ML041730182
| ML041730182 | |
| 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: ML041730182 (123) | |
Text
/Giventhe 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- Control Rod 2D5 indicates 200 steps withdrawn.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.
- 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?
1 Axial power distribution limits may be exceeded, but adequate SHUTDOWN MARGIN will be maintained as lona as the deviation remains less than 18 stem.
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 limits. Distractor D has one correct part and one incorrect part because peaking factors may be exceeded ISalem Tech Specs I I .. I I
Indication Systems, including:
a) The Limiting Condition(s) for Operation b) The Bases for the LCO(s) d) The LCO Action Statements(s) 1 Tuesday, May 18,2004 10:22:10 AM ] [ Page I of89 J
Given the following conditions for Unit 2:
- Reactor power - 14%.
- PZR Pressure control in AUTO.
j- 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.
Following isolation of steam flow, and with NO further operator action, key parameters are:
- PZR pressure lowered to I825 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 runnina. I Main Feedwater is supplying feedwater through - all- _BF40's,
_ _ __ __ - pumps Aux Feedwater
____ - are stopped.
1 Main Feedwater is sutmlvins feedwater throuah all BF40's. Aux Feedwater w m m are runnina. 1
%because there is NOT an auto start signal present for the AFP's, since level swelled with the rising
/steamflow 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 BF4O's. and no feed is beina sumlied to SG's.
(ReactorTrip Response, Basis Document I I i
CN&FDWEOO9 State the setpoints for the followina automatic actuations associated with the Condensate and Feedwater Svstem Condenser Hotwell Makeup and Rejection Condensate Pump Low Flow Recirculation I
I [Facility Exam Bank 1
~
IQUf3StiOn Source: ~ I[QuktionModification Method: [Significantly Modified II 1 Tuesday, May 18, 2004 10:22:11 AM 1 I Page2of89 I
Changed from "why" is FW isolated to a "is?" & isolated type question. - -
[ Tuesday, May 18,2004 10:22:11AM 1 I ~aae3of89 I
Given the following conditions:
- Salem Unit 2 is operating at 90% power.
- 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 rise?" - __ __ - . . . - -
,PRTpressure is allowed to rise to I O psig.
Pressurizer Spray is removed from service.
611112004
=\!Knowledge of the interrelations between Pressurizer Vapor Space Accident and the following:
enough to raise PRT pressure. C. is incorrect, Pressurizer Spray has no effect on PORV leakrate or on PRT Dressure. I
/Loss of Coolant Accident I I I
LOCAOI E008 Determine the indications that are monitored to ensure proper system/component operation for each step in 2-EOP-LOCA-1 1 PZRPRTEOOB including:
a) The Control Room location of Pressurizer and Pressurizer Relief Tank control bezels and indications b) The function of each Pressurizer and Pressurizer Relief Tank Control Room control and indication c) The effect each Pressurizer and Pressurizer Relief Tank control has upon Pressurizer and Pressurizer Relief Tank components and operation d) The plant conditions or permissives required for Pressurizer and Pressurizer Relief Tank Control Room controls to perform their intended function I Tuesday, May 18,2004 10:22:11AM ] 1 Page4of89 ]
The following plant conditions exist:
- - A small break LOCA has occurred.
!- The condenser is NOT available to receive steam.
'- All S/Gs have been determined to be intact.
- PZR level indicates zero and RVLIS indicates a bubble in the reactor vessel.
- Reactor vessel level is decreasing slowly.
- RCS pressure is greater than all S/G pressures.
- ECCS is running.
- RCPs are secured.
- - Natural Circulation cooling has stopped due to steam void in the S/G U-tubes.
Which of the following describes the primary __ - paths -___ of removing- _ _ core heat?
/Boilingremoving almost all heat from core, condensation of steam in U-tubes, SI flow and
'break flow removing heat from primaw.
Natural convection cooling removing all heat from core, break flow removing all heat from primary, S/Gs do NOT contribute due to steam void in U-tubes.
Boiling removing heat from core, break flow and SI removing heat from core, S/Gs do NOT contribute due to steam void in U-tubes.
'Radiative cooling removing almost all heat from-core, SI flow condensation of steam in U-tubes, and break flow removing heat from primary.
EK2. /,Knowledgeof the interrelations between Small -. Break LOCA and
- __ following:
. the -- - ..
EK2.03, /S/Gs
)
/3. 0 1
a is correct because at this point core is boiling; most heat removed by heat ofvaporization,heat is removed by SI injection; break flow and SlGs (reflux). Distractors b, c, and d are incorrect because b - S/Gs do contribute, natural circulation is not Drimaw heat removal. c - S/Gs do contribute.
,d - radiative cooling not primary means of heat ;emoval. I LOSS of Coolant Accident I A. Decay heai removal by natural circulation with pressurizer pressure controlling B. Decay heat removal by natural circulation with reactor vessel pressure controlling C Decay heat removal during transition from natural circulation to core boiling mode D. Decay heat removal by core boiling E. Decay heat removal during transition from core boiling to natural circulation mode F. Decay heat removal after re-establishing natural circulation with reactor vessel pressure controlling I I J I
1 I I Tuesday, May 18,2004 10:22:11 AM 1 1 Page5of89 1
I Tuesday, May 18,2004 10:22 11 AM- -- 1 I Page6of89 I 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.
28 minutes.
33 minutes.
EK2. 1 Knowledge of the interrelations between Large Break LOCA and the following: i m m Question 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=
1100gpm; 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' I equating to 370,000 gallons, and 15.2' level of 150,000, you need to pump in 220,000 gallons. That's J3.09 minutes. Distracter D is the time it would take to pump in the entire RWST volume. Distracter b 4s the time if CS DumD flow is not included.
[Loss of Coolant Accident
- - __ 1 ITank CaDacitv Data I I LOCAOIEOOI Given a plot of RCS pressure and other necessary plant parameterskonditionsvs. time dunng a LOCA, perform the following in accordance with the handout A Describe the response of the ECCS and subsystems B Identify important points on the cuye RWST Tank curve, page 28 of S2.0P-TM ZZ-0002 .
1 Tuesday, May 18.2004 10:22:11 AM i 1 Page7of89 1
'Salem Unit 2 has experienced a rupture of a RCS cold leg which has resulted in containment pressure peaking at 18 psig.
With all systems actuating as expected, which of the following choices identifies the containment isolations which have occurred, and the reason why they have occurred?
ALL feedwater to containment to preclude an excessive RCS cooldown event.
Phase B to isolate potential injection paths to containment; Containment Ventilation to ensure mon-essential containment ventilation penetrations are isolated.
I Main Steamline to minimize potential primary-to-secondary leakage; Feedwater to prevent uncontrolled filling of any SG.
Phase A to ensure non-essential containment penetrations are isolated; Phase B to isolate EK3. //Knowledgeof the reasons for the following responses as they apply to Large Break LOCA:
!EK3.06 IActuation of Phase A and B during LOCA initiation Distractor a Feedwater Isolation only isolates Main Feedwater, it does not isolate ALL feedwater. AFW ,
- isstill available for injection to SG's. Distractor b is incorrect because Phase B isolates leakage paths, !
/not iniection paths. Distractor c is incorrect because Main Steamline Isolation is desiqned to minimize (Reactor Trip or Safety Injection I 1 minimize the potential for a release
- - __ J I
1 1I I 8
. .. . .. .. .. ..-~_...
I- Tuesday, May 18,2004 10:22:12 AM __ 1 I Paae8of89 I
IWhich of the following choices describes conditions for potential damage to an operating Reactor
/Coolant Pump if NO operator action _ _ is taken? - - - - - - __ I 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.
Total Thermal Barrier CC return flow = 155 gpm and steady, any RCP seal injection flow = 14 gpm and lowering. - .
RCP seal leak-off flow = I.3 gpm and steady, any RCP motor winding temp 260 deg f and
[steady. - - - _ ........ - _ -
Malfunctions:
L . . - __
seal injection is high, but lowering, and will not affect RCP performance. D is incorrect because leak-off is within normal range, - as is RCP motor windinq temD..
L REACTOR COOLANT PUMP ABNORMALITY
~.. .
J I __ .........
I
, , .. ,-1 , I
'I . ., ,
........ 2.
ABRCPI EO01 a) Basic RCP Construction b) Seal Injection and Seal Water Configuration c) RCP CW Configuration I RCPUMPE008 Identify and describe the Control Room controls, indications, and alarms associated with the Reactor Coolant Pump, including:
a) The Control Room location of Reactor Coolant Pump control bezels and indications (N/A NEO) b) The function of each Reactor Coolant Pump Control Room control and indication (N/A NEO) c) The effect each Reactor Coolant Pump control has upon Reactor Coolant Pump components and operation (N/A NEO) d) The plant conditions or permissives required for Reactor Coolant Pump Control Room controls to perform their intended function j .... .-- ............. .. __ --.
Tuesday, May 18, 2004 10:22:12 I Paaegof89 1
Given the following conditions:
- Salem Unit 2 is operating at 100% power, steady state, with all controls in AUTOMATIC.
- The PZR level channel selected for ALARM fails LOW.
8
- Which of the following choices identifies how charging flow and PZR level will be affected?
Charging flow will , and PZR level will . I 1
rise, lower.
lower, lower.
El,lower, rise.
22K103 ,
8 AKI. IIKnoiedge of the operational implications of the following concepts as they apply to Loss of Reactor 1
8 level will rise since there is no letdown, and charging flow has a minimum flow stop of -46 gpm.
ICharging, Letdown, and Seal Injection system, including:
a) The Control Room location of Pressurizer Pressure and Level Control system bezels and indications b) The function of each Pressurizer Pressure and Level Control system Control Room control and indication c) The effect each control has upon Pressurizer Pressure and Level Control components and operation d) The plant conditions or permissives required for Pressurizer Pressure and Level Control system Control Room controls to perform their intended function I Tuesday, May 18, 2004 10:22:12 AM I I Paae 10of89 I
Given the following conditions:
,- Unit 2 is operating at 88% power.
j- The crew is attempting to isolate a Component Cooling Water leak in accordance with S2.0P-AB.CC-0001, COMPONENT COOLING ABNORMALITY.
- 22 Charging pump is operating.
- 21 and 23 Component Cooling pumps are operating.
- 22 Component Cooling pump and heat exchanger are isolated.
- Component Cooling Surge tank level indication, LI-628A, is 37% and lowering.
- Component Cooling Surge Tank Makeup is isolated.
- OHA Alarm C-2 CNTMT SUMP PMP START has actuated.
Which of the following identifies the location of the leak?
'21 Component Coolinq header.
22 Component Coolinn header.
Non-safeguards header.
LEither22 Component Cooling header OR Non-safeguards __ - -.
.. . -- -. header.
w . 0 1 [Location of a leak in the CCWS the leak is on either of the CC headers and non-safeguards header, then checks if any RCP OHAs 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..
/Component Cooling Abnormality J I - _ _____ - 1 b) Describe the plant response to actions taken in the abnormal procedure.
4 Tuesday, May 18,2004 10:22:12AM I Paae 11 of89 I
- Given the following conditions
- Salem Unit 2 is in Mode 6, with core offload in progress.
- Refueling cavity level is 24' over the reactor vessel 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 ABNORMALITYI which of the following actions iwill be Derformed FIRST?
lmmediatelv suspend core alterations.
'Close the Spent Fuel Pool Gate Valve.__ _ _
Initiate S2.OP-AB.RHR-0001 LOSS OF RHR. I Send an operator to adjust Service - Water
- flow controllers-for
- - ---____ - -in service
- - __ - HXs.
CCW ___ - - 1 2.1 1 Conduct Of Operations !
2.1.2 IKnowledge of operator responsibilitiesduring all modes of plant operation. JpTi3.oil4.01 action in AB.CC to close Gate Valve. 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.
I -
ABCCOlE005 __.
G 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.
ABRHRIE005 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.
Tuesday, May 18, 2004 10:22.12 AM 1 I Paae12of89 I
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.
I- All attempts to close 2PS3 have failed.
4 iIAW S2.OP-AB.PZR-0001, PRESSURIZER PRESSURE MALFUNCTION, which of the following choices identifies the actions required to be taken?
,EnergizeALL PZR heaters, commence a rapid power reduction in anticipation of tripping the
- reactor and stomina RCP's.
,Commence a power reduction, and raise charging flow to compress-the pressurizer bubble.
Trip the Reactor, confirm the trip, stop 23 RCP. If pressure reduction continues, initiate SI and GO TO EOP-TRIP-1.
Trip the reactor, confirm the trip, stop 21 and 23 RCP's. If pressure reduction continues, stop anv other RCP.
Emergency and Abnormal Plant Evolutions
-IROGroup] 1 - -
-11 1000027G431 !
027 . - Pressurizer Pressure Control Malfunction -_ - _ _ 11 2.4 1 Emergency Procedures I Plan
-2.4.31 /Knowledgeof annunciators alarms and indications, and use of the response instructions.
path. Stop 23 and 21 RCP's, then another, to reduce or stop rate of pressure reduction.
]Pressurizer Pressure Malfunction 1
I.. . ............ .......... .. . . . . .- .-.
I 1
[ Tuesday, May 18,2004__ -10:22 1 Page13of89 I
iGiven the following two sets of conditions:
'- Salem Unit Iis 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 2185 psig before the PORV was isolated.
- Pressurizer level was quickly recovered to 22% and pressurizer heaters remained in AUTO.
Which malfunction will take a longer time for pressure recovery from 2185 psig and why?
'The PORV because when the PORV opens the steam space needs to be reheated to raise Dressure.
ThePORV because the pressure reduction will be faster for the failed open 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.
AK3. I Knowledge of the reasons for the following responses as they apply to Pressurizer Pressure Control 1 Malfunction: 1 AK3.04 [Why, if PZR level is lost and then restored, that pressure recovers much more slowly 1rnrnI malfunction. Distracter B is incorrect because it's effect is inconsequential. Distracter C is incorrect because pressure reduction rate does not affect recovery time. D is the correct answer because after the 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.
/Pressurizer Pressure Malfunction I
II
.~
ABPZRI EO01 . . to S2 OP-AB PZR-OOOIKN Describe operation of the Pressurizer Pressure control svstem as amlied . I I
I
[ Tuesday, May 18,2004 __10 I Page 14of 89 1
I Tuesday, May 18, 2004 10,22:13AM 1 I Page15of89 1 Given the following conditions:
- Salem Unit 1 is in Mode 5.
- Rx power is 150 cps.
- RCS pressure is 300 psig.
8 Which of the following choices describes a condition which would cause OHA E-5, "SR DET VOLT TRBL" to alarm?
'Loss of Dower to either Source Range channel "C" or I'D" Gamma Metrics. I
'Deenergizing 1B 23OVAC bus without manually transferring 1N31, Source Range detector channel 1, to alternate power supply.
1 Removing 1A Vital Instrument Bus Inverter from service with I A VIB alternate source AC disconnected.
,Deenergizingthe IB 125VDC battery bus for maintenance.
A&. ];Ability to determine and interpret the following as they apply to Loss of Source Range Nuclear jlnstrumentation:
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 VI6 inverter will still be supplying power to 1B VIB.
ABNlSlEOOl Describe the operation of the following as applied to S2.OP-AB.NIS-O00I(Q), in accordance with this lesson plan:
a) source range instrumentation b) intermediate range instrumentation 1 Tuesday, May 18,2004 10:22:13 AM 1 I Paae 160f 89 I
Given the following conditions:
- A SGTR has occurred on 21 S/G.
,- Operators are preparing to initiate a RCS cooldown IAW step 15 of EOP-SGTR-I, STEAM
!GENERATORTUBE RUPTURE.
!- Ruptured SG pressure is 605 psig.
'- Containment pressure is 0.1 psig and steady.
Using the provided steam tables and assuming 25 degrees of instrument error, what must the final hottest CET temperature be after the cooldown to assure 20 degrees subcooling? - -
503 degrees. - _ _ .
475 degrees.
459 degrees.
441 degrees.
'Tube Leak:
Using provided steam tables 605 psig (620 psia) correlates to -490 deg. Subtract the 25 degree instrument error and the 20 degree subcooling, and the final temp required is -445 degree. The correct 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 iwouId NOT use adverse containment numbers. Distractor A is the choice for 1000 psig or greater, 8
'which is where the majority of SGTRs will fall. Distractors B & C were picked from Table D of SGTR-1
'as realistic numbers but for higher SG pressures.
Isteam Generator Tube Rupture I
- _ _ - _ _ J 1
I 17.2, 21, 25, and 27
[ Tuesday, May 18,2004
- _10_ I Paae17of89 I
Salem l Unit 1 is operating at 100% power steady state. During a board walkdown, the RO observes ithe following conditions:
- Rx power is 100.0% and stable
- 1IS/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.
Which of the following conditions is causing these indications?
!RCS leak from 11 loor, cold lea.
Main feed line break on 11 SG.
Steam line ruDture of 11 SG.
Tube rupture on I 1 SG.
I I SGTROIEOOI List 7 symptoms of a steam generator tube rupture. For each symptom:
A. Determine if it is unique to a steam generator tube rupture.
B. Determine if it can be used to positively identify a ruptured steam generator SGTROIE006 Determine the indications that are monitored to ensure proper system/component operation for each step in 2-EOP-SGTR-1 I Tuesdav. Mav 18.2004 10:22:13 AM I I Paae 18 of89 I
'Given the following plant conditions-
- 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.
I- Unit 2 ECAC fails to start.
Which of the following choices identifies the indication which would be present after 5 minutes with NO operator action? - - ____
All MSI 0 Atmospheric Relief Valves are failed open.
Letdown flow indicates __ 0 gpm. - -
21 RHR pump flow is 2,000 gpm.
21 RHR HX inlet temperature risina slowlv.
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 b) Emergency control Air Compressors c) Station Air Headers d) Control Air Headers e) Excess Flow Check Valves I
I 1 Tuesday, May 18,2004 10:22:13 AM ] I Page19of89 I
Given the following conditions:
- Salem Unit 2 is operating at 100% power when the entire 500KV electrical switchyard suddenly and unexpectedly becomes de-energized.
- The reactor trips automatically.
All EDGs start and load in Mode II.
,- SI is NOT required.
- All other expected automatic actions occur.
- The control room crew is recovering in 2-EOP-TRIP-2, REACTOR TRIP RESPONSE.
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.. . .. ___
will be operating in the NORMAL Mode. - -_
will have been MANUALLY initiated in Accident Pressurized-Mode.
'will need manual alianment due to the loss of Dower.
a __
1 .-_- .-.
Ir\nen 6/11/2004 (Emergency andAbnormai Pbnt Evolutions 1I bo#156A118 1- ' [LOSS of Off-Site Power /Record Number I 17 1
MI. Ability
- - to operate and / or monitor the following . . as they apply - to- Loss of OffISite Power MI.18 IControl room normal ventilation supply fan b is correct because the CAV system will not re-align due to loss of off-site power, all components that would cause a re-alignment are powered from vital AC or DC power supplies. Distractor a is incorrect 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-alianed. nor will it need to for a loss of off-site power. I IReactor Trip Response __ - _ _ I Control Area ventilation Operation CAVENTE008 Identify and describe the Control Room controls, indications, and alarms associatedwith the ControlArea Ventilation System, includina:
a) The Control Room location of Control Area Ventilation System control bezels and indications b) The function of each Control Area Ventilation System Control Room control and indication c) The effect each Control Area Ventilation System control has upon Control Area Ventilation System components and operation d) The plant conditions or permissives required for Control Area Ventilation System Control Room controls to perform their intended function e) The setpoints associated with the Control Area Ventilation System control room alarms Tuesday, May 18, 2004 10:22:14AM 1 1 Page20of89 I
Given the following conditions:
- Salem Unit 2 is operating at 100% power.
- PZR level channel I is selected for control 8- PZR pressure channel I is selected for control.
Which of the following choices identifies the IMMEDIATE ACTION that must be taken if a loss of the 2A Vital Instrument Bus were to occur?
Place the Charging System Master Flow Controller in MANUAL.
Place the PZR Master Pressure Controller __ in MANUAL. ____ __
,Place the Main Steam Dumps in MS-PRESSURE CONTROL-MANUAL.
57G123 18 Unwarrented Rod Motion, directs as an IMMEDIATE ACTION to place rod control in mANUAL. B,C,
!AND D are incorrect because they are not immediate actions.
bnwarrented Rod Motion I (Lossof 2A VIB I
I 1- 0003(Q), S1IS2.0P-AB.115-0004(q)and the bases for the actions.
115VACE004 I Describe the function and operating characteristics for the following 115 Vac Electrical Systems components:
a) A, 8, C, and D Vital Bus b) Essential Control Power System c) Essential Lighting Power System d) Unit 2 RMS Power System e) SPDS Power System I Tuesday, May 18, 2004 10:22:14AM 1 I Paae21 of89 1
Given the following conditions:
,- An accidental gaseous radwaste release has occurred in the 24 Gas Decay Tank room.
'- Radiation protection has determined radiation dose level in the room is 500 mr/hr.
A worker assigned to enter the room has a current yearly TEDE dose of 2.5 REM.
Assuming all required approvals have been obtained, what is the MAXIMUM total time the worker can be in the GDT room without exceedina their 10CFR20 TEDE dose limit?
30 minutes. 1 One hour.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 30 minutes.
5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
- d [Memory
- AK1.02 Biological effects on humans of the various types of radiation, exposure levels that are acceptable 12.5 for personnel in a nuclear reactor power plant; the units used for radiation intensity measurements and for radiation exposure levels I Radiation Protection Program Code of Federal Regulations I
RADCONE002 List the following external radiation exposure limits, in accordance with NC.NA-AP.ZZ-O024(Q), 10CFR20, and Reg. Guide 8.13-A. 10CFR20 dose limits for external, internal, and total whole body, skin, extremities, and eyes, as well as extension limits and requirements B. Administrative dose control levels for Category 1 and 2 Workers, as well as extension limits and requirements C. Reg. Guide 8.13 limits and administrative dose control levels for Declared Pregnant Women D. 10CFR2O and Administrative limits for members of the general public and minors E. Category 1 Radiation Worker F. Category 2 Radiation Worker RMSOOOEOOI State the purpose of the Radiation Monitoring System 1 Tuesdav. Mav 18.2004 10:22:14 AM I I Paae22of89 I
Given the following conditions:
- Salem Unit 2 is operating at 100% power steady state.
i- Service water (SW) pressure on both 21 and 22 headers drops from 105 psig to 95 psig.
i- 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.
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 21SW21 OR 22SW21, DIESEL CLG SW INLET VALVES, to isolate the leak, and lock out 2C EDG.
Shut 21SW37, C DIESEL CLG SW INLET VALVE, if leak is not isolated shut 22SW37, C DIESEL CLG SW INLET VALVES, declare 2C - EDG__ INOPERABLE.
Declare 2C EDG INOPERABLE,shut 21SW21 OR 22SW21, DIESEL CLG SW INLET VALVES, to isolate the leak. __ .-
I I
Lock out 2C EDG, isolate the leakby shutting 21SW37 AND 22SW37, C DIESEL CLG SW
- INLET VALVES.
L i I i L IApplication
-'I 1
20 AA1. ] Ability to operate and / or monitor the following - __as they apply to Loss of Nuclear Service Water: __ -
AA1.06 IControl of flow rates to components cooled bv the SWS Per step 3.1 50 of the AB,-lock out the EDG(s) that will be affected and isolate the leak. -The onlyway 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 &pplyik the leak. Isolating the
/SWZl'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.
IService Water Nuclear dwg I
a) The Control Room location of Service Water - Nuclear Header control bezels and indications b) The function of each Service Water - Nuclear Header Control Room control and indication c) The effect each Service Water - Nuclear Header control has upon Service Water - Nuclear Header components and operation d) -
The plant conditions or permissives required for Service Water Nuclear Header Control Room controls to perform their intended function e)
The setpoints associated with the Service Water Nuclear Header control room alarms I
I __
Tuesday, May 18,2004
- .10:22:14AM 1 1 Page23of89 1
,Giventhe 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 SUPP~Vfans ONLY.
- Place Control Room Ventilation in FIRE INSIDE CONTROL AREA on both Unit 2 and Unit 1,
- and ONLY secure __ ALL FHB supply - fans.
Place Control Room Ventilation in FIRE OUTSIDE CONTROL AREA on both Unit2 and Unit 1
'I. and secure ALL Unit 2 FHB SUPD~Vand exhaust fans ONLY.
- PlaceControl Room Ventilation in FIRE INSIDE CONTROL AREA on Unit 1 ONLY, and secure ALL FHB supply and exhaust fans.
d I 1 21 IAAl. ] Ability to- operate and I or monitor the following as they apply to Plant -_ -
Fire on
- Site:
For a fire in an area NOT served by the control room BOTH units CAV placed in fire outside control area.
operator
__ is directed to secure -ALL -
FHB supply
- - and exhaust fans.
/Control Room Fire - __
-- Response
- - I I
Obfe
^A_
Identify and describe the Control Room controls, indications, and alarms associated 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 (N/A NEO)
C. The effect each Fire Protection System control has upon Fire Protection System components and operation (N/A NEO)
D. The plant conditions or permissives required for Fire Protection System Control Room controls to perform their intended function FIRPROE012 1 Tuesday, May 18,2004~10:22:14AM- I I Page24of89 1
Which of the following choices contains ONLY instrumentation and controls that are available for use at the Remote Shutdown Panel 213?
I. CAV Supply Fans control switches.
II. RHR HX CCW outlet flow indications.
Ill. Letdown Orifices control switches.
IV. Component Cooling Surge Tank level.
V. Boric Acid Storage Tank levels.
VI. AFW Dum13 control switches.
I. II. 111. 1 Ill, IV, v. - _.
El I, v, VI.
M II. IV. VI.
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, FE601B is CC flow out of 22 RHR HX, and feeds F1601D which id on HSD Panel.
Hot-Shutdown Station Panel 213-1 drawin I Tuesdav. Mav 18. 2004 10:22:15 A M I I Paae25of89 I
Given the following conditions:
- Salem Unit 1 is operating at 100% power.
- All systems are aligned normally for 100% power.
i Which of the following choices, when taken by itself, identifies a condition which will NOT automatically close the ICV3, Letdown 45 GPM orifice Isolation Valve?
- Loss of 28VDC . - - _ _ control voltage. __ - - -
i -
~ :.~ ......... ............... .................... ... .......................... ..-.-
6-2 Hot Shutdown Panel 213 switch for 1CV3 momentarily taken to CLOSE.
l!k 'The-only _ operating__charging pump trips. - - -
=/,Knowledge of the interrelations between Control Room Evacuation and the following:
Logic drawing 224429 shows the either alarm or control channel will close CV3, loss of 28VDC will close lCV3, and no charging pp operating will close 1CV3. Taken by itself, trying to close lCV3 from ,
the HSD panel will not close 1CV3. The 1CV3 must be taken to local before the open/close switch will work. 1
/Logic Drawing i
r J
~.~
.............. i
[ Tuesday, May 18, 2004 10.22:15AM ] I Paae26of89 I
Given the following conditions:
- Salem Unit 2 has experienced a SBLOCA.
- Operators have transitioned from 2-EOP-TRIP-1 REACTOR TRIP RESPONSE, to 2-EOP-LOCA-6 LOCA OUTSIDE CONTAINMENT.
- The 21SJ49 COLD LEG ISOLATION VALVE has just been closed in an attempt to isolate the leak.
Which of the following choices describes the response which would indicate successful leak isolation, and the next__ action to be performed IAW- - -
2-EOP-LOCA-6?
RCS pressure__stable, open 21SJ49. _ _ - -
RCS pressure rising, stop 21 RHR pump.
,RCS pressure risina. close 21RHl9 RHR DISCHARGE X-CONN.
RCS pressure stable.-stor, 22 RHR pumr,.
Containment:
because the RHR system is split and there is no discharge path. Distractor c is incorrect because the 21RH1 9 would already have been closed at step 2. Distractor d is incorrect because both stable j Tuesday, May 18,2Oi4 10:22:15 AM 1 1 Page27of89 1
Given the following conditions:
- 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 second PORV will NOT open.
Which one of the following describes - _ _ _ _ the
. - consequences
- - __ __ - . . -of- the__ PORV- failure?
/Bleed and Feed cooling of the RCS must be terminated and secondary depressurization to Iiniect condensate pump flow must be immediatelv initiated.
'ALL SGs will reauire depressurization to iniect the alternate source of feedwater.
The RCS will rapidly re-pressurize when the SGS empty, resulting in a violation of the RCS Safety Limit.
The RCS may not depressurize quickly-enoughto ensure sufficientsl flow to provide RCS heat removal, and other RCS openings may have to be established.
EKI. ]:Knowledgeof the operational implications of the following concepts as they apply to Loss of Secondary
'Heat Sink:
,because SI flow may NOT be adequate at the PORV setpoint. Distractor a is incorrect because action j
!to align condensate pumps is already taken, and not as a contingency to Bleed and Feed. D is correct
- because 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 I
1- with no other operator action 1 Tuesday, May 18,2004 10:22:15 AM ] I Paae28of89 I
,Given the following conditions:
1- A LOCA has occurred during a cooldown while in MODE 3.
I- Operators have transitioned out of 2-EOP-TRIP-I.
- RCS pressure is 125 psig.
- 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 1150 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.. .
made and a RCS temperature soak for a ONE hour period will be completed.
made but NO actions are implemented 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 Shock concerns.
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.
Distractor c is incorrect because entry to FRTS will be made, THEN RCS pressure is evaluated.
Distractor a is incorrect because . . . no
. . soak
. . . is . . . . . prior to returning
. .required t procedure in effect.
I IAnateriai]-R q
e u
i -Figure ~
-Limit . -_ Curve -
1 Tuesday, May 18,2004 10:22:15 AM 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 dearees is Derformed?
- IO0 degrees / hr.
bnormal Plant Evolutions E l0 ,,NaturalCirculation with Steam Void in Vessel withlwithout RVLIS 2.2 I _Equipment
_ _ Control
,2.2.22 , /Knowledgeof limiting conditions for operations and safety limits.
Ijmj The Tech Spec LCO for RCS cooldown is 100 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
/d to 500 degrees. d is applicable in non-E tion step relative to a Natural Circulation Cooldown.
I Tuesday, May 18,2004 10:22:16 AM I I Page30of89 I
Given the following conditions:
- 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.
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. Thecrew will stabilize power c 1 49% and determine why the main turbine- __failed ____ - to trip.
An OT delta T turbine runback will occur, and will continue-untilthe Main Turbine is off line. 1
,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
- reactor.
Control rods must be_ _inserted MANUALLY, _ _ and
- - the Main - __ Turbine
__ must
- be
- manually
__ tripped.
' T I Ability to (a) predict the impacts of the following on the Control Rod Drive System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: - .
m __ - - _ _ - mj4.414.61 AMain 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 A M 2s 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 the rods will insert at max (72 spm) .~ in AUTO, and the crew will attempt to manually trip the Rx in iresponse to the ATWT.
IReactor Trip or Safety Injection A. RESPONSE TO NUCLEAR POWER GENERATION.
B. RESPONSE TO A LOSS OF CORE SHUTDOWN.
C. Shutdown Margin Status Tree
' FRSMOOTOOI Given a set of plant conditions, perform actions for a Response To Nuclear Power Generation in accordance with 2-EOP-FRSM-1 TRP001E016 State the Immediate Actions of EOP-TRIP-1 1 Tuesday, May 18,2Oi4 10:22:16 AM ] I Page31 of89
Given the following conditions:
- Salem Unit 2 is in MODE 6.
- - The Reactor vessel head is removed and on its storage stand.
1- The Refueling Cavity is being filled from the RWST IAW S2.OP-S0.SF-0003, FILLING THE a REFUELING CAVlTY.
- Refueling cavity water level is 1I 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'. I iPZR Cold Calibrated level is off-scale high RWST level is IO'.
,AI.I 1 Relative level indicationsin the RWST, the refueling cavity, the PZR and the reactor vessel during -1 2.7 [ 3.21 preparation for refueling I
I IDrainina the Reactor Coolant Svstem RCSOOOE008 Identify and describe the Control Room controls, indications, and alarms associated with the Reactor Coolant System, including:
a) The Control Room location of Reactor Coolant System control bezels and indications (N/A NEO) b) The function of each Reactor Coolant System Control Room control and indication (N/A NEO) c) The effect each Reactor Coolant System control has upon Reactor Coolant System components and operation (N/A NEO) d) The plant conditions or permissives required for Reactor Coolant System Control Room controls to perform their intended function (N/A NEO)
I . e) The setpoints associated with the Reactor Coolant System Control Room alarms I PZRP&LE008 .I system, including:
a) The Control Room location of Pressurizer Pressure and Level Control system bezels and indications b) The function of each Pressurizer Pressure and Level Control system Control Room control and indication c) The effect each control has upon Pressurizer Pressure and Level Control components and operation d) The plant conditions or permissives required for Pressurizer Pressure and Level Control system Control Room controls to perform their intended function e) The setpoints associated with the Pressurizer Pressure and Level Control system control room alarms 1 Tuesday, May 18,2004 10:22:16 AM 1 I Paae32of89 I
1 Tuesday, May 18,2004 10.22 16 AM I I Page33of89 I Given 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 P-8, and 24 SG pressure will be lower than the remaining SG's.
The reactor did NOT automatically trip because power is P-8, and 24 SG pressure will be higher than the remaining SG's.
The reactor tripped automatically on 214 Low RCS Flow on 1/4 RCS loops > P-10, and 24
'SG pressure will be higher than the remaininq SG's.
$Thereactor tripped automatically on 2/3 Low RCS Flow on 1/4 RCS loops > -P-lO, and 24 SG pressure will be lower than the remaining SG's.
[Plant Systems [002000A303 io02 i Reactor Coolant System 30 I
,A3. 1 Ability to monitor automatic operations of the Reactor Coolant
- . --- - - - - System including:
A3.03 IPressure, temperatures, and flows li4.4]
The logic for Rx trip shows neither 1/4 loops low flowbr 1/4 RCP bkrs open will causea Rx trip>P-lO but <P-8. As the flow lowers in 24 loop, its steam pressure will lower, as its DTT lowers. 24 loop steam flow will lower, and the other 3 loops steam flow will rise. Tc's n the 3 remaining loops will lower,
'and steam header pressure will lower.
I - - I Steam Dump System Operation I
I I
I I Tuesdav. Mav 18.2004 10:22:16 AM I I Page34of89 1
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 I
NO operator action, which of the following choices describes the RCP Seal Injection flow one
'minute after the turbine runback is completed?
kd Seal injection i o w will rise due to 1CV55, CENT-CHG PMP FLOW CONT VALVE, modulating in the oDen direction.
[Seal injection flow will lower due to 1CV7lj CHG HDR PCV, modulating in the open direction.
Seal injection flow will rise due to 1CV71, CHG
_ -HDR PCV, modulating in -the
__ closed
- - .- direction. I Seal injection flow will lower due to 1CV55, CENT CHG PMP FLOW CONT-VALVE, modulatina in the closed direction. I A3. 1 Ability to monitor automatic operations of the Reactor Coolant
- - _ - _Pump
_ System including: ..
mw32 .1 d is correct because after the runback is complete, programmed PZR level will be less than actual. This ill cause the CV55 to modulate 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.
Main Feedwater / Condensate Abnormality Charaina. Letdown. and Seal lniection a) The Control Room location of Reactor Coolant Pump control bezels and indications (N/A NEO)
Identify and describe the Control Room controls, indications. and alarms associated with the Reactor Coolant Puma includina b) The function of each Reactor Coolant Pump Control Room control and indication (N/A NEO) c) The effect each Reactor Coolant Pump control has upon Reactor Coolant Pump components and operation (N/A NEO) d) The plant conditions or permissives required for Reactor Coolant Pump Control Room controls to perform their intended function e) The setpoints associated with the Reactor Coolant Pump control room alarms 1
I Tuesday, May 18, 2004 10:22:16AM 1 I Page35of89 1
Given the following conditions:
- Salem Unit 2 is in MODE 5.
- 500 KV switchyard is aligned for normal operation with Unit 2 Main Generator Drops removed.
- 2B and 2C 4KV Vital busses are powered from 23 SPT.
- 2A 4KV Vital bus is powered from 24 SPT.
- RHR HX inlet temperature is 150 deg F.
All RCS Tc are 130 deg F.
All SGs are filed to 80% WR level.
- RCS pressure is 290 psig.
- PZR level is 20%.
- PZR temperature is 200 degrees.
- ALL S/G secondary side temperatures are 80 deg.
- POPS is in service.
/Giventhe above conditions, is it permissible to start 23 RCP IAW S2.0P-SO.RCP-001 REACTOR
!COOLANT PUMP OPERATION, and why?
Yes, because the H 4KV group bus and the 2A 4KV are powered from different SPT's which will prevent 2A vital bus under voltage protection - - actuation.
No, because a-PZR bubble is required to be drawn prior to starting any RCP.
Yes, because S/Gsecondary side temperatures at 80 deg F with RCS pressure of 290 psig will prevent exceeding fracture toughness stress limits in the SGs.
No, because PZR level is too low to prevent a RCS pressure transient and PZR heater isolation. --. . - . _.__ . ... . ..
a katem L- .
I &2 Knowledge of Reactor Coolant Pump System design feature(s) and or interlock(s) which provide for the
[following:
K4,02 IPrevention of cold water accidents or transients IEIm 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 23 RCP is powered from F 4kv ,
bus. Distractor b is incorrect because a PZR bubble is not required to start RCP. . Distractor c is L-I precautions associated with each operating procedure which are required to be considered by either Licensed or Non-Licensed Operators I
I I--
Tuesday, May 18, 2004 10 22.17 AM ] I Page36of89 I
1 Tuesday, May 18,2004 1022 I Page37of89 1 An electrical fault causes Auxiliary Building MCC I C West Valves and Misc. 230V Vital Control Center Motor Control Center (MCC-1CY2AX) to become deenergized.
With NO operator action, which of the following choices identifies a concern with this MCC de-energized while at 100% power?
Rundown of __the - 1A 28VDC
. __ battery due ___ to loss of -l A
. _ _2
_ 28V- battery charger.
,Havingto-declare 12 Boric Acid Storage Tank INOPERABLE at 63 deg. F due to the loss of
'tank heaters. . - - ___--
Lowering air pressure in the I C EDG Air Storage Reservoirs due to the loss of both I C EDG air system air compressors.
,M-e-m.o-~ ___
__ Salem IS 2 -1 .. .
611 112004 OOK204 33,
/Loss of 1C 460/230Vital Bus I
-~
CVCSOOE004 a) LetdownlCharging i) Letdown lsolaiton Valves, CV2, CV277 ii) Regenerative Heat Exchanger iii) Letdown Orifices iv) Letdown Orifice IsolationValves, 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 xvii) VCT IsolationValves, 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) Charging to Loop 3 Valve, CV77, Loop 4 Valve, CV79 xxiv) PZR Auxiliaty Spray Valve, CV75 xxv) CCP Flow Control Valve, CV55 b) RCP Seal Water i) Seal Water Injection Filters ii) Seal Bypass Flow Valve, CV114 iii) Seal Water Return Isolation Valve, CV104 I Tuesday, I y 18, 2004 10.22:17 AM ] I Page38of89 I
iv) Seal Water Return Relief Valve, CV115 v) Seal Return Cont. Is01 Valves, CV116, CV284 vi) Seal Return Filter vu) Seal Water Heat Exchanger Excess letdown I) Excess Letdown IsolationValves, CV278, CV131 ii) Excess Letdown Heat Exchanger 111) 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, CVI 85 xi) VCT Makeup IsolationValve, CV181 xii) Rapid Borate Stop Valve, CV175 i .. . .. ... .. . . .. . ... ... . . .... - - . .. .......... .... . .......... ..... .. ..
. . .. . .. . . ........ .. ~~ . . .~
1 I
I Tuesday, May 18,200410 I Page39of89 I
Given the following conditions:
- Unit 2 has experienced a Large Break LOCA.
EOP-LOCA-3, "TRANSFER TO COLD LEG RECIRCULATION" is complete with NO abnormalities encountered.
- Operators are currently at step 26, "Preparation for Hot Leg Recirc", of 2-EOP-LOCA-I , "LOSS OF REACTOR COOLANT".
- Off-site power is supplying all 4KV Vital busses.
If BOTH _ _ RHR
- pumps are operating, what __ would be the effect
-.._________._. - if
__ 22 RHR Pp were to trip? ---
21 RHR pump flow would increase to runout conditions. ._ __
21 and 22 SI pumps would begin to cavitate.
22 Containment Spray Pump would lose NPSH.
Containment Sprav flow would drop to zero. I
' K I ' K n o w l e d g eof the effect that a loss or malfunction of the Residual Heat Removal System will have on the following: - . .. .. .
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 I
other RHR pp were to trip will in cold leg recirc. Distracter b is incorrect because NPSH would still be sumlied bv 21 RHR OD 1 I2-EOP-LOCA-1 I I
I I
LOCAOIE008 Determine the indications that are monitored to ensure proper systemlcomponent operation for each step in 2-EOP-LOCA-I I Tuesdav. Mav 18.2004 10:22:17 AM I I Page40of89 I
Given the following conditions:
,- 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'.
With NO operator action, which of the following- choices contains ONLY automatic actions that will occur with the current RWST level?
a !21SJ44 opens, 21RH4 RHR PUMP SUCT VALVE shuts.
22CC16 opens, 22SJ113 SI CHG PUMP X-OVER VALVE opens.
m21SJ113 SI CHG PUMPX-OVER VALVE opens when 21RH4 is fully shut.
OOA308 35
,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 thev will start to close. The onlv answer which contains both a CC16 and a SJI 13 is c, all
!of the distracterscontain at least one SJ44, or RH4, ITransfer to Cold Leg Recirculation a) VCT Outlet Valves - CV40/41 b) -
SI Pump Mini-Flow Isolation Valves SJ67/68 c) -
RHR Hot Leg Suction Valves RH112 d) -
RHR Pump Suction Isolation Valves RH4s e) Containment Sump Suction Valves - SJ44s f) RHR Discharge to CharginglSl Pumps - SJ45s g) RHR Discharge to Containment Spray - CS36s 1
I Tuesday, May 18,2004 10:22:17 AM I I Paae41 of89 I
During a normal Unit Iplant 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 11 RHR pump running.
- 11 charging pump is in service 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.
If the SI Durnr, were to start. it miaht overr,ressurize the RCS.
/Cooling System: . . . . . . . . . .....
I because we are allowed to operate in this lineup IAW S I .OP-SO.RC-001 REACTOR COOLANT PUMP OPERATION. Distractor c is incorrect because the HU rate is less than I00 deg / hour allowed in TSAS 3.4.10.1.a. D is correct because only 1 charging/SI pp is allowed to be OPERABLE with loop Tc 1
..312 to prevent
. - exceeding relief
-_ - - - -capacity
__ .- - of POPS. . . . - . i I
ECCSOOEOI0 State the Technical Specifications associated with the components, parameters, and operation of the Emergency Core Cooling System, including:
a) The Limiting Condition@)for Operation (NIA NEO) b) The applicability of the LCO(s) c) The LCO Action Statement@)(N/A NEO) d) The Bases for the LCO(s)
' I 2/09/2002 DC COOK NRC Exam I Tuesday, May 18,2004 10.22:18AM I I Page42of89 I
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
/minuteafter the loss of off-site power? __
,21 t - and 22 CCW pumps are running , 2-3-CCW
- pump is stopped.
- -_- I No CCW pumps are running, all CCW pumps swap to MANUAL. I 21 and 22 CCW pps swap to AUTO, all CCW pumps--arerunning.
611 112004 A4. I Ability to manually operate and/or monitor in the __
control room:
I A4.01 ICCW indications and controls I[ 3.3![ 3.11 pumps will have their control circuit swap to manual, and manual operation is locked out. If a pump is selected to local manual at the HSD panel, it will still receive a start siqnal and manual control lockout.
ILogic Drawings I
CCWOOOE00616escribe the interlocks associated with the followinq Component Cooling Water Svstem components:
I Tuesday, May 18,2004 10.22.18 AM I Paae43of89 I
Given the following conditions:
- Salem Unit 2 is at 25% power.
- SSPS testing and troubleshooting are in progress.
- A Phase B Containment Isolation signal is generated and all related valves closed.
- The Phase B signal can NOT be reset.
Which of the following describes the required operator actions IAW S2.0P-AB.RCP-001 REACTOR COOLANT PUMP ABNORMALITY?
Restore CCW to the thermal barrier within five minutes or initiate a MANUAL reactor trip and stor, 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 trip and stop all RCP'S ONLY.
'K3. 1 Knowledge of the effect that a loss or malfunction of the Component Cooling Water System will have on I 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 1- t~; Describe the plant response to actions taken in the.abnormal procedure Describe the final plant
- -condition that is established by the abnormal procedure. I 1
1 Tuesday, May 18,2004- 10:22.18
- AM 1 I Paae44of89 I
Given the following conditions for Unit 2:
- Pzr pressure - 2235 psig.
'- RCS temperature - 547 deg F.
- Channel Ill (PT-457) has been selected as controlling channel.
- 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?
Pressurizer Dressure will rise until.. .
ONE Pressurizer Code Safetv lifts.
El ONE Pressurizer PORV opens.
BOTH Pressurizer Spray valves open.
i BOTH Pressurizer PORVs open.
K4. 1 Knowledge of Pressurizer Pressure ControlSystem design-feature(s) and or interlock(s) which provide for the following:
'correct it by energizing all available PZR heaters, and closing both PZR spray valves. This will cause
'pressure to rise. The spray valves will not open due to the MPC seeing the low pressure signal input.
The PZR PORV's are NOT controlled from the MPC, and independently open from direct pressure 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, Channel I l l PZR pressure is ready LOW.
/PR1,PR2 PZR Power Relief Valves I
PZRP&LE007 Identify and describe the local controls and indications associated with the Pressurizer Pressure and Level Control system, including:
a) The location of Pressurizer Pressure and Level Control system local controls and indications b) The function of Pressurizer Pressure and Level Control system local controls and indications c) The plant conditions or permissives required for Pressurizer Pressure and Level Control system local controls to perform their intended function d) The setpoints associated with the Pressurizer Pressure and Level Control system local alarm-Not applicable to this lesson
- PZRP&LEOO9 State the setpoints for automatic actuations associated with the Pressurizer Pressure and Level Control svstem 1 Tuesday,-May 18, 2004 <0:22:18 AM 1 1 Page45of89 1
Given 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%.
What effect will this have on PZR heaters and spray valves?
- B/U heaters energize, and spray valves open.
B/U heaters - - remain off, and spray __ valves
- - open.
B/U heaters energize, and spray valves remain closed.
B/U heaters remain - __ off, and spray valves- remain _ - closed. - - - __
Pressure Control System:
because it contains the correct actions that will take place. All the distracters are wrong combinations of snrav and heaters
[Pressurizer Pressure Control System Operation . - - - - _ _ J IPZR pressure and level control IPZR txessure and level control PZRP&LE008 Identify and describe the Control Room controls and indications associated with the Pressurizer Pressure and Level Control system, including:
a) The Control Room location of Pressurizer Pressure and Level Control system bezels and indications b) The function of each Pressurizer Pressure and Level Control system Control Room control and indication c) The effect each control has upon Pressurizer Pressure and Level Control components and operation d) The plant conditions or permissives required for Pressurizer Pressure and Level Control system Control Room controls to perform their intended function i ........
..T... ... .. . ...
1 Tuesdav. Mav 18,2004 10:22:18AM I 1 Page46of89 1
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? - -
125 VDC Vital batteries supply the Vital Instrument Busses, which allow SI circuitry actuation
/and reset capabilities from SSPS.
- 28 VDC Vital batteries maintain control power to valve control circuits, which keep the valves
'from moving without a control board manipulation.
The Reset "S" signal is locked in to prevent valves from moving when 230 VAC power is
'restored.
Depressing the CLOSE PB for all Phase Aisolation valves on the main contrbl board.
n Reactor Protection
!movingwhen power is restored to the vital busses. The POWER for the SI actuation and reset comes
!from the Vital Instrument Busses, which will be powered from their respective 125VDC battery following ia loss of off-site and on-site AC power. Distractor b is incorrect because control board PB are just one
/way of initiating valve movement, and merely maintaining control power to a valve bezel will not keep it
/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
/tend to make it's valves move, rather than not move, when a RWST level of 15.2 feet is reached.
- Distractor d is incorrect because just pushing the close PB would not override a valid demand for
'movement from any automatic signal.- .. - _ - - - __ . __ - .- . __
I LOPAOOE008 I Given a Step, Caution, Note, or Continuous Action Summary Item in EOP-LOPA-1, state its bases I _ Tuesday, May 18,2004 I O
What would be the effect on the Reactor Protection System if the 2 6 Vital Instrument Bus were to become deenergized with the unit at 100% power? _- - . ___ -
2RP4 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 Iof 2 45VDC power supplies to Train A I 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 113 due to channel II 1
,bistable tripped.
[Plant Systems IQl21)_C1OK201 ;
Reactor Protection System __ _ . .J . .
42i rains receive power 115 VIB. Therefore they will not disagree on contact status because the loss of power will affect them both the same. Distracter b is incorrect because SSPS train A will not be affected by the loss of 2B VIB except as noted above for distracter a, and that will not cause an OHA A-34. Distracter d is incorrect because 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 214 to 1213. C is correct because 2B VIB supplies power to the slave relays on SSPS Train B, so none of the have relavs will actuate.
loverhead Annunciators Window A
[Solid State Reactor Prot Train A AC Power Distributibn 1 i
RXPROTEOI1 State the power supply to the SSPS RXPROTEOZO including:
a) The Control Room location of Reactor Protection System control bezels and indications b) The function of each Reactor Protection System Control Room control and indication c) The effect each Reactor Protection System control has upon Reactor Protection System components and operation d) The plant conditions or permissives required for Reactor Protection System Control Room controls to perform their intended function e) The setpoints associated with the Reactor Protection System control room alarms
[ _-Tuesday, May 18,2004 10.22:19 AM - __
] I Page48of89 I
Given the following conditions:
- 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-I with RCS pressure at 1400 psig.
- RCS cooldown rate is 120 deg F/hr.
- 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, MULTIPLE STEAM GENERATOR DEPRESSURIZATION?
TriD all RCP's.
Minimize AFW flow.
Stop BOTH RHR pumps.
- Send L .. . . operators .... to . close all BFIS's,
.. ..... . ........ ..... ... . .. .. ... .. .. ___ - ._ . BF~O'S, ___ BF22's.
and .. ..
~ .. -.
........ . ... . I IComprehension 1 Salem 1 & 2
'TIAbility to (a) predict the impacts of the following on the Engineered Safety Features Actuation System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of 1 those abnormal__operation: - _ _ _ 1 A2.03 IRapid depressurization 11m 4.71 Once out of TRIP-1, no actions other than attemptingto close MSLl valve are taken in LOSC-1 prior to going to LOSC-2. Maintaining >I E4 lbmlhr to each S/Gkeeps tubes from drvinq out, amonu other things. Do not trip RCP's because pressure is dropping dueto cooldown. Doeskt matter ifit's 1
7 IDirect From Source I
[ Tuesday, May 18,200410:22 I ~age49of89 I
Which of the following choices identifies the purpose of the Containment Spray System?
Maintain containment pressure less than the design pressure of 47 psig following a Loss of Coolant
_ _ - Accident (LOCA). -- - __ - - __
Maintain containment pressure less than the test pressure of 54 psig following a Main Steam Line Break (MSLB) inside containment.
l!/ Inject a mixture of borated water and Sodium Chloride (NACI) into the containment atmosphere following a LOCA to minimize exposure to the public following a LOCA.
,Inject a mixture of borated water and Sodium Hydroxide (NAOH) into the containment
,atmospherefollowing a MSLB with failed fuel, to minimize exposure to the public.
I IConcept Used
[ Tuesday, May i8,2004 10:22:19AM 1 I Page50of89 1
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 hours later, containment H2 concentration has remained at 2%, and containment pressure has risen to 5.1 psig.
IAW S2.OP-SO.CAN-001 HYDROGEN RECOMBINER OPERATION, which of the following actions is correct with ____ regards
- to the operation of the H2 recombiners?
recombiner to assist in lowerina containment H2 concentration. 1 Lower the Recombiner Power Adjust Potentiometer setting by 4KW below the previous setting.
1 Secure 21 Recombiner due to containment - -_ - __ pressure - rising
- _ _ _above
- 5 psig.
. . . ___ - I A4. JiAbilityto manually - -operate and/or monitor in the
_-__ - control room: _ .
pm3.91 ause it Distracter c is incorrect because rising pressure > 5 psig does not require recombiner S/D. D is correct because if H2 concentration has remained stable or dropping, with a rise in containment pressure, ,
8
,recalculation of the Dower settina is rewired.
Hydrogen Recombiner Operation e) Reactor Shield Ventilation System f) Containment Pressure - Vacuum Relief System
- 9) ner CONTMTE008 ldent co ons inm ent Support Systems, including:
a) The Control Room location of Containment And Containment Support Systems control bezels and indications b) The function of each Containment And Containment Support Systems Control Room control and indication c) The effect each Containment And Containment Support Systems control has upon Containment And Containment Support Systems components and operation d) The plant conditions or permissives required for Containment And Containment Support Systems Control Room controls to perform their intended function e) The setp dS terns larms 1 Tuesday, May 18, 2004 10:22:19 AM 1 I Page 51 of89 1
r .... ..... . .
1
~
i Tuesday, May 18,2004 10.22:20 AM I I Page52of89 I
'Given the following conditions:
'- Salem Unit 2 is 7 days into a refueling outage.
- The core is partially offloaded with 7 bundles remaining in the Rx.
- Fuel movement is in progress, and S2.OP-I0.Z-0010 SPENT FUEL POOL MANIPULATIONS is in effect.
- SFP temperature is 120 deg. F.
- 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.
1'If 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.
Potential Rx cavity overflow due to rising SFP level - - ___ -if the Gate Valve - remains - - open.
Inability to place a raised Spent Fuel bundle into any location in the pool due to rising radiation level on 2R32 Fuel Handling Crane Area Monitor.
%K3. ] Knowledge of the effect that a loss or malfunction of the Spent Fuel Pool Cooling System will have on the 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 Abnormal Radiation i
SFPOOOE007 Identify and describe the local controls, indications, and alarmsassociated $h the Spent Fuel Pool Cooling System, including:
a) The location of Spent Fuel Pool Cooling System local controls and indications. (N/A STA) b) The function of Spent Fuel Pool Cooling System local controls and indications. (N/A STA) c) The plant conditions or permissives required for Spent Fuel Pool Cooling System local controls to perform their intended function. (N/A STA)
ABSFOlE002 1
...................... L................................................................ ............................................. ........... . . . . . .
1.. . Tuesday, May 18, 2004 10:22:20 AM 1 1 Paae53of89 1
[ Tuesday, May 18,2004 I O I Page54of89 1 As Unit 2 Main Turbine power is raised from 20% to 1OO%, Programmed Steam Generator Narrow Range Water Level will ... .. _ _ __ -
rise from 33% to 44%. -. __ -
Elremain constant at 33%
remain constant at 44%.
Istation Drawing - - I ADFWCSE003 I List the reasons for programming Steam Generator level I I I I
1 Tuesdav. Mav 18. ,200410:22:20 AM I I Page55of89 I
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 S2.0P-AB.STM-0001, the control room crew will.....
initiate a Main Steam Line Isolation__ (MSLI)
.- - __ ONLY.
initiate a Safety Injection (SI) ONLY.
trir, the reactor. confirm the trip and initiate a MSLI.
trip the reactor, initiate -a- SI. __ - -
p2. ]'Abilityto (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: i A2.04 ! [Malfunctioningsteam dump ]p2Fl 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-I. The Idistracters a and b are incorrect because they contain only 1 of the 3 required actions. Distracter d is incorrect because the operator attempts to isolate the steam leak prior to initiatina SI.
[ExcessiveSteam Flow I
ABSTMlE004 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.
Material Required for Examination ' . I- .~ ~- . _
QuestionSource:
I
~ 1 fiG Exam Bank 1 )QuestionModification Method: '
Question Source I
Tuesday, May 18,2004 10.22:20 AM 1 I Page56of89 I
Given the following conditions:
- Salem Unit 2 is operating at 12% power.
- Control Bank D rods are at 189 steps withdrawn in MANUAL.
- 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
- Reactor power assuming NO .
other operator action? _ _
,Tavewill lower, Rx power will lower.
Tave will lower, - _- Rx power will rise. __
Tave will rise, Rx power will rise.
Tave will rise, Rx power will lower.
a 63 F I ! K n o w l e d g e of the operational implications of the following concepts as they apply to the Main and Reheat Steam System: - _- - . ___ -- - _ _ . - _-
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.
[Steam Dump System Operation I i
STDUMPEOO8 Describe the purpose and operation of the following Steam Dump Controllers, a) Steam Pressure Controller b) Load Rejection Controller c) Plant Trip Controllers I STMGENE008 Identify and describe the Control Room controls, indications, and alarms associated with the Steam Generator, SG Blowdown and Drain Systems, including:
a) The Control Room location of Steam Generator, SG Blowdown and Drain Systems control bezels and indications (N/A NEO) b) The function of each Steam Generator, SG Blowdown and Drain Systems Control Room control and indication (N/A NEO) c) The effect each Steam Generator, SG Blowdown and Drain Systems control has upon Steam Generator, SG Blowdown and Drain Systems components and operation (N/A NEO) d) The plant conditions or permissives required for Steam Generator, SG Blowdown and Drain Systems Control Room controls to perform their intended function e) The setpoints associated with the Steam Generator, SG Blowdown and Drain Systems control room alarms I Tuesdav. Mav 18.2004 10:22:20 AM I I Page57of89 I
Salem Unit 2 is operating at 80% power with the following conditions:
- 21 condensate pump C/T.
- 23 Heater drain pump 01s.
- PT-506, Steamline Inlet Pressure Channel is 01s.
- Main condenser backpressure is 2.1"Hg.
I-I 21 and 22 SGFPs are in service.
I
'- 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 anv. is IMMEDIATELY reauired of the crew?
'TriD the Reactor. and GO TO 2-EOP-TRIP-1. I Verify
- Automatic
- Turbine Runback has
____ or is occurring.
Ensure Rod Bank Selector Switch is in AUTO.
No other IMMEDIATE___ ACTION is required. - . . ..
,2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
1 4
. 0 1
~
/Main Feedwater / Condensate System Abnormality I
[ Tuesday, May 18,2004 10:22:20AM I Page58of89 I
Given the following conditions:
- Salem Unit Iis operating at 80% power.
- A large quantity of river grass starts building up on the Circ water traveling screens and condenser waterboxes.
- A rapid power reduction is initiated IAW SI .OP-AB.LOAD-0001 RAPID LOAD REDUCTION, 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.
1- 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?
High power reactor trip (low range) at 25% on 214 PR Nl's.
High- power -
reactor trip (low--range)
- __ - at --
20% power equivalent IR current on ----
212 IRNIG.
i
,High power reactor trip at 109% on 2/4PR Nl's.
/OverDower rod block at 103% on 1/4PR Nl's.
I IApplication ;Salem 1 & 2 1/2004 118 51 K5.j Knowledge of the operational implications; of the following concepts as they apply to the Main Turbine 1 coincidence are correct and it would actuate before any high power trips. Distracter b is incorrect because the coincidence is 112 not 212 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.
Salem UFSAR I i - .... .....- .-- - .- - -- - -- ...... .- .-. . .................. ..... . .-- -. .. . . . ......... .... - .
~ __ -.
. .. . -~ .. .... . .. .. 1 tions. includina u a) Name of the trip or safetyhjiction '
b) Setpoint and Coincidence (NIA NEO) c) Any related permissives or block signals (N/A NEO)
.. .. - --~ ..... .
- 1. Tuesday, May 18, 2004 10:22:21 AM I I Page60of89 I
'Given the following conditions:
- Salem unit 2 is operating at 100% power.
I- All control systems are in AUTOMATIC.
- OHA G-7, ADFCS SWITCH TO MANUAL alarms.
- 24BF19 valve demand is steady and indicates 5% lower than the 21-23BFl9 valve demand.
- 24 SG NR level is 41% and dropping slowly.
IAW S2.0P-AB.CN-0001, which of the following describes __ __
the required actions?
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 t o maintain 24 SG level constant.
!Ensure 24BF19 and 24BF40 are in MANUAL and depress the open pushbutton on 24BF19 to
'raise 24 SG level.
Place the SGFP Master Speed Controller in MANUAL and adjust SGFP speed to raise 24 SG level.
611112004 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.
Main Feedwater/Main Condensate Abnormality I
I Describe, in general terms, the actions taken in S2.OP-AB.CN-0001and the bases for the actions in accordance with the Technical Bases Document.
............ ...... ......... .... J 1
I Tuesday, May 18, 2004 10:22:21 AM I Page61 of89 1
Given the following:
- Unit 2 is at 90% power.
- All four SG levels are being maintained in automatic.
- SGFPs are in AUTO.
- 24 SG NR level Channel I is failed high due to an apparent electronic problem. No corrective actions have been initiated yet.
Assuming no operator action, which one of the following describes plant response if 24 SG Narrow Range Level Channel - __ Ill D/P cell develops - - a leak, causing- Channel - Ill level _ - to read 23%?
[24BF19 and 24BF40_ _ shift__ to- MANUAL. 21 _ _and
- 22
- -SGFP swap to MANUAL.
j24BF19 and 24BF40 ONLY shift to MANUAL.
ALL BF19's and 40's swar, to MANUAL.
,No control response. The ARBITRATION feature of ADFWCS shifts to Channel II for the control input.
- Kl..... . ] Knowledge of the physical connections and/or cause-effect relationships between Main Feedwater System and the following
u the same IG-I 5 G-7 OHA ARP
.::.:I
................ -' I I
. . 1 I I
I Tuesday, May _18,2004 10:22:21 AM
.......................................... ........I I Paae62of89 I
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 I
/With NO operator action, which of the following choices describes the indications which would be jpresent 2 minutes after the reactor trip?
~ - - - _ _ - ____ - - __ - -
AFW flow indication reading 0 gpm for 21 and 22 SGs.
AFW flow indication reading - - 0 gpm__for - 23 and 24 __ -SGs. - __ - - - - __-
23 and 24 SG wide range (WR) levels rising slower than 21 and 22 SG-WR levels.
2land
_ _ SG wide range (WR) levels ___ __- rising - -slower than
- __ - __ - 23 and - _- -24 SG WR
- ____- levels.
'K6.I:Knowledge of the of the effect of a loss or malfunction on the following will have on the Auxiliary /
[Emergency_ _Feedwater
- System: - _ - - ____ . - - __ - _-___-- - .- -----
I
,K6.01 Controllers and Positioners -
112.51 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 AF1 I s .
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.
r - . . . . .. ..
[Reactor Trip or Safety Injection - __ - - _ _ _- J AFWOOOE006 Describe the interlocks associated with the following Auxiliary Feedwater System components:
a) Auxiliary Feedwater Pump Automatic Start b) Motor-Driven AFW Pump Recirculation Flow Control Valves c) Motor-Driven AFW Pump Discharge Flow Control Valves I
I I
[ Tuesday, May I 8 2004 10:22:21 AM ] I Page63of89 1
Given the following conditions:
i- 2C EDG is operating in parallel with the 5 0 0 W 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-
/ST.DG-O014,2C DIESEL GENERATOR ENDURANCE RUN.
!- 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.
Which of the following choices describes the consequences, if any, of continued EDG operation at this KW load?
Operation for the remaining 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> of the test ...
will result in exceeding the 30 minute load limitation for 2C EDG.
will result in exceedina the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> load limitation for 2 C EDG.
I F -
'will not have any adverse effect on 2CEDG.
A I . __ JiAbilityto predict and/or monitor changes in parameters associated with operating the A.C. Electrical Distribution controls including:
'the 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 oDeration between 2750-286OKW.
12C Diesel Generator Endurance Run -- __ - - - I and precautions associated witheach operating procedure which are requried to be considered by &hr; Licensed or Non-Licensed Operators.
1 Tuesday, May 18,2004 10 22 -21_AM 1 I Paae64of89 I
Given the following conditions:
- Salem Unit 2 is performing a reactor startup.
i- Reactor power is stable at 1x1OE-8 amps.
1- OHA B-3 2A VTL INSTR BUS INVRT FAIL alarms, accompanied by Aux Alarm Typewriter Point
'147 2A VITAL INSTR BUS INV TROUBLE.
Using Attachments 1 and 2 of S2.0P-S0.115-0011 2A VITAL INSTRUMENT BUS UPS INVERTER OPERATIONl determine the status of the 2A Vital Bus Inverter and its effect on the Rx startup.
The 2A Vital Bus Inverter... __ - -
has undergone a latched transfer, the reactor has tripped when IN35 Intermediate range detector momentarily lost power.
,has undergone a latched transfer, Rx startup may continue with no restrictions.
1 has undergone ___ an - unlatched
- transfer, Rx_ _startup may continue with no
_ _ restrictions. 1
[has undergone an unlatched-transfer, the reactor has tripped when 1N35lntermediate range
/detector momentarily lost power.
561 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal i 1
ON INVERTER lamp will be lit, indicating the static switch momentarily swapped to the alternate source and 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.
12A Vital Instrument Bus UPS System Operation e location of 115 Vac ElectricalSystems local controls and indications e function of 115 Vac Electrical Systems local controls and indications e plant conditions or permissives required for 115 Vac Electrical Systems local controls to perform their intended lMaterial Required for Exaaination 6 S2.0p-sO.115-0011 Attachments 1& 2 1
[:;-,-Tuesday,
.............. May__ 18,2004 10 I Page65of89 1
Which of the following choices contains ONLY components suppied power from the Unit 2 125VDC svstem? I Control Room Emergency Lighting, 13KV- switchgear, - SCADA system.
Pzr heater bus switchgear, EDG control power, 2RP4 status lights.
Alternate
__ Shutdown Syste-m (ASDS), IRPI, 460V switchgear. __
Vital Instrument Bus inverters, 4KV switchgear, SGFP Emergency Bearing Oil Pump.
223720. The2RP4 status panel receives 28VDC power from bkrs 2A11DE and 2B12E, shown on drawing 21 1357. The SCADA system ALSO receives 28VDC power. ASDS supply is also shown on 223720. IRPl is Powered from 23OVAC and 28VDC. SGFP EBOP is 250VDC load.
]Unit 2 125 V.D.C One Line I ..--
I Tuesday, May 18, 2004 10:22:22AM 1 1 Page66of89 I
Given the following conditions:
- Salem Unit 2 is operating at 30% power.
- 125 VDC breaker 2BDClAX12,ZG 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 t J PB on the breaker locally.
1
,The reactor is tripped, and 24 RCP breaker can only be tripped by depressing the STOP PB from the Control Room.
The reactor is tripped, and 24 RCP breaker is tripped. - __ - _ _ - .
The reactor i s at 30% power, and 24 RCP breaker is tripped.
following: I 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. Distracter c is incorrect because there is no demand for a reactor trip or breaker trip on a loss of control power to the pump breaker. Distracter d is I, I
,incorrect because the reactor is not tripped.
LRCS Reactor coolant - _ _pumps and lift oil pumps
- - - __ I
-.IRCS Reactor coolant pump
..... I DCELECE007 Identify and describe the local controls, indications, and alarms associated with the DC Electrical Systems, including:
A. The location of DC ElectricalSystems local controls and indications B. The function of DC ElectricalSystems local controls and indications C. The plant conditions or permissives required for DC Electrical Systems local controls to perform their intended function D. The setpoints associated with the DC Electrical Systems local alarms 4KVACOE007 Identify and describe the local controls, indications, and alarms associated with the 4160 Electrical System, including:
a) The location of 4160 ElectricalSystem local controls and indications. (N/A STA) b) The function of 4160 Electrical System local controls and indications. (N/A STA) c) The plant conditions or permissives required for 4160 ElectricalSystem local controls to perform their intended function.
(N/A STA) d) The setpoints associated with the 4160 Electrical System local alarms (N/A NEO)
I Tuesday, May 18,2004- - 10.22:22 AM I Page67of89 I
Given the following conditions:
- 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 22 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.
I Overfilling/overflow protection is provided by the DFO Day Tank high level overflow connection
'back to the DFOST. - __ - - - - __ - - - -
r - __-- - __ - - - - __ - - - -
'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 back to the DFOST.-
21 DFOTP must be manually started but the high level trip will automatically stop the pump to provide protection against overfilling/overflow of the DFO Day Tanks.
2.1 //Conduct Of Operations - - - ___ - ___ - - - - - -
,2.1.28 J IKnowledge of the purpose and function of major system components and controls.
113.21 Refer todrawing 223825 during this explanation. When the 21 FOST local EMERGENCY/NORMAL Switch (FTP-3) is taken to EMERGENCY, contacts 1,3,5,and 7 close, and 2,4,6,8 ouen. Contacts 1 and including:
a) The Limiting Condition(s) for Operation b) The Bases for the LCO(s) c) The applicability of the LCO(s) (N/A NEO) d) The LCO Action Statement@) (N/A NEO)
.. .... .. ... -- . ..... . . ... . -.. ... ... ___ .... ..... .. . . ....... .. ~--. . . . ~.--... . .. . . .
-... ...~. .-.... ....... -,
I Tuesday, May 18,2004 10:22:22AM ] I Page69of89 1
,Which 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-MONITORAND 2R9 FHB NEWFUELSTORAGE AREA BADIATION MONITOR, must be in alarm to cause FHB ventilation exhaust fans to start and iexhaust throuah a charcoal filter.
2R32A FUEL HANDLING CRANE AREA MONITOR in warning will cause all Fuel Handling Crane functions to be locked out.
EITHER 2 ~ 5SFP , 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.
,2R32A, Fuel Handling Crane Area Monitor in warning will cause-downward Fuel Handling Crane movement- to -be locked . . . . - out.
El j B 60 I
K4.1Knowledge of ARM system design feature(s) and or interlock(s) which provide for the following:
,K4.02 kuel building isolation 113.2*/13.4*1 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. Distracter a is incorrect because only 1 monitor is needed to be in alarm, not BOTH.
Distracters b and
......... .......d ...are ....... incorrect
...............because
...............the R32A
.......... only locks
................. out
......OUTWARD
......... rod motion Fuel Handling Building Ventilation - ..
r
' RMSOOOE006
- - __ Describe the interlocks associated with the following Radiation Monitoring System components:
A. R1B, Control Room Inlet Duct Monitor B. R5, FHB - SFP Area Rad Monitor C. R7, In Core Seal Table Area Rad Monitor D. R9, FHB - New Fuel Storage Area Rad Monitor E. RIOA, Personnel Hatch 100 Elev.
F. R1OB, Personnel Hatch 130 Elev.
G. R I I A , R12A, R12B, Containment Particulate, Noble Gas, and Iodine Monitor H. R13A,B,C,D,E CFCU Service Water Monitors I. R17A and B, Component Cooling Liquid Monitor J. R18, Liquid Waste Disposal K. R19A,B,C,D, Steam Generator Blowdown Liquid Monitors L. R32A, Fuel Handling Crane Area Rad Monitor M. R36, Evaporator and Feed Preheaters Condensate Monitor N. R41D, Plant Vent Radiation Monitor
- 0. 2R52, Liquid Pass Room Area Rad Monitor 1 FHVENTEOOG Describe the interlocks associated with the following Fuel Handling Area Ventilation System components:
a) Supply Fan Unit b) Exhaust Fan Controls
, .............................. ...... . ....... ............................................... ..................... ..... ... . . . . . . . . I I
i Tuesday, May........18,2004.....
10:22:22 AM
.......................... ................. 1 1 Page70of89 1
1 Tuesday, May 18,2004
-_ __ - AM
- 10.22'23 -
1 1 Page 71 of 89 1
Given the following conditions:
8- Salem Unit 1 is operating at 100% power.
'- Control room operators are preparing to perform a Containment Pressure Relief IAW S I .OP-1IS0.CAN-0002, CONTAINMENT PRESSURE-VACUUM RELIEF SYSTEM 0PERATI 0N.
'- Containment radiation levels are NORMAL for 100% power operation with no failed fuel.
After opening the 1VC5 and lVC6 to initiate the pressure relief, which of the following choices describes how the respective radiation monitors indication will respond?
1R12A- Containment Gas Effluent 1R41B- Plant Vent Noble Gas Intermediate Range -
!I R41D- Plant Vent Noble Gas Release Rate 1R12A rises; IR41B constant; 1R41D constant. ___ -
1R12A constant: 1R41B rises: 1R41D constant. I AI. 1 Ability to predict and/or monitor changes in parameters associated with operating the Process Radiation
,Monitoring System controls including:
flow will start when the lo range 1R41A 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 (uCilsec) by combining (product of) the on-range R41A through R41C with plant vent flow (cc/sec). It will rise when the pressure relief is initiated, and also A. The Control Room location of Radiation Monitoring System control bezels and indications (N/A NEO)
- 8. 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 operation (NIA NEO)
D. The plant conditions or permissives required for Radiation Monitoring System Control Room controls to perform their intended function I ...... ........................ J
'I 1 i Tuesday,
- - May 18,2004 10:22.23 AM ] [ Page72of89 1
I _Tuesday, May 18,2004 10:22:23 AM J I Page73of89 I
1
~
iWhich one of the choices describes a difference between Unit Iand Unit 2 SG Blowdown RMS ....
ichannel response? . . . .- ........................... ... ................. I
.On Unit 1, a WARNING on an R-I 9 closes the respective GB4, Blowdown Isolation Valve. On 1 Unit 2, it does not.
!On Unit 1, an ALARM on an R-19 closes only the respective GB4. On Unit 2, the same signal
/closes all interlocked blowdown valves.
'On Unit 2, an ALARM on an R-I 9 channel closes just the respective GB4. On Unit 1, the same signal _ _ - closes all interlocked blowdown _ _ - __ valves. - - __
On Unit 2, a WARNING on an R-19 closes the respective GB4, Blowdown Isolation-Valve. On Unit 1, it does not. -_ - -- __ -.- -
K1. IiKnowledge of the physical connections and/or cause-effect relationships between Process Radiation
,MonitoringSystem __ and the following: __ -
,K1.01 /Those systems served by PRMs l/3.6l NG. On Unit 1, there are ion monitors as a m I Radiation.:monitor : response to high radiation; :.-:..:.:including : that occur as aI result of the channel :...
a)__.-..- .................. .......... 1.. actions
...... ___._ in warning or alarm.
. J I Tuesday, May 18,2004i0.22 23 AM 1 I Page74of89 1
Given the following conditions:
'- A Loss of the 500kv switchyard has occurred.
- 4KV Vital bus 1A has been de-energized due to a Bus Differential relay actuation.
- Salem Unit 1 has initiated a MANUAL Safety Injection (SI).
Which of the following choices identifies the Service Water pumps that will be running 2 minutes AFTER the SI has been initiated?
14 and 15.
- I H ' l 2and 16.
will not start. C is correct because the C bus SEC will start the C bus lead pump ( I l ) , and the B SE will I start whichever pump is selected to "LEAD" (13 or 14). The "A"bus pumps (15 and 16) cannot start 1 because their resDective 4kv vital bus is deeneraized I Service Water Pump Operation a) Mechanical Trash Rake b) Traveling Screens c) Service Water Pumps d) Service Water Strainers e) SumpPumps Ventilation Fans & Heaters I
Tuesday, May 18, 2004 10:22.23AM ] I Page75of89 I
Given the following conditions:
- Salem Unit 1 is operating at 100% power.
- Unit 1 ECAC is shutdown and aligned for normal AUTOMATIC operation.
- 11 Chilled Water pump is in operation.
- 12 Chilled Water pump is in standby.
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? - - __
12 chilled water pump starts __ and I 1- chilled - ___ __water - -pump- -stops
- after a 2 second - __ time
- - delay.
12 chilled water pump remains in standby and IIchilled water pump remains running.
12 chilled water pump starts and I -Ichilled
- - ____- --- water
- pump _ _ - remains _ _ running. _ _ - _ -
12 chilled water pump is locked out from starting in automatic while I 1 chilled water pump is
'running.
nal to 11 CH pump. It does not trip 11 pump if running. So with the conditions in the stem, 11 will remain running
- ________ __ - _____ and- 12 will start. _____________ - - - - >
/ControlAir System Operation ._ _ -
I ,
I I I
[ __ 10.22.23 AM Tuesday, May 18,2004 1 I Page76of89 I
Given the following conditions:
- All 3 Salem Station Air Compressors have become unavailable.
- The NORMAL cooling water supply to the Unit IEmergency Control Air Compressor (ECAC) has been lost.
!Operation of the Unit 1 ECAC.. ..
can continue since cooling water will automatically swap to Demineralized Water through a I
icheck valve.
must be discontinued until cooling water can be manually aligned through a spool piece from Demineralized Water.
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.
[Exam Date:l ~- 6/11/2004
~078000K104 078 Instrument Air System 65 K1. ]'Knowledge of the physical connections andlor cause-effect relationships between Instrument Air System and the following: l a SUPP~Vand a return s ~ o otiece. l IService Water-Nuclear Drawing __ - --- ____ - I Control Air System Operation i) Redundant Air Supply Panels CONAIRE007 Identify and describe the local controls and indications associated with the ControlAir System, including:
a) The location of Control Air System local controls and indications b) The function of Control Air System local controls and indications c) The plant conditions or permissives required for Control Air System local controls to perform their intended function CONAlREOl2 Describe the procedures which govern the operation of the Control Air System including, significant prerequisites and 1 precautions associated with each operating procedure which are required to be considered by either Licensed Operators or Non- \
Licensed Operators. ...............
[ Tuesday, May 18,2004 10:22 23 AM I I Page77of89 I
From the 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 + I 1.O.
- Operation at 100% power for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> with
~ - RCS net unidentified leakage = 0.08 gpm.
I - 24___ SG I .......... . -... primary to secondary
_ .. ._............................ ___ ...leakage
= 0.4 gpm.
- Operationat 80% power for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with:
, - Power Range Channel N-41 inoperable and in tripped condition.
- -- QPTR = I.01 as
.............. .-. indicated by BEACON.
~
fp [o-.-. -............................ ____... ... ....... .....................
peration 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:
~
i 1 - Reactor coolant specific activity of 10E-2 uCi/gram DOSE EQUIVALENT 1-131.
i - Reactor coolant specific activity of 50/Ebar uCi/gram.
,2.1 I Conduct of Operations
,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
/HSBmust 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. Distracter C is incorrect because 1 power range 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 I
,QPTR is monitored 1x112 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
/to be in violation. Also the E-bar limit is 1 OO/E-bar.
I Salem Technical SPecifications ! ... I Core Operating Limits Report I
NFS-0231,Core Operating Limits Report, Unit 2,Cycle-14, Fiqure 2,"Axial Flux I..Tuesday, May 18,2004 10:22:24 AM
.................................... .... -_.............. ............ ] I Paae78of89 I
'Which of the following choices-contains ONLY-items that are described in I O CFR 50.46 as Acceptance Criteria for Emergency Core Cooling systems for light-water nuclear power reactors?
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.
,MI.The zirc-water reaction postulated to occur at 1800 deg. F will not become self sustaining under any circumstance.
I , Ill, v, VII.
L~
. .__ I
~ ..
~ CY ll\l \ I \/I \Ill U j l V , V ) VI, VII.
la I. II. Ill. IV.
2.1 1 Conduct of Operations 2.1.28 [Knowledge of the purpose and function of major system components and controls. 1Wb3 10CFR46 paragraph b delineate the Acceptance Criteria for a LOCA. Also Salem FSAR contains the same criteria copied from 10CFR46 in section 15.4.1.1. The only answer which contains ONLY correct 1
/is answer b. All the distracters contain either numeral I or VII. neither of which are correct. 1
[ Tuesday, May _ _
18, 2004 - 10 1 Page79of89 I
)All Precautions and Limitations for the procedure SHALL be reviewed and compared against kurrent plant conditions, to minimize the possibility that a system operating limit will not be exceeded
- __ inadvertently
___ - - during
__ __ procedure -- ---_ ---___ performance.
An Infrequently Performed Test or Evolution (IPTE) brief SHALL be held-if the procedure has not been performed within the past 6 months to familiarize operators with the procedure prior to use. __ __ - _- - - - -_ --__-
12.1 I'Conductof Operations I
,2.1.32 ' /Abilityto explain and apply all system limits and precautions. JW3.4113.81 because expedited performance is not a reason to N/A procedure steps, and the unit CRS must
,approve any such N/Ainq. Answer b is correct because P&L are required to be reviewed and signed
'p;ror to reaching any performance step, and they guard against inadvekently exceeding.system - limits.
lUse of Procedures I Nuclear Procedure Program 1 I
Conduct of lnfreauentlv Performed Tests or Evolutions
~~
- PROCEDEOOG
- State the requirements associated with the use of the follow%g types of proceduresin accordance with NC.NA-AP.22-0001, Nuclear Procedure System:
- a. Category I
- b. Category I I
- c. Category 111 L ............................. .......... .-..
. ~...... .____ . ......... ............ ,I 1- Tuesday, May 18,2004 10:22:24- AM- ___ 1 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.
,Which of the following choices identifies the actions required to raise power to 18% IAW S2.0P-
!SO.MS-0002 STEAM DUMP SYSTEM OPERATION?
DUMP PRESSURE SETPOINT DECREASE pushbutton, then withdraw Control Rods in MANUAL or dilute the RCS to establish 18% Dower.
- Raise the Main-Steam Dump PRESSURE MODE-AUTO setpoint by depressing-the-STM
/DUMP PRESSURE SETPOINT INCREASE pushbutton, then withdraw Control Rods in
'MANUAL
- - . - or borate the RCS ___ to
__ establish
- 18%
-- _________ - -power.
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 on programmed value. - __
-- - - __ ___ - - - . - _ - ~
I. _ _Y.._ _...._
..?. _
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
/Turbinehas not been synchronized. IOP-3 step 5.4.20 initiates a power ascension to 10-20 % using jMain Steam Dumps IAW SO.MS-002, Steam Dump Operation. Section 5.4 of SO.MS-0002 identifies ithe actions to be performed to raise Rx power (Step 5.4.1.B) 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 wrong control mode of MANUAL.
I . ...I STEAM DUMP SYSTEM OPERATION i
, IOP003E004 Recognize the actions which are required or may be taken when power reaches various levels listed below:
a) SR Permissive light is energized (P-6).
b) 214 PRs are greater that 10% (P-IO).
13 1 .O x 10-8 IR Amos i MSTEAMEOOB Identify and describe the Control Room controls, indications, and alarms associated with the Main Steam System, including:
a) The Control Room location of Main Steam System control bezels and indications b) The function of each Main Steam System Control Room control and indication c) The effect each Main Steam System control has upon Main Steam System components and operation
~ ....
d) The plant conditions or permissives required for Main Steam System Control Room controls to perform their intended function e) The setpoints associated with the Main Steam System control room alarms
.....RXOPERE021 Explain the relationship between steam flow and reactor power given specific conditions.
1 I I 1 Tuesday, May 18.2004 10 22 25 AM I I Paae82of89 I
On November 21st at 1300, with Unit 2 in Mode 1, it is discovered that the monthly surv&llance to verify that ECCS piping is full of water has NOT been performed since October 1Ith at 1300.
Which of the following choices identifies the LATEST time for satisfactorily performing the required surveillance to prevent taking action per the LCO?
1300 on November 22nd.
1300 on December 12th.
1300 on December 21st._ _ -
1300 on December 22nd. I TS Amendment 237, in section 4.0.3 allows a Grace period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OR up to the limit ofthe specified surveillance frequency, whichever is GREATER, from time of discovery to SAT performance 11
/of the surveillance to prevent entering 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.bS2.(instem) A
/"month" in Tech Specs is always 31 days. 31 days from the time of discovery (11/21 @ 1300) is 12/22 @ 1300
-- ~ p _ _
- ~ ~ ~
I Technical Specification Surveillance
- - Requirements
~ ~ ~ ~ - ~ ~ .
-~
~ - -__I I
1 Tuesday, May 18,2004 10:22:25
_ _ AM 1 I Paae83of89 I
I
!Which of the following parameter limits is established to ensure that radiation releases will remain
- within i..... - the limits of 10CFR20? ........
'Secondary
..... .....system activity.
iLiauid Waste discharae activitv. 1
!Primary system activity. . . _ .......... ---- .......... .... - - --------........... .......
Primaw to secondaw leakage. ...... I Radiation Control I
activity, primary activity and primary to secondary leakage limits ensure conformance to 10CFR100
/limits.
~~-
kF . _ R 2~-0 1 I I I I Tuesday, May 18,2004 . -
10:22:25 ..
A M ......... 1 Page84of89 I
While performing a review of 21 GDT release paperwork IAW S2.OP-SO.WG-0008, it is noticed that the calculated maximum release rate is 28 scfm.
Which of the following choices identifies whether the release may be initiated, and why?
/The GDT release cannot be initiated because flow rates 32 scfm cannot guarantee that the
'dose received by a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY will not exceed 500 mrem/yr whole body.
The GDT release cannot be initiated because flow rates 32 scfm cannot guarantee that the dose received by Site Staff inside the Protected Area will not exceed 3000 mrem/yr whole 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.
a- J b __ -1 IMiii!ii . ..-
Salem1 & 2 __
] 6/11/2003
$2.3 I Radiation Control ____ . . __ __ . . - . . . . - - - __ ___ __ 1 2.3.1 1- IAbility to control radiation releases.
gamma dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE
/BOUNDARYto less than or equal to 500 mremlyear to the whole body and 3000 mremlyear 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 mremlyear." The waste gas
,releaseprocedure 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."
, WASGASE002
[ Tuesday, May 18,2004 10 22 1 Paae85of89 1
The crew is performing EOP-FRCC-1, 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?
Safety Injection flow.
COGdelta temperature (ThotITcold).
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 ,
I CET's I lR.esponse to Inadequate Core Cooling i ................................... ............................ .- .......... --................... ....... ............. ......... __ __ ._.. . ................ ... ___ ............ .__
I A. EOP-CFST-1, Fiaure 2 B. 2-EOP-FRCC-1 -
C. 2-EOP-FRCC-2 D. 2-EOP-FRCC-3 1
1 i
1 .. .. __. ... --- . .... .... ............. .. __. .. .__.-- - ...... .-. ............ -. ....... .I 1
I I IDirect From Source I
I Tuesday, May 18,2004 - __10.22-25 __AM 1 I Page86of89 1
Given the following:
1 1- Salem Units 1 and 2 are operating at 100% power.
'- 1A EDG is C K for maintenance.
,- A loss of off-site power occurs, and both units trip.
- - 1B EDG fails to start.
- Which of the following choices describes how the control room response will differ if the crew
/directlyenters 1-EOP-LOPA-I LOSS OF ALL AC POWER, instead of I-EOP-TRIP-1 REACTOR
'TRIP OR SAFETY INJECTION? ... .......................
,In LOPA-I , a manual Safety Injection will NOT be initiated by the operators prior to power
'restoration. TRIP-I requires a Safetv lniection to be performed prior to transition out of TRIP-I.
iln LOPA-I , Functional Restoration Procedures (FRPs) will NOT be implemented until after a
% transition out of LOPA-I is performed. In TRIP-I , FRPs may be implemented prior to transition out of TRIP-I .
LOPA-I verifies AFW flow to be > 22E4 Ibm/hr. TRIP-I verifies flow to be >44E4 Ibm/hr. I
- LOPA-1does NOT confirm the reactor trip; TRIP-I does confirm the reactor trip.
[Generic Knowledge and Abilities [I94001G406 2.4 1 Emergency Procedures / Plan _ _ _ - __ ._ - ___ - -
E T 6 d is correct because FRPs are designed with the assumption that at least 1 4kv vitalbus is energized so that the operator can perform actions to trip the reactor or borate the RCS. Distractor c is incorrect because LOPA-I verifies 44E4 Ibm/hr to verify that the TDAFW pump is supplying the minimum safeguards AFW flow for heat removal. Distractor b is incorrect because FRP's are implemented PRIOR to leavina LOPA-I at step 20 when DromDt restoration of a vital bus has occurred.
IL.oss of All AC Power Basis Document
.. - ........ ........ ____-_................... 1 IReactor Trip Response I
,_I .. ..
~
Tuesday, May 18,2004 ............... 10:22 1 Pase87of89 1
Given 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 IRP4.
- 3/3 Auto Stop Oil Pressure Low Bistables are lit on IRP4.
- IR NI SUR is negative.
- Rx power is 1.2% and lowering.
- IR NI indicate lowering flux.
- No other action has been taken.
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.
'Rx trip is NOT confirmed, Turbine trip IS confirmed. GO TO FRSM-1 RESPONSE TO NUCLEAR POWER GENERATION.
,Rx trip IS confirmed,
~.
....................................... ....... Turbine
......................... ____ trip IS confirmed. Verify any
__ 4KV Vital bus is energized.
~...
/Rx trip is NOT confirmed, Turbine trip is NOT confirmed. Open Reactor Trip Breakers.
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 A M when no other action opens the reactor trip breakers.. Distractor c is incorrect because the turbine trip is not confirmed. The action is correct for the status in c. Distractor d is incorrect because
'the i __ Rx trip is confirmed.- The action is correct for an A M (first alternate action
--_--_ __ __ ________________ _____ - _____ - - __ ___ - __ - - __ __ ____to_____
trip__the________ reactor). - - - ____
AJmmediate Actions B.-Continuous Action Summaries C.-Communications D.-Log Keeping E.-Application of Notes and Cautions F.-Transitions
' TRPOOl E017 I
[- Tuesday, M_a y_i 8 , 2-o o i i o 22:26 AM J I Page88of89 I
Changed stem conditions, and added what action must be performed.
1 Tuesday, May 18,2004 10:22:26 AM I I Page89of89 I
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.
i- OHA E-48, ROD BOTTOM.
1- OHA E-39, PR CH DEV.
The following conditions exist:
- PR-N43, N44 indicate 101% power.
- PR N41 indicates 103% power.
i-I PR-N42 indicates 97% power.
!- All Shutdown Bank and Control Bank group demand counters are at 225 steps.
'- IRPl 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?
AB-ROD-0002. DROPPED ROD.
AB-ROD-0004 , ROD POSITION-INDlCAT6N FAILURE. I El :AB-NIS-0001, NI SYSTEM MALFUNCTIONS.
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 stucklmisaligned procedure is NOT correct for a dropped rod. i IDrowed Rod I
' ABROD2E003
....................... - ........ 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 Tuesday, May 18,2004 10:22:49 AM I Page Iof34 1
Unit 2 has experienced a Rx trip and SI during a SBLOCA. These conditions are present after checking PORVs 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.
onse and procedure transition?
/LeaveALL RCPs running, transition to 2-EOP-LOSC-1 at appropriate step. .
I
~
'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.
[s- -- I Salem I& 2
/Emergency and Abnormal Plant Evolutions 2.4 ]:Emergency -- - Procedures. - - / Plan __ - __ _. __ - ___ __ - _ _ _ _ _ _ _ _._ _ _
12.4.6 1/Knowledgesymptom based EOP mitigation strategies.
1 i 3 __. l i 1 4 . 0 1
55.4375) 22 RCP 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 i:
'Step 28.1 due to Containment pressure > 4 psig. In the basis document for EOP-TRIP-1 page 44-45
'for 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 injection into the RCS". LOSC entry is NOT indicated.
Reactor Trip or Safety Injection (Basis)
I ... . .. . - .. . ...... ,
the following analyzed transientslaccidents 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.
~
[ Tuesday, May 18,2004 10 22.49 AM I
I Tuesdav. Mav 1.
2004-10:22:50AM- I
Which of the following choices describes a condition in which Rapid Boration is NOT successfully
,established IAW S2.OP-SO. ........... -. .......... .___. CVC-0008,........ .. RAPID BORATlON? ... .... .................. ..... -. . . . . . . . .
/Chargingflow 90 gpm on 2FI-128B 2 CVC CHG SYS FLOW INDICATING BEZEL, 2CV175 IRAPID 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-I28B, 2CV174 BLENDER BYP VALVE open, 2CV172 BA FLOW CONTROL TO BLENDER open, boric acid flow 37 gpm on 2FI-I 1OA 2 CVC BORIC ACID
~.FLOW
........... _. ... INDICATING BZL
~ ..... --. ..... - ............................. _____ ... ................................................ __ ... _ ...... __ . ...... .-- .............................. ... ...........................
-4 .Charging flow 80 gpm on 2FI-l28B,2CV175 shut, 2CV172 and 2CV185 M/U FROM IBLENDER TO CHG PUMP SUCTION LINE in MANUAL and open, boric acid flow 39 gpm on
/2FI-11
.............. - OA.
~ ,c .................................................
harsins ..flow -9.5.gp.m. .on~2Fi~l.2..8.B~~
.2cv;...7.5.
...shut
.,.-.*sJl. .RWsTTO c.~~G.~puMp~s ST0 p..M-o.v
.open,
....... 2CV40 VCT OUTLET STOP shut.
AA2. 1 Ability to determine and interpret the following as they apply to Emergency Boration:
AA2.01 IWhether boron flow and/or MOVs are malfunctionina. from Dlant conditions gpm. Distractor b,c and d are incorrectbecause all theconditions are met IAW step 5.2. 5.3, and 5.4 respectively, of the procedure. I 1......... ............... J iI ................. - - .......... -.................................... ._
1I I
I 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 Dernin, CV27 xiii) Reactor Coolant Filter xiv) DiversionValve, CV35 xv) CVCS Holdup Tanks xvi) Volume Control Tank xvii) VCT IsolationValves, 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) Charging to Loop 3 Valve, CV77, Loop 4 Valve, CV79 xxiv) PZR Auxiliary Spray Valve, CV75 xxv) CCP Flow Control Valve, CV55 b) RCP Seal Water I 1
' Page 4 Of34 ..
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. IsoI. Valves, CVI 16, CV284 vi) Seal Return Filter vii) Seal Water Heat Exchanger Excess letdown i) Excess Letdown IsolationValves, 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 IsolationValve, CV181 xii) Rapid Borate Stop Valve, CV175 CVCSOOE012
-. ._ Describe the procedures which govern the operation of the Chemical and Volume Control System, including significant prerequisites and precautions associated with each operating procedure which are required to be considered by either Licensed or Non-Licensed Operators. ....
.................... .- --- .. ........ .i i
.............. 18, 2004 10:22:50 AM 1
Given the following conditions:
- 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.
With NO operator action, which of the following choices describes the condition of the plant 2 minutes after the feed pumps trip, and whether or not this condition exceeds any design basis?
,The reactor and turbine have tripped. Condition is within design basis parameters.
The turbine has tripped, the reactor has NOT tripped, AFW pumps started when both SGFPs tripped. Condition is within design basis parameters.
The reactor has tripped, the turbine has NOT tripped. Design basis parameters will be exceeded. __ _ _ _ _ _
The reactor AND turbine have NOT tripped, AFW pumps have started at 9%-NR levels in 2/4 SGs. Design
-. - - ___basis parameters will be exceeded. - - - _ _ _ _
AA2. I Ability to determine and interpret the following as they apply to Loss of Main Feedwater:
u uj4.314.41
\55.43(2)Even when the RPS system does not initiate a Rx trip, the AMSAC system will initiate a turbine trip and initiate AFW flow. Distracter d is incorrect because, in this case, AFW flow was established when both SGFPs tripped, not from SG NR level (normal OR AMSAC actuated), and 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 Dsiq will be met 1 I
I Tuesday, May 18,2064 10 2250 AM 1 I Page6of34 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.
/Ensurethe 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.
ight from section 3/4.OBases. Distracter a is incorrect because the VIB doesn't supply the FHB exhaust fans, the 230V I bus does. The stem asks for the bases for the VIB. Distracter b is incorrect because the bases states that 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.
Salem Tech Specs I.. ............... ........ . __.-- .................... ................. .. ......... ................... .~
1 Svsterns. includina I ................. - .........
Tuesday, May 18,2004 - _ _ -
10 22 50 AM ]
,Given the following conditions:
1
- Salem Unit 2 is in MODE 3, NOP, NOT.
i-i The control room receives OHA B-I 8 2C 125VDC CNTRL BUS VOLT LO 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 be taken?
I C 125VDC bus is ... ....
within the normal operating band, direct maintenance to raise the charger float voltage.
experiencing a minor short-to-ground, initiate S2.OP-SO. 125-0004 125VDC GROUND DETECT1ON.
iabove the Tech Spec minimum setpoint, ONLY continued monitoring for any indication of ifurther voltage degradation is reauired.
ibelow the Tech Spec minimum setpoint, secure the operating battery charger and place the Istandbv battery charger in service.
d ......
Ts ...................
[Memory /Salem 1 8 2
/,Lossof DC Power II lAA2-. I Ability to__ determine
] -1m i l and _- ___
interpret __ the- -following
________ __ -as ___they
________apply to Loss of DC- -Power:__ - ____ - -
AA2.02.II25V dc bus - - _ _
voltage, low/critical low, alarm
- --_--_ - _-- - - - -- - -- - -- - ____-- ------ - ----- - --- - -- - -- -- -- - ---- --- - . - . - 1
'55.43(5) A is the correct answer because the control band as specified in the NCOs logs is 125- i 1139.8V. Voltage is in the normal band, and the AR states to have maint adjust the float voltage.
'Distractor b is incorrect because there is no indication of a ground. Distracter c is incorrect because
-action __ IS.-required --____ IAW ARP. __ Distracter d is incorrect because voltage is above the TS limit.
Control room Logs - ___ __ -- - _ _ - _ _
Overhead Annunciators Window C A. Batteries I B. Battew Charaers -
I DCELECE008 Identify and describe the Control Room controls, indications, and alarms associated with the DC Electrical Systems, including:
A. The Control Room location of DC Electrical Systems control bezels and indications
- 6. The function of each DC Electrical Systems Control Room control and indication C. The effect each DC ElectricalSystems control has upon DC Electrical Systems components and operation D. The plant conditions or permissives required for DC ElectricalSystems Control Room controls to perform their intended function E. The setpoints associated with the DC.... Electrical Systems control..room alarms I
[ Tuesday, -- -
May 18,2004 10:22:50 AM - - -1 I Page8of34 1
Which of the following conditions, if left uncorrected, will require the Salem Operations Department to make an- Off-Site __ - notification?
UnDlanned loss of DID and CENTREX Phone svstems.
iA sinnle Unit 2 Hvdronen Recombiner declared INOPERABLE in MODE 3.
[Lossof OHA Window Boxes H,J.K.
bility to determine and interpret the following as they apply to Loss of Instrument Air:
- recombiners is only applicable in MODES 1 and 2.. Distractor c is incorrect because >75% of OHAs
- must be lost before any action will be taken. D is correct because the 2DR6 fails open and will overfill the AFWST resulting in a spill of Hydrazine that will drain into storm drains, requiring a spill report IAW j
- EAL
_. ._ 11.5.2.b ..... - I iI
/Salem ECG
-7 CONAIRE007 Identify and describe the local controls and indications associated with the Control Air System. including: I a) The location of Control Air System local controls and indications b) The function of Control Air System local controls and indications I
AFWOOOE012 1 autions associated with each operating procedure which are required to be considered by either Licensed or Non-Licensed 1 Tuesday, May 18,2004 .......... .___
10:22:51 AM
. . .J
Given the following conditions:
- Salem Unit 1 is in MODE 4.
- 21 RHR loop is in service supplying shutdown cooling.
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
ODenina of 2RH21 RHR TO RWST STOP VALVE.
Opening 2SJ69 RHR SUCTION FROM RWST with BOTH 2RH1 and 2RH2 RHR COMMON SUCTION VALVES oDen.
Continued operation of the 21 RHR loop in shutdown cooling when RCS WR Thots rise above 350 dea F.
- i-J [s 1
[Emergency - and - Abnormal Plant Evolutions
!- .... - .. /i ...................................
8:
!E04
~ ............................ i LOCA Outside Containment .
,2.4 I Emergency Procedures / Plan 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation
' /of-system components and controls.
I r4.Oi 4.01 1
'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 I
'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 1 t
'immediate mitigating action. Distracter c is incorrect because we routinely operate in this condition.
INRC Generic letter 98-02 address this loss of coolant outside containment.
I I
I RHROOOE013 For plant or industry events associated with the Residual Heat Removal System a) Summarize each event b) Identify the root cause(s) of the event (N/A NEO) c) Describe the events likelihood of occurrence at Salem Nuclear Generating Station (N/A NEO) d) Describe established or alternative actions which might prevent the events (re)occurrence at Salem Nuclear Generating Station (N/A NEO)
I I
~
Tuesday, May 18,2004 10,2251 AM 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-1, 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 rapid drop in RCS __ pressure. _ _ __ __ - __
A high core Delta T. 1 iA low core Delta T.
'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 iNat Circ settina UD.
/Loss of Heat Sink 1
1- ' thru 5 I I I...... Tuesday, May 18,2004 10:22:51 AM J 1 Page11 of34 1
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.
!Which of the following procedures must be implemented? i El~EOP-FRCE-I, RESPONSE TO EXCESSIVE - - ___ __CONTAINMENT
__ - PRESSURE.- --
EOP-FRCC-I RESPONSE TO INADEQUATE - - CORE - __ COOLING. 1 7EOP-FRCC-2,RESPONSE TO DEGRADED CORE COOLING.
$SaturatedCore Cooling E&. )Ability to determine and-interpret the following as they . . apply __ to.Saturated
- Core Cooling: - -
Is any RCP running (NO-stem), 5 or more CETs > 700 deg (YES-stem) RVLIS > 39% (NO-stem), GO 1
/TO FRCC-1 due to the RED PATH. All the distracters contain the wrong procedural transition.
/CriticalSafety __ Function Status Trees - -__ ___ __ - I I
FRCCOOEOOI I State the Red paths for the core cooling status tree
.. .- ~ .. ......................... .........
I Tuesday, May 18, 2004 10:22:51 AM I I Paae 12of34 I
Given the following conditions on Unit 2:
I
- 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.. .
LOCA-5 because it establishes minimum required containment spray flow and conserves RWST inventorv.
ILOCA-5 since FRPs are NOT implemented during the performance of LOCA-5.
2-EOP-FRCE-1 "RESPONSE TO EXCESSIVE CONTAINMENT PRESSURE" since restoration of the critical safetv function takes mecedence.
2-EOP-FRCE-1 because actions concernins Containment Smav oDeration are more restrictive. 1 lOOWE14A202 RWST level, containment pressure, and # of CFCU's operating. The less restrictive criteria in LOCA-5 is used because recirculation flow to the RCS is not available, and it is very important to conserve RWST water, if possible, by stopping containment spray pumps. So while the operator WILL enter l IFRCE-1 due to PURPLE path of containment pressure > 15 psig, the containment spray pumps will be loperated IAW LOCA-5.
Response to Excessive Containment Pressure Loss of Emergency Coolant Recirculation I !
Describe the basis for each step, caution, and note in 2-EOP-FRCE-1 thru 3 and EOP-CFST-1, Figure 5 Material Required for Examibtion 1L. . .. ...... _...... . ... . . . ...
Question Source: I [FacilityExam Bank Question Modification MeUlod: [Editorially Modified i Question Source Commentr I' Changed stem f;om given PURPLE path to containment pressure > I 5 psig entry point to FRCE-1. i I Tuesday, May 18, 2004 10.2251 AM 1
Given the following conditions:
- Salem Unit 2 has experienced a Large Break Loss of Coolant Accident.
- The Reactor trip and Safety Injection occurred sucessfully.
- 2-EOP-LOCA-1 LOSS OF REACTOR COOLANT is in effect.
- PZR pressure is 35 psig.
,- 1 CET is reading 900 degress, ALL other CET's are reading -550 degrees.
!- RVLIS Full Range is reading 74%.
- Containment pressure is 13 psig.
- Containment sump level is 62%.
- R44A radiation monitor is indicating 50 R/hr.
Which of the following choices identifies a procedural transition that is allowed under these
'conditions?
,FRCI-3, RESPONSE TO VOID IN REACTOR -. - _ _ VESSEL.
FRCC-2, RESPONSETO DEGRADED -___- - CORE
-__________ COOLING,
- -___ - _ _ __ 1 FRCE-1. RESPONSE TO EXCESSIVE CONTAINMENT PRESSURE.
[Emergency and Abnormal-Plant Evolution I
E l6 High Containment Radiation 55.43.(5) A is correct because the YELL0 FRCEOOTOOJ 1 Given a set of plant conditions, take corrective actions for an high containment radiation in accordance w i t h Z E O P - F R C E T 1 I
.. , . ~-. .......,
~~ ~
I Tuesdav. Mav 18.2004 10:22:52 AM I I Paae14of34 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.
IAW 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 Cold Leg Recirculation?
hosing 22SJ45 RHR to Charging/SI Pump Suction Valve.
Closing22CS36 RHR System to CS System Isolation Valve.
'Closing 22RH4 RWST to-RHR Pump Suction Valve.
Closing 22RH19 RHR HX Discharge- Cross-connect
_ - . ... __ - - __ -- - - . Valve. I zl n /Comprehension 1
[005000A204 ,
13 Ability to (a) predict the impacts of the following on the Residual Heat Removal System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
m _ _ _ -- - . mE[
55.43(5) Distracters a & b are incorrect because the SJ45 and CS36 do not require closing to swap over to CL recirc. C is correct because the RHR Dumo must be secured at steo 5.2 of LOCA-3 prior to m .
'closing RH4 and opening SJ44. Distracter d is incorrect because closing the RHI 9 will not require
[Transfer to Cold Leg Recirculation - - __ __ __ -.
I LCA3U2TOO1 I Given a set of Dlant conditions. Derforrn actions for a Transfer to Cold Lea Recirculation in accordance with 2-EOP-LOCA-3 1 1 Tuesdav. Mav 18.2004 10:22:52 AM I I Page15of34 I
Given the following conditions:
- Salem Unit 2 is operating at 100% power.
- An electrical fault has resulted in 2A 125VDC bus becoming deenergized.
- NO actions have been taken in response to the deenergized bus.
- Subsequently, a LOCA occurs, and operators manually initiate a Rx trip and Safety Injection.
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 NOTstart. After at least 1
'minute has elapsed since SI initiation, reset SI, reset 2A SEC, start ECCS pumps from the Control Room that did NOT start automaticallv.
The loss of 2A 125VDC bus will affect indication only. All ECCS equipment will start upon the SEC -Mode - Isignal 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 eneraize-to-actuate feature. and must be manuallv started.
Some ECCS pumps will NOT start, and operators will be unable to start those pumps from the control room. Local oeeration will be reauired to start those eumes.
and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: . -
55.43(5) The loss of the 2A 125VDC bus will not affect the SSPS cabinets. The dc power used in the SSPS logic cabinets is provided from 115VAC vital instrument power. As such the actuation of SSPS
'relays 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 oDerate these bkrs.
IReactor Trip or Safety Injection I I I I ECCSOOEO08 I identify and describe the Control Room controls, indications, and alarms associated with the Emergency Core Cooling System, including a) The Control Room location of Emergency Core Cooling System control bezels and indications (N/A NEO) b) The function of each Emergency Core Cooling System Control Room control and indication (N/A NEO) c) The effect each Emergency Core Cooling System control has upon Emergency Core Cooling System components and operation (N/A NEO) d) The plant conditions or permissives required for Emergency Core Cooling System Control Room controls to perform their intended function e) The setpoints associated with the Emergency Core Cooling System control room alarms
I Tuesday, May 18,2004 10:22 52 AM I Given the following conditions:
II
- 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 entered containment at 1415 to investigate 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 />.
- At 1416 22 CFCU breaker trips.
- I O minutes after the crew entered containment, the NCO reports that 2R12A CONTAINMENT GAS EFFLUENT is reading double what it was when the crew entered containment, and containment average air temperature is rising slowly.
Which of the following choices identifies how control room personnel will respond IAW SC.SA-
!ST.ZZ-0001 SALEM CONTAINMENT ENTRIES IN MODES ITHROUGH 4?
Allow personnel in containment to continue their inspection while monitoring for any continuing 1 rise in radiation level. If radiation levels on 2R12A increase by a factor of 4 from original level, I use -the- page system to direct personnel in containment to _exit.
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 fights, and direct them to exit the containment.
,Since the 2R12A is expected to rise with an RCS leak, the crew may remain in containment until their Heat Stress stav time is comdete.
&2 Ipiizq 611112004
~22000G114 ,
'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 SCSA-ST.ZZ-0001, 3.2.1 The ....I' containment lighting, when flashed, is the preferred method the Control Room uses for requesting communications with the work Party." I
]Salem Containment Entries in Modes 1 through 4 IContainment Entries at Power I I significant prerequisites and precautions associated with each operating procedure which are fequiredto be consider2 by either Licensed or Non-Licensed Operators Tuesday, May 18,
- 2004 10:22:52
- AM - 1 I Page 18of34 I
[ Tuesday, May 18, 2004 1022 Given the following conditions:
- Salem Unit 2 is in MODE 6.
- The Spent Fuel Pool (SFP) Gate Valve is open.
Which of the following choices describes the impact if SFP level drops 8 feet from normal, and what action will be performed?
Reduction in the amount of Iodine scrubbed by the water prior to bubbles reaching the surface, initiate makeup from WHUT . __ to SFP. - - __ _ -
RHR pump suction vortexing leading to possible loss of RHR cooling, close the SFP Gate Valve.__ - .- -
a Loss of Spent Fuel Pool cooling from uncovering suction piping, stop ALL SFP pumps.
- ____- 1 Rising radiation on FHB area radiation monitors 2R5 and 2R9, initiate S2.OP-AB.RHR-0002 LOSS OF RHR AT REDUCED INVENTORY.
those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
p 2 . 0 3 IAbnormal spent fuel pool water level or loss of water level I[ MI[ 3.5J 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 -401'. Distracter b effect will not occur with 15' of water still over the Rx vessel flange, and the closing of the gate valve is a correct action. - . ..---- .
LOSS of Refueling Cavity or Spent Fuel Pool Level I
\Spent Fuel Cooling dwg __ _ -
ABFUELOIEOO Describe, in general terms, the actions taken in S2.OP-AB.FUEL-O001(Q) and the bases for the actions.
2 ABSFOI EO01 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 1 Tuesday, May 18, 2004- 10:22:53 AM
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 INOPERABLE?
The time the plant is exposedto a LOCA event with less than two full trains of ECCS equipment available is limited by the ACTION time associated with each individual structure, svstem or comDonent.
Only ONE Train of ECCS components is necessary to prevent core damage during a DBA LOCA. Requiring both Off-Site power supplies to be OPERABLE whenever an EDG is INOPERABLE will ensure one comdete train oDerates.
The ACTION statements which permit limited variation from the basic requirements are accompanied by additional restrictions which
-- __ -- are more restrictive than the original criteria.
'As long as all required redundant systems and components are OPERABLE, a loss of off-site 1 bower will NOT result in a comdete loss of safetv function of critical svstems.
2.2.25 Knowledge of bases in technical specificationsfor limiting conditions for operations and safety limits. 1 (55.43 (2) Distracter c is the Bases for TSAS 3.1.3 MOVABLE CONTROL ASSEMBLIES, with a change. Distracters a L? b are compendiums of key words and tricky phrases that don't come from the Bases section of TECH SPECS. The correct answer d is found in the Bases section for TSAS 3.8.1 .I act. 82
[Salem Technical Specifications I
- - - . . __ 1 1 1 I
1 Tuesdav. Mav i 8.2004 10:22:53 AM I
'Given the following conditions:
- Salem U'nit 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 HIGH/LOW 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%.
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 2%
by volume since H2 concentration is always 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.
Reduce the oxygen concentration of the in service WGDT to prevent potential releases of radioactive materials due to explosion of the GDT. .. ___ -
Immediately 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.
I -
%c--
J Is- 1 [MemOrY-- Salem1
&2 ]
J Distracter d is incorrect because the immediate suspension of additions to the waste gas system is REQUIRED only when >4% 02. The reason is correct. Distracter b refers to a non-existent Table in 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. Distracter a is incorrect because while the addition of N2 will raise the total volume of the WGDT and lower the 0 2 concentration, the H2 concentration is monitored, and onlv assumed to be > 4% when monitorinq is unavailable.
/GaseousWaste Disposal System Operation I I Salem Tech Specs IWaste Disposal Gas Analyzer Panel 110 Alarm Response I
i WASGASE007
-Identify and describe the local controls, indications, and alarms associated with the Radioactive Waste Gas System, including:
a) The location of Radioactive Waste Gas System local controls and indications.
b) The function of Radioactive Waste Gas System local controls and indications.
c) The plant conditions or permissives required for Radioactive Waste Gas System local controls to perform their intended function.
b) The Bases f& the LCO(s) c) The applicability of the LCO(s) d) The LCO Action Staternent(s)
[ Tuesday, May 18, 2004 10:22:53 AM 1 I Page23of34 I
Given the following:
- Salem Unit 2 is in MODE 6 on day 19 of a refueling outage.
1- Fuel movement is NOT in progress.
- S/G nozzle dams are installed.
- 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:
- RCS temperature is 123 degrees.
I- RHR loop 21 is in service providing decay heat removal.
j- 21,23,24 S/G levels are 75% wide range.
- 22 S/G level is 100% wide range, and pressure is 210 psig.
- 21,23,24 S/G metal temps are all between 85-90 degrees.
- 22 S/G metal temp is reading 68 degrees.
IAW Tech Specs, what action is required based on these conditions, and why?
exceeding fracture toughness stress limits for the S/G.
'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 I pressure induced stresses remain. less than allowable limits.
'Raise 22 S/G 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.
2.1 lLConductof Operations
,2.1.11 [Knowledge of less than one hour technical specification action statements for systems. 3.0[ 3.81 pressure induced stresses in the steam generators do not exceed the maximum fracture toughness
,stress limits." The action time for the TS is 30 minutes. The correct answer of b contains both the laction 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. Red and Purple Paths J. TRIP-1 CASS K. Steamline Isolations L. Feedwater Isolations M. Feedwater Interlock N. Key Relief Valves
- 0. Tank Thumbrules I P. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less Technical Specifications STMGENEOI 0 Blowdown and Drain Systems, including:
a) The Limiting Condition(s) for Operation b) The Bases for the LCO(s) (N/A NEO) c) The applicability of the LCO(s) (NIA NEO) d) The LCO Action Statement(s) (N/A NEO)
IQuestion Modification Method:
1 /Significantly Modified-1 Added why? To question and completely replaced 2 distracters.
I Tuesday, May 18,2004 10:22:53 AM I I Page25of34 I
I Which of the following choices identifies a condition under which the CRS must use the Equipment Out Of Service (EOOS) Risk Assessment Tool IAW SH.OP-AP.ZZ-0027 ON-LINE RISK ASSESSMENT?
Whenever a safety related piece of equipment fails during normal operation.
Each time a risk significant Structure, System or Component (SSC) is removed from service.
Prior to removing Tech Spec related equipment from service for scheduled maintenance.
When a risk sianificant SSC scheduled outaae time is lonaer than one week.
,been included in the weekly PRA. Distracter d is wrong because if it is planned, then it is already
/includedin the next weeks PRA.
I - . .
[On-Line Risk Assessment J i I 1 b) Degraded Condition c) Nonconforming Condition d) Conditions Adverse to Quality (CAQ) e) Justification for Continued Operation (JCO) 9 Single Failure g) Consequential Failure SURVOOE009
~ ._ I WORKOOE008 1 Tuesday, May 18, 2004 10:22:53 AM 1 I Page26of34 I
Given the following:
- 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 E-24 ROD DEV OR SEQ.
- 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.
- Control rod 2D2 indicates 208 steps on IRPI.
- All other Control Bank D rods indicate between 208-215 steps.
I
- IAW Salem Tech Specs, which of the following choices identifies the condition described by the above indications, and what action is required?
- I 2 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 TSAS 3.1.3.2.1 POSITION INDICATING __ -- SYSTEMS - OPERATING. __ ..
Rod 2D2 is > I 2 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.
- d_ J is 1 bPPG?& __ ___ Salem1 t i 2
._- .. J - . 611112004 J
The greatest deviation between Group Demand and Individual Position is 13 steps between Group 2 and rod 2D2 of 13 steps, that is what is causing alarm. However, the IRPl indication of 200 steps IS Iwithin 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 IS 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 GDC deviation is not present. Distractor c is incorrect because AB ROD-1 will not be entered because the deviation doesn't warrant entrv.
]Hot Standbv to Minimum Load RODSOOE006 Describe the function and operating characteristics for the following Rod Control and Position Indication Systems components:
a) Rod Cluster Control Assembly (RCCA) b) Control Rod Drive Mechanism (CRDM) c) Rod Drive MG Sets
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 Steo Counters 1
RODSOOEOl3 State the Technical Specifications associated with the components, parameters, and operation of the Rod Control and Position Indication Systems, including a) The Limiting Condition(s) for Operation b) The Bases for the LCO(s) d) The LCO Action Statements(s) 1 Tuesday, May 18, 2004 10 I Page28of34 I
Given the following conditions:
I- Salem Unit Iis 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.
- 1R41D, Plant Vent Noble Gas Release Rate, has failed high and caused a Containment Ventilation Isolation (CVI) signal.
- 1R I I A , Containment Particulate Monitor, is blocked.
- 1R12A, 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 1R41A, 1R41D AND 1R12A must ALL be OPERABLE.
can be started after blocking both trains of 1R41, and resetting both trains of CVI.
can be started after resetting CVI, since the CVI signal can be reset without clearing the activation signal.
[194001G309 1
1
!an INOPERABLE R41D, it can be blocked and reset IAW step 5.1.1O.C-F, and the purge can kommence. Distractor a is incorrect because the procedure allows reset of CVI by blocking the
,affected rad monitor. Distractor b is incorrect because EITHER R41A,D OR R12A OPERABLE is
,enough to meet ODCM 3.3.3.9 requirements of Table 3.3-13. Distractor d is incorrect because the CVI Containment Purge to Plant Vent WASGASE012 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 Operators I Tuesday, May 18,2004 10:22:54 AM 1 I Page29of34 I
Tuesday, May 18,2004 10:22:54AM 1 1 Page30of34 I 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 Which of the following choices identifies the action that the crew will perform before the other a
!Evacuate non-essential Dersonnel ONLY from containment. I Close the Fuel Transfer Canal Gate Valve.
Initiate Containment Closure IAW S2.OP-AB.CONT-0001. CONTAINMENT CLOSURE.
i2.3 Radiation Control risen >
personnel will remain in containment to mitigate accident. Distractor c is incorrect because the gate valve will be closed later in the procedure a part of the effort to isolate containment in step 3.14.
Distractor d is incorrect because it will be performed later in the procedure at step 3.1 1.A The correct answer b is correct because it is the action which will be performed first of all the actions listed, to minimize personnel exposure.
Fuel Handling Incidents 1I - . . . . . . . . . . . . .
, ............................................................................................................ .-. . . . . . . . . . . . . . . . . . . . . I 1 Tuesday. May 18.2004 10:22:54 AM I I Page31 of34 1
Given the following conditions:
- Salem Unit Iis operating at 100% power.
- OHA G-24 TAC TEMP HI OR LO alarms and clears.
The OHA alarms and clears once every minute.
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.
1
'The RO requests permission to block OHA G-24.
Which of the following choices contains ONLY the actions and procedures that will be performed during the process to block the alarm?
hitiate a Level 2 significance Notification (NOTF), place an INFO sticker on the OHA box, log the notification number in .NC.DE-AP.ZZ-0030, "CONTROL OF TEMPORARY I
MODIFICAT10NS". ~
Initiate a Level 3 significance NOTF, place a single strip of red translucent tape diagonally across OHA box, log the notification number in Attachment Iof SC.OP-DL.ZZ-0010, CONTROL ROOM INSTRUMENTATION AND ALARMS
'Initiate a Level X significance NOTF, create a NU-IND task for an Operator Burden IAW 1NC.WM-AP.ZZ-0000 NOTIFICATION PROCESS, and place an INFO sticker near the OHA window box. I' Initiate a NOTF, place an INFO sticker in the Alarm Response Procedure for that OHA, place 2 strips of red translucent tape diagonally in an "X' across the OHA window box IAW SC.0P-
,DL.ZZ-0010 and SH.OP-AP.ZZ-0030, "OPERATOR BURDEN PROGRAM.
2.4 1 Emergency Procedures / Plan mm28 .1 55.43(5) Distractor A is incorrect because a significance level 2 notification is too high, the info sticker is placed near the OHA box, and the notification number will be logged in Att Iof DL.ZZ-0010. Distractor B is incorrect because for a completely INOPERABLEOHA, 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-0000, Notification Process, Att. 1, Classification Guidance, and also because the info sticker will be
- placed in the ARP, NOT near the OHA box.
Control Room Instrumentation and Alarms Operator Burden Program i
I
- 1. Industrial Safkty Practices
- 2. Radiation Worker Practices
- 3. Conservative Decision Making
- 4. Control Room 6At the Controls Areao
- 5. Communication
- 6. Shift Relief and Turnover I Page32of34 I
- 7. Procedure Use and Adherence
- 8. Alarm Response
- 9. Operator Appearance IO. Self AssessmenUCorrectiveAction
- 11. PlanffControlBoard Awareness and Maintenance of Critical Parameters
- 12. Housekeeping
- 13. Operator Rounds
- 14. Briefs
- 15. Human Error Reduction Techniques
- 16. Log Keeping
- 17. Training
- 18. Supervisor Involvement
- 19. Climbing on Equipment Attachment 1, Shift Briefing Format Attachment 2, Pre-Job Briefing Guidelines Attachment 3, Pre-Job Briefing Checklist Attachment 4, Pre-Job Briefing Checklist (Tagging)
Attachment 5. Human Performance, TORTen Human Error Traps I
Tuesday, May 18,2004 10
An event has occurred resulting in a Site Area Emergency declaration.
Health Physics assistance is required for a task being performed in the Auxiliary Building.
Which one of the following Emergency Response Facilities will dispatch Radiation Control personnel to support the task?
I Emergency Operations Center.
,Operations Support Center.
1I
,2.4.36 /Knowledgeof chemistry I health physics tasks during emergency operations. liz.o12.8/
,55.43.(4) Distractor d is incorrect because Control Room staff does NOT include RP techs. Distractor 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 Perform whatever work is necessary ...
in the plant.
\Shift Radiation Protection Technician Response I i I II HEAPHYEOI1 Describe the radiological responsibilities of the following:
- a. General Manager Hope Creek Operations
- b. Radiation Protection Superintendents
- c. Nuclear Technical Supervisors
- d. Department Managers
- e. Radiation Protection Personnel
- f. Job Supervisors
- g. Individual Workers Question Source: INPO Exam Bank 1 [QuesGon Modification Method: I /Editorially Modified 1.. ............................... ............. - . .. .- .__ . .
I Tuesdav. Mav 18. 2004 10:22:55 AM I r Page34of34 I