ML100850470
| ML100850470 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/16/2010 |
| From: | NRC Region 4 |
| To: | Entergy Operations |
| References | |
| 50-313/10-03, 50-368/10-03 license-operator, part 55 exam material | |
| Download: ML100850470 (154) | |
Text
ANO Unit 1 Initial Exam given March 16, 2010 Written RO exam contains Questions 1-75 Written SRO exam contains Questions76-100, located after the RO questions in this packet.
Reference files that were handouts are attached to back of exam.
The exam outlines are embedded within this file for each section of the exam.
Example, Tier 1, Group 1 topics for RO are the first several questions and therefore this section of the outline is located just in front of these questions in this package.
PWR Examination Outline Form ES-401-2 ES-401 Facility: Arkansas Nuclear One Unit I Date of Exam: 31512010 1 SRO-Only Points Tier Group G*
KKKKKKAAAAG A2 Total 1 2 3 4 5 6 1 2 3 4
- Total
- 1. 1 333 33 3 18 6 Emergency &
Abnormal 2 j N/A j_ N/A 1.
Tier Totals 5 5 4 4 5 4 27 10 Evolutkrns 1 33233222233 28 5 2.
2 011 ! 10 3 Plant Systems 3 4 4 38 8 Tier Totals 3 43 4 4 3 3 3
- 3. Generic Knowledge and Abilities Categories 1 ]I 2 I
3 4 10 1 2 3 4 7 1 2 1 4 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO only outlines (i.e., except for one category in Tier 3 of the SRO-oniy outline, the Tier Totals in each K/A category shall not be less than two).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the associated oufline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1 .b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1 .b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals G* on the SRO for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
ES-401 PWR Examination Outline Form ES-401-2 S-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group I (RO)
E/APE # / Name / Safety K/A Topic(s)
Function 000007 (BW/E02&E10; EK3.O1 Actions contained in EOP for CE/E02) Reactor Trip -
reactor trip Stabilization_-_Recovery /_1 2.2.37 Ability to determine operability 000008 Pressurizer Vapor -x and/or availability of safety related Space Accident /3 equipment 000009 Small Break LOCA / 3 EAI.02 RB Sump level 000011 Large Break LOCA /3 EK2.02 Pumps 000015/17 RCP Malfunctions/ AK2.1O RCP indicators and controls.
4 000022 Loss of Rx Coolant Not selected Makeup/2 AK1.01 - Loss of RHRS during all modes of 000025 Loss of RHR System /
operation.
4 I -
000026 Loss of Component I -
Not selected Cooling Water / 8 AK2.03 Controllers and positioners.
000027 Pressunzer Pressure Control System Malfunction /3 000029 ATWS / 1 I x EA2.02 Reactor trip alarm.
000038 Steam Gen. Tube 2.4.6 Knowledge of EOP mitigation upture / 3 strategies.
00040 (BW/E05; CE/E05; AKI .01 Consequence of PTS.
W/E12) Steam Line Rupture -
Excessive_Heat Transfer /4 AA2.06 AFW adjustments needed to 000054 (CE/E06) Loss of Main x maintain proper T-ave, and S/G level.
Feedwater / 4 EAI .05 Battery, when approaching fully 000055 Station Blackout / 6 discharged.
AK3.01 Order and time to initiation of 000056 Loss of Off-site Power /
power for the load sequencer.
AA2.05 S/G pressure and level meters.
000057 Loss of Vital AC Inst.
Bus/6 I-AKI .01 Battery charger equipment and 000058 Loss of DC Power I6 instrumentation.
000062 Loss of Nudear Svc AK3.02 The automatic actions Water / 4 (alignments) within the nucleai service water resulting from the actuation of the ESFAS.
000065 Loss of Instrument Air / Not selected 8
W/E04 LOCA Outside Not selected Containment / 3 ES-401 Form ES-401-2
ES-401 PWR Examination Outline Form ES-401-2 S-401 PWR Examination Outline Form ES-401-2 Emergencyc1 Abnormal Plant Evolutions Tier 1/Group I (RO)
E/APE #1 Name! Safety K/A Topic(s)
Function W/E1 1 Loss of Emergency Not selected Coolant Recirc. / 4 BW/E04; W/E05 Inadequate EAI .3 Desired operating results during Heat Transfer Loss
- abnormal and emergency situations.
of Secondary Heat Sink / 4 000077 Generator Voltage and 2.1.25 Ability to interpret reference material, Electric_Gild_Disturbances /6 such as .ra.hs, curves, tables, etc.
K/A Category Totals: Group Point Total:
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 770 Rev: 0 Rev Date: 9/2/2009 Source: Modified Originator: S. Pullin TUOI: Al LP-RO-EOPO1 Objective: 13 Point Value: 1 Section: 4.1 Type: Generic EPEs System Number: 007 System
Title:
==
Description:==
knowledge of the reasons for the following as the apply to a reactor trip: Actions contained in EOP for reactor trip.
KIA Number: EK3.0l CFR
Reference:
41 .5/41 .10/45.6/45.13 Tier: 1 RO Imp: 4.0 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 4.6 SRO Select: Yes Taxonomy: C Question: RO:j SRO:J The following conditions exist immediately after a reactor trip:
- Group 2, Rod 4 and Rod 5 failed to fully insert into the core
- RCS pressure is at 1730 psig
- Pressurizer level is at 50 inches
- A OTSG pressure is at 910 psig
- B OTSG pressure is at 905 psig
- CETs are 560°F and stable
- Turbine Trip Solenoid Power Available light is OFF Which action is the operator required to perform FIRST in response to the given information as well as the reason for the action?
A. Manually actuate MSLI for affected SG(s) and EFW due to loss of DOl.
B. Commence emergency boration per RT-12 due to stuck rods.
C. Trip all Reactor Coolant Pumps due to loss of subcooling margin.
D. Initiate High Pressure Injection per RT-2 due to low pressurizer level and low RCS pressure.
Answer:
B. Commence emergency boration per RT-12 due to stuck rods.
Notes:
(a) is incorrect since reactivity management is a higher priority and must be addressed first.
(b) is correct since two stuck rods require emergency boration.
(c) is incorrect because subcooling margin is adequate.
(d) is incorrect since pressurizer level is >30 inches and RCS pressure is >1700 psig.
References:
1202.001 Change 31 History:
This question modified from QID 0023 Selected for the 2010 RO/SRO Exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0771 Rev: 0 Rev Date: 9/03/09 Source: New Originator: S.PulIin TUOI: AILP-RO-AOP Objective: 2 PointValue: I Section: 4.2 Type: Generic APEs System Number: 008 System
Title:
Pressurizer (PZR) Vapor Space Accident
==
Description:==
Ability to determine operability and/or availability safety related equipment.
K/A Number: 2.2.37 CFR
Reference:
41 .7/43.5/45.13 I RO Imp: 3.6 RO Select: Yes Difficulty: 2 Tier:
Group: 1 SRO Imp: 4.6 SRO Select: Yes Taxonomy: K Question: RO: 2 SRO:1 Given:
stopped the
- Pressurizer Spray fails open and the ATC operator was able to close the Spray valve and Reactor Coolant system pressure decrease.
- Annunciator alarm PZR HEATER GROUND FAULT (K09-E3) comes in.
- RCS pressure response abnormally slow with Pressurizer heaters energized.
(1307.009) to
- Maintenance is requested to perform Unit I Emergency Powered Pressurizer Heater Checkout determine operability of vital powered pressurizer heaters the vital powered Which heaters groups are the vital powered pressurizer heaters, and which KW output of heaters will satisfy the operability requirements of Technical Specification 3.4.9?
A. Group 1 proportional heaters, Group 2 proportional heaters, Group 4 heaters, 124 KW output.
B. Group I proportional heaters, Group 2 proportional heaters, Group 5 heaters, 128 KW output C. Group I proportional heaters, Group 3 heaters, Group 5 heaters, 124 KW output D. Group 2 proportional heaters, Group 4 heaters, Group 5 heaters, 128 KW output Answer:
B. Group I proportional heaters, Group 2 proportional heaters, Group 5 heaters, 128 KW output Notes:
A. is incorrect wrong groups of heaters and KW output to low B. is the correct answer correct groups of heaters and KW meets operability requirements of TS 3.4.9 C. is incorrect wrong groups of heaters and KW output to low D. is incorrect wrong groups of heaters and KW meets operability requirements of TS 3.4.9
References:
1203.015 change 016 T.S. 3.4.9 amendment # 215 History:
New for the ROISRO 2010 exam
A ARKANSAS INITIAL ROISRO EXAM BANK QUESTION DAT NUCLEAR ONE - UNIT I Rev Date: 9/3/09 Source: New Originator: S.Pullin QID: 0772 Rev: 0 Objective: 5 Point Value: I TUOI: Al LP-RO-AOP Section: 4.1 Type: Generic EPEs System Number: 009 System
Title:
Small Break LOCA level ing as they apply to a small break LOCA: RB sump
==
Description:==
Ability to operate and monitor the follow KIA Number: EA1 .02 CFR
Reference:
41.7/45.5/45.6 3.8 RO Select: Yes Difficulty: 4 Tier: I RO Imp:
SRO Select: Yes Taxonomy: Ap Group: I SRO Imp: 3.8 Question: RO:I 3 SRO:F3 Given:
Small break LOCA has occurred.
The Reactor building sump is filling at a rate of 2%/minute.
Reactor Building sump level is 44%
ing steady how long can the Reactor building sump be What is the RCS leak rate and with the leak size remain used for an accurate leak rate calculation?
sump level can be used for 3 minutes.
A. RCS leak rate approximately 91 gpm, and the RB level can be used for 8 minutes.
B. RCS leak rate approximately 91 gpm, and the RB sump sump level can be used for 3 minutes.
C. RCS leak rate approximately 45 gpm, and the RB sump level can be used for 8 minutes.
level can be used for 3 minutes.
A. RCS leak rate approximately 91 gpm, and the RB sump Notes:
% and the RB sump can only be used for leak rate A. is the correct answer due to sump is 45.4 gallons per?
get an accurate leak rate due to volume determination up to 50% level after that level you can not uncertainties B. is incorrect due to the wrong Ieakrate and time.
C. is incorrect due to the wrong leakrate and time.
D. is incorrect due to the wrong leakrate.
References:
STM 1-08 Rev. 14 History:
New for the RO/SRO 2010 exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0198 Rev: 1 Rev Date: 8/9/05 Source: Direct Originator: J. Haynes TUOI: A1LP-RO-RBS Objective: 6 Point Value: 1 Section: 4.1 Type: Generic EPEs System Number: 011 System
Title:
Large Break LOCA
==
Description:==
Knowledge of the interrelations between the Large Break LOCA and the following: Pumps.
KIA Number: EK2.02 CFR
Reference:
41.7/45.7 Tier: 1 RO Imp: 2.6 RO Select: Yes Difficulty: 2 Group: 1 SRO Imp: 2.7 SRO Select: Yes Taxonomy: K Question: RO: 4 SRO:
Given:
- A large break LOCA has occurred.
- Offsite power has been lost.
Why must Reactor Building Spray flow be throttled to 1050-1200 gpm prior to transferring to Reactor Building sump suction?
A. To ensure adequate NPSH for ECCS pumps.
B. To prevent pump runout on the Spray pumps.
C. To prevent overloading EDG5 on transfer.
D. To reduce radiation levels near RB Spray piping.
Answer:
A. To ensure adequate NPSH for ECCS pumps.
Notes:
(a.) is correct.
(b.) is incorrect. The spray pumps are designed for the full flow that is achieved during ES conditions.
(C.) is incorrect. The EDGs are designed to handle the load of the spray pumps at full flow.
(d.) is incorrect. Flow rates will not effect radiation levels on sump recirc.
References:
1202.012, Chg. 008 History:
Developed for use in 98 RO Re-exam.
Selected for use in 2005 RO exam, replacement question.
Selected for the RO/SRO 2010 exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0609 Rev: 0 Rev Date: 8/9/05 Source: Direct Originator: Cork/Pullin TUOI: Al LP-RO-ARCP Objective: 19 Point Value: I Section: 4.2 Type: Generic APEs System Number: 015 System
Title:
Reactor Coolant Pump Malfunctions
==
Description:==
Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: RCP indicators and controls.
K/A Number: AK2.1 0 CFR
Reference:
41.7 / 45.7 Tier: I RO Imp: 2.8 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 2.8 SRO Select: Yes Taxonomy: K Question: RO:1 5 SRO:j Which of the following indications would require stopping a Reactor Coolant Pump?
A. Seal cavity pressures oscillating at 600 psi peak to peak B. Seal bleedoff temperature 160°F C. Seal beedoff temperature 60°F above 1st stage seal temperature D. Failure of one stage as indicated by zero stage DP Answer:
C. Seal beedoff temperature 60°F above 1st stage seal temperature Notes:
Answer °C is correct, this exceeds 40°F delta-T specified in section 1 of 1203.031.
Answers °A, B and D just indicate a need for increased monitoring frequency of an RCP.
References:
1203.031, Chg. 018 History:
New for 2005 RO exam, replacement question.
Selected for the RO/SRO 2010 exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0773 Rev: 0 Rev Date: 9/3/2009 Source: New Originator: S. Pullin TUOI: ANO-1-LP-RO-DHR Objective: 23 Point Value: I
- Section: 4.2 Type: Generic APE System Number: 025 System
Title:
Loss of Residual Heat Removal System
Description:
Knowledge of the operational implications of the following concepts as they apply to a Loss of Residual Heat Removal System: Loss of RHRS during all modes of operation.
K/A Number: AKI.01 CFR
Reference:
41 .8/41.10/45.3 Tier: I RO Imp: 3.9 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 4.3 SRO Select: Yes Taxonomy: a Question: RO:1 SRO:1 6 Given:
- The RCS is drained to 374 feet for seal replacement.
- RCS Temperature 140 F.
- RCS pressure is 10 psig.
- RCS leakage measured at 50 gpm.
- K Decay Heat Pump has been stopped and CV-1 050 Decay Heat Suction Valve has been closed per 1203.028, Loss of Decay Heat Removal AOP.
Per 1203.028, Loss of Decay Heat Removal AOP, what is the preferred makeup flow path for these conditions?
A. Gravity feed from the BWST.
B. Low Pressure Injection.
C. Spent Fuel Cooling Pump P-40A.
D. Borated Water Recirc Pump P-66.
Answer:
B. Low Pressure Injection.
Notes:
A. Gravity feed from the BWST is incorrect because the RCS is pressurized B. Low Pressure Injection is correct.
C. Spent Fuel Cooling Pump P-40A is incorrect because it is the least preferred method allowed.
D. P-66 is incorrect because it is low on the perfered list.
References:
1203.028 Change 21 History:
New for the RO/SRO 2010 exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0395 Rev: 0 Rev Date: 11/21/00 Source: Direct Originator: D.Slusher TUOI: A1LP-RO-NNI Objective: 14 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 027 System
Title:
Pressurizer Pressure Control Malfunction and the
==
Description:==
Knowledge of the interrelations between the Pressurizer Pressure Control Malfunction following: Controllers and positioners.
KIA Number: AK2.03 CFR
Reference:
41.7 / 45.7 Tier: 1 RO Imp: 2.6 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 2.8 SRO Select: Yes Taxonomy: C Question: RO:1 7 SRO:J The plant is shutdown and cooled down.
RCS pressure is 220 psig.
l&C is performing calibration checks on A RPS channel.
Why will l&C request the Pzr Control Pressure Selector, HS-1 038, be placed in the Y position?
A. To allow remote indications to be checked during calibration.
B. To prevent the ERV opening, causing a rapid depressurization of the RCS.
C. To maintain pressurizer heaters off during testing.
D. To allow the ERV low setpoint to be calibrated.
Answer:
B. To prevent the ERV opening, causing a rapid depressurization of the RCS.
Notes:
Answer [b] is correct, testing will cause ERV to open since the LTOP setpoint is in effect.
Answer [a] is incorrect, the selector switch does not select between local and remote indications.
Answer [c] is incorrect, PZR heaters are in manual control for cooldown.
Answer [d] is incorrect, l&C verifies the setpoint, it is undesirable to operate ERV at this point.
References:
1105.006, Chg. 010 STM 1-69, Rev. 13 History:
Direct from regular exambank QID#5545 for 2001 ROISRO Exam.
Selected for 2005 RO exam, replacement question.
Selected for the RO/SRO 2010 exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0582 Rev: 0 Rev Date: 9/3/2009 Source: New Originator: S. Pullin TUOI: AILP-RO-EFIC Objective: 26 Point Value: I Section: 4.1 Type: Generic EPEs System Number: 029 System
Title:
Anticipated Transient Without SCRAM (ATWS)
==
Description:==
Ability to determine or interpret the following as they apply to the ATWS: Reactor trip alarm.
KIA Number: EA2.02 CFR
Reference:
43.5 / 45.13 Tier: I RO Imp: 4.2 RO Select: Yes Difficulty: 4 Group: 1 SRO Imp: 4.4 SRO Select: Yes Taxonomy: An Question: Ro:j 8 SRO:1 8 Given:
- Plant startup is in progress.
- Reactor power is 20%.
- Total Main FW flow is 1.6 x e 6 Ibm/hr.
- Generator load is 1 80 Mwe.
Subsequently the following indications are observed:
- Reactor power dropping rapidly,
- Turbine Generator Lockout alarm is in,
- EFW actuated on both trains.
Which of the following annunciators, and reasons for the annunciator, could cause the above indications?
A. K08-A3 REACTOR TRIP because the in-service MFW pump has tripped causing a reactor trip with power >9%.
B. K08-F2 CRD MOTOR POWER FAILURE because a loss of transformer X8 has tripped the Regulating Groups.
C. K08-A5 AMSAC TRIP because both Gamma Metrics NI-501 and Nl-502 were not calibrated within 3% of heat balance as required.
D. K08-A3 REACTOR TRIP because the RPS anticipatory trip for Turbine has not been reset.
Answer:
C. K08-A5 AMSAC TRIP because both Gamma Metrics Nl-501 and Nl-502 were not calibrated within 3% of heat balance as required.
Notes:
A. Is incorrect because a reactor trip would have caused all of the control rods to insert not just the regulating groups.
B. Is incorrect because a loss of X8 would only lose one of the AC power supplies to the rods and no rods would trip.
C. Is correct, if gamma metrics indicated >45% with the given feedwater flow, an AMSAC Trip would be initiated.
D. Is incorrect because a reactor trip would have caused all of the control rods to insert not just the regulating groups.
References:
1102.002 Change 082 STM 1-59 Rev. 1
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
1102.002 Change 082 STM 1-59 Rev. 1 History:
New for the ROISRO 2010 exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0364 Rev: 0 Rev Date: 11/8/00 Source: Direct Originator: J.Cork TUOI: Al LP-RO-EOPO6 Objective: I Point Value: 1 Section: 4.1 Type: Generic EPEs System Number: 038 System
Title:
Steam Generator Tube Rupture
==
Description:==
Knowledge of EOP mitigation strategies K/A Number: 2.4.6 CFR
Reference:
41.10/43.5/45.13 Tier: 1 RO Imp: 3.7 RO Select: Yes Difficulty: 4 Group: 1 SRO Imp: 4.7 SRO Select: Yes Taxonomy: An Question: RO:I 9 SRO:
After a reactor trip, the following indications are observed:
- Makeup Tank level has lost 5 inches in the last 5 minutes
- RB and Aux. Bldg. Sump levels are stable
- A OTSG EFIC level is 35 and rising
- B OTSG EFIC level is 31 and stable
- A MFW Flow is 0.1 mlb/hr
- B MFW Flow is 0.3 mlb/hr Which of the following actions would be required to minimize the threat of a potential radioactive release to the public?
A. Initiate HPI per RT-2 B. Cooldown and isolate the B SG C. Cooldown and isolate the A SG D. Commence a rapid RCS cooldown at 240 °FIhr Answer:
C. Cooldown and isolate the A SG Notes:
Answer [c] is correct, the SG level parameters indicate a rupture on the A SG and a cooldown should be commenced to reduce RCS temperature to <500 F to minimize the possibility of lifting a secondary safety on the A SG.
[a] is incorrect, the leak size is about 30 gpm (30.86 gal/in. x 5 in./5 mm.). This is within the capacity of normal makeup.
[b] is incorrect, a cooldown and isolation is required but not on this SG.
[d] is incorrect, a rapid cooldown at this rate is not required until overfilling of ruptured SG is imminent.
References:
1202.006, Chg. 11 History:
Created for 2001 RQ/SRO Exam.
Selected for 2002 RO/SRO exam.
Selected for 2005 Jon Gray RO re-exam.
Selected for 2010 RO/SRO Exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0551 Rev: 0 Rev Date: 3-30-05 Source: Direct Originator: J.Cork TUOI: Al LP-RO-EOPO3 Objective: 10 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 040 System
Title:
Steam Line Rupture
==
Description:==
Knowledge of the operational implications of the following concepts as they apply to the Steam Line Rupture: Consequence of PTS.
K/A Number: AKI.0l CFR
Reference:
41.8/41.10/45.3 Tier: 1 RO Imp: 4.1 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 4.4 SRO Select: Yes Taxonomy: C Question: RO: 10 SRO:J 0 Which of the following would invoke Pressurized Thermal Shock (PTS) limits during a Steam Line Rupture?
A. HPI on with all RCPs off B. RCS cool down rate 105°F/hr with Tcold 360°F C. RCS cool down rate 55°F/hr with Tcold 310°F D. SG Tube to shell DT 150°F tubes colder Answer:
A. HPI on with all RCPs off Notes:
Answer A is correct per RT-1 4.
Answer B is incorrect, cooldown rate is >1 00°F/hr but Tcold >355°F.
Answer C is incorrect, cooldown rate is >50°F/hr but Tcold >300°F.
Answer D is incorrect, this is a limit but not a PTS limit.
References:
1202.012, Chg. 8 History:
New for 2005 RO exam.
Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0774 Rev: 0 Rev Date: 9/4/2009 Source: Modified Originator: S Pullin TUOI: Al LP-RO-EOPO2 Objective: 8 Point Value: I Section: 4.2 Type: Generic APEs System Number: 054 System
Title:
Loss of Main Feedwater (MFVV)
==
Description:==
Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MF.N): AFW adjustments needed to maintain proper T-ave and S/G level.
K/A Number: AA2.06 CFR
Reference:
43.5/45.13 Tier: 1 RO Imp: 4.0 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 4.3 SRO Select: Yes Taxonomy: C Question: RO:j 11 SRO:
A reactor trip has occurred from 100% power due to a loss of both MFW Pumps.
The following conditions have existed for three minutes:
- CET temperature = 580 degrees F.
- RCS pressure = 1600 psig.
Which of the following operator actions will be performed?
A. Trip all running RCPs.
B. Verify EFW flow to each Steam Generator is 320 gpm.
C. Verify Reflux Boiling setpoint is selected on both EFIC trains.
D. Verify EFW in hand and flow to each Steam Generator is 570 gpm.
Answer:
C. Verify Reflux Boiling setpoint is selected on both EFIC trains.
Notes:
A. Incorrect, this would be done for loss of subcooling margin but only if <2 minutes had expired without tripping the RCPs.
B. Incorrect this is done for loss of subcooling margin but only if EFW flow is less than adequate and the value given is similar but less than the minimum flow rate of greater than or equal to 340 gpm.
C. Correct since subcooling margin is lost and the Reflux Boiling setpoint is required to be selected in this situation.
D. Incorrect, this would be done if only one SG was available.
References:
1202.012 Change 008, RT-5 History:
Modified from QID 368.
Selected for the RO/SRO 2010 exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0496 Rev: 0 Rev Date: 12/8/2003 Source: Direct Originator: NRC TUOI: ELP-NLO-ELEC1 Objective: 29 Point Value: I Section: 4.1 Type: Generic EPEs System Number: 055 System
Title:
Station Blackout when
==
Description:==
Ability to operate and monitor the following as they apply to a Station Blackout: Battery, approaching fully discharged.
KIA Number: EA1 .05 CFR
Reference:
41.7 / 45.5 / 45.6 Tier: I RO Imp: 3.3 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 3.6 SRO Select: Yes Taxonomy: C Question: RO:I 12 SRO: I 2 Unit I has been in a station black-out for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> with battery bank D06 supplying bus D02 with power without a battery charger online for this entire time.
If the equipment on bus D02 does NOT change, which one of the following statements describes the batterys discharge rate (expressed as amperage) as the battery is expended?
A. The battery amperage will be fairly constant until the design battery capacity is exhausted.
B. The battery amperage will drop steadily until the design battery capacity is exhausted.
C. The battery amperage will rise steadily until the design battery capacity is exhausted.
D. The battery amperage will be fairly constant until the design battery capacity is exhausted and then will rapidly drop.
Answer:
C. The battery amperage will rise steadily until the design battery capacity is exhausted.
Notes:
P=lE; As the battery discharges under a constant load, battery voltage will drop and current (battery amperage) will rise.
References:
ELP-NLO-ELECI History:
Developed by NRC.
Used on 2004 RO/SRO Exam.
Selected for 2005 Jon Gray RO re-exam.
Selected for the RO/SRO 2010 exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0366 Rev: 0 Rev Date: 1/8/00 Source: Direct Originator: J.Cork TUOI: Al LP-RO-ESAS Objective: 5 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 056 System
Title:
Loss of Offsite Power
==
Description:==
Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Order and time to initiation of power for the load sequencer.
K/A Number: AK3.0l CFR
Reference:
41.5, 41.10 / 45.6 / 45.13 Tier: 1 RO Imp: 3.5 RO Select: Yes Difficulty: 2 Group: 1 SRO Imp: 3.9 SRO Select: Yes Taxonomy: K Question: RO: 13 SRO:1 3 An electrical storm has caused a Degraded Power situation with a spurious ES actuation of the even channels.
In which order will the following ES components be started automatically?
A. SW pump, HPI pump, LPI pump, RB Spray pump B. HPI pump, SW pump, LPI pump, RB Spray pump C. SW pump, HPI pump, RB Spray pump, LPI pump D. HPI pump, LPI pump, SW pump, RB Spray pump Answer:
D. HPI pump, LPI pump, SW pump, RB Spray pump Notes:
Answer [d] lists the correct order of load sequence with loss of offsite power and ES actuation.
The others are incorrect sequences of the correct components.
References:
1305.006, Chg. 030 History:
Created for 2001 RO/SRO Exam.
Selected for 2005 Jon Gray RO re-exam.
Selected for the 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0624 Rev: 0 Rev Date: 11/2/05 Source: Direct Originator: J.Cork TUOI: Al LP-RO-NNI Objective: 7 Point Value: I Section: 4.2 Type: Generic APEs System Number: 057 System
Title:
Loss of Vital AC Electrical Instrument Bus Loss of Vital AC Instrument
==
Description:==
Ability to determine and interpret the following as they apply to the Bus: S/G pressure and level meters.
KIA Number: AA2.05 CFR
Reference:
43.5 /45.13 I RO Imp: 3.5 RO Select: Yes Difficulty: 3 Tier:
Group: 1 SRO Imp: 3.8 SRO Select: Yes Taxonomy: C Question: RO: 14 SRO:I What would be the effect on the SG pressu re and level instruments on C03, if a loss of the RS-l bus occurred?
re and level A. Instrument power would automatically be transferred to YO-2 by the ABT, SG pressu instruments would not be effected.
pressure and level B. The NNI-X SI and S2 switches would open and SASS would transfer to NNI-Y, SG instruments would fail to mid scale.
pressure and level C. The NNI-X SI and S2 switches would open and SASS would transfer to NNI-Y, SG instruments would not be effected.
re and level D. Instrument power would automatically be transferred to YO-l by the ABT, SG pressu instruments would not be effected.
Answer:
re and level A. Instrument power would automatically be transferred to YO-2 by the ABT, SG pressu instruments would not be effected.
Notes:
logic power is A is correct, a loss of RS-1 would simply cause NNI-X to be powered from YO-2, -24vDC auctioneered and instrument power would transfe r by the ABT within 0.5 second s no effect on instruments.
to B is incorrect, it would take a loss of both RS-l and YO-2 to cause the SI and S2 switches open.
switch es to open.and SG NC is incorrect, t would take a loss of both RS-l and YO-2 to cause the SI and S2 r would make instruments work pressure does input to the BTU limit alarm but would not fail hgh due transfe correctly, D is incorrect, the alternate power to NNI-Y is from YO-l.
References:
STM 1-69, Rev. 13 History:
New for 2005 RO re-exam.
Selected for the 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0187 Rev: 1 Rev Date: 4/25/2002 Source: Direct Originator: S.Pullin TUOI: Al LP-RO-AOP Objective: 4.5 Point Value: I Section: 4.2 Type: Generic APE System Number: 058 System
Title:
Loss of DC Power
==
Description:==
Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation.
K/A Number: AKI .01 CFR
Reference:
41.8 / 41.10 / 45.3 Tier: 1 RO Imp: 2.8 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.1 SRO Select: Yes Taxonomy: C Question: RO:j15 SRO: J 5 Given the following indications at 100% power:
- Annunciator D02 UNDER VOLTAGE (K01-A8) in alarm.
- Annunciator D02 TROUBLE (K01-D8) in alarm.
- Annunciator D02 CHARGER TROUBLE (K01-E8) in alarm.
- The reactor has tripped.
- The turbine trip solenoid light is on.
- Breaker position lights on the RIGHT side of Cl 0 are off.
What are the actions required of the CBOT?
A. Trip the main generator output breakers.
B. Transfer Dl 1 to emergency supply DOl.
C. Trip all RCPs.
D. Transfer D21 to emergency supply DOl.
Answer:
D. Transfer D21 to emergency supply DOl.
Notes:
[d] is correct per 1203.036 as the conditions are indicative of a loss of D02.
[a] and [b] are incorrect due to this a loss of D02 not DOl these are actions for the loss of DOl.
[C] is incorrect due to we have not loss seal injection and seal cooling, this is an action in this procedure section if both of the before mentioned system functions are lost
References:
1203.036, Chg. 08 History:
Developed for use in 98 RO Re-exam Selected for use in 2002 RO/SRO exam, revised slightly.
Selected for 2005 Jon Gray RO re-exam.
Selected for the 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0281 Rev: 0 Rev Date: 9-3-99 Source: Direct Originator: D. Slusher TUOI: ANO-1-LP-RO-MSSS Objective: 3 Point Value: I Section: 4.2 Type: Generic AOP System Number: 062 System
Title:
Loss of Nuclear Service Water
==
Description:==
Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: The automatic actions (alignments) within the nuclear service water resulting from the actuation of the ESFAS.
KIA Number: AK3.02 CFR
Reference:
41.4, 41.8 I 45.7 Tier: I RO Imp: 3.6 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 3.9 SRO Select: Yes Taxonomy: C Question: RO:j 16 SRO:J 16 Service Water Pumps P-4A, P-4B (supplied from A-4), and P-4C are running.
An ES actuation channels (1-10) coincident with a loss of off-site power occurs.
Which service water pumps will autostart when A-3 and A-4 are re-energized and for what reason?
A. P-4A, P-4B and P-4C, due to high service water loads with all 10 channels actuated B. P-4A and P-4B, due to both being supplied from A-4 and #2 EDG tied on first C. P-4B and P-4C, due to B service water pump is the swing pump and its perferred to be running D. P-4A and P-4C, due to 3 service water pumps running prior to event to prevent EDG overloading Answer:
D. P-4A and P-4C, due to 3 service water pumps running prior to event to prevent EDG overloading Notes:
When ESAS actuates and the buses are re-energized the P-4A and P-4C handswitch position will interlock P 4B and keep P-4B from starting. Therefore, a, b, and c responses are incorrect.
References:
STM 1-42, Rev. 18, Service and Auxiliary Cooling Water, page 13, 14, 15 History:
Developed for 1999 exam.
Used in 2001 ROISRQ Exam.
Selected for the 2010 ROISRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0335 Rev: 0 Rev Date: 9-7-99 Source: Direct Originator: D. Slusher TUOI: ANO-1-LP-RO-EOPO4 Objective: 6 Point Value: I Section: 4.3 Type: B&W EPE/APE System Number: E04 System
Title:
Excessive Heat Transfer
==
Description:==
Ability to operate and I or monitor the following as they apply to the (Inadequate Heat Transfer):
Desired operating results during abnormal and emergency situations.
KIA Number: EAI .3 CFR
Reference:
CFR: 41.7 I 45.5 / 45.6 Tier: I RO Imp: 3.6 RO Select: Yes Difficulty: 2.5 Group: I SRO Imp: 3.8 SRO Select: Yes Taxonomy: C Question: RO:j 17 SRO: j 7 Given:
- Loss of all Feedwater
- HPI core cooling started What indications would you monitor to ensure adequate HPI core cooling?
A. CET temperatures stable at 100 minutes.
B. T-cold tracking associated SG T-sat.
C. T-hot tracking CET temperatures.
D. T-hotlT-cold differential temperature dropping.
Answer:
A. CET temperatures stable after 100 minutes.
Notes:
A is correct since the only criteria for evaluation of adequacy of core cooling via HPI is a decrease in CET temps.
B, C, and D are individual indications of adequate primary to secondary heat transfer.
References:
1202.004 Change 6 History:
Developed for 1999 exam.
Used on 2004 ROISRO Exam.
Selected for the 2010 ROISRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0775 Rev: 0 Rev Date: 9/8/2009 Source: New Originator: S. Pullin TUOI: Al LP-RO-GEN Objective: 7 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 077 System
Title:
Generator Voltage and Electrical Grid Disturbances
==
Description:==
Ability to interpret reference materials, such as graphs, curves, tables, etc.
K/A Number: 2.1.25 CFR
Reference:
41 .10/43.5/45.12 Tier: 1 RO Imp: 3.9 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 4.2 SRO Select: Yes Taxonomy: C Question: 8 SRO:
RO:1 REFERENCE PROVIDED Given:
Plant at 100% power Generator output 88OMWe Electrical storm caused a grid disturbance The Dispatcher calls Control Room and requests Unit I Generator be operated in the lagging mode at 180 Megavars.
What is the power factor for the above information?
A. 0.935 PF B. 0.955PF C. 0.97PF D. 0.98 PF Answer:
D. 0.98PF Notes:
Using Attachment N of Op-I 102.004 D. is correct A, B and C are associated with different generator loads.
References:
1102.004 Change 048 History:
Developed for the 2010 RO/SRO exam.
ES-401 PWR Examination Outline Form ES-401-2 5-401 PWR Examination Outline Form ES-401 -2 Emer enc and Abnormal Plant Evolutions Tier 1!Grou 2 RO E/APE # I Name / Safety Function K K K A A G K/A Topic(s) IR # QID Typ 12312 e 000001 Continuous Rod Withdrawal /1 Not selected NIA 000003 Dropped Control Rod /1 Not selected N!A 000005 Inoperable/Stuck Control Rod /1 Not selected N!A 000024 Emergency Boration /1 Not selected NIA 000028 Pressurizer Level Malfunction /2 x AK1 .01 PZR reference leak 2.8* 19 776 N abnormalities.
000032 Loss of Source Range NI /7 X AA2.04 Satisfactory source-range /
3.1 20 777 N intermediate-ran e overla 000033 Loss of Intermediate Range NI / 7 Not selected N!A 000036 (BW/A08) Fuel Handling Accident AK2.1 Changed to randomly selected NIA
/8 S stem 068 AK2.07 AAI.10 CVCS makeup tank level 000037 Steam Generator Tube Leak /3 2.9 21 778 N indicator 000051 Loss of Condenser Vacuum / 4 Not selected N!A 000059 Accidental Liquid RadWaste Rel.! Not selected N/A 9
000060 Accidental Gaseous Radwaste AKI .04 Changed to randomly N!A Rel./9 selected S tern 028 AKI .01 000061 ARM System Alarms / 7 x AK3.02 Guidance contained in alarm 3.4 22 634 D res onse for ARM s stem.
000067 Plant Fire On-site / 8 x AKI.02 Fire Fi htin 3.1 23 695 DR 000068 BW/A06 Control Room Evac. /8 x AK2.07 - ED/G 3.3 24 779 0 000069 (W!E14) Loss of CTMT Integrity / Not selected N/A 5
000074 (W/E06&E07) Inad. Core Cooling Not selected NIA 4
000076 High Reactor Coolant Activity! 9 Not selected NIA WIEO1 & E02 Rediagnosis & SI Not selected NIA Termination / 3 W/E13 Steam Generator Over-pressure / Not selected N!A 4
W/E15 Containment Flooding / 5 Not selected N/A W/E16 High Containment Radiation / 9 Not selected NIA BW/A01 Plant Runback!1 Not selected NIA BW!A02&A03 Loss of NNI-XIY! 7 Not selected NIA BW/A04 Turbine Trip / 4 Not selected NIA BW/A05 Emergency Diesel Actuation /6 x AK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual 4.0 25 349 D features.
AA2.2 Adherence to appropriate BW/A07 Flooding / 8 procedures and operation within the 3.3 26 780 N limitations in the facilitys license and amendments BW/E03 Inadequate Subcooling Margin / Not selected NIA 4
ES-401 PWR Examination Outline Form ES-401-2 BW/E08; W/E03 LOCA Cooldown - Not selected e.ress. /4 1E09; CE/A13; WIEO9&ElO Natural V Not selected Circ./4 IV 2.2.22- Knowledge of limiting conditions BW/E13&E14 EOP Rules and Enclosures x for operations and safety limits.
CE/All; W/E08 RCS Overcooling PTS /
- Not selected 4
CE/Al 6 Excess RCS Leakage / 2 Not selected V CE/E09 Functional Recovery Not selected K/A Cate.o Point Totals: 121 Group Point Total:
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0776 Rev: 0 Rev Date: 9/8/2009 Source: New Originator: S. Pullin TUOI: ASLP-RO-CMPO2 Objective: 9a Point Value: I Section: 4.2 Type: Generic APEs System Number: 028 System
Title:
Pressurizer (PZR) Level Control Malfunction
==
Description:==
Knowledge of the operational implications of the following concepts as they apply to pressurizer level control malfunctions: PZR refere.nce leg abnormalities.
K/A Number: AKI.01 CFR
Reference:
41 .8/41 .10/45.3 Tier: I RO Imp: 2.8 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.1 SRO Select: Yes Taxonomy: C Question: RO:J 19 SRO:I 19 Given:
Plant at 100% power Leak develops on the pressurizer reference leg What effect does this have on level indication and pressurizer level control valve, CV-1235?
A. Indicated level decreases and pressurizer level control valve, CV-1235, opens to control level.
B. Indicated level decreases and pressurizer level control valve, CV-1235, fails as is.
C. Indicated level increases and pressurizer level control valve, CV-1 235, fails as is.
D. Indicated level increases and pressurizer level control valve, CV-1235, closes to control level.
Answer:
D. Indicated level increases and pressurizer level control valve, CV-1 235, closes to control level.
Notes:
D. is correct, a leak in the reference leg would cause indicated level to increase. As a result of the level rise CV-1235 will close in orederto maintain level at setpoint.
A, B, and C are incorrect, using the different possible combinations. CV-1235 fails as is on a loss of Instrument Air not on a fialure of the reference leg.
References:
ASLP-RO-CMPO2 Rev 2 History:
New selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0777 Rev: 0 Rev Date: 9/8/2009 Source: New Originator: S. Pullin TUOJ: AILP-RO-NOP Objective: 4 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 032 System
Title:
Loss of Source Range Nuclear Instrumentation
==
Description:==
Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: Satisfactory source-range intermediate-range overlap K/A Number: AA2.04 CFR
Reference:
43.5/45.13 Tier: 1 RO Imp: 3.1 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 3.5 SRO Select: Yes Taxonomy: C Question: RO:J 20 SRO: I Given:
Source Range 5 E 4 counts Intermediate Range 1 X E -9 amps During the startup, the source range instruments fail to 3 counts per second.
What is the required operator action for the given condition?
A. Immediately suspend operations involving positive reactivity changes..
B. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> verify CRD trip breakers open.
C. Continue the startup..
D. Immediately initiate a shutdown and insert all control rods.
Answer:
C. Continue the startup..
Notes:
- c. is correct, procedure allows continuing with startup if intermediate range indicate >10 -10 amps.
A, B and D are incorrect due to these are the actions to take when both source range instruments fail and both intermediate range channels indicate <10 -10 amps.
References:
1203.021 Change 10 History:
New for the RO/SRO 2010 exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0778 Rev: 0 Rev Date: 9/8/2009 Source: New Originator: S. Pullin TUOI: Al LP-RO-ALEAK Objective: 3 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 037 System
Title:
Steam Generator (S/G) Tube Leak
==
Description:==
Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: CVCS makeup tank level indicator.
K/A Number: AA1.l0 CFR
Reference:
41 .7/45.5/45.6 Tier: I RO Imp: 2.9 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.1 SRO Select: Yes Taxonomy: K Question: RO:j 21 SRO:1 21 Given:
Plant at 100% power Makeup Tank level dropping at 1 inch every 2 minutes.
A OTSG N-i 6 TROUBLE (K07-A5)
PROC MONITOR RADIATION HI (Kl0-82)
What is the A OTSG Tube Leak rate?
A. 10.2 gpm
- 8. 15.4 gpm C. 20.4 gpm D. 30.8 gpm Answer:
B. 15.4 gpm Notes:
B. 15.4 gpm is correct based on makeup tank level is 30.86 gallons per inch, at a rate of change of 1 inch per 2 minutes equals 15.4 gpm leak.
A, C and D are incorrect.
References:
1203.039 Change 011 History:
New for the RO/SRO 2010 exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0634 Rev: 0 Rev Date: 11/8/05 Source: Direct Originator: J.Cork TUOI: AILP-RO-RMS Objective: 7 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 061 System
Title:
Area Radiation Monitoring (ARM) System Alarms
==
Description:==
Knowledge of the reasons for the following responses as they apply to the Area Radiation Monitoring (ARM) System Alarms: Guidance contained in alarm response for ARM system.
K/A Number: AK3.02 CFR
Reference:
41.5, 41.10 /45.6 /45.13 Tier: 1 RO Imp: 3.4 RO Select: Yes Difficulty: 4 Group: 2 SRO Imp: 3.6 SRO Select: Yes Taxonomy: C Question: RO:j 22 SRO: I Given:
- AREA MONITOR RADIATION HI (K10-B1) in alarm
- RADIATION MONITOR TROUBLE (Kb-Cl) in alarm In accordance with the alarm response procedure, the area monitors on C25 Bay 3 must be inspected.
What reason(s) would cause both alarms above to come into alarm?
A. WARNING and POWER ON lights on B. POWER ON light off C. HIGH ALARM light on and POWER ON light off D. FAILURE light on Answer:
B. POWER ON light off Notes:
B is correct, a loss of power will cause both the Hi Radiation and Trouble annunciators to come in.
A is incorrect, this would cause the Hi Radiation but not the Trouble annunciator.
C is incorrect, the POWER ON light off will cause both annunicators but the HIGH ALARM light will not be on with a loss of power.
D is incorrect, this will cause the Trouble annunciator but not the Hi Radiation annunciator.
References:
1203.0121, Chg. 046 STM 1-62, Rev. 11 History:
New for 2005 RO re-exam.
Selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0695 Rev: I Rev Date: 4/1/2008 Source: Repeat Originator: Steve Pullin TUOI: ASLP-R0-FRHAZ Objective: 4B Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 067 System
Title:
Plant Fire on Site
==
Description:==
Knowledge of the Operational implications of the following concepts as they apply to plant fire on site: fire fighting.
KIA Number: AK1.02 CFR
Reference:
41 .8/41 .10/45.3 Tier: 1 RO Imp: 3.1 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.9 SRO Select: Yes Taxonomy: K Question: RO:j 23 SRO:J Per 1015.007, Fire Brigade Organization and Responsibilities, which of the following describes the Ops Manning composition of the Fire Brigade for the initial response to a fire on Unit 1?
A. Unit 1 supplies the Fire Brigade Leader, Unit 2 supplies 3 Fire Brigade members, Security supplies one support member.
B. Unit I supplies the Fire Brigade Leader and 2 Fire Brigade members, Unit 2 supplies 1 Fire Brigade member, Security supplies one support member.
C. Unit 2 supplies the Fire Brigade Leader, Unit 1 supplies 3 Fire Brigade members, Security supplies one support member.
D. Unit 2 supplies the Fire Brigade Leader and 1 Fire Brigade member, Unit I supplies 2 Fire Brigade members, Security supplies one support member.
Answer:
A. Unit I supplies the Fire Brigade Leader, Unit 2 supplies 3 Fire Brigade members, Security supplies one support member Notes:
A is correct per the requirements of 101 5.007 B is incorrect. This answer was previously correct for a fire on Unit I prior to the latest revision.
C is incorrect. This is correct for a fire on Unit 2 D is incorrect This answer was previously correct for a fire on Unit 2 prior to the latest revision.
References:
101 5.007, Fire Brigade Organization and Responsibility Chg. 019 History:
Selected for 2008 RO Exam Selected repeat for the 2010 RO/SRO exam
INITIAL RO!SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0779 Rev: 0 Rev Date: 9/8/2009 Source: Direct Originator: S. Pullin TUOI: ANO-1-LP-RO-EDG Objective: 26 Point Value: I Section: 4.2 Type: Generic APEs System Number: 068 System
Title:
Control Room Evacuation
==
Description:==
Knowledge of the interrelations between the Control Room Evacuation and the following: ED/G K/A Number: AK2.07 CFR
Reference:
41.7/45.7 Tier: 1 RO Imp: 3.3 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.4 SRO Select: Yes Taxonomy: K Question: RO: 24 SRO:
Given:
Fire has occurred in the Cable Spread Room Performing 1203.002 Alternate Shutdown CRS follow-up actions are in progress
What condition can automatically trip the Emergency Diesel Generators?
A. Positive crankcase pressure trip B. Low lube oil pressure trip C. Mechanical over speed trip D. De-energized Governor Run Solenoid Answer:
C. Mechanical over speed trip Notes:
C will mechanically trip the fuel rack.
A and B require DC power to the emergency trip relay.
D is overriden on a No DC start
References:
1104.036, Emergency Diesel Generator Operation, Change 049 History:
Direct Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0349 Rev: 0 Rev Date: 9-7-99 Source: Direct Originator: D. Slusher TUOI: ANO-1-LP-RO-ELEC Objective: 1IJ Point Value: I Section: 4.3 Type: B&W EOP/AOP System Number: A05 System
Title:
Emergency Diesel Actuation.
Description:
Knowledge of the interrelations between the (Emergency Diesel Actuation) and the following:
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
K/A Number: AK2.1 CFR
Reference:
CFR: 41.7 /45.7 Tier: I RO Imp: 4.0 RO Select: Yes Difficulty: 3 Group: 3 SRO Imp: 3.8 SRO Select: Yes Taxonomy: C Question: RO:J 25 25 SRO:j Diesel Generator #1 is running for a surveillance test.
Low reactor coolant system pressure causes a reactor trip and ESAS actuation.
What will the ES Electrical response be?
A. A-3 and A-4 powered from SU #1, both diesel generators running unloaded.
B. A-3 and A-4 powered from SU #1, Diesel Generator # I tripped, Diesel Generator # 2 running unloaded.
C. A-3 powered from Diesel Generator #1, A-4 powered from SU #1, Diesel Generator # 2 running unloaded.
D. A-3 powered from Diesel Generator #1, and A-4 powered from Diesel Generator #2.
Answer:
A. A-3 and A-4 powered from SU #1, both diesel generators running unloaded.
Notes:
A is correct, electrical response should be the normal response for an ESAS.
B is incorrect, nothing should trip #1 EDG.
C is incorrect, the #1 EDG output breaker should open on an ES signal.
D is incorrect, both busses should be powered from SU #1.
References:
STM 1-32, Rev. 33 History:
Used in 1999 exam.
Modified from ExamBank, QID# 453.
Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0780 Rev: 0 Rev Date: 9/09/2009 Source: New Originator: S.Pullin TUOI: A1LP-RO-AOP Objective: 4 Point Value: I Section: 4.3 Type: B&W EPEs/APEs System Number: A07 System
Title:
Flooding
==
Description:==
Ability to determine and interpret the following as they apply to the (flooding): adherence to appropriate procedures and operation within the limitations in the facilities license and amendments.
K/A Number: AA.2.2 CFR
Reference:
43.5/45.13 Tier: 1 RO Imp: 3.3 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 3.7 SRO Select: Yes Taxonomy: C Question: RO:j 26 SRO:J 26 Given:
Plant power 100%
A Decay Heat pump QOS Dardanelle Lake Level 350 feet rising 1 ft/hr due to heavy rains Corps of Engineers predicts peak flood levels will reach 355 feet What action is required per Natural Emergencies procedure 1203.025 section 4 Flood?
A. Perform Rapid Plant Shutdown and align B Decay Heat pump for Decay Heat B. Perform Rapid Plant Shutdown and transfer plant auxiliaries to SU 2 transformer C. Trip Reactor and perform a Forced flow Cool Down.
D. Trip Reactor and perform a Natural Circulation Cool Down.
Answer:
B. Perform Rapid Plant Shutdown and transfer plant auxiliaries to SU 2 transformer Notes:
B. is correct due to 1203.025 directs you to perform a shutdown, and SU2 transformer is designed for flooding and should be used during a flood A. is incorrect 1203.025 directs you to perform a shutdown, and align a LPI pump for DH if both pumps are operable in this case A DH pump is QOS C. is incorrect the procedure does not call for a reactor trip but you should use rapid plant shut down and forced flow cool down D. is incorrect the procedure does not call for a reactor trip but you should use rapid plant shut down and forced flow cool down not Natural Circulation CID
References:
Natural Emergencies 1203.025 change 028 History:
New for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0595 Rev: 0 Rev Date: 9/09/2009 Source: New Originator: S.Pullin TUOI: A1LP-RO-RCS Objective: 26 Point Value: 1 Section: 4.3 Type: B&W EPEs/APEs System Number: E13 System
Title:
EOP Rules and Enclosures
==
Description:==
Knowledge of limiting conditions for operation and safety limits.
K/A Number: 2.2.22 CFR
Reference:
41.5/43.2/45.2 Tier: 1 RO Imp: 4.0 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 4.7 SRO Select: Yes Taxonomy: K Question: RO: 27 SRO:
In accorrJance with Technical Specification bases, what is the purpose of the Pressurizer Code Safeties and what is the design bases accident that defines their minimum capacity?
A. The Pressurizer Code Safeties prevent exceeding the safety limit of 2500 psig during a 100% load rejection without a reactor trip.
B. The Pressurizer Code Safeties prevent exceeding the safety limit of 2750 psig during a 100% load rejection without reactor trip.
C. The Pressurizer Code Safeties prevent exceeding the safety limit of 2750 psig during a startup accident.
D. The Pressurizer Code Safeties prevent exceeding the safety limit of 2500 psig during a startup accident.
Answer:
C. The Code Safeties prevent exceeding the safety limit of 2750 psig during a startup accident.
Notes:
Answer C is correct, it lists the proper safety limit and the design basis accident.
Answer A is incorrect, it lists the safety setpoint (not the safety limit) and a plausible, but incorrect, accident.
Answer B is incorrect, it lists the proper safety limit and a plausible, but incorrect, accident.
Answer D is incorrect, it lists the safety setpoint (not the safety limit) and the design basis accident.
References:
Technical Specifications bases B2.1 .2 amendment # 215 History:
New for 2010 RO/SRO exam
Form ES-401-2 ES-401 PWR Examination Outline PWR Examination Outline Form ES-401-2 Plant Systems Tier 2/Group I (RO)
K/A Topic(s) 1(5.05 The dependency of 003 Reactor Coolant Pump RCS flow rates upon the number of operating RCPs A4.08 RCP cooling water supplies 2.1.34 changed to 2.2.38 004 Chemical and Volume Knowledge of conditions and Control limitations in the facility license 1(4.03 Protection of ion exchangers (high letdown temperatures will isolate ion exchangers) 005 Residual Heat Removal 1(2.01 RHR Pumps 006 Emergency Core Cooling 1(6.10 Valves 007 Pressurizer Relief/Quench 1(5.02 Method of forming a Tank steam bubble in the PZR 008 Component Cooling Water A2.08 changed to A2.01 -
Loss of CCW Pum.
10 Pressurizer Pressure K3.02 RPS ontrol 012 Reactor Protection K6.10 Permissive circuits 2.1.32 Ability to explain and apply system limits and precautions 013 Engineered Safety Features K4.10 Safeguards Actuation equipment control reset 022 Containment Cooling A3.01 Initiation of safeguards mode of operation 025 Ice Condenser Not Selected 026 Containment Spray 1(1.01 - ECCS 039 Main and Reheat Steam I: A2.04 dump Malfunctioning steam 059 Main Feedwater A3.03 Feed water pump suction flow pressure 1(4.16 Automatic trips for MFW pumps 061 Auxiliary/Emergency Al .04 changed to Al .01 Feedwater S/G level ES-401 Form ES-401 -2
PWR Examination Outline Form ES-401 2 ES-401 PWR Examination Outline Form ES-401 -2 Plant S stems Tier 2/Grou I RO KIA Topic(s) IR # ID Ty K K K K K K A A A A G e
1 234561234 X 2.4.35 Knowledge 3.8 46 790 N 062 AC Electrical Distribution of local auxiliary operator tasks during an emergency and the resultant operational effects.
3.3 47 316 D K2.01 Major s stem loads 1(3.02 3.5 48 86 D 063 DC Electrical Distribution Components using DC control ower 1(2.01 Air 2.7 49 791 N 064 Emergency Diesel Generator X compressor N 3.4 50 792 X
1(1.05 Starting air 5 stern 1(5.01 Radiation 2.5 51 672 R 073 Process Radiation Monitoring X theory, including sources, types, units, and effects X A4.02 SWS valves 2.6 52 793 D 076 Service Water Al .02 Reactor and 2.6 53 794 N X
turbine building closed cooling water tern eratures Kl.03 changed to 2.7 54 535 D 078 lnstrumentAir X KI .02 Service air X A4.06 Operation of 2.7 55 795 D 103 Containment the containment ersonnel airlock 3 2 2 2 2 3 3 Grou Point Total: 28 K/A Catego Point Totals: 3 3 2 3 Form ES-401-2 ES-401
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0107 Rev: 1 Rev Date: 12/7/00 Source: Direct Originator: JCork TUOI: A1LP-RO-ICS Objective: 22 Point Value: I Section: 3.4 Type: Heat Removal From Reactor Core System Number: 003 System
Title:
Reactor Coolant Pump t as they apply to the RCP:
==
Description:==
Knowledge of the operational implications of the following concep The dependancy of RCS flow rates upon the numbe r of operati ng RCPs.
K/A Number: K5.05 CFR
Reference:
41.5 / 45.7 RO Imp: 2.8 RO Select: Yes Difficulty: 3 Tier: 2 Group: I SRO Imp: 3.0 SRO Select: Yes Taxonomy: C Question: RO:I 28 SRO: j 28 The plant is operating at 60% power with Delta Tc and SGIRX Master stations in Hand.
All other ICS stations are in Auto.
If one RCP has to be tripped due to high vibration, how will the ICS respond?
(Assume no operator action other than tripping the RCP.)
A. The ICS will runback the plant to 45% load at 50%/mm.
B. No change to FW will occur since the SG/RX Master is in Hand.
C. Demand is less than the RCP runback limit, no changes occur to FW.
D. The RC flow difference will re-ratio the FW flow demand.
Answer:
D. The RC flow difference will re-ratio the FW flow demand.
Notes:
(d) is correct. Answer (a) is Following an RCP trip Delta Tc will re-ratio feeciwater demands, therefore answer (c) are incorrect because they incorrect since the plant is operating below the runback setpoint, while (b) and state that ICS will not re-ratio feedwater demands.
References:
1203.012F, Chg. 028 STM 1-64, Rev.l1 History:
Modified QID 4408 for use on 1998 RO/SRO Exam.
Modified for use in 2001 RO Exam.
Selected for 2005 Jon Gray RO re-exam.
Selected for 2010 RO/SRQ exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0782 Rev: 0 Rev Date: 9/09/2009 Source: Modified Originator: S. Pullin TUOI: AILP-RO-RCS Objective: 23 Point Value: I Section: 3.4 Type: RCS Heat Removal System Number: 003 System
Title:
Reactor Coolant Pump
==
Description:==
Ability to manually operate and/or monitor in the control room: RCP cooling water supplies KIA Number: A4.08 CFR
Reference:
41.7/45.5 to 45.8 Tier: 2 RO Imp: 3.2 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 2.9 SRO Select: Yes Taxonomy: C Question: RO: 29 SRO: 29 Given:
- Plant heat up in progress from refueling outage.
- P-32C and P-32D RCP5 are running.
- Seal injection block CV-1206 is in override for testing
- Seal injection flow has been balanced and is in auto at 16 gpm total flow.
- Non-nuclear ICW to RCP motor cooling flow is 200 gpm.
- Nuclear ICW to RCP seal cooling flow is 35 gpm.
- RCS loop A & B cold leg temps are 275°F.
- RCP lift oil pressure is 1800 psig.
A start of RCP P-32A is attempted but is unsuccessful. Why?
A. Nuclear ICW to RCP seal cooling flow is low.
B. Seal injection flow is low.
C. RCP lift oil pressure is low.
D. RCP motor cooling flow is low.
Answer:
D. RCP motor cooling flow is low.
Notes:
D. is correct to satisfy the starting interlock RCP motor cooling flow needs to be >250 gpm A is incorrect, nuclear ICW to RCPS is greater than 30 gpm.
B is incorrect, seal injection flow is greater than 3 gpm to each RCP.
C is incorrect, RCP lift oil pressure is >1750 psig
References:
1103.006 change 032 History:
Modified from QID 559 Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
Rev: 0 Rev Date: 9/15/2009 Source: New Originator: S. Pullin QID: 0796 Objective: 5 PointValue: 1 TUOI: A1LP-RO-TS Section: 3.2 Type: Reactor Coolant System Inventory Control
)
System Number: 004 System
Title:
Chemical and Volume Control System (CVCS the facility license.
==
Description:==
Knowledge of conditions and limitations in K/A Number: 2.2.38 CFR
Reference:
41.7/41.10/43.1/45.13 3.6 RO Select: Yes Difficulty: 3 Tier: 2 RO Imp:
SRO Imp: 4.5 SRO Select: Yes Taxonomy: C Group: I Question: RO: 30 SRO:[ 3 REFERENCE PROVIDED concentration versus RCS Tave would require entry Which of the following Boric Acid Addition Tank level and into TRM 3.5.1 ?
A. 8,700 ppm Boron, BAAT level 36 inches 400 F Tave B. 9,500 ppm Boron, BAAT level 46 inches 450 F Tave C. 10,000 ppm Boron, BAAT level 50 inches, 500 F Tave D. 12,000 ppm Boron, BAAT level 56 inches, 550 F Tave Answer:
C. 10,000 ppm Boron, BAAT level 50 inches 500 F Tave Notes:
ce curve TRM figure 3.5.1-1 C. is correct due to the values fall below and to the right of referen to the left of reference curve TRM figure 3.5.1-1 A, B, and D are incorrect due to the values fall above and REFERENCE PROVIDED FOR THIS QUESTION
References:
1104.003 change 046 TRM 3.5.1 rev 16 History:
New for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
Rev: 0 Rev Date: 9-2-99 Source: Direct Originator: D. Slusher QID: 0259 TUOI: ANO-1-LP-RO-MU Objective: 07 Point Value: 1 Section: 3.2 Type: Reactor Coolant System Inventory Control System Number: 004 System
Title:
Chemical and Volume Control System provide for the following:
==
Description:==
Knowledge of CVCS design feature(s) and/or interlock(s) which gers)
Protection of ion exchangers (high letdown temperature will isolate ion exchan K!A Number: K4.03 CFR
Reference:
CFR: 41.7 2 RO Imp: 2.8 RO Select: Yes Difficulty: 2 Tier:
Group: I SRO Imp: 2.9 SRO Select: Yes Taxonomy: K Question: RO:I 31 SRO:T 31 n?
What is the function of the temperature interlock associated with RCS letdow A. Prevents letdown fluid from flashing to steam when pressure is reduced by closing CV-1221 (letdown isolation).
B. Prevents exceeding letdown piping thermal limits by shutting CV-1 213 & 1215 (letdown cooler inlet MOV).
C. Prevents degrading T36A/B resin by shutting CV-1221 (letdown isolation).
D. Prevents exceeding letdown cooler capacity by shutting CV-1 213 & 1215 (letdown cooler inlet MOV.
Answer:
C. Prevents degrading T36A/B resin by shutting CV-1 221 (letdown isolation).
Notes:
A is incorrect, this is the function of the letdown coolers.
exceeded before the resin is B is incorrect, interlock doesnt close the inlets and piping limits will not be damaged.
C is correct exceeded the interlock doesnt D although the letdown cooler capacity is exceeded when temperature is close the inlet valves.
References:
1104.002 Rev 051-02-0 STMI-04 Rev 5 History:
Used in 1999 exam.
Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0786 Rev: 0 Rev Date: 9/14/2009 Source: Modified Originator: S. Pullin TUOI: Al LP-RO-ELECD Objective: 11 Point Value: I Section: 3.4 Type: Heat Removal From Reactor Core System Number: 005 System
Title:
Residual Heat Removal System
==
Description:==
Knowledge of bus power supplies to the following: RHR pumps.
K/A Number: K2.0i CFR
Reference:
41.7 Tier: 2 RO imp: 3.0 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.2 SRO Select: Yes Taxonomy: K Question: RO: 32 SRO:
Given:
- Plant is in Mode 6
- P-34B Decay Heat pump is running Which of the following would cause a loss of Decay Heat Removal?
A. A-i voltage of 2475 volts B. A-2 voltage of 2475 volts C. B-5 voltage of 428 volts D. B-6 voltage of 428volts Answer:
D. B-6 voltage of 428volts Notes:
B Decay Heat Removal Pump is powered from A-4 via A-2. An undervoltage on the A buses or B buses will trip A-409 (A4 feeder breaker). The undervoltage setpoint for A-4 is 2450 volts. The undervoltage setpoint for B-6 is 429 volts. Therefore, a,b, and c are incorrect.
References:
OP-i 107.002 Change 025 STM 1-32 Figure 32.68 Rev 34 History:
Modified from QID 0293 Selected for 2010 RO/SRO Exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0783 Rev: 0 Rev Date: 9/10/2009 Source: New Originator: Possage TUOI: Al LP-RO-ESAS Objective: 20 Point Value: I Section: 3.3 Type: Reactor Pressure Control System Number: 006 System
Title:
Description:
Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Valves K/A Number: K6.1 0 CFR
Reference:
41.7/45.7 Tier: 2 RO Imp: 2.6 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 3.3 SRO Select: Yes Taxonomy: K Question:
RO: 33 SRO:
Given ESAS Channels 1-6 have actuated.
BWST Outlet Valve CV-1408 fails to open What effect will this have on the ECCS with no operator action?
A. A High Pressure Injection Pump AND A Low Pressure Injection Pump will be damaged due to loss of suction.
B. C High Pressure Injection Pump AND B Low Pressure Injection Pump will be damaged due to loss of suction.
C. A High Pressure Injection Pump AND A Reactor Building Spray Pump will be damaged due to loss of suction.
D. C High Pressure Injection Pump AND B Reactor Building Spray Pump will be damaged due to loss of suction.
Answer:
B. C High Pressure Injection Pump AND B Low Pressure Injection Pump will be damaged due to loss of suction.
Notes:
B. Is the correct answer. CV-1408 is ES actuated open to provide suction to the Green Train ECCS components A. Is incorrect, these are the RED Train ECCS Components and would not be effected by CV-1408.
C. Is incorrect, these are the RED Train ECCS Components and would not be effected by CV-1408.
D. Is incorrect, ESAS Channels 1-6 do not cause the Reactor Building Spray Pumps to start.
References:
STM 1-05 Rev. 16 STM 1-65 Rev. 5 History:
New for 2010 RO/SRO Exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0561 Rev: I Rev Date: 8/10/05 Source: Direct Originator: S.Pullin TUOI: AILP-RO-RCS Objective: 21 Point Value: 1 Section: 3.5 Type: Containment Integrity System Number: 007 System
Title:
Pressurizer Relief Tank/Quench Tank System
Description:
Knowledge of the operational implications of the following concepts as they apply to the PRTS:
Method of forming a steam bubble in the PZR.
KIA Number: K5.02 CFR
Reference:
41.5 / 45.7 Tier: 3 RO Imp: 3.1 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 3.4 SRO Select: Yes Taxonomy: Ap Question: 34 RO:1 SRO:1 A plant startup is in progress with a steam bubble being drawn in the Pressurizer.
- Initial Quench Tank pressure is 3 psig.
- RCS pressure 75 psig.
- Pressurizer temperature 320°F.
Which of the following assures that venting and steam bubble formation is complete in the Pressurizer?
A. Quench Tank pressure 7.6 psig after a 3 minute blow of the ERV.
B. Quench Tank pressure 6.2 psig after a 3 minute blow of the ERV.
C. Quench Tank pressure 4.8 psig after a 3 minute blow of the ERV.
D. Quench Tank pressure 3.5 psig after a 3 minute blow of the ERV.
Answer:
D. Quench Tank pressure 3.5 psig after a 3 minute blow of the ERV.
Notes:
D is correct with Quench Tank pressure rise less than or equal to 1 psig.
All other choices contain greater than I psig pressure rise which indicates nitrogen is still being vented to the Quench Tank.
References:
1103.005, Chg. 036 History:
New for 2005 RO exam, later modified for replacement.
Selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0787 Rev: 0 Rev Date: 9/14/2009 Source: New Originator: S. Pullin TUOI: A1LP-RO-MSSS Objective: 9 Point Value: 1 Sectiop: 3.8 Type: Plant Service Systems System Number: 008 System
Title:
Component Cooling Water System or operations on the CCWS; and
==
Description:==
Ability to (a) predict the impacts of the following malfunctions (b) based on those predictions, use procedures to correct , contro l, or mitigate the consequences of those malfunctions or operations: Loss of CCW Pump K/A Number: A2.01 CFR
Reference:
CFR: 41.5 / 43.5 / 45.3 / 45.13 Tier: 2 RO Imp: 3.3 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.6 SRO Select: Yes Taxonomy: C Question: RO:] 35 SRO:J 35 Given:
- 80% power,
- P33A and P338 ICW pumps in service.
- P33C (ICW Pump) out of service
- P33B (ICW Pump) trips d per 1104.028, ICW System What impact would this have on plant operations, and what actions are require Operating Procedure?
39, CV-2240 and CV-2241 A. Loss of Non-Nuc ICW, open ICW cross connect valves CV-2238, CV-22 38 and CV-2240 B. Loss of Non-Nuc ICW, close A to B cross connect valves CV-22 40 and CV-2241 C. Loss of Nuc ICW, open ICW cross connect valves CV-2238, CV-2239, CV-22 40 D. Loss of Nuc ICW, close A to B cross connect valves CV-2238 and CV-22 Answer:
40 and CV-2241 C. Loss of Nuc ICW, open ICW cross connect valves CV-2238, CV-2239, CV-22 Notes:
r open the suction and C is correct P33C supplies the Nuc ICW loads, OP-I 104.028 has the operato prior to reducin g loads.
discharge cross connect valves to supply both loops with one pump A is incorrect due to Non Nuc ICW loads were never lost B is incorrect due to Non Nuc ICW loads were never lost D is incorrect due to procedure has you open the valves and not close them
References:
OP-1104.028 Change 026 History:
New question, selected for 201 ORO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0788 Rev: 0 Rev Date: 9/14/2009 Source: New Originator: S. Pullin TUOI: A1LP-RO-RPS Objective: 5 Point Value: 1 Section: 3.3 Type: Reactor Pressure Control System Number: 010 System
Title:
Pressurizer Pressure Control System (PZR PCS)
==
Description:==
Knowledge of the effect that a loss or malfunction of the PZR PCS will have on the following:
RPS K/A Number: k3.02 CFR
Reference:
41.7 /45.6 Tier: 2 RO Imp: 4.0 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 4.1 SRO Select: Yes Taxonomy: K Question: RO:j 36 SRO: 136 Given:
- 100% power,
- A MFW Pump trips
- PZR Spray valve (CV-1 008) will not open.
What effect would this pressurizer control system malfunction have on the plant?
A Reactor trip due to AMSAC B. Reactor trip due to anticipatory trip from RPS on loss of MFW pumps C. Reactor trip due to High Power/Imbalance/Flow D. Reactor trip due to High RCS Pressure Answer:
D. Reactor trip due to High RCS Pressure Notes:
A is incorrect because total feedwater flow will remain above trip setpoint B is incorrect because only one MFW pump is tripped C is incorrect because the flow in this coice refers to RCS flow D is correct, without the spray valve opening RCS pressure will rise to the trip setpoint
References:
OP-1202.001 Change 31 History:
New selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0784 Rev: 0 Rev Date: 9/10/2009 Source: New Originator: Possage TUOI: AILP-RO-RPS Objective: II PointValue: I Section: 3.7 Type: Instrumentation System Number: 012 System
Title:
Reactor Protection System on the RPS:
==
Description:==
Knowledge of the effect of a loss or malfunction of the following will have Permissive circuits K/A Number: K6.10 CFR
Reference:
41.7/45.7 Tier: 2 RO Imp: 3.3 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 3.5 SRO Select: Yes Taxonomy: K Question: RO:j 37 SRO: 37 Given:
The plant is at 100% power I&C is troubleshooting RPS B RPS is in Manual Bypass The Power/Imbalance/Flow Trip bistable in Channel A has been pulled from the cabinet.
What would be the effect of a failure in the B RPS permissive circuitry that caused a short which de-energizes the B RPS Cabinet?
A. RPS would be in a 2 out of 3 coincidence trip logic B. RPS would be in a 1 out of 2 coincidence trip logic C. Reactor Trip would occur D. Only RPS Channel A will be tripped.
Answer:
C. Reactor Trip would occur Notes:
it C. Is correct. The conditions given would result in the A Channel being tripped, when B is de-energized would also be tripped and make up the logic to trip the reactor.
A and B are incorrect because the logic to trip the reactor has already been met.
D is incorrect, pulling the Power/Imbalance/Flow Trip bistable would trip the channel but due to de-energizing B RPS cabinet the reactor would trip.
References:
STM 1-63 Rev. 7 History:
Modified from Exam Bank ANO-OPSI-1 670 Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0785 Rev: 0 Rev Date: 9/10/2009 Source: New Originator: Possage TUOI: Al LP-RO-RPS Objective: 19 Point Value: I Section: 2.0 Type: Generic KIA System Number: 012 System
Title:
==
Description:==
Ability to explain and apply system limits and precautions.
KIA Number: 2.1.32 CFR
Reference:
41.10 / 43.2 / 45.12 Tier: 2 RO Imp: 3.8 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 4.0 SRO Select: Yes Taxonomy: K Question: RO:j 38 SRO:
Given:
The plant is at 100% power Which of the following is an applicable Limit & Precaution for the RPS System and why?
A. When testing an RPS protection channel any EFIC Channel can be placed in maintenance bypass simultaneously because RPS has no effect on an EFIC Channel.
B. Placing two RPS Channels in test simultaneously is allowed as long as Shift Manager permission is obtained because will have no effect on RPS operation but requries an LCO entry.
C. The key operated shutdown bypass switch associated with each RPS Channel can be used during power operation becuase this will have no effect on RPS operation but requries an LCO entry.
D. Only one RPS Channel Bypass Key shall be accessible for use in the control room because only one RPS Channel shall be key locked in the untripped state at any one time.
Answer:
D. Only one RPS Channel Bypass Keyshall be accessible for use in the control room because only one RPS Channel shall be key locked in the untripped state at any one time.
Notes:
D is correct, per limits and precautions of 1105.001 A, B and C are all incorrect versions of limits and precautions for 1105.001
References:
OP-1105.00l Change 024 History:
New Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0142 Rev: 0 Rev Date: 10/28/97 Source: Direct Originator: G. Giles TUOI: AA51002-012 Objective: 21 Point Value: 1 Section: 3.2 Type: RCS Inventory Control System Number: 013 System
Title:
Engineered Safety Features Actuation System(ESFAS)
==
Description:==
Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following:
Safeguards equipment control reset.
KIA Number: K4.10 CFR
Reference:
41.7 Tier: 2 RO Imp: 3.3 RO Select: Yes Difficulty: 2 Group: 1 SRO Imp: 3.7 SRO Select: Yes Taxonomy: K Question: RO:J 39 SRO:
Under what conditions can the Control Board Operator bypass or defeat a component automatically actuated by ESAS?
A. Bypassing or defeating a component automatically actuated by ESAS is not allowed.
B. The Control Board Operator, after careful consideration, determines that the component is no longer required.
C. ONLY when procedurally directed by the Emergency Operating or the Abnormal Operating procedures.
D. After it is determined that the component is no longer needed and approval is obtained from the SM/CRS.
Answer:
D. After it is determined that the component is no longer needed and approval is obtained from the SM/CRS.
Notes:
[A] is incorrect, provisions are made for this action.
[B] is partially correct, the component must not be needed but the CBO cannot make this decision on his own.
[C] is only one of the directions where a component can be bypassed/reset, CRS!SS permission is the other.
[D] contains all correct elements, lack of need and supervisory (SRO) permission.
References:
OP-i 202..012 Change 008 History:
Taken from Exam Bank QID # 4791 Used in A. Morris 98 RO Re-exam Previously used under K/A: 3.2 / Reactor Coolant System Inventory Control / 013 / Engineered Safety Features Actuation System / A4.02 / Ability to manually operate and/or monitor in the control room: Reset of ESFAS channels. / CFR: 41.7 / 45.5 to 45.8 I RO: 4.3 / SRO: 4.4 Used on 2004 RO/SRO Exam (K/A T2 Gi 013 K4.06)
Selected for the 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0135 Rev: I Rev Date: 4/7/05 Source: Direct Originator: B. Short TUOI: A1LP-RO-ESAS Objective: 20 Point Value: I Section: 3.5 Type: Containment Integrity System Number: 022 System
Title:
Containment Cooling System
==
Description:==
Ability to monitor automatic operation of the CCS, including: Initiation of safeguards mode of operation.
K/A Number: A3.01 CFR
Reference:
41.7 / 45.5 Tier: 2 RO Imp: 4.1 RO Select: Yes Difficulty: 2 Group: 1 SRO Imp: 4.3 SRO Select: Yes Taxonomy: K Question: RO:J 40 SRO:J 40 A LOCA has occurred.
Reactor Building (RB) pressure is 47 psia.
Which ESAS channels have actuated the RB cooling units and what is the correct RB cooling alignment?
A. ES channels 3 & 4, VSF-1A, IB, 1C, & ID running with service water aligned to the cooling coils.
B. ES channels 3 & 4, VSF-IA, 18, 1C, 10, & IE running with chilled water aligned to the cooling coils.
C. ES channels 5 & 6, VSF-1A, IB, 1C, & ID running with service water aligned to the cooling coils.
D. ES channels 5 & 6, VSF-1A, IB, 1C, 1D, & 1E running with chilled water aligned to the cooling coils.
Answer:
- c. ES channels 5 & 6, VSF-1A, IB, 1C, & ID running with service water aligned to the cooling coils.
Notes:
ESAS channels 5 & 6 actuate RB cooling fans VSF-IA through 1 D and also cause the bypass dampers to drop which allows air to bypass the retum air duct and chilled water coils and flow directly to the service water coils that were aligned by ES channels 5 & 6. Thus (C) is the correct answer. (a), (b) & (d) combine other ventilation alignments with other ES channels that are incorrect.
References:
STM 1-09, Rev. 9 History:
Developed for use in 98 RO Re-exam Selected for 2005 RO exam.
Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0078 Rev: 0 Rev Date: 6/29/98 Source: Direct Originator: JCork TUOI: Al LP-RO-ELECD Objective: 11 .e Point Value: 1 Section: 3.5 Type: Containment Integrity System Number: 026 System
Title:
Containment Spray System (CSS)
==
Description:==
Knowledge of the physical connections and/or cause-effect relationships between the CSS and the following systems: ECCS.
KIA Number: K1 .01 CFR
Reference:
41.2 to 41.9 / 45.7 to 45.8 Tier: 2 RO Imp: 4.2 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 4.2 SRO Select: Yes Taxonomy: K Question: RO:j SRO:J 41 If an ESAS occurs simultaneously with a Loss of Offsite Power, the start of RB Spray pumps is delayed by 35 sec. Why?
A. To allow the EDGs to come up to speed.
B. To allow SW pumps to start for spray pump cooling.
C. To prevent overload of the EDGs.
D. To prevent water hammer of the spray headers.
Answer:
C. To prevent overload of the EDGs.
Notes:
With an ES signal present, ES loads will sequence on to the EDG to prevent overload, therefore C is correct. (a), (b) and (d) are reasons for other aspects of RB spray operation but are not applicable to the basis for the time delay.
References:
1107.002, Chg. 025 History:
Developed for 1998 RO/SRO Exam.
Used in A. Morris 98 RO Re-exam Selected for 2005 Jon Gray RO re-exam.
Selected for the 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0202 Rev: 0 Rev Date: 11/23/98 Source: Direct Originator: R. Walters TUO1: A1LP-RO-EOP Objective: 9 Point Value: I Section: 3.4 Type: RCS Heat Removal System Number: 039 System
Title:
Main and Reheat Steam System
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Malfunctioning steam dump.
KIA Number: A2.04 CFR
Reference:
41.5 / 43.5 I 45.3 / 45.13 Tier: 2 RO Imp: 3.4 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 3.7 SRO Select: Yes Taxonomy: A Question: RO:J 42 SRO:J 42 Given:
- A plant startup is in progress with the reactor critical below the point of adding heat.
- B OTSG Turbine Bypass Valve (CV-6688) fails full OPEN and is unable to be closed with the handjack.
- Tave 524 degrees and dropping
- Pressurizer level 205 inches and dropping
- RCS pressure 2120 psig and dropping What is the proper course of action?
A. Close CV-6688 manual isolation valve MS-2A and maintain the reactor critical using A OTSG Turbine Bypass Valve to control RCS temperature and pressure.
B. Continue the reactor startup maintaining startup rate <1 DPM while continuing to monitor primary and secondary plant parameters.
C. Go directly to 1203.003, OVERCOOLING for actions to mitigate the oversteaming of the B OTSG.
D. Trip the reactor and follow the guidance of 1202.001 REACTOR TRIP.
Answer:
D. Trip the reactor and follow the guidance of 1202.001 REACTOR TRIP.
Notes:
(A.) is incorrect. You would not have time to take this actkand the operator should take conservative action of tripping the reactor..
(B.) is incorrect. With the reactor below the point of adding heat with a stuck open TBV, this would not be possible.
(C.) is incorrect. This will be the ultimate tab that you will end up in, however, it is necessary to trip the reactor first and progress through the Reactor Trip EOP.
(D.) is correct. Taking the conservative action of tripping the reactor is appropriate due to being below the minimum temperature for criticality and the inability to maintain SUR below I DPM.
References:
1102.008 (Rev 023), Approach to Criticality, pages 4&5
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0195 Rev: 0 Rev Date: 11/24/98 Source: Direct Originator: L. Kilby TUOI: A1LP-RO-FW Objective: 18 Point Value: 1 Section: 3.4 Type: RCS Heat Removal System Number: 059 System
Title:
Main Feedwater System
==
Description:==
Ability to monitor automatic operation of the MFW, including: Feedwater pump suction flow pressure K/A Number: A3.03 CFR
Reference:
41.7 / 45.5 Tier: 2 RO Imp: 2.5 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 2.6 SRO Select: Yes Taxonomy: K Question: RO:J 43 SRO:J Unit 1 is operating at 100% power with no abnormal conditions or alignments.
B MFP SUCT PRESS LO (K07-C8) annunciator is received.
Where can the Control Room Operators read the B MFW pump suction pressure WITHOUT leaving the control room?
A. The B MFP Lovejoy Operator Control Station (OCS).
B. B MFP Suction Pressure (P1-2830) indicator.
C. B MFP Suction Pressure computer point (P2830)
D. The Operator Information Touchscreen (OIT).
Answer:
C. B MFP Suction Pressure computer point (P2830)
Notes:
is not (a.) & (d.) are incorrect. These panels are located in the control room, however, MFP suction pressure available on these panels.
(b.) is incorrect. This indicator is located outside the control room.
are (c.) is correct. This computer point is found on the Plant Computer and the SPDS computer both of which available in the control room.
References:
STM 1-19, Rev. 11 History:
Developed for use in 98 RO Re-exam Selected for 2005 RO exam Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I OlD: 0789 Rev: 0 Rev Date: 9/14/2009 Source: New Originator: S. Pullin TUOI: AILP-RO-FW Objective: 6 Point Value: I Section: 3.4 Type: Heat Removal From Reactor Core System Number: 059 System
Title:
Main Feedwater (MFW) System the following:
==
Description:==
Knowledge of MFW design feature(s) and / or interlock(s) which provide for automatic trips for MFW pumps.
K/A Number: K4.16 CFR
Reference:
41.7 Tier: 2 RO Imp: 3.1 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 3.2 SRO Select: Yes Taxonomy: K Question: RO:1 44 SRO: I Given:
- 100% power Which of the following design features provide an automatic trip of the Main Feed Water Pump?
A. Main Feed Water Pump suction pressure reading 220 psig for 45 seconds B. Main Feed Water Pump bearing oil pressure reading 15 psig C. Main Feed Water Pump discharge pressure reading 1360 psig D. Main Feed Water Pump vibration reading 14 mils Answer:
C. Main Feed Water Pump discharge pressure reading 1360 psig Notes:
A is incorrect, suction pressure would have to be less than 200 psig for 40 seconds.
than 10 psig for B is incorrect, bearing oil pressure of 15 psig would cause an alarm but pressure must be less a trip.
C is correct, pump discharge pressure of 1350 psig would result in a pump trip D is incorrect, the high vibration trip is bypassed when the pump is in operation
References:
STM 1-24 Rev. 11 History:
New selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0270 Rev: 1 Rev Date: 11/8/05 Source: Direct Originator: D. Slusher TUOI: A1LP-RO-EFIC Objective: 29 Point Value: I Section: 3.4 Type: Heat Removal From Reactor Core System Number: 061 System
Title:
Auxiliary/Emergency Feedwater System
==
Description:==
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW controls including: S/G level.
KIA Number: Al .01 CFR
Reference:
41.5 / 45.5 Tier: 2 RO Imp: 3.9 RO Select: Yes Difficulty: 2.5 Group: 1 SRO Imp: 4.2 SRO Select: Yes Taxonomy: Ap Question: RO:J SRO:I 5 The EFIC automatic fill rate is designed to prevent overcooling.
With the plant in a degraded power condition and given a SG pressure of 885 psig, determine the proper OTSG fill rate by EFIC for the EFW system:
A. 3/min
- 8. 4/min C. 5/min D. -6/min Answer:
B. 4/min Notes:
OTSG fill rate is adjusted so that OTSG levels raise at 2 inches/minute at OTSG pressure of 800 psig and 8 inches/minute at OTSG pressure of 1050 psig. This limits the overcooling effects of feeding OTSGs with EFW. At 885 psig OTSG fill rate will be 4 inches/minute. b is the correct answer.
References:
1105.005, Chg. 032 History:
Used in 1999 exam.
Direct from ExamBank, QID# 92 used in class exam Modified for 2005 Jon Gray RO re-exam.
Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0790 Rev: 0 Rev Date: 9/14/2009 Source: New Originator: S Pullin TUOI: AILP-RO-EOP Objective: 9 Point Value: 1 Section: 3.6 Type: Electrical System Number: 062 System
Title:
A.C. Electrical Distribution
==
Description:==
Knowledge of local auxiliary operator tasks during emergency and the resultant operation effects.
K/A Number: 2.4.35 CFR
Reference:
41.10/43.5 / 45.13 Tier: 2 RO Imp: 3.8 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 4.0 SRO Select: Yes Taxonomy: C Question: RO:1 46 SRO:1 Given:
Unit I is in a Blackout condition.
Voltage has been recovered on SU#2 and is 155 kV To restore power to A-3 and A-4, what action along with its purpose is required by the Auxiliary Operator?
A. Perform Attachment 1, Blackout Breaker Alignment and UV Relay Defeat, to defeat UV interlocks to allow for starting of equipment necessary to protect the core.
B. Perform Attachment 1, Blackout Breaker Alignment and UV Relay Defeat, to prevent excess current during starting of the motors.
C. Perform Attachment 2, Recovery from Blackout Breaker Alignment and UV Relay Defeat, to allow for starting of equipment necessary to protect the core.
D. Perform Attachment 2, Recovery from Blackout Breaker Alignment and UV Relay Defeat, to allow Unit 2 to tie on non-vital loads on SU#2.
Answer:
A. Perform Attachment 1, Blackout Breaker Alignment and UV Relay Defeat, to defeat UV Close Permissive interlocks to allow for starting of equipment necessary to protect the core.
Notes:
A is correct, with degraded voltage on SU#2, Aft. I is required to defeat the UV interlocks.
B is incorrect, Aft. 1 would have no effect on actual starting current for motors C & D are incorrect, Aft 2 will only be performed when SU#2 voltage is greater than 158 kV.
References:
OP-1202.028 Change 010 History:
New selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0316 Rev: 0 Rev Date: 9/5/99 Source: Direct Originator: J Haynes TUOI: ANO-1-LP-RO-MU Objective: 3.5 Point Value: I Section: 3.6 Type: Electrical System Number: 062 System
Title:
A.C. Electrical Distribution
==
Description:==
Knowledge of bus power supplies to the following: Major system loads.
K/A Number: K2.0l CFR
Reference:
CFR: 41.7 Tier: 2 RO Imp: 3.3 RO Select: Yes Difficulty: 2 Group: 1 SRO Imp: 3.4 SRO Select: Yes Taxonomy: K Question: RO: 7 SRO:
Which of the following would explain why a loss of bus Al will cause CV-1 206 CRC Pump Seal Injection Block Valve) to close?
(Assume plant is at 100% power)
A. P36A (HPI) pump was the in-service pump.
B. Loss of instrument air to Seal Injection Control Valve, CV-l 207.
C. P36C (HP I) pump was the in-service pump.
D. Loss of instrument air to Pressurizer Level Control valve CV-1235.
Answer:
A. P36A (HPI) pump was the in-service pump.
Notes:
a is correct, if P36A was the in-service pump, then a loss of Al would cause a loss of A3, P-36A would cease to run, and CV-1206 would close when Seal Injection flow dropped to less than 22 gpm.
b is incorrect, CV-1 207 fails open on a loss of Instrument Air.
c is incorrect, a loss of Al would not affect P36Cs power supply, bus A4.
d is incorrect, CV-l 235 fails as-is on a loss of Instrument Air.
References:
1203.026, Change 11 History:
Used in 1999 exam.
Modified from ExamBank, QID# 3716.
Selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0086 Rev: 0 Rev Date: 7/11/98 Source: Direct Originator: JCork TUOI: Al LP-RO-ELECD Objective: 37 Point Value: I Section: 3.6 Type: Electrical System Number: 063 System
Title:
D.C. Electrical Distribution
Description:
Knowledge of the effect that a loss or malfunction of the dc electrical system will have on the following: Components using dc control power.
KIA Number: K3.02 CFR
Reference:
41.7 / 45.6 Tier: 2 RO Imp: 3.5 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.7 SRO Select: Yes Taxonomy: K Question: 48 RO:j SRO:J The plant is at 100% power.
Which of the following DC buses/panels, if de-energized, would cause a reactor trip?
A. Panel D41 B. Panel RA1 C. MCCD15 D. Panel D21 Answer:
B. Panel RA1 Notes:
Only B is capable of causing a reactor trip due to loss of two RCP contact monitors.
The others would cause a loss of vital equipment capability but as seen in Att. J of 1107.004, they would not cause a trip.
References:
1107.004, Chg. 016 History:
Developed for 1998 RO exam Used in A. Morris 98 RO Re-exam Selected for use in 2005 RO exam, but not used.
Selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0791 Rev: 0 Rev Date: 9/14/2009 Source: New Originator: S. Pullin TUOI: Al LP-RO-EDG Objective: I 9a Point Value: 1 Section: 3.6 Type: Electrical System Number: 064 System
Title:
Emergency Diesel Generators (ED/G)
==
Description:==
Knowledge of bus power supplies to the following: Air Compressor KIA Number: K2.01 CFR
Reference:
41.7 Tier: 2 RO Imp: 2.7 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 3.1 SRO Select: Yes Taxonomy: K Question: RO:j 49 SRO: 49 What is the power supply to Emergency Diesel Generator Starting Air Compressors, C4A1 and C4B2?
A. B31 and B41 B. 832 and B42 C. B51 and B61.
D. B52 and 862 Answer:
A. B31 and B41 Notes:
A is correct, the other choices are alternate possibilities.
References:
OP-1107.001 Change 73 History:
New for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0792 Rev: 0 Rev Date: 9/14/2009 Source: New Originator: S. Pullin TUOI: AILP-RO-EDG Objective: 19 Point Value: I Section: 3.6 Type: Electrical System Number: 064 System
Title:
Emergency Diesel Generators (ED/G)
==
Description:==
Knowledge of the physical connections and / or cause-effect relationships between the ED/G system and the following systems: Starting air system.
KIA Number: Ki .05 CFR
Reference:
41.2 to 41.9 I 45.7 to 45.8 Tier: 2 RO Imp: 3.4 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 3.9 SRO Select: Yes Taxonomy: C Question: RO:1 50 SRO:
Given:
Plant at 100%
Performing #1 EDG monthly surveillance per 1104.036 Supplement 1 The CBOT presses the start pushbutton on ClO K01-B2, EDG I OVERCRANK, alarms What is the cause of the alarm and how long did the starting air system attempt to start the engine?
A. #1 EDG did not exceed 300 rpm in 45 seconds and air start motors engaged for 8 seconds.
B. #1 EDG did not exceed 300 rpm in 8 seconds and air start motors engaged for 45 seconds.
C. #1 EDG did not exceed 30 rpm in 45 seconds and air start motors engaged for 2.5 seconds.
D. #1 EDG did not exceed 30 rpm in 8 seconds and air start motors engaged for 8 seconds.
Answer:
A. #1 EDG did not exceed 300 rpm in 45 seconds and air start motors engaged for 8 seconds.
Notes:
A is correct, due to meeting the annunciator logic B, C, and D are variations of the control logic for the starting air to the engine
References:
STM-1-31 rev 10 1203.012A change 038 History:
New 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0672 Rev: 0 Rev Date: 12/16/06 Source: Repeat Originator: Possage TUOI: AILP-RO-RMS Objective: 8 Point Value: I Section: 3.7 Type: Instrumentation System Number: 073 System
Title:
Process Radiation Monitoring System apply to the PRM
==
Description:==
Knowledge of the operational implications of the following concepts as they System: Radiation theory, including sources, types, units, and effects.
KIA Number: K5.01 CFR
Reference:
41.5 / 45.7 2 RO Imp: 2.5 RO Select: Yes Difficulty: 2 Tier:
Group: 1 SRO Imp: 3.0 SRO Select: Yes Taxonomy: K Question: RO:1 SRO:
monitor for steam What type of detector is used by the Main Condenser Air Discharge Radiation Monitor to generator tube leaks?
A. Scintillation Detector B. Geiger Mueller Detector C. Ion Chamber Detector D. Beta Radiation Detector Answer:
A. Scintillation Detector Notes:
detector.
A is correct. The Main Condenser Air Dischagre Radiation Monitor is a scintillation B is incorrect. Area Monitors are G-M Detectors C is incorrect. Ion chambers are used for RP surveys D is incorrect. The Penentration Ventilation Monitors are beta sensitive monitors.
References:
STM 1-62 Rev. 11 History:
New for 2007 RO Exam.
Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0793 Rev: 0 Rev Date: 9/15/2009 Source: Direct Originator: S Pullin TUOI: AILP-RO-MSSS Objective: I Point Value: 1 Section: 3.4 Type: Heat Removal From Reactor Core System Number: 076 System
Title:
Service Water System (SWS)
==
Description:==
Ability to manually operate and / or monitor in the control room SWS valves KIA Number: A4.02 CFR
Reference:
41.7/45.5 to 45.8 Tier: 2 RO Imp: 2.6 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 2.6 SRO Select: Yes Taxonomy: Ap Question: RO: 52 SRO:
When starting Service Water Pump P-4A after maintenance, you observe the following symptoms.
- Pump start is indicated by normal light indication above pump control HS on.
- Annunciator K10-B3 SW DISCH PRESS HI alarms.
- Valve position indication in the control room indicates proper valve alignment.
- SW Bay levels are 338 feet
- No change in SW flow or discharge pressure indications on the SPDS Diagnostics screen.
- No change in SW Loop pressure indications on control room panel C09.
Which of the following is the most likely cause of these symptoms?
A. The pump discharge valve was not opened when returned to service.
B. Warm weather conditions cause low demand from ACW/SW.
C. P-4A cannot pump into the system because of high system pressure from the other(running) pump.
D. P-4A is running without sufficient NPSH to pump water into the SW System.
Answer:
A. The pump discharge valve was not opened when returned to service.
Notes:
A is the correct answer. With the local discharge valve closed, the SW Pump would not be able to pump water to the loop, but since the discharge pressure switch is between the pump and discharge vlave, therefore a high discharge pressure would be seen.
B is incorrect, warm water would cause a high demand for flow not a low demand.
C is incorrect, if the maintenance performed had caused low discharge pressure such that the pump was unable to pump water to the loop, there would not be a high discharge pressure alarm.
D is incorrect, with a bay level of 338 feet, suction pressure would be (356.5-338)0.433= 8 psig which is adequate.
References:
OP-i 203.0121 Change 046 History:
Direct ANO Exam bank QID ANO-OPSI-3284 Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0794 Rev: 0 Rev Date: 9/15/2009 Source: New Originator: S. Pullin TUOI: Al LP-RO-ESAS Objective: 20 Point Value: I Section: 3.4 Type: Heat Removal From Reactor Core System Number: 076 System
Title:
Service Water System
==
Description:==
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: Reactor and turbine building closed cooling water temperatures.
K/A Number: Al .02 CFR
Reference:
41.5 / 45.5 Tier: 2 RO Imp: 2.6 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 2.6 SRO Select: Yes Taxonomy: K Question: RO:j 53 SRO:1 What would be the effect to service water pressure due to an inadvertent actuation of ES Channel 5?
A. Service Water Pressure would drop due to SW valves to the EDG Coolers opening.
B. Service Water Pressure would drop due to SW valves to the RB Coolers opening.
C. Service Water Pressure would rise due to ACW isolation valve closing.
D. Service Water Pressure would rise due to SW to ICW isolations closing.
Answer:
B. Service Water Pressure would drop due to SW valves to the RB Coolers opening.
Notes:
A is incorrect, SW to EDG Coolers open on diesel start. EDG starts on Channels I or 2 B is correct, ES Channel 5 will align SW to the RB Coolers C is incorrect, ACW isolation valve would close on ES Channel 2 D is incorrect, SW to ICW isoaltion valve will close on ES Channels I and 2
References:
STM 1-65 Rev. 5 History:
New, Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0535 Rev: I Rev Date: 10/13/200 Source: Direct Originator: J.Cork TUOI: Al LP-RO-AOP Objective: 3 Point Value: I Section: 3.8 Type: Plant Service Systems System Number: 078 System
Title:
Instrument Air System
==
Description:==
Knowledge of the physical connections and / or cause-effect relationships between the lAS and the following systems: Service Air KIA Number: Ki .02 CFR
Reference:
41.7 / 45.5 Tier: 2 RO Imp: 2.7 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 2.8 SRO Select: Yes Taxonomy: K Question: RO:J 54 SRO:j 54 Instrument Air pressure has dropped to 50 psig.
Which of the following manual or automatic actions should be performed or will occur in response to the low Instrument Air pressure?
Note: All actions for higher pressures have been completed at the required pressure and answer the question considering only the action for the current pressure.
A. Service Air to Instrument Air cross-connect automatically opens.
B. Unit I to Unit 2 Instrument Air cross-connect automatically opens.
C. Trip Reactor, actuate EFW and MSLI on both SGs.
D. Close Letdwon Cooler Outlet to isolate letdown Answer:
A. Service Air to Instrument Air cross-connect automatically opens.
Notes:
B is incorrect, this valve is closed when either unit IA pressure reaches 60 psig.
A is correct, this automatically occurs when pressure drops to 50 psig.
C is incorrect, this would not be done until pressure was less than 35 psig.
D is incorrect, this would not be done until pressure was less than 35 psig
References:
1104.025, Chg. 014 History:
Developed for 1998 RO exam (similar to QID 102)
Modified question for A. Morris 98 RO Re-exam Modified for J. Gray 2005 re-exam.
Selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0795 Rev: 0 Rev Date: 9/15/2009 Source: Direct Originator: S. Pullin TUOI: Al LP-RO-RBS Objective: 11 Point Value: 1 Section: 3.5 Type: Containment Integrity System Number: 103 System
Title:
Containment System
==
Description:==
Ability to manually operate and / or monitor in the control room: Operation of the containment personnel airlock door K/A Number: A4.06 CFR
Reference:
41.7/45.5 to 45.8 Tier: 2 RO Imp: 2.7 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 2.9 SRO Select: Yes Taxonomy: C Question: RO:J 55 SRO:J Given:
Plant refueling is in progress The Reactor Building Coordinator calls the control room and reports the following:
The inner door of the reactor building personnel hatch will not close The outer door is operable In accordance with Technical Specifications for Refueling Operations, how does this affect fuel movement?
A. Irradiated fuel movement in the reactor building and auxiliary building must be suspended.
B. Irradiated fuel movement in the reactor building must be suspended.
C. Irradiated fuel movement in the auxiliary building must be suspended.
D. Irradiated fuel movement may continue without restriction.
Answer:
D. Irradiated fuel movement may continue without restriction.
Notes:
D is correct, fuel movement may continue in both the Reactor Building and Aux Building provided one of the air lock doors is capable of being closed.
A, B, and C are incorrect due to the outer door being operable.
References:
T.S. 3.9.3 Amendment No. 215 History:
Direct from ANO exam bank ANO-OPSI -6622 Selected for 2010 ROISRO exam.
m C) rTI 0
C C
0 m
x 3
0 0
C (fl D CD (D
o L C)
CD ,
0
= g . CD W
CD ir
-n 0 -T,
- 0 3
m 3 C) m C
C 1%)
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0429 Rev: 0 Rev Date: 4/30/2002 Source: Direct Originator: S.Pullin TUOI: Al LP-RO-CRD Objective: 8 Point Value: I Section: 3.1 Type: Reactivity Control System Number: 001 System
Title:
Control Rod Drive System
==
Description:==
Knowledge of bus power supplies to the following: One-line diagram of power supply to trip breakers KIA Number: K2.02 CFR
Reference:
41.7 Tier: 2 RO Imp: 3.6 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.7 SRO Select: Yes Taxonomy: C Question: RO:1 5R01 If breaker 8631 opened while operating at 100% power, the response of the Control Rod Drive system would be:
A. A ratchet trip of all regulating rods since half of the power supply has been removed.
B. No effect on regulating rods, safety rods are held by a single phase (CC) energized.
C. A ratchet trip of the safety rods due to a single phase remaining energized.
D. A trip of all safety rods since the main power has been removed.
Answer:
B. No effect on regulating rods, safety rods are held by a single phase (CC) energized.
Notes:
B is correct. The one-line diagram shows the power supply configuration from A-501 providing power to the CC phase on the DC hold bus which will maintain the safety rods out. Regulating rods are not effected normal movement will be supplied by the Bus 2 power supplied by A-501.
References:
STM 1-02, Control Rod Drive System, page 9, step 2.4 History:
Direct from regular exambank OlD 4208.
Selected for use in 2002 RO/SRO exam.
Selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0604 Rev: 0 Rev Date: 6/30/05 Source: Direct Originator: S.PulIin TUOI: Al LPR-RO-RCS Objective: 5 Point Value: I Section: 3.2 Type: Reactor Coolant System Inventroy Control System Number: 002 System
Title:
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the RCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of coolant inventory.
KIA Number: A2.0l CFR
Reference:
41.5 / 43.5 I 45.3 / 45.5 Tier: 2 RO Imp: 4.3 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 4.4 SRO Select: Yes Taxonomy: C Question: RO:j 57 SRO:I 57 A reactor trip has occurred and the CRS is directing actions per 1202.001, Reactor Trip.
Assume all actions have been performed when required by system parameters.
5 minutes later the following is reported:
The CBOR reports that Pressurizer level has fallen to 30 and continuing to drop.
Pressurizer Level Control (CV-1 235) is in Auto and fully open.
Which of the following is the proper response?
A Initiate HPI per Repetitive Task (RT-2).
B. Reduce Letdown by closing Orifice Bypass (CV-1223).
C. Isolate Letdown by closing Letdown Cooler Outlet (CV-I 221).
D. Operate CV-1235 in HAND to control PZR level 90 to 110.
Answer:
A Initiate HPI per Repetitive Task (RT-2).
Notes:
Answer A is correct, this is done when level is < 30 per 1202.001.
Answer B is incorrect, this was done early in the procedure, shortly after immediate actions.
Answer C is incorrect, this was done earlier when level was < 50.
Answer D is incorrect, CV-1235 is operating properly in Auto, taking it to hand would not help.
References:
1202.001, Chg. 031 History:
New for 2005 RO exam, modified as a replacement question.
Selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0797 Rev: 0 Rev Date: 9/15/2009 Source: New Originator: S. Pullin TUOI: A1LP-RO-MU Objective: 4 Point Value: I Section: 3.2 Type: Reactor Coolant System Inventory Control System Number: 011 System
Title:
Pressurizer Level Control System (PZR LCS)
==
Description:==
Ability to monitor automatic operation of the PZR LCS, including: Charging and letdown.
K/A Number: A3.03 CFR
Reference:
41.7/45.5 Tier: 2 RO Imp: 3.2 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 3.3 SRO Select: Yes Taxonomy: Ap Question: RO: 58 SRO:
Given:
Plant at 100%
Letdown flow 80 gpm indicated on Fl-1236 Letdown pressure 50 psig on P1-1237 CV-1244 and CV-1245 Letdown Dl Inlet Isolation valves lose power.
With no operator action what would be the expected automatic response of the pressurizer level control system?
A. P1-1237 would read 50 psig and Pressurizer level control valve CV-1 235 position would close.
B. P1-1237 would read 150 psig and Pressurizer level control valve CV-1 235 position would open.
C. P1-1237 would read 50 psig and Pressurizer level control valve CV-1235 position would open.
D. P1-1237 would read 150 psig and Pressurizer level control valve CV-1 235 position would close.
Answer:
B. P1-1237 would read 150 psig and Pressurizer level control valve CV-1 235 position would open.
Notes:
B is correct, due to letdown Dl Inlet Isolation Valves fail closed on a loss of power. Which would isolate letdown, letdown pressure would rise to the letdown relief setpoint of 150 psig, causing a LOCA. Pressurizer level would go down causing CV-1235 to open.
A,C, and D are variations of these possible combinations.
References:
STM 1-04 Rev. 9 History:
New for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0308 Rev: 0 Rev Date: 9-5-99 Source: Direct Originator: J. Cork TUOI: ANO-1-LP-RO-CRD Objective: 16 Point Value: 1 Section: 3.1 Type: Reactivity Control System Number: 014 System
Title:
Rod Position Indication System
==
Description:==
Knowledge of RPIS design feature(s) and/or interlock(s) which provide for the following: Rod hold interlocks.
KIA Number: K4.05 CFR
Reference:
CFR: 41.5 / 45.7 Tier: 2 RO Imp: 3.1 RO Select: Yes Difficulty: 2.5 Group: 2 SRO Imp: 3.3 SRO Select: Yes Taxonomy: C Question: RO:1 59 SRO:I 59 Given:
- Plant is at 100% power.
- ICS is in full automatic.
Subsequently, annunciator K07-B3 ASYM ROD RUNBACK IN EFFECT alarms.
A check of the P1 panel shows that Rod 6 in Group 5 has dropped.
Which of the following alarms or indications would you expect to see on the diamond panel?
A. Sequence Inhibit lamp ON B. Out Inhibit lamp ON C. Auto Inhibit lamp ON D. Manual lamp ON Answer:
B. Out Inhibit lamp ON Notes:
A is incorrect because the sequence inhibit is generated from relative position indications which do not use absolute position indications.
B is correct because the rods are interlocked so that they cannot move outward with an asymmetric rod fault with power greater than 40%.
C is incorrect because the rods are in auto and dropped rod is not a condition which will place the CRD system in manual.
D is incorrect because the diamond will not automatically revert to manual.
References:
1105.009 Change 32 History:
Developed for 1999 exam.
Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0299 Rev: 0 Rev Date: 9-5-99 Source: Direct Originator: J Haynes TUOI: ANO-1-LP-RO-Nl Objective: 10 Point Value: 1 Section: 3.7 Type: Instrumentation System Number: 015 System
Title:
Nuclear Instrumentation System
==
Description:==
Knowledge of the effect that loss or malfunction of the NIS will have on the following: ICS KIA Number: K3.04 CFR
Reference:
41.7/45.6 Tier: 2 RO Imp: 3.4 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 4.0 SRO Select: Yes Taxonomy: An Question: RO: 60 SRO: 60 Given:
- The plant is at 80% power.
- The NI SASS mismatch alarm is bypassed due to a mismatch.
What would be the predicted plant response if Nl-6 failed to 125%?
A. Control rods move inward, feedwater flows go up.
B. Control rods move inward, feedwater flows go down.
C. Control rods move outward, feedwater flows go up.
D. Control rods move outward, feedwater flow go down.
Answer:
A. Control rods move inward, feedwater flows go up.
Notes:
The mismatch alarm disables the SASS module automatic operation. When NI-6 fails to 125% power, ICS will see Nl-6 as the input power. ICS will generate an error to drive rods in. At the same time a cross-limit is generated to keep feedwater balanced with reactor power. Feedwater will go up. Therefore, B, C, and D are incorrect.
References:
STM 1-64, Integrated Control System, rev 10, page 33, step 2.6.1, page 43, step 2.7 History:
Used in 1999 exam.
Direct from ExamBank, QID# 3723 Selected for 2002 RO exam.
Used on 2004 SRO/SRO Exam.
Selected for 2010 ROISROexam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0077 Rev: 0 Rev Date: 9/29/98 Source: Direct Originator: JCork TUOI: ANO-1-LP-RO-NNI Objective: 5 Point Value: 1 Section: 3.7 Type: Instrumentation System Number: 016 System
Title:
Non-Nuclear Instrumentation System (NNIS)
Description:
Knowledge of the physical connections and/or cause-efffect relationships between the NNIS and the following systems: RCS KIA Number: KI .01 CFR
Reference:
41.2 to 41.9/45.7 to 45.8 Tier: 2 RO Imp: 3.4 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 3.4 SRO Select: Yes Taxonomy: C Question: 61 RO:1 SRO:1 61 Given:
- Loop A RCS flow 70 E6 Ibm/hr
- Loop B RCS flow 63 E6 Ibm/hr
- Loop A Tave 578°F
- Loop B Tave 580°F
- Unit Tave 579°F Which Tave will be selected by the SASS Auto/manual transfer switch and why?
- a. Unit Tave due to Loop B flow
- b. Loop A Tave due to Loop B flow
- c. Loop B Tave due to Loop B flow ci. Unit Tave, flows are within tolerances Answer:
- b. LoopATave due to Loop Bflow Notes:
SASS will automatically select the Loop Tave for the Loop with the highest RCS flow should either flow drop below 95%. Normal RCS loop flow is 70 E6 Ibm/br, therefore Loop B flow is <95% and SASS will select Loop A flow for Tave control, this control function protects the core from excessive heat transfer based upon flux to flow, therefore, (b) is the only correct response.
References:
STM 1-69 (Rev 13), Non-Nuclear Instrumentation System page 12 step 3.3.5 History:
Modified QID 2517 for 1998 RO/SRO Exam.
Used in A. Morris 98 RO Re-exam Selected for 2002 RO/SRO exam.
Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0240 Rev: 0 Rev Date: 8-17-99 Source: Direct Originator: Don Slusher TUOI: ANO-1-LP-RO-NNI Objective: 25 Point Value: I Section: 3.7 Type: Instrumentation System Number: 017 System
Title:
In-Core Temperature Monitor (ITM) System
Description:
Knowledge of the effect of a loss or malfunction of the following ITM system components:
Sensors and detectors.
KIA Number: K6.01 CFR
Reference:
CFR: 41.7/45.7 Tier: 2 RO Imp: 2.7 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.0 SRO Select: Yes Taxonomy: C Question: 62 RO:J SRO:J Given:
- Plant is at 100% power
- All CET5 indicate 602 °F ICC train B Core Exit Thermocouple TE-1 152 fails to 900 °F.
What is the effect of this failure?
A. Core Exit Thermocouple TE-1 152 will be removed from the average.
B. ICC Core Exit Thermocouple indication will go to 627 °F.
C. TRAIN B SUBCLG MARG LO annunciator will alarm.
D. B SPDS will switch from ATOG to the ICC display.
Answer:
A. Core Exit Thermocouple TE-1 152 will be removed from the average.
Notes:
CETs are averaged together to generate alarms, indication, or action. Therefore, b, c, and d are incorrect and a is correct since ICCMDS will determine that TE-1 152 is unreliable and remove it from the average.
References:
1105.008 Rev 17 History:
Developed for 1999 exam.
Used on 2004 RO/SRO Exam.
Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0138 Rev: 0 Rev Date: 12/02/98 Source: Direct Originator: B. Short TUOI: AA51002-013 Objective: 9 Point Value: 1 Section: 3.4 Type: RCS Heat Removal System Number: 045 System
Title:
Main Turbine Generator System
Description:
Ability to manually operate and/or monitor in the control room: Turbine stop valves.
KIA Number: A4.06 CFR
Reference:
41.7 / 45.5 to 45.8 Tier: 2 RO Imp: 2.8 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 2.7 SRO Select: Yes Taxonomy: K Question: 63 63 RO: SRO:
During the performance of Main Turbine Governor Valve testing, while governor valve #1 was closed in the test position governor valve #3 fails closed. What turbine problems does this impose?
A. Moisture impingement on the turbine blading.
B. Thermal shock to the turbine rotor.
C. Turbine will trip due to low load.
D. Turbine overspeed condition.
Answer:
B. Thermal shock to the turbine rotor.
Notes:
(A) is incorrect. The closure of both valves does not change the quality of the steam.
(B) is correct. Closure of GVI and GV3 with GV2 & GV4 open or closure of GV2 & GV4 with GV1 & GV3 open causes thermal shock on the turbine rotor.
(C) is incorrect. The load shifts through the two valves that remain open.
(D) is incorrect. The load will stay essentially the same so that an overspeed condition should not occur.
References:
1106.009 (Change 37)
History:
Developed for use in A. Morris 98 RO Re-exam Selected for 2010 ROISRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0798 Rev: 0 Rev Date: 9/16/2009 Source: New Originator: S. Pullin TUOI: A1LP-RO-MSSS Objective: 4 Point Value: I Section: 3.8 Type: Plant Services System System Number: 075 System
Title:
==
Description:==
Knowledge of abnormal condition procedures.
KIA Number: 2.4.11 CFR
Reference:
41.10 / 43.5 / 45.13 Tier: 2 RO Imp: 4.0 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 4.2 SRO Select: Yes Taxonomy: C Question: RO:I 64 SRO:
Given:
- Plant at 100% power
- Lake Temperature is stable at 65 F
- P-3A, P-3B, and P-3C Circulating Water Pumps are running
- P-3A Circulating Water Pump trips.
- P-3D Circulating Water Pump standby pump was started.
- It is noticed that the condenser waterbox discharge temperature is 10 degrees higher and condenser vacuum is dropping.
- AOP 1203.016, Loss of Condenser Vacuum, has been entered.
Which of the following is the cause for these conditions?
A. The stopping and starting of a circ pump caused fouling to be removed from the tube sheet promoting better heat transfer capabilities.
B. The discharge valve on the tripped pump did not go completely closed and circulating water is short cycling.
C. The debris on the bar grates of the circulating water bays was stirred up during the circ pump swap causing reduced flow.
D. Lake temperature is too high for 3 circulating water pump operation per 1104.008, Circulating Water and Water Box Vacuum System Operation.
Answer:
B. The discharge valve on the tripped pump did not go completely closed and circulating water is short cycling.
Notes:
(A.) is incorrect. Although some fouling can be removed during pump rotations, it should not result in a 10 degree change in waterbox discharge temperature.
(B.) is correct. The discharge valve on an idle pump can allow a significant amount of backflow from the operating pumps if it is not closed completely.
(C.) is incorrect. This condition is normal for a circ pump swap and may contribute to waterbox fouling, however, the service water system would be affected by this condition as well.
(D.) is incorrect. 1104.008 states that 4 CW Pumps are needed when lake temperature is above 67 F
References:
1104.008, Circulating Water System, change 27, pagel 3, Caution
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I-History:
New for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0542 Rev: 0 Rev Date: 12/8/2003 Source: Direct Originator: NRC TUOI: Objective: Point Value: I Section: 3.8 Type: Plant Service Systems System Number: 086 System
Title:
Fire Protection System
==
Description:==
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fire Protection System controls including: Fire header pressure.
K/A Number: Al .01 CFR
Reference:
41.5/45.5 Tier: 2 RO Imp: 2.9 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.3 SRO Select: Yes Taxonomy: K Question: RO:j 65 65 SRO:J You are on watch in the Control Room when the following annunciator alarms:
- K12-Ai, FIRE As Fire Water Header pressure drops from 110 psig to 80 psig select the order that fire pumps would start.
A. Jockey FWP P-il; Diesel Fire Pump P-6B starts second; Electric Fire Pump P-6A starts last.
B. Electric Fire Pump P-6A; Diesel Fire Pump P-6B starts second; Jockey FWP P-i I starts last.
C. Electric Fire Pump P-6A; Jockey FWP P-il starts second; Diesel Fire Pump P-6B starts last.
D. Jockey FWP P-Il; Electric Fire Pump P-6A starts second; Diesel Fire Pump P-6B starts last.
Answer:
D. Jockey FWP P-i 1; Electric Fire Pump P-6A starts second; Diesel Fire Pump P-6B starts last.
Notes:
D is correct, Jockey FWP P-li; Electric Fire Pump P-6A starts second; Diesel Fire Pump P-6B starts last.
The other choices are incorrect based on pressure to start for each one.
References:
STM 1-60, Fire Protection System, rev 8, page 2.
History:
Developed by NRC.
Used on 2004 RO Exam.
Selected for 2010 RO/SRO exam
m -4 om -n C,, CD 00* 0.0 C) 0 m B 0 C, DOCD D 0. DC Cl)
C) CD CD *00. CD 0 1, 0. 2.3 CDC
- 0 o 0 0 D CD -
CDC D
-1 0 0 D 0 Ci)
C)
C)
C) ci) C) Cl, C,) C C C C C C) 0* C C. C. CD 0 0 0 0 C) 0 -I 0 SQ. C) 0 CD CD
> 00> > =0> >
C D D 3D 0-C C* C*
0 0 C CD =
C .
CD CD CD p o*
0 0. 0. 0. CDO 0 CD CD CD D o
- 0. CD CD CD CD 0 CD-CD . CD C C)
D 0 0 - 0 :i. *0 CD
- 8. 3 CDc,)O 0 CO m CD 00 3_CD CD D
-o D 0 . 5. 00 5.
CD -U C) 0 DC) 0 5 j CD D Ci) 0 D 0.
3 0 C) (0 C)
C) CD CD 3C)
C CD O
- 0. C) C C) CDO 0 0. C. *< 0 0.
CD 0 -C CD 0 0)
D D C,) Ci)
D m CD Ci) C C,) C) a 00 CD -1 x C,) C) D 0 CD 3 0
- 0. CDCD 3 0 CD 00 CD 3 0.0. DD C) 3 3 m CD 0 3 CD5 ci CD 0 U, 0. 0 D C C) CDO D
-o D CD I.
C) C 0. 0 C) gg 0 Cn D O CD 0 CD 0. 0 C) 0 m D CD C* CD D DO C C 0 00 0.-
- 0. C)
.0 - CD 0 CC) C) 3 C) 0 C) SQ. D 0
- 2: C D 0 Ci) 0. D C) 0 CII D
5 C,) CD C) 0
-a 0 0 D
C)
CD
-n 0
C.) C.) . C.) C.) .D r C.) . .
3 J 0) 0 J 0) m -
C,)
-.J - 4 - -
0 C..fl . C.) I) -. - 0) 0) 0) 0) 0 0 CD 0) -4 0)
C,.)
co
-n
- 0) 0) 0 0 CD 0 0) 0 -
F) 0) C.) 0) CD 0 F.) ci 3 m
-I Cl) ci z ci z z z ci z z ci CD 0
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0482 Rev: 0 Rev Date: 10/7/2003 Source: Direct Originator: J.Cork TUOI: Al LP-WCO-CZ Objective: 13 Point Value: I Section: 3.9 Type: Radioactivity Release System Number: 068 System
Title:
Liquid Radwaste System (LRS)
==
Description:==
Ability to perform specific system and integrated plant procedures during all modes of plant operation.
KIA Number: 2.1.23 CFR
Reference:
41.10 / 43.5 /45.2 / 45.6 Tier: 3 RO Imp: 4.3 RO Select: Yes Difficulty: 2 Group: SRO Imp: 4.4 SRO Select: Yes Taxonomy: K Question: RO:166 5R0 Which of the following must be performed when you are releasing an Liquid radwaste tank and its Process Monitor is inoperable?
A. Estimate radiation level every four hours during the release.
B. Have an independent sample obtained and analyzed prior to release.
C. Estimate flow rate at least once every three hours during release.
D. The Tank can NOT be released iwhen its process monitor is inoperable.
Answer:
B. Have an independent sample obtained and analyzed prior to release.
Notes:
Answer B contains the requirement from Att. BI of 1104.020. The other answers are incorrect.
2004 Exam Development Note: Randomly selected alternate K/A 2.1.23 to replace 2.1.31 due to lack of CR controls at ANO for the Liquid Radwaste system.
References:
1104.020, Change 49, Aft. BI, section 2 History:
Modified regular exambank QID #2765.
Used on 2004 RO/SRO Exam.
Selected for 2010 RO/SRO
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0800 Rev: 0 Rev Date: 9/16/2009 Source: New Originator: S Pullin TUOI: A1LP-RO-ESAS Objective: 20 Point Value: 1 Section: 2 Type: Generic K&A System Number: 2.1 System
Title:
Conduct of Operations
==
Description:==
Ability to locate control room switches, controls, and indications, and to determine they correctly reflect the desired plant lineup.
KIA Number: 2.1.31 CFR
Reference:
41.10/45.12 Tier: 3 RO Imp: 4.6 RO Select: Yes Difficulty: 3 Group: SRO Imp: 4.3 SRO Select: Yes Taxonomy: C Question: RO:j SRO:
Given:
LOCA in progress has caused ESAS actuation of Channel 1-4 Which of the following combinations of indications and locations are correct for the given condition?
A. CV-3820, SW TO ICW, green light, on C16; CV-1 270, RCP SEAL BLEEDOFF FROM D RCP, red light, on C18; CV-1 053, QUENCH TANK DRAIN, green light, on C16 B. CV-1233, RCS MAKEUP, red light, on C16; CV-1441, BWST PURIF RECIRC ISOL, green light, on C13; CV-5612 ,FIRE WATER TO RB, green light, on C18.
C. CV-1 285, HIGH PRESSURE INJECTION, red light, on C16; CV-1407, BWST OUTLET, red light, on C18; CV-3841, LPI PUMP BRG CLR E-50 INLET, red light, on C16 D. CV-1408, BWST OUTLET, red light, on C18; CV-7402, RB PURGE INLET, green light, on C18; CV-4804, RB VENT, red light, on C16 Answer:
C. CV-1285, HIGH PRESSURE INJECTION, red light, on C16; CV-1407, BWST OUTLET, red light, on C18; CV-3841, LPI PUMP BRG CLR E-50 INLET, red light, on C16 Notes:
C is correct in that it has the correct indications and panel locations.
A is incorrect in that it has the incorrect indications and correct panel locations.
B is incorrect in that it has the correct indications and incorrect panel locations.
D is incorrect in that it has the incorrect indications and incorrect panel locations.
References:
STM 1-65 Rev 5 ESAS STM 1-05 Rev 16 DHR History:
New selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
History:
New for 2010 RO/SRO exam QID: 0799 Rev: 0 Rev Date: 9/16/2009 Source: New Originator: S Pullin TUOI: AILP-RO-ICS Objective: 11 PointValue: 1 Section: 2.0 Type: Generic K&A System Number: 2.1 System
Title:
Conduct of Operations
==
Description:==
Ability to explain and apply system limits and precautions.
K/A Number: 2.1.32 CFR
Reference:
41.10 /43.2/45.12 Tier: 3 RO Imp: 3.8 RO Select: Yes Difficulty: 4 Group: SRO Imp: 4.0 SRO Select: Yes Taxonomy: Ap Question: RO: 68 SRO:I Procedure 1105.004, Integrated Control system limit and precaution states do not operate Reactor Demand H/A station in Auto with both S/Gs on low level limits.
What is the reason for this precaution and does any exception apply?
A. Due T-ave reduction as power lowers rods will pull to maintain T-ave at setpoint, you can operate with Reactor Demand H/A station in Auto with both S/Gs on low level limits if you adjust T-ave setpoint to match reactor power B. Due T-ave reduction as power lowers rods will not move due to T-ave error, you can not operate with Reactor Demand H/A station in Auto with both S/Gs on low level limits C. When S/Gs are on Low Level Limits, the Tave calibrating integral is blocked,, you can operate with Reactor Demand H/A station in Auto with both STGs on low level limits providing you verify calibrating integral is blocked on PDS.
D. When S/Gs are on Low Level Limits, the Tave calibrating integral is released, you can not operate with Reactor Demand H/A station in Auto with both S/Gs on low level limits.
Answer:
A. Due T-ave reduction as power lowers rods will pull to maintain T-ave at setpoint, you can operate with Reactor Demand H/A station in Auto with both S/Gs on low level limits if you adjust T-ave setpoint to match reactor power Notes:
A is correct, due to lowering power with S/G on LLL will cause Tave to ramp down. The Rx Demand station will try to pull rods to maintain 579 F. Limit & Precaution allows this mode of operation only if you reduce Tave setpoint to match Rx power.
B, C and D are incorrect
References:
OP-i 105.004 Change 20 History:
New selected for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0116 Rev: 0 Rev Date: 7/14/98 Source: Direct Originator: JCork TUOI: AILP-RO-NOP Objective: 7 Point Value: 1 Section: 2.0 Type: Generic K/As System Number: 2.2 System
Title:
Equipment Control
==
Description:==
Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.
KIA Number: 2.2.1 CFR
Reference:
45.1 Tier: 3 RO Imp: 3.7 RO Select: Yes Difficulty: 2 Group: SRO Imp: 3.6 SRO Select: Yes Taxonomy: K Question: RO:J 9 SRO:j During an INITIAL approach to criticality, if criticality is NOT achieved within of the ECC, then insert and A. Plus or minus 1.0% delta k/k control rods to achieve 1.5% SD margin establish hot shutdown conditions B. Plus or minus 1.0% delta k/k regulating groups to achieve 1.0% SD margin notify Reactor Engineering C. Plus or minus 0.5% delta k/k control rods to achieve 1.5% SD margin verify calculation D. plus or minus 0.5% delta k/k regulating groups to achieve 1.0% SD margin verify calculation Answer:
C. plus or minus 0.5% delta k/k control rods to achieve 1.5% SD margin verify calculation Notes:
Answer C is correct per 1102.008.
References:
1102.008, Chg. 023 History:
Used in 1998 RO exam Used in NRC developed RO exam 8/24/92, no. 88 Used in A. Morris 98 RO Re-exam Used in 2001 RO Exam Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0801 Rev: 0 Rev Date: 9/17/2009 Source: New Originator: S Pullin TUOI: AILP-RO-TS Objective: 7 Point Value: 1 Section: 2.0 Type: Generic K&A System Number: 2.2 System
Title:
Equipment Control
==
Description:==
Ability to determine operability and / or availability of safety related equipment.
K!A Number: 2.2.37 CFR
Reference:
41.7 /43.5 /45.12 Tier: 3 RO Imp: 3.6 RO Select: Yes Difficulty: 2 Group: SRO Imp: 4.6 SRO Select: Yes Taxonomy: Ap Question: RO:j 70 SRO: 70 REFERENCE PROVIDED Which of the following plant conditions would require entry into LCO 3.2.1 due to exceeding Regulation Rod Insertion Limits per the COLR?
A. 80% Power, 4 RCPs in service, 150 EFPD, Rod Index of 250%
B. 70% Power, 4 RCPs in service, 300 EFPD, Rod Index of 220 %
C. 60% Power, 3 RCPs in service, 100 EFPD, Rod Index of 265 %
D. 50% Power, 3 RCPs in service, 350 EFPD, Rod Index of 255 %
Answer:
B. 70% Power, 4 RCPs in service, 300 EFPD, Rod Index of 220 %
Notes:
Per the graphs in the COLR answer (B) falls within the Operation Restricted area of the figure and would require entry into LCO 3.2.1.
A, C, and D do not require entry into LCO
References:
ANO-1 Cycle 22 COLR Figures 3-A through 4-B History:
New for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0802 Rev: 0 Rev Date: 9/17/2009 Source: New Originator: S. Pullin TUOI: ASLP-RO-RADP Objective: 15 Point Value: I Section: 2.0 Type: Generic K&A System Number: 2.3 System
Title:
Radiation Control
==
Description:==
Knowledge of radiation exposure limits under normal or emergency conditions.
KlANumber: 2.3.4 CFR
Reference:
41.12/43.4/45.10 Tier: 3 RO Imp: 3.2 RO Select: Yes Difficulty: 3 Group: SRO Imp: 3.7 SRO Select: Yes Taxonomy: Ap Question: RO: 71 SRO:
Given:
- A General Emergency has been declared on Unit 1.
- A Maintenance crew must enter a radiological area with a dose rate of 150 Rem/Hr to protect valuable property.
Which of the following is the MAXIMUM time an individual team member can stay in this area?
A. 4 minutes B. 6 minutes C. 8 minutes D. 10 minutes Answer:
A. 4 minutes Notes:
A is correct, for protecting valuable property 10 Rem is the does limit.
B, C and D exceed 10 Rem limit.
References:
OP-i 903.033 Change 01 9-01-0 History:
New for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0803 Rev: 0 Rev Date: 9/17/2009 Source: New Originator: S Pullin TUOI: AILP-RO-EOP Objective: 2 Point Value: I Section: 2.0 Type: Generic K&A System Number: 2.4. System
Title:
Emergency procedure / plan
Description:
Knowledge of EOP mitigation strategies.
KIA Number: 2.4.6 CFR
Reference:
41.10 /43.5/45.13 Tier: 3 RO Imp: 3.7 RO Select: Yes Difficulty: 2 Group: G SRO Imp: 4.7 SRO Select: Yes Taxonomy: K Question: RO: 72 SRO:r 72 General rules of the Generic Emergency Operating Guidelines are that symptoms are treated whenever they occur based on priorities.
Which of the following transients has top priority per the GEOG?
A. Overheating B. Overcooling C. Loss of Subcooling Margin
. D. Steam Generator Tube Rupture Answer:
C. Loss of Subcooling Margin Notes:
C is correct per the GEOG LOSM has top priority.
References:
Volume 1 GEOG Part 1, Introduction History:
New for 2010 RO/SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0161 Rev: I Rev Date: 4/24/2002 Source: Direct Originator: J. Cork TUOI: Al LP-RO-AOP Objective: 4 Point Value: I Section: 2.0 Type: Generic K&A System Number: 2.4 System
Title:
Emergency procedure / plan
==
Description:==
Knowledge of abnormal condition procedures.
KIA Number: 2.4.11 CFR
Reference:
41.10 / 43.5 / 45.13 Tier: 3 RO Imp: 4.0 RO Select: Yes Difficulty: 3 Group: G SRO Imp: 4.2 SRO Select: Yes Taxonomy: C Question: RO:j73 SRO: 73 Given:
- Power escalation is in progress following a shutdown.
- Reactor power is 35%.
- Rod6 ofGroup7drops.
Which of the following actions should be taken?
A. Insert all regulating rods in sequential mode.
B. Trip the reactor and go to Reactor Trip, 1202.001.
C. Verify plant stabilizes at 320 MWe after ICS runback.
D. Verify SDM within COLR limit within one hour.
Answer:
D. Verify SDM within COLR limit within one hour.
Notes:
[a] would only be performed if power was <2%.
[b] would not be done because only one rod dropped.
[c] power is <360 MWe so there wouldnt be any runback, the value given would require a power increase.
[d] is the correct answer per TS.
References:
1203.003, Control Rod Drive Malfunction Action, change 023, page 12, step 4 History:
Developed for use in 98 RO Re-exam.
Used in 2001 RO/SRO Exam.
Selected for 2002 RO/SRO exam. Revised to agree with ITS.
Selected for 2010 RO/SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 818 Rev: 0 Rev Date: 2/5/2010 Source: New Originator: S.Pullin TUOI: A1LP-RO-EOP Objective: 2 Point Value: I Section: 2.0 Type: Generic K&A System Number: 2.4 System
Title:
Emergency procedure / plan
==
Description:==
Knowledge of general guidelines for EOP usage K/A Number: 2.4.14 CFR
Reference:
41 .10/45.13 Tier: 3 RO Imp: 3.8 RO Select: Yes Difficulty: 2 Group: SRO Imp: 4.5 SRO Select: Yes Taxonomy: K Question: RO: 74 SRO:
The EOP/AOP user guide procedure states Reactor Trip (1 202.001) is the entry point for all EOPs with one exception.
Which of the following is the exception?
A. Loss of SCM B. Overcooling C. Overheating D. Tube Rupture Answer:
D. Tube Rupture Notes:
D. is correct this is the only EOP that would be entered in the absence of Reactor trip A., B., and C, are incorrect these procedures are only entered after a Reactor trip
References:
1015.043 ANO-1 EOP/AOP user guide change 003 History:
New selected for 2010 RO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0242 Rev: 0 Rev Date: 9-1-99 Source: Direct Originator: D. Slusher TUOI: ANO-1-LP-RO-NNI Objective: 3 Point Value: I Section: 2 Type: Generic System Number: 2.4 System
Title:
Emergency Procedures/Plan
==
Description:==
Ability to identify post-accident instrumentation.
KIA Number: 2.4.3 CFR
Reference:
CFR: 41.6/45.4 Tier: 3 RO Imp: 3.5 RO Select: Yes Difficulty: 2 Group: G SRO Imp: 3.8 SRO Select: Yes Taxonomy: K Question: RO:175 SRO:
What instruments are marked with a green dot?
A. Instruments designated for use during an alternate shutdown.
B. Instruments that should be reliable during accident conditions.
C. Instruments the Shift Engineer uses after a reactor trip.
D. Instruments designated for use during a loss of NNI-Y power.
Answer:
B. Instruments that should be reliable during accident conditions.
Notes:
Instruments which are to be used for accident conditions are marked by a green dot as required by Reg Guide 1.97. Therefore, b is the correct answer.
(a) is incorrect because SPDS is designated for the alternate shutdown.
(c) is incorrect because SE instruments used after a reactor trip are designated by the 1015.037.
(d) is incorrect because NNI-X instruments are available in a loss of NNI-Y.
References:
1305.028 change 12 page 2 History:
Developed for 1999 exam.
Selected for the 2010 RO/SRO exam
ES-401 Record of RO Rejected KIAs Revision I Form ES-401-4 Tier / Randomly Selected K/A Reason for Rejection Group 1/2 036 Fuel Handling Could not write a credible question since only RO Accident AK2.01 Fuel action is to suspend fuel movement and exit.
Handling Equipment.
Randomly selected new system 068 Control Room Evacuation AK2.07 ED/C.
1/2 060 Accidental Gaseous The RO would not perform this function.
Radwaste Release Randomly selected new system 028 Pressurizer AK1 .04 Calculation of Level Malfunction AK1 .01 Pressurizer reference Offsite Doses due to release from the power leak abnormalities.
plant.
2/1 004 Chemical and Volume Could not write a credible question to match the K/A Control 2.1.34 and tie to that system.
Knowledge of primary and Randomly selected 2.2.38 Knowledge of conditions secondary chemistry and limitations in the facility license.
limits.
2/1 008 Component Cooling No credible tie for this K/A exists for the System.
Water A2.08 Effects of Loss of CCW Pump.
Randomly selected A2.01 shutting (automatically or otherwise) the isolation valves of the letdown cooler.
2/1 061 Emergency/Auxiliary Not possible to prepare a psychometrically sound Feedwater Al .04 AFW question related to the subject K/A.
source tank level Randomly selected Al .01 S/G level.
2/1 078 Instrument Air System Not possible to prepare a psychometrically sound (lAS) KI .03 question related to the subject K/A.
Containment Air.
Randomly selected Kl .02 Service Air.
2/2 001 Control Rod Drive No credible tie for this K/A exists for the System.
K2.05 M/G Sets Randomly selected K2.02 One-line diagram of power supply to trip breakers.
ES-401 Record of RO Rejected KIAs Revision I Form ES-401 -4 Tier / Randomly Selected K/A Reason for Rejection Group 2/2 068 Liquid Radwaste The RO would not perform this function.
K4.01 Safety and environmental precautions Randomly selected new system 014 Rod position indication K4.05 Rod hold interlocks.
for handling hot, acid, and radioactive liquids.
2/2 071 Waste Gas Disposal Not possible to prepare a psychometrically sound K3.05 ARM and PRM question related to the subject K/A.
systems.
Randomly selected new system 015 Nuclear Instrumentation K3.04 ICS.
3 2.3.11 Ability to control NRC comment that the ODCM was too cueing, radiation releases. therefore rejected 2.3.11.
Randomly selected new Generic K/A 2.4.11.
ES-401 PWR Examination Outline Form ES-401-2 Facility: Arkansas Nuclear One Unit I Date of Exam: 31512010 RO K/A CatpoyPoints SRO-Only_Poipts Tier Group KKKKKKAAAAG A2 G* Total 1 2 3 4 5 6 1 2 3 4
- Total
- 1. 1 3 3 6 Emergency &
Abnormal Plant 2 3 1 4 Evolutions Tier Totals N/A N/A 6 4 10 1 3 2 5 2.
Plant 2 0 1 2 3 Systems Tier Totals 4 4 8
- 3. Generic Knowledge and Abilities I 2 3 4 1 2 3 4 7 Categories 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.
The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1 .b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (lR) of 2.5 or higher shall be selected.
Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7* The generic (G) K/As in Tiers I and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1 .b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRS) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is Sampled in other than Category A2 or G on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
PWR Examination Outline Form ES-401-4 ES-401 Tier I Randomly Selected K/A Reason for Rejection Group 1/1 065 Loss of Instrument Air Could not write a credible SRO level question.
2.4.18 Knowledge of Randomly selected new system 038 Steam the specific bases for Generator Tube Rupture 2.4.18 EOPs 1/2 001 Continuous Rod Could not write a credible SRO level question.
Withdrawal AA2.05 Randomly selected new system 005 Inoperable stuck Uncontrolled Rod control rod AA2.03 Required actions if more than withdrawal from available one rod is stuck or inoperable.
indications 2/2 028 Hydrogen Could not write a credible SRO level question.
Recombiner and Purge Randomly selected new system 016 Non-Nuclear Control 2.4.23 Instrumentation 2.2.40 Ability to apply Technical Knowledge of the bases Specifications for a system.
for prioritizing emergency procedure implementation during emergency operations.
ES-401 PWR Examination Outline Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions Tier 1/Group 1 (SRO)
E!APE # / Name / Safety Function K/A Topic(s) 000007 (BW/E02&E10; CE/E02) Reactor EA2.1- Facility conditions and selection of Trip Stabilization Recovery /1 appropriate procedures during abnormal and emergency operations 000008 Pressurizer Vapor Space Not selected Accident / 3 000009 Small Break LOCA /3 Not selected 000011 Large Break LOCA / 3 Not selected 000015/17 RCP Malfunctions/4 Not selected 000022 Loss of Rx Coolant Makeup /2 AA2.04- How long PZR level can be maintained within limits 2.4.31 Knowledge of annunciator alarms, 000025 Loss of RHR System / 4 indications, or response procedures.
000026 Loss of Component Cooling AA2.O1- Location of a leak in the CCWS Water / 8 000027 Pressurizer Pressure Control Not selected System Malfunction / 3 00029 ATWS / I Not selected 000038 Steam Gen. Tube Rupture /3 2.4.18 Knowledge of the specific bases for EOP5.
000040 (BW/E05; CE/E05; W/E12) 2.4.6- Knowledge of symptom based EOP Steam Line Rupture Excessive Heat
- mitigation strategies Transfer / 4 000054 (CE/E06) Loss of Main Not selected Feedwater / 4 000055 Station Blackout /6 Not selected 000056 Loss of Off-site Power / 6 Not selected 000057 Loss of Vital AC Inst. Bus / 6 Not selected 000058 Loss of DC Power /6 Not selected 000062 Loss of Nuclear Svc Water / 4 Not selected 000065 Loss of Instrument Air / 8 2.4.18 Knowledge of the specific bases for EOP5 Rejected system to 038 Steam Gen Tube Rupture W/E04 LOCA Outside Containment /3 Not selected ES-401 Form ES-401-2
Form ES-401-2 ES-401 PWR Examination Outline PWR Examination Outline Form ES-401-2 ES-401 Em encv and Abnormal Plant Evolutions Tier 1/Group 1 (SRO)
K K K A: A G K/A Topic(s) IR # OlD T E/APE#/Name/Safety Function y 123 12V.
V p
V e
NIA V
W/E1 1 Loss of Emergency Coolant V
Not selected V
Recirc. /4 Not selected NIA BW/E04; W/E05 Inadequate Heat V Transfer Loss of Secondary Heat Sink / 4 Not selected N!A 000077 Generator Voltage and Electric = V Grid Disturbances / 6 =
V K/A Category Totals: =
= 1 3j Group Point Total: 6 ES-401 Form ES-401-2
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0588 Rev: 0 Rev Date: 6/1/05 Source: Direct Originator: J.Cork TUOI: Al LP-RO-EOPO4 Objective: 11 Point Value: I Section: 4.3 Type: B&W EPEs/APEs System Number: ElO System
Title:
Post-Trip Stabilization
==
Description:==
Ability to determine and interpret the following as they apply to the (Post-Trip Stabilization):
Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
KIA Number: EA2.1 CFR
Reference:
43.5 /45.13 Tier: 1 RO Imp: 2.5 RO Select: No Difficulty: 3 Group: I SRO Imp: 4.0 SRO Select: Yes Taxonomy: C Question: SRO:.1 76 RO:j Given:
- Reactor tripped due to a loss of both MFWP5 approximately 15 minutes ago.
- Annunciator K02-B6 A3 L.O. RELAY TRIP is in alarm.
- AFW pump, P-75, is tagged out for maintenance.
- Steam Driven EFW Pump, P-7A, has tripped on overspeed.
- RCS pressure is 2000 psig.
- GETs are 612°F.
- Both OTSG levels are 30.
Which of the following procedures should be in use for the above conditions?
A. 1202.002, Loss of Subcooling Margin B. 1202.004, Overheating C. 1202.011, HPI Cooldown D. 1203.037, Abnormal ES Bus Voltage Answer:
B. 1202.004, Overheating Notes:
Answer B is correct, the Overheating EOP should be entered with CETs> 610°F and all MFW and EFW lost during loss of adequate Subcooling Margin.
Answer A is incorrect, this procedure would have been in use up to the point where CETs became > 610°F.
Answer C is incorrect, this procedure is entered from Loss of Subcooling Margin.
Answer D is incorrect, this procedure is used when ES bus voltage is low but not de-energized.
References:
1202.004, Chg. 006 History:
New for 2005 SRO exam.
Selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
Rev: 0 Rev Date: 9/21/2009 Source: New Originator: S. Pullin QID: 0805 TUOI: AILP-RO-TS Objective: 13 Point Value: I Section: 4.2 Type: Generic APEs System Number: 022 System
Title:
Loss of Reactor Coolant Makeup to Reactor Coolant Makeup: How
==
Description:==
Ability to determine and interpret the following as they apply long PZR level can be maintained within limits.
KIA Number: AA2.04 CFR
Reference:
43.5 / 45.13 RO Imp: 2.9 RO Select: No Difficulty: 4 Tier: 1 Group: 1 SRO Imp: 3.8 SRO Select: Yes Taxonomy: Ap Question: RO:j SRO:I Given:
- RCS Cooldown in progress
- Tave is 295 F
- RCS Pressure is 440 psig.
- Pressurizer level is 65 inches
- All makeup has been lost
- Pressurizer level is dropping at 5 inches per minute
- Assuming pressurizer level rate of change remains the same When will LCO 3.4.9 Pressurizer, be entered due to low Pressurizer level and what is the bases per Technical Specification for the low level?
A. 2 minutes and to maintain the minimum ES bus powered pressurizer heaters OPERABLE.
- 8. 2 minutes and to maintain on scale pressurizer level indication.
C. 4 minutes and to maintain the minimum ES bus powered pressurizer heaters OPERABLE.
D. 4 minutes and to maintain on scale pressurizer level indication.
Answer:
D. 4 minutes and to maintain on scale pressurizer level indication.
Notes:
water level limit has D is correct, the limit per LCO 3.4.9 is less than or equal to 45 inches and the minimal been established to ensure that water level is above the minimum detecta ble level.
per OP-i 102.010 A is incorrect, due to PZR level would be 55 inches which is below the administrative limit for PZR level, but does not require entry into the LCO.
per OP-1102.010 B is incorrect, due to PZR level would be 55 inches which is below the administrative limit for PZR level, but does not require entry into the LCO.
which would C is incorrect, due to PZR level would be 45 inches which is at the pressurizer heater cutoff level deenergize the ES powered heaters.
References:
T.S. 3.4.9 Amendment 215 History:
New selected for 2010 SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
Rev: 0 Rev Date: 9/21/2009 Source: New Originator: S. Pullin QID: 0806 TUOI: AILP-RO-AOP Objective: I Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 025 System
Title:
Loss of RHR System procedures.
==
Description:==
Knowledge of annunciator alarms, indications, or response K/A Number: 2.4.31 CFR
Reference:
41.10 / 45.3 RO Imp: 4.2 RO Select: No Difficulty: 3 Tier: 1 Group: I SRO Imp: 4.1 SRO Select: Yes Taxonomy: Ap Question: RO: SRO:F Given:
- Mode 5
- RCS temperature 170 F
- CV-1050 and CV-1410 interlocks are not bypassed
- RCS pressure 0 psig
- A RCP seal removed for maintenance
- A Decay Heat in service
- Following alarms are received
- DECAY HEAT FLOW HI/LO (K09-A8)
- DECAY HEAT VORTEX WARNING (K09-D8)
- ISOL VLV OPEN RC PRESS LO (Ki 0-E5) the given conditions?
Which section of OP-1203.028, Loss of Decay Heat Removal, will be entered for A. Section 6, Decay Heat Pump Trip B. Section 7, Suction Valve Closure C. Section 9, Loss of Both DH Systems RCS Pressure Boundary Intact D. Section 10, Loss of Both DH Systems RCS Pressure Boundary Open Answer:
B. Section 7, Suction Valve Closure Notes:
valve to close.
B is correct, with the given alarms K10-E5 would automatically cause the DHR Suction for the given condition. The pump does not automatically A is incorrect, the DHR Pump would still be running stop on valve closure.
loss of both DHR C is incorrect, although the RCS is still intact with an RCP seal removed, the transition to Pumps does not occur until RCS temperature is greater than 280 F until RCS D incorrect, the RCS is not open and the transition to loss of both DHR Pumps does not occur temperature is greater than 280 F
References:
OP-I 203.028 Change 021 op-1203.0121 Change 046 History:
New selected for 2010 SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0807 Rev: 0 Rev Date: 9/21/2009 Source: New Originator: S Pullin TUOI: AILP-RO-AOP Objective: 3 PointValue: I Section: 4.2 Type: Generic APEs System Number: 026 System
Title:
Loss of Component Cooling Water.
to the Loss of Component Cooling
==
Description:==
Ability to determine and interpret the following as they apply Water: location of a leak in the CCWS.
K/A Number: AA2.01 CFR
Reference:
43.5 / 45.13 RO Imp: 2.9 RO Select: No Difficulty: 3 Tier: 1 Group: I SRO Imp: 3.5 SRO Select: Yes Taxonomy: C Question: RO:1 SRO:1 Given:
- Plant at 100%
- The following alarms are received
- ICW COOLER OUTLET TEMP HI (K12-E4)
- RCP BEEDOFF TEMP HI (K08-C7)
- A RCP seal temperature rising
- Skewed RCP Seal Injection flows indicated on C04
- RCS leak rate is 50 gpm abnormal operating condition?
Which of the following procedures provide the actions necessary to mitigate the A. OP-I203.039, Excess RCS Leakage B. OP-1203.026, Loss of Reactor Coolant Makeup C. OP-I 203.031, Reactor Coolant Pump and Motor Emergency D. OP-1102.016, Power Reduction and Plant Shutdown Answer:
A. OP-1203.039, Excess RCS Leakage Notes:
that combats an intersystem A is correct, because Excess RCS Leakage procedure is the only procedure LOCA.
leaks, but with the B is incorrect, OP-1203.026 has a section to address makeup & purification system indications given this is not considered a makeu p & purific ation system leak.
as a seal failure issue, C is incorrect, with the given indications the student could misdiagnose this a rapid plant shutdo wn would be necess ary.
D is incorrect, with the given leak rate,
References:
OP-1203.039 Change II History:
New selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0585 Rev: 0 Rev Date: 9/21/2009 Source: New Originator: B. Possage TUOI: AILP-RO-EOP Objective: 9 Point Value: I Section: 4.1 Type: Generic EPEs System Number: 038 System
Title:
Steam Generator Tube Rupture
==
Description:==
Knowledge of the specific bases for EOPs.
K/A Number: 2.4.18 CFR
Reference:
41.10 / 43.1 / 45.13 Tier: 1 RO Imp: 3.3 RO Select: No Difficulty: 3 Group: 1 SRO Imp: 4.0 SRO Select: Yes Taxonomy: Ap Question: RO: SRO:j Given:
- SGTR in progress
- Rxistripped
- RCS pressure 1350 psig
- RCS Thot 540°F
- Projected dose rate at site boundary at NUE criteria
- B SG level at 395 and rising rapidly
- A° SG level stable at 40° Considering the above conditions, which of the following procedural actions will cause higher tube stresses than normal limitations but is acceptable during a SGTR per the EOP technical bases document?
A. Perform a cool down to less than 500°F at 100°F/hr and isolate bad SG.
B. Steam bad SG to maintain bad SG Tube-to-Shell DT <1 50°F (tubes colder).
C. Steam bad SG to maintain bad SG Tube-to-Shell DT <100°F (tubes hotter).
D. Establish a cool down rate of 250°F/hr to 500°F Thot.
Answer:
B. Steam bad SG to maintain bad SG Tube-to-Shell DT <150°F (tubes colder).
Notes:
B is correct, per Technical Bases during emergency cool downs the tube to shell delta T limits are relaxed.
With the given information an emergency cool down is required at the rate of </ 240 F/hr.
A is incorrect, this rate is the normal cool down rate.
c is incorrect, this is the normal tube to shell delta T limit.
D is incorrect, this exceeds the allowed emergency cool down limit.
References:
OP-1202.006 Change 11 B&W EOP Technical Bases Document History:
New selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0584 Rev: 0 Rev Date: 5/20/05 Source: Direct Originator: J.Cork TUOI: A1LP-RO-EOPO3 Objective: 10 Point Value: I Section: 4.1 Type: Generic EPEs System Number: 040 System
Title:
Steam Line Rupture
==
Description:==
Knowledge of symptom based EOP mitigation strategies.
K/A Number: 2.4.6 CFR
Reference:
41.10 / 43.5/45.13 Tier: 1 RO Imp: 3.7 RO Select: No Difficulty: 4 Group: I SRO Imp: 4.7 SRO Select: Yes Taxonomy: An Question: RO: SRO: F present:
A steam fine rupture has occurred in the Reactor Building with the following conditions now
- ESAS actuated on channels I thru 6.
- RB pressure 19 psig and dropping.
- HPI throttled due to existence of adequate SCM.
- RCS pressure is 1050 psig.
- T-hot is 490°F.
- EOP actions have terminated the overcooling.
ESAS and The SE recommends to the CRS to restore normal operating pressure per RT-14 in order to reset re-start RCP5.
As CRS, does this recommendation follow the EOP mitigation strategies?
A. Yes, overcooling event has been terminated.
B. No, this could overstress reactor vessel.
C. Yes, adequate SCM has been restored.
D. No, RB pressure is not within normal limits.
Answer:
B. No, this could overstress reactor vessel.
Notes:
limits B is correct, trainee must recognize that with RCPs secured and HPI having been initiated that PTS pressure. PTS limits prevent overstressing apply until an evaluation is performed prior to returning to normal reactor vessel.
violate A is incorrect, yes the overcooling has been terminated but normal operating pressure would procedure.
procedure.
C is incorrect, subcooling margin was never lost but normal operating pressure would violate D is incorrect, although RB pressure is a concern the overriding concern is with PTS concerns.
THIS QUESTION IS TIED to 43.1
References:
1202.012, chg. 004-03-0, RT-14 History:
New for 2005 SRO exam.
Selected for the 2010 SRO exam
m m Cl)
Cl) 0 0
m 3
0 0
-I.
z
.1 0
-I 3
m Cl) 0 0
-I 3
m Cl) 0
m m Cl)
Cl, 0
0 -
m x
- ** 3 0 00 C .
0 0 g g -a D) 0 z
3(0(0 CD V Cl) 0 0 0
- g 0.0. (0 g 0.0 CD CD CD CD CD CD 0 Cn5fl CD CD 0
- 0. Z V Q)
Cl)
.0 0.
CD .CD
-n *0*
0 3
m .p.
Cl) 0 (-I
-I, (I
0 CD -I (0 3 m
ci Cl) 0
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I Rev: 0 Rev Date: 6/1/05 Source: Direct Originator: S.PuIlin QID: 0589 Objective: 4 Point Value: 1 TUOI: A1LP-RO-TS Section: 4.2 Type: Generic APEs System Number: 005 System
Title:
Inoperable/Stuck Control Rod as they apply to the Inoperable/Stuck Control
==
Description:==
Ability to determine and interpret the following Rod: Required actions if more than one rod is stuck or inoperable.
K/A Number: AA2.03 CFR
Reference:
43.5 / 45.13 3.5 RO Select: No Difficulty: 4 Tier: I RO Imp:
2 SRO Imp: 4.4 SRO Select: Yes Taxonomy: An Group:
Question: RO:J SRO:j Given:
- Plant is at 40% power.
- Group 4, Rod 4 is stuck and is mis-aligned from the group by 7.5%.
- The rod can not be re-aligned with the group.
Subsequently Group 7 Rod 6 drops to 0% withdrawn.
above conditions?
What are the required action(s) per Technical Specifications for the A. Open Control Rod Drive breakers, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Rate surveilllance, B. Borate to restore SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and perform Linear Heat SR 3.2.5.1, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
rod worth is within C. Borate to restore SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and verify the potential ejected the assumptions of the rod ejection analysis within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. Borate to restore SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and place the plant in Mode Answer:
3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. Borate to restore SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and place the plant in Mode Notes:
Answer D is correct per TS 3.1.4 action C for two inoperable rods.
Answer A is incorrect, this action is performed for two dropped rods.
Answer B is incorrect, this action is performed for one inoperable rod and the time given for the stated condition is incorrect.
Answer C is incorrect, this action is performed for one inoperable rod and the time given for the stated condition is incorrect.
References:
T.S. 3.1.4 amendment 215 Do not include this spec in the student handout!!!
History:
New for 2005 SRO exam.
Selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT 1 -
QID: 0808 Rev: 0 Rev Date: 9/22/2009 Source: Modified Originator: S.Pullin TUOI: A1LP-RO-POISN Objective: 14 Point Value: 1 Section: 4.2 Type: Generic Abnormal Plant Evolutions System Number: 0024 System
Title:
Emergency Boration Boration:
==
Description:==
Ability to determine and interpret the following as they apply to the Emergency Amount of boron to add to achieve the required SDM.
K/A Number: AA2.05 CFR
Reference:
43.5 / 45.13 Tier: I RO Imp: 3.3 RO Select: No Difficulty: 4 Group: 2 SRO Imp: 3.9 SRO Select: Yes Taxonomy: Ap Question: RO: SRO:1 REFERENCE PROVIDED
- Rx has tripped with three CRDs stuck full out.
- Core lifetime = 150 EFPD
- RCS initial Boron concentration = 810 ppm
- Chemistry reports that the RCS boron concentration is 2200 ppm.
Which of the following contains guidance that must be used, for the above conditions?
A. No action required, SDM is adequate B. 1202.012, RT-12 Emergency Boration C. 1203.017, Moderator Dilution D. 1103.015, Reactivity Balance Calculation Answer:
B. 1202.012, RT-12 Emergency Boration Notes:
adequate Answer B is correct, using Att. 8-16 from the plant data book, the examinee should determine that Boration must be performed until adequate SDM is established.
SDM has not been established and Emergency Answer A is incorrect, SDM is not adequate.
for Answer C is incorrect, although this might seem like a logical choice, this procedure should not be used these conditions.
Answer D is incorrect, although this might seem like a logical choice, use of the Reactivity Balance Calculation procedure does not have any plant actions in it.
References:
1202.012 RT-12, Chg. 008 CALC-ANO1 -NE-08-00007 NOTE: CALC Att. 8-16 must be in SRO handout!!!!
History:
Modified from QID 678 Selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
Rev: 0 Rev Date: 6/6/05 Source: Direct Originator: S.Pullin QID: 0591 Objective: I Point Value: 1 TUOI: Al LP-RO-ANNI Section: 4.3 Type: B&W EPEs/APEs System Number: A02 System
Title:
Loss of NNI-X they apply to the (NNI-X): Facility conditions
==
Description:==
Ability to determine and interpret the following as ncy operations.
and selection of appropriate procedures during abnormal and emerge K/A Number: AA2.1 CFR
Reference:
43.5/45.13 3.6 RO Select: No Difficulty: 4 Tier: I RO Imp:
SRO Imp: 4.0 SRO Select: Yes Taxonomy: An Group: 2 Question: RO: SRO:. 84 Given:
- Pressurizer Level Control Valve CV-1 235 indicates 50% open.
- RC Pump Seals Total lnj Flow valve CV-1207 indicates 50% open.
- Letdown flow indication is zero.
- Letdown pressure indication is zero.
- Letdown Orifice Bypass valve CV-1223 indicates 50% open.
- RCS pressure is 2210 psig and slowly rising.
- Pressurizer Spray valve CV-1 008 indicates closed.
What procedure should be in use due to the above conditions?
A. 1203.015, Pressurizer Systems Failure B. 1203.024, Loss of Instrument Air C. 1203.047, Loss of NNI Power D. 1203.012B, ACA for K10-A8 LETDOWN TEMP HI Answer:
C. 1203.047, Loss of NNI Power Notes:
of a loss of NNI X and Y power.
Answer C is correct since the conditions given are representative due to something other than a loss of Answer A is incorrect, this would be in use if Spray valve was failed NNI power.
to loss of IA, but the positions given are Answer B is incorrect, this would be in use for failed valves due different than for loss of air alone.
n flow would still be indicated while the Answer D is incorrect, this is chosen for hi letdown temp but letdow question states there is none.
References:
1203.047, Chg. 000-01-0 History:
New for 2005 SRO exam.
Selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0592 Rev: 0 Rev Date: 6/6/05 Source: Direct Originator: S.Pullin TUOI: Al LP-RO-ASDCD Objective: 2 Point Value: 1 Section: 4.3 Type: B&W EPEs/APEs System Number: E08 System
Title:
LOCA Cool down other support procedures or
==
Description:==
Knowledge of EOP implementation hierarchy and coordination with guidelines such as, operating procedures, abnormal operating procedures, and severe management guidelines KIA Number: 2.4.16 CFR
Reference:
41.10 /43.5 /45.13 I RO Imp: 3.5 RO Select: No Difficulty: 4 Tier:
Group: 2 SRO Imp: 4.4 SRO Select: Yes Taxonomy: Ap Question: RO:i SRO:j 85 Given:
- Rx was shutdown using 1203.045 Rapid Plant Shutdown,
- Due to a RCS leak
- RCS pressure 1720 psig and lowering slowly
- HPI flow 150 gpm
- A & B SG pressure 910 psig
- RCS cool down rate 35°F per hour
- All Turbine bypass valves closed Which procedure should be in use?
A. 1202.001, Overcooling B. 1203.041, Small Break LOCA cool down C. 1203.040, Forced Flow cool down D. 1202.01 0, ESAS Answer:
B. 1203.041, Small Break LOCA cool down Notes:
Answer B is correct with an uncontrolled cool down continuing due to break/HPI flow, regardless of SG status.
Answer A is incorrect, Overcooling entry conditions have not yet been met Answer C is incorrect, although RCPs are running, there is no control of the cool down.
Answer D is incorrect, although parameters are close to ES actuation setpoints, the ESAS procedure would eventually transition to 1203.041.
References:
1203.039, Chg. 011 History:
New for 2005 SRO exam.
Selected for 2010 SRO exam
PWR Examination Outline Form ES-401-2 ES-401 PWR Examination Outiine Form ES-401-2 Pntystsms-Trpf(RO)
K/A Topic(s) IR # QID T K K K K K K A A A A G 1234561234 y p
e X A2.02 Conditions which 3.9 86 809 N 003 Reactor Coolant Pump exist for an abnormal shutdown of a RCP in comparison to a normal shutdown of RCP Not Selected N/A 004 Chemical and Volume Control Not Selected N/A 005 Residual Heat Removal Not Selected N/A 006 Emergency Core Cooling Not Selected N!A 007 Pressurizer Relief/Quench Tank Not Selected N/A 008 Component Cooling Water X A2.02 Spray failures 3.9 87 762 R 010 Pressurizer Pressure Control Not Selected N/A 012 Reactor Protection X A2.06 Inadvertent ESFAS 4.0 88 812 N 013 Engineered Safety Features Actuation actuation Not Selected NIA 022 Containment Cooling Not Selected N/A 025 Ice Condenser Not Selected N/A 026 Containment Spray Not Selected N/A 039 Main and Reheat Steam Not Selected N/A 059 Main Feedwater X 2.2.22 Knowledge of 4.7 89 811 N 061 Auxiliary/Emergency Feedwater limiting conditions for operations and safety limits Not Selected N/A 062 AC Electrical Distribution X 2.2.42 Ability to recognize 4.6 90 810 N 063 DC Electrical Distribution system parameters that are entry-level conditions for Technical Specifications Not Selected N!A 064 Emergency Diesel Generator Not Selected N/A 073 Process Radiation Monitoring Not Selected N!A 076 Service Water Not Selected N/A 078 Instrument Air Not Selected N/A 103 Containment 3 2 Group PointTotal: 5 K/A Category PointTotals:
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I Rev: 0 Rev Date: 9/23/2009 Source: New Originator: S Pullin QID: 0809 Objective: 6 Point Value: I TUOI: A1LP-RO-AOP Section: 3.4 Type: Heat Removal from Reactor Core System Number: 003 System
Title:
Reactor Coolant Pump System (RCPs) ctions or operations on the RCPs; and
==
Description:==
Ability to (a) predict the impacts of the following malfun mitigate the consequences (b) based on those predictions use procedures to correct, control, or abnorm al shutdown of an of those malfunctions or operations: Conditions which exist for an RCP in comparison to a normal shutdown of an RCP K/A Number: A2.02 CFR
Reference:
41 .5/43.5/45.3/45/13 RO Imp: 3.7 RO Select: No Difficulty: 3 Tier: 2 1 SRO Imp: 3.9 SRO Select: Yes Taxonomy: C Group:
Question: RO:j SRO:j 86 Given:
- 100% Power,
- C RCP seal bleed off temperature 210 F.
- C RCP motor bearing temperature 185 F.
- C RCP motor inboard vibration alert alarm,
- C RCP seal cavity pressure oscillating from 650 to 1250 psig.
t Pump What is the appropriate section and action of 1203.031, Reactor Coolan te the conseq uences of these malfun ctions?
and Motor Emergency which will mitiga ty of unaffected A. Section 2, Seal Failure, Reduce reactor power to within the capaci Operation, OPIIO3.006.
RCP combination and stop the affected RCP per Reactor Coolant Pump B. Section 2, Seal Failure, Trip the Reactor and trip the affected RCP.
ty of unaffected C. Section 5, Motor / Bearing Trouble, Reduce reactor power to within the capaci per Reacto r Coolan t Pump Operat ion, OP1 103.006.
RCP combination and stop the affected RCP d
D. Section 5, Motor / Bearing Trouble, Trip the Reactor and trip the affecte RCP.
Answer:
B. Section 2, Seal Failure, Trip the Reactor and trip the affected RCP.
Notes:
in cooling (seal injection or ICW B is correct, a seal bleedoff temperature of greater than 200 F with no change flow) meets the requirements to trip the RCP due to seal failure section.
instead of a normal shutdown A is incorrect. The given conditions require an abnormal shutdown of an RCP of an RCP.
of a normal shutdown C is incorrect. The given conditions require an abnormal shutdown of an RCP instead ofanRCP.
ts stopping the RCP.
D is incorrect. The given conditions do not indicate a bearing problem that warren
References:
OP-1203.031 Change 018 History:
New selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0762 Rev: 0 Rev Date: 11/11/200 Source: Repeat Originator: Steve Pullin TUOI: ANO-1-LP-RO-RCS Objective: 6 Point Value: 1 Section: 3.3 Type: Reactor Pressure Control System Number: 010 System
Title:
Pressurizer Pressure Control System (PZR PCS)
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Spray valve failures KIA Number: A2.02 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.13 Tier: 2 RO Imp: 3.9 RO Select: No Difficulty: 4 Group: 1 SRO Imp: 3.9 SRO Select: Yes Taxonomy: An Question: RO:j SRO:I 87 Given:
-Unit I is operating at 40% power.
-The Unit is in three pump ops due to the failure of P-32B.
-The Pressurizer Spray Control valve (CV-1 008) fails open.
The ATC attempts to close the Pressurizer Spray Isolation valve (CV-1 009) and it will NOT close
-Reactor Coolant Pressure is at 2100 psig and slowly lowering with all Pzr Heaters on.
What is the correct procedure and correct action for this condition?
A. 1202.001 Reactor Trip, and trip the Reactor.
B. 1202.001 Reactor Trip, and stop P-32C.
C. 1203.015 PZR System Failure, and trip the Reactor.
D. 1203.015 PZR System Failure, and stop P-32C.
Answer:
D. 1203.015 PZR System Failure, and stop P-32C.
Notes:
A is incorrect. Since the Power to Pump trip entry conditions are not met.
B is incorret with the correct action but with the incorrect procedure since the Power to Pump trip entry conditions are not met.
C is incorrect with the correct procedure but incorrect action.
D is correct.
References:
1203.015 Pzr System Failure Chg 16 History:
New for the 2009 Retake SRO Exam Selected for 2010 SRO exam REPEAT
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0812 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S. Pullin TUOI: A1LP-RO-ESAS Objective: 6 Point Value: I Section: 3.2 Type: Reactor Coolant System Inventory Control System Number: 013 System
Title:
Engineered Safety Features Actuation System
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based ability on those predictions use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadvertant ESFAS actuation.
K/A Number: A2.06 CFR
Reference:
41.5 /43.5 / 45.3 / 45.13 Tier: 2 RO Imp: 3.7 RO Select: No Difficulty: 3 Group: I SRO Imp: 4.0 SRO Select: Yes Taxonomy: Ap Question: RO:j SRO:I 88 Given
- Plant at 100% power
- P-2B Condensate Pump OOS
- Inadvertent actuation of ES Channel #1
- S/U #1 QOS for maintenance LCO 3.8.1 .A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Time Clock in effect What would be the impact to the plant due to this malfunction and what procedure would be used to mitigate the effects?
A. #1 Emergency Diesel Generator would start and use OP-1105.003, Engineered Safeguards Actuation System to reset the tripped channel.
B. Red Train High Pressure Injection would occur and use 1202.01 0, ESAS EOP to override HPI C. Loss of power to A-i bus and use 1202.001, Reactor Trip EOP D. All Seal Return isolates and use 0P1203.031, Reactor Coolant Pump and Motor Emergencies to realign seal bleed off.
Answer:
C. Loss of power to A-i bus and use 1202.001, Reactor Trip EOP Notes:
C is correct, the Unit Aux supply breaker to A-i would open on ES Channel #1 actuation and would result in a reactor trip due to a loss of all Condensate pumps resulting in a Iosss of Main Feedwater.
A is incorrect, although the EDG would start with a reactor trip the EOP would have priority over securing the EDG B is incorrect, although HPI would occur the ESAS EOP would not be utilized to secure HPI for an inadvertant actuation.
D is incorrect, seal return would be realigned to the Quench Tank rather than isolate.
References:
STM 1-32 Rev 33 OP-1107.001 Change 073 History:
New selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0811 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S Pullin TUOI: A1LP-RO-EFIC Objective: 43 Point Value: 1 Section: 3.4 Type: Heat Removal from Reactor Core System Number: 061 System
Title:
Auxiliary / Emergency Feewater System
==
Description:==
Knowledge of limiting conditions for operations and safety limits K/A Number: 2.2.22 CFR
Reference:
41.5 / 43.2 / 45.2 Tier: 2 RO Imp: 4.0 RO Select: No Difficulty: 3 Group: I SRO Imp: 4.7 SRO Select: Yes Taxonomy: C Question: RO:F SRO:J 89 Given
- A SG Low level transmitter feeding the D EFIC Channel failed Lo
- B SG Pressure transmitter feeding the C EFIC Channel failed Hi What operator actions are required per Technical Specifications?
A. Place D channel in bypass per 3.3.11 .A B. Place C channel in bypass per 3.3.11 .B C. Trip D channel per 3.3.11 .B D. Trip C channel per 3.3.11 .A Answer:
B. Place C channel in bypass per 3.3.11.8 Notes:
B is correct, the Low Level transmitter failing low will result in a trip of the D Channel, 3.3.11 .B requirements for two inoperable channels requires one to be placed in bypass and the other one tripped.
A is incorrect, D Channel is already tripped and placing in bypass would have no effect. TS 3.3.11 .A is only applicable to one inoperable channel. The question asks what to do for two inoperable channels C is incorrect, because it is only half of the action required by 3.3.11 .8 D is incorrect because tripping C Channel would result in an EFIC actuation.
References:
TS 3.3.11 Amendment 215 History:
New selected for 2010 SRO exam
A ARKANSAS INITIAL RO!SRO EXAM BANK QUESTION DAT NUCLEAR ONE - UNIT I Rev Date: 9/23/2009 Source: New Originator: S. Pullin QID: 0810 Rev: 0 Objective: 5 Point Value: 1 TUOI: A1LP-RO-TS Section: 2 Type: Generic Knowledge and Abilities System Number: 063 System
Title:
DC Electrical Distribution meters that are entry-level conditions for Technical
==
Description:==
Ability to recognize system para Specifications K/A Number: 2.2.42 CFR
Reference:
41.7/41.10143.2143.3/45.3 3.9 RO Select: No Difficulty: 3 Tier: 2 RO Imp:
SRO Select: Yes Taxonomy: C Group: I SRO Imp: 4.6 Question: RO:1 SRO:l entry into Tech nical Specification 3.8.4, DC Sources, Operating Which of the following conditions requires tion 3.8.4?
and what is the bases for Technical Specifica Battery operable.
A. DO4A, Battery Charger inoperable and D06, Bases is to insure reactor cool ant press ure boun dary limits are not exceeded as a resul t of abno rmal ities
, Battery Charger inoperable.
B. DO4B, Battery Charger inoperable and DO3B dary limits are Bases is to insure reactor coolant pressure boun not exceeded as a result of abnormalities
, Battery Charger inoperable.
C. DO4A, Battery Charger inoperable and DO4B ided, and reactor building operability Bases is to insure adequate core cooling is prov a postulated DBA and other functions are maintained in the event of Battery operable.
D. DO3B, Battery Charger inoperable and D07, uate core coolin g is prov ided, and reactor building operability Bases is to insure adeq a postulated DBA and other functions are maintained in the event of Answer:
, Battery Charger inoperable.
C. DO4A, Battery Charger inoperable and DO4B
, and reactor building operability Bases is to insure adequate core cooling is provided a postulated DBA and other functions are maintained in the event of Notes:
le train being inoperable, the subsystem is inoperab C is correct, with both battery chargers on the same ing is prov ided , and 3.8.4 is to insure adequate core cool requiring entry into TS 3.8.4. The bases for TS are maintained in the event of a postulated DBA reactor building operability and other functions subsystem.
g inoperable does not affect the operability of the A is incorrect, Only one of the two charges bein The bases used for this option is partially correct. affect the le but since they are on different trains they do not B is incorrect, two battery chargers are inoperab ally correct.
this option is parti operability of either subsystem. The bases used for .
char ges bein g inop erable does not affect the operability of the subsystem D is incorrect, Only one of the two The bases used for this option is partially correct.
References:
T.S. 3.8.4 Amendment 215 History:
New selected for 2010 SRO exam
Form ES-401-2 PWR Examination Outline ES-401 Form ES-401-2 ES-401 PWR Examination Outline ystems Tier 2/Group 2 (SRO)
TYPO G K/ATopic(s) IR # QID System#/Name K K K K K K A A A A 1234561234 Not selected NIA 001 Control Rod Drive Not selected NIA 002 Reactor Coolant Not selected N/A 011 Pressurizer Level Control Not selected N/A 014 Rod Position Indication Not selected NIA 015 Nuclear Instrumentation X 2.2.40 Ability to apply 4.7 91 599 D 016 Non-nuclear Instrumentation technical specifications for a system Not selected N/A 017 In-core Temperature Monitor Not selected NIA 027 Containment Iodine Removal 2.4.23 Knowledge of the NIA 028 Hydrogen Recombiner bases for prioritizing and Purge Control emergency procedure implementation during emergency operations.
Rejected system replaced with 016 Non-Nuclear Instrumentation Not selected N!A 029 Containment Purge Not selected N/A 033 Spent Fuel Pool Cooling X 2.1.40 Knowledge of 3.9 92 600 0 034 Fuel Handling Equipment refueling administrative requirements.
A2.01 Faulted or ruptured 4.6 93 813 N 035 Steam Generator X S/Gs.
Not selected NIA 041 Steam Dumplrurbine Bypass Control Not selected NIA 045 Main Turbine Generator Not selected N!A 055 Condenser Air Removal Not selected N/A 056 Condensate Not selected N/A 068 Liquid Radwaste Not selected N!A 071 Waste Gas Disposal Not selected N/A 072 Area Radiation Monitoring Not selected N/A 075 Circulating Water Not selected NIA 079 Station Air Not selected NIA 086 Fire Protection K/A Category PointTotals: = [ =
i ((2[ Group PointTotal: 1 Form ES-401-2 ES-401
A ARKANSAS INITIAL ROISRO EXAM BANK QUESTION DAT NUCLEAR ONE - UNIT I Rev Date: 6/27/05 Source: Direct Originator: J.Cork QID: 0599 Rev: 0 Objective: 35 Point Value: I TUOI: Al LP-RO-NNI Section: 3.7 Type: Instrumentation System Number: 016 System
Title:
Non-Nuclear Instrumentation for a system.
==
Description:==
Ability to apply technical specifications K/A Number: 2.2.40 CFR
Reference:
41.10/43.2/43.5 /45.3 RO Select: No Difficulty: 4 Tier: 2 RO Imp: 3.4 SRO Select: Yes Taxonomy: Ap Group: 2 SRO Imp: 4.7 Question: RO: SRO:1 91 REFERENCE PROVIDED The plant is operating at 100% power.
have failed LOW.
Both PZR level transmitters LT-1 001 and LT-1 002 cal Specification 3.3.15 and Table 3.3.15-1?
Which of the following actions are required by Techni A. Be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
B. Both channels must be restored within 7 days.
30 days or be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
C. Restore one channel to operable status within days or be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. Restore one channel to operable status within 7 Answer:
days or be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. Restore one channel to operable status within 7 Notes:
-1 and actions C and E.
Answer D is correct in accordance with Table 3.3.15 of 7 days per action C.
Answer A is incorrect, there is still an allowance be restore d.
Answer B is incorrect, only one channel must ation of A and E.
Answer C is inocrrect, this is a combin
References:
T.S. 3.3.15 Amendment 232 Note: T.S. 3.3.15 must be in students handout.
History:
Direct from regular exam bank QID#ANO-OPSI-6623 Selected for 2005 SRO exam.
Selected for 2010 SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
Rev: 0 Rev Date: 6/27/05 Source: Direct Originator: J.Cork QID: 0600 TUOI: Al LP-RO-FH Objective: 16 Point Value: I Section: 3.8 Type: Plant Service Systems System Number: 034 System
Title:
Fuel Handling Equipment
==
Description:==
Knowledge of refueling administrative requirements.
K/A Number: 2.1.40 CFR
Reference:
41.10 / 43.5 / 45.13 RO Imp: 2.8 RO Select: No Difficulty: 3 Tier: 2 2 SRO Imp: 3.9 SRO Select: Yes Taxonomy: K Group:
Question: RO: SRO: r 92 Given:
- Plant is in a Refueling outage.
- Core re-load is in progress.
- Approximately 90% of the core is in the Reactor vessel.
the process of indexing over the The Main Fuel Handling Bridge has a once-burned fuel assembly and is in specified core location when Nl-502 fails to 0.1 cps.
What action should be taken?
ments A. No action necessary because with Nl-501 operating, Tech Spec NI require for operablility are met.
location B. Contact the Main Fuel Bridge operator and place the assembly in a core without any adjacent fuel assemblies.
d unless C. Halt operations on the Main Fuel Bridge. Core geometry cannot be change two neutron flux monito rs are operab le.
and D. Verify boron concentration in the Refueling Canal is greater than 2326 ppm then continue fuel load.
Answer:
two neutron flux C. Halt operations on the Main Fuel Bridge. Core geometry cannot be changed unless monitors are operable.
Notes:
Answer C is correct per 1502.004, 5.3, and T.S. 3.9.2 alterations.
Answer A is incorrect, although only one is required in Mode 6, two Nis are required during core Answer B is incorrect, this is still a core alterati on.
Answer D is incorrect, this is simply a requirement for refueling.
References:
1502.004, Chg. 041 T.S. 3.9.2 Amendment 215 History:
Direct from regular exam bank QID#3178 Selected for 2005 SRO exam.
Selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0813 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S Pullin TUOI: Al LP-RO-EOPO6 Objective: 4 Point Value: I Section: 3.4 Type: Heat Removal from Reactor Core System Number: 035 System
Title:
Steam Generator System (S/GS)
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the S/G and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulted or wptured S/Gs.
KIA Number: A2.01 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.5 Tier: 2 RO Imp: 4.5 RO Select: No Difficulty: 4 Group: 2 SRO Imp: 4.6 SRO Select: Yes Taxonomy: An Question: RO:j SRO:l 93 Given:
- Plant at 100% power Simultaneously the following occurs:
- Reactor trips on low RCS Pressure
- N-16 alarm on A Steam Generator
- Steam Line High Range Radiation monitor RI-2681 in alarm.
- RCS pressure drops to 1300 psig
- CETs indicate 550°F
- Reactor Building and Aux Building sump levels are stable.
Starting with 1202.001, Reactor Trip EOP, which of the following lists the order of EOPs to mitigate this event?
A. 1202.002 Loss of Subcooling Margin and 1202.006 Tube Rupture B. 1202.002 Loss of Subcooling Margin and 1202.010 ESAS C. 1202.006 Tube Rupture and 1202.010 ESAS D. 1202.006 Tube Rupture and 1202.012 RT-l0 Answer:
A. 1202.002 Loss of Subcooling Margin and 1202.006 Tube Rupture Notes:
A is correct, The Reactor Trip EOP immediate actions will send the operator to Loss of Subcooling margin, with the only LOCA being a tube rupture the Loss of Subcooling Margin procedure will send the operator to Tube Rupture.
B is incorrect, ESAS would only be entered if RCS pressure dropped below 150 psig.
C and D are incorrect, Reactor Trip would send the operator to Loss of Subcooling Margin EOP first.
References:
OP-1202.00l Change 031 OP-1202.002 Change 006 History:
New selected for 2010 SRO exam
0 d) z C
z D z Cl) I LU Ei E CD 0 0 a* CD C) CD CD LL 4 C?
CD F CD 0) 0 CD 4 0) 0) 0) 0) 0) 0) f.
0 0 LU 0
F-- 0. c
- a -. Q E I...
. C) a* C) C) 0 U
-o o 0 .U) C C 5 -Q) .92 a)
Co C) G) 9 C
C4 In -° E
C) .-
0 . C 0 ECo x . . a)0 D
. 0 a)
Iii CU ci
- N V C
10
.- V 0 a)
Q 0 )
0 E a) .* C .
Co Ca) 0 C
Q._
E 0 cU o
CU-S aE5 a 0=
E 0 C (U
- C) ..
0 .
C a)
>. C a)
Ci) (U.S Co C) 0)
x .C) 0CC .5 0 Ca) 00 0_
LU V 0jCU CU *D C O C (U CCU
- CU C DC)
.2 th.50 a C-
- 0 0) .2 0).
a)
C .
0 C C
.5
- j
- 0) 0 0.- 0 0 000 O
> E 0 a) (1) a)E a)
- 0) Cl) 0) 0 0 - 0) 0 0 C 2 V V V V C V 0 V 0)0 a) 0)0 a) a) ..0CU a)0) c C
0)0):;
.OVC C 0 C C CD CDD CC
<2.co C) SV Eci, co a) 0 I 44 a. -
CU C C-1 C) C)
- 0) < r-C) - - Cl) c-a 0 c-a Ci) e Cl) r c-0 (0
0 C Cl) -
(0 g 5
C)
I a)_ 0... 0 a
0 .
C I. C.) . SQ 5C 0C coo Q)OC E2.!
C)
- 0) 0 0 CU CU *0 o jZ LU LU L1 C.) -Qo c.LIJO .jW0..Q..
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0492 Rev: I Rev Date: 12/4/06 Source: Direct Originator: S.PulIin TUOI: A1LP-RO-EOPO8 Objective: 7 Point Value: 1 Section: 2.0 Type: Generic Knowledges and Abilities System Number: 2.1 System
Title:
Conduct of Operations operating
==
Description:==
Ability to evaluate plant performance and make operational judgments based on characteristic s, reactor behavior, and instrument interpretation .
KIA Number: 2.1.7 CFR
Reference:
41.5 / 43.5 / 45.12 / 45.13 3 RO Imp: 4.4 RO Select: No Difficulty: 4 Tier:
Group: SRO Imp: 4.7 SRO Select: Yes Taxonomy: An Question: RO:j SRO:J Given the plant conditions following a reactor trip:
- RCS temperature: 605 degrees stable
- RCS pressure: 2300 psig slowly dropping
- ERV: open in AUTO
- OTSG shell temperature: 558 degrees
- OTSG levels 20 inches, steady
- PZR level 180 inches, rising Which of the following actions are required?
A. Trip the running RCP per 1202.002, Loss of Subcooling Margin.
B. Isolate the ERV per 1202.001, Reactor Trip.
C. Select the reflux boiling setpoint per RT-5.
D. Initiate Full HPI per RT 3.
Answer:
B. Isolate the ERV per 1202.001, Reactor Trip.
Notes:
RCS pressure Answer B is correct. A pressurizer steam space leak is indicated by PZR level rising with have closed at 2395 psig.
dropping and no rise in RCS temperature. ERV is open and should however the Answer A is incorrect, Tube to Shell delta T of 60 degrees tubes hotter would require this action delta T is only 47 degrees in the question.
subcooling Answer C is incorrect, although RCS temperature/pressure conditions are close to a loss of margin which would require selection of Reflux Boiling but SCM is still adequate.
EOP in Answer D is incorrect, Full HPI would be required if the ERV opened in Auto with the Overheating effect but the Overheating entry conditions are not met.
References:
1202.001, Chg. 031 History:
Modified from regular exambank QID#3314.
Used on 2004 SRO Exam.
Modified for use on 2007 SRO Exam.
Selected for 2010 SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0814 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S Pullin TUOI: AILP-RO-FH Objective: 4 Point Value: I Section: 2 Type: Generic Knowledge and Abilities System Number: 2.1 System
Title:
Conduct of Operations
==
Description:==
Knowledge of the fuel-handling responsibilities of SROs K/A Number: 2.1.35 CFR
Reference:
41.10/43.7 Tier: 3 RO Imp: 2.2 RO Select: No Difficulty: 3 Group: G SRO Imp: 3.9 SRO Select: Yes Taxonomy: K Question: RO: SRO:I Which of the following conditions would require the SRO in charge of fuel handling to order a stop to fuel movement in the Reactor Building?
A. Outage Control Center reports that the reactor has been subcritical for 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br />.
B. National Weather Service declares a Tornado Watch in effect for Conway County.
C. One Control Room Emergency Air Conditioning System (CREACS) inoperable for the past 5 days.
D. Reactor Building Radiation monitor RE-8017 inoperable, and portable survey instrument is being monitored on the fuel handling bridge.
Answer:
A. Outage Control Center reports that the reactor has been subcritical for 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br />.
Notes:
A is correct, the reactor must be subcritical for greater than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to fuel movement.
B is incorrect, Pope, Johnson, Yell and Logan counties in a tornado watch would require stopping fuel movement. Conway county is immediately east of Pope county.
C is incorrect, with one CREACS channel inoperable we have 30 days to repair prior to stopping fuel movement.
D is incorrect, RE-8017 is desired to be operable for monitoring radiation levels on the bridge, however if it becomes inoperable any portable survey instrument is allowed for monitoring rad levels and continue fuel movement.
References:
OP-I 502.004 Change 041 History:
New selected for 2010 SRO exam.
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0646 Rev: 0 Rev Date: 10/23/200 Source: Direct Originator: CorkfPossage TUOI: A1LP-RO-EDG Objective: 2 Point Value: 1 Section: 2 Type: Generic Knowledge and Abilities System Number: 2.2 System
Title:
Equipment Control
==
Description:==
Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.
K/A Number: 2.2.25 CFR
Reference:
41.5 /41.7 I 43.2 Tier: 3 RO Imp: 3.2 RO Select: No Difficulty: 3 Group: G SRO Imp: 4.2 SRO Select: Yes Taxonomy: C Question: RO:J SRO:I 96 Given:
- #1 EDG has one Air Start Compressor and its associated Air Receiver Tanks tagged out.
- The remaining Air Start Compressor on #1 EDG trips while EDG is running for a surveillance.
- The Air Receiver Tanks pressure is 145 psig.
In accordance with Technical Specifications, what is the required action for the above conditions?
A. No actions are necessary since the EDG is running and an air start system is not needed.
B. Restore required starting air receiver pressure to within limits in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
C. Declare #1 EDG inoperable immediately.
D. Be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Answer:
C. Declare #1 EDG inoperable immediately.
Notes:
Answer C is correct, with only one receiver bank and pressure <158 psig the EDG must be declared inoperable per 3.8.3.E.1.
Answer A is incorrrect, although the EDG is running, if it tripped there would not be enough air for a re-start.
Answer B is incorrect, this is the action from 3.8.3.D and would be applicable if pressure was between 158 and 175 psig.
Answer D is incorrect, this action is from 3.8.1 .F and would be applicable if the EDG was not made operable within 7 days.
References:
3.8.3 and Bases Amendment 215 History:
Uses QID 447 stem with some modifications, all answers are different, therefore it is a new question.
New question for 2007 SRO exam.
Selected for the 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0815 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S Pullin TUOI: ASLP-SRO-MNTC Objective: 2 Point Value: 1 Section: 2 Type: Generic K&A System Number: 2.2 System
Title:
Equipment Control
==
Description:==
Knowledge of maintenance work order requirements K/A Number: 2.2.19 CFR
Reference:
41.10 / 43.5 / 45.13 Tier: 3 RO Imp: 2.3 RO Select: No Difficulty: 3 Group: G SRO Imp: 3.4 SRO Select: Yes Taxonomy: K Question: RO: SRO: 97 Given:
- Power 100%
- All equipment operable and there are no Tech Spec LCOs in effect
- Annunciator K12-B5, P-7A Turbine Trip alarm is received
- WCO reports that the linkage for the trip throttle valve has broken.
You are the Shift Manager, Per EN-WM-100, Work Request (WR) Generation, Screening and Classification, which work order process should be used to correct this condition.
A. Emergency Maintenance B. Priority One Work Order C. Priority Two Work Order D. Priority Three Work Order Answer:
B. Priority One Work Order Notes:
B is correct, since P-7A inoperability is a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Time Clock, a Priority I work order would be initiated to begin maintenance and work around the clock to completion.
A is incorrect, Emergency Maintenance is only used to protect the public or prevent serious injury/death C is incorrect, Priority 2 work orders are entered into the T-3 week schedule and would exceed TS time limit D is incorrect, Priority 3 work orders timing would exceed TS time limit
References:
EN-WM-1 00 Rev 3 History:
New selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0816 Rev: 0 Rev Date: 9124/2009 Source: New Originator: S Pullin TUOI: A1LP-RO FH Objective: 4 Point Value: 1 Section: 2 Type: Generic Knowledges and Abilities System Number: 2.3 System
Title:
Radiation Control
==
Description:==
Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.
KIA Number: 2.3.14 CFR
Reference:
41.12 /43.4 / 45.10 Tier: 3 RO Imp: 3.0 RO Select: No Difficulty: 3 Group: G SRO Imp: 3.8 SRO Select: Yes Taxonomy: C Question: RO: SRO:
During a fuel handling accident Krypton-85 is the major source of gaseous activity released from a damaged Fuel assembly that has decayed for >190 days.
Which portion of the body will receive the highest dose after a fuel handling accident?
A. Skin dose from Beta B. Whole body dose from Gamma C. Extremities dose from Beta D. Internal Organ dose from Gamma Answer:
A. Skin dose from Beta Notes:
A is correct, skin dose rates from K-85 are 100 times higher than the whole body gamma dose rates.
B, C, and D are all incorrect.
References:
OP-1203.042 Change 005-03-0 History:
New selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0411 Rev: 0 Rev Date: 12/1/00 Source: Direct Originator: E-Plan TUOI: ASLP-RO EPLAN Objective: 7 Point Value: 1 Section: 2 Type: Generic Knowledges and Abilities System Number: 2.4 System
Title:
Emergency Procedures/Plan
==
Description:==
Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the transmission system operator.
It/A Number: 2.4.30 CFR
Reference:
41.10/43.5/45.11 Tier: 3 RO Imp: 2.7 RO Select: No Difficulty: 2 Group: G SRO Imp: 4.1 SRO Select: Yes Taxonomy: C Question: RO:j SRO:I A fire was reported at 0844 in the vicinity of the Old Radwaste Building.
It is now 0920 and the fire is still burning.
What is the Emergency Plan time requirement for notification to the NRC?
A. Notification to the NRC is required within 15 minutes of the declaration of an emergency class.
B. Notification to the NRC is required immediately following notification of the ADH and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the declaration of an emergency class.
C. Notification to the NRC is required immediately following declaration of an emergency class and notify the ADH within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
D. Notification to the NRC is required within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the declaration of an emergency class.
Answer:
B. Notification to the NRC is required immediately following notification of the ADH and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the dedaration of an emergency class.
Notes:
Answer [B] is correct since this is the procedural requirement.
Answer [A], [C], [D]are incorrect, these are not in accordance with 1903.011.
References:
1903.OIIY, Emergency Initial Notification Message Change 036 History:
Modified E-Plan exam bank QJD#61 for use in 2001 SRO Exam.
Selected for use in 2002 SRO exam.
Selected for 2010 SRO exam
INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -
QID: 0750 Rev: 2 Rev Date: 6/23/08 Source: Direct Originator: Spullin TUOI: A1LP-RO-AOP Objective: 5 Point Value: 1 Section: 2.0 Type: Generic K/As System Number: 2.4 System
Title:
Emergency Procedures / Plan
==
Description:==
Knowledge of local auxiliary operator tasks during an emergency and the operational resultant effects K/A Number: 2.4.35 CFR
Reference:
41.10/43.5/45.13 Tier: 3 RO Imp: 3.8 RO Select: No Difficulty: 3 Group: SRO Imp: 4.0 SRO Select: Yes Taxonomy: C Question: RO:J SRO: I 00 Given:
- Severe Fire on 335 Auxiliary Building on Unit I
- Reactor has been tripped Which of the following actions would the CRS direct the Outside AC to perform and what procedural guidance would be used?
A. Fire fighting tasks per Fire or Explosion procedure 2203.034.
B. Securing Polishers per Reactor Trip/Outage Recovery procedure 1102.006.
C. Placing the Startup Boiler in service per Startup Boiler Operation procedure 1106.022.
D. Throttle CV-2627 EFW Supply to A SG per Fires in Areas Affecting Safe Shutdown procedure 1203.049 Answer:
D. Throttle CV-2627 EFW Supply to A SG per Fires in Areas Affecting Safe Shutdown procedure 1203.049 Notes:
A. is incorrect; due to recent procedures changes have the opposite Unit ACs fighting the fire, but the WCO non licensed operator has fire fighting duties B. is incorrect; under normal Reactor trip conditions this would be an Outside AC action promptly following Rx trip, but it is not in the Reactor procedure C. is incorrect; under normal Reactor trip conditions this would be an Outside AC would perform following Rxtnp D. is correct; this is a new procedure action for the non licensed operators
References:
1203.049 Fires in Areas affecting Safe Shutdown Change 005 History:
Selected for 2010 SRO exam
SRO NRC Exam HANDOUT FOR 2010 EXAM
RCS Pressure (100 psig Increments) 0 (TI 0 I)
(TI 0 01 0 0 0 o o 0 0 a 0 o e 0 rs) 0 0
ni 0
01 0 0
Cl) 11 0
C C.)
0 0 m C) Co Ci Cl)
C, -I C.)
0 (Ii A C) CD 0
-0.o 3 e 0 1 0 n CD
-I 0
!. 0 0 0
Q -I 0
C (TI 0
- m 0
> 0
- 0. 1 (TI m 0 0
IV IV IV 4 Cli Ci. C 000 Dl C, 0 0 0 -
-n,.l-l1 CD (TI (I) (TI 0
C) 3 CD 0) 0 0
m 0)
(TI 0
0 0
0 ii 0
-I 0)
0 SG Pressure (10 psig Increments) 43 0
- 1.) F3 C C) .h . (Yi C) 0) 0) J 1 0) 0) CD CO 0 0 - C..3 Ch 0 C) 0 C) 0 C) 0 C) 0 C) 0 C) 0 (11 0 C) 0 () 0 C) 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0
o m
0 (I
0 C m
L C.)
0 0
G)
(1) 6)
cJ1 -
o 1 0 mm C)
CD I
I I I I (I) I
- 0) I I
I C) 0 o I 1x I im I II I I 1<
I I C)
C) 0 L I>
IG) 0)
Im 0
0 It43 lo I-.
Io
CD
.4-.
0 0 ,
is rE flWfl PWC*flflfl ww WflW W*VV?W WnrCcaWP ?aWPO iOW flXVAW Wfl FW rrw. Wnc rvr,ZPW iiwte wttaUwr4.2 n i, .uqw%4dafl ctn Wfl WI. Fb JWr 4IW%4dt? ItJ1t% tft .d%tWt4% ettJ CD
-, m %kflWn.v.TC w.a1wrww WflW W.rg .fl Wi C%VPC S
&r4Ct Ltit,4tt V4Ctti flt4 *.fltr4tt tJ2CWI2 CC.r41tv
, . fl
%kflW tJ4It tPJCtW tttL$ WZtV. JCt tCWItZt
.fl ,lwn a 19 CA Th qzr fl Ib r CIW A UA %PØ lC flIA. ICrnJW. V 4W d .cCI1& l.flA .ra V
- rn 3
L *yn twrIxa.
t wLrrWn4
.. rrce r
bflW fl 1 L I c r- CD u -wCa rnt etfltt.r@,l
- n::::::::. r t E it r - °
< 4zt,:!
uSJ43 I***n*** nn* ****
SaflflflflSaflUUflfl.. -E* HflI&
- sna**
t_
- _ w,arg WC,i4W t - *4 t 102 Y Mt SJ*t 44
.CøCo øn
? &
tCOa %
Y WZ M.Y agn t23fl. tt fl5J S t!$fl%fl W 0 CD
- a. aCfr.nr.ae 4i h 3*St. .ICU34O4 % & W .flIC 4f.AYk&u.P4% % rEWt D4t r9 *Ia JrrzR rt rMr ct rmc
.;n;n;;;:;;::;;;:;;;;;;::;:::;;;;:;:;;:::;i.;r $r$ tfl I.*m*** fl*a* ._._. *fl**fl* **** ** **fl ***fl_-z flY mwn r- ---i øWflt.Q nrs C.... an c.
q- !W%L ffi CCLU!. JO & ti 4 V JtJ%tIJ WW %V & J LLtJ tt J It t:a*:ILOC flfla fl*** **** *na* on. n** **fl***** ***fl**flfl*S*flat2 tV9! tfiC,.fl I :. .. tt.R.$ Or,R.IX eX.j
- 2*t_ 1.1 nnnn*.n****n*n**n*nn*nnnn***n*nns*n*n***n. -t 1+ fl 2 .- t *% t %t* ttt e .
.% .;;;;n:;;:;::;::;;:;;;:;::::;::;;::::;;;;;;:;:;;:;;z;;;;;;;;;nit. 4 .
LU 0 I
PRZC .n..**n. .mm *** *flS*
twor, ...fln*** ..... fl... ......... ..n..*. ..... ........................ **nn F av..z I*** .n.* **n. *.n......nn* nn. ..... .n... nPaC.h n Yw I
P Rwn%2 t n tait ow .rw
.r g fl.tr. ..Itrt.t w c. vg? tg.t
._ SIrJZVbI ..***n Sfl ** **fl a***S*N *aS . %nntt.I!nrnwnra3nntn )IVflt %%iVJI
., 4PJW *nnmnm *fln n.m. nan ..... ...n .....**** ..fl.n... .n.. ******** fl _-fl e$ . S U3 fl tBS t$ V 6* CD rdiI.nIII nnnn* *flfl nan an.. *n** flflflnI **fl* **n*nnmnn* ****n*** ...uu & G E
a ltttt!. IP4 ,:tI!a;, iO C , <t s?Jov r ssys 4
tt;t;.ttw ttiI rtht q
It SC I xrju ***fl*flfl u.n. ***n **fl* ***** .*fl**fl SS*** *...n******n..Sfl..* ..... *fl** *****
.nn..... *fl ****D ne.. *fl kt**.*s**n**n**.*n***n***as..**.*n*n*nN*n.**a**a*n**mannn**n***n****ns***nh S*** *** ***fl *nmnn*nn*e*n*nmm ..n. nn* ***fl U*b t:%ntn.nti%gv LPJ2t3 wt.aI.-tr4ft -ç4 ran.!t o
a
- R***** ...n **S **** **** **** **S******* ***N***RRfl***flS* **K** R*** **** **b L
Wt t Iu3CNuy t CLfl
- P
!alCt J ytt 4
IC!
cr E V CSfl *t O2 t?
D tPMN)i 4I V 34 WtI fl t C) ttz,fl
. Ch1 Cu &k U t e o t__ r4 ZN *S** ..a..nn. a.... ..n. ......n** i..n.*........nn. ..... .n.. nn. ..n. .n.*n..L th pCc,CaLl *2:qt2 ***S* *** ** ***** ***** ******S******** **fl N*** ***** *** **** *fl*flnr hOt C M tara,r4hnNt k.tn .,rntwe tCC ... : ... ...fl Nfl.. ***** .*****e*....******* ..._. on. *.... ...*u *.... ...** mm*. .I t fi n.muszt.*
Cit C b2 0 w At.n.nn.n.n..n ...n.nn..n.a..n .n...n.nn.n..n. .nn.n...nnaIi ncItotflmr flfg Vrt a.r%kflflh* flC I ru F fl aF4fl t )JJ 3O. ........n.n.....n...................n..n..n...n....a...n.na . . fl Aøtp%;w..M4I 1/4.5r.4%v %flflt% 1C%t. it S ..n.nn* ***** ****fl ***** .n.. .n.......n ,%.tC eQ%Z *4tt P ItL tOLt4*. VJ.tt.. UI to n... nn. ..... ..n. e iw
ø ...fl.........fl...fl........fl..............fl.fl...fln.n.
E jet
,, t ia W tct 4,.;,
w ntrfloz J.ar .1%
gi,rnasc J nt.nrd *nfl%fl-*
4zt
.I..
rn,
.r.
.1!.
r..
4ø%mrr, ILL ...
a 2.
r .
at w
i ez Cw&Ci.r 9cM. tie....i.
F..e.cf9 C99.
t.
- 9. 99 9 7.
.t.i.
ft Ar.CCn
-.*n.s.....s....a.......n.s...n.n..........sn,a
........s**..a.... ......***......***sss**s**s *h .9 7.9 fl D i_ d.l.
t wvn 7tt 71171119499
.r .
.7 1.
r:::::::::::::::::::::::::::::::::::::: ,,. ci I
- 1. 74 9 ,7c .9
- 741 949W449 *t2i7
.174 7.17,. 1749,. 2 * .2 19791119249 11224i 7
9249*7111242 17714 774,2 9W7,11....
Cd 7 7 & 7 .2
.5*5* *fl* a.... ..saa *****
555***5R**a55**fl**555*asfla*
- a*** *a5**ahiia* 447.72271 41194 1551 41 .94 0 vs 7,.4144t11 717.
9.78 11774 1771 M.C1.74127 114 .1741.
17S,97ta4918 749 C
9e1424Wo 77174. 87, 17 tiflCt 29 n,,
17 4*71 931.7 74
.74 i n79 771.. 74474 7277tS749 7717 7.94919,!.
C 7.
Ia%
S 947.7,9 4771
.1. 5fl*a*555a5555IIn
........._:..............a 7 0 0
7 9,,!,
7171.7, O771r7 CltiJJi7liItlikN 4 .97 971 7191471 17 1, 11 17411% 7411 77 447 9i%uW 77,1 .7 2 9.
9.1 274 we iCe 9727S81C
%.C4454t 4141911 17.9 7,20 9.1 .191. 14.717,1 17.194 411 71%. 7474 97 74 C1194t4477i, .9 93*7 21.1 717
.774 s...ss.nnni *WSUb9.1E ws*****sasa**ss**saasai au... c: 9.9 7117 LU 7
17774871747127,72196 9114711197712111971727,191.91745 1I.77 891291. 914271 177,177121* 2.11199121919$1x9 9174712912.191747,2712,1
...............i 1974*4..,.... S*55a555*I 41 555 .97 .7 79) 0.
C)
I.. s t, a. 2 t; a .. 5
.:::::::::u::::
4979794117 CN
- ait. rcu.eti .izi7)*. roui v* **zi*e. ci iM.va41 11 .1711 271. &*. 1971*1 7, 2419*7 .941 7l.1 .277 .211 4,97519744 417. il i...*ssssa ia nra. It, .477
- .at r. .99.1.9. .8 C :::z::::: ::::: ,j LU 11 49 797979.772 .99114947 47,7,2 9471 7 .6 97l11tfl i92919,7W611. 9.9421.979 9.749. 791147977 7, 41i949 9.9776 1. .9 7,79 9, 5 g479n 7941 7 *55555S 55W*5 CD C117157,2247717.4 947.277121714497,19741791 7411777, 2248717.491771 9217.27717,1991871*481.7.12 C5.911*2 1447197 12411.7.7. ,1.l714 9911241.777,121711, aassss IS 5F4fl *i 7,447 C .974,47.. 499 17 127711 17,49 C 7 .9 99..,.,,, .71479.1744, 447, 87 771 7 4,994,,,99 7174 79,9 7,4 11,69 474. 47,.. .4 55555554 ii5lfl55t C 1, 71.. tltat 47941 97,9717797t, .a. rLn94.77flt. .17 811n97.77. r9h.in 477.
°$J it i1.4: 4 79 91.
t 9
- n::iu:::::: . g U) 7120 82147C2091177C7.1.7W1.71 114C7.C 5.,.40t.197 494 .74 1IC91.772,71191%93 41417t.7491 7,47197717911.V4%.194tI.9 79.411 47 9.477
°L ;:::.
947.7,94 55flhI55s5ssssal4 ci C..)
,t,,t,.9o1, 17e*,
)*,) as,. L. r 9 J j ..
it ,l1. 7777*. C711. a t
t1a 74171. 4i7474itit 77711.1171*. r17491,71 91O7 11776711 7,4499.5449749594 797aftc ,l944 i3.120t .1599 9487971141 .9.724491174*. 97,.*e.d7,wtP l. 477744 4 93$. 4,777 .7 si.
19.42
- a. Ia..r.i 917.
1i,747,6.91)J 71,. . 1174
.79
- I&7t t n ::;::::::
dt,4s 44 ii.7 ...i S I5S 55559 7.2917 1 8IrC727.,Cw 7*9 4.7 lap. 17 8117 22 71w1rCw?r444 !7475,.1l79*.847, 27 71727 7117 774 U,71l f175 9, flnlarnnarn::rna CD 7 1. 41 1711. 7793 9817 877 71. 44747. .91174 954 .919 8111.98194 171s. 7914712 474 177. 351 96744 ,97114 7 914914. 474 *17, 7 .917 sill sasssss*sa F
- 7. lIt 74,97,. 47,97.7440 .t7f 9.1949 7.80 2 CWt %7tt172OLL 7.1,7, 91.90 7.17 779992 91 9.64 P .t 16 4Cc 7,47 174 11.4 5 i ....ss..sai 941199489111717731717997744979$9442 79917979$fl97.949 727,119179 7419117971719917777297271791794412911719191787119914 i%1991199117146$1147172$it$141991197119,,19744t.1i4117479919 11171*71179914911 flu is inn ....i 2 77.9 7579991749 779 487t84i24+t7917 47217 044
.17 271 .719? .78,. 7.9527 718 77.98 8 577 7? 9 .12 717 98 ,28?!.2487 118 27, 7 2,, 9 78 .2 2 807,. 7 2 7 49 1:511 ::::u 2 *111 sssssss5S 0
CD 7,727791197949919C2119I,IC7, 77295,776.*.Il96, 1.17Cfl4917791190 49711*7C91917C911917779177976.117118.977119981912 799111.91C71177.49119477491i c,W9.9.%1991717,7149747,11 747.14744 15 11755555555555 7 .7 77 77 877*, 71947,911 27, 16 11.78 817*, 70977,790 17 1. 491.8711.849 *7 4 91475 21.1770 .7 isiiIsnsaa*asn 744 7 1. 7 7WV!.r.49,r.97r)
J 2 fl.OlflI3c 9.779917.79 941744 9,594 947441177 9,14* 94976 7.1 7714*799 4 9.547499.26 CD 94177,4114*7.1 fl7t1a7,2#,47%.t4j77b4l37%4l,jA* 17% A%71317% %493t%V21*4I7 7 7. *817174 .2 %4t%914t%t12Ii%FI7 7.974 48 lSliIiflSSSSSSS5S 7797.919.17 79177177141.877 687741.12 574 711. 497. *17 rA19%t7777%7C91 08194177, 4717, 74474. 4714 179.7 .9494 jl9 7217 5 41777 47,4 41 isiiIssssa*55555 27I128tlret.14W8t17t917,4.,,4ra 79F 69,L* 8 7 tL1,l7a7t493$, 7427,9.77 7.607 77 779. 80179014277.4% 97,. 2, 4 448e0,4P 7 . *.
I 51.. 7741 492 747747. 1717
I3 RCS Pressure (50 psig Increments)
- - F) F) F) F.,) F) C,) b F) . 0) 0) 0 4,) . 0) 0) 0 4%) . 0) 0 0) 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 F,)
0 0 =
m C,) 0 0
0 00
_c*)
0 o
0 m
.Q.
,, (DC C C) 0-
-os 0 3 00 0
0.
I-..
rn 0
0 UI 0
- 0) m 0
0 -
0
-4, 0)
1202.013 EOP FIGURES REV 4 PAGE 5 of 6 FIGURE 5 SG Pressure to Establish 40° to 60°F Primary to Secondary AT 1100 1050 1000 Desired operating region 950 900 850 40°F Primary to Secondary 800 750 w
E 700 650 U,
a O 550 500 D 60°F Cl) 450 Primary to Secondary w 400 a
C! 350 Cl, 300 250 200 150 100 50 0
250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 GET TEMPERATURE (5°F Increments)
.3 0
0 m
0 Cl) 0 G) 1
-o G)
C C,
C, m Cl) 0 m
C, D) 0*
C, CDl 0
oG)
I-I.
oc a)
- 0 0
Cl)
CD C,
0 0.
m
-t
-U 0
m 0*)
0 0)
0 C)
(Under-excited) MEGAVARS (Over-excfted) g g C C C 0 0 0 0 0 0 0
roO iir Irt 1flr[rirt[i((r rr r
1 Y1TTtt4t111 Tt T T1i1111i111I411 11.%. 111111111 C rErrrrrrrr,TTTTTT,,, r TT T T T TTT111111I 1JT 1)1 m 1T,TTTIjrr t tttiiiit i i 41tH}tN FH IH)H 30c) 1 E1 0 4-Th(1)NN4 CJ)UN -HfH-ft-IHI-FU tttttf1tl1ltlI l1i o 11iiI11f CD -
CD CD 0
tUUh1tt1iN(1tJli bJH 0 a 0 -* -
1L1 +1I1 + 1 F [H-I- + IIt + -I +It T9-1 t+/-1
[ [tt1tt11.fl1k1I tIN .l CD t[tLtLtMiT4t11 11L liii D C) 0 - - -
I ft hiit1tIit1l 1l Ili N CD r 1 D)CD C, *D CD 0
o -II - -
Lt -
Co___ co CD CD c_____
- 101 CD m CO 01 o C)
F.) o________ -
I- I 01 01
-n -n if. M
ATTACHMENT G Page 1 of 1 BAAT Volume and Concentration Vs. RCS T-ave (Ref. TRM Figure 3.5.1-1) 00 00
- 00 5000 00
.00 2000 1000 0
200 300 400 500 600 8700 ppm 09500 ppm 10,000 ppm 12,000 ppm
I PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE:
I PAGE: 67 of 127 1104.003 CHEMICAL ADDITION CHANGE: 046 ATTACHMENT H Page 1 of 1 Volume of BAAT vs. Depth of Liquid 8000 7000 6000 rr - - -- - -
- 4__
4:/ :
E-5000 (0
C
- E i-- -____
4000 0
(0 L -- I H L J H 3000 2000 <
1000 H H: :::: :::: ::::
-:h tz,__
H:: HH r
0 i;;
0 20 0 40 50 60 70 80 90 100 Depth in Inches (LT-1 604) 1.0 To calculate the BAAT (T-6) level drop corresponding to a certain feed volume:
1.1 Read initial BAAT level and determine initial volume from graph.
1.2 Subtract feed volume from initial tank level.
Example: It is desired to feed 530 gallons of boric acid.
A. Initial BAAT level = 82. (From graph, - 7100 gal.)
B. Initial volume - feed volume = 7100 - 530 = 6570 gal.
C. Final level, from graph, corresponding to 6570 gal. = - 74.
ANQ-1 CYCLE 22 COLR CALC-ANO1 -NE-08-00006 Figure 3-A Regulating Rod Insertion Limits for Four-Pump Operation From 0 to 200 +/- 10 EFPD (Figure is referred to by Technical Specification 3.2.1)
-U 0
0
-b C,
a, a,
50 100 150 200 250 300 Rod Index, % Withdrawn 0 20 40 60 80 100 GROUP 7 0 20 40 60 80 100 L I I I I I GROUP 6*
0 20 40 60 80 100 I I GROUP 5*
Technical Specification 3.5.2.5(2) requires that operating rod group overlap be 20% +/- 5%
between two sequential groups, except for physics tests.
ANO-1 Rev. 0
ANO-1 CYCLE 22 COLR CALC-ANOI -NE-08-00006 Figure 3-B Regulating Rod Insertion Limits for Four-Pump Operation From 200 +/- 10 EFPD to EOC (Figure is referred to by Technical Specification 3.2.1) 110.0 0
2.
- 3 C,
Rod Index, % Withdrawn 0 20 40 60 80 100 I I I I I I GROUP 7*
0 20 40 60 80 100 I I I I I I GROUP 6*
0 20 40 60 80 100 I I I I I GROUP 5*
Technical Specification 3.5.2.5(2) requires that operating rod group overlap be 20% +/- 5%
between two sequential groups, except for physics tests.
ANO-1 Rev. 0
ANO-1 CYCLE 22 COLR CALC-ANOI -NE-08-00006 Figure 4-A Regulating Rod Insertion Limits for Three-Pump Operation From 0 to 200 +/- 10 EFPD (Figure is referred to by Technical Specification 3.2.1) 11(
100.0 90.0 OPERATION IN THIS 80.0 REGION IS NOT ALLOWED (105.5, 77) 0 70.0 OPERATION (248.5, 67)
RESTRICTED 60.0 0 SHUTDOWN a, 50.0 MARGIN a,
a, 40.0 (206.5, 43) 30.0 PERMISSIBLE OPERATION 20.0 REGION 10.0 0.0 0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 20 40 60 80 100 I I I I I GROUP 7*
0 20 40 60 80 100 I I I I I GROUP 6*
0 20 40 60 80 100 I I I GROUP 5 Technical Specification 3.5.2.5(2) requires that operating rod group overlap be 20% +/- 5%
between two sequential groups, except for physics tests.
ANO-1 10 Rev. 0
ANO-1 CYCLE 22 COLR CALC-ANOI -N E-08-00006 Figure 4-B Regulating Rod Insertion Limits for Three-Pump Operation From 200 +/- 10 EFPD to EOC (Figure is referred to by Technical Specification 3.2.1) 0 0
-I C,,
0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 20 40 60 80 100 I I I GROUP 7*
0 20 40 60 80 100 L I I GROUP 6*
0 20 40 60 80 100 I I GROUP 5*
Technical Specification 3.5.2.5(2) requires that operating rod group overlap be 20% +/- 5%
between two sequential groups, except for physics tests.
ANO-1 Rev. 0
Attachment B-16: Boron Concentration for 1.5% Shutdown Margin During Emergency Boration 2800 2700 2600 2500 2400 2300 E
0.
- 0. 2200 0
2100 I 2000 1900 1800 1700 1600 1500 1400 1300 1200 1100 0 50 100 150 200 250 300 350 400 450 500 Cycle Lifetime, EFPD CALC-ANO I -NE-08-00007 ANO-I Cycle 22 Plant Data Book Page 53 of 53 Rev. 000
PAM Instrumentation 3.3.15 3.3 INSTRUMENTATION 3.3.15 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.15 The PAM instrumentation for each Function in Table 3.3.15-1 shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTIONS
.--------NOTE----
Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required channel to 30 days with one required channel OPERABLE status.
B. Required Action and B.1 Initiate action to prepare Immediately associated Completion and submit a Special Time of Condition A not Report.
met.
C. One or more Functions C.1 Restore one channel to 7 days with two required channels OPERABLE status.
D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C not Table 3.3.15-1 for the met. channel.
ANO-1 3.3.15-1 Amendment No. 24&,222,232
PAM Instrumentation 3.3.15 CONDITION REQUIRED ACTION COMPLETION TIME E. As required by Required E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action D. I and referenced in Table 3.3.15-1. AND E.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> F. As required by Required F.1 InWate action to prepare Immediately Action D. I and referenced and submit a Special in Table 3.3.15-1. Report.
SURVEILLANCE REQUIREMENTS
These SRs apply to each PAM instrumentation Function in Table 3.3.15-1.
SURVEILLANCE FREQUENCY SR 3.3.15.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized.
SR 3.3.15.2 --------------NOTE----------.-------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION. 18 months ANO-1 3.3.15-2 Amendment No. 24,222
PAM Instrumentation 3.3.15 Table 3.3.15-1 Post Accident Monitoring Instrumentation CONDITIONS REFERENCED FROM FUNCTION REQUIRED CHANNELS REQUIRED ACTION D.1
- 1. Wide Range Neutron Flux 2 E I
- 2. RCS Hot Leg Temperature 2 E I
- 3. RCS Hot Leg Level 2 F I
- 4. RCS Pressure (Wide Range) 2 E I
- 5. Reactor Vessel Water Level 2 F I
- 6. Reactor Building Water Level (Wide Range) 2 E
- 7. Reactor Building Pressure (Wide Range) 2 E I
- 8. Penetration Flow Path Automatic Reactor 2 per penetration flow E Building Isolation Valve Position path
- 9. Reactor Building Area Radiation (High Range) 2 F I
- 10. Deleted I
- 11. Pressunzer Level 2 E I
- 12. a. SG A Water Level Low Range 2 E I
- b. SG B Water Level Low Range 2 E I
- c. SG A Water Level High Range 2 E I
- d. SG UB Water Level High Range 2 E I
- 13. a. SGAPressure 2 E I
- b. SG B Pressure 2 E I
- 14. Condensate Storage Tank Level 2 E I
- 15. Borated Water Storage Tank Level 2 E I
- 16. Core Exit Temperature (CETs per quadrant) 2 E I
- 18. High Pressure Injection Flow 2 E I
- 19. Low Pressure Injection Flow 2 E I
- 20. Reactor Building Spray Flow 2 E I (a) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
(b) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.
ANO-1 3.3.15-3 Amendment No. 24,222