP pumps is contained in step 3.53 of AB.FUEL-2. Distracters a & d both have possible effects, but both actions in these distracters are wrong, in (a) because m/u is NEVER from WHUT, and in (d) initiation of AB.RHR-2 would only be performed if at reduced inventory of 4 01'. Distracter b effect will not occur with 15' of water still over the Rx vessel flanae. and the closina of the aate valve is a correct action.
I I
IL oss of
........ Refueling
__.. Cavity
-..... __....... or - Spent Fuel
............. Pool.-
Level J
[Spent Fuel Coolina dwa
~
ABSFOIEOOI Evaluate the relationship of the following systems as they apply to a loss of spent fuel cooling:a. Service Waterb. Component Cooling Waterc.
Fuel Handling Building Ventilation Tuesday, April 20, 2004 9:23:27 AM I
1 I
.Which of the following choices describes the Tech Spec Bases for NOT having to declare a
!component INOPERABLE solely because the EDG capable of supplying its Vital Bus is il NO PERABLE?
~
k3,The time the plant is exposed to a LOCA event with less than two full trains of ECCS
,equipment available is limited by the ACTION time associated with each individual structure, Isvstem or component.
'Only ONE Train of ECCS components is necessary to prevent core damage during a DBA
- INOPERABLE will ensure one complete train operates.
limits.
!The ACTION statements which permit limited variation from the basic requirements are accompanied bv additional restrictions which are more restrictive than the oriainal criteria.
As long as all required redundant systems and components are OPERABLE, a loss of off-site power will NOT result in a complete loss of safetv function of critical svstems.
April 21, 2004 7:33:46 AM 1
,Given the following conditions:
1
! - Salem Unit 2 is operating at 100% power when OHA C-I GAS ANLY TRBL is received in the control room.
- The NE0 sent to investigate reports local alarm B-3 OXYGEN HIGHILOW on Waste Disposal
/Gas Analyzer PNL 110 is in alarm.
- Local indication for in service Waste Gas Decay Tank (WGDT) 0 2 concentration is 2.1%.
'Reduce the oxygen concentration of the in service WGDT to prevent potential releases of jlmmediately suspend all additions to the in service WGDT to prevent the concentration of potentially explosive gas mixtures from exceeding the flammability limits of hydrogen and oxygen.
radioactive
- materials
__ - --- - - due
--- ____- to -- explosion
- - - - of the GDT.
IAW Tech Specs, which of the following choices describes what action is required to be performed,
/and why?
Perform an inert gas (N2) addition to the waste gas system to lower 0 2 concentration to c 2%
jbv volume since H2 concentration is alwavs assumed to be > 4% in GDT's.
/!!! Refer to Table 1 H2/02 FLAMMABILITY CONCENTRATION PRESSURE CURVE of S2.0P-SO.WG-0003 GASEOUS WASTE DISPOSAL SYSTEM OPERATION to determine at what
,rate 0 2 concentration needs to be reduced to prevent possible explosive mixture.
55.43(2) 55.43(5) Tech e waste gas holdup system to be <2%. The action
/REQUIRED only when >4% 02. The reason is correct.
a real procedure, and a correct reason. C is the correct answer because the Tech Spec REQUIRES the reduction of 0 2 from 2-4% to less that 2% within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Also, the bases section for this tech specs describes that a potential explosion and release of radioactive materials from this explosion
'would not be IAW GDC 60, 10CFR50 Appx. A.
IN2 will raise the total volume of the WGDT and lower the 02 concentration, the H2 concentration is
- monitored, and only assumed to be > 4% when monitoring is unavailable.
<2% in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Distracter b refers to a non-existent Table in Distracter a is incorrect because while the addition of rn local controls to perform their intended I a)
The Limiting Condition(s) for Operation 1 Page22of34 I
.~
Tuesday, April 20, 2004 9:23:30 AM 1::::
1 Paae23of34 1
Given the following:
1-Salem Unit 2 is in MODE 6 on day I 9 of a refueling outage.
I-Fuel movement is NOT in progress.
- - S/G nozzle dams are installed.
i-A secondary side hydro is being performed on 22 S/G
'While taking Control Room logs IAW S2.OP-DL.ZZ-0002, the RO observes the following:
i 1
!- RCS temperature is 123 degrees.
- RHR loop 21 is in service providing decay heat removal.
- 21,23,24 S/G levels are 75% wide range.
,- 22 S/G level is 100% wide range, and pressure is 210 psig.
1-21,23,24 S/G metal temps are all between 85-90 degrees.
- 22 S/G metal temp is reading 68 degrees.
ilAW I...
Tech
- Specs, what action is required based on these
- conditions, and
._ why?
- Reduce 22 S/G pressure to less than or equal to 200 psig within 30 minutes to ensure that S/G
/pressure induced stresses remain less than allowable limits.
- Raise 22 S/G secondary side temperature to > 70 deg F within 15 minutes to ensure that S/G ipressure induced stresses remain less than allowable limits.
'Raise 22 SIG primary side temperature to > 70 deg F within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to prevent exceeding
/fracture
_._-- toughness stress limits
~ for
~ _ the S/G.
~
~~~
~
I 12.1 1 Conduct of Operations I
2.1.lj. [Knowledge of less than one hour technical specification action statements for systems.
1 1_3?01 3.81 jstress limits." The action time for the TS is 30 minutes. The correct answer of b contains both the action required, the right time period, and the correct reason. All the distracters contain the correct
- reason
--. -. - written
-- 2 -.-
different
-- ways,
__- but either
have
.. the wrong time
_-__ period
-- or ----..
the ---
wrong
---. action.
I alem Tech Specs 3.7.2 IS r
L I L.
nd ControlGrade Interloc& -
1:::
Wednesday,
~.
I.
Red and Purple Paths K.
Stearnline Isolations L.
Feedwater Isolations M.
Feedwater Interlock N.
Key Relief Valves
- 0.
Tank Thumbrules J.
TRIP-I CASS State the Technical Specifications associate onents, parameters, and operati enerat Blowdown and Drain Systems, including:
a) b)
c) d)
The Limiting Condition(s) for Operation The Bases for the LCO(s) (N/A NEO)
The applicability of the LCO(s) (N/A NEO)
The LCO Action Statement@) (N/A NEO)
_ _ _ ~
[-..-.-Wednesday, April 21, 2004 7:36:41 AM ]
I Page25of34 I
Which of the following choices describes a condition which is considered Minor Maintenance IAW
.- NC.WM-AP.ZZ-0001 WORK MANAGEMENT PROCESS?
Yearlv insr>ection of the Auxiliarv Buildina elevator.
, PROCEDE002 i
21 Heater Drain wmr> motor oil chanae at 100% Dower.
1 Given a list of purposes, select the purpose of the following types of procedures in accordance with NC.NA-AP.ZZ-0001 (a),
Nuclear Procedure System: SELECT the purpose of the following types of procedures in accordance with NC.NA-APZ-OOOl(Q), Nuclear Procedure System:
- a.
Abnormal Operating Procedures
- b.
Administrative Procedures
- c.
Alarm Response Procedures
- e.
Emergency Operating Procedures
- f.
Integrated Operating Procedures
- g.
In-service Test Procedures
- h.
Operating Procedures
- i.
MMlS Work Standards
/Regularly L..-.___
...... scheduled
.............. - calibration
___.... of N41 power range. nuclear instrument.
- Welding repair of a snubber in the outer mechanical penetration area.
E li 3
NTENANCE CRITE not be considered minor maintenance. Distracter b irequires a power reduction from 100% power. Distracter c requires entry into an LCO. Distracter d is
'welding on a safety related SSC. All distracters do not meet criteria in att 1, Work Management Process 1
J II Tuesday, April 20, 2004 9:23:31 AM I.......................
J 1 Page26of34 I
Given the following:
I
'- Salem Unit 2 is operating at 90% power.
- Power is being raised at 10% per hour.
- With Rod Control in AUTOMATIC, and control rods NOT moving, operators receive OHA E24
'ROD DEV OR SEQ.
I-Control Bank D Group 1 demand is 21 1 steps.
- Control Bank D Group 2 demand is 210 steps.
- Control rod 2D2 indicates 197 steps on the P-250 computer.
i-Control rod 2D2 indicates 208 steps on IRPI.
1-All other Control Bank D rods indicate between 208-215 steps.
ilAW Salem Tech Specs, which of the following choices identifies the condition described by the
- above i-indications, and what action, if any, -__.
that
. is ___.
required?
,Rod 2DZis >I2 steps below its Bank Demand, if after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> soak rod does NOT return to 112 step deviation, action must be taken IAW TSAS 3.1.3.1 MOVABLE CONTROL
'ASSEMBLIES GROUP
- ......... HEIGHT.
Control Bank D Group demand deviation is outside the allowable band, action must be taken
/IAW
- 8.
TSAS 3.1.3.2.1 POSITION INDICATING SYSTEMS OPERATING.
- Rod 2D2 is >I2 steps below its Bank Demand, actions must be taken IAW TSAS 3.1.3.5
- CONTROL ROD INSERTION LIMITS.
!Control Rods are all within the Bank Demand to Individual Rod deviation setpoints, but 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
'control rod position verification readings must be taken due to OHA E-24 being INOPERABLE.
I Salem1 & 2
]
0611 1120041
~.
'd 1 s
- within the +/- 12 step limit, and according to the Tech Spec Bases for MOVABLE CONTROL
'ASSEMBLIES, EITHER the control console indication OR P-250 computer is sufficient to comply with
/spec (Page B 314 1-5 halfway down the page) Since ONE of the TWO indications is SAT, NO ACTION IIS REQUIRED, since the TS is not being entered. Distracter a is incorrect because it is the correct action if > 85% power and a VALID deviation exists.
Distracter b is incorrect because the +/- 2 step IGDC deviation is not present. Distractor c is incorrect because AB ROD-1 will not be entered because Control Rod Drive Mechanism (CRDMj c)
Rod Drive MG Sets..
Page 27 of 34
[I:
...... Tuesday, April 20, 2004 9:23:32 AM
Reactor Trip and Trip Bypass breakers Reactor Control Unit Power Cabinets Logic Cabinet components:
Pulser ii) Master Cycler iii) Slave Cyclers iv) Bank Overlap Unit DC Hold Cabinet Rod Position Indicator (RP Coils Signal Conditioning Modules Pulse-to-Analog (P to converters Rod Bottom Bistables Rod Insertion Limit Comparator n).
Step Counters State the Technical Specifications associated with the components, parameters, and operation of the Rod Control and Position Indication Systems, including:
a) b)
RODSOOE013 The Limiting Condition(s) for Operation The Bases for the LCO(s)
~ 1
..... d)
The LCO Action Statements@)
I r-
. ]
April 20, 2004 9:23:33 AM I
Paae28of34 I
!Given the following conditions:
1-Salem Unit I is in MODE 5.
- - The control room is preparing to perform a containment purge IAW S I.OP-SO.WG-0006, CONTAINMENT PURGE TO PLANT VENT.
- 1 R41 D, Plant Vent Noble Gas Release Rate, has failed high and caused a Containment Ventilation Isolation (CVI) signal.
- 1 R1 IA, Containment Particulate Monitor, is blocked.
- 1 R12A, Containment Gas Effluent, and 1R12B, Containment Gas Effluent-Iodine, are OPERABLE and NOT blocked.
Which of the following choices identifies whether or not the containment purge can be started IAW S I.OP-SO.WG-0006, and why?
The Containment Purge.....
'CANNOT be started until the R41 is repaired or jumpered to allow reset of CVI signal.
- CANNOT
... be started.
because
.-. I R41A, 1._
R41
. D AND 1 R12A must ALL -.-----.
be OPERABLE.
can be started after blocking both trains of I R41, and resetting both trains of CVI.
can be started after resetting CVI, since the CVI signal can be reset without clearing the activation signal.
940.0.1..G~3.0~g....
I_::-_
- an INOPERABLE R41D, it can be blocked and reset IAW step 5.1.10.C-F, and the purge can commence.
Distractor a is incorrect because the procedure allows reset of CVI by blocking the
- enough to meet ODCM 3.3.3.9 requirements of Table 3.3-13. Distractor d is incorrect because the CVI CANNOT
/______
be reset until the high rad signal....... is cleared/blocked per logic diagram 221057 i
lcontainment Purge to Plant Vent
[.........................................................................................................................................................................................................................
RPS Safeguards Actuation Signals I
Identify and describe the Control Room controls, indications, and alarms associated with the Radioactive Waste Gas System, including:
WASGASE012 a) b)
c) operation d) intended function Describe the procedures which govern the operation of the Radioactive Waste Gas System, including significant prerequisites and precautions associated with each operating procedure which are required to be considered by either Licensed or Non-Licensed Ooerators The Control Room location of Radioactive Waste Gas System control bezels and indications The function of each Radioactive Waste Gas System Control Room control and indication The effect each Radioactive Waste Gas System control has upon Radioactive Waste Gas System components and The plant conditions or permissives required for Radioactive Waste Gas System Control Room controls to perform their
........ e) -. The setpoints associated
. with
.. the Radioactive Waste Gas
........ System..............
control room alarms Tuesday, April 20, 2004 9:23:33 AM 1
L I
Page29of34 1
Tuesday, April 20, 2004 9:23:34 AM
]
c:
1 Page30of34 1
- Given the following conditions
.- Salem Unit 2 is in MODE 6.
'- 32 Fuel bundles have been transferred to the Spent Fuel Pool.
!- While being raised from its location in the reactor, a spent fuel assembly has been damaged, and
/bubbles are coming to the surface.
- - The assembly can NOT be moved further in either direction.
- Radiation levels in containment are 500 mrem/hr.
I Fuel Handling Incidents
/Which of the following choices identifies the action that the crew will perform before the other
'actions IAW
... S2.OP-AB.FUEL-0001, FUEL HANDLING INCIDENT?
1
]Evacuate ALL personnel from containment.
Evacuate non-essential personnel ONLY from containment.
/Close
............... the Fuel Transfer Canal Gate Valve.
- Initiate Containment Closure IAW S2.OP-AB.CONT-0001, CONTAINMENT CLOSURE.
Wednesday, Aprll21, 20047 I
Page31 of34 I
Given the following conditions:
I-Salem Unit 1 is operating at 100% power.
- OHA G-24 TAC TEMP HI OR LO alarms and clears.
- The OHA alarms and clears - once every minute.
1-The Secondary NE0 reports that TAC system temperatures are in the middle of the control band and steady. He also reports no abnormal condition is apparent with the TAC system.
/The RO requests permission to block OHA G-24.
I
- Which of the following choices contains ONLY the actions and procedures that will be performed
- during the process to block the alarm?
,Initiate a Level 2 significance Notification (NOTF), place an INFO sticker on the OHA box, log Ithe notification number in.NC.DE-AP.ZZ-0030, "CONTROL OF TEMPORARY
'MODIFICATIONS".
Initiate a Level 3 significance NOTF, place a single strip of red translucent tape diagonally
'across OHA box, log the notification number in Attachment 1 of SC.OP-DL.ZZ-0010 CONTROL ROOM INSTRUMENTATION AND ALARMS hitiate a Level X significance NOTF, create a NU-IND task for an Operator Burden IAW NC.WM-AP.ZZ-0000 NOTIFICATION PROCESS, and place an INFO sticker near the OHA
'window box."
Initiate a NOTF, place an INFO sticker in the Alarm Response Procedure for that OHA, place 2 30, "OPERATOR BURDEN PROGRAM.
,2.4.33 ! /Knowledge of the process used track inoperable alarms.
0010. Distractor B is incorrect because for a completely INOPERABLE OHA, section 5.1.3 of AP.ZZ-30 states to place 2 strips diagonally in an X across the window box. Distractor C is incorrect because a significance level X is used for enhancements and non-plant affecting systems IAW NC.WM-AP.ZZ-10000, Notification Process, Att. 1, Classification Guidance, and also because the info sticker will be placed in the ARP, NOT near the OHA box.
I Control Room Instrumentation and Alarms Operator Burden Program Standards Reactivity Management
- 1.
industrial Safety Practices
- 2.
Radiation Worker Practices
- 3.
Conservative Decision Making
- 4.
- 5.
Communication
- 6.
Shift Relief and Turnover Control Room 6At the Controls Area6 1
Page32of34 I
-~
- 7.
Procedure Use and Adherence
- 8.
Alarm Response
- 9.
Operator Appearance
- 10.
Self AssessmenVCorrective Action
- 11.
- 12.
Housekeeping
- 13.
Operator Rounds
- 14.
Briefs
- 15.
Human Error Reduction Techniques
- 16.
Log Keeping
- 17.
Training
- 18.
Supervisor Involvement
- 19.
Climbing on Equipment, Shift Briefing Format, Pre-Job Briefing Guidelines, Pre-Job Briefing Checklist, Pre-Job Briefing Checklist (Tagging), Human Performance, Top Ten Human Error Traps PlanffControl Board Awareness and Maintenance of Critical Parameters J
fied Tuesdav. Aoril 20. 2004 9:23:36 AM 1
An event has occurred resulting in a Site Area Emergency declaration.
iHealth Physics assistance is required for a task being performed in the Auxiliary Building.
l ~ h i c h one of the following Emergency Response Facilities will supply Radiation Control personnel
[to support the task?
~ __._
IEmergency Operations Center.
'Operations Support Center.
'Technical Sumort Center.
1
/Control Room.
c is incorrect because the TSC provides technical guidance. Distractor a is incorrect because it is located off-site. Crews are sent out from OSC to Derform whatever work is necessarv in the Dlant.
I
[Shift Radiation Protection Technician Response I
i..............
I J
~
~~
Senior Reactor Operator Answer Key
- 1. b 2, d 3. a 4. b 5. c 6. a
- 7. d
- 8. b 9. d
- 10. c
- 11. a
- 12. a
- 13. c
- 14. d
- 15. c
- 16. c
- 17. d
- 18. c
- 19. b
- 20. a
- 21. d
- 22. c
- 23. b
- 24. d
- 25. b Page I
I rnU Material Required for E;raminati~n~Administration ExnmLevei K4 MaterialRequiredfo rExamination Exam section Wednesday, April 21,2004 Page I of I
SRO Answer Distribution Answer Number of Questions Tuesday, April 20, 2004 9:24:23 AM
SRO Cognitive Level Cogmitive Level Number of Questions Application Comprehension Memory 6
7 12 Tuesday, April 20, 2004 9:24:29 AM
Question Source-SRO Question Source ModiQkation Method SRO Number Facility Exam Bank Direct From Source Facility Exam Bank Editorially Modified Facility Exam Bank Significantly Modified INPO Exam Bank Editorially Modified New 2
3 4
1 15
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20040407 UNIT 2 REFUELING WATER STORAGE TANK 4
40 36 ti30 2
z 4
- 2.
J n
W I-
$ 2 0 E
z 15 10 5
0 CONTAINED VOLUME IN GALLONS X 1000 Salem Unit 2 Page 28 of 33 Rev. 7
3000.
2700 2400 2100 c3 1800 I
' m
- a.
W E
3 m
m W
CIL Q
5 00 2 00 900 600 300 0
FIGURE 4 A THERMAL SHOCK L I M I T A CURVE PTS PLANT OPERATIONAL LIMITS CURVE 2560 PSIG L I M I T A-TI-I 230°F 280°F 1
I 1 i o 200 TEMPERATURE O F EOP-CF ST-1 300 400 SALEM UNIT 2 REV. 23
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20040407 NFS-023 1 Revision 1 March 2004 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 14 COLR FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED TKERMAL POWER Page 11 of 12 40 50
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20040405 S 2. OP-SO. 115-0011 (Q)
DEVICE 2AVII2A1 (CB201) 2AVII2A2 (CB25 1)
RECTIFIER DC OUTPUT ammeter DC BUS voltmeter DC INPUT ammeter ATTACHMENT 2 (Page 1 of 3) 2A VITAL INSTRUMENT BUS UPS AS-FOUND STATUS SHEET STATUS 50 amps I36 volts 0 amps 1.0 2A INSTRUMENT BUS RECTIFIER PANEL:
BLOWN F251 lamp (clear lens)
BLOWN F252 lamp (clear lens)
ON @
BLOWN F1 lamp (clear lens)
DC BUS AVAILABLE lamp (red)
DC INPUT AVAILABLE lamp (red)
REG. AC OUTPUT AVAILABLE lamp (red)
INV. AC OUTPUT AVAILABLE lamp (red)
NORMAL AC INPUT AVAILABLE lamp (red}
DC BUS LOW/FAIL lamp (white)
DC INPUT LOW/FAIL lamp (white)
REG. AC OUTPUT LOW/FAIL lamp (white)
Salem L Page 46 of 51 Rev. 11 ON
@ OFF ON @
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20040405 COMMON LOW AIR FLOW lamp (white)
LOW AIR FLOW lamp (white)
(1)
(2)
Cabinet Internal BLOWN FUSE lamps (record fuse number(s) of any illuminated lamp(s):
S 2.OP-SO.115-0011 (Q)
ON (OFF)
ON GFF)
OFF W
1 DEVICE STATUS 2AVII2A5 (CBlO1)
N)OFF TRIPPED CHARGE ON toggIe switch CHARGED Iamb (red)
LOW AIR FLOW lamp (white)
BLOWN F102 lamp (clear lens)
BLOWN F103 lamp (clear lens)
Cabinet Internal BLOWN FUSE lamps (record fuse number@)
of any illuminated lamp(s):
a F
T (1)
If lamp is ON, then ensure at least one fan in each energized panel is operating and associated filter is clean.
(2)
If lamp is ON, then ensure at least one fan in the panel is operating and associated filter is clean.
Salem 2 Page 47 of 51 Rev. I1
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20040405 S 2. OP-SO. 115-0Oll(Q)
ATTACHMENT 2 (Page 3 of 3) 2A VITAL INSTRUMENT BUS UPS AS-FOUND STATUS SHEET 3.0 2A VITAL INSTRUMENT BUS REGULATOR & STATIC SWS PANEL:
STATIC SWITCH ON INVERTER lamp (red) rEST TRANSFER toggle switch STATIC SWITCH ON ALTERNATE lamp (white)
DEVICE STATUS 2AVI12A3 (CB301)
OFF TRIPPED 2AVII2A4 (CB 15 1)
OFF TRIPPED MAN.BYPASS switch (record position)
~ C U f i $ I L.
OFF ON ~ F F >
~~~
I1
-fn amps AC LOAD ammeter I
I i 20 volts II AC OUTPUT voltmeter Hertz AC OUTPUT frequency SOURCE SELECTOR switch LINE LOW AIR FLOW lamp (white)
(1)
~
SYNCHRONIZED lamp (red)
@ OFF SYNCH MONITOR lamp (clear Iens)
REG. AC INPUT AVAILABLE lamp (red)
BLOWN F151 lamp (clear lens)
BLOWN F152 lamp (clear lens)
\\ /
RETURN MODE switch Jabinet Internal BLOWN FUSE lamps (record use number(s) of any illuminated lamp(s):
(1)
If lamp is ON, then ensure at least one fan in the panel is operating and associated filter is clean.
Salem 2 Page 48 of 51 Rev. 11
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20040407 s2.OP-so. 115-00ll(Q)
DEVICE 2AVIIZA1 (CB201)
ATTACHMENT 1 (Page 1 of 3) 2A VITAL INSTRUMENT BUS UPS STATUS SHEET STATUS ON STATUS ON STATUS ON NORMAL ALT AC DC SOURCE SOURCE (1)
SOURCE ON OFF OFF 1.0 2A INSTRUMENT BUS RECTIFIER PANEL:
DC INPUT ON lamp (red)
BLOWN EO1 lanip (white)
ON (3)
OFF ON OFF OFF OFF
~
~
2AVI12A2 (CB25 1)
BLOWN F251 lamp (clear lens)
BLOWN F252 lamp (clear lens)
I ON 1
OFF 1
ON OFF OFF OFF OFF OFF OFF BLOWN F1 lamp (clear lens)
DC BUS AVAILABLE lamp (red)
DC INPUT AVAILABLE lamp (red)
REG. AC OUTPUT AVAILABLE lamp (red)
OFF OFF OFF ON OFF ON ON (3)
OFF ON ON ON OFF (2)
INV. AC OUTPUT AVAILABLE lamp (red)
NORMAL AC INPUT AVAILABLE lamp (red)
DC BUS LOW/FAIL lamp (white)
DC INPUT LOW/FAIL lanip (white)
REG. AC OUTPUT LOW/FAIL lamp (white)
ON OFF ON ON OFF OFF OFF OFF OFF OFF (3)
OFF OFF OFF OFF ON (2)
(1)
With lnverter/Rectifier de-energized.
(2)
With AC Line Regulator de-energized, (3)
With DC Power Supply aligned to 2A Vital Instrument Bus Inverter.
Lamps will be in the opposite state when DC Power Supply is removed from the inverter.
Salem 2 Page 43 of 51 Rev. 11
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20040407 DEVICE INV. AC OUTPUT LOW/FAIL lamp (white)
NORMAL AC IN, LOW/FAIL lamp (white)
INVERTER TRIPPED lamp (white)
STATIC SWITCH TRANSFERRED lamp (white)
COMMON LOW AIR FLOW lamp (white)
LOW AIR FLOW lamp (white)
S 2.OP-SO, lIS-OOll(Q)
STATUS ON STATUS ON STATUS ON NORMAL ALT AC DC SOURCE SOURCE (1)
SOURCE OFF OFF OFF OFF OFF ON OFF OFF OFF OFF OFF OFF (2)
(2)
(2)
(3)
(3)
(3)
ATTACHMENT 1 (Page 2 of 3) 2A VITAL INSTRUMENT BUS UPS STATUS SHEET DEVICE 2AVII2A5 (CB101)
STATUS ON STATUS ON STATUS ON NORMAL ALT AC DC SOURCE SOURCE (1)
SOURCE ON OFF ON CHARGE ON toggle switch CHARGED lamp (red)
LOW AIR FLOW lamp (white)
BLOWN FlOl lamp (clear lens)
BLOWN F102 lamp (clear Iens)
BLOWN F103 lamp (clear lens)
(1)
With InverterlRectifier de-energized.
(2)
If lamp is ON, then ensure at least one fan in each energized panel is operating and associated filter is clean.
ON OFF ON ON OFF ON (3)
(3)
(3)
OFF OFF OFF OFF OFF OFF OFF OFF OFF (3)
If lamp is ON, then ensure at least one fan in the panel is operating and associated filter is clean.
Salem 2 Page 44 of 51 Rev. 11
A L L ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20040407 S2. OP-SO. 1 E-OOll(Q)
ATTACHMENT 1 (Page 3 of 3) 2A VITAL INSTRUMENT BUS UPS STATUS SHEET 3.0 2A VITAL INSTRUMENT BUS REGULATOR & STATIC SWS PANEL:
STATUS ON STATUS ON STATUS DEVICE NORMAL ALT AC ON DC SOURCE SOURCE (1)
SOURCE
~
~
2AVII2A3 (CB301)
ON ON OFF (3) 2AVII2A4 (CB 1 5 1)
ON I
ON ON MAN,BYPASS switch I NORMAL I BYPTO ALT 1 pREI; I(
BYPTO SOURCE SELECTOR switch I OUTPUT I OUTPUT 1 OUTPUT 11 AC OUTPUT voItmeter I
115-130 1
115-130 VAC VAC
~-
~- -
~
AC OUTPUT frequency meter I 59.5-60.5 I 59.5-60.5 I 59.5-60.5 I/
Hz HZ HZ STATIC SWITCH ON ALTERNATE lamD (white)
RETURN MODE toggle switch AUTO AUTO I AUTO (1)
With InvertedRectifier de-energized.
(2)
(3)
If lamp is ON, then ensure at least one fan in the panel is operating and associated filter is clean, With AC Line Regulator de-energized.