ML072960536
| ML072960536 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/31/2007 |
| From: | - No Known Affiliation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 50-324/07-301, 50-325/07-301, ES-401-5 50-324/07-301, 50-325/07-301, NUREG-1021, Rev 9 | |
| Download: ML072960536 (190) | |
Text
Draft Submittal (Pink Paper)
Reactor Operator Written Exam BRUNSWICK JULY-AUG EXAM - 325, 324/20Q7-301 DRAFT RO WRITTEN EXAM WITH ANSWERS
ES-401 Sample Written Examination Question Worksheet Form ES-401-5
--Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
Importance Rating RO
-SRO 2
1 203000 A3.09 3.6 Ability to monitor automatic operations of the RHRlLPCI: INJECTION MODE (PLANT'SPECIFIC) including: Emergency generator load sequencing Proposed Question:
Common 1 Unit Two (2) is in Mode 1.
NO.4 Diesel Generator Monthly Load Test is in progress with OG # 4 loaded to 3000 KW.
A seismic event occurs causing a complete Loss of Off-site Power and a LOCA on Unit Two (2) with the following times (in seconds):
- Time 0 = Seismic event and Loss of Off-site Power
- Time 5 = LOCA signal on Low Reactor water level Which one of the following describes the RHR pump start sequence?
A.
All four RHR pumps start at Time=15 seconds B.
All four RHR pumps start at Time=20 seconds C.
'B' RHR pump starts at Time=15 seconds
'A', 'C', and '0' RHR pumps start at Time=20 seconds..
D.
'A', 'C', and '0' RHR pumps start at Time=15 seconds
'B' RHR pump starts at Time=20 seconds.
Proposed Answer:
C NUREG-1 021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 A.
Incorrect - B will start in 15 seconds. Pumps A,C and D in 20 seconds B.
Incorrect - B will start in 15 seconds.
C.
Correct Response D.
Incorrect - B wi*1I start in 15 seconds. Pumps A,C and D in 20 seconds The STR relays for the RHR pump start must meet the following two conditions before starting their ten second time delay.
- 1. IF power is from the normal source, THEN RHR Loop A(B) pumps start 10 seconds after initiation signal is received.
- 2. IF power is from the emergency diesel generators, THEN RHR Loop A(B) pumps start 10 seconds after the diesel ties to the emergency bus.
The LOOP signal at Time=O gives all DGs a start signal. Since DG 4 was already running, it will not take 10 seconds for the diesel to come up to speed and energize E4.
The LOOP signal will cause the DG4 output breaker to open, mode will shift to Isochronous and the DG4 output breaker will reclose. Since this is effectively instantaneous, bus E4 is energized at Time=O. 10 seconds from the LOCA signal is Time 15 seconds for the start of RHR pump B.
RHR pumps A, C, and 0 will start at Time 20. 10 seconds for the Diesels to start and energize their respective E bus and then 10 seconds for the STR relays to time out.
Per 20P-17, Step 5.2 Automatic Startup for LPCI Mode 5.2.1 Initial Conditions
- 1. RHR has received an automatic initiation signal from the following:
- a. Reactor low level three OR
- b. Both of the following:
- High drywell pressure AND
- Low RPV pressure 5.2.2 Procedural Steps
- 1. WHEN automatic initiation occurs, THEN OBSERVE the following:
- a. LPGIINITIATION SIGNAURESET indicating light is on.
- b. IF power is from the normal source, THEN RHR Loop A(B) pumps start 10 seconds after initiation signal is received.
- c. IF power is from the emergency diesel generators, THEN RHR
.Loop A(B) pumps start 10 seconds after the diesel ties to the emergency bus.
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet 20P-17, Section 5.2 (Attach if not previously provided)
"Automatic Startup for LPCI Mode", Steps 1.b and 1c Proposed references to be provided to applicants during examination:
NONE Learning Objective:
____________ (As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam LOR-CLS-LP-017-20-#2 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination-Outline Cross-reference:
Level Tier #
Group #
KIA #
1 205000 K2.02 2.5 Knowledge of electrical power supplies to the following: Motor operated valves Proposed Question:
Common 2 Unit 1 is shutdown. Power has been lost to Panel 1-XDA. This affects the---
A~
Normal power supply to RHR Valve - F008 B.*
A$SDpower supply toRHRValve ~F008 C.
Normal power supply to RHR Valve - F009 D.
ASS'D power supply to RHR Valve - F009 Proposed Answer:
B NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Explanation (Optional): The ASSD supply to F008 is 1-XDA A.
Incorrect - Normal Supply to F008 is 1(2)-XDB/B50 B.
Correct Response C.
Incorrect - Normal Supply to F009 is 1(2)XAlDH3 D.
Incorrect - ASSD supply to F009 is 1(2)XD/DX5 50-17, Table 17.4 F008 RHR SID Cooling Suction Isolation Valve-Outboard20" Anchor Darling Flex Wedge Gate Valve with SB-3 Actuator 1(2)-A71-CSS1 0 Normal:
1(2)-XDB/B50 Alt. A55D Feed:
1(2)-1 XDA/B26 F009 RHR Shutdown Cooling Suction Isolation Valve-Inboard 20"Anchor Darling Split Wedge Gate Valve with SB2-80 operator A71-S9 Normal:
1(2)XAlDH3 Alt. ASSD Feed:
1(2)XD/DX5 Technical Reference(s):
SD-17 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
Question Source:
Question History:
NUREG-1021, Revision g (As available)
Bank#
Modified Bank #
(Note changes or attach parent)
New X
Last NRC Exam
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5
Exa-minatio-n-Gutli-ne-Cross-reference:-- -
--level---
- -----------R-O------------------SR-Q Tier #
2----
Group #
1----
KIA #
205000 A4.05 Importance Rating 3.2 Ability to manually operate and/or monitor in the control room: Minimum flow valves Proposed Question:
Common 3 lAW procedure1(2)OP-17, when RHR is in the Shutdown Cooling mode of operation, RHR Minimum Flow Bypass Valve E11-F007A(B) is A.
expected to exhibit some leakage past its valve seat.
B.
expected to exhibit some leakage through its valve disk.
C.
expected to exhibit some leakage along its valve disk and past its valve seat.
D.
not expected to exhibit any leakage.
Proposed Answer:
B Explanation (Optional):
A.
Incorrect - designed leakage is through valve disk only B.
Correct Response C.
Incorrect - designed leakage is through valve disk only D.
Incorrect - designed leakage is through valve disk.
Per 50-17 3.4 Minimum Flow Bypass Control (Figure 17-13)
The minimum flow bypass valves power supplies are found in Table 17-4.
Containment-s~g-~_.-9J§9s for 1(2)-E11-F007A(B) have been drilled with a vent hole to prevent thermally induced pressure-locking when required to open.
Per 10P-17, Step 5.7.2.12 NOTE: E11-F007A(B) has a hole drilled in its valve disc. During SDC operations, leakage from 2 to 5 gpm can pass through the disc to the Torus when E11-F007A(B) is closed.
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet SD-17, 10P-17 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
____________ (As available)
Question Source:
Question..History:
Bank#
Modified Bank #
New Last NRC.Exam..
x (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
Importance Rating RO
--BRO 2
1 206000 K4.18 3.2 Knowledge of HIGH PRESSURE COOLANT INJECTION SYSTEM design feature(s) and/or interlocks which provide for the following: Pump minimum flow: BWR-2,3,4 Proposed Question:
Common 4 The HPCI pump is being placed in service for quarterly testing. When the HPCI pump
_(1)_, the HPCI Minimum Flow Valve E41-F012 will auto close when discharge flow reaches _(2)_,
A.
(1) Turbine Steam Supply Valve is fully open (2) 400 gpm B.
(1) Turbine Steam Supply Valve is fully open (2) 800 gpm C.
(1) reaches 120 psig discharge pressure (2) 400 gpm D.
(1) reaches 120 psig discharge pressure (2) 800 gpm Proposed Answer:
0 NUREG-1021, Revision 9
ES-401' Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 A'.
Incorrect - Flow must be 800 gpm. Pump running is based on discharge pressure.
B.
Incorrect - Pump running is based on discharge pressure.
C.
Incorrect - Flow must be 800 gpm.
O.
Correct response Per 50-19, Section 3.3.3
- 3. Minimum Flow Control (Figures 19-16 and 17)
As mentioned earlier, under no-flow or low-flow conditions, the HPCI Pump is protected from overheating by maintaining a minimum flow through the pump by diversion of water to the Suppression Pool. This diversion of flow is automatically controlled ~y the Minimum Flow Bypass To Suppression Pool Valve, E41-F012. E41-F012 is normally closed whenthesyste'mis i'n'astandby lineup.
If the HPCI pump is running and system flow is less than 400 gpm, the minimum flow valve will automatically open. A pressure switch, located at the HPCI Pump di,scharge, closes when the discharge pressure reaches 120 psig to indicate that the pump is running. When system flow reaches 800 gpm, the minimum flow valve will automatically close.
In addition, this valve will automatically close if either the Turbine Stop Valve or the Turbine Steam Supply Valve is fully closed.
SO-19 Technical Reference(s):
(Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
Question Source:
____________ (As available)
Bank#
Modified Bank #
(Note changes or attach parent)
New X
10 CFR Part 55 Content:
55.41 Last NRC Exam'
'Question History:
Question Cognitive Level:
Memory or Fundamenta'i Knowledge Comprehension or Analysis NUREG-1021, Revision 9 X
ES-401 Comments:.
NUREG-1021, Revision 9 Sample Written Examination Question Worksheet 55.43 Form ES-401-5
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 217000 A1.05 3.7 Ability to predict and/or monitor changes in parameters associated with operating the RCIC system controls including: RCIC Turbine Speed Proposed Question:
Common 5 Unit Two (2) is in Mode 1.
The RCIC System Operability Test has just been completed and RCIC has been shutdown. The Reactor Operator reports that Turbine Steam Supply Valve, E51-F045, has no indication. '
Investigation reveals that E51-F045 is CLOSED but the FULL CLOSED limit switch has failed to actuate.
What is the impact on the RCIC system if an automatic initiation were to occur prior to any repairs being performed.
The RCIC system will:
A.
start and operate normally.
B.
NOT automatically initiate and cannot be manually started.
C.
automatically initiate and then trip on mechanical overspeed.
D.
NOT automatically initiate but will operate normally if manually started.
Proposed Answer:
C NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 A.
Incorrect - RCIC will trip on overspeed B. Incorrect - RCIC will auto initiate, the Turbine Steam Supply Valve, E51-F045 position indication has no impact on the auto start.
C. Correct response D. Incorrect - RCIC will auto initiate and trip on overspeed.
Per 50-16, Step 3.6.3 Ramp Generator and Signal Convertor (RGSC) Box The Ramp Generator for RCIC is reset to idle speed by closing the E51-F045. If the Ramp Generator is not reset to idle, it will NOT assume control of RCIC Turbine speed when E51-F045 is opened on the initiation and RCIC begins to accelerate. With no limit on acceleration, the RCIC Turbine will trip on mechanical overspeed.
The ramp generator and signal convertor box (RGSC) provides the speed reference voltage to the "speed reference section" of the EGM and consists of the following three subsystem circuits:
D The Low Signal Selector This circuit is continuously sensing, and it selects the lowest (least positive polarity) signal output from either the "ramp generator" or the "signal convertor" circuits, and transmits that low signal to the of the EGM.
D The Ramp Generator
, The ramp generator circuit outputs a continuous "idle" (low voltage) signal of approximately 1600 rpm until the "ramp" function is initiated.
The "ramp" function is initiated by closing of contacts on the Turbine Steam Supply Valve, E'S1-F045 when the valve leaves the full closed position. The time for this ramp increase from the idle speed setting to the rated high speed setpoint iS,approximately 22 seconds. On each occasion that the ramp circuit is de-energized by closing E51-F045, the circuit is automatically resetto the "idle" output signal.
Technical Reference(s):
SO RCIC system (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
NUREG-1021, Revision 9 Bank#
LOR-CLS-
ES-401 Question History:
Sample. Written Examination
.Que'stion Worksheet LP-016-15-#2 Modified Bank #
New Last NRC Exam Form ES-401-5 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis
_X 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample 'Written Examination Question Worksheet Form ES-401-5
--Examination-Outline Cross-reference:
Level Tier #
Group #
KIA #
1 209001 K1.09 3.2 Knowledge of the physical connections and/or cause-effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following: Nuclear boiler instrumentation Proposed Question:
Common 6 A feedwater line rupture has occurred on Unit Two and the following conditions exist:
Drywell pressure Reactor water level N036/N037 Reference leg temp RPV pressure 8.8 psig
- 35 inches 225 0 F 500 psig What is the configuration of both loops of Core Spray?
A.
Shutdown with 2E21-FO~1A(B), and 2E21-F004A(B) open and 2E21-F005A(B) is closed.
B.
Running with 2E21-F004A(B) and 2E21-F005A(B) injection valves open and flow to the vessel.
C.
Running with 2E21-F031A(B), and 2E21-F004A(B) open and 2E21-F005A(B) is closed.
D.
Running with 2E21-F004A(B) and 2E21-F005A(B) injection valves open but no flow to the vessel Proposed Answer:
C NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 A.
Incorrect - Core Spray will be running due to Hi DW pressure signal.
B.
Incorrect - No injection flow because F005 valves are closed and pressure is > pump shutoff head.
C.
Correct Response - the F031A(B) will be open for min flow protection of the pump, F004A(B) are normally open and the FO.05A(B) are closed because the low pressure permissive of 410 psighas not been met.
D.
Incorrect - FOD5 valves are closed. Rx pressure is > pump shutoff head Per 50-18, Step 3.1.3 Core Spray Inboard and Outboard Injection Valves The Core Spray Inboard and Outboard Injection Valves, E21-F005B(A) and E21-F004B(A), respectively, have automatic and manual control functions. Normally, the valves are in the automatic mode as dictated by the Control Switches, E21-S1 BfA and S2BfA (CLOSE-AUTO-OPEN, spring return to AUTO), for each valve.
Both valves will automatically open provided the following conditions are met:
o An initiation signal present (Low reactor water level #3 or low reactor vessel pressure coincident with a high drywell pressure).
(K10) o Low reactor pressure permissive satisfied (K20) o 10 second start timer relay timed out (timer starts once the E-bus is energized) (K3fK4)
If the inboard (E21-F005BfA) valve control switch is turned to CLOSE while an initiation signal is present, the valve automatic opening function will be disabled. The manually initiated CLOSE signal will
. override the automatic OPEN signal thus allowing the valve to be throttled as the Operator desires. The valve's automatic opening function will remain disabled until the system initiation signal is cleared (initiation logic is reset, reactor pressure increases above 410 psig, or power is lost to the associated Emergency Bus). A white light will illuminate "CLOSE SIG SEALED VLV E21-F005" on P601 which indicates that the automatic opening is disabled.
Technical Reference(s):
OP-18, 1.5.5.2 /R19 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: ---------
Learning Objective:
NUREG-1021, Revision 9 (As available)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier#
Group #
KIA #
1 209001 A4.09 3.6 Ability to manually operate and/or monitor in the control room: Suppression Pool level Proposed Question:
Common 7 During a Main Steam line break outside the drywell the suppression pool level had lowered below -31 inches.
Which ONE of the following systems and lineups is used by direction of EOP-02-PCCP, Primary Containment Control* Procedure, to restore suppression pool level?
A.
Residual Heat Removal by unlocking and opening the A(B) RHR Keepfill Station Bypass Valve only.
B.
C.
Core Spray by unlocking and slowly opening the Core Spray Pump 1A(1 B) suction valve from the CST only.
Residual Heat Removal by unlocking and opening the A(B) RHR Keepfill Station Bypass Valve and slowly opening the Torus Discharge Isolation Valve F028A(B)
D.
Core Spray by unlocking and slowly opening the Core Spray Pump 1A(1 B) suction valve from the CST, starting a Core Spray pump and opening the Full Flow Test Bypass Valve F015A(B).
Proposed Answer:
B NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Question Source:
Question History:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge X '
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1 021, Revision 9
.Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 PCCP directs makeup to the torus using Core Spray from CST per OP-18, this line up uses only one valve CORE SPRAY PUMP 1A(1B) SUCTION VALVE FROM THE CONDENSATE STORAGE TANK, E21-F002A(B).
a.
is incorrect because two other valves are required to be open the torus cooling isolation valve and the suppression pool cooling valve.
c.
is incorrect because two other valves are required to be open the torus cooling isolation valve and the suppression pool cooling valve.
d.
is incorrect because there is no need to start the pump there are no check valves on the CS pump suctions.
From OP-18.0 8.4 Filling the Suppression Pool with Core Spray In Standby 8.4.1 Initial Conditions
- 1. Core Spray Loop A(B) is in Standby in accordance with Section 5.1.
NOTE: A 1 foot change in CST level is a change of approximately 15,900 gallons..
- 2. CST level is sufficient to provide the desired volume to the Suppression Pool AND to maintain CST level above 12 ft.
8.4.2 Procedural Steps Loop A(B)
- 1. STATION a second operator to specifically monitor Suppression Poo'l Level.
- 2. DIRECT AO to perform the following:
- a. ESTABLISH continuous communications with Control Room Personnel monitoring the Suppression Pool level and CST level.
NOTE: Overfilling the Suppression Pool will cause level to be above the Tech Spec limit of -27 to -31 in.
CAUTION IF CRD pump suction is from the CST, THEN CST level should be maintained above 11 ft. to prevent CRD pumps from tripping on low suction pressure.
- b. UNLOCK AND SLOWLY THROTTLE OPEN CORE SPRAY PUMP 1A(1B) SUCTION VALVE FROM THE CONDENSATE STORAGE TANK, E21-F002A(B).
- 3. MONITOR Suppression Poo*1 level AND CST Level.
- 4. WHEN desired level in Suppression Pool has*been reached, THEN CLOSE AND LOCK CORE SPRAY PUMP 1A(1B) SUCTION VALVE FROM THE CONDENSATE STORAGE TANK, E21-F002A(B).
- 5. ENSURE CST level is above 12 ft.
- 6. IF CST level drops below 12 ft, THEN GO TO 00P-31.2 to refill CST, AND PERFORM CONCURRENTLY with this procedure.
- 7. ENSURE Suppression Pool level is within Tech Spec limits of -27 to -31 in.
NUREG-1 021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 211000 A2.04 3.1 Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Inadequate system flow Proposed Question:
Common 8 During an ATWS on Unit One (1), the SLC control switch is placed in the Pump B Run position with MCC 1XG de-energized. SLC indications are:
Pump A red light OUT Pump B red light LIT Squib valve A light OUT Squib valve B light LIT SLC tank level 53%
SLC discharge pressure 1450 psig.
RWCU Inboard Isolation Valve, G31-F001 RWCU Outboard Isolation Valve, G31-F004 What operator action is required?
A.
Perform alternate boron injection.
B.
Monitor for lowering SLC tank level.
C.
Close the inboard RWCU isolation valve.
D.
Place the SLC control switch in Pump A Run.
Proposed Answer:
A NUREG-1021, Revision 9 OPEN CLOSED
ES-401 Explanation.(Optional):
Sample Written Examination Question Worksheet Form ES-401-5 A.
Correct Response B.
Incorrect - With no flow, tank level will not be lowering,
C.
Incorrect - Inboard valve didn't need to close because squib valves did not fire.
O.
Incorrect - No power is available to SLCpump The B squib valve did NOT fire (light is lit). The A squib valve has no power and will not fire due to MGG 1XG being de-energized. The B SLG pump is recirculating through the relief valve. Therefore, there is no flow and per EOPs alternate injection is required.
SO-05 Technical Reference(s):
(Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Q,uestion History:
Bank#
Modified Bank #
New Last NRC Exam LOR-CLS-LP-005-A*006-#1 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier#
Group #
KIA #
1 212000 K6.05
';t 3.5 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR PROTECTION SYSTEM: RPS Sensor Inputs Proposed Question:
Common 9 Following a refueling outage Unit Two (2) is conducting a plant startup. The reactor mode switch has just been placed to RUN with reactor power at 10%. All MSIVs are open, however the limit switch, 821-F022D-LS-3, on MSIV 2821-220 is currently seen as CLOSED by RPS.
Which of the following MSIVs, if closed, will cause a RPS scram actuation from MSIV valve position directly?
A.
2821-F022A and 2821-F028D 8.'
2821-F028~and 2821-F022A C.
2821-F028D and 2821-F022C D.
2821-F022C and 2821-F028D Proposed Answer:
B Explanation (Optional):
Isolating any combination of three Main Steam Lines will cause a Reactor Scram.
From SD-03:
As with the Turbine Stop Valves, two Steam Lines may be isolated (S 10% closed) without causing.a reactor scram.
- Isolating one steam line does not cause a half-Scram.
- Isolating Main Steam Lines A and D or 8 and C does not cause a half-Scram.
- Isolating any other combination of two Main Steam Lines will cause a half-Scram.
- Isolating any combination of three Main Steam Lines will cause a Reactor Scram.
a.
is incorrect because this combination only isolates two steam lines (A &D) c.
is incorrect because this combination only isolates two steam lines (C & D) d.
is incorrect because this combination only isolates two steam lines (C & D)
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet SO-03 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
CLS-LP-03, Obj. 9. Given any (As available) scram signal, describe the logic arrangement for the signal including what combination of signals will cause a full scram.
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam x
(Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 212000 A2.09 3.6 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal cpnditions or operations: High containmenUdrywell pressure.
Proposed Question:
Common 10 Unit Two (2) has inserted a manual reactor scram due to a loss of RBCCW. Several control rods failed to insert on the scram. Plant conditions:
APRM indicated power 3%
Reactor pressure 960 psig, controlled by EHC Drywell pressure 2.1 psig Mode Switch Shutdown SDV Hi Hi Wtr Trip Bypass
- Normal, Which ONE of the following is required to reset RPS?
A.
Install jumpers per LEP-02, Section 3, Reset ARI and Verify the SDV Vent and Drain Valves are OPEN.
B.
Install jumpers per LEP-02, Section 3, Inhibit ARI and Verify the SDV Vent and Drain Valves are CLOSED.
C.
Manually insertcontrol rods using the RMCS until A-06 Annunciator APRM DOWNSCALE clears.
D.
Place the SDV Hi Hi Wtr Trip Bypass Keylock Switcb to BYPASS then wait until A-O~
Annunciator SDV HI-HI LEVEL RPS TRIP clears.
Proposed Answer:
B NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 LP-02 Section 3 - The purpose of this section is to insert control rods by repeated manual scram, overriding RPS if required.
CO: _
- 1. Unit 1 Only: ENSURE the REACTOR MODE SWITCH, C71-S1, is in "SHUTDOWN."
CO: _
- 2. Unit 2 Only: IF steam flow is less than 3 X 106 Ib/hr, THEN ENSURE the REACTOR MODE SWITCH, C72-S1, is in "SHUTDOWN".
NOTE Steps 3 and 4 may be performed concurrently.
- 3. IF an automatic scram signal is present AND power is available to the RPS bus, THEN INSTALL the following jumpers to bypass the reactor scram:
CO: _
- a. Jumper 15 in Panel H12-P609, Terminal Board DD, from the right side of Fuse C71A(C72A)-F14A to Terminal 4 of Relay C71A(C72A)-K12E.
CO: _
- b. Jumper 16 in Panel H12-P609, Terminal Board BB, from the left side of Fuse C71A(C72A)-F14C to Terminal 4 of Relay C71A(C72A)-K12G.
CO: _
- c. Jumper 17 in Panel H12-P61*1, Terminal Board DD, from the right side of Fuse C71A(C72A)-F14B to Terminal 4 of Relay C71A(C72A)-K12F.
CO: _
- d. Jumper 18 in Panel H12-P611, Terminal Board BB, from the left side of Fuse C71A(C72A)-F14D to Terminal 4 of Relay C71A(C72A)-K12H.
- 4. INHIBIT ARI by performing the following steps:
CO: _
- a. PLACE ARI AUTO/MANUAL INITIATION switch, C11(C12)-CS-5560, to "INOP".
CO: _" b. PLACE ARI RESET switch (spring return), C11 (C12)-CS-5562, to "RESET" and MAINTAIN for a minimum of five (5) seconds, THEN RELEASE.
- CO: _
- c. VERIFY the red "TRIP" light located above ARI INITIATION, C11 (C12)-CS-5561 is off.
CO:
- 5. ENSURE the DISCH VOL VENT & DRAIN TEST switch is in "ISOLATE".
- 6. VERIFY the following valves are closed:
CO: _
- a. DISCH VOL VENT VLV C11(C12)-V139 CO: _
- b. DISCH VOL VENT VLV C11(C12)-CV-F010 CO: _
- c. DISCH VOL DRAIN VLV C11 (C12)-V140 CO: _
- d. DISCH VOL DRAIN VLV C11(C12)-CV-F011 CO:
- 7. RESET RPS.
a.
c.
d.
is incorrect because ARI must be inhibited or scram air pressure will not recover to close the Scram valves, also the SDV vent and Drain valves must be closed.
is incorrect because clearing the APRM downscale annunciator will have no effect on the RPS system, the RPS trip in SHUTDOWN is set at 15°J'c>.
is incorrect because bypassing the SDV and draining the SDV will not allow RPS to be reset with a high drywell pressure scram in.
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet LEP-02, Section 3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
none Learning Objective:
- 6. Given plant conditions, the (As available)
Local Emergency Procedures, and which steps have been completed, determine required operator actions. (EOP-01-LEP-01,02,03)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam x
(Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 215003 K6.05 3.1 Knowledge of the effect that a loss or malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM)
SYSTEM: Trip Units Proposed Question:
Common 11 The 'F' IRM mode switch has inadvertently been placed in the "standby" position.
Which ONE of the following describes the affect this will have on the IRM?
A.
The.input signal from output amplifier is removed and an "INOP" trip will be generated.
B.
The 12V input from the range switch is removed which deselects all ranges and, in turn, causes no input to be sent to attenuator.
C.
It provides an "INOP" trip to yield maximum design protection before the channel is removed from service. '
D.
It removes the input signal that is sent to attenuator.
Proposed Answer:
C NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Per S009.1 Sample Written Examination Question Worksheet Form ES-401-5 3.3.2 The IRM controls located on backpanel drawer P606 are described below (Figure 09.1-12)
Each channel is provided with the following:
Mode Switch Changes mode of IRM operation to allow for maintenance or calibration. Each switch has six positions.
OPERATE IRM channel functions as described in previous sections.
, STANDBY Same as "operate", except that it gives Inop trip to yield maximum design protection before channel is removed from service.
ZERO 1 Removes signal from output ampHfier so that output amplifier, local meter, and recorder can be zeroed ZERO 2 Removes 12V from range switch which deselects all ranges and, in turn, causes no input to be sent to attenuator. This allows setting the zero adjust on the output amplifier DC input circuits 125 Removes input to be sent to attenuator and generates a calibration signal to yield 125 on the black scale of the IRM channel meter for setting the gain on the output amplifier.
40 Same a*s "125" position except that it produces 40 on the black scale A.
Incorrect - for ZERO 1 position B.
Incorrect - for ZERO 2 position O.
Incorrect - for 125 position Technical Reference(s):
NUREG-1021, Revision 9 SO-09.1, section 3.3.2 (Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be provided to applicants during examination:
NONE Learning Objective:
____________ (As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam x
(Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written E~amination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 215004 K5.01 2.6 Knowledge of the operational implications of the following concepts as they apply to SOURCE RANGE MONITOR (SRM) SYSTEM:
Detector operation Proposed Question.:
Common 12 The Source Range Monitor (SRM) detectors are fission chambers that have an applied voltage to an electrode. The applied voltage to the SRM detector is A.
higher than the applied voltage used for the IRM detector and the SRM electrode generates an electrical signal inversely proportional to neutron flux in the core.
B.
lower than the applied voltag~ used for the IRM detector and the SRM electrode generates an electrical signal inversely proportional to neutron flux in the core.
C.
higher than the' applied voltage used for the IRM detector and the SRM electrode generates an electrical signal proportional to neutron flux in the core.
D.
lower than the applied voltage used for the IRM detector and the SRM electrode generates an electrical signal proportional to neutron flux in the core.
Proposed Answer:
C NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Per SO 09.1 section 2.1 Sample Written Examination Question Worksheet Form ES-401-5 2.1 Neutron Detectors (SRM & IRM)
The neutron detectors are fission chambers which generate an electrical signal proportional to the neutron flux level in the core.
There are 4 SRM detectors, one per core quadrant, and 8 IRM detectors, located in gaps between fuel bundles on an. opposite corner from the control rod (wide-wide gap) (Figure 09.1-2).
The SRM (IRM) detector is a fission chamber that has an applied voltage to the electrode of approximately 600 (100) volts. The operating chamber is pressurized. with Argon to about 213 (17) psia. This configuration of voltage and Argon gas pressure produces very little secondary ions; therefore, the chamber operates as an ion chamber. This is true for both the SRM and IRM detectors.
A.
Incorrect - the signal is not inversely proportional.
B.
Incorrect - SRM voltage is higher and the signal is not inversely proportional.
O.
Incorrect - SRM voltage is higher Technical Referenc~(s):
SO 09.1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
Question Source:
Question History:
(As available)
Bank#
Modified Bank #
(Note changes or attach parent)
New X
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 NUREG-1 021, Revision 9
ES-401 Comments:
NUREG-1021, Revision 9 Sample Written Examination Question Worksheet Form ES-401-5
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
2 215005 2.4.10 3.0 Emergency Plan/Procedures: Knowledge of annunciator response procedures Proposed Question:
Common 13 With Unit Two (2) at 96% power the following are received:
A-06 6-7, FLOW REF OFF NORMAL A-06 3-8, APRM UPSCALE TRIP / INOP A-OS 2-2, ROD OUT BLOCK Affected APRM ODA shows FLOW (47°A>>
Affected RBM ODA shows FLOW COMPARE alarm The cause of the conditions above is that the affected APRM:
A.
B.
has a-eritical self test fault.
has a non-critical self test fault.
C.
has a recirc flow transmitter input failed to zero.
D.
Aas a recirc flow transmitter input failed upscale.
Proposed Answer:
C NUREG-1021, Revision 9
ES-4D1 Explanation (Optional):
Per A-06 3-7 and A-06 3-8:
Sample Written Examination Question Worksheet Form ES-401-5 C.
Correct - A flow transmitter failure to zero does not cause a critical or non-critical fault. It will in this case cause a rod block and upscale alarm because the power level is above the flow-biased trip set point that results from the reduction in flow from 94% to 47%. It should be recognized that the FLOW(%) is much lower than that necessary to achieve 96°/b power. If the alarm is due to the comparator (greater than or equal to 10o/b) then FLOW COMPARE is displayed as indicated in the question.
A. & B.
Incorrect - A critical self test fault causes both alarms. A non-critical self-test fault only causes the trouble alarm. Because of the absence of the trouble alarm the self test faults can be eliminated as the cause.
D.
Incorrect - If the recirc flow transmitter failed upscale, the FLOW(%) would be at least 110%. A rod block would result from a total" recirc flwo upscale condition, however, this rod block is a result of the lower flow-biased trip set points.
Technical Reference(s):
APP A-D6 3-7 and 3-8 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Question History:
Bank #
Modified Bank #
New Last NRC Exam LOR-CLS-LP-009.6*141-
- 5 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Unit One (1) is operating at full power. The CRO is performing 101-03.1, CONTROL OPERATOR DAILY SURVEILLANCE REPORT.
A value of -23" is recorded for item no. 114, RCIC Steam Li~e Flow.
AllACf"IME NT' 1 Paige 59 :of 6,3 rr~M Si~r CHECK tls'r~
- NOTEiS,
<:PER IF-REO, r.~iE NO.
~c fSJO'~£~
Ui~TS
'113R,ecOiID; HPC,11i1.Pl m'l fJawhVl E4:tJPiO'~~~...,t 'fe~~
,-g llsb!~J:3,,3;If.t..1!t.m 30"
$,,,3,,6~1. ~
Q 11.1$
~~O!RD ROIC: :s~ar:n f~flowbigh ES'1..partd--iN'O'11..1 'tectln~
Sp~~tu1 T.lble3.3.al",~ :rr.m,j~,~
SR, 3,.3.1..'1.,1
'11.5 REC:OiRO, HPCR s,tf:ern ~!e.~h~
6i1,'1,.p;O~NOO5*~,
- 'f.e~~
~.)tl T$~e 3,3.6t 1-~t~m 3a" SR,3,;3.,6..1.1 o
t,,2,,3 Q
'116*
RE;CO'iRD: ReIC: ~.gn ~~ftow'tlgh E5'~..p~NO'18-l Tedhn~
Sprd"1C8'1lorr! 'iahe,3.3.6..1*1 t~m Q..
6R 3~~.tt't~
BRU NS~tt.CK S1TIEAjMB..iEC'fRICPilANT iDAJlY :SUtRV8LJLANiOEREiPORT CONTR'Cl. OPERATOR,S Q.
IF the instrument indication is below the specified limits, DECLARE the instrument inoperable. IF the instrument indication is above the specified limits, CONTACT the System Engineer.
Which one of the following actions, if any, is required?
A.
Notify the System Engineer.
B.
No actions required, this value is within the limits.
C.
Determine if a Technical Specifications LCO exists.
D.
Initiate actions to perform OOP-10.1.1 RCIC Operability Test.
Proposed Answer:
C*
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 2170002.1.33 3.4 Conduct of Operations: Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications Proposed Question:
NUREG-1021, Revision 9 Common 14
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Explanation (Optional):
The value of -23" is below and outside the allowable range of -22" to +22" therefore the instrument is inoperable. This is a Technical Specifications required instrument check. The next step is to check Technical Specifications.
a.
is incorrect because this reading is below the acceptable instrument range.
b.
is incorrect because the value is below the limits.
d.
is incorrect because there is no requirement to immediately run a RCIC operability test.
Technical Reference(s):
101-03.1, TS 3.3.5 (Attach if not previously provided)
Proposed references to be provided to applicants during-examination:
NONE Learning Objective:
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam x
(Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision-9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 218000 K4.02 3.8 Knowledge of AUTOMATIC DEPRESSURIZATION SYSTEM design feature(s) and/or interlocks which provide for the following:
'Allows manual initiation of ADS logic Proposed Question:
Common 15 Unit Two (2) is in Mode 1 when a LOCA occurs resulting in the following plant conditions:
44 inches and decreasing
, o'perating unloaded tripped on overcurrent failed to auto initiate operating in the minimum flow condition tripped on overcurrent in AUTO in alarm in ~Iarm in alarm clear' clear Which one of the following describes the status of the Automatic Depressurization System (ADS)?
A.
actuate automatically in'~ely-83seconds.
B.
actuate automatically after reactor level decreases to -4.5 inches.
C.
not-al!~'Fflatically actuate--and the SRVs must be ma-nually opened.
D.
n'ot automatically aetl:J*ate,u-At-il additional ECCS pumps are started.
NUREG-1021, Revision 9
ES-401 Proposed Answer:
C Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 ADS requires One Core Spray pump or two RHR pumps in one loop to be running before it will initiate. The status if this interlock is indicated via the CS or RHR pumps running alarm.
Starting additional ECCS pumps will have no effect.
ADS must also see LL3 and a confirmatory LL1 alarm. These conditions have been met so the timer should have started their 83 second countdown and annunciated the timerstartalarm and ADS relays energized alarm. Since these alarms have not annunciated, ADS will not auto initiate.
A. and B. - Incorrect - ADS will not auto initiate.
D.
Incorrect - Starting additional pumps will have no effect.
Technical Reference(s):
SD-20 permissive tables Annunciator Response:
CORE SPRAY OR RHR PUMPS RUNNING (A-03 2-1)
REACTOR ADS LO WATER LEVEL (A-03 4-2)
REACTOR LOW WTR LEVEL INITIATION (A-03 6-9)
AUTO DEPRESS TIMERS INITIATED (A-03 5-1)
AUTO DEPRESS RELAYS ENERGIZED (A-03 3-2),
(Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
NUREG-1021, Revision 9 Bank#
Modified Bank #
New LOR-CLS-LP-020-11 (Note changes or attach parent)
ES-401 Question History:
.Sample Written Examination Question Worksheet Last NRC Exam Form ES-401-5 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form' ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 218000 2.4.49 4.0 Emergency Procedures / Plan Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
Proposed Question:
Common 16 Unit 1 is operating at 100°A> power when the control room receives the following alarms..
SAFETY/RELIEF VALVE OPEN (A-03 1-10)
SAFETY OR DEPRESS VLV LEAKING (A-03 1-1)
Additional the operators note that a steam flow/feed flow mismatch with feed flow greater than steam flow and a decrease in generator power.
The operators immediately cycle the control switch of the affected safety/relief valve to OPEN and CLOSE OR OPEN and AUTO several times. If the SRV fails to close, A.
the operators must manually trip the turbine a'nd scram the reactor.
B.
when suppression pool temperature reaches 110 degrees F., the operators must manually scram the reactor.
\\
C.
the operators must immediately scram the reactor.
D.
when suppression pool temperature reaches 110 degrees F., the operators musU manually trip the turbine then scram the reactor.
Proposed Answer:
B Explanation (Optional): 3.0 OPERATOR ACTIONS - OAOP-30.0 Rev. 12 Page 4 of 9 Step 3. IF it is determined that the affected safety/relief valve can NOT be closed, THEN IMMEDIATELY PERFORM the following:
- INSERT a manual reactor SCRAM A.
Incorrect - manual turbine trip not required.
C.
Incorrect.- suppression pool temperature must reach 110 degrees prior to scram.
D.
Incorrect - manual turbine trip not required.
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet OAOP-30.0 Rev. 12 Page 4 of 9 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam x
(Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge X---
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comment's:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 223002 K4.05 2.9 Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) and/or interlocks which provide for the following: Single failures will not impair the function ability of the system.
Proposed Question:
Common 17 RPS MG Set A has tripped. RPS Distribution Panel A has NOT yet been transferred to its alternate source.
The LL3 instrument providing input to PCIS Channel 82 fails downscale.
Which of the following describes the response ofMSIVs and Steam Line Drains?
A.
Only the Inboard Steam Line Drain valve and all MSIVs close.
8.
Only the Outboard Steam Line Drain valve and all MSIVs close.
C.
Inboard and Outboard Steam Line Drain valves and all MSIVs close.
D.
Inboard and Outboard Steam Line Drain valves close, and all MSIVs remain open.
Proposed Answer:
C Explanation (Optional):
Channel B2 tripped would give a Group 1 logic BID tripped, loss of RPS A would remove power from Group 1 logic AlCand result in a full MSIV isolation. A2 (Loss of RPS) and B2 closes outboard steam line drain. Loss of A logic power from RPS A will close the Inboard steam line drain.
See Figure 25.7 in SD-25 Technical Reference(s):
NUREG-1021, Revision 9 SD-025 (Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be provided to applicants during examination:
NONE Learning Objective:
____________ (As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam LOI-CLS-LP-025*08A-#3 (Note changes or attach pa'rent)
Question Cognitive Level:
Memory or Fundamental Knowledge X---
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 239002 K1.07 3.6 Knowledge of the physical connections aQd/or cause-effect relationships between RELIEF/SAFETY VALVES and the following:
Suppression pool Proposed Question:
Common 18 When steam is disch"arged from the safety relief valves (SRV) to the suppression pool water, which ONE of the following describes the flow path taken by the SRy discharge and the reason for the path?
A.
The SRV discharges to a tailpipe which terminates in a manifold (T-Quencher) approximately seven feet above the bottom of the suppression pool. This provides even heat distribution in the suppression pool, but does not significantly reduce the dynamic forces.
B.
The SRV discharges to a tailpipe which terminates in a manifold (T-Quencher) approximately seven feet below normal suppression pool water level.
This provides even heat distribution in the suppression pool and reduced dynamic forces on the suppression chamber.
C.
The SRV discharges to a tailpipe which connects to a ring header, the ring header terminates in a manifold (T-Quencher) approximately seven feet below normal suppression pool water level.
This provides even heat distribution in the suppression pool, but does not significantly reduce the dynamic forces.
D.
The SRV discharges to a tailpipe which connects to a rin'g header, the ring header terminates in a manifold (T-Quencher) approximately seven feet ~_Qqye the bottom of the suppression pool.1 This provides even heat distribution in the suppression pool and reduced dynamic forces on the suppression chamber.
Proposed Answer:
B NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Per SD-04, Section 2.2 Sample Written Examination Question Worksheet Form ES-401-5 The steam discharged from the safety relief valves (SRV) is condensed by the suppression pool water. The SRV tailpipes are routed to the suppression pool through the vent pipes. The SRV tailpipe continues downward and terminates in a manifold (T-Quencher) approximately seven feet below normal water level. Each T-Quencher is approximately twenty feet long and has holes along both sides of its length. The sparging effect of the T-Quencher provides even heat distribution in the suppression pool and reduced dynamic for~es on the suppression chamber upon SRV actuation.
The T-Quenchers are restrained (along the length) by gussets fixed to the Quencher Support Structure; the Quencher Support Structure is mounted to the bottom of the suppression 9hamber.
A.
Incorrect - the tailpipe terminates 7 feet below, not above water level, to reduce dynamic forces C,D.
Incorrect - the tailpipe does not terminates in a manffold;-At>ta ring header, below the water level to significantly reduce dynamic forces Technical Reference(s):
SO-04 Rev. 4 Section 2.2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective:
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam 239002A108-
- 1 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 NUREG-1021, Revision 9 Sample Written E*xamination Question Worksheet Form ES-401-5
Examination Outline Cross-reference:
259002 K4.06 ES-401 Sample Written Examination Question Worksheet Level Tier #
Group #
KIA #
Importance Rating RO 2
1 3.1 Form ES-401-5 SRO Knowledge of REACTOR WATER LEVEL CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following:
Control Signal Failure Proposed Question:
Common 19 Unit Two (2) is in Mode 1 at 100% rated thermal power when the Reactor Operator notes the following plant conditions:
- Digital Feedwater Level Control System
- Feedwater Control System Trouble (A-7 4-2)
- AlB Level Selector Switch transferred to 1-Element Control annunciated.
selected to 'A' Which ONE of the following in*aications resulted in the transfer to 1-Element Control?
'B' Feed Flow indicator has. failed upscale.
B.
'e' Main Steam Line Flow indicates 5% lower.
C.
'A' Narrow Range Level indicator has failed downscale.
D.
'A' Reactor Pressure transmitter fails upscale.
Proposed Answer:
A NUREG-1021, Revision 9
ES-401 Explanation (Optional):
SD 32.2 Section 4.2.4 Sample Written Examination Question Worksheet Form ES-401-5 4.2.4 Losos of Any Feed Flow Input There are 2 normal feed flow inputs that are, summed to provide an output signal to the following and dependent upon initial power level and severity of failure the following may occur:
DAuto transfer to 1 element operation resulting from real alarm block criteria being exceeded OR individual feed flow not within 10% of average feed flow OR total feed flow now < 20%.
D Recirc pump runback if total feedwater flow goes < 16.4%
D Hydrogen Water Chemistry injection solenoids trip if total feed flow < 17.3°A>.
D Hydrogen Water Chemistry may trip on external setpoint step change (>5 SCFM)
°4.2.6 Loss of Reactor Pressure Input For a failed pressure transmitter, the system will use the non-failed pressure transmitter regardless of switch position.
B.
is wrong since a difference of >1 O°A> is required between steam flow indicator to cause a transfer to single element control.
C. is wrong since a level transmitter failure will not impact 3-element control.
D. is wrong since DFCS would look at the non failed channel.
Technical Reference(s):
SO-32.2 APP A-0704-2, FW eTl SYS TROUBLE (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE learning Objective:
(As available)
Question Source:
NUREG-1021, Revision 9 Bank#
Modified Bank #
New x
(Note changes or attach parent)
ES-401 Question History:
Sample Written Examination Question Worksheet Last NRC Exam Form ES-401-5 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 261000 K1.09 3.2 Knowledge of the physical connections and/or cause-effect relationships between SGTS and the following: PCIS Proposed Question:
NUREG-1 021, Revision 9 Common 20
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Unit Two (2) is in Mode 1 when a small LOCA begins inside the Drywell. Plant conditions are:
Assumi,ng no equipment malfunctions are present, which one of the following describes the status of the Standby Gas Treatment System to these conditions?
A.
Both SBGT trains should have automatically started.
B.
Neither,SBGT train should have automatically started.
c.
Only the 'A' SBGT train should have automatically started.
D.
Only the '8' SBGT train should have automatically started.
NUREG-1021, Revision 9
ES~401 Proposed Answer:
A Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 Per 20P-01 0, Standby Gas Treatment System Operating Procedure SBGT will start on a high Drywell pressure signal (1.7 psig) alone and does not require a LOCA initiation signal. The A SBGT isolation damper will automatically open on the initiation and does not impact the auto start of SBGT.
B, C, D.
Incorrect - both trains start on the High DW Press signal Technical Reference(s):
20P-010, SBGT (Attach if not previously provided).
Proposed references to be provided to applicants during examination: _N_O_N_E Learning Objective:
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exa-m LOR-CLS-LP-010 #1 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1 021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 262001 A3.03 3.4 Ability to monitor automatic operations of the A.C. ELECTRICAL DISTRIBUTION including: Load shedding Proposed Question:
Common 21
'Unit 2 has a load shed scheme which automatically trips selected loads on either a Generator Lockout or a LOCA.
Which ONE of the following describes loads and their associated signals included in the Full Time Load Shed scheme for Unit 2 during normal full power operation?
A.
1 Turbine Building Chiller (on LOcA Signal) 1 Running Circ Water Intake Pump (on LOCA Signal) 2 Heater Drain Pumps (on LOCA Signal)
B.
1 Turbine Building Chiller(on LOCA Signal) 1 Running Circ Water Intake Pump (on LOCA Signal) 2 Heater Drain Pumps (on Generator Lockout Signal)'
C.
2 Turbine Building Chillers (on LOCA Signal)
,1 Running CircWater Intake Pump (on LOCA Signal) 2 Heater Drain Pumps (on Generator Lockout Signal)
D.
1 Turbine Building Chiller (on Generator Lockout Signal) 2 Running Circ Water Intake Pumps (on LOCA Signal) 2 Heater Drain Pumps (on Generator Lockout Signal)
Proposed Answer:
B NUREG-1021, Revision 9
ES-401 Explanation.(Optional):
Sample Written Examination Question Worksheet Form ES-401-5 A.
Incorrect - Htr Drain Pumps affected by Generator Lockout C.
Incorrect - Only one TB chiller is affected D.
Incorrect - TB Chiller is affect by LOCA and only one Circ Water Intake pump is affected Per 80-50.1, Section 4.1.3 Both Brunswick Units have a load shed scheme which automatically trips selected auxiliary loads on either a Generator Lockout or a LOCA.This scheme lowers the minimum required switchyard voltage necessary for proper'operation of plant emergency auxiliaries. For some loads, the scheme is ena'bled on a FULL TIME basis. For other loads, Brunswick Plant Operations will enable ~he scheme on a PART TIME basis per plant procedures, when requested by the System Operator.
Where necessary to maintain LOCA voltage support, the System Operator will request the Brunswick Plant Operator to ENABLE the PART TIME Load Shed Scheme. When no longer required to maintain LOCA voltage support, the System Operator will notify the Brunswick Plant Operator that PART TIME Load Shed may be DISABLED.
The Full Time loads for each unit inctude:
- 1 Turbine Building C~iller (on LOCA Signal)
- 1 Running Circ Water Intake Pump (on LOCA Signal)
- 2 Heater Drain,Pumps (on Generator Lockout Signal)
Technical Reference(s):
80-50.1, Section 4.1.3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
Question Source:
Question History:
NUREG-1021, Revision 9 (As available)
Bank#
Modified Bank #
(Note changes or attach parent)
New X
Last NRC Exam
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension.or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1 021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 262002 K4.01 3.1 Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C.lD.C.) design feature(s) and/or interlocks which provide for the following: Transfer from preferred power to alternate power supplies Proposed Question:
Common 22 On the vital UPS, the rectifier section supplies regulated DC power from an AC source to the inverter section.
In a normal alignment if _(1)_, the UPS will transfer to its _(2)_ supply.
A.
(1) the inverter output is lost
.(2) 120 VDC B.
(1) the rectifier output is lost (2) 120 VDC (1) the inverter output is lost C.
(2) 250 VDC D.
(1) the rectifier output is lost (2) 250 VDC Proposed Answer:
D NUREG-1 021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 Technical Reference(s):
SO-52, section 2.2.1 A.
Incorrect - the transfer is affect by the rectifier voltage B.
Incorrect - the transfer will be to a 250 VOC supply C.
Incorrect - The transfer is affected by rectifier voltage SO-52, section 2.2.1 The rectifier section supplies regulated DC power from an AC source to the inverter section. Three-phase 480 VAC power is connected'through the input circuit breaker to a power transformer. This circuit breaker provides overload and fault protection as well as a means for manually disconnecting the rectifier. The power transformer converts the 480 VAC input to an appropriate voltage level for the rectifier. The rectifier uses solid state components to convert the incoming AC voltage to a regulated DC voltage which is higher than the maximum voltage of the alternate DC source, approximately 280 VDC.
An alternate source of DC voltage from the 250 VDC system is supplied in parallel with the rectifier through a blocking diode. Normally the blocking diode will be back-biased by the output of the rectifier preventing the DC source from supplying the inverter and the rectifier from supplying the DC source. In the event the rectifier output is lost due to a rectifier failure or loss of the AC' input, the bias will be removed from the blocking diode permitting the DC source to supply the input to the inverter.
(Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Question Cognitive Level:
Memory or Fundamental Knowledge Learning Objective:
Question Source:
Question History:
(As available)
Bank#
Modified Bank #
(Note changes or attach parent)
New X
Last NRC Exam NUREG-1021, Revision 9
ES~401 Sample Written Examination Question Worksheet Comprehension or Analysis x
Form ES-401-5 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 263000 K2.01 3.1 Knowledge of electrical power supplies to the following: Major D.C. loads Proposed Question:
Common 23 Unit One (1) is operating at rated power when Division II DC Switchboard is lost.
What immediate impact will this power loss,have on the Reactor?
A.
Inboard MSIVs close, SRVs are available to control RPV pressure using their normal power source.
B.
Outboard MSIVs close, SRVs are available to control RPV pressure using their normal power source.
C.
Inboard MSIVs close, SRVs are available to control RPV pressure using their alternate power source.
D.
Outboard MSIVs close, SRVs are available to control RPV pressure using their alternate power source.
Proposed Answer:
D NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Explanation (Optional):
SO-51, section 4.3.6.5, SO 25, section 2.5 Table 25-3 A.
Incorrect - the SRVs lost their normal power source B.
Incorrect - the SRVs lost their normal power source C.
Incorrect - Inboard MSIVs affected by Division I Automatic Depressurization System (ADS)
ADS system logic is designed such that an entire division of DC power can be lost and ADS will still function. ADS logic A is powered from DivisioryjDpanels 3B(4B)for units 1(2), and will not initiate without power. ADS logic. B is normally powered from~ivision II DC panels 3B(4B) with an-auto trans~to 3A(4A). The level initiation relays for the B logic are powered from 3A(4A) only, therefore ADS logic B will not initiate on a loss of Division I power. As long as EITHER ADS logic initiates, ADS will perform its intended function.
The SRVs are normany powered from DC panel 3B(4B) and auto transfer to 3A(4A) on loss of power. An alarm will be received on the P601 panel on loss of DC power to any portion of ADS logic' or to any SRV.
Technical Reference(s):
SO-51, section 4.3.6.5 O-ASSD-01 SD-25, section 2.5 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis Question Source:
Question History:
Bank #
Modified Bank #
New Last NRC Exam LOI-CLS-LP-051.0*08E-(Note changes or attach parent) x NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
Examination Outline Cross-reference:
ES-401 Sample Written Examination Question Worksheet Level Tier #
Group #
KIA #
Importance Rating Form ES-401-5 RO SRO 2
2 264000 K1.07 3.9 Knowledge of the physical connections and/or cause-effect relationships between EMERGENCY GENERATORS (DIESEUJET) and the following: Emergency core cooling systems Proposed Question:
Common 24 During accident.;<conditiqns, the following sequence of events occurs:
Th~ Core Spray Pumps will auto start at:
A.
T =14 seconds.
T =0 seconds T =4 seconds T =5 seconds T =14 seconds T = 1'6\\ seconds Drywell pressure rises above -the scram setpoint Off-site power is lost Reactor pressure drops below. the low pressure initiation setpoint for ECCS Diesel Generators energize their respective E Buses Reactor level drops below LL3 B.
T =24 seconds C.
T =29 seconds.
D.
T =31 seconds.
.Proposed Answer:
C NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 Sequential timing relays (15 seconds) will not initiate until a LOCA signal is present AND the associated E bus has power.
SD-18 Rev. 3 Page 24 of 53 3.2.2 Response to Core Spray Initiation Signal (Figures 18-7 and 18-8)
.Satisfying a system initiation signal from the above logic will result in the following:
D The Emergency Diesel Generators (all four) automatically start immediately upon the receipt of a Core Spray System initiation signal.
D The Core Spray pumps automatically start 15 seconds from receipt of the initiation signal if the Emergency Buses are energized (off-site power available). If off-site power is not available, the pumps automatically start 15 seconds from the time the Emergency Diesel Generators re-energize the buses.
1 Technical Reference(s):
SD-18 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam CLS-LP-18*09B-#4 x
(Note changes or attach parent)
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES~401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
1 300000 K3.02 3.3 Knowledge of the effect that a loss or malfunction of the Instrument Air System will have on the following: Systems having pneumatic valves or controls.
Proposed Question:
Common 25 Unit 2 is in Mode 3 following a seismic event and plant scram.
Unit 2 conditions are:
- Reactor level
- Reactor pressure
- Drywell. pressure
- Division I PNS header pressure
- Division II PNS header pressure
+55 inches 500 psig 9 psig 93 psig 98 psig Which one of the following describes the status of the Backup Nitrogen Rack Isolation Valves?
A.
Div I Nitrogen Backup Isolation Valves, RNA-SV-5482 is open..
Div II Nitrogen Backup Isolation Valves, RNA-SV-5481 is open B.
Div I Nitrogen Backup Isolation Valves,. RNA-SV-5482 is open Div II Nitrogen Backup Isolation Valves, RNA-SV-5481 is closed
- c..
Div I Nitrogen Backup Isolation Valves, RNA-SV-5482 is closed Div II N.itrogen Backup Isolation Valves, RNA-SV-5481 i.s open D.
Div I Nitrogen Backup Isolation Valves, RNA-SV-5482 is closed Div II Nitrogen Backup Isolation Valves, RNA-SV-5481 is closed Proposed Answer:
B NUREG-1 021, Re\\(ision 9
ES-401 Explanation (Optional):
Sample-Written Examination Question Worksheet Form ES-401-5 No LOCA signal is present so the Backup N2 valves will not be open on a Core Spray initiation signal. The Backup N2valves open at 95 psig or lower in the PNS header.
This would result in Division I Backup N2 valve being open and Division II closed.
2.0 AUTOMATIC ACTIONS OAOP-20.0 Rev. 30 Page 4 of ~ 8 2.4 IF Division I (II) non-interruptible instrument air header pressure decreases to 95 psig, THEN the following will occur:
Division I
- RB INSTR AIR RECEIVER 1A(2A) PRESS LOW (UA-01 1-1) alarms
- DIV I BACKUP N2 RACK ISOL VLV, RNA-SV-5482, opens to align the respective backup nitrogen bank
- RB Stby Air Camp, PI-3785, indicates less than or equal to 95 psig
- Reactor Building Standby Air Compressors start.
Division II
- RB INSTR AIR RECEIVER 1B(2B) PRESS LOW (UA-01 1-2) alarms
- DIV II BACKUP N2 RACK ISOL VLV, RNA-SV-5481, opens to align the respective backup nitrogen bank
- RB Stby Air Camp, PI-3786, indicates less than or equal to 95 psig
- Reactor Building Standby Air Compressors start.
2.5 IF Division 1(11) PNS header pressure decreases below 95 psig, THEN the following occur:
- DIV I BACKUP N2 RACK ISOL VLV, RNA-SV-5482, opens to align the respective backup nitrogen bank
- DIV II BACKUP N2 RACK ISOL VLV, RNA-SV-5481, opens to align the respective backup nitrogen bank A.
Incorrect - Div II setpoint not reached <<95psig), No LOCA signal reached C.
Incorrect - Div I setpoint reached <<95 psig)
D.
Incorrect _ Div II setpoint not reached <<95 psig)
Technical Reference(s):
AOP-20, Pneumatic System (Attach if not previously provided)
Failures NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be provided to applicants during examination:
Learning Objective:
____________ (As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam LOR-CLS-LP-046-07-#1 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1 021, Revision 9
-Examination Outline Cross-reference:
400000 A3~01 ES-401 Sample Written Examination Question Worksheet Level Tier #
Group #
KIA #
Importance Rating RO 2
1 3.0 Form ES-401-5
- SRO, Ability to monitor automatic operations of the CCWS including: Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the CCWS Proposed Question:
Common 26 Unit 2 is in Mod'e 1 when a Loss of Coolant Accident occurs resulting in the following plant conditions:
- Reactor level
-30 inches and rising rapidly
-'-*'---.'*-.--~R~e*,a,G*tof_pressu+e~------I-1O.O-psig-andJowering-slnw1¥
- Drywell pressure 28 psig and rising slowly Assuming all systems respond as designed, which one of the following describes the effect of these conditions on the Unit 2 Service Water System? '
TBCCW Heat Exchanger supply valves, 2-SW-V3 and 2-SW-V4, will close.
B.
Normal NSW supply valves for Diesel Generators'3 and 4, 2-SW-V681 and 2-SW-V682, will close-and the Unit One (1) supply valves will open.
C.
CSW Isolation Valve to the Vital Header, 2-SW-V111, will open unless NSW Isolation,Valve to the Vital Header, 2-SW-V117, is' open.
D.
RBCCW Heat Exchanger supply valves, 2-SW-V103 and 2-SW-V106, will close.
Proposed Answer:
0 NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 RBCCW valves close on low pressure after a time delay not on a LOCA signal.
NSW OG supply valves will not swap to the alternate unit unless header pressure is low.
The vital header supply valve from CSW does not automatically open.
The RBCCW isolation valves will close on a LOCA initiation signal.
Per SO-43
- 2. RBCCW Heat Exchanger Supply Valves Under normal plant operating conditions, the only flow o,n the NSW header is through the RBCCW heat exchanger supply Valves, SW-V106 and SW-Y103, to the R~QCW h~at Exchangers.
SW-V106 and SW-V103 each have a three position CLOSE/NEUT/OPEN spring return to neutral switch on Panel XU-2.
Operation of SW-V1 06 and SW-V103 is either 100 percent open or closed. When placed in OPEN, the valves travel full open.
Upon receipt of a LOCA or LOOP signal, Valves SW-V106 and SW-V103 will automatically close to ensure sufficient service water supply to the Emergency Diesel Generator System, the vita,l header equipment (i.e., RHR pump sea,l heat exchangers, RHR pump room coolers, and core spray pump room coolers), and the RHRSW cooling load. Valves SW-V106 and SW-V10,3 may be opened after
,'- the LOCA and/or LOOP initiation signals have been reset. On a LOCA signal the,close signal to V103 and V106 may be over-ridden by use of an over-ride switch located in the Electronic Equipment Room (Panels 1(2)-XU-24(30) for 1(2)-SW-V1 03, 1(2)-XU-7(29) for 1(2)-SW-V1 06). This will allow restoration of cooling water to the RBCCW heat exchangers during a LOCA event. LOOP closure signal is over-ridden by wire lift only during Station Blackout per AOP-36.2.
Technical Reference(s):
Core Spray System 1(2)
Actuated alarms: A-01 2-6 and A-03 2-6 NUREG-1021, Revision 9 (Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam LOR-CLS-LP-043-16-#1 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
2 201002 A4.02 3.5 Ability to manually operate and/or monitor in the control room: Emergency in/notch override switch Proposed Question:
Common 27 During the approach to criticality, the operator is withdrawing rods using single notch withdrawal.
The operator initiates a notch withdrawal of rod 18-19 from 02 to 04 and notes faster than normal rod speed. The Withdraw light above the Rod Movement Control switch is still lit when rod p()sTfion--04 comes on*,lnengoes out.
How can the operator prevent the rod from double notching?
A.
Place Rod Select Power to Off.
B.
Place the Timer Test Switch to the Reset position.
C.
Place the Rod Movement Control Switch to Insert.
D.
Place the Emergency Rod In Notch Override switch to Emergency In.
Proposed Answer:
0 NUREG-1021, Revision 9
ES-401 Sample Written Examination
. Question Worksheet Form ES-401-5 Explanation (Optional): Placing the Emergency Rod In Notch Override switch to "Emergency In" will interrupt power to the timer and directly energize the drive in bus.
A. Incorrect - because rod selected and driving signal bypasses Rod Select switch.
B. Incorrect - this is used to reset a timer malfunction.
C.
Incorrect - because the insert select relays are locked out with the timer running on a withdraw signal.
SO-7 Section 3.1.2 Emergency rod insertion is activated by holding the Rod Out Notch Override Switch in the EMERGENCY IN position. Switch contacts remove power to the solid state timer and associated foUower relays an-d-a-PlJties-~ro'weT-dire-ctlyio-th-e-IJ-Brive-=in-"-b~Becattse power is removed from the solid state timer, emergency insert will interrupt any operations by that circuitry which might be in progress and allow the selected control rod to be immediately inserted.
Also, because the timer is de-activated, no settle operation will take place when the Rod Out Notch Override Switch is returned to the neutral position. A "rod selected and driving signal" is generated while the switch is in EMERGENCY IN to bypass the drift alarm.
However since there is no settle function, the rod drift indications will be received as soon as the switch is released.
Technical Reference(s):
SD-7 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Question History:
NUREG-1 021, Revision 9 Bank#
Modified Bank #
New Last NRC Exam LOI-CLS-LP-007-A*07E-#2 (Note changes or attach parent)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
2 201003 K5.01 2.6 Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD DRIVE AND MECHANISM:
Hydraulics Proposed Question:
Common 28 The reactor is at rated power.
Control Rod individual scram time testing is being performed per PT-14.2.1. -When Control Rod 10-11 i's scrammed,*the b~1I check internal to the mechanism insert port fails to reposition. This failure. will:
A.
prevent the rod from fully inserting.
B.
cause a slower than normal scram time.
(
C.
result in a faster than normal scram time.
D.
have no effect on scram time at high reactor pressures.
Proposed Answer:
B Explanation (Optional):
A.
Incorrect - the rod will insert on reactor pressure C.
Incorrect - will result in slower scram time without accumulator o.
Incorrect - will result in slower scram time without accumulator Per SO-8, Section 2.11.3 Without the accumulator pressure assist, the scram time increases significantly at reactor pressures less than 800 psig and at lower pressures the control rod may not scram at all. "The
'scram time with the accumulator only (vessel ports plugged) increases as reactor pressure increases and may not fully insert at reactor pressures above 1000 psig. See OSO-08.1 for additional information on,scram response characteristics of CROMs.
Technical Reference(s):
NUREG-1021, Revision 9 8D-8 (Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5
-Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam LOI-CLS-LP-008-A*08B-#2 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part '55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5
. Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
Importance Rating RO
-SRO 2
2 201006 K6.01 :
2.8 Knowledge of the effect that a loss or malfunction of the following will have on the ROD WORTH MINIMIZER SYSTEM (RWM)
(PLANT SPECIFIC) : RWM power supply:- P-Spec(Not-BWR6)
~
Proposed 'Question:
Common 29 A loss of UPS 120V panel V-1 OA has occurred. The _(1)_ power supply to _(2)_ has been lost.
Proposed Answer:
A Explanation (Optional): 2V-10A is the primary power supply to the RWM and RPIS on Unit 2 B. Incorrect -RMCS primary power supply is from 2--V8A C. Incorrect - RWM and RPIS alternate power is from is from 2-V8A D. Incorrect - RPIS alternate power is from is from 2-V8A Technical Reference(s):
SD-07.1, step 3.1.2 BN-52.0.01 (Attach if not previously proVided)
Proposed references to be provided to applicants during examination:
None Learning Objective:
NUREG-1 021, Revision 9
____________ (As available)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam x
(Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
Gf--=.ommerns-:---
NUREG-1021, Revision 9 55.41 55.43
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
2 202001 K3.06 3.7 Knowledge of the effect that a loss or malfunction of the RECIRCULATION SYSTEM will have on following: Low pressure coolant injection logic: Plant-Specific Proposed Question:
Common 30 A valid LPGI initiation has occurred and cannotp~~.reset at this time. The Recirculation System input signal from the LPGI Initiation Logic Division I has failed high.
Actual Reactor Pressure is 300 psig and lowering.
How is the Recirculation system affected?
A.
ONLY Recirculation pump Discharge Valve, B32-FO~11? and Recirculation Pump Discharge Bypass V~lve, B32-F032S will have auto closed.
)
B.,
Recirculation Pump Discharge Valves, B32-F031A and B, and Recirculation Pump Discnarge Bypass Valves, B32-F032A and B will have auto closed.
G.
ONLY Recirculation Pump Discharge Valve, B32-F031A and Recirculation Pump Discharge Bypass Valve, B32-F032A will have auto closed..
D.
Recirculation Pump Discharge Valves, B32-F031A and Band Reeirqulatiorr-Pttmp Di~charge Bypass Valves, B32-F032B will still be op~n Proposed Answer:
B NUREG-1 021, Revision 9
ES-401 Explanation (Optional):*
Sample Written Examination Question Worksheet Form ES-401-5 A
Incorrect - Both pump valves are affected C.
Incorrect - Both pump valves are affected D.
Incorrect - Valves will close Per SD 17.0 section 3.2.2 To aid in maintaining adequate core cooling, the following Reactor Recirculation System valves
. receive automatic close signals IF reactor pressure is reduced to less than 310 psig, and are prevented from being reopened until the LPCI Initiation Logic is reset:
Either Division I or II logic will close these valves.
- Recirculation Pump Discharge Valve, B32-F031A(B)
- Recirculation Pump Discharge Bypass Valve, B32-F032A(B)
Technical Reference(s):
SD 17.0 section 3.2.2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam x
(Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge X----
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
Importance Rating RO
-SRO 2
2 202002 K4.02 3.0 Knowledge of RECIRCULATION FLOW CONTROL SYSTEM design feature(s) and/or. interlocks which provide for the following:
Recirculation pump speed control: Plant-Specific Proposed.Question:'
Common 31 Unit One (1) is, (jper~ting with the following conditions:
RFP 1B Recirc*Pump Speeds Reactor Power RFP 1A 55%
.....__.._....._....__..._._.....~E.~~at.~~..__,.__. _
idling 58%
I
.,RFP 1A trips and R'EACTOR LEVEL HilLa alarms. Reactor Level drops to-the scram setpo'int and continues to lower to +1 ~ 0 inches before the operator brings the idling RFP on line (30 seconds after the trip of RFP 1A) to restore Reactor Level.
What is the current status of the Recirc Pumps?
A.
Running at 58% speed.
B.
Running on limiter #2 C.
Running on limiter#1.
D.
.Tripped on ATWS ARI/RPT.
Proposed Answer:'
C NUREG-1.021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 A, B,D.
Incorrect -
Both Recirc Pumps should be operating on the #1 limiter (28°A>> due to total FW Flow <16.4°A>, LL2 setpoint changed from 118" to 105" on U1 due to power uprate plant mod.
SD-32.2 Section 3.1.3 The PCTPWR block uses the inputs from total steam flow and total feed flow signal to establish the following:
o Input to 3 element permissive to DFCS (> 20%
feed fl'ow) o HWC injection permissive (>17.3% feed) o Input to SF, FF and RPV Pressure low out of limit real alarm block
(> 10% feed flow) o Permlsslves on ~~~~1~~rTlno~~~~e~ts-a~t~~~~~~~
19.1 %)
o Permissives on RWM LPAP (29.80/0 total steam flow, resets at 27.8°1'<>>
o Feed Flow/Steam Flow mismatch to LEVELERR o Recirc limiter #1 (34%) if total feed flow is <16.4°1'<>
SO-32.2 Technical Reference(s):
(Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective:
____________ (As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam X-6281 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge
-Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 NUREG-1021, Revision 9
ES-401 Comments:
NUREG-1021, Revision 9 Sample Written Examination Question Worksheet Form ES-401-5
ES-401 Sample Written Examination Question Worksheet Form ES~401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
2 204000 K1.05 2.7 Knowledge of th~ operational implications of the following concepts as they apply to RWCU: Flow Controllers Proposed Question:
Common 32 Unit Two is operating under the following conditions:
2A RWCU pump is in service.
A RWCU filter is operating at 100gpm.
B RWCU filter is operating at 105 gpm.
--..-..--..-------f\\JRl1X Quflellemp: 11-g-F------------------
Reject flow: Ogpm A complete loss of air to the RWCU system occurs.
How will the 2A and 28 RWCU pumps and the demins be affected?
A.
The 2A RWCU pump will trip due to loss of cooling water. The 28 RWCU pump will automatically start when system flow reaches 60gpni:~th*e-*demin-filters-will shutdown.
B.
The 2A RWCU pump will trip due to loss of cooling water. The 2B--RWCU-*pumJ3 will not automatically start, demin filters_will J-0*-into HOLD.
C.
The 2A RWCU pump will trip due to low system flow. T-he 28 RWCU pump wiH~n-ot automatically start,the demin filters shatctown.
D.
The 2A RWCU pump will trip due to low system flow. The 28 RWCU pump will automatically start, the demin filters will go into HOLD.
Proposed Answer:
C NUREG-1 021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 A.
Incorrect - pump trip caused by low flow B.
Incorrect - pump trip caused by low flow, demins shutdown o.
Incorrect - demins shutdown Loss of air causes the running RWCU pump to trip on low system flow due to the F066, filter discharge valve, failing shut. The filters shutdown, not go into hold, due to the loss of air to the hold pump discharge valve which fails shut and does 'not ?lIow hold pump operation.
Per 50-14, Section '2.5.2 Air Operated Valves FlO Flow Control Valve G31~Z002-66A(B)(Figure 14-4) controls flow through individual filter/demineralizer units. The controller is located locally on the Main Programmer Control Panel. The 66A (B) valves are
:ddesig-n-e-d-to-fail-close'd-o'n'-'a-ioss-of-i-n-sfftlm-ent-a+r.
Technical Reference(s):
SO-14 Proposed references to be provided to applicants during examination:
Learning Objective:
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam LOI-CLS-LP-o14-A-09H-#2 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
2 241 000 A1.14 3.4 Ability to predict and/or monitor changes in parameters associated with operating the REACTORITURBINE PRESSURE REGULATING SYSTEM controls including: Pressure setpointlpressure demand Proposed Question:
Common 33 Unit Two (2) is in Mode 1 performing a plant startup with the following conditions:
2 bypass valves are open Max combined flow dial is adjusted to 110%
IbeaJ-Hm,it-a~a-l-is-,a.djuste.d-to-1-0D%
Medium speed acceleration rate selected 1800 speed select depressed Turbine speed is 900 rpm.
Speed Increasing light illuminated Pressure Regulator A is in control Pressure Regulator B In standby* (3 psig bias)
Pressure Set 945 psig If the pressure set DECREASE pushbutton is depressed and held depressed, the Turbine Control Valves wiU'
?
A.
remain throttled to regulate turbine*~speed. The Bypass Valves will close as necessary to maintain reactor pressure at -942 psig.
B.
remain at the current throttled position. The Bypass Valves will open further as necessary to maintain the lowering pressure setpoint.
C.
open as limited by the speed limiter. The Bypass 'Valves will close slightly and maintain reactor pressure approximately 3 psi higher.
D.
clo$.e after the Bypass Valvesh,Cl'v',eclosed in response to the lowering pressure setpoint.
Proposed Answer:
8 NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 A.
Incorrect - the bypass valves open, not close to maintain pressure C.
Incorrect - The TCVs remain at their current position. The bypass valves open o.
Incorrect - The TCVs remain at their current position Per SO 26.2 section 4.1.3.2 Pressure setpoint will give an open signal to the TCVs and BPVs, but the TCV position will be limited by the LVG from the speed control network.
During the reactor startup as reactor pressure is increased general procedures direct the increase of the pressure circuit set point. This in combination with the reduced load limit circuit setpoints cause the bypa~s valves to be sent an open signal as pressure in the steam lines exceeds the pressure control pressure setpoint.
Proposed references to be provided to applicants during examination:
NONE SO 26.2 - section 4.1.3.2 Technical Reference(s):
Learning Objective:
(Attach if not previously provided)
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam LOI-CLS-LP-26.3*10G-#2 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1 021, Revision 9
ES-401 Sample Written Examination Q*uestion Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
2 202001 A1.09 3.3 Ability to predict and/or monitor changes in parameters associated with operating the RECIRCULATION SYSTEM controls including: Recirc pump seal pressures Proposed Question:
Common 34 Unit Two' (2) is at 100% power.
PUMP B SEAL STAGING FLOW HI/LO annunciator is received.
Recirc Pump 28 seal #2 pressure lowering towards zero-;.
Recirc Pump 28 seal #1 pressure stable at 1000 psig.
AU other indications are normal.
Which ONE of the following are these conditions indicative of?
A.
Failure of seal #1.
B.
failure of seal #2.
C.
plugging of internal restricting coil #1.
D.
plugging of internal restricting coil #2.
Proposed Answer:
C NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 C.
Correct - plugging of coil #1 will not increase any leakage, but will cause the upper seal to loose pressure.
SO-2, Section 4.2.2 Recirculation Pump Seal Failure (Figure 02-18 and 02-19)
Plugging of the number one restricting coil would result in a reduction of the number two seal pressure. Plugging the number two restricting coil would result in a high pressure condition in the number two seal. In both cases of restricting.coil plugging, flow through the seal staging line would decrease, causing a low flow alarm at 0.6 gpm (Unit 2 only).
Technical Reference(s):
SO-2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE (As available)
Learning Objective:
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam LOR-CLS-LP-002*15A-
- 26 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES~401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KIA #
2 268000 A4.01 3.4 Ability to manually operate and/or monitor in the control room: Sump integrators Proposed Question:
Common 35 Which ONE of the following describes how Drywell Equipment Drain and Drywell Floor Drain flow monitoring is accomplished and monitored.
A.
'Flow elements are in~talledoutside the Drywe,lI in the ~ischarge line from each sump. They provide a signal wh.ich drives flow integrators and a recorder in. the
IMa-m--e-o-ntral-R-oom-=-~otai_B,rywe*H-E-qtJ1-p-me-r-lt-f)f-a+n-f~ew-is-c-efls-iGe-Fed-iJe-R-t~fieG leakage and total Drywell Floor Drain flow is considered unidentified leakage.
B.
Flow el.~ments are installed inside the Drywell in the discharge line from each sump'LThey provide a signal which dri-ves flow integrators and a recorder ifl the Radwaste Control Room. Total Drywell Equipment Drain flow is considered
-identified leakage and total Drywell Floor Drain flow is considered unidentified I~akage.
C.
Flow elements are installed outside the Drywell in the discharge line from each sump. They provide a signal which drives flow integrators and a recorder in the Main Control Room. Total Drywell Equipment Drain flow is considered unidentified leakage and total Drywell Floor Drain flow is considered identified leakage.
D.
Flow elements are installed inside the Dryw.eJI in the discharge line from each sump. They provide a 'signal which drives flow integrators and a recorder in the Radwaste Control Roon-i. Total Drywell Equipment Drain flow is considered unidentified leakage and'-'total Drywell Floor Drain flow is considered identified leakage.
Proposed Answer: '
A NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Sample Written Examination Question Worksheet Form ES-401-5 B.
Incorrect - Flow elements are installed outside the drywell C.
Incorrect - Identified & Unidentified leakage defined wrong D.
Incorrect - Flow elements are installed outside the drywell, Identified & Unidentified leakage defined wrong SD-47 3.6 Drywell Equipment Drain and Drywell Floor Drain Flow Monitoring The flow elements are installed outside the Drywell in the discharge line from each sump. The flow elements and the associated electrical circuitry calculate the discharge flow of each sump pump. The calculated flowrate is sent toa three pen recorder located on the vertical section of Panel P-603 in th'e Main Control Room (recorder G16-FR-R600).
The flow signaI aISOcfrivesflow Inlegratorsillth-el'J1crtrteuntrot--Ruullrthat-are---.**.*****
used to calculate the total identified and unidentified leakage in the Drywell.
The flowrate of the Equipment Drain Pump is used to calculate the identified leakage from components inside the Drvwell. The flowrate from the Floor Drain Pump is used to calculate the unidentified leakage within the Drywell..
3.7 Control Room Annunciation Control Room annunciation occurs in two places. The Radwaste Control Room has annunciator panels which provide alarms for individual sump operation. The Main Control Room is provided with annunciators for high leak rates from individual Reactor Building sumps. The timers which control the leak rate annunciation is separate from the Radwaste Control Room run timers Technical Reference(s):
SO-47, section 3.6 and 3.7 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective:
Question Source:
Question History:
NUREG-1021, Revision 9
____________ (As available)
Bank#
Modified Bank #
(Note changes or attach parent)
New
. X Last NRC Exam
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Question Cognitive Level:
Memory or Fundamental Knowledge X---
Comprehension or Analysis 10 CFR Part 55 Content:*
55.41 55.43 Comments:
.. _.~----_.~---------------
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination -Outline Cross-reference:
Level Tier#
Group #
KIA #
2 286000 K4.01 3.4 Knowledge of FIRE PROTECTION SYSTEM design feature(s) and/or interlocks which provide for the following: Adequate supply of water for the fire protection.
Proposed Question:
Common 36 During Station Blackout conditions, the Diesel Driven Fire Pump is providing inventory makeup to the Unit Two (2) reactor vessel due to failure of RCIC and HPCI.
Fire Water Storage Tank level is rapidly lowering due to a breach of tank integrity.
.---W-I1e-r:e-shQu.ld-th.eJ)jeseLDLiven£iIa£uill.p.-.s_u.ctioJ]j)_e_aJj.gned to lAW 0 P-41 ?
A.
County Water Storage Tank.
B.
Unit One (1) Condensate Storage Tank.
C.
Unit Two (2) Condensate Storage Tank.
D.
Makeup Demineralized Water Tank.
Proposed Answer:
0 NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Explanation (Optional):
A,B,C Incorrect - Not specified in OP-41 MUD tank is the alternate source for fire water and can be aligned per OP-41.
SO-41, Section 4.2.4 Fire Protection Alternate Water Supply OOP-41, Fire Protection and Well Water System, provides a method to supply backup fire protection water in the event the normal supply becomes unavailable.
The normal water supply is a 300,000 gallon Fire Protection Water Tank with a 200,000 gallon backup from the Makeup Demineralized Water Storage Tank (MUD).
The procedure provides direction for controfroom operations to maintain fire protection water under the following conditions:
- With no fire pumps running when a breech or loss of the normal supply tank occurs, shift the fire pump suction to the MUD Tank.
- When the fire pumps are running and a lowering level in the normal supply tank occurs, shift the fire pump suction to the MUD Tank on the fly.
Technical Reference(s):
OP-41 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective:
(As available)
Question Cognitive Level:
Memory or Fundamental Knowledge Question Source:
Question History:
NUREG-1 021, Revision 9 Bank#
Modified Bank #
New Last NRC Exam LOI-CLS-LP-041-A*013-#5 x
(Note changes or attach parent)
ES-401 Sample Written Examination Question Worksheet Comprehension or Analysis Form ES-401-5 10 CFR Part 55 C.ontent:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
KJA#
2 290001 A3.02 3.5 Ability to monitor automatic operations of the SECONDARY CONTAINMENT including: building differential pressure: Plant-Specific Proposed Question:
Common 37 A valid Reactor Water Low Level 2 has occurred in Unit 2.
Which ONE of the following describes the events that occur to maintain secondary containment differential pressure?
A.
'(1) The Reactor Building Ventilation System Fans stop.
(2) ONLY One of Two SBGT fans starts (3) The isolation valves close and the fans stop in the Purge System.
(4) The SBGT inlet and outlet isolation valves open B.
(1) The Reactor Building Ventilation System Fans stop.
(2) Both SBGT fans start.
(3) The isolation valves close and the fans sto'p in the Purge System.
(4) The SBGT inlet and outlet isolation valves open C.
(1) Both SBGT fans start.
(2) The Reactor BUilding"'\\7entiiaffon System Fans stop.
(3) The isolation valves close and the fans stop in the Purge System.
(4) ONLY the outboard supply and exhaust BFIVs close D.
. (1 ) ONLY One of Two SBGT fans start.
(2) The* Reactor Building Ventilation System Fans stop.
(3) The isolation valves close and the fans stop in the Purge System.
(4) ONLY the outboard supply and exhaust BFIVs close Proposed Answer:
B
,NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Per SD-04.1, Section 2.6 Sample Written Examination Question Worksheet Form ES-401-5 Low Reactor Water Level #2 High Drywell Pressure Main Stack High Radiation Reactor Building Ventilation Exhaust High Temperature Reactor Building Ventilation Exhaust Rad High Any of the above signals initiate the following sequence of events:
- Closes the isolation dampers which stops the fans in the Reactor Building Ventilation System.
- Starts both SBGT fans simultaneously.
- Closes the isolation valves and stops the fans in the Purge System.
- Opens the SBGT inlet and outlet isolation valves (U2 only).
- Opens the SBGT Reactor Building suction valves.
- Closes the SBGT Primary Containment suction valves.
Secondary Containment Integrity is ensured by the SBGT System maintaining a minimum negative 0.25" H20 pressure at a flowrate of 3000 scfm.
Per SO 37.1, section 2.3 Reactor Building Isolation Dampers Four Reactor Building isolation dampers, two in the discharge duct from the supply fans and two in the discharge duct from the exhaust fans provide isolation capability of the Reactor Building ventilation to prevent the possible release of radioactivity under varying conditions. The dampers can be operated manually, remotely and automatically.
Additionally, the following signals automatically close all the dampers and start the Standby Gas Treatment System upon receipt of an isolation signal from the Primary Containment Isolation System (PCIS) System APP Tech Spec
- 1) Reactor low water level #2 105" ~ 101"
- 2) High Drywell pressure 1.7 psig S 1.8 psig
- 3) High Reactor Building exhaust temperature 135°F N/A
- 4) High Reactor Building exhaust radiation 4 mr/hr s 16 mr/hr
- 5) Main stack radiation high level lAW ODCM A.
Incorrect - Both SBGT fans start B.
Correct C.
Incorrect -
The inboard BFIVs also close D.
Incorrect - The inboard BFIVs also close. Both SBGT fans start.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s):
SD-04.1, section 2.6.
(Attach if not previously provided)
SD-37.1, section, 2.3 Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank#
Modified Bank #
New x
(Note changes or attach parent)
---~Q-u.esti-onJ:ilsioJ:Y---* --- -.---- -
L.as!.-NB.C_._~~~_m Question Cognitive Level:
Memory or Fundamental Knowled.ge X---
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
Examination Outline Cross-reference:
ES-401 Sample Written Examination Question Worksheet Level Tier #
Group #
KIA #
Importance Rating Form ES-401-5 RO SRO 2
2 214000.K1.05 3.3 Knowledge of the physical connections and/or cause-effect relationships between ROD POSITION INFORMATION SYSTEM and the following: Full core display.
Proposed Question:
Common 38 Rod 46-39 was selected Just prior to a reactor scram.
No operator action has been taken.
IT-he-RQobs_entE~that the "FULL-IN" light is 'Ul!Jmirrated _for rod 46-39 on the Full Core Display while the Four Rod Display Selected Rod is not indicating "00".
Which of the following describes the correct reason for the observation?
A.
RPIS Rod Drift circuitry bypasses the Four Rod Display.
B.
UPS power to the RPIS 24 VDC power supply has tripped.
C.
A reactor scram automatically deselects any selected rod.
D.
The scram must be reset to allow the rod to settle to position "00".
Proposed Answer:
D NUREG-1021, Revision 9
ES-401 Explanation (Optional):
Per SD-07 Sample Written ~xamination" Question Worksheet F'orm ES-401-5 2.0 COMPONENT DESCRIPTION/DESIGN DATA 2.1 Position Indicating Probes All of the rod position data is derived from position probes that are located "
internal to the control rod drive mechanisms. Within the position probes are hermetically sealed" reed switches that are placed at specific locations within the probe (Figure 07-3). There are 53 reed switches in every probe and one probe within each rod drive. A magnet which moves with the drive piston actuates the reed switches as the CRDM moves the control rod.
Forty-nine of the reed switches are spaced at equal 3-inch intervals, providing an indication signal at each rod latching position and at the
jA-a+fwa-yi3e~*R-t-betweer:l-e.acb_Oflb-os-e_P-o-sJtion-sJbe_s~switches are closed one at a time to transmit signals to the Rod Position Information System which, in turn, energizes the corresponding digital indicator on the four rod display in the Control Room. The even numbered switches correspond to the latched positions and the odd numbered switches correspond to the intermediate positions. If a reed switch should fail open or if no reed switch is closed, the 4-rod group will have no display for the rod position (blank).
The RWM and process computer will also in"a"icate the failure of the reed'--
switch. If the correct conditions are met, the RWM will provide an inferred position for the control rod.
Switch 5? is installed on the opposite side of the switch support from Switch 00, but slightly higher. Itclos~s at the same time as switch 00 and
- prov"ide~ a "rod full in" signal. Switch 51, installed immediately above 52, is r",¢lqs:ed'at the extreme upper e'nd of control rod insertion travel to maintain
".th,e "rod full in" signal when the rod is inserted past the 00 latched position "such as on a scram. Switch 49, located near the bottom of the probe near switch 48, closes simultaneously with 48 and generates a "rod full out" signal.
A is incorrect because RPIS Rod Drift circuitry does not bypass the Four Rod Display..
B is incorrect since a loss of this power supply would also result in a loss of the "Full In" green light.
C is incorrect since a reactor scram does not deselect any selected rod.
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet SD-07 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: __N_O_N_E Learning Objective:
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam CLS-LP-07*06A-#1 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO#39 Level Tier #
Group #
KJA#
1 295001 AA2.05 3.1 (K&A Statement) Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:
Jet pump operability Proposed Question:
Unit one (1) was initially at 91 0A> power when an unexplained
~~~~~~~~~~~~~~~~ctorRower occurred. The following cond~ions have been noted:
Generator electrical output decrease Indicated core flow increase Core plate differential pressure decrease Recirculation Flow which had been balanced now reads:
Reactor Recirculation Loop A indicated flow 31 x 106 Ibm/hr Reactor Recirculation Loop 8 indicated flow 33 x 106 Ibm/hr Reactor Recirculatton Pump.~peeds ha*ve not changed What action is required?
a.
Lock the "A" Reactor Recirculation Pump scoop tube.
b.
Perform PT-13.1, Reactor Recirculation Jet Pump Operability.
c.
Lower the speed of the "8" Reactor Recirculation Pump to within 5°A> of the speed of the "A" Reactor Recirculation Pump d.
Raise the speed of the "A" Reactor Recirculation Pump to within 5% of the speed of the "8" Reactor Recirculation Pump Proposed Answer:
NUREG-1021, Revision 9 b.
Perform PT-13.1, Reactor Recirculation Jet Pump Operability.
Sample Written Examination Question Worksheet Form ES-401-5 Explanation: Core flow has risen with no change of recirc pump speeds this does NOT comply with the reactor power and core plate dIp lowering; which indicates a lowering of core flow.
Recirc pump speed has not changed therefore it appears recirc pump B flow has risen while its speed remained the same this indicates a failed jet pump. For these indication AOP - 04.4 directs performing PT-13.1 to check jet pump operability.
b.
is incorrect because there has been no change in recirc pump speeds. In accordance with 1AOP-3.0 step 3.1.2, the scoop tube is locked when recirc pump speed is rising.
c.
is incorrect because the requirement is to balance speeds is within 10% when core flow is greater that or equal to 58 x 106 Ibs/hr. There is not a 10% imbalance.
d.
is incorrect because the speed should be lowered on the faster pump additionally the requirement is to balance speeds is within 10% when core flow is greater that or equal to 58 x 106 Ibs/hr. There is not a 10% imbalance.
Technical Reference(s):
AOP - 04.4 and SD - 02 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
None Learning Objective:
CLS-LP-302-C (As available)
- 13. Given plant conditions and OAOP-04.4, determine the required supplementary actions.
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam
. LOI-CLS-LP-302-C*013 002 N/A (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
Original question was LOI AopBnk LOI-CLS-LP-302-C*013 002 NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO#40 Level Tier #
Group #
KIA #
1 295003 AK1.03 2.9 (K&A Statement) Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Under voltage/degraded voltage effects on electrical loads.
~'
i Proposed Question:
A DBA LOCA occurs on Unit One concurrent with a Loss of Offsite Power on both Units.
A 4TS-0\\7AC-I5TISE3lOCRout occurs.due to-proie-ctive relay actuation.
What is the response of the Unit One Low Pressure ECCS systems?
Proposed Answer:
a.
One loop of Core Spray and 2 RHR pumps in one loop will inject.
b.
One loop of Core Spray and 3 RHR pumps in two loops will inject.
c.
Both Loops of Core Spray and 2 RHR pumps in one loop will inject.
d.
Both Loops of Core Spray and 3 RHR pumps in two loops will inject.
c.
Both Loops of Core Spray and 2 RHR pumps in one loop will inject.
Explanation: 4 KV Bus E3 supplies power to bus E7 which supplies power to RHR Loop A injection valve F017A. With this valve de-energized only loop B will inject. Both Core Spray loops are powered from Unit 1 buses E1 and E2.
a.
is incorrect because both loops of core spray have power from buses E1 and E2 c.
is incorrect because both I'oops of core spray have power from buses E1 and E2 and the Unit 1 A side RHR injection valve is closed with a loss of power.
d.
is incorrect because the Unit 1 A side RHR injection valve is closed with a loss of power.
Technical Reference(s):
APP - 17 2-1 Station Electrical Print or load lists NUREG-1021, Revision 9 (Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be provided to applicants during examination:
None Learning Objective:
CLS-LP-302-G
- 2. List the automatic actions expected to occur in accordance with the fOllowing AOPs:
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam x
(Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge
C-ompre-hensiof1--or-ArTa~y-s-i-s----
X:-------_
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1 021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
Level Tier #
Group #
K1A#.
2 214000 K1.05 3.3 Knowledge of the physical connections and/or cause-effect relationships between ROD POSITION INFORMATION SYSTEM and the following: Full core display.
Proposed Question:
Common 38 Rod 46-39 was selected just prior to a reactor "scra"m.
No operator action has been taken.
f-he-R-G-ebs-e"FVe-s-t-ha-t-tne--~!-~_lJ-l_b_--I-N!.!._J"igh-t-is--illu-mi"rtated--forJ-O-dA£~illhe_EuJli~_Q[e_
Display while the Four Rod Display Selected Rod is not indicating "00".
Which of the following describes the correct reason for the observation?
A.
RPIS Rod Drift circuitry bypasses the Four Rod Display.
B.
UPS power to the RPIS 24 VDC power supply has tripped.
C.
A reactor scram automatically deselects any selected rod.
D.
The scram must be reset to allow the rod to settle to position "00".
Proposed Answer:
0 NUREG-1021, Revision 9
ES-401' Explanation (Optional):
Per SD-07 Sample Written, Examination Question Worksheet,
Form ES-401-5 2.0 COMPONENT DESCRIPTION/DESIGN DATA 2.1 Posi~ion Indicating Probes All of the rod position data is derived from position probes that are located internal to the control rod drive,mechanisms. Within the position probes are hermetica,lIy sealed reed switches that are placed at specific locations within the probe (Figure 07-3). There are 53 reed switches in every probe and one probe within each rod drive. A magnet which moves with the drive piston actuates the reed switches as the CRDM moves the control rod.
Forty-nine of the reed switches are spaced at equal 3-inch intervals, providing an indication signal at each rod latching position and at the
..-.....-IhaIfwaTP-oinrbetwe-e-n-e-a-c'h-of-th-ose-pos1ticfls-:-l-M'es-e-swit-e-Aes-ar-e-Gto-seG---
one at a time to transmit signals to the Rod Position Information System which, in turn, energizes the corresponding digital indicator on the four rod display in the Control Room. The even numbered switches correspond to the latched positions and the odd numbered switches correspond to the intermediate positions. If a reed switch should fail open or if no reed switch is closed, the 4-rod group will have no display for the rod position (b'lank).
The RWM and process computer will also in'cl'icate the failure of the reed'*--
switch. If the correct conditions are met, the RWM will provide an inferred position for the control rod.
Switch ~2 is installed on the opposite side of the switch support from Switch 00, but slightly higher. It closes at the same time as switch 00 and provid~s a i'rod full 'in" signal. Switch 51, installed immediately above 52, is closed at the extreme upper end of control rod i,nsertion travel to maintain the "rod full in" signal when the rod is inserted pa~t the 00 latched position such as on a scram. Switch 49, located near the bottom of the probe near switch 48, closes simultaneously with 48 and generates a "rod full out" signal.
A is incorrect because RPIS Rod Drift circuitry does not bypass the Four Rod Display.
B is incorrect since a loss of this power supply would also result in a loss of the "Full In" green light.
C is incorrect since a reactor scram does not deselect any selected rod.
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet SO-07 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam CLS-LP-07*06A-#1 (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO#39 Level Tier #
Group #
KIA #
1 295001 AA2.05 3.1 Proposed Question:
(K&A Statement) 'Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:
Jet pump operability Unit one (1) was initially at 91 % power when an unexplained
**----.jEle6fe*ase-i-n-r-e-at-eF-f30wer-eGGtMr-e~_J::le_folIQwingcond~av_e--.----------~
been noted:
Generator electrical output decrease Indicated core flow increase Core plate differential pressure decrease Recirculation Flow which had been balanced now reads:
Reactor Recirculation Loop A indicated flow 31 x 106 Ibm/hr Reactor Recirculation Loop 8 indicated flow 33 x 106 Ibm/hr Reactor Recirculali-on Pump speeds have not changed What action is required?
a.
Lock the "A" Reactor Recirculation Pump scoop tube.
b.
Perform PT-13.1, Reactor Recirculation Jet Pump Operability.
c.
Lower the speed of the "B" Reactor Recirculation Pump to within 5% of the speed of the "A" Reactor Recirculation Pump d.
Raise the speed of the "A" Reactor Recirculation Pump to within 5% of the speed of the "8" Reactor Recirculation Pump 1
Proposed Answer:
NUREG-1021, Revision 9 b.
Perform PT-13.1, Reactor Recirculation Jet Pump Operability.
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Explanation: Core flow has risen with no change of recirc pump speeds this does NOT comply with the reactor power and core plate dIp lowering; which indicates a lowering of core-flow.
Recirc pump speed has not changed therefore it appears recirc pump B flow has risen while its speed remained the same this indicates a failed jet pump. For these indication AOP - 04.4 directs performing PT-13.1 to check jet pump operability.
b.
is incorrect because there has been no change in recirc pump speeds. In accordance with 1AOP-3.0 step 3.1.2, the scoop tube is locked when recirc pump speed. is rising.
c.
is incorrect because the requirement is to balance speeds is within 10% when core flow is greater that or equal to 58 x 106 Ibs/hr. There is not a 10°A> imbalance.
d.
is incorrect because the speed should be lowered on the faster pump.additionally the requirement is to balance speeds is within 10% when core flow is greater that or equal to 58 x 106 Ibs/hr. There is not a 10°A> imbalance.
Technical Reference(s):
AOP - 04.4 and SD - 02 (Attach if not previously provided)
===============::::::::::::::======::::::::::::::::;:::;;;:;;::;;;====~-------~-_. -. _.. ---_.._----------_.-
Proposed references to be provided to applicants during examination: _N_o_ne Learning Objective:
CLS-LP-302-C (As available)
- 13. Given plant conditions and OAOP-04.4, determine the required supplementary actions.
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam LOI-CLS-LP-302-C*013 002 N/A (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
Original question was LOI AopBnk LOI-CLS-LP-302-C*013 002 NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO#40 Level Tier #
Group #
KIA #
Importance Rating RO
~~SRO 1
1 295003 AK1.03 2.9 (K&A Statement) Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Under voltage/degraded voltage effects on electrical loads.
I
'j Proposed Question:
A DBA LOCA occurs on Unit One concurrent with a Loss of Offsite Power on both Units.
A 4160 VAC bus E3 lockout occurs due to protective relay actuation.
What is the response of the Unit One Low Pressure ECCS systems?
'\\
Proposed Answer:
a.
One loop of Core Spray and 2 RHR pumps in one loop will inject.
b.
One loop of Core Spray and 3 RHR pumps in two loops will inject.
c.
Both Loops of Core Spray and 2 RHR pumps in one loop will inject.
d.
Both Loops of Core Spray and 3 RHR pumps in two loops will inject.
c.
Both Loops of Core Spray and 2 RHR pumps in one loop will inject.
Explanation: 4 KV Bus E3 supplies power to bus E7 which supplies power to RHR Loop A injection valve F017A. With this valve de-energized only loop B will inject. Both Core Spray loops are powered from Unit 1 buses E1 and E2.
a.
is incorrect because both loops of core spray have power from buses E1 and E2 c.
is incorrect because both loops of core spray have power from buses E1 and E2 and the Unit 1 A side RHR injection valve is closed with a loss of power.
d.
is incorrect because the Unit 1 A side RHR injection v"alve is closed with a loss of power.
Technical Reference(s):
APP - 17 2-1 Station Electrical Print or load lists NUREG-1021, Revision 9 (Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be provided to applicants during examination: _N_on_e Learning Objective:
CLS-LP-302-G
- 2. List the automatic actions expected to occur in accordance with the following AOPs:
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam x
(Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9 x
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Exami-n-ation--Outline Cross-reference:
R0#41 Level Tier#
Group #
KIA #
1 295004 2.1.30 3.9 (K&A Statement) Conduct of Operations: Ability to locate and operate components, including local controls.
Proposed Question:
During an outage, 125 DC Battery and Battery Charger 1B-2 is to be taken out of service, ~ultlngjn_JQ.~s_of the normal source of power to DC Distribution Panel 9A.
Which ONE of the following describes the transfer of DC
---"---~Distrr6Lffr(frf--P-aner9AlOifS-a1ternate sourcer a.
At DC Distribution Panel 9A close the alternate supply breaker, then open the normal supply breaker.
b.
At DC Distribution Panel 9A open__ttle o_Q[malsupply breaker, then close the alt~n_ate supply breaker.
c.
In the battery room remov~ the cover from the box on the wall that contains the manual transfer switches then ~Iose the--a-Ite-rnate supply breaker, then open the normal supply breaker.
d.
In the battery room remove the cover from the box on the wall that contains the manual transfer switches then operi the normal supply breaker, then close the alternate supply breaker.
b.
Proposed Answer:
Procedure guidance requires opening the normal supply breaker, then closing the alternate supply breaker to maintain divisional_separation. The breakers a re mechanically interlocked and located on DC Panel 9A.
Explanation: Per OWP-51/1. Interlocks prevent cross connecting OC power supplies when alternate power sources are selected (SO-51). See -procedure and SO sections below.
NUREG-1021, Revision 9
ES-401 Sample Written' Examination Question Worksheet Form ES-401-5 a.
is incorrect because closing the alternate breaker while the normal breaker is closed will cro'sstiethe divisions and is not permitted, furthermore its blocked by the mechanical interlocks.
b.
correct c.
is incorrect because this is description of shifting the Battery Charger power supplies, additionally closing the alternate breaker while the normal breaker is closed will crosstie the divisions and is not permitted, furthermore its blocked by the existing mechanical interlocks.
d.
is incorrect because this is description of shifting the Battery Charger power supplies, the DC Panel 1B power supplies are located on the panel not the, battery room wall.
Technical Reference(s):
OWP-51//1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: _N_on_e
--'-------*-L-~'~rning 0 bjective-:-C[s::tP=5-1~t3:--rJe-s-crtb-e-th-elo-c-ation-a'nd-op'erat+on-of'~As avanable)
Battery Chargers 1B-1, 1B-2, 2B-1, and 2B-2 AC Power Transfer Switches.
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam LOI SYTEMS 295004 K1.03 001 M (Note, changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge X---
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
Original question was LOI AopBnk LOI-CLS-LP-302-C*013 002 From OWP-51/1 3.4.3 TRANSFER DC Distribution Panel 9A, located in Unit 1 4160V BOP bus area, to alternate as follows:
- 1. NOTIFY System Dispatcher transfer of DC Distribution Panel 9A will cause a momentary loss of power to the Power System Stabilizer.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5
- 2. OPEN normal power supply incoming feeder breaker, NORM.
- 3. CLOSE---alternate power supply incoming feeder breaker, ALT.
From SD - 51 The battery chargers are powered from the 480 VAC emergency distribution system. This ensures a reliable source of AC power to maintain charger operation and minimize battery depletion during various losses of AC power. Additionally, each unit's division 2 battery chargers have another source of emergency AC power via manual transfer switches mounted on the wall in their respective battery rooms. These manual transfer switches are inside boxes that require cover removal, and are rotated to align the alternate source. These alternate sources of power and transfer switches were installed to satisfy 1O-CFR-50 Appendix R requirements in the event of a fire in the normal power source motor control center.
NUREG-1021, Revision 9
ES-401 S.ampleWritten Examination Question Worksheet Form ES-401-5 Prop Examination Outline Cross-reference:
R0#42 Level Tier #
Group #
KIA #
1 295005 AK2.01 3.8 (K&A Statement) Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: RPS.
Proposed Question:
Unit One (1) 'is operating at full power when a Turbine Control Valves (TCV) fast closure occurs. Which one of the following at a MINIMUM must be se-nsed by the RPS ~ystem to result-in a full reactor scram? '
Low disc dump pressure sensed on:
a.
10 L b.
c.
d.
TCV2 any two TCVs TCVs 1 or 3 and 2 or 4.
any three of the four TCVs Proposed Answer:
c.
TCVs 1 or 3 and 2 or 4.
Explanation:
A load reject signal will energize the fast acting Solenoid Valves on the control valve actuators, which removes hydraulic trip fluid pressure. The trip signal comes from pressure switches on the fast acting trip control (FATC) supply to the control valve disc dumps. Circuitry is designed such that the pressure switch on either control valve. 1 or 3 will trip RPS Trip System A. Either control valve 2 or 4 will trip RPS Trip System B. These switches will also provide a Scram signal on loss of hydraulic trip fluid pressure when a load reject signal is not present.
a.
is incorrect because the full scram signal requires inputs from valves 1 or 3 and 2 or 4.
b.
is incorrect because specifically valves 1 or 3 and 2 or 4 must sense a trip signal.
d.
is incorrect because the MINIMUM number of full scram signal requires inputs from 2 valves (1 or 3 and 2 or 4). This statement is more than the MINIMUM.
SO -26, Table 26-4 Technical Reference(s):
NUREG-1021, Revision 9 (Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination-Outline Cross-reference:
R0#43 Level Tier #
Group #
KIA #
1 295006 AA2.01 4.5 (K&A Statement) Ability to determine and/or interpret the following as they apply to SCRAM:
Reactor Power Proposed Question:
Following a loss of the Uninterruptible Power Supply system, a Reactor scram occurs. Indications on P603:*
APRM/IRM Re.C.orders AERM Downscale SRM Recorders SRM--PerTods-.---------- -..
Full Core Display Blank Lights illuminated Blank
---;;-I-nd---::-:i~c-at::-:-in--g-------=8:::-=0-s-e-co-n---=-ds---.L--.---.-.. -_ -..-
Rod full-in/full-out indications unavailable ERFIS and the Process Computer are not available. What is the current status of the Reactor?
a.
Reactor ppwer cannot be determined to be <2°k.
}j.
b.
Reactor power can be determined to be <2°k from the RTGS.
c.
Reactor power can be determined to be <2% Q-nly. in the back panels.
d.
Reactor Ga-n be determined to be shutdowA-under all conditions'without boron.
Proposed Answer:
b.
Reactor power can be determined to be <2°k from the RTGB.
Explanation:
APRM apron lights operate independent of UPS power. With downscale lights lit, operator can determine power <2°k using these indications on P603, with loss of
- UPS, a.
is incorrect because reactor power can be determined from the illuminated APRM downscale lights.
b.
correct c.
is incorrect reactor power can be determined from the RTGB where the downscales lights are located.
d.
is incorrect because control rod positions cannot be determined, therefore reactor cannot be determined to be shutdown under all conditions without boron.
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination' Form ES-401-5 Question Worksheet 001-37.5 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
none Learning Objective:
CLS-LP-300-E, 14. Given plant conditions and the Level/Power Control Procedure, determine the following:
- e. If the reactor is shut down.*
(As available)
Question Source:
Bank#
Modified Bank #
New LOIEopBnk LOI-CLS-LP-300-C*007 005 (Note changes or atlacff-"---."--'-_.._..__._--._-_.-._.._
parent)
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
Original question was LOI AopBnk LOI-CLS-LP-302-C*013 002 NUREG-1021, Revision 9
ES-401 Sample Written Examination
. Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
R0#44 Level Tier #
Group #
KIA #
Importance Rating
- RO SRO 1
1 295016 AK2.01 4.4 (K&A Statement) Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following: Remote shutdown panel: Plant Specific Proposed Question:
The main control room has been evacuated in accordance with OAOP-32. Control has been transferred to the Remote Shutdown Panel (RSDP). While operating in this condition, which one of the following ~9_n be perform~d per OAOP-32 from the RSDP and r:e!atecLstations,-.-?--__.
A.
Operate RCIC in full flow test for reactor pressure control.
I B.
Reject coolant to Rad Waste while RHR Loop B is in Shutdown Cooling.
C.
Perform emergency depressurization by taking seven ADS control switches to OPEN.
D.
Spray the Drywell with RHR Loop A while in Suppression Pool Cooling with RHR Loop B.
Proposed Answer:
B.
Reject coolant to Rad Waste while RHR Loop B is in Shutdown Cooling.
Explanation:
- a. is correct because you can operate RCIC in the Pressure Control Mode and the procedure allows for RCIC injection in this mode. You cannot operate RHR Loop A from the RSDP and there are no controls for rejecting water from the RSDp
. a.
is incorrect RHR Lo op B is operated locally and only monitored from the RSDP c.
is incorrect because seven ADS cannot be manually initiated when control is transferred to the RSDP, there are only three SRV controls on the panel.
d.
is incorrect there are no controls on the RSDP for RHR Loop A.
Technical Reference(s):
SD-62 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
_n_o_n_e NUREG-1021, Revision 9
ES-401 Learning Objective:
Sample Written Examination Question Worksheet CLS-LP-62, 3. List the systems that can be cont[o_lJeq from the Remote Shutdown Panel or local control stations.
Form ES-401-5 (As.available)
Question Source:
Bank#
Modified Bank #
New LOI AopBnk LOI-CLS-LP-302-E*007 001 (Note changes or attach parent)
Question History:
Last NRC Exam LOI
--- _----- --------------============---------
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:.
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination-Outline Cross-reference:
R0#45 Level Tier #
Group #
KIA #
Importance Rating RO----- --
SRO 1
1 295018 AA2.04 2.9 (K&A Statement) Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: System flow:.
Proposed Question:
FO'lIowing a loss of Off-Site Power on Unit One (1), Diesel Generators #1 and #2 are tied to their respective Emergency Buses. A rupture of the Unit One (11 Nuclear Service Water Header in the Service Water Building rest1lts in the following indications:
Nuclear Header Pressure (XU-2)
Nuclear Hdr Serv Wtr Press-Low 20 psig Alarm sealed in Nuclear Service Water pumps are manually tripped.
Diesel Generators #1 and #2:
A.
have no available cooling water and will trip.
B.
have no available cooling water but will continue to run.
C.
cooling water supply will automatically transfer to the Unit Two (2) Nuclear Service Water header.
D.
cooling water supply will automatically transfer to the Unit One (1) Conventional Service Water header.
c.
Proposed Answer:
cooling water supply will automatically transfer to the Unit Two (2) Nuclear Service Water header Explanation:
Should the service water pressure upstream of the jacket water heat exchanger remain below 5.6 psig for 30 seconds when the valve is open the alternate unit supply valve will open, then the normal supply valve will close.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 a.
is incorrect because Section 2.0 of OAOP~18, Nuclear Service Water System Failure, If the EDGs are operating the cooling water valves will shift to the oppositeUDitJ~$W b.
is incorrect the EDGs will trip on either high lube oil temperature 190°F or high jacket water temperatur.e 200°F d.
is incorrect beca.use the loss of cooling water was caused by rupture of the Nuclear Service Water header. If the Conventional Service Water Pump starts it will only pump into the break. The NSW Alternate valve at the EDG uses the local NSW pressure to determine a loss of NSW.
Technical Reference(s):
SD-39 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
None Learning Objective:
CLS-LP-43, 3. List the loads that can be s~pplied by the (As available)
- ***********----------Nuclear:-S.ervlce-Water_S-¥_stem. ~
~
Question Source:
Bank#
Modified Bank #
New LOI AopBnk LOI-CLS-LP-302-H*02B 003 (Note changes or attach parent)
Question History:
Last NRC Exam used on Dec. 1995 NRC exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments: AOP-18.0 PG 3 From 50-39, Section 2.7 Diesel Generator Service Water (Figure 39-7)
Two service water supply lines provide service water to the tube side of each EDG set jacket water cooler. Each unit's Nuclear Service Water (NSW) System provides an independent source to all four Diesels. Diesel generator start and speed increase above 500 rpm opens the valve NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 from the respective unit's NSW header. Should the service water pressure upstream of the jacket water heat exchanger remain below 5.6 psig for 30 seconds when the valve is open the alternate unit supply valve will open, then the normal supply valve will close. When the engine is shutdown and speed drops below 500 rpm the open valve will close. This switching sequence is initiated any time service water flow is lost when an EDG set is operating. Return flow of service water from all four jacket water coolers is routed to a common return line which discharges to the intake canal.
Automatic Stop Control (Figure 39-14)
Under conditions where continued Diesel Generator operation may cause damage to the Diesel itself automatic shutdowns are provided. The shutdown signals will vary dependent upon whether the engine has been started manually or automatically.
When operating due to receipt of an automatic start signal the following trips and lockout are provided:
- Low lube oil pressure 27 psig
- Overspeed 590 rpm
-_._---------------~-----------------
When operating as a result of an initiation from a normal non-emergency start the following trips and lockouts are enforced in addition to those listed above:
- High lube oil temperature 190°F
- High jacket water temperature 200°F
- Jacket Water Low pressure 12 psig The low lube oil pressure, low jacket water pressure, and overspeed shutdowns are blocked for the first forty-five second on initiation of an engine start sequence (auto or manual). This permits the conditions to be established which will prevent these shutdowns during engine operation (Le., jacket water and lube oil pumps running.)
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-.5 Examination-Outline Cross-reference:
R0#46 Level Tier #
Group #
KIA #
Importance Rating RO
---S*R-O**
1 1
295019 AA1.01 3.5 (K&A Statement) Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Backup air supply Proposed Question:
The RX Bldg Standby Compressors have been returned to service..
During rated power operation on Unit.One (1) the Division II RNA supply header develops a leak. The following is observed by the CO:
~~~~I~~~~~~E~~ L~~~~~~----------~-
.----- *1B Reactor Building't\\ir-Compressor is Running Instrument air pressure.a.t the RTGS is 115 psig Service air pressure at the RTGB is 115 psig Based on these plant conditions what automatic action(s) are verified?
A.
1A Reactor ~uilding Air Compressor st~rts.
B.'
Service air valves PV-706-1 and PV-706-2 close.
C.
High Pressure Bottle Rack Isolation Valve, RNA-SV-5481, opens.
D.
Secondary Containment Isolation Valves 1B-BFIV-RB and 1D-BFIV-RB close.
Proposed Answer:
c.
High Pressure Bottle Rack Isolation Valve, RNA-SV-5481, opens.
Explanation:
Outboard MSIVs receive pneumatics from Div I & II. Backup Nitrogen aligns due to low RNA header pressure (auto action of AOP-20.0) The low pressure is indicated by the annunciator RB INSTR AIR RECEIVER 1B PRESS LOW (UA-01 1-2) which is set at 95 psig.
a.
is incorrect because the 1A compressor is the DIV I backup and would not have any effect on DIV II.
b.
is incorrect because the PV-706-1 and PV-706-2 valves isolate the Service Air Header. This isolation removes non-essential Service Air loads and is not associated with the RNA supply header.
d.
is incorrect because these are secondary containment ventilation isolation valves and are not supplied by RNA.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s):
OAOP-20.0, SD-46 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
None Learning Objective:
CLS-LP-46, 5. Given plant conditions, determine if the following automatic actions should occur:
- d. Nitrogen Backup Initiation (As available)
LOI AopBnk LOI-CLS-LP-302-K*02 002 Bank#
Modified Bank #
(Note changes or attach
pafent1-'--~'--"-'-
Question Source:
New Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X----
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
From SD-46, If the PNS/RNA header pressure drops to 95 psig, the Nitrogen Backup System valves will open to supply the Drywell loads, but the PNS/RNA isolation valves will not close.
The common pneumatic loads will be supplied by the system that has the highest pressure.
Check valves are installed to prevent the Nitrogen Backup System from supplying the other PNS/RNA loads.
4.2 Abnormal,Operation 4.2.1 Loss of Pneumatic Systems If the Service Air pressure lowers to S 105 psig, Service Air Isolation Valves PV-706-1 and PV-706-2 automatically close to isolate the Service Air Header. This isolation removes non-essential Service Air loads in an effort to'restore air header pressure to normal which will preserve pneumatics to essential loads - those required for safe plant operation. The setpoint of 105 psig for the isolation provides a sufficient margin for the service air compressors to start and load to restore header pressure before the isolation occurs. If service air header pressure NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 decreases to 98 psig, then Service Air Dryer A Bypass Pressure Control Valve, SA-PV-5067, begins to open.
NUREG-1 021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination-OutlineC-ross-reference:
R0#47
-Level Tier #
Group #
KIA #
Importance Rating RO---------------S-R-O----
1 1
295021 AK2.03 3.6 (K&A Statement) Knowledge of the interrelations between LOSS,OF SHUTDOWN COOLING and the following: RHR/shutdown cooling Proposed Question:
When establishing Alternate Shutdown Cooling per AOP-15.0, is it preferred to establish injection with:
r'
-a.
RHR due to the"injection path through the jet pumps.
b~
Core Spray due to the ability to take a suction from the CST:
c.
Core Spray due to the injection path through the spray header.
d.
RHR due to the ability to take a suction from the reactor vessel.
Proposed Answer:
a.
RHR due to the injection path through the jet pumps.
Explanation:
During Alternate Shutdown Cooling using SRVs, a note in the procedure identifies RHR as the preferred pump for injection into the reactor vessel. This is because the flow path of the RHR system provides flow up through the core, therefore providing better decay heat removal than can be provided by the Core Spray system.
b.
is incorrect during a loss of shutdown cooling Core Spray takes a suction from the suppression pool.
c.
is incorrect because Core Spray discharges into the top of the core.
d.
is incorre'ct during a loss of shutdown cooling RHR takes a suction from the suppression pool.
Technical Reference(s):
OAOP-15.0 & CLS-LP-302-L (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
None NUREG-1021, Revision 9
ES-401 Learning Objective:
Sample Written Examination Question Worksheet CLS-LP-302-L, 7. State the reason(s) for the following actions taken during the use of AlternCite Shutdown Cooling:
- g. Using RHR as preferred pump over Core Spray in Alternate Shutdown Cooling.
Form ES-401-5 (As available)
Question Source:
Bank#
Modified Bank #
New LOI AopBnk LOI-CLS-LP-302-L*07G 001 (Note changes or attach parent)
Ouesti-o-n-History-"--:---'------Il::a-st~-RG-e~-m---NJ.t=\\-A.------------
Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 NUREG-1021, Revision 9
ES-401 Sample Written Examin~tion Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
R0#48 Level Tier #
Group #
KIA #
1 295023 AK2.03 3.4 (K&A Statement) Knowledge of the interrelations between REFUELING ACCIDENTS and the following: Radiation monitoring equipment.
Proposed Question:
Unit Two (2) is in a refueling outage with Primary Containment Ventilatio.n During Personnel Entry in progress.-(Containment CAC Group 6 vent and purge valves open). A spentfuel bundle is dropped resulting in the following ala~rms being received:
Area Rad Refuel Floor High (UA-03 3-7)
Process Rx Bldg Vent Rad Hi (UA-03 4-5)
Process Rx Bldg Vent Rad Hi-Hi (UA-03 3-5)
The sca enters AOP-05.0 and vefi.fies that SBGT initiates and SeCQ.o9aryContainment=isolates.
Which ONE of the following additional automatic actions, if any, occur at this time?
A.
NONE.-
B.
and CAC vent and purge valves isolate ONLY.
C.
and Control Room Emergency Ventilation System is in operation ONLY.
D.
and CAC vent and purge valves' isolate, and Control Room Emergency Ventilation System i~ in operatio.n..
Proposed Answer:
B.
and CAC vent and purge valves isolate.
Explanation:
Per auto and immediate actions of AOP-05.0
.NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 a.
is incorrect because a Group 6 isolation is required and because a fuel assembly has been damaged Control Room Emergency Ventilation System must beverfiedJQ _b~jn__ QJ~~ration.
b.
correct because a Group 6 isolation is required and because a fuel assembly has been damaged Control Room Emergency Ventilation System is must be ensured to be in operation.
c.
is incorrect because a Group 6 isolation is required and because a fuel assembly has been damaged Control Room Emergency Ventilation System must be verified to be in operation d.
is incorrect because CREV does not Auto Start Technical Reference(s):
OAOP-05.0-(Attach if not previously provided)
Proposed references to be provided to applicants during examination:
None
____________L_~~rn!!19__Qbjective:
CLS-LP-302-J, 8. List the immediate operator actions (As available) re-qutre-d-to-b-e-pe-rfofmed--i-n--aec-efia-Ase-w~t-h-Q,,",4~Or-rP--=...-5...,,.~O'----
"Radioactive Spills, High Radiation, and Airborne" Activity".
Question "Source:"
Bank#
Modified Bank #
New LOI-CLS-LP-302-J*002 002 (Note changes or attach parent)
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
Per OAOP-5 2.1 IF PROCESS RX BLDG VENT RAD HI-HI (UA-03 3-5) is in alarm, THEN the following actions occur:
- Reactor Building Ventilation isolation
- SBGTS auto start
- Group 6 Isolation.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 295024 Gen 2.2.25 Examination Outline Cross-reference:
R0#49 Level Tier #
Group #
KIA #
Importance Rating
.RD.
1 1
2.5
_.S.RO (K&A Statement) 2.2.25 Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. (High Drywell Pressure)
Proposed Question:
Which one qf the following prevents exceeding the maximum allo\\Nable drywell pressure during a Design Basis Accident?
The Technical Specifications requirement that:
A.
average drywell air temperature of be maintained <150 0
F.
r Proposed Answer:
B.
the unit will be shut down when suppression pool temperature> 110 0
F..
C.
primary containment isolation valves be operable and close within specified times.
D.
The suppression chamber-to-drywell vacuum breakers are all required to be closed.
B.
the unit will be shut down when suppression pool temperature
> 110 0
F.
Explanation:
This limitation ensures that peak primary containment pressures and temperatures do not exceed maximum allowable values during a postulated DBA a.
is incorrect because this limit prevents exceeding the primary containment design temperature.
c.
is incorrect this limit insures that the release of radioactive materials is limited to design limits.
d.
is incorrect because this requirement insures that the drywell does not exceed the design negative differential pressure across the suppression chamber-drywell boundary.
Technical Reference(s):
Reference LCO T.S Bases B 3.6.2.1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
none Le.arning Objective:
(As available)
Question Source:
NUREG-1021, Revision 9 Bank#
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Modified Bank #
New New (Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X----
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
___ Reference T.S. Bases B 3.6.2.1 A Iim-ii8tio-n on ttl-e suppression pool ave-rag-e temperalTIfeiStequireatopTovia-e-assurarrce-that----------------------------
the containment conditions assumed for the safety analyses are met. This limitation subsequently ensures that peak primary containment pressures and temperatures do not exceed maximum allow~ble values during a postulated DBA or any transient resulting in heatup of the suppression pool.
Bases 3.6..1.3 The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) to within limits. Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DBA.
3.6.1.4 In the event of a DBA, with an initial drywell average air temperature less than or equal to the LCO temperature limit, the resultant peak accident temperature is maintained below the drywell design temperature. As a result, the ability of primary containment to perform its design function is ensured.
3.6.1.6 The function of the suppression chamber-to-drywell vacuum breakers' is to relieve vacuum in the drywell. There are 10 internal vacuum breakers located on the vent header of the vent system between the drywell and the suppression chamber, which allow flow from the suppression chamber atmosphere to the drywell when the drywell is at a negative pressure with respect to the suppression chamber. Therefore, suppressionchamber-to-drywell vacuum breakers prevent an excessive negative differential pressure across the suppression chamber-drywell boundary.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO#50 Level Tier #
Group #
KIA #
1 295025 2.1.23 3.9 (K&A Statement) Ability to perform specific system and integrated plant procedures during different modes of plant operation.
Proposed Question:
Following a Unit One (1) reactor scram conditions require the operating crew to perform Reactor Scram Procedure, EOP-01-RSP.
Plant conditions are:
--~.. ----,----~..._'..-- -----Reactof-wat-e!-JeveJ Reactor pressure 187 inches,_ste~d~y~
-1100 psig', SRV/ADS valves are opening on their auto-openfng setpoints EOP-01-RVCP requires that SRV/ADS valves be opened until reactor pressure decreases to: '
A.
950 psig. The CO must adhere to a specified SRV/ADS valve opening sequence..
B.
1050 psig. The CO must adhere to a specified SRV/ADS valve opening sequence.
C.
950 psig. The CO is not required to adhere to a specified SRV/ADS valve opening sequence.
D.
1050 psig. The CO is not required to adhere to a specified SRV/ADS valve opening sequence.
Proposed Answer:
c.
950 psig. The CO is not required to adhere to a specified SRV/ADS valve opening sequence.
Explanation: EOP-01-RVCP requires pressure be reduced below 950 psig when ADS/SRV valves are cycling. No pressure control band or SRV/ADS* valve opening sequence is specified.
a.
is incorrect because there is no specific opening sequence.
b.
is incorrect because RPV pressure must be lowered below 950 psig.
c.
correct.
d.
is incorrect because RPV pressure must be lowered below 950 psig.
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet EOP-01-RVCP,001-37.4 (Attach if not previously--provided)
Proposed references to be provided to applicants during examination:
none Learning Objective:
CLS-LP-300D, 9. Given plant conditions, the Reactor Vessel Control Procedure, and which steps have been completed, determine the required operator actions.
(As available)
Question Source:
Question History:
Bank#
Modified Bank #
New Last NRC Exam EOP Bank LOI-CLS-LP-300-D*009 N/A (Note changes or atta-ch-------
parent)
Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 X---
55.43 Comments:
001-37.4 If SRVs are used, they should be opened one at a time until a sufficient number have been opened to reduce reactor pressure at least to the value at which steam flow through the main turbine bypass valves is at 100% of bypass capacity (950 psig}. "
No pressure control band or SRV opening sequence is specified here, since the purpose of the instruction is simply to reduce reactor pressure quickly and effect direct, positive control of the SRVs.
Original question was N/A NUREG-1021, Revision 9
Sample Written Examination Question Worksheet Form ES-401-5 Examination-Outline Cross-reference:
RO#51 level Tier #
Group #
KIA #
Importance Rating RO
5-RO 1
1 295026. Gen 2.1.2 3.0 (K&A Statement) Knowledge of operator responsibilities during all modes of plant operation.
(Suppression Pool High Water Temp)
Proposed Question:
Following a Group 1 isolation, control rods failed to insert on Unit two (2). The crew is executing EOP-01-lPC and EOP-02-PCCp.
Current plant conditions are:
S-tiPPr-pool-temop Suppr pool level Reactor pressure Reactor level Reactor power SRVs
--- -------1e4°F---
-29" 950 psig-TAF lOA>.
One open Which ONE of the following actions is required by Emergency Operating Procedures?
A.
Perform emergency depressurization. _
B.
Equalize around and open MSIVs to establish the main condenser as a heat sink.
C.
Open an adqitional SRV to reduce reactor pressure, 1OO°F/Hr cooldown rate may be*
exceeded.
D.
Open an additional SRV to reduce reactor pressure, 1OO°F/Hr cooldown rate-may NOT be exceeded.
Proposed Answer:
A.
Perform emergency depressurization.
Explanation: Conditions are unsafe on HCTl requiring Emergency Depressurization.
b.
is incorrect because MSIVs cannot be opened because of the exceeding HCTl requires immediate."emergency depressurization.
c.
is incorrect because Opening additional SRV to reduce pressure is directed only before exceeding HCll.
d.
is incorrect because Opening additional SRV to reduce pressure is directed only before exceeding HCll.
NUREG-1021, Revision 9
ES-401 Technica-I Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet EOP-02-PCCP and the HCTl (Attach if not previously provided) graph, 01-37.8 Proposed references to be provided to applicants during examination:
Provide EOP-02-PCCP and the HCTl graph as reference learning Objective:
lOI-ClS-lP-300-l*05A (As available)
Question Source:
Bank#
Modified Bank*#**
New X lOI SYSTEMS 295026 A2.03 001 (Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X---
10 CFR Part 55 Content:
55.41 55.43 Comments:
From 01-37.8 As long as suppression pool water level remains at or above the elevation of the downcomer vent openings (-5.5 feet), the need to emergency depressurize the reactor due to suppression pool heatup is dictated by the Heat Capacity Temperature Limit. Actions associated with the Heat Capacity Temperature Limit are described under the discussion of Steps SPIT-10 through SP/T-12.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO#52 level Tier #
Group #
KIA #
1 295028 EK3.01 3.6 (K&A Statement) Knowledge of the reasons for the following responses as they apply to HIGH DRYWEll TEMPERATURE: Emergency depressurization.
Proposed Question:
During accident conditions the following conditions exist in Unit two (2):
RPV pressure Drywell pressure Drywell-temperaturee Suppr chamber press Suppr pool level Suppr pool temp 500 psig 14 psig 300~F-13 psig
- 5.0 feet 160°F Why must emergency. depressurization be immediately performed?
A.
Tehere is steam in the suppression chamber free air space.
B.
Containment parameters have ~ntered the unsafe region of PSP.
C.
Prevent.exceeding the suppression chamber boundary design load.
D.
'Prevent prolonged operation above the drywell and SRV design temperatures.
Proposed Ans:wer:
D.
Prevent prolonged operation above the drywell and SRV design temperatures.
Explanation: Extended operation above 300°F is not permitted and the temperature should not be allowed to exceed the SRV maximum qualification temperature of 340°F.
a.
is incorrect b~cause suppr pool level is above -5.5 feet.
b.
is 'incorrect because containment parameters are still within PSP.
c.
is incorrect because suppr pool parameters are within HCTl limits.
001-37.8 and SD-04 Technical Reference(s):
NUREG-1021, Revision 9 (Attach,if not pre~iously.provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be provided to applicants during examination:
none Learning Objective:
CLS-LP-300L, 4. State tre effect on Primary -
Containment if the following limits are exceeded:
- h. Drywell Design Temperature Limit (As available)
Question Source:
Question History:
Bank#
Modified Bank #
(Note changes or attach parent)
New X
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
001-37.8 If drywell temperature is already above 30QoF when Step DWIT-16 and DWIT-19 are reached,
.drywell sprays may still be used, if available, in preference to emergency depressurization. If sprays are effective in reducing the drywell temperature, emergency depressurization need not be performed. Extended operation above 30QoF is not permitted and the temperature should not be allowed to exceed the SRV maximum qualification temperature of 340°F.
Original question was N/A NUREG-1021, Revision 9
ES-401 Sample Written Examination
, Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO #53,
Level Tier #
Group #
KIA #
Importance Rating RO-SRO 1
1 295028. EA1.04 3.9 (K&A Statement) Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE: Drywell pressure Proposed Question:
Following a loss of drywell,cooling a,Unit One (1) scram occurs due to high drywell pressure. A steam lea-j( in the drywell subsequently results in the following plant conditions:.
_.. RPVleveL-----..
RPV pressure Drywell pressure Drywell average temp Suppr pool level
.+tOO" 350 psig 7.5 psig 295°F and steady
'-5.5 inches Regarding using RHR to spray the drywell:
A.
Spray the drywell to lower Drywell pressure below the LOCA setpoint.
B.
Do NOT spray the drywell. Defeat drywell cooler LOCA lockout per SEP-10.
C.
Do NOT spray the drywell. Open 7 ADS valves for emergency depressurization.
D.
Spray the drywell to restore and~maintain drywell temperature below design limit Proposed Answer:
d.
Spray the drywell to restore and maintain drywell temperature below design limit.
Explanation:
BEFORE drywell temperature reaches 300 degrees drywell spray is required although it is close to the unsafe region of DWSIL.
a.
is incorrect because there is no requirement at this time to spray the drywell because of drywell pressure (suppression pool sprays are not required until 11.5 psig and drywell sprays are secured before drywell pressure lowers to 2.5,p~ig).
c.
is incorrect because OW cooler LOCA L/O cannot be defeated if actual LOCA conditions exist.
d.
Is incorrect because emergency depressurization is not required.
NUREG-1 021, Revision 9
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s):
EOP-02-PCCP and the DWSIL (Attach if not previously -provided)
- graph, Proposed references to be provided to applicants during examination:
Provide EOP-02-PCCP and the DWSIL graph as reference Learning Objective:
LOI-CLS-LP-300-L*05g (As available)
Question Source:
Bank#
Modified Bank #
-New--
x-(Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content:
55.4.1 55.43 Question History:
Last NRC Exam N/A x
Comments:
From 01-37.8. Consistent with the meaning of "before," the drywell sprays should be actuated before the temperature reaches 300°F (drywell design temperature).
From SEP-10 IF:
CO: _
- a. Directed to defeat the drywell cooler LOCA lockout logic due to low reactor water level, AND CO: _
- b. Actual LOCA conditions do not exist in the Drywe 11-, AND CO: _
- c. RBCCW is operating and supplying the Drywell, CO: _
- d. THEN PERFORM Section 4 on page 7 of this procedure.
Basis for DWSIL A typical Drywell Spray Initiation Limit (Figure 5) is defined to be the highest drywell temperature at which initiation of drywell sprays will not result in an evaporative cooling pressure drop to below either
- 1. The drywell-to-torus chamber differential pressure capability or
- 2. the high drywell pressure scram setpoint (for actual Drywell Spray Initiation Limit curve, refer to EOP-01-UG).
This temperature i~ a function of drywell pressure, and the Limit is utilized to preclude NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 containm.ent failure or de-inertion following initiation of drywell sprays.
NUREG-1 021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examin?tion Outline" Cross-reference:
RO#54 Level Tier#
Group #
KIA #
Importance Rating
~-RO
.SRO 1
1 295030. EK1.01 3.8 (K&A Statement) Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Steam condensation Proposed Question:
During accident conditions, suppression pool level cannot be maintained above -6.5 ft. Which ONE of the following actions is directed by EOP-02-PCCP, Primary Containment Control Procedure, and the basis for that'action?
A.
Terrrifriatef HP*CI*operation"irrespectiveof-adequate-corecool*ing--te-'F>r-event~exceeding Pressure Suppression Pressure Limit.
B.
Terminate HPCI operatiorL irrespective of adequate core cooling to prevent exceeding Primary Containment Pressure Limit-A.
D.
Terminate HPCI. operation only if adequate core* cooling can be assured"s+nce co,nta.~r:l,rn*eRt..~integrity can-be maintained by venting containment.
C.
Terminate HPCI operatio~_only ifadequate core cooling can be assured, since core cooling--must take priority over containment integrity.
Proposed Answer:
B.
Terminate HPCI operation irrespective of adequate core cooling to prevent exceeding Primary Containment Pressure Limit-A.
Explanation:
Operation of the HPCI System with its exhaust discharge device not submerged will directly pressurize the primary containment because the downcomers are uncovered (The Pressure Suppression Pressure Limit was exceeded when suppression poollevelexceeded.*-
5.5 feet} This rise in pressure could exceed the Pressure Suppression Pressure Limit.
a.
is incorrect because the Pressure Suppression Pressure Limit has already been exceeded.
c.
is incorrect because securing saving the containment takes precedent over operating HPCI.
d.
is incorrect because primary containment venting is performed only, as necessary, to restore and then maintain pressure below the limit.
Technical Reference(s):
001-37.8, OEOP-02-PCCP (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Learning Objective:
LOI-CLS-LP-300-L*05g (As-available)
Question Source:
Bank#
Modified Bank #
New LOI SYSTEMS LOI-CLS-LP-300-L*013 002 (Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
From 001-37.8, Operation of the HPCI System with its exhaust discharge device not submerged will directly pressurize the suppression chamber. HPCI operation is therefore secured as required to precl~ude the occurrence of this condition. The consequences of not doing so may extend to failure of the primary containment from overpressurization, and thus HPCI must be secured irrespective of adequate core cooling concerns.
Venting of the primary containment is performed irrespective of the off-site radioactivity release rate that will occur, and defeating isolation interlocks if necessary, because the consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled radioactive release much greater than might otherwise occur. Note that primary containment venting is performed only, as necessary, to restore and then maintain pressure below the limit.
NUREG-1021, Revision 9
ES-401 SampleWritteri Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO#55 Level Tier#
Group #
KIA #
Importance Rating RO BRO 1
1 295031 EK1.02 3.8 (K&A Statement) Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL: Natural circulation: Plant-Specific.
Proposed Question:
Unit 1 has achieved cold shutdown when an inadvertent shutdown cooling isolation occurs. None of the shutdown cooling isolation valves can be re-opened. 'Nbich one of the following methods of decay heat removal J§_required for these conditions?
A.
Start a Recirculationpum-p-and us-eR*WCU**withth-e--regenerat-ive-heat--ex-e~aflgers bypassed.
B.
Raise and maintain reactor vessel water level to between 200" and 220" and monitor reactor coolant heatup.
C.
Ensure Main Steam line inboard and outboard isolation valves are open, then raise reactor vessel water level to >254" and raise CRD flow.
D.
Monitor reactor coolant heatup and, prior to exceeding 212°F, ensure inboard and outboard reactor vent valves are ope.n and raise CRD flow.
Proposed B.
Answer:
Raise and maintain reactor vessel water level to between 200" and 220" and monitor reactor coolant heatup.
Explanation:
Per AOP-15 During conditions in which there is no circulation,.the reactor.
vessel water level, should be maintained between 200" and 220", or as directed by the Shift Superintendent based on plant conditions, a.
is incorrect because the recirculation pumps would be shutdown if operating and there is no direction to restart them. AdditionaUy there is no direction to use RWCU without some form of feed system.
c.
is incorrect because this method would only be used if. reactor ve~sel coolant temperature cannot be maintained <212°F.
d.
is incorrect because these vent valves are closed prior to exceeding 212°F Technical Reference(s):
NUREG-1021, Revision 9 OAOP-15, CLS-LP-302-L (Attach if not previously provided)
ES-401 Sample Written "Examination Question Worksheet Form ES-401-5 Proposed references to"be proVided to applicants during examination:
Learning Objective:
CLS-LP-302-L, Obj 5b (As available)
Question Source:
Bank#
Modified Bank #
New x
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory orFLJn~dame-n-falkI16wleage Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 x
Comments:
From OAOP-15, During conditions in which there is no circulation, the reactor vessel water level, as read on 821-LI-R605A(B), should be maintained between 200" and 220", or as directed by the.Shift Superintendent based on plant conditions, until forced circulation is restored.
MONITOR reactor coolant heatup/cooldown in accordance with 1(2)PT-01.7 for any unexpected trends.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Exam-ih-atibh Outline Cross-reference:
RO#56 Level Tier #
Group #
KIA #
Importance Rating RO~---
---~----SRO 1
1 295037 EA2.06 4.0 (K&AStatement) Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor pressure.-
Proposed Question:
During an ATWS, EOP-01-LPC is being executed.
Plant conditions are:
--Reactor-power Suppr pool temp Drywell pressure RPV water level SRV# F013E RPV pressure SLC
___3%
95°F 1.7 psig
+85" Failed open 410 psig Injecting Which one of the following actions is required by LPC at this time?
A.
Raise reactor water level to >90 inches.
I
-B.
Prevent injection from any RHR and Core S-pray not required,~for injection.
C.
Open seven SRVs to emergency depressurize.
D.
Operate the condensate system to prevent uncontrolled injection.
Proposed
. Answer:
b.
Prevent injection from any RHR and Core Spray not required for injection.
Explanation:
During an ATWS with reactor pressure less than normal actions must be taken to prevent the uncontrolle.d injection of ECCS injection systems when a LOCA initiation signal is present. In the conditions above SLC is injecting and reactor power is 3olb, reactor pressure is 440 psig and an SRV is failed open OEOP-01-LPC requires preventing injection from RHR and Core Spray not required for adequate core cooling. This makes the correct answer "Prevent injection from any RHR and Core Spray not' req~ired for injection."
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form'ES-401-5 a.
Incorrect because level must be maintained below 90".
b.
-Cbr-recf~-
c.
. Incorrect because emergency depressurization is not require because all the conditions in Table 3 are met.
d.
Incorrect because the step for preventing RHR and CS spray injection comes before preventing uncontrolled injection for the condensate system.
Technical Reference(s):
001-37.5, OEOP-01-LPC (Attach if not previously provided)
Proposed references to be provided to applicants during examination:'
OEOP-01-LPC L-earn*I**n,g **O***b,iectl*ve".-*-
C_LS:L,P~3QO~_E_#J9~..._Gi.Y_e.n__ pl~!JJ_fQQcJ i!j_Q-'J.~_..~_O(:t th~
(As
'1 bl )
J
--- -aval a
.. e Level/Power Control Procedure, determine the required operator actions.
Question Source:
Bank#
Modified Bank #
New x
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
, 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO#57 Level Tier #
Group #
KIA #
Im.portance Rating RO
- SRO 1
1 295038 EK1.01 2.5 (K&A Statement) Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE:.Biological effects of radioisotope ingestion.
Proposed Question:
Which one of the following is the bases for the Technical Specification limit on reactor coolant specific iodine activity?
Limits the:
A.
off-site thyroid dose during a design basis loss of coolant accident.
B.
on-site thyroid dose during a design basis main steam line accident.
C.
off-site whole body dose during a design basis main steam line accident.
D.
on-site whole body dose during a design basis loss of coolant accident.
Proposed' Answer:
C.
off-s*ite whole body dose during a design basis main steam line accident. '
Explanation:
The reactor coolant specific iodine activity limit is based on limiting the total effective dose equivalent (TEDE) to less than 25 rem total. TEDE is based on a combination of both the internal and external dose. Although iodine is a threat to the thyroid it is the 'total dose (TEDE) that is the bases for the limit. The design basis.accident of concern for this limit is a main steam line break (MSLB).
a.
Incorrect because the total effective dose is basis and the design basis accident is a MSLB.
b.
Incorrect because the limit is based on off-site exposure (at the site boundary) and the design bases accident is a MSLB.
c.
Correct d.
incorrect because the limit is based on off-site exposure (at the site boundary) and the design bases accident is a MSLB.
Technical Reference(s):
TS Bases B 3.4.6 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Learning Objective: ____________________ (As available)
Question Source:
Bank#
Modified Bank #
New x
(Note changes or attach parent)
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge X----
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
TS BASES During circulation, the reactor coolant acquires radioactive materials due to release of fission products from fuel leaks into the reactor coolant and activation of corrosion products in the reactor coolant. These radioactive materials in the reactor coolant can plate out in the RCS, and, 'at times, an accumulation will break away to spike the normal level of radioactivity. The release of coolant during a Design Basis Accident (DBA) could send radioactive materials into the environment.
Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure that in the event of a release of any radioactive material to the environment during a DBA, radiation doses are maintained within the limits of 10 CFR 50.67 (Ref. 1).
This LCO contains iodine specific activity limits. The iodine isotopic activities per gram of reactor coolantare expressed in terms of a DOSE EQUIVALENT 1-131. The allowable levels are intended to limit the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> radiation dose to an individual at the site boundary to a small fraction of the 10 CFR 50.67 limit.
LCO The specific iodine activity is limited to S 0.2 JjCi/gm DOSE EQUIVALENT 1-131. This limit ensures the source term assumed in the safety analysis for the MSLB is not exceeded, so any release of radioactivity to the environment during an MSLB is less than a small fraction of the 10 CFR 50.67 limits.
10 CFR 50.67 The NRC may issue the amendment only if the applicant's analysis demonstrates with NUREG-1021, Revision 9
ES-401 reasonable assurance that:
Sample Written Examination Question Worksheet Form ES-401-5 (i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following th*e onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem)2 total effective dose equivalent (TEDE).
(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
NUREG-1 021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examiriclti6ri--'Outline Cross-reference:
RO#58 Level Tier #
Group #
KIA #
Importance Rating RO'-
"'-'-- -- SRO 1
1 600000 AK1.02 2.9 (K&A Statement) Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: Fire Fighting.
Proposed Question:
A report of a fire in the Unit 1 HPCI room has been received. The installed fire suppression system has initiated. Which one of the following is the type of installed system and what precautions are necessary?
A-.-
B.
C.
D.
C02"fit-e soppres"sion system;atmosphericmonitoringisrequired-priorto*ent-ry~
C02 fire suppression system; HPCI room ventilation must be started before the C02 system can be reset.
Wet pipe sprinkler system supplied by a deluge valve; notify radwaste to ensure they have adequate capacity for the water.
Dry pipe sprinkler system supplied by a deluge valve; notify radwaste to ensure they have adequate capacity for the water.
A.
Proposed Answer:
C02 fire suppression system; atmospheric monitoring is required prior to entry.
Explanation:
The High Pressure Coolant Injection (HPCI) pump room in each Reactor Building is protected by an automatic, total flooding, fixed carbon dioxide fire suppression system. Each system consists of nozzles attached to a piping network which is connected to the main and reserve C02 manifolds.
HPCI Room entry following C02 discharge without 'adequate atmospheric monitoring is a significant hazard to personnel.
a.
correct.
b.,
is incorrect because the C02 system must be reset before the ventilation can be restarted.
c.
is incorrect since the HPCI primary and reserve systems are both C02.
d.
is incorrect since the HPCI primary and reserve systems are both C02.
Technical Reference(s):
OOP-41 Section 8.54 & SD-41 (Attach if not previously provided)
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 t
Proposed references to be provided to applicants during examination:
Learning Objective:
Question Source:
CLS-LP-41, 12. Describe the sequence of events following these fire protection system initiations: a) HPCI (C02) initiation Bank#
Modified Bank #
(As available)
(Note changes or attach parent)
New New Question History:
LastNRCExam__
__N/A Question Cognitive Level:
Memory or Fundamental Knowledge X---
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Exa-mlnation-Outline Cross-reference:
RO#59 Level Tier#
Group #
KJA#
Importance Rating
- --RG-------~-S*R.O---.--
1 2
295010 AK2.05 3.7 (K&A Statement) Knowledge of the,tnterrelations between HIGH DRYWELL PRESSURE and the following: Drywell cooling and ventilation.
Proposed Question:
Unit Two (2) was operating at rated power. Both Reactor Feed Pumps tripped. HPCI and RCIC both failed. CRD flow has been maximized and SLC is injecting demin water to stabilize RPV level.
Current plant conditions:
What action is required to control containment parameters?
A.
RPV level RPV pressure Drywell pressure Dry'well temp Vent the drywell per OP-10.
+30" 950 psig 2.0 psig 152uF B.Vent the drywell per SEP-01.
C.
Spray the suppression pool per SEP-03.
D.
-Defeat drywell cooler LOCA lockout per SEP-10.
Proposed Answer:
D.
Defeat drywell cooler LOCA lockout per SEP-1 O.
Explanation:
High energy conditions do not exist in the drywell, drywell pressure/temp are elevated due to trip of drywell coolers wtth reactor water level below LL3, drywell coolers should be started per PCCP and SEP-10, LOCA logic can be defeated.
a.
is incorrect because venting per OP-1 0 cannot be performed above 1.7 psig drywell pressure b.
is incorrect venting per SEP-01 is not performed unless PCPL-A challenged c.
is incorrect because SEP-03 is not performed unless suppression chamber pressure is > 2.7 psig d.
correct NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet 001-37.8, OEOP-02-PCCP, (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
OEOP-02-PCCP - no entry conditions Learning Objective:
CLS-LP-300-L, 11. Given* the Primary Containment Control Procedure, which steps have been completed and plant parameters, determine the required operator actions.
(As available)
Question Source:
Question History:
Modified Bank #
New Last NRC Exam N/A (Note changes or attach parent)
.Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination--Outline Cross-reference:
RO#60 Level Tier #
Group #
KIA #
Importance Rating R-O---------------S-RO 1
2 295015 AK1.02 3.9 (K&A Statement) Knowledge of the operational implications of the following concepts as they apply to INCOMPLETE SCRAM: (CFR 41.8 TO 41.10) Cooldown effects on reactor power.
Proposed Question:
During an ATWS, EOP-01-LPC directs the operator that a cooldown can NOT be initiated prior to the Cold Shutdown Boron Weight 9CSBW) being injected into the reactor vessel.
--Which--ONE--of-thefoUowil1g.Jsthe-xeas-o.o--foI-this--re-quire.m_ent?
A.
In a partially borated core the reactivity effects of a cooldown are unpredictable.
B.
The cooldown will reduce steaming rates resulting in stagnant core flow which may lead to neutron flux oscillations.
C.
Initiating a cooldown while injecting boron is an uncontrolled reactivity manipulation and it will prevent the boron from being uniformly mixed.
D.
Cooldown is not allowed at this time to ensure that low pressure injection systems cannot inject into the vessel and add positive reactivity.
Proposed A.
In a partially borated core the reactivity effects of a cooldown are Answer:
unpredictable.
Explanation:
If any amount of boron less than the CSBW has been injected into the reactor vessel, cooldown is not permitted 'unless it can be determined that control rod insertion alone ensures the reactor will remain shut down under all conditions. The core reactivity response from cooldown in a partially borated core is unpredictable and subsequent steps may not prescribe the correct actions for such conditions if criticality were to occur.
b.
is incorrect since this action will not cause a reduction in core flow or flux oscillations.
c.
is incorrect since it does not prevent the boron from being uniformly mixed.
d.
is incorrect since EOP guidance directs the prevention of injection systems into the vessel and the consequent addition of positive reactivity.
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet 001-37.5, Level/Power Control (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
P-01-LPC - no entry con itions Learning Objective:
CLS-LP-300-E, 6. Explain the reason for lowering reactor (As available) water level while performing the Level/Power Control Procedure.
Question Source:
Bank#
Modified
.Sa_nk#
New New (Note changes or attach parent)
Question History:
Last NRC Exam N/A Question CognItive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 C.FR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO#61 Level Tier #
Group #
KIA #
2 295020 AK1.01 3.7 (K&A Statement) Knowledge of the operational implications of the following concepts as they apply to INADVERTENT CONTAINMENT ISOLATION: Loss of normal heat sink.
Proposed Question:
Following a Unit Two (2) Reactor scram, control rods have failed to fully insert. MSIVs have closed due to misoperation of the Mode Switch.
EOP-01-LPC, Level Power Control, is being executed. Prior to
-re~fchit'-g theste"plcfeqUa-lize ancfop-en--MSIVs,a--step-in--LPCasks "Is Main Condenser Available As A Heat Sink?"
For which of the following conditions would the question be required to be answered NG?
A.
Condenser vacuum is 0 inches.
B.
The Condensate system is not available.
C.
The Cire Water Intake Pumps have all tripped and are not available.
D.
Mechanical Vacuum Pumps cannot be op~rated.
Proposed Answer:
C.
The Circ Water Intake Pumps have all tripped Explanation: Equalizing and opening MSIVs authorizes low condenser vacuum to be bypassed.
Condenser vacuum can be re-established if either SJAEs or vacuum pumps are operable. With the MSIVs open the SJAE will be available. Per the EOP-UG step clarification, condensate does not have to be available, but circ water to the condenser must be available.
a.
is incorrect since Condenser vacuum can be re-established if either SJAEs or vacuum pumps are operable.
b.
is incorrect since Per the EOP-UG step clarification, condensate does not have to be available c.
Correct answer.
d.
is incorrect since Condenser vacuum can be re-established if either SJAEs or vacuum pumps are operable; the SJAE will be available"when the MSIVs are bypassed.
NUREG-1021, Revision 9-
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet 001-37.5, Level/Power Control (Attach if not previously--provided)
EOP-01-UG Propo~ed references to be provided to applicants during examination:
NONE Learning Objective:
_C_L_S_-_L_P-_3_0_0-_E_#_5 (As available)
Question Source:
Bank#
Modified Bank #
New EopBnk LOI-CLS-LP-300-E*005 001 (Note changes or attach
-parent)
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55.43 Comments:
Per 01-37.5 LPC; STEPS RC/P-12 through RC/P-19 (continued)
STEP BASES:
If the MSIVs are closed, it is prudent to attempt to reopen them in order to reject as much heat as possible to the main condenser. Reopening the MSIVs will limit the heatup of the Suppression Pool and minimize the challenges to Primary Containment.
The MSIVs are only reopened if the MSIVs are not required to be closed by OEOP-04-RRCP, Radioactivity Release Control Procedure (Le., no fuel failure indicated by an Abnormal Core Conditions and Core Damage Unusual Event EAL classification), boron injection is required, the main condenser is available as a heat sink, and there is no indication of a steam line break.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 ExaminatibnOutline Cross-reference:
RO#62 Level Tier#
Group #
KIA #
2 295022 AA1.01 3.1 (K&A Statement) Ability to operate and/or monitor the fol.lowing as they apply to LOSS OF CRD PUMPS: CRD hydraulic system.
Proposed Question:
Unit Two (2) is operating at 100°A, power when the operating CRD Pump trips on low suction pressure. The operator attempts to start the standby CRD Pump, which fails to start.
Which ONE of the following actions is required lAW OAOP~02?
A.
If a CRD pump cannot be returned to service within 20 minutes initiate a reactor scram.,
B.
Immediately scram th*e reactor if two or more HCU low pressure alarms are received and confirmed.
C.
If aCRD pump has not been restored within 20 minutes of the initial CR.D pump trip lower power to less than 26°k.
D.
Monitor CRD system parameters and continue attempts to start a CRD no immediate actions are required at this time.
Proposed Answer:
d.
Monitor CRD system parameters and continue attempts to start a CRD no immediate actions are required at this time.
Explanation: If there are no HCU low pressure alarms, there is no requirement to scram the reactor.
a.
is incorrect because the time limit is based on inoperable accumulators.
b.
is incorrect because there is 20 minutes allowed after the second HCU alarm to restore CRD system pressure.
c.
is incorrect because there is no requirement to lower power and APP A-05, 3-1 states reactor power should remain constant while the CRDHS is not operating.
Technical Reference(s):
NUREG-1021, Revision 9 OAOP-02, 2APP-A-05.
(Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be provided to applicants during examination:
None Learning Objective:
CLS-LP-302 B; 4. Given plant conditions and OAOP-2.0, determine the required supplementary actions.
(As available)
Question Source:
Bank#
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X---
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
From*OAOP-02 IF the operating CRD Pump has failed, THEN RESTART the CRD Hydraulic System following loss of a CRD Pump in accordance with1 (2)OP-08.
b.
IF reactor pressure is greater than or equal to 950 psig, AND two or more HCU low pressure alarms (A-O? 6-1) are received (confirmed by amber light on Full Core Display), THEN* ENSURE CRD pressure is restored to greater than or equal to 940 psig within 20 minutes.
- 6. MONITOR the following CRD System parameters for possible system leakage or flow control valve failures:
a.
CRD Drive Water Pressure, C11(C12)-PDI-R602.
b.
CRD Cooling Water Pressure, C11(C12)-PDI-R603.
c.
CRD Drive Temperature, C11(C12)-TR-R018.
d.
CRD Charging Water Header Pressure, C11(C12)-PI-R601.
From 2APP-A05 Reactor power and generator load should be maintained constant during the time period that the CRDHS is not operating (no CRD pumpsTunning).
NUREG-1021, Revision 9
ES-401 Original question was N/A NUREG-1021, Revision 9 Sample. Written Examination Question Worksheet Form ES-401-5
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 ExamiffatiorrOutline-Cross-reference:
RO#63 Level Tier #
Group #
KJA#
Importance Rating
-SRO 1
2 295029 EA2.02 3.5 (K&A Statement) Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: Reactor pressure.
Proposed Question:
During an accident, plant conditions are:
Primary Containment level Drywell pressure Dryyve~l_p~~ssure Per EOP-02-PCCP, 38 feet 74 psig (CAC-PI-4176) 70 psig (CAC-PI-1230)
A vent the primary containment within ODCM release rate limits.
B.
vent the primary containment irrespective of off-site release rates.
C.
establish drywell sprays with RHR pumps NOT required to assure adequate core cooling.
D.
terminate injection into the reactor from all sources external to the primary containment irrespective of adequate core cooling.
Proposed Answer:
B.
vent the primary containment irrespective of off-site release rates to maintain the ability to operate SRVs and containment vent valves.
Explanation:
PCPL-A has been reached, PCCP requires venting irrespective of rad release.
a.
is incorrect since the containment must be vented irrespective of release rates.
c.
is incorrect drywell sprays must be secured when primary containment level exceeds 21".
d.
is incorrect since sources required for core cooling should NOT be secured until primary containment water level reaches 68.5 feet.
Technical Reference(s):
NUREG-1021, Revision 9 001-37.5, Level/Power Control EOP-01-UG (Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be p~ovided to applicants during examination:
OEOP-02-PCCP, PCPL-A curve - no EOP entry conditions CLS-LP-300-L # 5h Learning Objective:
(As available)
Question Source:
Bank#
Modified Bank #
New LOI-CLS-LP-300-L*04G (Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X----
10 CFR Part 55 Content:
55.41 55~43 Comments:
Per 01-37.8 PCCP The'directions to vent "before drywell pressure reaches PCPL-A" allows, but does not require, venting at significantly lower pressures. Early or extended venting can permit primary containment pressure reductions before significant fuel damage occurs, thereby increasing the capacity of the contain'ment to retain fission products and reducing the radioactivity released to
,the environment. If the primary containment has failed, venting may also reduce the offsite dose by directing fission products through an elevated release point.
At PCPL-A venting of the primary containment is performed irrespective of the off-site radioactivity release rate that will occur, and defeating isolation interlocks if necessary, because the consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled radioactive release much greater than might otherwise occur. Note that primary containment venting is performed only, as necessary, to restore and then maintain pressure below the limit.
Terminating drywell sprays prior to +21 inches minimizes the potential for containment damage due to submerging the vacuum breakers.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 ExaniTnati6nDutiine Cross-reference:
RO#64 Level Tier #
Group #
KIA #
Importance Rating RO
SRO 1
2 295033 EK3.04 4.0 (K&A Statement) Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Personnel evacuation.
Proposed Question:
Emergency conditions on Unit Two (2) require performance of EOP AEDP, Alternate Emerge~cy Depressurization Procedure. '
Which ONE of the following describes the proper method to meet the evacuation requirements---ofEOP-Ot..AEDP?
A.
Sound the Unit 2 Turbine B.uilding evacuation alarm only and announce the evacuatien.
Direct all operations personnel to report to the OSC.
B.
Sound the Unit 1 and Unit 2 Turbine Building evacuation alarms and--aftflotmee-tAe evacuation. Direct all operations personnel to report to the OSC.
C.
Sound the Unit 2 Turbine Building evacuation alarm only and---announ~tiQn.
Request SCO to notify the TSC that the Turbine Building is being evacuated.
D.
Sound the Unit 1 and Unit 2 Turbine Building evacuation alarms aM al1110unce tAe evaCt1at;O'~. Request SCO to notify the tsc that the-Turbine Building is being evacuated.
Proposed Answer:
D.
Sound the Unit 1 and Unit 2 Turbine Building evacuation alarms and announce the evacuation. Request-SCO to notify the TSC that the Turbine Building is being evacuated Explanation:
There is a potential that the radiation levels in the turbine building will change significantly during the rapid depressurization since it is being performed irrespective of offsite radioactivity release rates. The method of evacuation selected was to sound the Unit 1 and Unit 2 Turbine Building evacuation alarms. Shielding does not separate the two turbine buildings so that a significant radiation level in one may affect the other. The TSC is notified of the evacuation so that they may effectively evaluate any missions that are in progress in the buildings.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 a.
is incorrect because both units must be evacuated and the OSC is not immediately acthiafed:- 'It will be activated later by the TSC.
b.
is incorrect because the OSC is not immediately activated. It will be activated later by the TSC.
c.
is incorrect because both units must be evacuated.
d.
correct EOP-01-AEDP Technical Reference(s):
(Attach if not previously provided)
Proposed references to be provided to applicants during examination:
LearningObjective-: _C__LS_.._L_~-_--3_0_0- L_#_5__h
{As available)
Question Source:
Bank#
Modified Bank #
New LOI-CLS-LP-300-H*005 (Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
Per EOP-01-AEDP;
- 1. EVACUATE the Unit 1 and 2 Turbine Buildings using the following actions:
- a. SOUND the Unit 1 and Unit 2 Turbine Building evacuation alarms and announce the evacuation.
- b. REQUEST the sca to notify the TSC that the Turbine Building is being evacuated due to potential high radiation conditions during the alternate emergency depressurization.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examiriatibfl-Outline Cross-reference:
RO#65 Level Tier #
Group #
KIA #
Importance Rating
-RG--*-----
--- - -. -----SRO 1
2 295035* EK3.01 2.8 (K&A Statement) Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Blow-out panel operation Proposed Question:
E>'ete-rmining that a positive pressure exists in the reactor Quildingwhen Seco-ndary Containment is required is an entry condition for OEOP secp, Secondary Containment Control Procedure. Which one of the following is the reason for this requirement?
Thi~ entry condition:
A.
anticipates the effects of a loss of SBGT B.
anticipates the automatic or manual isolation of Reactor Building HVAC.
C.
directs actions to control reactor building pressure after tAe-release of the secondary eontainmeRt blowout panels.
D.
directs actions to control reactor building pressure before the release of the secondary containment blowout panels.
Proposed Answer:
D.
directs actions to control reactor building pressure before the release of the secondary containment blowout panels.
NUREG-1021, Revision 9
ES-401 Sampl.e Written Examination Question Worksheet Form ES-401-5 Exp*l~rnation-:----This'directs the operator to*ensure the,properoperation--of-tbeR-BJ::I1{ACJ:o_attempt _
to prevent a loss of secondary containment prior to the building going positive and blowing out the panels.
STEP SCCP-13 STEP BASES:
The Secondary Containment Control Procedure (SCCP) is structured along three parallel action paths. Actions taken to control one parameter may directly affect control of the other parameters and thus all three sections are executed concurrently. Current values and trends of parameters and the status of plant systems and equipment during a transient dictate the order and priority with which specific actions are executed. This approach precludes the assignment of any particular priority to the execution of any of the individual sections.
. ---, '---it~r~lrle~¢r~ht~~~-tl$0a-Ry--atr-eGt4Y~s-S0Gia-tetl~pel=a-t6tia-ctio.rjssSirii~c-e-onJ¥_-a_-smaJl operating margin exists between its normal operating value and the pressure at which the Reactor Building (secondary containment) blowout panels release. Once secondary containment integrity has been lost by the blowout panels functioning to relieve high reactor building pressure, no mechanism is available through which reactor
~~iI~!~.~>e~~~~.~~~>~~~~ny longer be effectiveIy controlied.~!>~:fflt*~~~il~:~I~:;:~~~f~~fte4
~~s4per~i~~~~~I~~g~~i~~~.~ffi~~n
\\i~H~}~p~qq~at2Jr~. This adequately addresses reactor building pressure control to the e'xtent that procedural instructions are appropriate.
A. Incorrect because this ?tep is to provide immediate attention to prevent the loss of Secondary Contain'ment. This action will insure that reactor building ventilation (not SBGT) is operating correctly to control the reactor building dIp (step SCCP-11).
B. Incorrect since this condition is designed to prevent the loss of secondary containment and directs determining if an isolation has occurred and if so restoring reactor building ventilation.
C. Incorrect since there are no steps in SCCP to control reactor building pressure.
Technical Reference(s):
001-37.9 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective:
CLS-LP-300-M, 3 Given plant conditions, determine if the (As.available)
Secondary Containment Control Procedure should be entered.
NUREG-1021, Revision 9
ES-401 Question Source:
Sample Written Examination Question Worksheet Bank #
Modified Bank #
Form ES-401-5 (Note changes or attach parent)
New x
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or 'Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
Per 001-37.9 Although reactor building pressure is an entry condition to this procedure, it is not a parameter which has any directly associated operator actions since only a small operating margin exists between its normal operating value and the pressure at which the Reactor Building (secondary containment) blowout panels release. Once secondary containment integrity has been lost by the blowout panels functioning to relieve high reactor building pressure, no mechanism is available through which reactor building pressure can any longer be effectively controlled. When available and while radiation levels permit, direction to operate the Reactor Building HVAC is specified in the procedure. This adequately addresses reactor building pressure control to the extent that procedural instructions are appropriate.
NUREG-1021, Revision 9
ES-40t Sample Written Examination Question Worksheet Form ES-401-5 Exarrrin-ation Outline Cross-reference:
RO#66 Level Tier #
Group #
KIA #
Importance Rating RO 3
1 2.1.32 3.4
'--SRO (K&A Statement) Ability to explain and apply all system limits and precautions.
Proposed Question:
During rated power operation, a loss of drywell cooling requires entry into AOP-14.0 due to rising drywell temperature and pressure.
The crew is preparing to vent primary containment per OP-10.
OP-10 prohibits venting the drywell and the. suppression chamber simultaneously because this..actioncould resuilln:
A.
bypassing the pressure suppression -function,
B.
operation of torus to drywell vacuum breaker C.
operation of reactor building to torus vacuum breakers D.
excessive release of radioactivity through the main stack Proposed answer:
Explanation:
A.
bypassing the pressure suppression function Per OP-1 0, torus and drywell cannot be vented at the same time in Modes 1, 2 or 3. per 01-37.1, this could result in bypassing pressure suppression function a.
is correct b.
is incorrect because this lineup equalizes pressure between the drywell and the suppression pool free air space since the vacuum breakers operate on a dIp between the spaces this would bypass them, not open them.
c.
is incorrect because these vacuum breakers prevent drawing a negative pressure in the suppression pool. Cross connecting the drywell and the suppression pool free air space !~ill not cause a negative pressure in the suppression pool.
v d.
is incorrect because this action would only be done when LOCA conditions do NOT exist NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401-5 Question Worksheet OP-10, AOP-14, 001-37.1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
N/A Learning Objective:
CLS-LP-302-D, 2. Given plant conditions and AOP-14.0, determine the required supplementary actions.
(As available)
Question Source:
Bank#
Modified Bank #
-New LOI-CLS-LP-302-D*002 005 (Note changes or attach parent)
,Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Question History:
Last NRC Exam N/A x
Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Exaniinafidn-OutlineCross..;reference:
RO#67 Level Tier#
Group #
KIA #
Importance Rating
RO--
3 1
2.1.1 3.7
~----------SRG- -
(K&A Statement) Knowledge of conduct of operations requirements.
Proposed.Que*stion:
Units One (1) and Two (2) are both performing a startup following a forced shutdown for hurricane conditions.
Unit One (1) is at 25% power with Turbine roll in progress.
Unit Two (2) is at 800 psig performing Reactor heatup and pressurization.
The MINIMUM number of Control Operators required by 01-01.02 to be in the Main Control Roo~(as defined by 01-01.01) is:
A.
2 B.
C.
3 4
D.
5 C.
4 Proposed Answer:
Explanation:
Per 01-01.02, Section 5.1, Unit startup (or scheduled shutdown) requires 2 COs in the control room (as a minimum) that's two per unit for a total of 4 operators.
a.
is incorrect because both units are involved in,a startups with neither turbine synchronized to the grid.
b.
is incorrect because both units are involved in a startups with neither turbine synchronized to the grid.
d.
is incorrect because only four operators (two per unit are required) by 001-01.02 Proposed references to be provided to applicants during examination:
N/A Technical Reference(s):
001-01.02, Section 5.1.1.4.
(Attach if not previously provided)
NUREG-1021, Revision 9
ES~401 Sample Written Examination Question Worksheet Learning Objective:
LOI-CLS-LP-201-D, 001 Form ES-401-5 (As available)
Question Source:
Bank#
Modified Bank #
New LOI-Admin LOI-CLS-LP-201-D*001 (Note changes or attach parent)
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge C0r1'lp~~h_ension or Analysis
_X 10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
Sample Written Examination Question Worksheet ES-401 Examination--Outline Cross-reference:
RO#68 Level Tier#
Group #
KIA #
Importance Rating
- RO 3
1 2.2.34 2.8 Form ES-401-5
~---SRO (K&A Statement) Knowledge of the process for determining the internal and external effects on core reactivity.
Proposed Question:
Unit Two is operating at end of Cycle 16 with coast-down in progress. GP-13 has been implemented by bypassing feedwater flow around high pressure feedwater heaters 1Q achieve an 84°F
.equivalent reduction in final feedwater temperature.
Which ONE of the following describes the effects on Unit Two (2) operation?
A.
B.
During reduced feedwater heating:
the changes in core inlet sub-cooling will make the core flux more top peaked.
a trip of a single recirculation pump will increase the margins to thermal limits.
C.
a trip of a single recirculation pu~p will increase the possibility.of aA-1'nstability event.
D.
The changes in core inlet subcooling will require adjustments to the core thermal limits.
Proposed C.
a trip of a single recirculation pump will increase the possibility of an Answer:
instability event.
Explanation:
3.10 WHEN operating near regions of possible thermal hydraulic instability (when operating at reduced flow such as with a single recirculation pump), reduced feedwater heating
,increases the possibility of an instability event.
a.
is incorrect because the reduced feedwater heating increases core inlet subcooling which shifts the flux shape lower in the core.
b.
is incorrect because a trip of a recirculation pump will d.
is incorrect because although single loop operation required adjusting recirculation flWp there are no technical specifi'cations for operating with reduced feedwater heating.
Technical Reference(s):
NUREG-1021, Revision 9 GP-13 (Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Propose-a references to be provided to applicants during examination:
--N/A Learning Objective: ____________________ (As available)
Question Source:
Bank#
Modified Bank #
New New (Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge
. -Co"mpr-ehel1sion or AnalysIs x-_*_-
10 CFR Part 55 Content:
43.6 Comments:
NUREG-1021, Revision 9
Sample Written Examination Question Worksheet ES-401 Exa-mination-Outline Cross-'reference:
RO#69 Level Tier #
Group #
KIA #
Importance Rating
--RG----
3 2
2.2.3 3.1 Form ES-401-5
SRO, (K&A Statement) Knowledge of the design! procedural! and operational differences between units.
Proposed Question:
Which ONE of the following is the difference between the Condensate and Feedwater systems between Unit 1 and Unit 2?
A.
Power operation up to 75°A> with a single feedwater pump is permitted on Unit 1, verses 69% on Unit 2 because'Unit 1 has larger feedwater pumps.
B.
Power operation up to 75% with a single feedwater pump is permitted on 'Unit 2, verses
~~% on Unit 1 because Unit 2, has larg'er feedwater pumps.
D.
, To manually start a condensate pump on Unit 2 the Mode Switch must be in AUTO and the discharge valve must be closed. On Unit 1 the mode switch must be in MANUAL or the discharge valve closed.
To manually start a condensate pump on Unit 1 the Mode Switch must be in AU,TO or the discharge valve must be closed. On Unit 2 the mode switch must be in MANUAL and the discharge valve closed.
Power operation up to 75°A> with a single feedwater pump is permitted on Unit 1, verses 69% on Unit 2 because Unit 1 has larger feedwater pumps.
Proposed A.
Answer:
Explanation:
GP-05, 3.21, Reactor power is limited to 75°A> (Unit 1), 69% (Unit 2) with a single RFP in service.
a.
is correct b.
is incorrect because Unit 1 can operate at a higher power because it has larger feedwater pumps c.
is incorrect,because the manually start a condensate pump on Unit 2 the mode switch must be AUTO,,0the discharge valve is closed. The Unit 1 mode switch must be in MANUAL and the discharge valve closed.
d.
is incorrect because the manually start a condensate pump on Unit 1 mode switch must be in MANUAL and the discharge valve closed. The Unit 2 the mode switch must be AUTO or the discharge valve is closed.
NUREG-1021, Revision 9
ES-401 Sample. Written Examination Form ES-401-5 Question Worksheet Technical Reference(s):.
GP-04 04 05, SO-32 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective:
Question Source:
CLS-LP-32, 4. List the automatic starting permissives for:
(As available)
- a. Condensate Pumps Bank #
Modified Bank #
New X
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X---
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
From GP-05, 3.21 Reactor power is limited to 75% (Unit 1), 69°A> (Unit 2) with a single RFP in service.
SO-32 and GP-05, The reactor feed pumps (RFP) receive their suction from the #3 FWHs and the HOPs. Each RFP is a horizontal, single-stage, centrifugal pump operating at a flow rate of
-13,877 Unit 1 a'nd 13,000 Unit 2 gpm at full power.
NUREG-1021, Revision 9
ES-401 Sample Written Examinatio.n Question Worksheet Form ES-401-5 ExaminatienOutlineCross-reference:
RO#70 Level Tier #
Group #
KIA #
Importance Rating
.RO__. _ ---.-- __.
.S.RO 3
3 2.3.10 2.9 (K&A Statement) Ability to perform procedures to reduce excessive levels of radiation and
. guard against personnel exposure.
Proposed Question:
Unit One is performing a plant shutdown per GP-05 to meet Technical Specificatron requirements due to excessive unidentified leakage. The drywell will be entered shortly to identify the leakage source.
TIPtracsswere.completedJustpriof.to.tbeshutdown.J.eq_uirement.
TIP probes are currently at the index position with the Mode switches in Manual.
Which ONE of the following describes the requirement.s of 001-01.03 regarding the TIPs for drywell entry?
TIPs must be withdrawn to the in shield position with a clearance on each ball valve closed and each Mode switch in Off.
A.
TIPs may be placed under clearance in their present configuration.
B.
TIPs may be left at the index position but each Mode switch must be placed under clearance in the Off position.
C.
TIPs must be withdrawn to the in shield position with a clearance on each ball valve closed and each Mode switch in Off.
D.
TIPs must be withdrawn to the in shield position with a clearance on each ball valve closed and each Mode switch in Manual.
Proposed C.
Answer:
Explanation:
From 01-01.03, Attachment 11:
The TIPs shall be in the stored position (in shield) in the TIP Room, and a clearance placed on each TIP ball valve (Closed) and each TIP machine AUTO-MANUAL mode switch (OFF).
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 a.
is incorrect because the must be withdrawn to the in shield position to prevent the overexposure of anyone inthelower~levelsofthe drywellintheevent-a-TI.8--leaves-the cask.
There must be a clearance on each ball valve closed and each Mode switch in Off.
b.
is incorrect because in the index position they are out of the cask the indexers are located in the overhead of the lower level of the drywell. The TIPs must be withdrawn to the in shield position to prevent the overexposure of anyone in the lower levels of the drywell. There must be a clearance on each ball valve closed and each Mode switch in Off.
c.
is correct d.
is incorrect because placing the mode switch in manual does not stop the movement of the TIP, it could still be manually initiated.
Technical Reference(s):
01-01.03, Attachment 11, SD-.
(Attach if not previously provided) 9.5 Proposed references to be provided to applicants during examination:
Learning Objective: ____________________ (As available)
Question Source:
Bank#
Modified Bank #
New S215001 GEN 2.3.10 (Note changes.
or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 43.4 Comments:
From NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examinatio-n Outline Cross-reference:
RO#71 Level Tier #
Group #
KIA #
Importance Rating RO 3
1 2.3.1 2.6
-*SRO During, emergency conditions, it has been determined that exposures in excess of 10CFR20 limits may be required for protection of valuable property.
(K&A Statement) Knowledge of 10 CFR: 20 and related facility radiation control requirements Proposed Question:
In accordance with PEP-03.7.6, the Emergency Worker Dose Limit is:
A.
10 Rem TEDE, 30 Rem to extremities.
B.
10 Rem TEDE, 100 Rem to extremities.
C.
25 REM TEDE, 75 Rem to extremities.
25 Rem TEDE, 250 Rem to extremities.
Proposed B.
10 Rem TEDE, 100 Rem to extremities.
Answer:
Explanation:
10 REM limit for protection of valuable property with 10 times TEDE limit for extremities Uust like normal federal limits of 5 and 50), per PEP-03.7.6 Attachment 2 a.
is incorrect because the extemities limit is 10 times the TEDE limit.
b.
correct c.
is incorrect because the limit for property is 10 REM.
d.
is incorrect because the limit for property is 10 REM.
Technical Reference(s):
PEP-03.7.6 Attachment 2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
LOI-CLS-LP-1 02-A*11 B Learning Objective:
(As available)
NUREG-1021, Revision 9
Sample Written Examination Question Worksheet ES-401 Question Source:
Bank#
Modified Bank #
New LOI-CLS-LP-1 02-A*11 B Form ES-401-5 (NoteChanges or attach parent)
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge X---
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 X
Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examin-ation-0utline Cross-reference:
RO#72 Level Tier #
Group #
KIA #
Importance Rating RO-------------------------SR0 3
4 2.4.27 3.0 (K&A Statement) Knowledge of fire in the plant procedure.
Proposed Question:
While on watch in the Control Room as the Unit One (1) Control Operator you are informed of a class C fire in a cabinet near the center of the Unit One (1) Breezeway. According to PFP-13, General Fire Plan, which one of the following actions is required?
Sound the Fire Alarm and make the appropriate announcement 1nclud-ing-:--
H._---._-------** _.*
A.
that the use of the PA and radio is restricted to fire communications, then monitor the RTGB, particularly the feedwater system.
B.
that the use of the PA and radio is restricted to fire corpmunications, then monitor the RTGB, particularly the standby liquid control system.
C.
that the use of the PA system is restricted and all communications must be over radios, then monitor the RTGB, particularly the feedwater system.
D.
that the use of the PA system is restricted and all communications must be over radios, then monitor the RTGB, particularly the standby liquid control system.
Proposed A.
. that the use of the PA and radio is restricted, then monitor the RTGB, Answer:
particularly the feedwater system.
Explanation:
OPFP-013 b.
is incorrect because there are no controls or wiring related to the standby liquid control system in the breezeway.
c.
is incorrect because the use of radios is restricted during a fire.
d.
is incorrect because the use of radios is restricted during a fire and there are no controls or wiring related to the standby liquid control system in the breezeway.
Technical Reference(s):
NUREG-1021, Revision 9 OPFP-13, Attachment 1.
(Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be provided to applicants during examination:
None
_Learning Objective:
_L_O_I_-C_L_S_-L_P_-_30_6_-A_*_O_06 (As available)
Question Source:
Bank#
Modified Bank #
New-New (Note changes or attach parent)
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis
)(-- -
10 CFR Part 55 Content:
55.41 55.43 Comments:
NUREG-1021, Revision 9
ES~401 Sample Written Examination Question Worksheet Form" ES-401-5 Examination Outline Cross-reference:
RO#73 Level Tier #
Group #
KIA #
Importance Rating
-----R-O---
3 4
2.4.21 3.7
- SRO (K&A Statement) Knowledge of the parameters and logic used"to assess the status of safety functions including:
1.
Reactivity control 2.
Core cooling and heat removal 3.
Reactor coolant system integrity 4.
Containment conditions 5.
Radioactivity release control.
Proposed Question:
UriifTwo--(2) **C-6nlroIO-~j"eratbr-has-in-serted-a-manua-1reactor scram.
The plant status matrix scram status box in the lower right hand corner of the ERFIS screen has changed from "NO SCRAM" (in green) to "SCRAM RODS" (in red).
This ERFIS screen indication means that a SCRAM signal has:
A.
oCg[I~d and at least one control rod has NOT fully inserted.
B.
oCCYffed and all control rods are inserted to beyond position 00.
C.
NOT occurred and all control rods ~re inserted beyond position 00.
D.
NOT occurred but the scram valves are OPEN and all control rods are full-in.
Proposed Answer:
A.
occurred and at least one control rod has NOT fully inserted.
'Explanation:
Alarm SCRAM RODS Red indicates a Scram signal is present and control rod insertion time limit met, however any rod not past shutdown position b.
is incorrect because at least one control rod is not fully inserted.
c.
is incorrect because a scram signal has occurred and at least one control rod is not fully inserted.
d.
is incorrect because a scram signal has occurred and at least one control rod is not fully inserted.
01-60 Technical Reference(s):
NUREG-1021, Revision 9 (Attach if not previously provided)
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Proposed references to be provided to applicants during examination~:~
Learning Objective: ____________________ (As available)
Question Source:
Bank#
Modified Bank #
New LOI-CLS-LP-060-A*02E 002 (Note changes or attach parent)
Question History:
Last NRC Exam Questior1Cognitive Level:
Memory o~r Fundamental Knowledge
- X---
Comprehension or Analysis 10 CFR Part 55 Content:
Comments:
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:
RO#74 Le-vel Tier #
Group #
KIA #
Importance Rating
RQ--- -- -----------------SR0 3
4 2.4.23 2.8 (K&A Statement) Knowledge of the bases for prioritizing emergency proce'dure implementation during emergency operations.
Proposed Question:
After entering the Reactor Scram Procedure, 1(2)EOP-01-RSP, the crew has progressed to inserting the nuclear instruments. While executing this step it is determined that rea-cfor~-waterlevel can NOT be maintained above 170".
. WMicMoneoftMe-fellewing-is-required-by-this-*stepand-wt1at-~sthe bases?
A.
MAY return to "Can Reactor Water Level be Restored or Maintained at 170" for record keeping.
It is optional after the crew enters 1(2)EOP-01-RVCP since RVCP's entry conditions are the same.
B.-
MUST return to "Can Reactor Water Level be Restored or Maintained at 170" and answer the question NO and execute that path to insure other procedures are entered to correct the water level problem.
C.
MUST return to "Can Reactor Water Level be Restored or Maintained at 170" and answer the question NO. This insures entry into 1(2)EOP-01-RVCP to restore level. If RVCP was previously entered there is no need for re-entry.
D.
MAY return to "Can Reactor Water Level be Restored or Maintained at 170" and answer the question NO based upon the conditions that exist at that time. This provides a guide to significant parameters but allow flexibility in determining EOP actions.
Proposed Answer:
b.
MUST return to "Can Reactor Water Level be Restored or Maintained at 170" and answer the question NO and execute that path td.jhsure other procedures are entered to correct the water level problem.
NUREG-1021, Revision 9
ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Explanation:
Per 001-37.3 and OEOP-01-UG, STE P--Q~-4 BASES -If reactor waterlev-elcann_otb_e__m_ai ntaloe-'i~_b~_Ye-+/-_170 jnCbJ~§L ~_ggJtion~1 level control measures must be taken. These additional measures will be taken once the operator enters the Reactor Vessel Control Procedure or Level/Power Control procedure.
The purpose of the critical step is to allow the operator to reassess limiting parameters should they degrade or change during the execution of the procedure. While reassessing the parameter, the EOP is re-entered to direct the operator to a different mitigation strategy or to transfer the control of that parameter to another portion of the EOP. If the status of a critical step changes, the operator shall return to that step and take the action required.
a.
is incorrect because the step must be re-entered, where the entry conditions are similar they are NOT the same.
b.
is correct c.
is incorrect because when step 14 is answered no RVCP must be re-entered.
d.
is incorrect because the step must be re-entered.
Technical Reference(s):
001-37.3 and OEOP-01-UG (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
2EOP-01-RSP Learning Objective:
CLS-LP-300-B, 5. Given plant conditions and the EOPs, determine the correct operator response for the following conditions:
a) A Critical Step changes after it has been assessed.
(As available)
Question Source:
Bank#
Modified Bank #
New x
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
55,.41 55.43 NUREG-1021, Revision 9
ES-401 Comments:
From---------------
NUREG-1021, Revision 9 Sample Written Examination Question Worksheet Form ES-401-5
ES-401 Sample Written Examination Question Worksheet Form ES-401-S
- Examination*Outline.Cross-reference:
RO#7S Lev.el Tier #
Group #
KIA #
...-----~Q__..._...
S_RO 3
4 2.4.1S*
Importance Rating 3.0 (K&A Statement) Knowledge of communications procedures associated with EOP implementation.
proposed Qu.estion:
During emergency operations you are required to make emergency notifications to State and County agencies using the Selective Signaling System.
How is this system operated?
Lift the handset, determine you have a dial tone then dial the access number, pause for a "beep" then dial the designated number.
A.
Lift the handset, determine you have a dial tone then dial the access number, pause for a "beep" then dial the designated number.
B.
Lift the handset;.verify there is no dial tone, dial the three digit location number, when you hear a "beep" dial the designated number.
C.
Press and hold the MONITOR button and listen for voice traffic. When voice traffic is clear press and h~ld the PTT bar, then announce your location and state your message.
D.
Verify the Power Supply light is lit, the Volume knob is set on "I" or "II" and the "Mode" is correct. Press and hold the PTT bar, announce our call sig.n state the intended parties and state your message.
Proposed A.
Answer:
Explanation:
OPEP-3.1.3, Attachment 1 contains instructions for using this dial up system where after verifying there is a dial tone the access number (10) is dialed a "beep" verified then the appropriate number dialed.
a.
is correct b.
is incorrect because the Selective Signaling system has a dial tone and uses a two digit access number, additionally there is no "beep" in this system. This distracter is based on the Decision line.
c.
is incorrect because this distracter is based on the VHF radio operating instruction.
d.
is incorrect because this distracter is based on the VHF radio TSC Base Station operating instruction.
NUREG-1021, Revision 9
ES-401 Technical Reference(s):
Sample Written Examination Form ES-401~5 Question Worksheet OPEP-03.1.3, Attachment 1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective:
Question Source:
CLS-LP-48, 5. Describe the use of each of the following types of communications equipment:
a) Two-way Radio b) Sound-powered Phone c) Rolm Telephone d) Public Address Station e) Emergency Telephone
-Bank#
Modified Bank #
(As available)
(Note changes or attach parent)
New x*
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge
_X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 Comments:
From OPEP-03.1.3 ATTACHMENT 1 Page 1 of 1 Use of Selective Signaling 1.0 To Contact Warning Points 1.1 Lift handset of selective signaling phone and listen for dial tone.
1.2 Dial 10, pause for a beep, then Dial 22.
2.0 To Contact Emergency Operations Centers 2.1 Lift handset of selective signaling phone and listen for dial tone.
2.2 Dial 10, pause for a beep, then Dial 33.
NUREG-1021, Revision 9