ML080720496
| ML080720496 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/08/2008 |
| From: | Division of Reactor Safety II |
| To: | |
| References | |
| 50-327/08-301, 50-328/08-301 | |
| Download: ML080720496 (106) | |
See also: IR 05000327/2008301
Text
Final Submittal
(Blue Paper)
FINAL SRO
WRITTEN EXAMINATION
AND REFERENCES
SEquDY/lH
0{ 00 8-dtJj
SRO WRITTEN EXAM REFERENCE
LISTING
1. ECA-1.1, Curve 9
2.
0-SO-78-1, Spent Fuel Pit Coolant System, Appendix C
3.
Technical Specification 3.8.2.3
4.
EPIP -1 Page 9 and 10.
5.
Steam Tables
1/2008 Sequoyah SRO NRC Exam
2/8/2008
1
007 EA1.08 1
A
2
008 AA2.29 2
D
3
009 EA2.24 3
C
4
011 EK3.13 4
D
5
015/017 AK2.08 5
D
6
022 AA2.03 6
B
7
025 AK2.05 7
D
8
026 AK3.03 8
D
9
027 AAl.02 9
B
10
038 G2.1.3 10
A
11
040 AKl.03 11
C
12
055 G 2.4.18 12
B
13
056 AKl.03 13
B
14
058 AKl.01 14
C
15
062 AAl.01 15
C
16
W/E04 G2.4.28 16
B
17
W/E05 EK3.1 17
C
18
W/E11 EK2.2 18
A
19
001 AK2.05 19
C
20
033 G2.4.21 20
C
21
036 AKl.01 21
A
(
22
060 AA1.02 22
C
{
13
069 G2.4.45 23
B
24
074 EK3.07 24
B
25
076 AK 2.01 25
C
26
W/E09 EA2.1 26
D
27
W/E14 EA2.2 27
D
28
003 K3.04 28
A
29
003 G2.3.10 29
D
30
004 K5.26 30
B
31
004 A3.08 31
C
32
005 A4.03 32
D
33
006 Kl.02 33
D
34
007 K3.01 34
B
35
007 G2.1.1 35
D
36
008 K4.07 36
A
37
010 K6.03 37
D
38
012 K4.04 38
B
39
013 K2.01 39
B
40
022 A2.04 40
C
41
025 K5.02 41
C
42
026 K4.07 42
C
43
026 G2.4.48 43
D
44
039 Al.05 44
A
\\5
059 A2.07 45
C
46
059 A4.01 46
B
Page 1
1/2008 Sequoyah SRO NRC Exam
2/8/2008
(
\\
47
061 K5.01 47
B
48
062 K1.02 48
A
49
063 K2.01 49
C
50
064 K6.08 50
D
51
073 A1.01 51
A
52
076 A2.02 52
C
53
076 A4.02 53
B
54
078 K3.01 54
A
55
103 A3.01 55
A
56
011 K6.05 56
C
57
015 K2.01 57
A
58
016 K1.l0 58
D
59
017 K4.03 59
D
60
045 K5.17 60
C
61
055 A3.03 61
A
62
068 A4.02 62
D
63
071 K3.04 63
C
64
075 G 2.2.13 64
B
65
086 A1.01 65
A
66
G 2.1.3 66
C
67
G 2.1.27 67
A
(
68
G 2.1.2868
D
j9
G 2.2.2669
B
70
G 2.2.3370
C
71
G 2.3.271
C
72
G 2.3.972
B
73
G 2.3.1073
C
74
G 2.4.16 74
D
75
G 2.4.2275
A
76
008 G2.4.41 76
A
77
029 EA2.09 77
A
78
054 AA2.05 78
B
79
057 AA2.17 79
C
80
065 AA2.01 80
D
81
W/E11 2.4.48 81
A
82
W/E12 G2.1.32 82
B
83
003 G2.2.22 83
A
84
005 AA2.03 84
D
85
051 AA2.02 85
D
86
W/E02 EA2.2 86
B
87
W/E16 G 2.2.31 87
A
88
008 A2.09 88
B
89
010 G 2.4.38 89
B
90
013 A2.05 90
D
11
103 G 2.2.14 91
A
92
029 G2.4.46 92
D
Page 2
(
(
93
034 A2.02 93
94
G2.1.1294
95
G 2.1.3395
96
G 2.2.8 96
97
G2.3.l97
98
G 2.3.698
99
G 2.4.999
100
G 2.4.40100
Page 3
1/2008 Sequoyah SRO NRC Exam
2/8/2008
D
C
C
A
A
C
A
B
ES*401
Site-Specific SRO Written Examination
Cover Sheet
Form ES-401-8
(
u.s. Nuclear Regulatory Commission
Site-Specific SRO Written Examination
Applicant Information
Name:
Date:
Facility/Unit:
Sequoyah 1 & 2
Reqion:
10
II[{] 1110
IVO
Reactor Type:
W[{] CE 0
BW 0
GEO
Start Time:
Finish Time:
Instructions
Use the answer sheets provided to document your answers. Staple this cover sheet
on top of the answer sheets. To pass the examination you must achieve a final grade
of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items
if given in conjunction with the RO exam; SRO-only exams given alone require a final grade
of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination,
and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.
Applicant Certification
All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature
Results
RO/SRO-Only/Total Examination Values
-- / -- / --
Points
Applicant's Scores
-- / -- / --
Points
Applicant's Grade
-- / -- / --
Percent
ES-401, Page 31 of 33
1/2008 Sequoyah SRO NRC Exam
0210812008
1. Given the following plant conditions:
-
Unit 1 tripped from 10% power due to trip of both MFPTs.
-
The AFW pumps start signal was generated from the trip of both MFPTs.
-
Steam generator levels dropped to 12% (Lowest Level) post trip, and are slowly
rising.
Which ONE (1) of the following describes the operator response to enable the operator
to MANUALLY control AFW flow?
A. MO AFW LCVs must be taken to Accident Reset before the valves can be
controlled manually.
NO reset is required to control TO AFW LCVs manually.
B. TO AFW Pump must be taken to Accident Reset before the associated AFW flow
can be controlled manually.
NO reset is required to control MO AFW LCVs manually.
C. Both the TO AFW Pump and the MO AFW LCVs must be taken to Accident Reset
before AFW flow can be controlled.
O. Neither the TO AFW Pump nor the MO AFW LCVs must be taken to Accident Reset
before AFW flow can be controlled.
Page 1
1/2008 Sequoyah SRO NRC Exam
02/08/2008
2. Given the following:
Unit 1 Reactor Trip ISafety Injection initiated due to a pressurizer
relief valve being stuck partially open.
Reactor Coolant Pumps have been stopped.
Pressurizer level is 60% and slowly rising.
RCS pressure is 1310 psig and stable.
Core Exit Thermocouple temperature is 580°F.
What would be the effect on Reactor Vessel Level and Pressurizer Level if
the RCS pressure is lowered due to increased pressurizer relief valve
leakage?
Reactor Vessel
Pressurizer
Level
Level
A.
Rising
Continue to rise at the same rate
B.
Rising
Rising faster
C.
Dropping
Continue to rise at the same rate
D.
Dropping
Rising faster
Page 2
1/2008 Sequoyah SRO NRC Exam
02/08/2008
3. Unit 1 Reactor has been tripped with plant status as follows:
RCS pressure is 1325 psig and slowly dropping.
RCS temperature is 530°F and stable.
Pressurizer level is 17% and slowly dropping.
Containment pressure is 2.1 psig and slowly rising.
SG pressures are - 1005 psig and stable.
-
SG levels are - 40% and stable.
-
Annunciator TS-62-43 REAC COOL PMPS SEAL WATER TEMP
HIGH is LIT and indicator for RCP #3 reads 227°F and steady.
-
Annunciator TS-62-42 REAC COOL PMPS LOWER BEARING TEMP
HIGH is lit and indicator for RCP #1 reads 201°F and steady.
Which ONE (1) of the following is the correct action for the RCP(s) in these
conditions?
A. All RCPs are required to be tripped.
B. ONLY RCP #1 is required to be tripped.
C. ONLY RCP #3 is required to be tripped.
D. ONLY RCPs #1 and #3 are required to be tripped.
Page 3
1/2008 Sequoyah SRO NRC Exam
02/08/2008
4. E-1, Loss of Reactor or Secondary Coolant, directs the operator to perform a transfer
to Hot Leg Recirculation.
Which ONE (1) of the following statements correctly describes the basis for this
realignment?
A. Realigns the ECCS flow to reduce Core Exit Thermocouple temperature to
acceptable values.
B. Realigns the ECCS flow to cool the reactor vessel upper internals package.
C. Realigns the ECCS flow to pass through the RHR heat exchangers to remove
decay heat.
D. Realigns the ECCS flow to prevent boron precipitation.
Page 4
1/2008 Sequoyah SRO NRC Exam
02/08/2008
5. Unit 1 is operating at full power. Given the following events and conditions on the
RCPs:
Alarms indicate a loss of ALL CCS flow to the RCPs.
Seal injection flow rate to each RCP is 8 gpm.
Which ONE (1) of the following identifies how the operation of the RCP's will be
affected if the operators do not respond to this alarm?
A. The RCPs should operate without CCS indefinitely.
B. The RCPs will experience seal failure.
C. The RCP stator windings will overheat.
D. The RCP motor bearings will overheat.
Page 5
1/2008 Sequoyah SRO NRC Exam
02/08/2008
6. Given the following plant conditions:
Reactor power is steady-state at 75%.
Pressurizer Level Control Selector Switch (XS-68-339E) is in the 339/335
position, and level control is in automatic.
Temperature input to the pressurizer level control system fails to 530°F.
Which ONE (1) of the following describes the effect this condition would have on the
pressurizer level control system? (Assume NO operator action)
A. Charging initially increases to 120 gpm then returns to normal and pressurizer level
stabilizes at a higher value.
B. Charging initially decreases to minimum flow and indicated pressurizer level lowers
to 25% where it stabilizes.
C. Charging decreases to minimum and indicated pressurizer level lowers to 17%,
then level rises to the high level reactor trip setpoint.
D. Charging increases to 120 gpm and the pressurizer level rises to the high level
reactor trip setpoint.
Page 6
1/2008 Sequoyah SRO NRC Exam
02/08/2008
7. Unit 1 at 100% power. All conditions are normal with the following exceptions:
07:00, 1A RHR Pump is tagged due to motor repair and will be returned
to service in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
08:00, Unit 1 has a reactor trip and SI due to a LOCA.
08:01, E-O, Reactor Trip or Safety Injection, is entered followed by a transition
to E-1, Loss of Reactor or Secondary Coolant.
08:08, The 1B RHR Pump trips on overcurrent due to a locked rotor.
08:14, Step 15 of E-1, transitions crew to ECA-1.1, Loss of RHR Sump
Recirculation. The conditions at the transition are:
RCS pressure is 160 psig
Containment pressure is 9.7 psid and stable
RWST level is 58% and dropping
Containment sump level is 25% and rising
Assuming the plant condition remain unchanged, and ALL ECA-1.1 steps to
address Containment Spray Pump operation are completed, how many
containment spray pumps will be running and what will be the suction source
to the pumps?
Number of
Pumps Running
Suction Source
A.
1
B.
1
Containment sump
C.
2
D.
2
Containment sump
Page 7
1/2008 Sequoyah SRO NRC Exam
02/08/2008
8. Following a total loss of all AC power, the operator is directed to isolate CCS
to the RCP thermal barriers per ECA-O.O, "Loss of Shutdown Power". When
offsite power is restored, the operator is directed in ECA-0.1, "Recovery from
Loss of Shutdown Power without SI Required", to ensure that the RCP
thermal barrier isolation is complete prior to restarting a CCS pump.
Which ONE (1) of the following describes the basis, for isolating the CCS
thermal barrier return prior to restarting the CCS pump?
A. To prevent thermal shock to the RCP pump impeller upon restart of the
CCS system during recovery.
B. To prevent thermal shock to the RCP seal packages upon restart of the CCS
system during recovery.
C. To ensure elevated heat loads as a result of the loss of all AC power are
within the design cooling capacity of CCS prior to starting a CCS pump.
D. To prevent steam from forming and circulating in the CCS system and
ensures the CCS system is available to cool equipment necessary for
recovery.
Page 8
1/2008 Sequoyah SRO NRC Exam
02/08/2008
9. Given the following:
Reactor at 85% RTP stable conditions for 10 days.
-
Both Pressurizer Spray Valve Controllers in MANUAL and output set to "0".
-
All other systems and controllers in normal alignment.
Which ONE (1) of the following would be the immediate effect if the Pressurizer
Master Pressure Controller was placed in MANUAL and the output is raised to 100%?
A. Pressurizer Pressure HI alarm, and Actual pressurizer pressure starts to rise.
B. Pressurizer Pressure HI alarm, and Actual pressurizer pressure starts to drop.
C. Pressurizer Pressure LO alarm, and Actual pressurizer pressure starts to rise.
D. Pressurizer Pressure LO alarm, and Actual pressurizer pressure starts to drop.
Page 9
1/2008 Sequoyah SRO NRC Exam
02/08/2008
10. The CRO is performing an RCS cooldown in response to a Steam Generator tube
rupture. The oncoming shift has arrived in the control room for turnover.
In accordance with OPDP-1, which ONE (1) of the following identifies turnover
requirements?
A. An operator is NOT permitted to simultaneously turnover and perform the
cooldown;
A Control Board Walkdown and a Log Review by the oncoming operator are
required to be completed.
B. An operator is NOT permitted to simultaneously turnover and perform the
cooldown;
A Control Board Walkdown by the oncoming operator is NOT required provided the
Log Review is completed.
C. An operator is permitted to simultaneously turnover and perform the cooldown;
A Control Board Walkdown and a Log Review by the oncoming operator are
required to be completed.
D. An operator is permitted to simultaneously turnover and perform the cooldown;
A Control Board Walkdown by the oncoming operator is NOT required provided the
Log Review is completed.
Page 10
1/2008 Sequoyah SRO NRC Exam
02/08/2008
11. Given the following
A Reactor Trip/Safety Injection was initiated on Unit 2 due to a steam
line break on Steam Generator #1 outside containment but upstream of
the MSIV.
The crew responded in accordance with the emergerency operating
procedures.
When Steam Generator #1 blowdown was complete the RCS
conditions were as follows:
Pressurizer pressure 1880 psig.
Pressurizer level 9%.
Lowest RCS Tcold temperature reached was 474°F.
Which ONE (1) of the following identifies the concern when the steam
generator blowdown was complete following isolation and the condition(s)
required to re-energize the pressurizer heaters?
Concern when blowdown complete
To re-energize heaters
A. Steam Generator tube failure due
Restore level to >17%
to high differential pressure.
and Reset SI
B. Steam Generator tube failure due
Restore level to >17%
to high differential pressure.
SI reset NOT required
C. Pressurizer overfill and RCS
Restore level to >17%
pressurization.
and Reset SI
D. Pressurizer overfill and RCS
Restore level to >17%
pressurization.
SI reset NOT required
Page 11
1/2008 Sequoyah SRO NRC Exam
02/08/2008
12. Given the following Unit 1 plant conditions:
ECA-O.O, Loss of All AC Power is in effect.
Depressurization of all intact SGs is in progress.
The SG are to be depressurized at the "maximum rate".
Which ONE of the following identifies the parameter used to determine the maximum
rate and the basis for the parameter being the limiting parameter?
A. Ability to maintain pressurizer level above 10% to prevent loss of pressure control
resulting in reactor vessel head voiding.
B. Ability to maintain at least one SG narrow range level above 10% to prevent loss of
sufficient heat transfer capability from the RCS.
C. Ability to maintain RCS Tcold cooldown equal to or less than 100°F/hour to prevent
transition to FR-P.1, Pressurized Thermal Shock.
D. Ability to maintain RCS pressure equal to or greater than 200 psig to prevent
nitrogen from entering the the RCS from CLAs.
Page 12
1/2008 Sequoyah SRO NRC Exam
02/08/2008
13. Given the following conditions:
-
A Loss of Offsite Power occurred
RCS pressure is 2085 psig
Thot is 573°F
Core exit thermocouples are 583°F
Which ONE (1) of the following is the closest approximation of RCS subcooling?
Page 13
1/2008 Sequoyah SRO NRC Exam
02/08/2008
14. Given the following plant conditions:
Units 1 and 2 are operating at 100% power with No Tech Spec LCO
actions in effect.
The 125 V DC Power System is normally aligned with the exception of the
Vital Battery Board IV, which is being supplied from the Vital Battery V and
and 2-S Vital Battery Charger.
Offsite power is lost.
1A-A and 2B-B diesel generators start and load.
1B-B and 2A-A diesel generators fail to start.
If offsite power has been lost for longer than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, which ONE (1) of the
following statements identifies the condition of the 125V Vital DC batteries?
(Assume NO operator action is taken)
A. All four 125v Vital DC Batteries would be at their normal voltage requirements.
B. All four 125v Vital DC Batteries would be discharged beyond design limits.
C. 125v Vital DC Batteries I and IV would be at their normal voltage requirements.
125v Vital DC Batteries II and III would be discharged beyond design limits.
D. 125v Vital DC Batteries II and III would be at their normal voltage requirements.
125v Vital DC Batteries I and IV would be discharged beyond design limits.
Page 14
1/2008 Sequoyah SRO NRC Exam
02/08/2008
15. Which ONE (1) of the following conditions would require entry into Tech Spec 3.7.5,
A.
Water level of 675 ft and ERCW supply temperature of 81.5°F.
B.
Water level of 677 ft and ERCW supply temperature of 82.5°F.
C.
Water level of 679 ft and ERCW supply temperature of 83.5°F.
D.
Water level of 681 ft and ERCW supply temperature of 84.5°F.
Page 15
1/2008 Sequoyah SRO NRC Exam
02/08/2008
16. Given the following plant conditions:
-
A Loss Of Coolant Accident (LOCA) outside containment has resulted in RCS
Subcooling dropping to 0 degrees F.
-
The operating crew is performing the actions of ECA-1.2, LOCA Outside
Containment.
-
Actions are being taken to isolate the leak.
In accordance with ECA-1.2 "LOCA Outside Containment," which ONE (1) of the
following is the initial mitigating strategy and what would be the indication used to
confirm the LOCA has been isolated?
A. Ensure RHR suction from RCS Isolated
Pressurizer level rising
B. Ensure RHR suction from RCS Isolated
RCS pressure rising
C. Isolate RHR Cold Leg Injection
Pressurizer level rising
D. Isolate RHR Cold Leg Injection
RCS pressure rising
Page 16
1/2008 Sequoyah SRO NRC Exam
02/08/2008
17. Which ONE (1) of the following is an adverse consequence that could result from
delaying feed and bleed cooling if the conditions to initiate feed and bleed are met in
FR-H.1, "Loss of Secondary Heat Sink"?
A. High Temperature induced failure of S/G U-tube bends.
B. An overpressurization challenge to the Reactor Vessel.
C. Inability to prevent or minimize core uncovery due to high RCS pressure.
D. Inability to recover the S/Gs without damage from high thermal stresses caused by
cold water refill.
Page 17
1/2008 Sequoyah SRO NRC Exam
02/08/2008
18. Given the following conditions:
At 0900 the Reactor Trips due to 2 dropped rods.
At 0920 a small break LOCA occurs.
At 0950 the crew transitioned to ECA-1.1, "Loss of RHR Sump
Recirculation", due to the failure of both RHR pumps.
Crew has reduced ECCS flow to 1 SIP and 1 CCP per ECA-1.1.
SI flow cannot be terminated due to lack of subcooling.
At 1030 the crew is performing ECA-1.1 Step 20 RNO to establish the
minimum required ECCS flow to remove decay heat.
Which one of the following correctly describes the flow rate that meets the intent of
ECA-1.1, Step 20 RNO, AND requirements of using ECCS pumps in meeting this flow
rate?
REFERENCE PROVIDED
A. Establish 325 gpm ECCS flow. ECCS pumps may be started and stopped as
necessary to accomplish the desired flow rate.
B. Establish 325 gpm ECCS flow. ECCS pumps are NOT permitted to be started and
stopped as necessary to accomplish the desired flow rate.
C. Establish 400 gpm ECCS flow. ECCS pumps may be started and stopped as
necessary to accomplish the desired flow rate.
D. Establish 400 gpm ECCS flow. ECCS pumps are NOT permitted to be started and
stopped as necessary to accomplish the desired flow rate.
Page 18
1/2008 Sequoyah SRO NRC Exam
02/08/2008
19. Given the following:
Unit 1 operating at 50% power
Rod control in AUTO with Bank 0 at 176 steps
-
Tavg auctioneering unit fails LOW
Which ONE (1) of the following identifies how the rod control system will
respond AND what the indication will be on 1-M-4?
A. Rods inserting as indicated by RODS IN 'RED' light LIT
B. Rods inserting as indicated by RODS IN 'GREEN' light LIT
C. Rods withdrawing as indicated by RODS OUT 'RED' light LIT
D. Rods withdrawing as indicated by RODS OUT 'GREEN' light LIT
Page 19
1/2008 Sequoyah SRO NRC Exam
02/08/2008
20. Given the following plant conditions for Unit 2:
Intermediate Range N36 failed high.
Operators placed the level trip bypass switch for N36 to the bypass position.
Subsequently the Reactor trips due to Large Break LOCA.
Which ONE (1) of the following describes the operation of source range instruments
to be used to monitor the Subcriticality Status Tree?
A. Source Range channel N31 and N32 are automatically reinstated when power
decreases below P-1O.
B. Source Range channel N31 and N32 are automatically reinstated when power
decreases below P-6.
C. Both Source Range channels, N31 and N32, must be manually reinstated
when the operable Intermediate Range channel (N35) decreases below the
P-6 setpoint.
D. Both Source Range channels, N31 and N32, must be manually reinstated
when the operable Intermediate Range channel (N35) decreases below the
P-10 setpoint.
Page 20
1/2008 Sequoyah SRO NRC Exam
02/08/2008
21. Given the following:
-
Unit 1 is in Mode 6 and currently loading fuel into the core.
-
Ice is being blown into the Ice Condenser Baskets.
-
A previous cycle fuel assembly is dropped during transfer to its core
location.
-
Bubbles can be seen rising from the dropped fuel assembly and Area
Radiation Monitors go into high alarm.
In accordance with AOP-M.04, Refueling Malfunctions, all of the following would be
required EXCEPT?
A. Initiate Control Room Isolation.
B. Evacuate el. 734 Refuel Floor.
C. Initiate Auxiliary Building Isolation.
D. Close 1-78-610, Transfer Tube Wafer Valve.
Page 21
1/2008 Sequoyah SRO NRC Exam
02/08/2008
22. Waste Gas Decay Tank J contains high activity gas. Waste Gas Decay Tank J relief
valve develops a large flange leak.
Which ONE (1) of the following would identify the response of the Rad Monitors listed?
A. Waste Gas Rad Monitor (RE-90-118) would alarm, but the Auxiliary Building
Ventilation Monitor (RE-90-1 01) would NOT alarm..
B. Waste Gas Rad Monitor (RE- 90-118) would alarm, and the Auxiliary Building
Ventilation Monitor (RE-90-1 01) would alarm.
C. Waste Gas Rad Monitor (RE- 90-118) would NOT alarm, but the Auxiliary Building
Ventilation Monitor (RE-90-1 01) would alarm.
D. Waste Gas Rad Monitor (RE- 90-118) would NOT alarm, and the Auxiliary Building
Ventilation Monitor (RE-90-1 01) NOT alarm.
Page 22
1/2008 Sequoyah SRO NRC Exam
02/08/2008
23. Given the following for Unit 1:
- Plant is at 100% power.
- Containment entry is in progress for maintenance.
- The following annunciators are locked in.
- LWR PERS ACCESS OUTER DR LOCK
- LWR PERS ACCESS INNER DR LOCK
- UPRILWR AIR LOCK BREACH
- Containment pressure has rapidly equalized with the Aux Building.
- Containment pressure currently indicates 0.18 PSID.
In addition to TS 3.0.3, which ONE (1) of the following identifies both, a required TS
entry AND its associated action statement?
A. Enter TS 3.6.1.4 "INTERNAL PRESSURE" and restore Containment to annulus DP
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B. Enter TS 3.6.1.1 "CONTAINMENT INTEGRITY" and restore Containment integrity
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
C. Enter TS 3.6.1.3 "CONTAINMENT AIR LOCKS" and restore both Containment
Access doors to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
D. Enter TS 3.6.1.2 "SECONDARY CONTAINMENT BYPASS LEAKAGE" and restore
Secondary Containment Bypass Leakage within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Page 23
1/2008 Sequoyah SRO NRC Exam
02/08/2008
24. During the performance of FR-C.1, "Inadequate Core Cooling" the operators
are directed to start the RCPs if Incore TCs remain >1200°F.
Which ONE (1) of the following describes the purpose of starting the RCPs?
A. Provides temporary cooling to the core by forcing single-phase flow through
the core.
B. Provides temporary cooling to the core by forcing two-phase flow through
the core.
C. Ensure steam bubbles in the SG U-tubes are removed to maximize heat
transfer for cooling the core.
D. Ensure steam bubbles in the RCS intermediate legs are collapsed to minimize RCS
flow restrictions for cooling the core.
Page 24
1/2008 Sequoyah SRO NRC Exam
02/08/2008
25. Given the following:
-
The unit has been at 100% power
-
Total primary to secondary leakage is 29 gpd
-
Identified RCS leakage is 0.02 gpm
-
Unidentified RCS leakage is 0.02 gpm
-
Pzr level, VCT level, charging flow, and letdown flow are all stable
If fuel failure resulted in a significant increase in RCS activity, which ONE (1) of the
following identifies how the identified process rad monitors would respond?
1-RM-90-106 - Lower Containment Radiation Monitor
1-RM-90-119 - Condenser Vacuum Pump Exhaust Rad Monitor
1-RM-90-106
1-RM-90-119
A.
Stable
Rising
B.
Stable
Stable
C.
Rising
Rising
D.
Rising
Stable
Page 25
1/2008 Sequoyah SRO NRC Exam
02/08/2008
26. Given the following plant conditions:
-
Reactor trip occurred with subsequent loss of RCPs.
-
Operators have implemented ES-0.2, "Natural Circulation Cooldown" to go to
Cold Shutdown.
-
A cooldown rate of 25°F/hour has been established.
-
RCS depressurization has been initiated.
-
RVLlS upper plenum range - 98%.
-
The Shift Manager has determined that cooldown shall proceed as quickly as
possible due to reduced CST inventory.
Which ONE (1) of the following correctly states the procedure that maximizes the
allowable cooldown rate for the provided circumstances and maximum cooldown rate
allowed by the procedure?
A. Use ES-0.2, Natural Circulation Cooldown
The cooldown limit is 50°F/hr
B. Use ES-0.2, Natural Circulation Cooldown
The cooldown limit is 100°F/hr
C. Use ES-0.3, Natural Circulation Cooldown With Steam Voids in Vessel (with
RVLlS)
The cooldown limit is 50°FIhr
D. Use ES-0.3, Natural Circulation Cooldown With Steam Voids in Vessel (with
RVLlS) The cooldown limit is 100°F/hr
Page 26
1/2008 Sequoyah SRO NRC Exam
02/08/2008
27. While performing actions of FR-Z.1, "Response to High Containment Pressure", what
steps are taken to limit the peak pressure rise in containment in the event of a Main
Steam Line Break in containment?
A. Start Ice Condenser Air Handling Units.
B. Dispatch personnel to open Air Return Fan breakers.
C. Throttle all AFW Flow to less than 25 gpm per Steam Generator.
D. Isolate AFW Flow to any Steam Generator depressurizing in an uncontrolled
manner.
Page 27
1/2008 Sequoyah SRO NRC Exam
02/08/2008
28. Given the following:
-
Unit 1 is at 32% power.
-
1B Start Bus trips out on differential relay actuation.
Which ONE (1) of the following describes the plant response?
A. Reactor trips due to the loss of power to RCP #2 and RCP #4.
B. Reactor trips due to the loss of power to RCP #1 and RCP #3.
C. Only the 1B-B DIG starts and connects to the 1B-B 6.9 KV SO Bd.
D. All 4 DIGs start but ONLY the 1B-B DIG connects to the 6.9 KV SO Bd.
Page 28
1/2008 Sequoyah SRO NRC Exam
02/08/2008
29. During a plant shutdownlcooldown, which ONE (1) of the following describes the
earliest condition required to perform Hydrogen Peroxide (H202) addition to the RCS
with one RCP running, and the reason for the addition of Hydrogen Peroxide?
Condition
A. RCS Temp Less Than 250°F
provided RHR is in service
B. RCS Temp Less Than 250°F
provided RHR is in service
C. RCS Temp Less Than 180°F
D. RCS Temp Less Than 180°F
Reason
Lower the corrosion rate of the RCS
Minimize Dose in Steam Generator
area
Lower the corrosion rate of the RCS
Minimize Dose in Steam Generator
area
Page 29
1/2008 Sequoyah SRO NRC Exam
02/08/2008
30. Given the following:
Unit 1 at 100% power
Letdown in service at 75 gpm
Which ONE (1) of the following lists two condition's where BOTH would result
in a decrease in the available NPSH to the CCPs?
(Evaluate each condition separately)
Condition 1
A. VCT Level transmitter,
1-LT-62-130A fails Low
B. VCT Level transmitter
1-LT-62-130A fails High
C. Raising setpoint on
1-TCV-70-192, Letdown
Heat ExchangerTCV
D. Lowering setpoint on
1-TCV-70-192, Letdown
Heat Exchanger TCV
Condition 2
Loss of air to 1-FCV-62-77,
Letdown Flow Isolation
Loss of air to 1-FCV-62-93,
Charging Flow Control
Loss of air to 1-FCV-62-118,
Letdown Divert to HUT LCV
Loss of air to 1-FCV-62-79,
Mixed Bed High Temp Bypass
Page 30
1/2008 Sequoyah SRO NRC Exam
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31. Given the following plant conditions:
Reactor power is stable at 66% with Tave on program
-
The pressurizer level control system is in Automatic
Which ONE (1) of the following identifies the pressurizer level control setpoint?
A. 41%
B. 45%
C.48%
D.60%
Page 31
1/2008 Sequoyah SRO NRC Exam
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32. Unit 1 is currently in Mode 3 with a shutdown to Mode 5 underway. Which
ONE (1) of the following correctly identifies the administrative temperature
limit and the electrical interlock for opening 1-FCV-74-1 & 2, RHR Loop 4
Hot Leg Isolation valves in accordance with 0-SO-74-1, Residual Heat
Removal System?
A.
<235°F administrative limit, 350 psig electrical interlock
B. <235°F administrative limit, 380 psig electrical interlock
C. <350°F administrative limit, 350 psig electrical interlock
D. <350°F administrative limit, 380 psig electrical interlock
Page 32
1/2008 Sequoyah SRO NRC Exam
02/08/2008
33. Which ONE (1) of the following correctly describes the response of the
Eagle-21 system and SSPS if containment sump swapover level criteria is
met with a valid Safety Injection signal present?
Eagle-21
SSPS
A. bistables de-energize
input relays de-energize
B. bistables de-energize
input relays energize
C. bistables energize
input relays de-energize
D. bistables energize
input relays energize
Page 33
1/2008 Sequoyah SRO NRC Exam
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34. Given the following:
Unit 1 was operating at 100% power when a Reactor Trip and
Safety Injection occurred.
Shortly after the the SI actuation the following conditions were observed:
-
Annunciator 1-RA-90-59A RX BLDG AREA RAD MON HIGH RAD
in alarm.
-
Radiation rising on:
- Rx Bldg Access Hatch - Upper, 1-RM-90-59
- Rx Bldg Personnel Lock - Upper, 1-RM-90-60
- Rx Bldg Instr Rm - Lower, 1-RM-90-61
Additionally the following annunciators were noted as being in alarm:
-
TS-68-309 PRESSURIZER RELIEF TANK TEMP HIGH
-
TS-30-31
LOWER COMPT TEMP HIGH
-
TS-30-241 LOWER COMPT MOISTURE HI
-
LS-63-104 CONTAINMENT SUMP FULL
-
PS-68-301 PRESSURIZER RELIEF TANK PRESSURE HIGH
alarmed and cleared.
Which ONE (1) of the following could result in these conditions, assuming
NO operator action was taken?
A. #2 seal on RCP #4 failed
B. Pressurizer safety valve, 1-68-568, failed open
C. Reactor Vessel Head inner O-ring fails
D. Reactor head vent throttle valve, 1-FCV-68-396, failed open
Page 34
1/2008 Sequoyah SRO NRC Exam
02/08/2008
35. Given the following:
Unit 2 is operating at 100% RTP.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> into the shift annunciator WINDOW "LS-68-300A/B
PRESSURIZER RELIEF TANK LEVEL HI-LOW" ALARMS and locks
in on 2-XA-55-5A annunciator panel.
The level indicator on M-5 indicates the level to be 72%.
Redundant instrumentation in the Aux Control room shows the PRT
level to be 72%.
PRT temperature and pressure are verified to be the same as at
shift turnover.
Which ONE (1) of the following identifies the condition of the PRT and the
required crew actions as a result of the level alarm?
A. The PRT level alarm is valid, entry into AOP-R.05, RCS Leak and Leak
Source Identification is required.
B. The PRT level alarm is valid, entry into AOP-R.05, RCS Leak and Leak
Source Identification is NOT required.
C. The PRT level alarm is false. According to OPDP-1, the maximum time
that maintenance has to correct the condition causing the invalid alarm is
7 days, at which time the alarm is required to be cleared/disabled.
D. The PRT level alarm is false. According to OPDP-1, the maximum time
that maintenance has to correct the condition causing the invalid alarm is
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, at which time the alarm is required to be cleared/disabled.
Page 35
1/2008 Sequoyah SRO NRC Exam
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36. Given the following;
Both Units in service at 100% RTP.
C-S Component Cooling Water (CCS) Pump is in service.
Which ONE (1) of the following identifies requirements prior to, and after,
transferring C-S CCS Pump power supply manual throwover switch to
ALTERNATE?
A.
B.
C.
D.
Prior to the
transfer
Pump must be shutdown
Pump must be shutdown
Alternate feeder supply
breaker must be closed
Alternate feeder supply
breaker must be closed
After the
transfer
Page 36
1/2008 Sequoyah SRO NRC Exam
02/08/2008
37. Given the following:
Unit 2 was operating at 100% power.
-
The Loop 1 pressurizer spray valve controller failed causing the spray
valve to fully open.
Which ONE (1) of the following describes the response of the PZR pressure
control system to these conditions? (Assume No Operator Actions taken)
A. Master controller output would INCREASE.
PZR pressure would be maintained above the Reactor Trip setpoint.
B. Master controller output would INCREASE.
PZR pressure would decrease to the Reactor Trip setpoint.
C. Master controller output would DECREASE.
PZR pressure would be maintained above the Reactor Trip setpoint.
D. Master controller output would DECREASE.
PZR pressure would decrease to the Reactor Trip setpoint.
Page 37
1/2008 Sequoyah SRO NRC Exam
02/08/2008
38. Given the following events and conditions:
Unit 1 was operating at 60% power.
Pressurizer pressure decreased to 1940 psig.
The SSPS Train A Low Pressurizer Pressure trip logic failed to actuate.
Which ONE (1) of the following identify the status of the trip coils on the reactor trip
breakers RTA and RTB when the reactor trips?
A.
B.
C.
D.
RTA
48v UV
trip coil
Energized
Energized
De-energized
De-energized
RTB
120v Shunt
trip coil
De-energized
Energized
De-energized
Energized
Page 38
1/2008 Sequoyah SRO NRC Exam
02/08/2008
39. The Unit 1 operating crew is responding to a reactor trip due to a loss of
120V AC Vital Instrument Power Board 1-1.
Which ONE (1) of the following describes the plant response if PZR pressure
transmitter 1-PT-68-334 (Channell!) failed LOW with no operator action?
A.
SI master relays on both trains of SSPS would actuate AND both trains of
ECCS equipment would start.
B. Sl master relays on both trains of SSPS would actuate AND only "B" train
ECCS equipment would start.
C. Only the "B" train SSPS SI master relays would actuate AND both trains
of ECCS equipment would start.
D. Only the "B" train SSPS SI master relays would actuate AND only "B"
train ECCS equipment would start.
Page 39
1/2008 Sequoyah SRO NRC Exam
02/08/2008
40. Given the following:
Both Units in service at 100% RTP.
Upper Compartment Cooling Units A-A and B-B are in service on
both units.
Compare the effects of the inadvertent closing of the Lower Compartment
Cooling Unit (LCCU) valves listed:
Unit 1, LCCU 1A-A ERCW Inlet FCV (Outboard), 1-FCV-67-107
Unit 2, LCCU 2A-A ERCW Inlet FCV (Outboard), 2-FCV-67-107
If both of the valves were closed, which ONE (1) of the following identifies
the effect on the Upper Containment temperature on the respective unit(s)
and the mitigation strategy, if any?
A. Upper containment temperature would RISE on Units 1 and 2.
No additional cooling unit would be available to be placed in service for
either Unit.
B. Upper containment temperature would RISE on Units 1 and 2.
0-SO-30-4, Upper Compartment Cooling Units, would be used to place
additional coolers in service on both Units.
C. Upper containment temperature would RISE on Unit 1 only.
No additional cooling unit would be available to be placed in service for
this Unit.
D. Upper containment temperature would RISE on Unit 1 only.
0-SO-30-4, Upper Compartment Cooling Units, would be used to place
additional coolers in service for this Unit.
Page 40
1/2008 Sequoyah SRO NRC Exam
02/08/2008
41. Which ONE (1) of the following identifies a provision of the containment ice
condenser design that increases the ice condenser's ability to remove heat
during a small break LOCA?
A. Melted ice is directed away from the inlet doors and toward drain lines by
turning vanes.
B. One out of every two ice bays has a drain which directs melted ice away
from the inlet doors. (Only half of the total have drains)
C. Inlet doors have proportioning springs to modulate door opening which
equalizes air/steam flow through each ice bay.
D. Intermediate deck doors have proportioning springs to modulate door
opening which equalizes air/steam flow through each ice bay.
Page 41
1/2008 Sequoyah SRO NRC Exam
02/08/2008
42. Given the following:
Unit 1 has experienced a Reactor Tripi Safety Injection due to a LOCA.
The operating crew is implementing the Emergency Instructions and
preparing to align the suction of the Containment Spray Pumps to the
containment sump.
Before 1-FCV-72-23, Containment Spray Pump 1A Suction From
Containment Sump, would be opened, the water level in the containment
sump would have to be at least __ because of
_
A. 8%;
the interlock associated with 1-FCV-74-3, RHR Pump 1A Suction
Isolation.
B. 8%;
the interlock associated with 1-FCV-72-22, Containment Spray Pump 1A
Suction From RWST.
C. 11%;
the interlock associated with 1-FCV-74-3, RHR Pump 1A Suction
Isolation.
D. 11%;
the interlock associated with 1-FCV-72-22, Containment Spray Pump 1A
Suction From RWST.
Page 42
1/2008 Sequoyah SRO NRC Exam
02/08/2008
43. Given the following plant conditions:
Unit 1 in service at 100% power.
Due to the Normal feeder breaker to 6.9kV Shutdown Board 1B-B inadvertently
opening, 1B-B Diesel Generator started and energized the board.
One hour later, the following occurs:
-
At 1405 a steamline break occurs inside containment.
Containment pressure is 3 psig and rising.
At 1407 the CRO observes that 1B-B Containment Spray Pump ammeter reads
'0' amps and the only light LIT above the handswitch is the GREEN light.
Which ONE (1) of the following describes the current status of the 1B-B Containment
Spray Pump AND the correct operator response or reason for current status?
A. Pump should be running but has failed to auto start.
Immediately start 1B-B Containment Spray Pump after verifying sequencer has
timed out.
B. Pump should be running but has failed to auto start.
Verify 1B-B DIG loading available, then start 1B-B Containment Spray Pump.
C. Pump status is currently correct.
UVX and UVY relays are preventing start of the 1B-B Containment Spray Pump.
D. Pump status is currently correct.
Blackout timer has NOT timed out to start the 1B-B Containment Spray Pump.
Page 43
1/2008 Sequoyah SRO NRC Exam
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44. Given the following:
- Unit 1 is 1% power.
- The OATC inadvertently lowers the Steam Dump controller setpoint.
Which ONE (1) of the following identifies a parameter that requires monitoring for Tech
Spec LCO entry conditions, AND if reached, what the required action would be as
stated in the Tech Spec action?
Parameter
Tech Spec Action
A. RCS Tave < 541°F
Restore within its limit within 15 min
B. RCS Tave < 541°F
Restore within its limit within 30 min
C. RCS Pressure < 2200 psia
Restore within its limit within 5 min
D. RCS Pressure < 2200 psia
Restore within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
Page 44
1/2008 Sequoyah SRO NRC Exam
02/08/2008
45. Given the following:
Unit 2 operating at 55% power.
-
All CBPs and HDTPs in service.
Both MFPs in service.
If the Main Feed Pump Turbine(MFPT) 2B trips, which ONE (1) of the
following identifies the correct status of MFPT 2B Condenser Inlet and Outlet
Valves(FCVs) following the trip both before any Operator action is taken and
after the AOP-S.01, Loss of Normal Feedwater, has been completed?
Before any
AfterAOP
Operator Action
Completion
A.
Closed
Open
B.
Closed
Closed
C.
Open
Open
D.'
Open
Closed
Page 45
1/2008 Sequoyah SRO NRC Exam
02/08/2008
46. Given the following:
Unit 1 operating at 100% power.
-
A main feedwater transient occurs resulting in several alarms.
The Operator notes the following:
-
Turbine VPL is approximately 68%.
Control Rods inserting at 72 step/min.
MFPT 1B SPEED CONTROLLER at 100% and pump discharge flow
increased.
MFPT 1A indicates 0 flow.
MFW reg valves demand at 100% and S/G levels are below setpoint.
Steam Dump valves open.
Which ONE (1) of the following identifies the status of MFP 1A and the
position of the low pressure (LP) control valves in the steam supply to the
MFP 1A turbine?
A. MFP 1A has tripped;
LP Control valves OPEN.
B. MFP 1A has tripped;
LP Control valves CLOSED.
C. MFP 1A is NOT tripped but has unloaded;
LP Control valves OPEN.
D. MFP 1A is NOT tripped but has unloaded;
LP Control valves CLOSED.
Page 46
1/2008 Sequoyah SRO NRC Exam
02/08/2008
47. Given the following;
Unit 1 experienced a Reactor trip from 100% power.
The Operators have not operated any controls post-trip.
The crew completed E-O, Reactor Trip and Safety Injection, and has
entered ES-0.1, Reactor Trip Response.
Pressurizer level is 25% and slowly decreasing.
All Steam Generator levels are between 12% and 18% narrow range
and slowly rising.
Steam Generators pressures are approximately 990 psig and slowly
decreasing.
Tavg is 545°F and slowly decreasing.
RCS pressure is 2020 psig and slowly decreasing.
Which ONE (1) of the following actions should be the first priority of the crew
in accordance with ES-0.1 to address the conditions?
A. Establish Emergency Boration.
B. Throttle Auxiliary Feedwater Flow.
C. Close MSIVs and bypass valves.
D. Initiate Safety Injection and Return to E-O.
Page 47
1/2008 Sequoyah SRO NRC Exam
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48. Given the following plant conditions:
Unit 1 &2 are operating steady-state at 100%.
-
All systems are normally aligned.
Voltage on 6.9 kV Shutdown Board 1B-B instantaneously drops to
5400 volts.
Which ONE (1) of the following describes the plant response to these
conditions?
A. After 1.25 seconds all Diesel Generators will auto start.
B. After 1.25 seconds only the 1B-B Diesel Generator will auto start.
C. After 300 seconds all Diesel Generators will auto start.
D. After 300 seconds only the 1B-B Diesel Generator will auto start.
Page 48
1/2008 Sequoyah SRO NRC Exam
02/08/2008
49. Given the following:
6.9kv Shutdown Board 1A-A control power is being supplied from the
BACKUP bus NORMAL feeder in support of a maintenance activity.
Which ONE (1) of the following identifies the source of the control power to
the shutdown board and how the control power transfer would occur if the
normal source was lost?
Source of Control Power
A. 125v Vital Battery Board I
B. 125v Vital Battery Board I
C. 125v Vital Battery Board III
D. 125v Vital Battery Board III
Control Power Transfer
Manual
Automatic
Manual
Automatic
Page 49
1/2008 Sequoyah SRO NRC Exam
02/08/2008
50. The 2A-A DG is running. A 2A-A DG Fuel Oil Transfer Pump automatically starts due
to level in one of the day tanks and the pump shaft completely shears.
Which ONE (1) of the following correctly describes the operation of the 2A-A DG Fuel
Oil Transfer pumps?
A. The second Fuel Oil Transfer pump will be running due to low discharge pressure
on the first pump that started.
B. The second Fuel Oil Transfer pump will be running ONLY if both day tanks reach a
low level.
C. The second Fuel Oil Transfer pump will NOT be running and will not start as the
tank level continues to lower.
D. The second Fuel Oil Transfer pump will NOT be running, but will start as its
associated tank level continues to lower.
Page 50
1/2008 Sequoyah SRO NRC Exam
02/08/2008
51. Given the following plant conditions:
Both Units operating at 100% RTP.
-
All systems are in a normal alignment
-
A leak occurs in the Unit 2 Letdown Heat Exchanger.
.
Assuming no action by the crew, which ONE (1) of the following describes the effects
on the CCW System?
A. Radiation Monitor 2-RM-90-123A indication rises. Surge Tank vent
valves 1-FCV-70-66 and 2-FCV-70-66 close when the monitor reaches
the HIGH rad setpoint.
B. Radiation Monitor 2-RM-90-123A indication rises. Surge Tank vent
valves 2-FCV-70-66 closes when the monitor reaches the HIGH rad
setpoint. The 1-FCV-70-66 will remain open.
C. Unit 2 Surge Tank level lowers. The 2A and C-S CCS pumps will both
lose suction when the surge tank empties.
D. Unit 2 Surge Tank level lowers. The 2A CCS pump will lose suction when the surge
tank empties.
Page 51
1/2008 Sequoyah SRO NRC Exam
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52. Given the following plant conditions and information:
Both Units operating at 100% power
All systems aligned normally
If a large ERCW leak developed at the inlet to the 1A1 CCS Heat
Exchanger, which ONE (1) of the ERCW Headers would have increased
flow, and when the leak was isolated, how would the 1A and 2A ERCW
header pressures be effected?
A.
B.
C.
D.
Header with
Flow Increase
1A
1A
2A
2A
Effect on
Header Pressure
Both 1A and 2A header
pressures would increase
Only the 1A header
pressure would increase
Both 1A and 2A header
pressures would increase
Only the 2A header
pressure would increase
Page 52
1/2008 Sequoyah SRO NRC Exam
02/08/2008
53. Given the following:
-
0-FCV-67-152 is the CCS HX 081 and 082 outlet valve
1-FCV-67-146 is the CCS HX 1A1 and 1A2 outlet valve
-
2-FCV-67-146 is the CCS HX 2A1 and 2A2 outlet valve
Which ONE (1) of the following identifies which valve(s) receive(s) an
automatic reposition signal as a result of a Unit 1 Safety Injection?
O-FCV-67-152
1-FCV-67-146
2-FCV-67-146
A.
Yes
Yes
No
8.
Yes
No
No
C.
No
Yes
Yes
D.
No
Yes
No
Page 53
1/2008 Sequoyah SRO NRC Exam
02/08/2008
54. Which ONE (1) of the following identifies the air supply header pressure at which the
Train A essential air valve, 1-FCV-32-80, Aux Cmpsr A-A Rx Bldg U-1 Isol, would
automatically CLOSE and the effect the loss of air to containment would have on the
Pressurizer Spray Valve?
Spray Valve
Pressure
Position
A.
50 psig decreasing
Fails CLOSED
B.
50 psig decreasing
Fails OPEN
C.
69 psig decreasing
Fails CLOSED
D.
69 psig decreasing
Fails OPEN
Page 54
1/2008 Sequoyah SRO NRC Exam
02/08/2008
55. Given the following plant conditions:
Unit 2 experienced a Manual Safety Injection (SI).
Containment Purge Rad Monitors, 2-RM-90-130, was in high rad
alarm but has been reset.
Containment Purge Rad Monitors, 2-RM-90-131, has indicated
normal.
Containment Vent Isolation occurred on Train A.
Containment Vent Isolation did NOT occur on Train B.
Phase A has been RESET.
SI signal has NOT been RESET.
Which ONE (1) of the following describes the status of the Containment Vent
Isolation (CVI) system?
The CVI
and
A. should have occurred on A and B Train;
the CVI can be reset with the SI signal present.
B. should have occurred on A and B Train;
the SI signal must be reset before the CVI can be reset.
C. should NOT have occurred on B Train;
the CVI can be reset with the SI signal present.
D. should NOT have occurred on B Train;
the SI signal must be reset before the CVI can be reset.
Page 55
1/2008 Sequoyah SRO NRC Exam
02/08/2008
56. Given the following:
Unit 1 in MODE 4
RCS temperature at 240°F
Pressurizer level in automatic at 25% and a steam bubble is formed
1-XS-68-339E, LEVEL CONTROL CHANNEL SELECTOR, is in
the 335/320 position
Which ONE (1) of the following identifies which channel is selected as
backup AND whether or not it is a Post Accident Monitoring (PAM)
instrument?
A. 1-LT-68-339, Pressurizer Level Transmitter Channell
This is a PAM instrument
B. 1-LT-68-339, Pressurizer Level Transmitter Channel I
This is NOT a PAM instrument
C. 1-LT-68-335, Pressurizer Level Transmitter Channel II
This is a PAM instrument
D. 1-LT-68-335, Pressurizer Level Transmitter Channel II
This is NOT a PAM instrument
Page 56
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57. Which ONE (1) of the following sets of Nuclear Instruments would lose
power if Vital Instrument Power Board 1-11 were to be deenergized?
A. Power Range N42, Source Range N32, and Intermediate Range N36.
B. Power Range N43, Source Range N31, and Intermediate Range N35.
C. Power Range N44, Source Range N32, and Intermediate Range N36.
D. Power Range N41, Source Range N31, and Intermediate Range N35.
Page 57
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58. Which ONE (1) of the following will occur as a result of containment
pressure increasing to 1.54 psid?
A. CRDM coolers trip
Upper compartment coolers trip
B. CRDM coolers trip
Ice Cond Floor Cooling Isolation valves isolate
C. Incore Instrument Room Chillers trip
Upper Compartment Coolers trip
D. Incore Instrument Room Chillers trip
Ice Cond Floor Cooling Isolation valves isolate
Page 58
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59. Which ONE (1) of the following is the highest core temperature that can be
indicated by an Incore Thermocouple while remaining in the upper indicated
physical limits in accordance with GOI-5, Apparatus Operations?
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60. Which ONE (1) of the following identifes how (a) main steam header pressure
will respond as turbine load is raised from 50% to 70% and (b) which method
of maintaining Tavg matched with Tref would result in the value for MTC
being the MOST negative as turbine load was raised?
A. (a) Main steam header pressure would decrease.
(b) Rods are withdrawn to maintain Tavg on program, with Boron
concentration held constant.
B. (a) Main steam header pressure would increase.
(b) Rods are withdrawn to maintain Tavg on program, with Boron
concentration held constant.
C. (a) Main steam header pressure would decrease.
(b) Rod position is held constant, while Boron concentration is lowered to
maintain Tavg on program.
D. (a) Main steam header pressure would increase.
(b) Rod position is held constant, while Boron concentration is lowered to
maintain Tavg on program.
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61. Given the following:
Unit 1 is at 60% RTP with shutdown in progress due to Steam
Generator #3 tube leakage identified.
-
The Condenser Vacuum Pump Discharge filters have been installed.
1-HS-2-255, COND VAC PUMP EXH FILTER BYPASS, is in P-auto.
-
A leak on the condenser vacuum breaker develops leakage equal
to 30 scfm.
-
The following alarms are received at approximately the same time:
-
PDIS-2-255 COND VAC PMPS EXH FILTER DIFF PRESS HI.
-
1-RA-90-99A CNDS VAC PMP LO RNG AIR EXH MON HIGH RAD.
Which ONE (1) of the following identifies the status of 1-FCV-255, Condenser
Vacuum Pump Exhaust Filter Bypass Flow Control Isol?
The bypass valve
_
A. would have opened AUTOMATICALLY due to the high .6.P setpoint
across the filter.
B. would have opened AUTOMATICALLY due to the high pressure setpoint
in the exhaust stack.
C. would be prevented from opening AUTOMATICALLY or MANUALLY.
D. would be prevented from AUTOMATICALLY opening, however valve
could be opened MANUALLY using control switch.
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62. With the pump control handswitches in AUTO, which ONE (1) of the
following correctly describes operation of the Reactor Coolant Drain Tank
(RCDT) pumps when 2-FCV-68-310, PRT Drain to RCDT, is opened to
reduce the level in the Pressurizer Relief Tank (PRT)?
A. RCDT Pump A will auto start ONLY if the level in the RCDT is above the
low level pump cutoff level switch.
B. RCDT Pump A will auto start EVEN if the level in the RCDT is below the
low level pump cutoff level switch.
C. RCDT Pump B will auto start ONLY if the level in the RCDT is above the
low level pump cutoff level switch.
D. RCDT Pump B will auto start EVEN if the level in the RCDT is below the
low level pump cutoff level switch.
Page 62
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63. Given the following plant conditions:
Gas Decay Tank 'C' release is in progress with Train A ABGTS
running for dilution flow.
Irradiated Fuel assembly insert shuffles are being conducted in the
Spent Fuel Pit.
Fuel Handling Area Exhaust Fan A is running.
A leak occurs on the waste gas compressor which results in a gas
release to the Auxiliary Building.
0-RE-90-1 01, Auxiliary Building Vent Monitor, alarms due to High
radiation.
0-RE-90-102 and 0-RE-90-103, Fuel Storage Pool Area Monitors,
indicate NORMAL radiation levels and are NOT in alarm.
Which ONE (1) of the following indicates the effect this would have on the
running Fuel Handling Area Exhaust Fan and on the ABGTS?
(Assume no operator action has been taken)
A.
Fuel Handling Area Exhaust Fan would remain in service;
Train B ABGTS would start.
B.
Fuel Handling Area Exhaust Fan would remain in service;
Train B ABGTS would NOT start.
C.
Fuel handling Area Exhaust Fan would be stopped;
Train B ABGTS would start.
D.
Fuel handling Area Exhaust Fan would be stopped;
Train B ABGTS would NOT start.
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64. Given the following:
Unit 1 is in service at 70% RTP.
A tube leak is suspected in Condenser A.
1A1 and 1A3 waterboxes must be drained and tagged with a
clearance to inspect for a condenser tube leak.
After the Amertap system is shutdown in automatic, the Amertap is to
be restored to the waterboxes on other side of Condenser A.
Which ONE of the following identifies (1) the requirement for tagging the
Amertap flow paths into and out of the 1A1 and 1A3 waterboxes in
accordance with SPP-10.2, Clearance Procedure to Safely Control Energy
and (2) how having the CCW inlet and outlet MOVs closed one one side of
Condenser A effects the ability to operate a waterbox MOV on Condensers B
or C with the unit at power?
A. (1) Single valve isolation of flow path is acceptable.
(2)
An Inlet MOV on Condenser B or C could NOT be operated
electrically from the MCR, but could be electrically operated locally at
the valve.
B. (1) Single valve isolation of flow path is acceptable.
(2) An Inlet MOV on Condenser B or C could NOT be operated
electrically neither from the MCR nor locally at the valve.
C. (1)
Two closed valves in series is required unless lack of two valve
isolation is documented and communicated to the clearance holder.
(2)
An Inlet MOV on Condenser B or C could NOT be operated
electrically from the MCR, but could be electrically operated locally at
the valve.
D. (1)
Two closed valves in series is required unless lack of two valve
isolation is documented and communicated to the clearance holder.
((2) An Inlet MOV on Condenser B or C could NOT be operated
electrically neither from the MCR nor locally at the valve.
Page 64
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65. Given the following
The Electric Fire Pump is out of service for maintainence.
A scrap lumber fire exist on site but not affecting any plant equipment
or structure.
The Fire Brigade Leader calls the MCR and request the Diesel Fire
Pump be started.
Which ONE (1) of the following identifies the actions required by the MCR
operator to start the diesel fire pump and the effect on the High Pressure Fire
Protection system if the pump were to overspeed after starting with the
overspeed trip mechanism failing to function?
A. The MCR operator could start the pump from the MCR.
HPFP header pressure would increase and remain high.
B. The MCR operator could NOT start the pump from the MCR, an AUO
would have to start the pump locally.
HPFP header pressure would increase and remain high.
C. The MCR operator could start the pump from the MCR.
The pressure would increase until the pump automatically shuts down
after a time delay.
D. The MCR operator could NOT start the pump from the MCR, an AUO
would have to start the pump locally.
The pressure would increase until the pump automatically shuts down
after a time delay.
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66. Which ONE (1) of the following identifies actions, in the list below, that are
included in the listing of activities required to be reviewed during Shift
Turnover in accordance with OPDP-1, Conduct of Operations, by an
on-coming OAC at Sequoyah?
1. Radiological changes in plant
2. PERs generated since last shift worked
3. Standing Orders
4. Temporary Alteration Control forms (TACFs)
5.
LCOs
6. Priority 1 and 2 Operator Workarounds
A. All EXCEPT 1 and 2
B. All EXCEPT 1 and 4
C. All EXCEPT 2 and 6
D. All EXCEPT 4 and 6
Page 66
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67. Given the following:
Unit 2 at 35% RTP with power ascension in progress in accordance
with 0-GO-5, Normal Power Operation, following a refueling outage.
A loss of feedwater results in a reactor trip.
The level in all 4 Steam Generators drops to < 0% Narrow Range
and starts recovering.
25 seconds after the trip the Steam Generator levels are:
SG #1
-
6%
SG #2
-
9%
SG #3
-
4%
SG #4
-
3%
Which ONE (1) of the following identifies the expected response of the
ATWS Mitigation System Actuation Circuitry(AMSAC)?
A. Arming conditions for AMSAC were not present.
B. AMSAC actuated.
C. Arming conditions were present, but actuation conditions were not
present.
D. AMSAC has not actuated, but will actuate after the appropriate time delay
if SG levels stabilize at their current values.
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68. Which ONE (1) of the following best describes the function of the "0" solenoids in the
steam dump system when the "0" solenoids are energized?
A. Prevents the steam dump valves from opening if the condenser is not available.
B. Allows the steam dump cooldown valves to open when Tavg is below the Lo-Lo
Tavg interlock.
C. Aligns control air to supply for modulation control when the steam dumps are
armed.
O. Allows the steam dump valves to respond more quickly when needed..
Page 68
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69. Given the following:
Unit 1 is in MODE 6 with core reload in progress.
LS-78-3, SPENT FUEL PIT LEVEL HIGH-LOW is in alarm and an
AUO has been sent to investigate.
Source Range counts indicate N31=2cpm, and N32=3cpm.
The 17th fuel assembly to be loaded is in the Refuel Machine Mast
and is being lowered into the core.
Which ONE (1) of the following identifies a condition that would require fuel
movement to be stopped?
REFERENCE PROVIDED
A. During shift rounds an AUO reports the Spent Fuel Pit Temperature to be
69°F.
B. Source Range counts during Fuel Assembly insertion indicate
N31=4cpm, and N32=7cpm.
C. AUO reports Spent Fuel Pit level is stable approximately half way
between the EI 725.5 rung and the EL 726 rung.
D. The MCR RO maintaining the Fuel Assembly Transfer Forms (FATF)
determines that the 18th fuel assembly is in the Pit Side Upender before
the 17th assembly is set in the core.
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70. Given the following plant conditions:
Unit 1 is at 40% power, steady-state, MOL
Rod control system is in MANUAL
All process parameters are on program
A malfunction in the steam dump system causes three steam dump
valves to fail partially open, increasing reactor power to 50%
Which ONE (1) of the following identifies the value the Tref would indicate
after the plant stabilizes AND the speed at which the control rods would
move as they were being withdrawn to restore Tavg?
A. Tref would indicate approximately equal to 559°F.
Control rods would move at 64 steps/minute.
B. Tref would indicate approximately equal to 562°F.
Control rods would move at 64 steps/minute.
C. Tref would indicate approximately equal to 559°F.
Control rods would move at 48 steps/minute.
D. Tref would indicate approximately equal to 562°F.
Control rods would move at 48 steps/minute.
Page 70
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71. Given the following:
Unit 2 is stable at 3% RTP.
-
An RCS sample line fitting has a small leak inside the Polar Crane
Wall.
-
A work plan is being prepared to repair the leak and the work is being
merged into the schedule.
Which ONE (1) of the following identifies the the lowest level of approval
required to make the entry inside the Polar Crane Wall at the current plant
conditions?
A. Plant Manager
B. Site Vice President
C. Rad Protection Manager
D. Rad Ops Shift Supervisor
Page 71
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72. Given the following:
Unit 1 is in Mode 1.
Operators are in the process of placing Train A Containment Purge in
service to lower containment.
Which ONE (1) of the following damper(s) is opened LAST to ensure that the
lower ice doors remain closed during startup of the purge?
A. 1-FCO-30-1A, Purge Air Supply Fan 1A Suction Isolation Damper
B. 1-FCV-30-2, Purge Air Supply Fan 1A Discharge Isolation Damper
C. 1-FCV-30-61, Purge Air Exhaust Fan 1A Suction Isolation Damper
D. 1-FCV-30-213, Purge Air Exhaust Fan 1A Discharge Isolation Damper
Page 72
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73. Following completion of EA-0-3, Minimizing Secondary Plant Contamination,
during a SGTR event, which ONE (1) of the following identifies the alignment of the
Low Volume Waste Treatment (LVWT) Pond?
A. The LVWT pond would be aligned to the diffuser pond and aligned to recieve the
discharge of the Turbine Building Sump.
B. The LVWT pond would be aligned to the diffuser pond but isolated from the
discharge of the Turbine Building Sump.
C. The LVWT pond would be isolated from the diffuser pond but aligned to recieve the
discharge of the Turbine Building Sump.
D. The LVWT pond would be isolated from the diffuser pond and isolated from the
discharge of the Turbine Building Sump.
Page 73
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74. Given the following:
-
A reactor trip and safety injection has occurred.
E-O, Reactor Trip Or Safety Injection, is entered.
-
While the Unit Supervisor (US) is reading Step 3, an ORANGE Path
condition on the CONTAINMENT Critical Safety Function (CSF) Status
Tree is identified by the STA.
-
There are NO other ORANGE path conditions present.
Which ONE (1) of the following identifies when the transition to the ORANGE
Path procedure is to take place:
A. Immediately after the US confirms the ORANGE Path condition is
present.
B. As soon as verification that NO RED Path conditions exist on remaining
status trees.
C. As soon as the reading of the E-O IMMEDIATE ACTION steps has been
completed by the US.
D. When transition is made from E-O and verification that NO RED Path
condition exist.
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75. Which ONE (1) of the following explains the BASES for the prioritization of the HEAT
SINK safety function compared to the PRESSURIZED THERMAL SHOCK safety
function?
A. HEAT SINK safety function is prioritized HIGHER than the PRESSURIZED
THERMAL SHOCK safety function because HEAT SINK has a larger impact on the
barriers to a fission product release.
B. HEAT SINK safety function is prioritized HIGHER than the PRESSURIZED
THERMAL SHOCK safety function because Heat Sink must exist before a
PRESSURIZED THERMAL SHOCK could occur.
C. HEAT SINK safety function is prioritized LOWER than the PRESSURIZED
THERMAL SHOCK safety function because a Heat Sink must exist before a
PRESSURIZED THERMAL SHOCK could occur.
D. HEAT SINK safety function is prioritized LOWER than the PRESSURIZED
THERMAL SHOCK safety function because PRESSURIZED THERMAL SHOCK
has a larger impact on the barriers to a fission product release.
Page 75
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76. Given the following:
Unit 1 experienced a reactor trip I safety injection due to the
inadvertent opening of a pressurizer safety valve.
Reactor Coolant Pumps were stopped due to Phase B after verifying
Containment Spray in service.
Plant conditions when the first Emergency Plan classification was made:
Pressurizer pressure
1635 psig and stable
Pressurizer level
100%
High Quad. Avg. Incore Thermocouple
572°F and stable
Lowest Tcold
544°F and stable
RVLlS
87% and dropping
Steam Generator levels
33% NR and stable
Highest Containment Radiation
1.1E+01 Remlhr and stable
Containment Hydrogen
0.0%
Containment Pressure
4.8 psid and dropping
Containment Sump Level
14% and rising
RCS Activity
< Tech Spec limit
Which ONE (1) of the following identifies the required initial Emergency Plan
classification and the classification required fifty (50) minutes later if the High
Quad. Avg. Incore Thermocouple is 743°F and rising with RVLlS at 48%?
(SED Judgement is NOT to be used as criteria for the declaration)
Reference Provided
Initial
50 minutes
Classification
Later
A.
Alert
Site Area Emergency
B.
Alert
General Emergency
C.
Site Area Emergency
Site Area Emergency
D.
Site Area Emergency
General Emergency
Page 76
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77. Given the following:
Unit 1 operating at 100% power when an ATWS occurred following the
trip of both MFPTs.
All S/G water levels are approximately 4% narrow range and dropping
with ONLY the MDAFW pump 1A-A in service supplying 260 gpm to
S/G #1 and 190 gpm to S/G #2.
At the time of entry into FR-S.1 the RCS temperature and
pressure are increasing rapidly.
Subsequently, while perfoming FR-S.1, the reactor is successfully
tripped and a safety injection occurs on low pressurizer pressure.
Which ONE (1) of the following identifies BOTH the status of the turbine at
the time FR-S.1 was entered and the correct procedure implementation
as a result of the safety injection?
A. The turbine is tripped.
Perform steps 1 through 4 of E-O, Reactor Trip orSafety Injection, and
ES-0.5, Equipment Verifications, while continuing in FR-S.1.
B. The turbine is tripped.
Performance of steps 1 through 4 of E-O, Reactor Trip or Safety Injection,
and ES-0.5, Equipment Verifications, is NOT permitted until FR-S.1 is
completed.
C. The turbine is NOT tripped.
Perform steps 1 through 4 of E-O, Reactor Trip or Safety Injection, and
ES-0.5, Equipment Verifications, while continuing in FR-S.1.
D. The turbine is NOT tripped.
Performance of steps 1 through 4 of E-O, Reactor Trip or Safety Injection,
and ES-0.5, Equipment Verifications, is NOT permitted until FR-S.1 is
completed.
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78. Given the following:
Unit 1 at 4% power, startup in progress in accordance with GO-3,
Power Ascension from Reactor Critical to Less than 5 Percent Reactor
Power.
The MFW Bypass Reg valves are controlling steam generator levels in
AUTO.
Main Feed Pump (MFP) 1A in service.
Motor Driven Auxiliary Feedwater Pumps have been stopped and
placed in A-P AUTO.
Following an inadvertent Feedwater Isolation, the crew enters
AOP-S.01, Loss of Normal Feedwater
If all 4 SG levels were at 28% and dropping, which ONE (1) of the following
identifies the status of the MFW Bypass Reg Valves and the procedure to be
in effect when the plant is stabilized?
A. MFW Bypass Reg Valves would be closed;
AOP-S.01, Loss of Normal Feedwater.
B. MFW Bypass Reg Valves would be closed;
ES-0.1, Reactor Trip Response.
C. MFW Bypass Reg Valves would be opening;
AOP-S.01, Loss of Normal Feedwater.
D. MFW Bypass Reg Valves would be opening;
ES-0.1, Reactor Trip Response.
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79. Given the following:
Unit 2 is operating at 100% RTP when a transient causes a reactor trip
-
The Operating Crew implements E-O, Reactor Trip or Safety Injection,
then transitions to ES-0.1, Reactor Trip Response, and stabilizes the
plant.
-
The operating crew observes the following:
Steam dump valves CLOSED.
-
S/G Atmospheric Relief Valves Operating.
First Out annunciators DARK.
-
All SSPS status lights DARK.
Which ONE (1) of the following identifies the procedural requirements to
place TID AFW pump in service and the required time to make the first
1OCFR5(}.72 report to the NRC in accordance with SPP-3.5?
A. Transfer TID AFW Pump Turbine Trip & Throttle valve power supply in
accordance with AOP-P.02, Loss of 125v Vital DC Battety Board;
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
B. Transfer TID AFW Pump Turbine Trip & Throttle valve power supply in
accordance with AOP-P.02, Loss of 125v Vital DC Battety Board;
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
C. Transfer TID AFW Pump Turbine control power in accordance with
AOP-P.04, Loss of Unit 2 Vital Instrument Power Board;
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
D. Transfer TID AFW Pump Turbine control power in accordance with
AOP-P.04, Loss of Unit 2 Vital Instrument Power Board;
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
Page 79
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80. Given the following:
Unit 1 is in MODE 4 following a Refuel Outage.
-
ABGTS Train B is tagged for heater bank inspection.
-
Annunciator PS-32-104 TRAIN A AUX CONTROL AIR PRESSURE
LOW alarms.
-
AUO reports the Train A Aux air is isolated from station air, Train A
Aux Air Compressor running and the Train A header is 67 psig and
slowly dropping.
Which ONE (1) of the following identifies the action required per Unit 1
Technical Specifications regarding the status of ABGTS?
A. ABGTS Train A remains OPERABLE until the Train A Containment Air
Isolation valve automatically closes, at which time LCO 3.0.3 entry would
NOT be required.
B. ABGTS Train A remains OPERABLE until the Train A Containment Air
Isolation valve automatically closes, at which time LCO 3.0.3 entry would
be required.
C. Both ABGTS Trains would be INOPERABLE with the current conditions.
LCO 3.0.3 entry would NOT be required.
D. Both ABGTS Trains would be INOPERABLE with the current conditions.
LCO 3.0.3 entry would be required.
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81. Given the following:
Unit 2 is responding to a LOCA with various Aux Building Radiation
Monitors in alarm.
The operators reach the last step in the procedure in effect and
observes the following conditions:
RCS pressure 1040 psig and dropping.
RWST Wide Range level 72% and dropping.
Containment Sump level is 0%.
Which ONE (1) of the following statements identifies...
1) The procedure transition required to assure continued removal of decay
heat under these conditions.
and
2) How the termination of ECCS flow would be procedurally accomplished
when termination criteria was met?
A. 1)
Transition to ECA-1.1, Loss of RHR Sump Recirculation;
2)
ECCS flow would be terminated in ECA-1.1.
B. 1) Transition to ECA-1.1, Loss of RHR Sump Recirculation;
2)
ECCS flow would be terminated in ES-1.1.
C. 1)
Transition back to E-1, Loss of Reactor or Secondary Coolant;
2)
ECCS flow would be terminated in E-1.
D. 1) Transition back to E-1, Loss of Reactor or Secondary Coolant;
2)
ECCS flow would be terminated in ES-1.1
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82. Given the following:
The crew is performing actions of ECA-2.1, "Uncontrolled Depressurization of
All Steam Generators."
The RCS subcooling is 48 degrees F.
RCS pressure is 1540 psig and rising.
Pressurizer level is 24% and rising.
While performing Step 14, MONITOR SI Termination Criteria, the SIG #4
pressure suddenly begins to rise in an uncontrolled manner.
What action is required for this condition?
A. Terminate SI in ECA-2.1, then transition to E-2, Faulted Steam Generator Isolation.
B. Transition to E-2, Faulted Steam Generator Isolation, then terminate SI.
C. Remain in ECA-2.1, SI termination criteria are not currently met.
D. Transition to E-2, Faulted Steam Generator Isolation, SI termination criteria are not
currently met.
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83. Given the following:
Unit 1 is at 100% RTP.
Control Rod D Rods are at 220 steps.
A single rod in Control Bank D drops to 115 steps.
Which ONE (1) of the following statements describes the BASES for restricting thermal
power to less than or equal to 75%?
A. Ensures Fuel Rod Integrity is maintained during continued operation.
B. Minimize AFD swings due to xenon redistribution.
C. Removes requirement to determine QPTR using incore probes.
D. Eliminates the requirement to re-evaluate specific accident analysis.
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84. Given the following:
Unit 1 operating at 86% power when MFP 1A trips.
-
As the rods are inserting, the OATC observes the following:
Bank D Group 1 remains at 190 steps while the Bank D Group 2 rods
are inserting as indicated by RPls and step counters.
-
The OATC places the rod control to MANUAL stopping rod motion with
Bank D Group 2 rods at 148 steps.
Which ONE of the following identifies the correct response to the rod
misalignment per AOP-C.01, Rod Control System Malfunction?
A. Make repairs then realign the Group 1 rods to the Group 2 rods per
B. Make repairs then realign the Group 2 rods to the Group 1 rods per
C. Remove the Unit from service within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> in accordance with Tech
Specs using GOs.
D. Trip the Reactor and Go to E-O, Reactor Trip or Safety Injection.
Page 84
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85. Given the following:
-
Unit 1 turbine load is at 40%.
-
Condenser vacuum is being lost due to air in-leakage.
-
Operators have entered AOP-S.02, Loss of Condenser Vacuum.
-
Condenser pressure is 1.9 psia and increasing at 0.1 psia/minute.
Which of the following are the correct actions in accordance with AOP-S.02 for the
current conditions?
A. Trip the turbine and GO TO AOP-S.06, Turbine Trip.
B. Trip the Reactor and GO TO E-O, Reactor Trip or Safety Injection.
C. Continue in AOP-S.02, if condenser pressure exceeds 2.7 psia and can NOT be
restored within 5 min, Trip the turbine and GO TO AOP-S.06, Turbine Trip.
D. Continue in AOP-S.02 if condenser pressure exceeds 2.7 psia and can NOT be
restored within 5 min, Trip the Reactor and GO TO E-O, Reactor Trip or Safety
Injection.
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86. Given the following:
An inadvertent Reactor Trip occurred on Unit 1.
A Safety Injection was initiated due to SG #2 Atmospheric Relief
Valve sticking open.
The operating crew is currently performing E-2, Faulted Steam
Generator Isolation and the following conditions exist:
RCS subcooling is 64 degrees F.
S/G levels in #1, #3 and #4 SGs are stable at 33% narrow range.
S/G #2 level is 0% wide range.
RCS pressure 1760 psig and rising.
Pressuizer level 22% and rising.
Which ONE (1) of the following identifies the procedural flow path to
implement ES-1.1, SI Termination and, the required action in ES-1.1 if the
RCS pressure starts to drop and continues to drop following the action to
stop the first centrifugal charging pump (CCP)?
A. Transition to ES-1.1 would be made DIRECTLY from E-2.
Restart the CCP and transition to E-1, Loss of Reactor or Secondary
Coolant.
B. Transition to ES-1.1 would be made DIRECTLY from E-2.
Transition to ES-1.2, Post LOCA Cooldown.
C. No direct transition from E-2 to ES-1.1 can be made, a transition must
first be made to E-1, Loss of Reactor or Secondary Coolant.
Restart the CCP and transition to E-1, Loss of Reactor or Secondary
Coolant.
D. No direct transition from E-2 to ES-1.1 can be made, a transition must
first be made to E-1, Loss of Reactor or Secondary Coolant.
Transition to ES-1.2, Post LOCA Cooldown.
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87. Given the following conditions:
Unit 1 core reload in progress.
An irradiated assembly is in the manipulator crane and is being transported
to a core location.
An additional irradiated fuel assembly is in the RCCA change fixture.
The Refueling SRO determines a leak exists in the Reactor Cavity Seal and
RADCON tech reports that radiation levels are rising as the water level is
dropping.
Which ONE (1) of the following describes the action that is REQUIRED to be taken
with the 2 fuel assemblies in accordance with AOP-M.04, Refueling Malfunctions?
A.
B.
C.
D.
Irradiated Assembly
In Manipulator Crane
Transport back to Spent
Fuel Pit side
Transport back to Spent
Fuel Pit side
Place in any core location
Place in any core location
Irradiated Assembly
In RCCA change Fixture
Place in any core location
Transport back to Spent
Fuel Pit side
Transport back to Spent
Fuel Pit side
Place in any core location
Page 87
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88. Given the following:
Unit 1 restarted 10 days ago following a refueling outage and is
operating at 100% power.
Control Bank D is currently 185 steps withdrawn and Rod control isin
AUTOMATIC.
Subsequently, the following occurs;
A controller problem results in the Letdown Hx Temperature
Control Valve slowly drifting closed.
The operating crew enters AOP-C.02, Uncontrolled RCS
Boron Concentration Changes.
Which ONE (1) of the following correctly describes the direction the control
rods would move AND how rod control is to be operated in accordance with
A. OUT
Control rods are to be withdrawn in Manual
B. OUT
Control rods are to be operated in Automatic
C. IN
Control rods are to be inserted in Manual
D. IN
Control rods are to be operated in Automatic
Page 88
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89. If the Reactor Cooling System pressure exceeded the Technical
Specification Safety Limit in Mode 3, which ONE (1) of the following identifies
the maximum time allowed for each of the following?
1. Pressure restored to less than Safety Limit.
2. Operations Duty Manager (ODS) Notification.
A.
B.
C.
D.
Pressure Restored
5 minutes
5 minutes
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
1 hour
ODS Notification
Within 5 minutes after event
initiation
Within 5 minutes after REP
declaration
Within 5 minutes after event
initiation
Within 5 minutes after REP
declaration
Page 89
1/2008 Sequoyah SRO NRC Exam
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90. Given the following:
Unit 1 at 100% power with the TD-AFW pump out of service for
maintenance.
-
Annunciator "125V DC VITAL CHARGER II FAILURE OR
VITAL BAT II DISCHARGE" ON 1-M-1 panel is in alarm and the crew
is investigating the cause.
Following a reactor trip due to a spurious main turbine trip, the crew is
performing ES-0.1, Reactor Trip Response, when the OATC
determines the Train B AFW pump is NOT running.
Which ONE (1) of the following identifies a reason the AFW pump B is not
running AND the allowed procedure use sequence provided by EPM-4,
User's Guide, to allow placing the AFW pump B in service?
A. Loss of control power to SSPS Train B slave relays.
AOP-P.02, Loss of 125v DC Vital Battery Board, can be implemented
ONLY after ES-0.1 has been completed.
B. Loss of control power to SSPS Train B slave relays.
Implement AOP-P.02, Loss of 125v DC Vital Battery Board while
continuing in ES-0.1.
C. Loss of control power to the 6900v Shutdown Board.
AOP-P.02, Loss of 125v DC Vital Battery Board, can be implemented
ONLY after ES-0.1 has been completed.
D. Loss of control power to the 6900v Shutdown Board.
Implement AOP-P.02, Loss of 125v DC Vital Battery Board while
continuing in ES-0.1.
Page 90
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91.
Given the following:
Unit 1 is being shutdown for a refueling outage.
Operating crew is performing 0-GO-7, Unit Shutdown From Hot
Standby To Cold Shutdown.
Which ONE (1) of the following identifies when Containment Closure Control
is required to be implemented in accordance with 0-GO-15, Containment
Closure Control, AND who will maintain the listing of Containment Closure
Exceptions in effect in accordance with 0-GO-15?
A.
B.
C.
D.
Required
When
When Unit enters
Mode 5
When Unit enters
Mode 5
When RHR is placed
in service
When RHR is placed
in service
Containment Closure
Exceptions Maintained by
Operations Work Control
Center(WCC) US
Unit 1 SRO
Operations Work Control
Center(WCC) US
Unit 1 SRO
Page 91
1/2008 Sequoyah SRO NRC Exam
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92. Given the following:
Unit 1 is in MODE 4 with Heatup in Progress.
Unit 2 is in MODE 6 with core reload in progress.
Which ONE (1) of the following identifies a condition where the associated
Unit's Containment Purge would be required to be shutdown, but would
continue to run until an Operator MANUALLY shutdown the purge?
A. Annunciator, 1-RA-90-131A CNMT PURGE AIR EXH MON HIGH RAD,
alarms with Unit 1 Upper Containment Purge in progress.
B. Annunciator, 0-RA-90-101A AUX BLDG VENT MONITOR HIGH RAD,
alarms with Unit 1 Lower Containment Purge in progress.
C. Annunciator, 2-RA-90-131A CNMT PURGE AIR EXH MON HIGH RAD,
alarms with Unit 2 Lower Containment Purge in progress.
D. Annunciator, 0-RA-90-101A AUX BLDG VENT MONITOR HIGH RAD,
alarms with Unit 2 Upper Containment Purge in progress.
Page 92
1/2008 Sequoyah SRO NRC Exam
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93. You are the Fuel Handling SRO for receiving new fuel. As a new fuel cask is being
lifted from the transport truck, the lifting slings break and the cask drops to the ground.
Initial inspections indicate extensive damage to one fuel assembly.
Per AOP-M.04, Refueling Malfunctions, which ONE (1) of the following correctly
describes the MINIMUM level required for approval of recovery instructions?
A. Fuel Handling SRO.
B. Shift Manager.
C. Operations Manager.
D. Plant Manager.
Page 93
1/2008 Sequoyah SRO NRC Exam
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94. Given the following:
Both units in MODE 1.
The following data resulted from the weekly testing on 125v DC
Vital Battery III.
Battery on float charge.
Battery Terminal Voltage 133v dc.
Battery charging current is 2.2 amps.
Battery Room and Electrolyte temperature is 68°F.
Cell electrolyte level is between minimum and maximum marks.
Cell voltage is 2.11v dc.
Corrected Specific Gravity is 1.197.
Which ONE (1) of the following identifies the Unit 1 TIS required actions
regarding the 125v DC Vital Battery III?
REFERENCE PROVIDED
A. Immediately declare INOPERABLE due to Category A criteria, restore to
OPERABLE status within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT
STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the
following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
B. Immediately declare INOPERABLE due to Category B criteria, restore to
OPERABLE status within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT
STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the
following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
C. May be considered OPERABLE provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> all the
Category B measurements are taken and found to be within their
allowable values, and provided all parameter(s) are restored to within
limits within the next 6 days.
D. May be considered OPERABLE provided all Category B measurements
are within their allowable values and provided the parameter(s) are
restored to within limits within the next 7 days. No requirement exists to
verify Category B measurements within allowable values within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Page 94
1/2008 Sequoyah SRO NRC Exam
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95. Given the following:
-
Unit 2 is in Mode 5.
-
At 0130 the Control Room Emergency Ventilation system (CREVS) Train B was
started.
-
At 0200 the CREVS Train A was removed from service for heater repair, with
repairs expected to be completed in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
-
At 0530 the CREVS Train B breaker tripped on overload.
Which ONE (1) of the following describes Unit 2 Tech Spec 3.7.7 "Control Room
Emergency Ventilation System" applicability for this condition AND the basis for this
Tech Spec?
A. Tech Spec 3.7.7 would NOT be entered in this mode.
Limits the radiation exposure to personnel in the control room to 5 rem or less
whole body following all credible accident conditions.
B. Tech Spec 3.7.7 would NOT be entered in this mode.
Limits the radiation exposure to personnel in the control room to 1 rem or less
whole body following all credible accident conditions.
C. Tech Spec 3.7.7 would be entered in this mode.
Limits the radiation exposure to personnel in the control room to 5 rem or less
whole body following all credible accident conditions.
D. Tech Spec 3.7.7 would be entered in this mode.
Limits the radiation exposure to personnel in the control room to 1 rem or less
whole body following all credible accident conditions.
Page 95
1/2008 Sequoyah SRO NRC Exam
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96. An engineer has submitted a design change request to replace the controller for
FCV-62-93, "Charging Flow Control," with a different type of controller to
improve pressurizer level control at low power.
Which ONE (1) of the following correctly describes the MINIMUM required
qualifications for the person(s) PREPARING the safety evaluation paperwork in
accordance with SPP-9.4, 10CFR50.59 Evaluations of Changes, Tests, and
Experiments?
A. 50.59 qualified ONLY.
B. SRO Licensed Operator ONLY.
C. Degreed Engineer AND 50.59 qualified.
D. SRO Licensed Operator AND 50.59 qualified.
Page 96
1/2008 Sequoyah SRO NRC Exam
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97. Given the following:
-
Unit 2 is in Mode 4
-
The Chemistry Lab Technician notifies the control room that both the Inside and
Outside Containment RCS Hot leg Sample isolation valves failed to close after
drawing a sample.
Which of the following is correct concerning the initial action time requirement
and
the applicability of the associated 31 day penetration isolation verification requirement
due to the failure listed above (incorporating ALARA considerations), if the selected
isolation valve is in a High Radiation Area?
Action
A. Isolate within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
B. Isolate within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
C. Isolate within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
D. Isolate within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
31 day Penetration Isolation Verification
Is required every surveillance period
and can be verified Closed by
administrative means
Is required every surveillance period
and can be verified Closed by
administrative means
Is NOT required if the selected
penetration isolation valve is locked
in the Closed position
Is NOT required if the selected
penetration isolation valve is locked
in the Closed position
Page 97
1/2008 Sequoyah SRO NRC Exam
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98. Given the following conditions on Unit 1:
You are the Unit Supervisor.
Rad Waste water inventory is approaching storage capacity.
A release of the Monitor Tank is planned.
Sample results indicate non gaseous activity in the tank is slightly higher
than the 3.0E-5 uci/ml value listed in 0-SI-CEM-077-400.1, Liquid
Waste Effluent Batch Release, for opening a Batch Release Permit.
The source check on 0-RM-90-122, Liquid Radwaste Release Monitor
has failed.
The Shift Manager has declared 0-RM-90-122, Liquid Radwaste
Release Monitor Inoperable.
Which ONE (1) of the following statements is correct regarding
approving/disapproving the permit for releasing the Monitor Tank in accordance
with 0-SI-CEM-077-400.1?
A. Approval not permitted. Cannot release Monitor Tank until contents is reprocessed
to lower activity.
B. Approval not permitted. Cannot release monitor tank until 0-RM-90-122 has been
returned to OPERABLE status.
C. Approval is permitted. Provided it is released in accordance with ODCM. Shift
Manager approval is required for release due to activity level.
D. Approval is permitted. Provided it is released in accordance with ODCM. Shift
Manager approval is NOT required for release at this activity level.
Page 98
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99. Given the following:
-
Mode 5.
-
RCS Boron Concentration = 2210 ppm.
-
Required Shutdown Margin Boron Concentration = 2100 ppm.
-
RWST Boron Concentration = 2600 ppm.
-
Chemistry reports Volume Control Tank Boron Concentration = 2000 ppm due to
Primary Water inleakage into the VCT which has now been stopped.
-
RCS is being drained to 696' for maintenance, per 0-GO-13, "Reactor Coolant
System Drain and Fill Operations."
-
RCS level is inadvertently lowered to 695'.
-
RHR pump amps, discharge pressure, and flow begin to fluctuate excessively.
Which ONE (1) of the following describes the required action(s) and mitigating
response in accordance with AOP-R.03, RHR System Malfunction, and Technical
Specifications?
Action
A. Stop RHR Pump
B. Stop RHR Pump
C. Lower RHR Pump Flow
rate to 1000 - 1500 gpm
D. Lower RHR Pump Flow
rate to 1000 - 1500 gpm
Mitigating Response
gravity fill.
Raise RCS level using charging
pump taking suction from VCT
gravity fill
Raise RCS level using charging
pump taking suction from VCT
Page 99
1/2008 Sequoyah SRO NRC Exam
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100. Given the following:
-
An event is occurring on Unit 1 that requires an ALERT emergency be
declared in accordance with the Radiological Emergency Plan (REP).
The Shift STA does NOT have an active SRO license.
Prior to making the declaration the Shift Manager becomes
incapacitated due to a medical problem.
Which ONE (1) of the following identifies both the individual who is
responsibile for the making the REP declaration in accordance with EPIP-1,
Emergency Plan Classification Matrix, and the requirement for conducting
Assembly and Accountibility in accordance with EPIP-3, Alert?
A.
B.
C.
D.
Responsible for
making REP Declaration
Unit Supervisor
Unit Supervisor
Shift Technical Advisor
Assembly and Accountibility
Required
NOT required but can be conducted.
Required
NOT required but can be conducted.
Page 100