ML082210499
| ML082210499 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 06/05/2008 |
| From: | Constellation Energy Group |
| To: | David Silk Operations Branch I |
| Shared Package | |
| ML073040307 | List:
|
| References | |
| TAC U01633 | |
| Download: ML082210499 (197) | |
Text
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 1 ID: (250230 Points: 1.00 Unit 1 was operating at 100% power MOC when an event occurred. The appropriate AOP was entered, appropriate actions were taken and the reactor was manually tripped. No other actions have been taken. The following conditions now exist:
- 1. Pressurizer pressure is1724 PSIA and lowering
- 2. RCS Tcold is 530°F and steady
- 3. S/G Press is -880 PSIG in both S/Gs
- 4. Pressurizer level is 1 8 5 and rising
- 5. All CHG pumps are operating
- 6. Letdown is isolated
- 7. CNTMT Pressure is 1.7 PSlG and slowly rising Which of the following actions and reasoning are correct for the conditions given?
A. TRIP 11A & 12B RCPs to minimize heat input into the RCS
- 6. TRIP 11B & 12B RCPs to minimize coolant inventory loss from the RCS C. TRIP 11A & 128 RCPs to minimize coolant inventory loss from the RCS D. TRIP 11B & 12A RCPs to minimize heat input into the RCS Answer: C Answer Explanation:
Choice C is correct based on EOP-5 Basis Step 1V.E. The EOP basis says that the RCP trip strategy ensures RCPs are secured for a small break LOCA (the hot leg case being more restrictive), while at the same time allowing two or more RCPs to remain running in the event of a non- LOCA. The incentive for stopping all RCPs during a LOCA is to minimize coolant inventory loss from the RCS. Further, it states that if RCS pressure drops to 1725 PSIA, as a result of the event, then trip RCPs so that either of the following pairs of RCPs remain running: 11A and 12B (21A and 22B) RCPs, or 11B and 12A (21B and 22A) RCPs, and to trip all RCPs if CIS has actuated, Component Cooling flow can not be verified to the RCPs, or if RCS temperature and pressure are less than the minimum pump operating limits.
The conditions given support the conclusion that the trip strategy should be implemented.
Distractor A is incorrect since it states an incorrect basis for the trip two leave two strategy, the strategy is not based on minimizing heat input into the reactor.
Distractor B is incorrect since it lists an incorrect pair of RCPs to trip Distractor D is incorrect since it lists an incorrect Basis for the trip two leave two strategy OPERATIONS Page: 1 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM MultiDle Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to Complete:
Difficulty:
System ID: 50230 User-DefinedID: Q50230 Sross Reference Number:
Topic: Correct Actions , RCP Trip Strategy RO Importance: 4.2 SRO Imnortance: 4.6
.<A Number: 42008AK304 2omments: New Question : Comprehensive/Analysis Reference EOP-5 Basis page 27 & 28 ( Basis for step 1V.E EOP-5 Unit 1 page 11 000008 Pressurizer AK3.04 Knowledge of 4.
Vapor Space Accident I 3 the reasons for the 2 following responses as they apply to the Pressurizer Vapor Space Accident (CFX 41.5,41.10 / 45.6 /
45.13): RCP tripping rea uirements This question gives indications typical of a pressurizer vapor space accident in the stem, along with a RCS pressure which should be interpreted to conclude that a SlAS has or should have occurred. Based on this the RCP trip strategy should be implemented. The candidate must determine which two RCPs should be tripped for the trip strategy, and recall the basis for this action.
OPERATIONS Page: 2 of 146 11June2008
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EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Reference EOP-5 Basis EOP 5 basis block step 1V.L for restoring S/G level LOCA / 3 interrelations between the small break LOCA and the following: (CFR 41.7 / 45.7)
OPERATIONS Page: 4 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 3 ID: Q50232R : Points: 1.00 Unit 1 has been in a Station Blackout for the last hour. Power has just been restored to 11 4KV Bus via high line 5051. The crew is evaluating restoring CCW and restarting RCPs. 11A RCP temperatures were hotter than 11B, 12A and 12B RCP temperatures by 5°F. Given 11A RCP lower seal temperature reached 284°F and Controlled Bleed-off (CBO) temperature of 260°F, which one of the following actions is acceptable?
A. Reestablish CCW flow to all RCPs by manually throttling 1-CC-284, CC CNTMT SUPPLY HDR ISOL valve, and NO RCPs should be restarted.
B. Reestablish CCW flow to all RCPs by manually throttling 1-CC-284, CC CNTMT SUPPLY HDR ISOL valve, and start 11B, 12A, OR 12B RCPs.
C. Reestablish CCW flow to all RCPs by opening 1-CC-3832-CV, CC CNTMT SUPPLY valve, and NO RCPs can be started.
D. Reestablish CCW flow to all RCPs by opening 1-CC-3832-CV, CC CNTMT SUPPLY valve, and start 11B, 12A, OR 128 RCPs.
Answer: A Answer Explanation:
B. is incorrect, since lower seal temperature reached 284"F, CCW must be throttled on reinitiation.
A. is correct per EOP-7.
C. & D. are incorrect because with CBO temperature exceeding 250°F, the seals must be rebuilt.
NO-1-200 recommends that if CCW can be restored it should, even though you may not be able to run RCPs Based on AOP-7C block step VI.E.4 on page 30 OPERATIONS Page: 5 of 146 11 June 2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM auestion Type: Multiple Choice Status: Active 4lways select on test? No 4uthorized for practice? No
'oink: 1.oo rime to Comdete: 3 2.00 System ID:
Jser-Defined ID:
Zross Reference Number:
rooic:
3 0 Importance:
SRO Importance: L.V
.<A Number: 42015AK208 Zomments: Reference AOP7C Block step VI.E.4 page 30 Modified from bank question 201237S01 :
Cornprehensive/Analysis 000015/17 RCP AK2.08 -Knowledge of the Malfunctions / 4 interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following:
ccws OPERATIONS Page: 6 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 4 ID: Q50250R Points: 1.00 Unit 1 is operating at 100% power with the following conditions:
0 11 Charging Pump is running.
13 Charging pump is tagged out for maintenance 0 12 Charging Pump hand switch in NORMAL.
0 Backup Charging Pump Selector Switch is in the 12-13 position.
A loss of offsite power occurs. 1A D/G starts and energizes 11 4KV Bus. 1B D/G trips immediately upon starting. 20 minutes later RCS temperatures are being maintained constant by the ADVs. Which of the following describes the response of the pressurizer? (Assume No other operator actions)
A. Pressurizer level will be slowly rising with charging flow but NO letdown flow.
B. Pressurizer level will be constant with charging and letdown at minimum C. Pressurizer level will be lowering with letdown but NO charging flow.
D. Pressurizer level will be slowly lowering with NO charging flow OR letdown flow.
Answer: D Answer Explanation:
A. Incorrect, Charging pumps will not start and will not operate without operator action under these conditions B. Incorrect, Charging pumps will not start and will not operate without operator action under these conditions.
C. Incorrect, Letdown would have isolated due to the loss of power effects D. Correct, with no operator actions no charging pumps will be running and pzr level will be slowly lowering due to 6 GPM seal leakage.
OPERATIONS Page: 7 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 4 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? NO Points: 1.oo Time to Complete: 3 2.00 Svstem ID: 50530 User-Defined ID: -50R Cross Reference Number:
Given a malfunction of the CVCS system, identify the Topic:
effects on pzr level and determine the actions RO Importance:
SRO Importance: V . .
KA Number: 42022AK103 Comments: New Question : ComprehensivdAnalysis Reference RCS Lesson Plan Alarm Manual for 1C07 F-01 ( Page 6), F-05 (Page014), F-06 (Page16) 000022 Loss of Rx AK1.03 - Knowledge of the 3.
Coolant Makeup / 2 operational implications of the 0 following concepts as they apply to Loss of Reactor Coolant Makeup [(CFR41.8 /
41.10 / 45.3): Relationship between charging flow and PZR level OPERATIONS Page: 8 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 5 ID: Q50610 Points: 1.00 Unit-1 in mode 5 on SDC with the RCS capable of being pressurized. The following conditions exist:
RCS Temperature is 180°F 0 RCS Pressure is 180 psia 0 11 & 12 SGFP are secured and tagged out 0 Main & Auxiliary Feedwater is tagged out to 11 S/G for maintenance A loss of both LPSl pumps occurs. Due to a malfunction, no charging pumps are available. Which of the following is the next course of actions for these conditions?
A. Feed and bleed the RCS using the HPSl pumps and pressurizer PORVs.
- 6. Align condensate to 12 S/G and bleed steam from 12 steam generator.
C. Align a containment spray pump to provide flow through the shutdown cooling heat exchanger.
D. Feed and bleed the RCS using the CS pumps and pressurizer PORVs.
Answer: 6 Answer Explanation:
In this condition, with a loss of both LPSl pumps the preferred order would be to align a Containment spray pump, followed by steaming using available S/Gs, then followed by once through core cooling. However RCS pressure needs to be reduced to less than 170 PSIA to use the CS pumps. Since without Auxiliary Spray (No charging pumps) this is not possible, other means must be used.
A. Feed and bleed the RCS using the HPSl pumps and pressurizer PORVs. Is incorrect since the RCS is capable of being pressurized. This would be the last course of action of the available choices. A candidate might think that the other methods are not available since both S/Gs are not available and main feedwater is not available. A candidate might think that you need both S/Gs available for H/R.
- 6. Align condensate to 12 S/G and bleed steam from the 12 steam generator. Is correct since you have one S/G available for heat removal, and RCS pressure is too high for Containment spray pumps, and without a charging pump auxiliary spray is not available this is the next course of action.
C. Align a containment spray pump to provide flow through the shutdown cooling heat exchanger. Is incorrect since RCS pressure is greater than 170 PSIA, and auxiliary spray is not available to lower pressure (No charging pumps), this course of action is not correct for the conditions given. A candidate might not realize that RCS pressure is too high to use the CS pumps.
OPERATIONS Page: 9 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM D. Feed and bleed the RCS using the CS pumps and pressurizer PORVs. Is Incorrect since the RCS is capable of being pressurized. This would be the last course of action of the available choices. A candidate might think that the other methods are not available since both S/Gs are not available and main feedwater is not available, and may believe that since a CS pump is the first alternative to a LPSl pump, that the CS pumps should be used first for OTCC Status:
No I I Points: 1 .oo 4
4.00 V.L 42025AK202 Basis: SDC
References:
Modified from 202-3B-S-1 : Comprehensive/Analysis References AOP-3B Section IV page 22-25,section VI pages 42 - 57 of Residual Heat Removal System and the following: LPI or OPERATIONS Page: 10 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 6 I D Q15864R Points: 1.00 Unit 1 is operating at 100% power when the following indications are observed:
CC Head Tank Level is 42 inches and steady RCP CCW TEMP HI Alarm CCW PP RM LVL HI Alarm 11/12 CC HX CC TEMP HI Alarm 13 SW PP Bkr VU IMPR Alarm Which of the following is occurring?
A. A loss of CCW due to a Salt Water Leak B. A loss of CCW due to CCW leak C. A loss of CCW due to a CCW Pump loss D. A loss of CCW due to a SW pump trip Answer: A Answer Explanation:
A. Correct - Saltwater leak in the CCW PP room CCHX heatup B. Incorrect, CCW head tank is steady. This distractor is plausible since a person could misconstrue the alarm for CCW pump room level, to indicate that CCW leakage is causing this alarm. They must combine this information with the other information in the stem, such as head tank level, to rule out a leak in the CCW header.
C. Incorrect, Alarms/lnfo given do not support loss of 11 or 12 CCW PPS. This distractor is plausible if a person misconstrues the meaning of the 13 CCW pump bkr alignment alarm to imply that they have lost a CCW pump. The alarm comes in on an improper bkr lineup, but is not the alarm you would get if a pump tripped, you would get D. Incorrect, A SW pump trip is not indicated based on parameters shown OPERATIONS Page: 11 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 6 Info I Question Twe: I MultiDle Choice I Status: 1 Active Always select on test? I No References CCW Lesson Plan Slide 20 000026 Loss AA2.02 -Ability to determine of Component and interpret the following as Cooling Water / 8 they apply to the Loss of Component Cooling Water:
(CFR: 43.5 / 45.13): The cause of possible CCW loss Question 6 Table-Item Links A / 0 Trainina Proaram License Operator Initial Training (LOIT)
System Desianations Electrical 480V Motor Control Centers Electrical 480V Transformers and Buses OPERATIONS Page: 12 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 7 ID: Q50251 Points: 1.00 Given the following conditions:
Unit-1 is at 100% power Pressurizer pressure is 2250 PSlA PZR backup and proportional heater control is in AUTO 1-HS-100 (PZR pressure control) is in the "Y" position 1-HS-lOO-3 (PZR htr cutoff) is in the "X+Y" position 1-PT-1OOY fails high Select the expected PROPORTIONAL HEATER & Pressurizer Spray response. Assume no operator action A. Proportional heaters will operate at approximately 1/3 higher power level than before the failure, and the red lights will be illuminated. Pressurizer Spray will be at 50% of maximum flow.
B. Proportional heaters will operate at approximately 1/3lower power level than before the failure, and the green lights will be illuminated. Pressurizer Spray will be at maximum flow.
C. Proportional heaters will be at minimum power level, and the red lights will be illuminated. Pressurizer Spray will be at maximum flow.
D. Proportional heaters will be at minimum power level, and the green lights will be illuminated. Pressurizer Spray will be at 50% of maximum flow.
Answer: C Answer Explanation:
C . Proportional heaters will be at minimum power level, and the red lights will be illuminated.
Pressurizer Spray will be at maximum value. Is correct, heaters will respond as if RCS pressure were high. At +25 psia above the normal setpoint (2250) Minimum Power signal is applied to the heaters.
A. Proportional heaters will operate at approximately 1/3 higher power level than before the failure. Pressurizer Spray will be at 50% of maximum flow. Is incorrect, power to the heaters lowers on high pressure.
B. Proportional heaters will operate at approximately 1/3 lower power level than before the failure. Pressurizer Spray will be at maximum flow. Is incorrect, pressure difference will be outside the control band, causing minimum power to be sent to the heaters, and Pressurizer Spray to be at maximum.
D. Proportional heaters will be at minimum power level, and the green lights will be illuminated. Pressurizer Spray will be at 50% of maximum flow. Is incorrect, greerdred lights are function of breaker positions, supply breakers remain shut.
Modified from question used on 712002 NRC exam OPERATIONS Page: 13 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 7 Info Question Type: Multiple Choice Status: Active Alwavs select on test?
Authorized for practice? No Points: 1.oo Time to Complete: 4 Difficulty: 2.00 System ID: 50251 User-Defined ID: (250251 Cross Reference Number:
Predict response of Pzr htrs t& Spray o a PPCS Topic:
Malfunction RO Importance: 4.0 SRO Importance: 3.9 KA Number: 42027AA101 Comments: Bank question used on 2002 NRC Exam :
Comprehensive/Analysis Reference RCS- InstrumentationLesson LOI-064A2 Slides 69-77, 110 -116 000027 Pressurizer AAl.01 - Ability to operate Pressure Control System Malfunction / following as they apply to the Pressurizer Pressure Control Malfunctions: PZR heaters, s ra s, and PORVs OPERATIONS Page: 14 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 8 ID:Q50712 Points: 1.00 A loss of load event occurred from 100% power and the reactor did not trip. Which of the following is correct?
A. Diverse Scram System (DSS) will trip the Reactor by opening the MG Set Output breakers.
B. Diverse Scram System (DSS) will trip the Reactor by opening the load contactors from the MG sets.
C. Diverse Scram System (DSS) will trip the Reactor by energizing the RPS Matrix trip relays D. Diverse Scram System (DSS) will trip the Reactor by de-energizingthe RPS Matrix Relays.
Answer: E3 Answer Explanation:
DSS Trip from ESFAS will trip the Reactor on 2 out of 4 High Pressurizer Pressures by opening the load contactors from the MG sets. This is an Anti-ATWS trip feature in the event of an RPS failure.
The MG sets are started and stop from the local control panel in the 45' or 27' switchgear room (either location can control both MG sets). The local panels have controls to close and open the MG set Output breakers as well as the load contactor which sends the output of the MG set to the CEDMs. DSS performs its function by opening the load contactor. The output breakers are manually opened. When DSS actuates it opens the MG set load contactor which removes power from the CEDM and they insert via gravity.
A. Diverse Scram System (DSS) will trip the Reactor by opening the MG Set Output bkr.
Is incorrect, the output breaker is not opened by DSS, only the load contactors.
B. Diverse Scram System (DSS) will trip the Reactor by opening the load contactors from the MG sets. Is correct C. Diverse Scram System (DSS) will trip the Reactor by energizing the RPS Matrix Relays. Is incorrect because DSS does not use the Matrix Relays it uses the MG set load contactor.
D. Diverse Scram System (DSS) will trip the Reactor by de- energizing the RPS Matrix Relays. Is incorrect since DSS does not use the Matrix trip Relays.
OPERATIONS Page: 15 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 029 ATWSll 2.4.31 Knowledge of 4.2 annunciator alarms, indications, or response procedures.
OPERATIONS Page: 16 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 9 ID: Q50252 Points: 1.00 When responding to a SGTR which of the fallowing set of parameters should be used to identify the most affected S/Gin accordance with the EOP-6 basis document?
A. Steam Flow-Feed Flow Mismatch, Main Steam Line RMS B. S/G Level Trends, S/G blowdown RMS C. Steam Flow-Feed Flow Mismatch, S/G blowdown RMS D. S/G Level Trends, post trip bypass feed regulating valve position Answer: A Answer Explanation:
Choice A is correct per EOP-6 Basis Choice B is incorrect since per EOP-6 basis the blowdown RMS cannot be used since it measures common RMS Choice C is incorrect since it also uses S/G Blowdown monitor which is common Choice D is incorrect since post trip the bypass valve goes to 56%
OPERATIONS Page: 17 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM I
Multiple Choice No No 1.oo 3
3.00 50252 I User-Defined ID: 050252 I Topic: Given conditions, parameter values associated with the Steam Flow indicators, determine that a SGTR 3.7 3.6 41038EA113 I Comments: New Question: FundamentaVMemory References : EOP-6 basis for step 1V.J on page 36, EOP-6 Lesson Pian -Slide 43 Main Steam and Blowdown LP Slide 86 , LOI-083-1 000038 Steam Gen. EA1.13 -Ability to operate 3.
Tube Rupture / 3 and monitor the following 7" as they apply to a SGTR:
Steam flow indicators LOlT Learning Objective for ROs and SROs # 2.0 Recall the strategy and basis for the major actions performed in EOP-6, SGTR and what actions are required and safety functions are in jeopardy of being lost.
OPERATIONS Page: 18 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 10 ID: Q50253 Points: 1.00 A manual reactor trip from 100% power was initiated due to an event in progress. Upon entry into the appropriate optimal recovery procedure the following parameters are observed:
11 S/G level is -50 inches and lowering 12 S/G level is -30 inches and lowering CNTMT pressure is 4.25 psig and rising All electrical busses are energized from their normal power supplies.
Pressurizer level is 160 inches and rising slowly Pressurizer pressure is 2250 psia and rising slowly Tavg is 545°F and rising slowly RCS subcooling is 40°F and lowering slowly Based on these conditions, which of the following has occurred?
A. A steam line rupture on 11 S/G has occurred
- 6. A steam line rupture on 12 S/G has occurred C. A condensate header rupture has occurred D. A feed line rupture has occurred Answer: D Answer Explanation:
Lowering levels are caused by loss of feedwater to the S/Gs A steam header rupture would result in an excessive cooldown and Subcooling would be higher than 40°F and increasing. A condensate header rupture would not result in containment pressure increasing.
A. A steam line rupture on 11 S/G has occurred. Is Incorrect. A steam line rupture will have some of the same symptoms as a feedline rupture including, lowering S/G levels due to the increased rate of mass removal, and if in containment it will cause an increase in containment pressure. However a steam line rupture will cause a lowering of RCS temperature, Pzr level and increase in subcooling. Because 11 & 12 S/G have different levels a candidate might choose the one with the lowest or highest level believing that a steam rupture exist on one or the other due to shrink or swell effects.
B. A steam line rupture on 12 S/G has occurred. Is incorrect. A steam line rupture will have some of the same symptoms as a feedline rupture including, lowering S/G levels due to the increased rate of mass removal, and if in containment it will cause an increase in containment pressure. However a steam line rupture will cause a lowering of RCS temperature, Pzr level and increase in subcooling. Because 11 & 12 S/G have different levels a candidate might choose the one with the lowest or highest level believing that a steam rupture exist on one or the other due to shrink or swell effects.
OPERATIONS Page: 19 of 146 11June2008
I I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM C.A condensate header rupture has occurred. Is Incorrect. A condensate header will give some of the same symptoms as a feedwater line rupture outside of containment, including a lowering of S/G levels due to reduction of feedwater available to the S/Gs.
Also due to the lowering of S/G levels RCS temperature will increase causing Pzr level to increase. The temperature increase will cause subcooling to lower. However, a condensate header rupture will not cause an increase in containment pressure.
D. A feed line rupture has occurred - Is correct. A feedwater line rupture will result in a lowering of S/G levels due to reduction of feedwater available to the S/Gs. Also due to the lowering of S/G levels RCS temperature will increase causing Pzr level to increase.
The temperature increase will cause subcooling to lower. A FW rupture in Cntmt will cause containment pressure to rise SRO Importance: 4.3 KA Number: 42054AK101 Comments: New Question : ComprehensivdAnalysis References EOP-4 basis page 9 & 10 AOP3G Basis page 3 000054 (CUEOG) AK1.O1- Knowledge of the Loss of Main operational implications of Feedwater / 4 the following concepts as they apply to Loss of Main Feedwater (MFW): MFW line break depressurizes the S/G(similar to a steam line break)
OPERATIONS Page: 20 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 11 ID: (350254 Points: 1.00 A station blackout has occurred on Unit 1. After the appropriate Optimal Recovery Procedure has been entered, AFW flow is established to 11 & 12 S/Gs and steaming has commenced via the ADVs. The CRS has directed you to verify natural circulation flow. Which of the following groups of parameters/indicationsindicates that natural circulation flow has been established?
A. Tcold 525°F and constant Thot 580°F and constant CET 585°F and constant B. Tcold 535°F and constant Thot 585°F and constant CET 580°F and constant C. Tcold 535°F and lowering Thot 580°F and lowering CET 585°F and constant D. Tcold 535°F and constant Thot 580°F and constant CET 585°F and constant Answer: D Answer Explanation:
D is correct since per EOP-7 Basis CET should be consistent with Th; if Th is constant CET temperature should be constant A is incorrect since the delta T between Th and Tc is higher than 50°F B is incorrect since CET is not consistent with Th it is lower C is incorrect since CET is not consistent with Th OPERATIONS Page: 21 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 11 Info I Question TvDe: I Multide Choice 1I References : EOP-7 Step 1V.K Page 26 EOP-7 Basis for 1V.K.page 30 and 31 circulation cooling I I I OPERATIONS Page: 22 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 12 ID: Q50255 Points: 1.00 Upon entry into EOP-2 for LOOP, #23 ARN pump is started (due to problems with the steam driven AFW pumps) to establish feedwater flow to #21 and #22 S/Gs with the following flow values:
-- #21 S/G 1-FIC-4525A indicates 290 GPM
-- #22 S/G 1-FIC-4535A indicates 300 GPM Based on these flow values which statement below is the correct operator response?
A. Maintain flow values ensuring that RCS cooldown rate is < 100°F per hour.
B. Reduce AFW flow as the common suction flow limit is exceeded.
C. Reduce AFW flow to 300 GPM to protect the 2B DG from overloading.
D. Maintain flow values ensuring as CST level lowers pump cavitation does not occur.
Answer: C Answer Explanation:
A & B are incorrect because EOP-2 basis states to limit flow to 300 gpm when 23 AFW pp powered from DG to protect DG from overloading and 575 gpm all other times D is incorrect because suction line limit of 1200 gpm is not being violated C is correct because flow limited to 300 gpm when on DG per EOP 2 Basis Step IV G.2 OPERATIONS Page: 23 of 146 11June2008
I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM References EOP-2 pages 13 Step IV G.2 EOP-2 page 24- Basis for IV.G.2 000056 Loss AK3.02- Knowledge of the of Off-site Power reasons for the following 16 responses as they apply to the Loss of Offsite Power: Actions contained in EOP for loss of offsite power OPERATIONS Page: 24 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 13 ID: Q50256 Points: 1.00 Unit one is operating at 100% power when the following indications are noted:
Pressurizer pressure is 2250 PSlA Pressurizer level is rising All B/u Htrs are ON AFAS Loss of Power Alarm Actuation SYS loss of Power alarm RAS Actuation Sys tripped alarm SlAS Actuation Sys tripped alarm CSAS Actuation Sys tripped alarm 11, 12, & 13 Charging pumps are operating Letdown is at minimum Based on these indications which of the following is correct?
A. 1YO1 has been lost B. 1YO2 has been lost C. 1YO3 has been lost D. 1YO4 has been lost Answer: A Answer Explanation:
A is correct. Based on the indications listed in AOP -75. All others are not consistent with the indications of AOP 7J OPERATIONS Page: 25 of 146 11June2008
EXAMINATION ANSWER KEY Question 13 Info 1Question Type:
CCNPP 2008 NRC RO EXAM Alwa s select on test?
Authorized for practice?
Points:
RO Importance:
SRO Importance:
KA Number:
Multiple Choice Active No No 42057AA204 Comments: New Question: ComprehensivdAnalysis Reference AOP 7J Section V actions for a loss of 1YO1-Pages 12 -15 1
AA2.04 -Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: ESF system panel alarm annunciators and channel status indicators OPERATIONS Page: 26 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 14 ID:Q50474 Points: 1.00 Unit 1 is operating at 100% power with the following conditions:
0 Outside Air Temperature = 100°F 0 Bay Temperature = 80°F The following are noted 0 TURB BLDG SRW HDR PRESS LO Alarm 0 11 SRW HDR PRESS LO Alarm 0 12 SRW HDR PRESS LO Alarm 0 11 SRW HEAD TK LVL Alarm 0 12 SRW HEAD TK LVL Alarm 0 STEAM LINE DRAIN PANEL Alarm (a) Which of the following caused these indications? (b) What actions are required?
A. (a) A leakhupture of the Service water header in the Turbine Building, (b) Shut the Turbine Building isolation valves, and ensure SRW pressure returns to normal.
- 6. (a) A leakhupture of the Service water header in the Turbine Building, (b) Shut the Turbine building isolation valves, trip the reactor and implement EOP-0, C. (a) A leakhupture of 11 or 12 Service water header in the Auxiliary building, (b)
Shut the Turbine Building isolation valves and ensure that one header pressure restores to normal.
D. (a) A leakhupture of 11 or 12 Service water header in the Auxiliary building, (b)
Shut the Turbine Building isolation valves, trip the reactor implement EOP-0.
Answer: B Answer Explanation:
- 6. A rupture of the Service water system in the Turbine Building as evidenced by HDR and tank alarms, location is turbine building as indicated by 1122 alarms, Trip the reactor and turbine, shut the turbine building isolation valves is the correct answer per AOP-76 basis.
A. Incorrect because it does not say to trip the reactor as is required by AOP7B C. Incorrect because of no reactor trip and steam line drain panel alarm which means the leak is in the turbine building D. Incorrect because leak is in the turbine building not auxiliary building because of 1122 alarms Plausibility analysis OPERATIONS Page: 27 of 146 11 June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM C. A candidate could choose distractor C or D since a leak or rupture of the Service Water header in the Auxiliary building would give a majority of the same symptoms as a IeaMrupture in the turbine building, including All SRW HDR Press lo alarms. This is because the SRW system in the auxiliary building and the turbine building are normally cross connected on the supply and return side. They are isolated from each other either on SlAS or manually by the operator. This means a rupture in a either Turbine building or Auxiliary building header will result in lowering pressure and head tank levels. In order to eliminate the Auxiliary building header as a source, the candidate must recognize that the steam drain panel alarm will come in on high level in any of the turbine building sumps. A high level in the turbine building sump means that the leak is in the turbine building. If a candidate did not recall that that this alarm is associated with a turbine building sump being high, then he will not be able lo eliminate distractor C or D Question 14 Info 3uestion Type:
Status:
4lways select on test?
Ptuthorized for practice? No Points: 1.oo rime to Complete: 3 Difficulty: 4.00 Svstem ID: 50474 User-Defined ID: Q50474 Zross Reference Number:
Identify via indications and alarms the cause of loss of Topic:
Service Water RO Importance:
SRO ImDortance:
.<A Number: 42062AA202 Zomments: New Question : Comprehensive/Analysis References AOP 7B basis page 4 Alarm manual for 1C-13 pages 11, 16,34 and 42.
Alarm manual for 1C 03 page 93 Alarm manual for 1T22, page 8 and 9 000062 Loss AA2.02- Ability to of Nuclear Svc Water determine and interpret the I4 following as they apply to the Loss of Nuclear Service Water: The cause of possible sws loss OPERATIONS Page: 28 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 15 ID: Q50258 Points: 1.00 AOP 7D, "Loss of Instrument Air", requires a reactor trip if IA pressure lowers to 50 PSIG in MODE 1. Which of the following is the reason for this?
A. Protect the Turbine/Generator from damage due to the loss of Service Water cooling B. Protect the RCS from overfeeding effects due to S/G level control concerns C. Protect the RCPs from damage due to CCW Containment CVs failing shut.
D. Protect the SGFP from damage due to miniflow valves failing shut Answer: B Answer Explanation:
Per the AOP 7D basis Step VA, B is the correct answer The FRVs fail as is at 40 psig. This would prevent S/G level control The 50 PSIG trip value was chosen to enable FRV's and TBVs response post trip. The objective is to allow the FRVs to ramp shut, removing the immediate need to trip the SGFPs due to overfeeding effects on the RCS.
A is incorrect since the SRW valves do not fail shut until a complete loss of IA occurs. Since at 50 PSIG we have not completely loss IA the SRW turbine building isolation valves will be open.
C is incorrect since the Containment CCW valves are 50% open at 40 PSIG which supplies sufficient flow. Per AOP7D these valves do not go full shut until you get to 13 psig IA pressure.
The steps for addressing the effects of loss of CCW are in the section for Modes 3,4,5,6.
D is not correct since the S/G miniflow valves fail open on loss of air.
OPERATIONS Page: 29 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 15 Info Question TvDe: MultiDle Choice Status: Active ~~
I Always select on test? No Authorized for oractice? No Points: 1.oo Time to Complete: 4 2.00 System ID: 50258 User-Defined ID: Q50258 Cross Reference Number:
Topic: ldentifv basis for actions for loss of instrument Air svstem RO Importance: 3.7 SRO Importance: 3.9 KA Number: 42065AK308 Comments: New Question : FundamentaVMemory References : AOP7D Basis page 7 of Instrument Air / 8 reasons for the following responses as they apply to the Loss of Instrument Air:
Actions contained in EOP for This question is a basis question for AOP actions. As such it is reasonable to expect RO candidates to know the basis for the trip.
OPERATIONS Page: 30 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 16 ID:Q50690 Points: 1.00 1A DG is paralleled on 11 4 KV bus for routine surveillance testing when a major grid disturbance occurs and the following alarms annunciate:
0 1A POTVOLT FREQ LO 0 GENERATOR LOW FREQUENCY Which of the following system responses or operator actions is correct?
A. 1A Diesel Generator Trips on underfrequency
- 6. Reduce 1A generator load until the alarm clears.
C. 1A Diesel output breaker trips on underfrequency.
D. Raise 1A Diesel Generator speed to clear the alarm.
Answer: C Answer Explanation:
Candidate should recognize from the stem that with the D/G running paralleled to the bus that it is not in reset mode.
1A Diesel Generator Trips on underfrequency -- Is Incorrect, because the D/G output breaker trips but 1A D/G does not trip on low frequency.
Reduce 1A generator load until the alarm clears.-- Is Incorrect since the D/G output breaker will trip. If the D/G was in the reset mode (Emergency) then this would be correct.
1A Diesel output breaker trips on underfrequency.-- Is Correct the D/G breaker will trip Raise 1A Diesel Generator speed to clear the alarm.-- Is Incorrect. since the D/G output breaker will trip.
OPERATIONS Page: 31 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 16 Info Question Type: Multiple Choice Status: Active Alwavs select on test? No Authorized for Dractice? No Points: 1.oo Time to Complete: 3 2.00 System ID: 50690 User-Defined ID: Q50690 Cross Reference Number: FA1RBANKS ToDic:
RO Importance: 4.1 SRO Importance: 4.3 KA Number: 2445 Comments: New Question : ComprehensivdAnalysis Ref. AOP 7M, Alarm manual for 1C18A and 1C62/2C62/2C61 000077 Generator 2.4.45 Ability to prioritize 4.1 Voltage and Electric and interpret the significance Grid Disturbances / 6 of each annunciator or alarm.
Question 16 Table-Item Links A / 0 Trainina Proaram Licensed Operator Requal Training (LOR)
System Desianations Abnormal Operating Procedures (AOPs)
OPERATIONS Page: 32 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 17 ID: Q20628 Points: 1.00 A loss of load transient resulted in a plant trip with PORVs lifting. What would indicate that the quench tank rupture disk has ruptured?
A. "CNTMT NORMAL SUMP LVL HI" alarm actuates B. "QUENCH TK TEMP LVL PRESS" alarm clears C. "QUENCH TK LVL PRESS" alarm actuates D. Indicated rate at which RCS pressure lowers will decrease.
Answer: A Answer Explanation:
A."CNTMT NORMAL SUMP LVL HI" alarm actuates. Is correct, the sump alarm along with quench tank pressure lowering are indications that the rupture disk has ruptured.
6."QUENCH TK TEMP LVL PRESS" alarm clears. Is incorrect, level and temperature would remain high if the rupture disk blows.
C."QUENCH TK LVL PRESS" alarm actuates. Is Incorrect, since the alarm will not re-actuate due to a possible high level situation in the Quench Tank.
D.Rate at which RCS pressure lowers - decreases. Is incorrect, the small range of back pressure associated with the intact or open quench tank would have little effect of P O W relief capacity.
OPERATIONS Page: 33 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 17 Info auestion TvDe: MultiDle Choice Status: Active Always select on test? . .-
Nn Authorized for practice? No Points: 1.oo rime to Complete: 5 Difficulty: 2.00 System ID: 50260 User-Defined ID: Q20628 Cross Reference Number:
Topic: Identify indications of a ruptured quench tank rupture disk.
RO Importance: 3.7 SRO Importance: 3.7 (A Number: 44E02EA11 Zomments: Used 7/2002 NRC exam : ComprehensivdAnaIysis
References:
Alarm Response Manual for 1C10 page 49 and 50 e
Alarm response manual for 1C06 page48 -50 CFR 55.41.3. 55.41.7 EA1.l - Ability to operate and I Trip- or monitor the following as they Stabilization - apply to the (Reactor Trip Recovery/l Recovery) Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Question 17 Table-Item Links System Desianations Reactor Coolant OPERATIONS Page: 34 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 18 ID: (250711 Points: 1.00 Unit 1 is operating at 90% power preparing for Main Turbine Valve testing when the following indications are observed:
0 Tavgnref alarm 0 Hi POWER TRIP RESET DEMAND alarm 0 11 S/G Pressure is 810 PSlA and lowering 0 12 S/G Pressure is 805 PSlA and lowering 0 Tcold is 515°F and lowering 0 Generator Megawatts are lowering Which of the following set of actions is correct?
A. Lower turbine load to restore Tcold to program and implement AOP 7K, Overcooling Event B. Insert CEAs OR Borate the RCS to lower power and implement AOP 7K, Overcooling Event C. Adjust the setpoint on TBV controller to maintain S/G pressure and Tcold and implement AOP 7K D. Implement AOP-7K, Overcooling Event, trip the Reactor and implement EOP-0 Answer: D Answer Explanation:
Lower turbine load to restore Tcold to program and implement AOP 7K, Overcooling Event. Is incorrect, Tcold has lowered to the trip setpoint of 515" per AOP7K the reactor should be tripped. These actions would be correct if Tcold were not at trip criteria for AOP- 7K.
Insert CEAs OR Borate the RCS and implement AOP 7K, Overcooling Event. Is Incorrect, Tcold has lowered to the trip setpoint of 515" per AOP-7K the reactor should be tripped. These actions would be correct if Tcold were not at trip criteria for AOP-7K.
Adjust the setpoint on TBV controller to maintain S/G pressure and Tcold and implement AOP 7K. Is Incorrect, these actions are associated with a loss of load event in which Tcold is lowering and are actions per AOP-7F. AOP-7F shares some symptoms with AOP-7K such as lowering megawatts, Tavgnref alarm and abnormal S/G pressure.
However the lowering Tcold should indicate this is an excess stream demand and not a loss of load.
Implement AOP-7K, Overcooling event, trip the Reactor and implement EOP-0. Is correct, from the symptoms given with Tavgnref alarm and the High Power demand reset alarm, an excess steam demand is causing overcooling of the RCS, increasing reactor power, and since Tcold has reached the trip criteria, a reactor trip should be initiated.
OPERATIONS Page: 35 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM RO Importance: 4.0 SRO Importance: 4.3 KA Number: 244 Comments: New Question: Comprehensive/Analysis qeferences -AOP 7K E itry condition and step IV B actions CUE05 Steam 2.4.4 - Ability to recognize 4.5 Line Rupture - abnormal indications for Excess Heat system operating parameters Transfer that are entry-level conditions for emergency and abnormal operation procedures OPERATIONS Page: 36 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 19 ID: Q50262 Points: 1.00 Due to continuous CEA withdrawal event, a reactor trip has occurred on Unit-1. Immediately after the reactor trip occurs, MCC 104R feeder breaker trips. While implementing EOP-0 the following indications are noted:
2 stuck CEAs.
11 Boric acid pump trips when started Main Turbine and Generator are tripped.
Pressurizer level indication is 40 inches and slowly rising RCS subcooling is 0°F.
RCS pressure is 1600 psia and dropping.
S/G levels are both -40" and stable.
S/G pressures are both 880 psia.
Containment press. is 2.0 psig and rising.
Containment temp is 215°F and rising.
All electrical buses are energized.
So far, no actions (other than the actions for reactivity control) have been taken. Which one of the following groups of safety functions should be reported as "cannot be met"?
A. Reactivity Control and RCS Pressure/lnventory Control.
- 6. RCS Pressure/lnventoryControl and Containment Environment.
C. Core/RCS Heat Removal and RCS Pressure/lnventory Control.
D. Containment Environment and Reactivity Control.
Answer: B Answer Explanation:
Candidate needs to recognize that even with both boric acid pumps not operating ( Loss of MCC 104R loses 12 boric acid pump) there is boric acid flow available via gravity feed valves A is incorrect, Boration is in progress means Reactivity is complete.
B is correct, PRZ level is low, RCS pressure is decreasing and SC is unsat, CNTMT temp. and pressure exceeding limits and increasing.
C is incorrect, HR is complete.
D is incorrect, Boration is in progress means Reactivity is complete.
OPERATIONS Page: 37 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 19 Info Question Type: Multide Choice 1I Status: Active Always select on test? No Authorized for Dractice? No Points: 1.oo Time to Complete:
System ID:
User-Definedm: Q50262 Cross Reference Number: 201-0-84-02 RO Importance:
SRO Importance: 3.6 KA Number: 21 17 Comments: Modifiedfrom Q25093 : Comprehensive/Analysis References EOP-0 pages 6 -17 AOP 71 page 148 and 149 000001 Continuous 2.4.14 Knowledge of general 3.
Rod Withdrawal / 1 guidelines for EOP usage. 8 OPERATIONS Page: 38 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 20 ID: GI50264 Points: 1.00 Given the following initial conditions:
0 Unit 2 is at 100% power.
0 All CEAs are fully withdrawn.
0 CEDS Control System is in OFF 0 Turbine Control is in OPER AUTO 0 21& 22 Digital Feedwater Control Systems (DFWCS) are in AUTO The following events occur:
0 ONE Reg Group 2 CEA drops to the bottom of the core.
0 The reactor does NOT trip.
0 TM/LP Pre-Trip is received on 1 channel 0 The crew enters AOPl B, CEA Malfunction.
Which ONE of the following describes a correct action for the conditions listed above in accordance with AOP 1B?
A. Decrease Turbine load by momentarily depressing the "DOWN" Button to restore Tcold within 2°F of program value.
B. Decrease Turbine load by depressing the Turbine Manual Button, then depress the "DOWN" button to restore Tcold within 2°F of program value.
C. Decrease Turbine load by depressing the Turbine Manual Button, then depress the "DOWN" button to restore Tcold within 5°F of program value.
D. Decrease Turbine load by momentarily depressing the Reference Signal Decrease Button , then depress the "GO" button to restore Tcold within 5°F of program value.
Answer: B Answer Explanation:
A. Decrease Turbine load by momentarily depressing the "DOWN" Button to restore Tcold within 2°F of program value. This is not correct since the turbine is in OPER AUTO mode, pressing the Down button will have NO effect.
6.Decrease Turbine load by depressing the Turbine Manual Button, then depress the "DOWN" button to restore Tcold within 2°F of program value. This is correct since placing the unit in manual and using the down button is acceptable and by procedure 2°F of program is the correct Tcold value C. Decrease Turbine load by depressing the Turbine Manual Button, then depress the "DOWN" button to restore Tcold within 5°F of program value.. This is incorrect since Tcold should be brought within 2°F of program value.
D. Decrease Turbine load by momentarily depressing the Reference Signal Decrease Button , then depress the "GO" button to restore Tcold within 5°F of program value. This is incorrect since Tcold should be brought within 2°F of program value.
OPERATIONS Page: 39 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Ruestlon 20 Info 2uestion Type: Multide Choice Status: Active 4lways select on test? No 4uthorized for practice? No
=mints: 1.oo rime to Complete: 4 2.00 System ID: 50264 Jser-Defined ID: (250264 3oss Reference Number:
Topic: Given a drop rod event determine if turbine actions correct 3 0 Importance: 4.1 SRO Importance: 4.1
<A Number: 42003AA105 Zomments: New Question : ComprehensivdAnalysis Reference 01-43A page 50 -52 AOP 18 page 7 section IV. A.2, OP-3 page 35 000003 AA1.05 -Ability to operate and I 4.
Dropped or monitor the following as they 1 Control Rod / 1 apply to the Dropped Control Rod: Reactor power - turbine Dower OPERATIONS Page: 40 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 21 ID: Q50265 Points: 1.00 Given the following:
Unit 1 is operating at 100% power.
A group 5 CEA (#34)drops to bottom of the core. The appropriate AOP was entered and stabilizing actions are completed. It has been determined that the CEA # 34 is inoperable. Power has been reduced to 69% power.
Which ONE of the following describes the correct actions and reasons based on the conditions above?
A. AS1 is monitored using Excore Nls. Excore NI limits are more conservative than the lncore detectors B. AS1 is monitored using Excore Nls because the DAS must be declared inoperable under these conditions C. AS1 is monitored using lncore Nls because the Excore detectors are not reliable under these conditions D. AS1 is monitored using lncore Nls. lncore NI limits are more conservative than the excore detectors Answer: A Answer Explanation:
Per AOP 1B Basis for step VI.Cl . the Better Axial Shaping Selection System (BASS) is declared inoperable when any CEA is mispositioned by more than 15 inches from its respective group. With BASS operable, AS1 monitoring is performed by BASS using the incore detectors which are able to provide a more accurate picture of the core power distribution. When we must use the excore detectors to monitor ASI, because of their conservatism, if we were close to the AS1 limit prior to the event, a power reduction may be necessary to stay within the limits of the excore detectors.
With BASS inoperable the AOP directs AS1 monitoring using Excore and it states that these limits are more conservative than those associated with the incores which are normally used for AS1 monitoring A. AS1 is monitored using Excore Nls. Excore NI limits are more conservative than the lncore detectors -- is correct per AOP 1B basis, the limits are more conservative since the Excore detectors monitor a limited area of the core.
B. AS1 is monitored using Excore Nls because the DAS must be declared inoperable under these conditions-- Is incorrect per AOP I B basis. DAS is not declared inoperable under theses conditions. No alarms or indications that ( Data Acquisition System). DAS should be declared inoperable were provided. A candidate might mistakenly assume that DAS should be declared inoperable instead of BASS.
C.ASI is monitored using lncore Nls because the Excore detectors are not reliable under these conditions-- Is incorrect. Since BASS must be declared inoperable under the given conditions, The incore detectors are not used to monitor ASI. Under these conditions the lncore detectors are less reliable than the Excores.
OPERATIONS Page: 41 of 146 11 June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM D. AS1 is monitored using lncore Nls. lncore NI limits are more conservative than the excore detectors- Is incorrect. AS1 is monitored using the Excore detectors since with a CEA inserted more than 15" from its group position, BASS is declared inoperable and the excore detectors are used, and the Excore detectors are more conservative than the incores is inoperable RO Importance: 3.9 SRO Importance: 4.2 KA Number: 42005AK306 Comments: New Question : Comprehensive/Analysis AOP-1B Basis for IV. C page 14 1C06 alarm manual page 80-83 Loss of Plant Computer Lesson Plan LOI-202-7HR13slide 28 000005 AK3.06 - Knowledge of the 3.9 Inoperable/Stuck reasons for the following Control Rod / 1 responses as they apply to the Inoperable / Stuck Control Rod: Actions contained in EOP for inoperablehtuck control rod OPERATIONS Page: 42 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 22 ID: Q50266R Points: 1.00 Unit 1 is at 10 % power conducting a S/U after a shutdown to repair a steam leak. Which of the
~
following correctly describes the response of Wide Range Channel A outputs on 1CO5 if its high voltage power supply degrades slightly ( 10%) over 5 minutes?
A. Source Range Counts per second (CPS) ONLY will decrease in proportion to the decrease in voltage B. Source Range Counts per second (CPS) And Wide Range YOpower will remain the same.
C. Wide Range Yopower ONLY will decrease in proportion to the decrease in voltage.
D. Source Range Counts per second (CPS) And Wide Range Yo power will decrease in proportion to the decrease in voltage.
Answer: B Answer Explanation:
The Wide range (WR) Logarithmic nuclear instruments consist of four redundant channels that provide reactor power level indication from the source range ( CPS) to full power (% power). The wide range Log channels use fission chambers that operate in the ionization chamber region of the gas filled detector curve. This provides a constant output over varying voltages. In this region there is no appreciable increase in the number of ion pairs collected as voltage is increased. The operating range of the Wide Range detectors is from .1 CPS to 200 Yopower. The displays on 1CO 5 indicate Source range counts per second (CPS) on the left side of the indicator and ?Lo power on the right side of the indicator. Both the CPS and % power outputs are driven from the same circuitry. The candidate must recall the region of the gas filled detector curve that the Wide Range detectors operate in, and recall that in this region the ion pairs collected is not a function of applied voltage. The candidate must also recall that the Wide range detectors produces two output indicators ( CPS & % Power) and that both of these are driven by the same detector output, even though the circuitry operates under two different principles. The effect of voltage change is the same for both outputs.
A. Source Range Counts per second (CPS) ONLY will decrease in proportion to the decrease in voltage -- is incorrect since these detectors operate ( Both CPS and % power mode) in the ionization chamber region where small changes in voltage will have little effect on the detector out.
B. Source Range Counts per second (CPS) And Wide Range Yo power will remain the same.-- is correct since the detectors operate in the ionization region of the gas filled detector curve both CPS and Yo power indication will not be affected by small changes in high voltage.
C. Wide Range % power ONLY will decrease in proportion to the decrease in voltage.-- Is incorrect since the detectors operate in the ionization region of the gas filled detector curve both CPS and Yopower indication will not be affected by small changes in high voltage.
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EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM D. Source Range Counts per second (CPS) And Wide Range % power will decrease in proportion to the decrease in voltage. -- is incorrect since the detectors operate in the ionization region of the gas filled detector curve both CPS and % power indication will not be affected by small changes in high voltage.
Question22 Info 2uestion Type: Multiple Choice Status: Active 4lways select on test?
4uthorized for Dractice? No Points: 1.oo rime to Complete: 4 3.00 System ID:
Jser-Defined ID: Q50266R 3ross Reference Number:
Topic: Evaluate effects of voltage change on Source Range 7 0 Importance:
3RO Importance: 3.1
.(A Number: 42032AK101 3omments: New Questions : FundamentaVMemory Reference Excore NI Lesson Plan LOI-78A-1-2 slides 24-30 also alarm manual 1C05- page 36 I 000032 Loss I AK1.01- Knowledge of the I 2.
following concepts as they apply to Loss of Source Range Nuclear Instrumentation: Effects of OPERATIONS Page: 44 of 146 11June2008
I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 23 ID: Q50713 Points: 1.00 Which of the following events would require that AOP-GD, Fuel Handling Incident be entered?
A. A piece of wire from the refueling machine drops into the Refueling Pool.
- 6. A new fuel assembly is dropped during insertion in the Spent Fuel Pool.
C. Dropping the dry transfer Cask in the Auxiliary Building on its way to the Spent Fuel Pool.
D. A new fuel assembly is dropped in the New Fuel Storage Area Answer: B Answer Explanation:
A piece of wire from the refueling machine drops into the Refueling Pool.-- Is not correct, per AOPGD basis an item such as a piece of wire falls the AOP is not entered, the wire is retrieved but: the AOP nor the ERPlP are entered for this incident.
A new fuel assembly is dropped during insertion in the Spent Fuel Pool --Is correct, since the dropped fuel assembly has the potential to damage irradiated fuel in the Spent Fuel pool.
Dropping the dry transfer Cask in the Auxiliary Building on its way to the Spent Fuel Pool.-- Is Incorrect, Since the dry Transfer Cask would not have a fuel assembly in it at this time, per the AOP 6D basis , although it is heavy , since it does not have a fuel assembly in it is not classified as a Fuel Handling incident, the response will be per the AOP of any system that is damaged.
A new fuel assembly is dropped in the New Fuel Storage Area- Is Incorrect. Per AOPGD, if a new fuel assembly is damaged in the New Fuel Storage area, and only Alpha being a concern, a Fuel Handling Incident has not occurred.
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EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM IQuestlon 23 Info I Question Type: Multiple Choice Status: Active No I Authorized for Dractice? No 1.oo A
RO Importance:
SRO Importance:
KA Number:
Comments:
Reference Per AOP 6D and AOPGD basis page 5
~
036 Fuel 2.4.4 Ability to recognize Handling Accident abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
Question 23 Table-Item Links A / 0 Traininq Proqram Licensed Operator Requal Training (LOR)
OPERATIONS Page: 46 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 24 ID: Q50268 Points: 1.00 Unit 1 was operating at 100% power when a tube leak in 11 Steam Generator occurred. The appropriate AOP was entered, the reactor was manually tripped and the appropriate EOP has been entered. At 100°/oPower the following indications were noted:
11 S/G Pressure 860 PSlA 12 S/G Pressure 860 PSlA Calculated leakage 60 GPM 11 S/G has been isolated and the RCS cooled down to 500°F. Due to problems with the TBVs the temperature is holding at 500" F.
Per the applicable optimal recovery procedure the pressure differential between RCS and the S/G should be minimized. Approximately what pressure does the RCS need to be depressurized to reduce the leak rate to 25 GPM?
A. 1261 PSlA B. 1252 PSlA C. 921 PSlA D. 901 PSlA Answer: C Answer Explanation:
m l = 60 GPM, m2 = 25 GPM Diff Press1 = 2250 PSI - 860 psi Diff Press 2 = desired RCS pressure - Sat press for 500°F A. 1261 is incorrect. This is answer you get if you do not take the square root of the pressure differential 6.1252 is incorrect. This is answer you get if you do not use square roots and use gauge instead of absolute pressure for S/G pressure ( 875 instead of 860)
C. 921 is correct D. 901 is what you get if you take the square roots of the flow rates and don't add the SG pressure to DP calculation and subtract 15 for absolute pressure.
OPERATIONS Page: 47 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question24 Info Question Type:
Status: Active I Always select on test?
Authorized for practice? No Points: 1.oo rime to Comdete: 3 I Svstem ID:
User-Defined ID:
Sross Reference Number:
Topic: What pressure to depressurize RCS for SGTR RO Importance: Q C J .J SRO Importance: 3.9
.<A Number: 42037AK102 Zomments: New Question : FundamentaWMemory Reference Steam tables and knowledge of relationship between DP and flow AK1.02 - Knowledge of the 3.5 Generator Tube operational implications of Leak I 3 the following concepts as they apply to Steam Generator Tube Leak: Leak rate vs. Dresswe droD OPERATIONS Page: 48 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 25 ID: Q50475 Points: 1.00 A fire in the Unit1 Cable Spreading room has occurred. The SM has determined that a Control Room evacuation is necessary and AOP-SA should be implemented. Which of the following sets of actions are required to be completed within the first 30 minutes of CR Evacuation to prevent damage to plant equipment?
A. Trip the RCPs AND start the OC Diesel Generator
- 6. Start the OC Diesel Generator AND Establish Charging flow C. Trip the RCPs AND Trip MCC-104 load center D. Establish AFW flow AND Establish Charging flow Answer: A Answer Explanation:
Trip the RCPs AND start the OC Diesel Generator - correct per AOPSA basis 1V.C and notes Ill C. 2 Start the OC Diesel Generator AND Establish Charging flow - Not correct, Charging flow not required until 60 minutes Trip the RCPs AND Trip MCC-104 load center Not correct, trip MCC-104 load center does not have a time limit Establish A R N flow AND Trip MCC-104 load center - Not correct , charging flow not required for 60 minutes OPERATIONS Page: 49 of 146 11June2008
I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM I Question 25 Info I Question Type: Multiple Choice Active No No 1.oo 3
4.00 50475 (250475 1 Topic: plant equipment
? I 4.3 42067AA204 New Question : Fundamental/Memory Reference AOPSA Basis page 1 AOPSA page 5 000067 Plant AA2.04 - Ability to determine and Fire On-site / interpret the following as they apply to the Plant Fire on Site: The fire's extent of potential operational WA Match Analysis This question matches the important aspect of this WA since a fire in the cable spreading room at CCNPP could result in damage to the RCPs due to the potential loss of CCW due to the fire effecting the close circuitry for the CCW containment isolation valves. the RCPs are tripped to prevent the damage. Also, the OC Diesel Generator can be potentially damaged since with a loss of the 07 bus the OC D/G will loose its pre-lubrication and can be potentially damaged if its is not started within 30 minutes, or air bottles connected to perform a pneumatic prelubrication. MCC 104 is stripped due to the potential damage to the PORV control circuitry which could result in a loss of RCS inventory.
OPERATIONS Page: 50 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 26 ID: Q50270 Points: 1.00 A loss of offsite power has occurred and 1A Diesel Generator has failed to start, followed by a complete loss of all feedwater. Once Through Core Cooling (OTCC) was initiated due to a loss of all feedwater. Which of the following is true concerning OTCC?
A. OTCC will be successful B. OTCC will be not be successful C. OTCC will only be successful if 13 CHG pp is aligned to 14B 480V Bus D. OTCC will only be successful if 12 HPSl pump is started Answer: B Answer Explanation:
PORV 1-RC-402-ERV will not operate on high RCS pressure, its "manual open" handswitch position will not function to open the PORV-correct, power from MCC-114 is required to allow the PORV to open automatically or manually. With only one PORV available, OTCC will not be successful per attachment 17 even with all HPSl and Chg pumps.
All other answers are incorrect based on the need to have both PORVS operational for OTCC to be successful OPERATIONS Page: 51 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 26 Info Question Type: Multiple Choice Status: Active Alwavs select on test? No Authorized for practice? No Points: 1.oo Time to Complete: 3 Difficulty: 2.00 System ID: 50270 User-Defined ID: Q50270 Cross Reference Number:
ToDic: Effect of PORVs on OTCC with failure RO Importance: 3.4 SRO Importance: X
-. A.
KA Number: 42056AA132 Comments: Modified from Q 20589 used on 2002 NRC Exam :
ComprehensivdAnalysis
References:
EOP Attachments - Attachment 17 AOP-71 pages 18 ,50 000074 EK2.02 - Knowledge of the 3.9 (W/E06&E07) Inad. interrelations between the Core Cooling / 4 PORV and the following :
Inadequate Core Cooling Question 26 Table-Item Links Operations Procedure References (from Nucleis)
EOP EOP-02-1 LOSS OF OFFSITE POWER OPERATIONS Page: 52 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 27 ID: Q50273 Points: 1.00 Unit one is operating at 100% power with the following conditions:
0 13 SRW Pump tagged out for maintenance A large Loss of Coolant Accident (LOCA) occurred on Unit-1, followed by a loss of feedwater. The following conditions are noted:
0 12 SRW PMP - Tripped CNTMT PRESS - 15 PSlG Approximately 30 minutes later it was determined that 12 SRW Pump trip was caused by a blown fuse and has been replaced.
The CRS has directed you to restore SRW.
Which of the following sets of actions is correct based on the conditions listed?
A. Isolate SRW to CACs on 12 SRW SUBSYS and start 12 SRW Pump.
B. Isolate SRW to CACs on 12 SRW SUBSYS and start 12 SRW Pump, then throttle SRW to 13 & 14 CACs C. Place 12 SRW Pump in PTL, do not start 12 SRW until Technical support has provided guidance..
D. Place 12 SRW Pump in PTL, place 1B D/G OUT BKR in PTL, locally trip 1B D/G Answer: A Answer Explanation:
Reference Reference EOP-8 appendix 2, Vital Auxiliaries.
Isolate SRW to CACs on 12 SRW SUBSY and start 12 SRW Pump to support 1B D/G operation - correct per EOP-8 and basis allows restarting 12 SRW pump with SRW to 13
& 14 CAC isolated.
Isolate SRW to CACs on 12 SRW SUBSY and start 12 SRW Pump, then throttle SRW to 13 & 14 CACs - incorrect per the basis restoration via throttling was not evaluated because these are butterfly valves Place 12 SRW Pump in PTL, do not start 12 SRW until Technical support has provided guidance..- incorrect, EOP-8 allows you to attempt a start of 12 SUBSYS in this condition and later contact TSC for continued operation.
Place 12 SRW Pump in PTL, place 1B D/G OUT BKR in PTL, locally trip 1B D/G -
incorrect, these actions are for the case that 12 SRW subsys cannot be restarted.
OPERATIONS Page: 53 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to Complete:
3.00 System ID:
User-Defined ID:
Cross Reference Number:
ToDic: Actions for idle CAC with high CNTM pressure RO Importance: 3.3 SRO Importance: 3.7 KA Number: 44A16AK33 Comments: New Question : ComprehensivdAnalysis Reference EOP-8 appendix 2 Vital Auxiliaries and EOP-5 Basis page 33 CE/A 16 Excess EK3.3 -Knowledge of the reasons 3.2 RCS Leakage for the following responses as they apply to the Excess RCS Leakage (CFR:41.5 / 41.10,45.6,45.13):
Manipulation of controls required to obtain desired operating results during abnormal and emergency situations -
OPERATIONS Page:54 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 28 ID: Q50476 Points: 1.00 Using Provided Reference and given the following conditions on Unit-2 Pressurizer Pressure = 315 PSlA RCS Tcold = 140°F S/GTemperature = 90°F Pressurizer Level = 160 inches 4KV Bus Voltage = 4130 V 13.8 KV Bus Voltage = 14.2 KV Which of the following conditions would prever.. s irting RCP 21A per plar.. operating procedures A. A pressurizer level control malfunction causes pressurizer level to rise to 172 inches and stabilizes.
B. A heatup causes RCS Temperature to rise to 155°F and stabilizes C. A voltage regulator perturbation causes 4 KV Bus voltage to lower to 41 10 Volts and stabilizes.
D. An electrical perturbation causes 13 KV Bus voltage to rise to 14.8 KV and stabilizes.
Answer: A Answer Explanation:
Per OI-1A Section 6.1 .B Starting requirements for an RCP, S/G Temp no more than 60°F below RCS temperature, Pressurizer level less than 170 inches, RCS Pressure and Temperature within the limits of Flgurel7,4KV Buss voltage is greater than 4100 Volts, and 13.8 KV bus voltage less than or equal to 14.8KV A. A pressurizer level control malfunction causes pressurizer level to rise to 172 inches and stabilizes.-- Is Correct since pressurizer level has to be less than 170 inches B. A heatup causes RCS Temperature to rise to 155°F and stabilizes-- Is Incorrect since RCS temperature has to be less than 60°F above S/G temperature 90 + 60 = 150.155°F is 55" less than RCS temperature and still within the limits C. A voltage regulator perturbation causes 4 KV Bus voltage to lower to 41 10 Volts and stabilizes.-- Is incorrect since bus voltage is greater than 4100 volts.
D. An electrical perturbation causes 13 KV Bus voltage to rise to 14.8 KV and stabilizes.--
Is incorrect since the limit is less than or equal to 14,8KV . 14.2 KV is less than 14.8 KV OPERATIONS Page: 55 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 3uestion Type: Multiple Choice Status: Active 4lways select on test? No 4uthorized for practice? No
'oints: 1.oo rime to Complete: 3 Iifficulty: 3.00 System ID: 50476 Jser-Defined ID: Q50476 3oss Reference Number:
rooic: ldentifv the effect on startina an RCP 3 0 Importance:
SRO Importance: 2.9
.<A Number: 34003K614 Zomments: New Question : ComprehensivdAnalysis 01-1A pages 5-13 OI-1A figure 17 NOTE Figure 17 is provided as Reference 003 Reactor K6.14 - Knowledge of the effect 2.6 Coolant Pump of a loss or malfunction on the following will have on the RCPS: Starting requirements OPERATIONS Page: 56 of 146 11June2008
I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 29 ID:Q50275 Points: 1.00 The Letdown Backpressure Control Valves, CV-11OP and CV-11OQ, prevent saturated conditions between the and - to prevent erosion damage. The Letdown Backpressure Control Valves - to decrease pressure and to increase pressure..
A. LD Heat Exchanger, LD Flow control valves, close, open B. LD Heat Exchanger, LD Flow control valves, open, close C. Regenerative heat exchanger, Letdown Flow Control Valves, close, open D. Regenerative heat exchanger, Letdown Flow Control Valves, open, close Answer: 6 Answer Explanation:
Per SD-41 ( CVCS). the purpose of the Backpressure Control Valves are to:
Prevent flashing of hot liquid to steam between the Letdown Flow Control valves and the Letdown Heat Exchanger, Control letdown system pressure at 460 + 40 psi (NOT/NOP)
Prevents erosion damage from the letdown control valve to the inlet of the heat exchanger. Valves open to decrease system pressure and close to increase system pressure A. LD Heat Exchanger, LD Flow control valves, close, open --- Incorrect, valves open to decrease pressure and close to increase pressure.
B. LD Heat Exchanger, LD Flow control valve, open, close -- Correct C. Regenerative heat exchanger, Letdown Flow Control Valves, close, open- Incorrect, Prevent flashing of hot liquid to steam between the Letdown Flow Control valves and the Letdown Heat Exchanger.
D. regenerative heat exchanger, Letdown Flow Control Valves, open, close- Incorrect, Prevent flashing of hot liquid to steam between the Letdown Flow Control valves and the Letdown Heat Exchanger, and valves open to decrease pressure and close to increase pressure OPERATIONS Page: 57 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 29 Info 3uestion TvDe: Multble Choice Status: Active Always select on test? No ~
Authorized for practice? No Points: 1.oo Time to Complete: 3 Difficulty: 2.00 50275 User-Defined ID: (250275 Cross Reference Number:
Topic:
RO ImDortance:
SRO ImDortance:
KA Number: 32004K543 Comments: New Question : FundamentaVMemory References CVCS Lesson Plan LOI-041-1-1(LD) slides 72
-74 Control following concepts as they apply to the CVCS: Saturation subcooling, superheat in steadwater OPERATIONS Page: 58 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 30 ID: (250276 Points: 1.00 Unit 1 is operating at 100% power when the following sequence of events occurs:
Time 0 11 SG Pressure - P1013A, P1013B, P10136, P1013D = 860 PSlA 12 SG Pressure - P1023A, P1023B, P10236, P1023D = 860 PSlA 11 SG Level - LTl14A, LTl14B, LT114C, LT114D = 0 inches 12 SG Level - LT124A, LT124B, LT124C, L'T124D = 0 inches Time 1 Min 11 SG Pressure - P1013A, P1013B, P1013C = 865 PSI , P1013D = 856 PSlA 12 SG Pressure - P1023A, P1023B, P1023C, P1023D = 740 PSlA 11 SG Level - LT114A, LTll4B, L T l l 4 C = -180 inches , LT114D= -120 inches 12 SG Level - LT124A, LTl24B, LT124C = -100 inches, LT124D = -1 10 inches Time 2 Min 30 seconds 11 SG Pressure - P1013A, P1013B, P1013C = 805 PSI , P1013D = 800 PSlA 12 SG Pressure - P1023A, P1023B, P1023C, P1023D = 740 PSlA 11 SG Level - LTll4A, LTll4B, L T l l 4 C = -180 inches , L T l l 4 D = -100 inches 12 SG Level - LTl24A, LTl24B, LTl24C = -100 inches , LTl24D = -180 inches Based on these conditions this of the following is currently true:( Assume NO operator actions)?
A. AFW is supplying 11 S/G ONLY B. AFW is supplying 12 S/G ONLY C. AFW is isolated to 11 & 12 S/Gs D. AFW is supplying 11 & 12 S/G Answer: D Answer Explanation:
AFAS start signal is initiated when EITHER S/G has two of its four wide range level transmitters sensing e -170(170 inches from normal level of 0 inches)
Starts AFW after a 20 second time delay to prevent spurious actuation.
If the S/G level rises above -17 0 prior to the 20 seconds the AFAS start signal will drop out. However, the sensors will have to reset.
Once initiated AFAS start stays locked in.
AFAS BLOCK Secures AFW flow to the SG that has been identified as RUPTURED by shutting four blocking valves (2 motor and 2 turbine) to the S/G Monitors four pressure channels for each SG.
Differential pressures of sensors (one from each SG) provides input to two bistables which trip at 115 psid (TS setpoints are 135 and 130).
When there are 2 of 4 D/P signals from a SG the logic matrix from either SG sends a RUPTURE signal to BOTH actuation logic subsystems (ZA and ZB).
OPERATIONS Page: 59 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM AFAS Block isolates the generator with the lowest pressure.
When the differential pressure condition clears the block valves will Reopen unless the handswitches have been taken to close.
~
M a p l e Choice ~
I Status: Active NO I No 1.oo 5
J.J 32013K502 New Question: ComprehensivdAnalysis Reference AFAS Lesson plan LOI-036B-1-1 Slides 1-14, 37-39,60-65 013 Engineered K5.02 -Knowledge of the 2.
Safety Features operational implications of the 9 following concepts as they apply to the ESFAS: Safety system I logic and reliability I I OPERATIONS Page: 60 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 31 ID: Q50290 Points: 1.00 Which of the following is the most likely reason for this condition?
"SI PPS RECIRC MOV 659 CLOSED RAS BLOCKED" Alarm is ON A. MINI FLOW RETURN TO RWT ISOL ,1- Sl-659 MOV, is shut with an inadvertent RAS present B. MINI FLOW RETURN TO RWT ISOL MOV, 1- Sl-659 MOV is shut with no RAS present C. SI PP RECIR LOCKOUT handswitch, l-HS-3659A, is in ON and RAS present D. MINI FLOW RETURN TO RWT ISOL, 1-SI-659-MOV shut and SI PP RECIR LOCKOUT handswitch, 1-HS-3659A in ON Answer: B Answer Explanation:
Per Alm Manual for 1CO9 window H-55 Different sets of conditions will give the alarm.
B. MINI FLOW RETURN TO RWT ISOL MOV, 1- SI-659 MOV is shut with no RAS present will give this alarm A, C,D have conditions that do not fully satisfy any of the three requirements to get the alarm OPERATIONS Page: 61 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM
[Question 31 Info Question Type: Multiple Choice Status: Active LAlways select on test? No No
. .,. ...,. 1.oo I Time to Complete: 3 3.00 50290 Q50290 Reason for SI PPS RECIRC MOB 659 ALM ON RO Importance: 39 L.O SRO Importance: 2.8 I KA Number: 34005A405 Comments: New Question: FundamentaVMemory Reference Alarm Manual 1C-09 window H-55 page 86 room: Position of RWST recirculation valve (locked when not in use, continuously monitored when in use)
OPERATIONS Page: 62 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP2008 NRC RO EXAM 32 ID: Q50350 Points: 1.00 Using provided references:
U-1 is at 100% power, when one RPS matrix logic channel is declared inoperable. The present time is 1300. When must the unit be in Hot Standby?
A. 1900 today
- 6. 1300 two days from now C. 1300 three days from now D. 1900, two days from now Answer: D Answer Explanation:
Per TS 3.3.3.E, Initially we have 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to restore matrix logic channel, and if this is not met we have 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to be in Mode 3.
OPERATIONS Page: 63 of 146 11June2008 8
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM IQuestion 32 Info IQuestion Type: Multiple Choice Status: Active No 1.oo 4 ,
3.00 System ID: 50350 User-Defined ID: Q50350 for one RPS matrix logic channel inop RO Importance: 3.4 SRO Importance: 3.8 KA Number: 2132 Comments: Modified from Q20207: ComprehensivdAnalysis
References:
Tech Spec 3.3 Provide A copy of Tech Spec 3.3 012 Reactor 2.1.32 Ability to explain and 3.8 Protective System apply system limits and precautions LOlT Learning Objective for ROs and SROs Given the RM-1-101 and/or Technical Specifications, Determine whether the Limiting Condition for RPS Operation is Met, and Identify Required ACTIONS for:
- a. RPS Trip Units (A8.01,02, K1O.O1)
- b. Manual Trip C. Logic Matrices and Relays (A8.03)
- d. TCBs (A8.04, K13.04)
OPERATIONS Page: 64 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 33 ID: Q50613 Points: 1.00 Using Provided
Reference:
A LOCA occurred on Unit one 20 minutes ago. Concurrent with the LOCA, 14 4KV bus tripped due to a fault on the bus. As a result of the LOCA Containment pressure rose to 3.0 PSlG and operators verified the appropriate system actuations. RCS pressure is now 225 PSlA and the CRS has directed you to verify safety injection flow. Which of the following is the correct Safety Injection flow for these conditions, and what if any actions should be taken?
A. 1375 GPM, open 1-SI-653, HPSl HDR XCONN valve B. 1375 GPM, no additional actions required C. 690 GPM, open 1-SI-653, HPSl HDR XCONN valve D. 690 GPM, no additional action required.
Answer: D Answer Explanation:
A. 1375 GPM, open 1-SI-653, HPSl HDR XCONN valve -- Is Incorrect, the flow rate is not correct since 14 4KV bus tripped on fault, 1B D/G will not repower the bus so only one HPSl pump will be running so flow will be 690 GPM.
6.1375 GPM, no additional actions required-- Is incorrect, since 14 4KV bus tripped on fault, 1B D/G will not repower the bus so only one HPSl pump will be running so flow will be 690 GPM.
C. 690 GPM, open 1-SI-653, HPSl HDR XCONN valve -- Is Incorrect, the flow is correct but 1-Sl-653 is only opened if the flow is unacceptable, also since 14 4KV bus tripped on fault, 1B D/G will not repower the bus so no power is available for 12 HPSl pump or 1-SI-653.
D. 690 GPM, No additional action required..-- Is Correct since flow is acceptable Candidate has to use the conditions in the stem ( LOCA, RCS pressure and 14 4KV Bus tripping on fault) and interpret this to determine that 1B D/G will start, but due to the fault will not pick up the bus. This will leave him with one HPSl pump running and no power for 12 HPSl pump and no power for 1-SI-653 ( HPSl HDR cross-connect). He should determine that the flow rate is acceptable by evaluating the graph provided ( attachment 10 & 11) with one pump running and no other actions are required. If he assumes that 1B picks up the bus then he will assume two pumps. He may have the misconception that he needs to open the HPSl HDR X Conn since he only has one pump running, or assumes that he needs it open to start 12 HPSl pump.
OPERATIONS Page: 65 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM IQuestion 33 Info 1I I Question TvDe: MultiDle Choice Status: Active ~ _ _
I Always select on test? No Authorized for practice? No Points: 1.oo 3.2 32006A109 New Question : Comprehensive/Analysis References : EOP-5 Page 10 EOP-5 Basis Reference Provided is Attachment 10 & 11 of EOP attachments design limits) associated with operating the ECCS controls I
OPERATIONS Page: 66 of 146 11June2008
I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 34 ID: Q50336 Points: 1.00 Given the following conditions:
-Unit 2 is at 100% power
-Quench tank pressure is 15 psig
-Quench tank temperature is 120°F
-Pressurizer temperature is 653 degrees F
-Pressurizer pressure is 2250 psia Which ONE of the following identifies the expected reading of the leaking pressurizer safety valve tailpipe temperature instrument, and the expected steam condition?
A. 215 O F; superheated B. 250 O F; wet vapor C. 280 O F; superheated D. 585 O F; wet vapor Answer: B Answer Explanation:
A. Is incorrect, temp wrong and it is not superheated. Will get this if you don't convert to absolute pressure B. Is correct, using the mollier diagram and that this is an isenthalpic process, you add 14.7 to 15 psig =29.7 psia then run that line up to saturation line and you get 250, or use steam tables.
C. Is incorrect, temp is wrong and it is not superheated. Will get this if you use wrong entering enthalpy (extrapolate wrong)
D. Is incorrect, temp is too high. Would get this if he used 2250PSIA and the intersection of 1 118 enthalpy on the mollier diagram OPERATIONS Page: 67 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Status: Active No No Points: 1 .oo Q50336 2.7 35007A103 Comments: Bank Question : Comprehensive/Analysis 007 Al.03 - Ability to predict and/or 2 Pressurizer monitor changes in parameters (to .
Relief/Quench prevent exceeding design limits) 6 Tank associated with operating the PRTS controls including: Monitoring quench tank temperature OPERATIONS Page: 68 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 35 ID: (250337 Points: 1.00 Unit 1 is in Mode 1 at 100% power when a loss of Component Cooling occurs. Which of the following resulting condition(s) from this event alone would require a Reactor trip?
A. CCW Head tank level lowers to and stabilizes at 5 inches.
B. RCP upper thrust bearing temperature rises to and stabilizes at 196°F.
C. RCP lower seal cavity temperature rises to and stabilizes at 205°F D. CCW Header pressure lowers to and stabilizes at 60 PSIG.
Answer: B Answer Explanation:
A Incorrect, AOP7C would address this but it is not trip criteria B. Correct, this satisfies trip criteria per AOP 7C- Step A C. Incorrect, This is not a trip criteria, 200°F is a start criteria for the RCP D. Incorrect this is not a trip criteria, it is only 5 psig lower than the 65 # exit criteria for AOP 7C.
Status:
No No 1.oo 3
2.00 I Topic:
RO Importance:
SRO Importance:
KA Number:
Comments:
008 K3.01- Knowledge of the effect 3.
Component that a loss or malfunction of the 4 Cooling Water CCWS will have on the following:
Loads cooled by CCWS I
OPERATIONS Page: 69 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 36 ID: Q50338R Points: 1.00 Unit 2 is operating at 100% power with the L.etdown Hx Temperature Controller, 2-TIC-223, in MANUAL. While operating 2- TIC-223 the Reactor Operator adjusts the output of 2- TIC-223 to 100%. Which of the following could result from this action?
A. Reactor Power decreases i3nd Radiation Monitor isolation, 1-CV-521-CV shuts
- 9. U D HX CCW Diff Press High Alarm C. REGEN HX OUT TEMP HIGH Alarm D. RCS boron concentration decreases and reactor power increases Answer: A Answer Explanation:
Raising the output of the TIC to 100% will shut the TCV therefore causing temperature to rise which will cause boron to be sloughed which is a negative effect. However, temperature out of the letdown Ht Exchanger will rise and at 140°F 1-CV-521-CV will shut.
U D HX CCW Diff Press High Alarm will not happen because this would mean max flow (
Valve wide open)
REGEN HX OUT TEMP HIGH Alarm. Is not affected by CCW flow RCS boron concentration decreases and reactor power increases is incorrect (This would mean increased flow, lower temperature)
OPERATIONS Page: 70 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 36 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? NO Points: 1.oo Time to Comdete: 3 3.00 System ID:
User-Defined ID: Q50338R Cross Reference Number:
Topic: Adjusting TIC-223 effects on plant RO Importance: 3.0 SRO Importance: 2.9 KA Number: 38008A409 Comments: New Question : ComprehensivdAnalysis CVCS Lesson Plan Lot-041-1-1 (LD)- Slides 63-68 1C07 Alarm manual page 14
, 008 Component A4.09 - Ability to manually 3.
Cooling Water operate and/or monitor in the 0" control room: CCW temperature control valve OPERATIONS Page: 71 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 37 ID: Q50339 Points: 1.00 Unit 1 was operating at 100% Power MOC, with the following conditions:
Pressurizer pressure is 2250 PSlA PZR backup and proportional heater control is in AUTO 1-HS-100 (PZR pressure control) in the "X" position A loss of 1YO9 has occurred. What affect will this have on the operation of the Pressurizer Pressure Control System?
A. Pressurizer Spray Valve will not operate in MANUAL OR AUTO B. Pressurizer Spray Valve will not operate in AUTO, but will operate in MANUAL C. All Pressurizer Heaters will be OFF D. Pressurizer Heaters will not energize on high pressurizer level (+13")
Answer: A Answer Explanation:
A. Correct per AOP71 section of loss of 1 YO9 page 57 B. Incorrect answer not supported by AOP 71 C. Incorrect Pzr htrs are not powered by 1YO9 , however PZR htrs will be off on a loss of 1Y10 due to the low level cutoff coming from 1Y10.
D. Incorrect Answer not supported by AOP 71 OPERATIONS Page: 72 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 010 Pressurizer K2.02 - Knowledge of bus 2.
Pressure Control power supplies to the following: 5 Controller for PZR spray valve OPERATIONS Page: 73 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 38 ID: Q50340 Points: 1.00 U-1 is in Mode 3 performing a reactor startup per OP-2, "Plant Startup from Hot Standby to Minimum Load". The following initial conditions exist:
0 Digital Feedwater Control System (DFWCS) is in AUTO 0 Turbine Bypass Valves (TBV) in AUTO with setpoint of 900 PSI 0 ALL Reactor Trip Circuit Breakers (TCBs) are shut 0 Crew is preparing to withdraw shutdown CEAs Due to an electrical malfunction developing on the CEDM bus the following indications are noted:
0 Four reactor trip bus UV relay indicator lights energize on the Trip Status Panel above 1C15 What impact will this have on DFWCS?
A. DFWCS will receive a signal from ESFAS AL ONLY to shift to Post Reactor Trip Mode B. DFWCS will receive a signal from ESFAS BL ONLY to shift to Post Reactor Trip Mode C. DFWCS will receive a signal from ESFAS AL & BL to shift to Post Reactor Trip Mode D. DFWCS will remain in its present mode and continue to feed the S/Gs Answer: B Answer Explanation:
B. Correct - DFWCS will receive a signal from ESFAS BL ONLY to shift to Post Reactor Trip Mode is the correct answer.
A. Incorrect as the signal to shift to post trip comes from ESFAS BL ONLY C. Incorrect the signal to shift to post trip mode comes from ESFAS BL only D. Incorrect- with UV on Rx trip bus DFWCS will shift to post trip mode.
The signal to shift the DFWCS to the post trip mode uses only ESFAS B logic channel.
This is activated when it receives 2 of four reactor trip bus undervoltage inputs. A candidate could have the misconception that this signal comes from A logic channel or both A and B logic channels. He could also have the misconception that you need to have a turbine trip and Rx trip bus undervoltage to get DFWCS into the post trip mode.
This is plausible since this was the configuration that existed some years ago (Turbine trip was an input to get DFWCS into post trip mode), but is no longer the case.
OPERATIONS Page: 74 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM IQuestion 38 Info I Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to ComDlete: 3 012 Reactor K3.04 - Knowledge of the effect 3.8 Protection that a loss or malfunction of the
- RPS will have on the following:
ESFAS OPERATIONS Page: 75 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 39 ID: Q50341R Points: 1.00 U-1 is in Mode 3 returning from a maintenance outage. ESFAS sensor cabinet ZF (1C93) has been shutdown for cleaning and inspection. (All other channels are energized and signals are not bypassed).
The SIAS Containment pressure transmitter (1PT5313B) fails high, which of the following describes a correct response for this condition?
A. 1-SRW-l597-CV, 11 SFP HX SRW Inlet valve closes.
- 6. 1-SI-4151-CV, Containment Spray Valve Opens.
C. 11 Penetration Room Exhaust Fans Starts.
D. 13 Containment Iodine Removal Unit Starts.
Answer: D Answer Explanation:
Containment Pressure transmitter 1PT5313 A - 6, supply a signal for SlAS on Hi containment pressure., when 1PT5313B fails with sensor cabinet ZF de-energized you have met the condition to actuate SIS on Containment pressure . Equipment actuated by a SIAS A or SlAS B will go to its actuated condition. You have not met the conditions for CIS, or CSAS in order to have equipment associated with these actuations to reposition.
Since 1PT 5313 6 does not feed CIS or CSAS. Those ESFAS signals will not actuate.
A. is incorrect SRW-1597-CV, 11 SFP HX SRW Inlet valve closes. This will happen on a CSAS B. is incorrect SI-4151-CV, Containment Spray Valve Opens.. This will happen on a CSAS C. is incorrect - 11 Pen Room Exhaust Fans Starts. This will happen on a CIS D. is Correct - 13 Containment Iodine Removal Unit Starts. This will happen on a SlAS OPERATIONS Page: 76 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM hestion 39 Info Nestion Type: Multiple Choice Status: Active Jlways select on test? No Juthorized for practice? No
'0 ints :
rime to Comdete:
Iifficulty: 3.vv System ID: 50591 Jser-Defined ID: (250341 R 2ross Reference Number: Q20783 When a high containment pressure signal is generated, ropic:
which components receive this signal?
3 0 Importance: 97 3RO Importance: 9 1 J . I (A Number: 32013K601 Zomments: Modified from Bank Q20783 : Fundamental/Memory ESFAS Subsystems Receiving High Containment Pressure Signal
References:
ESFAS Lesson Plan LOI-048-1-2 slides 17
,19,47,49, 50, 54 -56, 65-66, 69, 73-74, 76, 104-107 LD058 (6158ASH0001), 61058 013 Engineered K6.01- Knowledge of the effect of Safety Features a loss or malfunction on the following will have on the ESFAS:
(CFR: 41.7 / 45.5 Sensors and
- detectors OPERATIONS Page: 77 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 40 ID: (250342 Points: 1.00 A SIAS has occurred on Unit 1. Which of the following is a correct statement for CAC operation?
A. The CACs can be started in Fast Speed at 1CO9 AND at the load contactor panel.
B. The CACs can be shifted to Fast Speed at the load contactor panel ONLY C. The CACs can be stopped from the load contactor panel ONLY D. The CACs can be stopped at 1CO9 and at the load contactor panel.
Answer: 6 Answer Explanation:
The CACs can be shifted to Fast Speed at the load contactor panel ONLY A is incorrect, with a SIS present CACS can not be started in fast speed from the CR per LD 76 sheet 11 B is correct since CACs can be shifted to Fast Speed at the load contactor panel ONLY C is incorrect, SlAS signal seals in per LD 76 sheet 11 D is incorrect SIAS signal seals in per LD 76 Sheet 11 All of the answers require the candidate to be familiar with the logic sheets and/or control drawings for the CACS. If a candidate does not know the logic he could have the misconception that the CACs can be shifted to fast or stopped with a SlAS present since the H/S at 1CO9 have a pull to lock feature to start them in slow. He could confuse this with the ability to pull to lock and stop the CACs. Some pumps ( CCW, SW, SRW) can be pulled to lock and will not start on SIAS. This is not true for the CACs. CACs are manually started in fast speed from 1CO9 when containment environment is degraded.
OPERATIONS Page: 78 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM ECCS Lesson Plan Containment Spray & Cooling LOI-052-3-3 slides 56-58, 60-61, 65, LD 76 Sheet 11 AOP-SA page 53 022 Containment Cooling physical connections and/or cause effect relationships between the CCS and the following systems: SEChemote monitorine svstems OPERATIONS Page: 79 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 41 ID: Q50343 Points: 1.00 Which of the following combinations of conditions/equipment would maintain the pressure and temperature of the containment within design parameters following a Design Basis LOCA ( With a Loss of offsite power)?
A. 11 CAC is OOS, and CS PMP 11 fails to start and CS PMP 12 seizes and trips when started.
B. 13 CAC is 00s and 1A DG fails to start on SlAS C. 11 CS pump is 00s , 13 SRW pump is 00s. 11 SRW pump fails to start on SlAS and 14A 480V breaker trips D. 13 CAC is OOS, and CS PMP 12 fails to start and 1A D/G trips when started on SlAS Answer: A Answer Explanation:
11 CAC is OOS, and CS pump A fails to start and CS pump B trips when started - this would leave you with three CACs which will maintain CNTMT temperature/pressure below design. Correct 13 CAC is 00s and 1A DG fails to start on SlAS this would leave one CAC and one spray pump which would not meet minimum DBA. Incorrect 13 CAC is OOS, and CS pump B fails to start and 1A D/G trips when started on SlAS -
this would leave No Spray pumps and 1 CAC which would not meet requirements.
Incorrect 11 CS is 00s , 13 SRW pump is 00s. 11 SRW pump fails to start on SlAS and 14A 480V breaker trips - this would leave 1 Spray pump and one CAC which is not above minimum. Incorrect OPERATIONS Page: 80 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 41 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points:
Time to Complete: 3 Difficulty: 3.00 System ID: 50343 User-Defined ID: Q50343 Cross Reference Number:
Which combination of CACs & CS Pumps will maintain Topic:
Containment Temp & Press?
RO Importance: 3.7 SRO ImPortance: A..1.
KA Number: 35026K404 Comments: Modified from Q 20397 : Comprehensive/Analysis References ECCS Lesson Pan LOl-052-3-3 Slides 73-77 Basis: Ensure Design Pressure and Temperature of Containment / UFSAR Chapter 14.20 026 K4.04 - Knowledge of CSS design Contain feature(s) and/or interloek(s) which ment provide for the following: Reduction of Spray temperature and pressure in containment after a LOCA by condensing steam, to reduce radiological hazard, and protect equipment from corrosion damage (spray)
OPERATIONS Page: 81 of 146 11June2008
, I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 42 ID: Q50345 Points: 1.00 Following a LOCA, instrument air has been isolated to the containment due to a CIS. What effect does this have on the continued use of the Containment Spray System to cool the containment?
(Assume no operator actions)
A. Both spray flowpaths are no longer available as the spray header CVs have failed closed on a loss of air.
B. No effect since the Containment Spray CVs would be supplied by Saltwater Air Compressors.
C. No effect, each spray header CV fails open on a loss of air to ensure the flowpath remains available.
D. No effect, the spray header CVs have keyswitches that override valves open to ensure the flowpath remains available.
Answer: C Answer Explanation:
A is incorrect: Containment Spray CVs fail open on loss of IA B is incorrect: The SWAC will not supply the spray CVs unless action is taken to override IA-2085 open C. Correct, since each spray CV fails open D, Incorrect, there is no override switch for the spray CVs , there is an override switch for the IA containment isolation valve The candidate must know that the CVs fail open on loss of air. Also, the candidate needs to know that there is no override switch for theses CVs. A misconception is plausible about this since there are override switches for various components that receive a CIS including the IA - containment isolation valve. However action is required by the operator to override the containment air isolation valve open. The SWAC will start on SlAS and supply air to components in containment, but only if the IA containment isolation vale is overridden open.
OPERATIONS Page: 82 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 42 Info Question TvDe: MultiDle Choice Status: Active Always select on test? No Authorized for Dractice? No Points: 1.oo Time to Complete: 3 Difficulty:
System ID: 50345 User-Defined ID: Q50345 Cross Reference Number: Q20409 What effect does lost of Air have on Containment Spray Topic:
CVS?
RO ImDortance: 2.7 SRO Importance: 2.9 KA Number: 35026K202 Comments: Bank Question : FundamentaUMemory
References:
DWg 60-617-B Sht. 33 AOP 7D attachment 1, 2, 3 Lesson Plan LOl-052-3-3 Slides 16, 21 , 24 Basis: Instrument Air Isolated to Containment Due to CIA 026 Containment K2.02 -Knowledge of bus 2 Spray power supplies to the following: MOVs 7 WA Match analysis. The WA addresses knowledge of bus power supplies for Containment Spray MOVs. Since CCNPP has AOVs rather than MOVs for containment spray, the important aspect of the WA as it applies to CCNPP is to have knowledge of the effects of the air supply to these AOVs. Based on CCNPP configuration this question addresses the knowledge of the air supply to the AOVs, and thereby the intent of the WA.
OPERATIONS Page: 83 of 146 11June2008
001 : W ! O d LWSO :ai w WVX3 O t l 3 t l N 8002 d d N 3 3 M Y W M S N V NOIlVNIWVX3
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 43 Info Duestion TvDe: MultiDle Choice Status: Active Always select on test? No 4uthorized for Dractice? No Points: 1.oo rime to Complete: 2 1.oo 50347 User-Defined ID: Q50347 Cross Reference Number: Q20605 Topic:
RO Importance:
SRO Importance: 2.7 U Number: 34039A109 Comments: Bank Question Used 7/2002 NRC exam :
FundamentaVMemory References Lesson Plan LOI-077-1-2Slides 66-71 OM-98 sh2.
039 Main A1.09 - Ability to predict andor 2.5 and Reheat monitor changes in parameters
- Steam (to prevent exceeding design limits) associated with operating the MRSS controls including:
Main steam line radiation monitors OPERATIONS Page: 85 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 44 ID: Q50348 Points: 1.00 Unit 2 is starting up from a refueling outage and has stabilized power at 440 MWE to perform an NI calibration prior to raising power to full power. The following indications are noted:
TBV 2-MS-3946-CV fails open Electrical MWs lowering Tave-Tref Alarm Which of the following actions AND reasons are correct for the conditions given:
A. Raise turbine load then insert CEAS, to stop reactor power rise.
B. Lower turbine load then withdraw CEAS, to restore Tcold to program.
C. Raise turbine load then insert CEAs, to maintain power < 5%
D. Lower turbine load then withdraw CEAS, to maintain reactor power Answer: D Answer Explanation:
A. Incorrect, (+ MTC) exists and TBV failing open would lower power B. Incorrect, withdrawing CEAS to control temperature is not sanctioned by AOP 7K C. Incorrect, Per AOP7K, caution, for + MTC overcooling event a positive reactivity addition will be needed to maintain power after temperature has been stabilized with turbine D. Correct per AOP 7K The conditions given in the stem state that this is a startup after a refueling outage. With the current fuel loading that CCNPP uses and has been using for several years, a positive MTC will exist up to about 70% power (620 MWE). With a positive MTC a TBV failing open will cause Tcold to lower which will cause power to lower. Per our procedures once the temperature mismatch has been corrected by lowering turbine load, a positive reactivity addition will be necessary to stabilize power.
It is plausible for a candidate to chose to maintain power c 5% since a step in AOP7K says that if power lowers to less than 5% then maintain it there. Since we are at 49%
power, a failed open TBV should not cause power to go to 5%. It is also plausible to chose withdraw CEAs to restore Tcold, since Tcold will be lowering and if a candidate does not understand that he must match secondary power to primary power to stabilize temperature then he will get this incorrect. If a candidate does not recognize that for the conditions given he has a positive MTC, then he might choose A or B.
OPERATIONS Page: 86 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 44 Info Question Type:
Status: Active Alwavs select on test? No Authorized for practice? No Points: 1.oo Time to Complete: 4 Difficulty: 2.00 System ID: 50348 User-Defined ID: Q50348 Cross Reference Number:
Given conditions , parameter values and/or indications ,
Topic:
determine the effects on plant operations a RO Importance: 4.1 SRO Importance: .
A!?.-
KA Number: 2143 Comments: New Question : ComprehensivdAnalysis Reference AOP-7K page 5,9, 10 039 Main 2.1.43 - Ability to use procedure to and Reheat determine the effects on reactivity Steam of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.
OPERATIONS Page: 87 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM I Question 45 Info Multiple Choice Active No No 1.oo 4
2.00 50349 Q50349 CRO-103-2-4-12 Evaluate conditions for 1 SGFP oDeration RO Importance: 3.0 n n SRO Importance: J.J KA Number: 34059A207 Comments: Modified from Q24634: ComprehensivdAnalysis
References:
01-12A page 10 & 11, AOP3G page 9 Basis: Continued Power Operation above 440 MWe per AOP-3G A2.07 - Ability to (a) predict the Feedwater 1 impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Tripping of MFW pump turbine WA Match Analysis- The important aspects of this WA as applies to CCNPP are to be able to predict those operations and/ or malfunctions that will result in tripping a Main Feedwater pump and determine the actions to take to prevent this from occurring. In the conditions given in the stem of this question there are two items which if not corrected could directly lead to trip of the Main Feedwater pump, and the other two could indirectly lead to a trip. They are speed and suction pressure, While neither is at the trip setpoint as of yet, if they continue to degrade the impact of continued operation in this condition could be a trip of the feed pump. In this case for these conditions the candidate needs to determine that the speed is above the maximum for the conditions and needs to adjust the speed to less than or equal to 5100 RPMs to preclude the possibility of a feed pump tripping.
OPERATIONS Page: 89 of 146 llJune2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 46 ID: (250351 Points: 1.00 When a SGFP is in Direct Governor Valve control what does the Operator Control Station (OCS) demand indicate?
A. Displays % output demand to the HP & LP governor valves of that SGFP.
B. Displays speed demand signal from DFWCS to the Lovejoy system for that SGFP C. Displays % output demand to the HP governor valve ONLY of that SGFP.
D. Displays speed demand signal from Lovejoy for that SGFP.
Answer: A Answer Explanation:
A. - Correct, Per 01-12A Sec 6.1 8 B. - Incorrect, These indications are seen when in HIC control C.- Incorrect, Much of SGFP operating range utilizes only LP valves so both valves must receive control signal D.- Incorrect, Lovejoy does not send speed demand signal to DFWCS OPERATIONS Page: 90 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM IQuestion 46 Info Question Type: Multiple Choice Status: Active Alwavs select on test? No Authorized for practice? I No Points: 11.00 Time to Complete: 14 Difficultv: 12.00 valve control RO Importance: 2.5 SRO Importance: 2.6 KA Number: 34059A304 Comments: Modified from Q24612: FundamentaVMernory
References:
LOI-045E-1-1 Slide 131 059 Main A3.04 - Ability to monitor 2.
Feedwater automatic operation of the MFW, 5 including: Turbine driven feed
- Pump OPERATIONS Page: 91 of 146 11June2008
I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 47 ID: Q50354 Points: 1.00 Unit 2 was operating at 100% power when a Condensate Header rupture occurs which requires a reactor trip. After the trip S/G levels are being controlled with #22 AFW pump. Subsequently a loss of Instrument Air occurs. What is the effect on the plant over the next hour ? Assume no operator action taken A. RCS temperatures will lower due to the operating AFW pump speed rising to the maximum governor setting AND the FCV going to full open B. RCS temperatures will rise due to the operating AFW pump tripping on overspeed.
C. RCS temperatures will rise due to S/G levels lowering due ONLY to AFW pump Steam Supply valves shutting.
D. RCS temperatures will lower due to S/G levels rising due ONLY to the AFW flow control valves failing full open.
Answer: A Answer Explanation:
A. Correct, Per 01-32A and OM 801. AFW pump goes to maximum speed and the flow control valves fail open.
B. Incorrect, Per 01-32A Sect. 5.0 the AFW pump does not overspeed on loss of air, only goes to maximum speed.
C. Incorrect, ANV steam driven pumps steam valves fail open , not shut per OM 801 Sht 1
D. Incorrect, While the control valves do fail open, they are not the ONLY reason for the overfeed, the AFW pump at maximum speed is also a factor.
OPERATIONS Page: 92 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM MultiDle Choice I Status: Active No No 1.oo 3
2.00 50354 I Topic:
Evaluate the effect of a loss of operating air on the AFW system components.
4.4 4.6 34061K301 Comments: New Question : ComprehensivdAnalysis Reference AFW Lesson Plan LOI-036A-1 LP slides 117 ,
118,119,125 AOP-SA page 21 & 22,23, OM 801 Sheet 1 and 2,0132A Section 5.0 061 K3.01 -Knowledge of the effect 4 Auxiliary/Emergen that a loss or malfunction of the .
cy Feedwater AFW will have on the following: 4 RCS OPERATIONS Page: 93 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 48 ID: Q50614 Points: 1.00 Unit one is in Mode 6 performing a fuel shuffle in accordance with the refueling procedures. The refueling machine is being moved to the upender to retrieve a fuel assembly when MCC-105 is lost. Which of the following is a correct action for the refueling machine?
A. The bridge and trolley brakes are automatically released, then move the bridge
& trolley by handwheel to the upender.
B. The bridge and trolley brakes must be manually released, then move the bridge
&trolley by hand to the South pool.
C. The hoist brakes must be manually released, then manually lower the hoist by handwheel to pick up the fuel assembly.
D. The hoist brakes are released automatically, then manually lower the hoist by handwheel to pick up the fuel assembly.
Answer: B Answer Explanation:
A. The bridge and trolley brakes are automatically released, then move the bridge &
trolley by handwheel to the upender.-- Incorrect, the brakes must be manually released, also manual operation of the hoist should only be used to place a fuel assembly in a safe location.
B. The bridge and trolley brakes must be manually released, then move the bridge &
trolley by handwheel to the South pool. --- Correct per 01 25C, the brakes must be manually released, then the bridge and trolley can be moved by handwheel C. The hoist brakes must be manually released, then manually lower the hoist by handwheel to pick up the fuel assembly.- Incorrect, manual operation of the hoist should only be used to place a fuel assembly in a safe location D. The hoist brakes are released automatically, then manually lower the hoist by handwheel to pick up the fuel assembly.-- Incorrect, the brakes must be manually released, manual operation of the hoist should only be used to place a fuel assembly in a safe location OPERATIONS Page: 94 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM I Question 48 Info Question Type: Multiple Choice I
Status: Active Alwavs select on test? No 062 AC Electrical 2.1.41 -Knowledge of the 2.8 Distribution refueling process OPERATIONS Page: 95 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 49 ID: Q50358 Points: 1.00 Unit-1 has experienced a Loss of Offsite Power. Operators have implemented the appropriate Emergency Operating procedure. During plant stabilization a SlAS actuates due to low pressurizer pressure, and currently D/G 1B is loaded to 3630 KW.
Which of the following is correct for these conditions? (Assume NO operator actions)
A. GEN FLD LOSS REVERSE PWR UNDER FREQ alarms AND the D/G output breaker trips.
B. 1B DG POT VOLT FREQ LO alarms due to low frequency C. 1B DG POT VOLT FREQ LO alarms due to low voltage D. 1B Diesel Generator ENGINE EXCTR SHUTDOWN alarms due to exciter shutdown.
Answer: B Answer Explanation:
B is correct per EOP-2 step IV.P.1 and basis A, C, D not consistent with EOP 2 Basis, however action must be taken immediately to bring load less than 3600 KW A. GEN FLD LOSS REVERSE PWR UNDER FREQ alarms AND the D/G output breaker trips.. -- Incorrect, the output breaker will not trip on underfrequency because the D/G is running in emergency mode.
B. 1B DG POT VOLT FREQ LO Alarm due to low frequency- Correct, as the D/G is overloaded and running in the isynchronous mode, it will try to maintain load up to the limits of its fuel rack setting, then it will lower in speed and the low frequency alarm will come in.
C. 1B DG POT VOLT FREQ LO Alarm due to low voltage-- Is Incorrect, the D/G voltage regulator will maintain voltage at the setpoint. Since no operator action is taken to adjust voltage it will maintain voltage.
D. 1B Diesel Generator ENGINE EXCTR SHUTDOWN alarm due to exciter shutdown--
Is incorrect, since the diesel will lower its speed, the frequency will lower but the exciter will maintain excitation so the exciter will not shutdown.
OPERATIONS Page: 96 of 146 11June2008
, I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM ComprehensivdAnalysis References : EOP-2 page 38 EOP -2 Basis page 37 Electrical connections andlor cause effect Distribution relationships between the ac distribution system and the OPERATIONS Page: 9 7 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 50 ID: Q50359 Points: 1.00 Given the following:
Unit 1 is operating at 100% power A loss of 12 DC bus occurs Which of the following describes the effect of the loss of 12 DC Bus? ( Assume No operator actions)
A. 1B D/G start solenoids fail open AND SRW CV fails open B. 1B D/G control power loses power AND TCBs 1, 2, 5, 6 trip C. 1B D/G field flash AND control power lose power.
D. 1B D/G start solenoids fail open AND TCBs 3 , 4 , 7,8 trip Answer: C Answer Explanation:
A. 1 6 D/G start solenoids fail open AND SRW CV fails open - Not Correct Start Solenoids fail shut B. 1B D/G control power loses power AND TCBs 1,2, 5, 6 trip Not Correct- TCBs 3,4,7,8trip C. 1B D/G field flash AND control power lose power. Correct D. 1B D/G start solenoids fail open AND TCBs 3 , 4 , 7, 8 trip - Not Correct - start solenoids fail shut OPERATIONS Page: 98 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 50 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to Complete: 3 Difficultv: 4.00 System ID: 50359 User-Defined ID: Q50359 Cross Reference Number:
Totic: Affect on D/G for loss of DC power RO Imoortance: 2.7 n n SRO Importance:
KA Number: 36063K102 Comments: NEW Question : FundamentaVMemory
Reference:
AOP7J page 80 2
Electrical connections andfor cause effect Distribution relationships between the DC 7 electrical system and the following OPERATIONS Page: 99 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 51 ID: Q50361' Points: 1.00 Unit 1 was operating at 100% power when a SlAS occurred with a loss of 11 4KV BUS. Which of the following statements is correct for these conditions?
A. 1A D/G will trip on Low -Low lube oil pressure OR High Crankcase pressure B. 1A D/G will trip on Lube Oil Temp High-High OR Engine Overspeed C. 1A D/G will trip on HT Cooling Water Pressure Low OR generator overvoltage D. 1A D/G will trip on Low - Low lube oil pressure OR Generator Differential Protection Answer: D Answer Explanation:
A. Incorrect, While lube oil pressure low is a trip under Emergency start, Hi crank case pressure is NOT B. Incorrect, While Engine Overspeed will under Emergency start, Lube oil temperature hi will NOT C. Incorrect, HT water pressure is not a trip during emergency start D. Correct, per system description 24C page 63 OPERATIONS Page: 100 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 1 Question 51 Info Multiple Choice Active No No 1.oo 13 3.8 4.1 36064K401 Modified from Q15961 : FundamentaVMemory Reference Diesel Generator Lesson Plan LOI-024A-1-1 Slides 84 -90 Drawing 61086 Sheet 19A interlock(s) which provide for the following: Trips while loading the ED/G(frequency, voltage, OPERATIONS Page: 101 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 52 ID: Q20392 Points: 1.00 Unit 1 has tripped and EOP-0 is implemented. The Condenser Off-Gas (1-RE-1752), S/G Blowdown Recovery (1-RIC-4095), S/G Blowdown Recovery (1-RE-4014) radiation monitor meter indications are pegged LOW and all lights on their panels are out/dark.
What action should be performed to support the Radiation Levels External to Containment (RLEC) safety function?
A. Shut the Steam Generator Blowdown Control Valves and report RLEC cannot be met, due to loss of power effects B. Shut the Steam Generator Blowdown Control Valves and report RLEC is complete C. Restart the sample pumps and re-evaluatethe indications D. Attempt to clear all RMS alarms and re-evaluate the indications Answer: A Answer Explanation:
A. Shut the Steam Generator Blowdown Control Valves and report RLEC cannot be met due to loss of power effects per EOP-0.- Correct 9.Report "Radiation Levels External to Containment is complete" to the CRS--incorrect, the safety function should be reported as "cannot be met, due to loss of power effects".
NO-1-201, pages 25-26 C. Restart the sample pumps and re-evaluate the indications-incorrect, pumps cannot be restarted with power unavailable, as indicated by no lights on the panels energized.
D. Attempt to clear all RMS alarms and re-evaluate the indications--incorrect, alarms and indications cannot be restored with power unavailable, as indicated by no lights on the panels energized.
The cause of these indications is a loss of power to the Rad Monitors Distractor C is plausible since the candidate could have the misconception that the lights are out and indicators low due to a sample pump failing, which could be corrected by restarting it. He could also misconstrue theses indications as those of a saturated detector. 01-35 page 15 gives actions for resetting a saturated detector which includes taking the operate switch to reset , removing a fuse and taking the switch to operate.
Distractor D is plausible since the candidate could have the misconception that a spike has occurred on or near the detector and that by clearing the alarms the system will reset and he could revaluate the conditions. 01-35 gives actions for an RMS alarm due to a failed or degraded detector which include depressing the bypass pushbutton on the alarm display.
OPERATIONS Page: 102 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM IQuestion 52 Info Question Type: Multiole Choice Status: Active Always select on test? No Authorized for Dractice? No I Points: 1.oo 2.00 50362 (220392 I Topic: Given conditions associated with RMS, identify the actions to support RLEC Safety Function RO Importance: 2.5 SRO Importance:
KA Number: 37073A201 Comments: Bank Question : Comprehensive/Analysis Used 7/2002 NRC exam
References:
EOP-0 page 17 EOP-0 basis page 28 NO-1-201, pages 25-26 073 Process A2.01 - Ability to (a) predict the Radiation impacts of the following Monitoring malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Erratic or failed power supply OPERATIONS Page: 103 of 146 11June2008
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EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 53 Info Question Type: Multiple Choice Status: Active NO
~~
1.oo n
1.oo 50363 Q50363R CRO-113-2-5-11 Unit 1 is at 75% power, when 12 SW Pump trips. What actions are required
- 3. I KA Number: 34076A201 Comments: Modified from CRO-113-2-5-11: ComprehensivdAnalysis References : AOP-7A pages 12, 19 AOP-7A Basis page 10 Action Taken When Unitlat 75% Power When 12 SW Pump trips and one header operating 1
076 Service A2.01 - Ability to (a) predict the Water impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Loss of sws OPERATIONS Page: 105 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 54 ID: Q50364 Points: 1.00 Unit -1 is operating at 100% power when Instrument Air System pressure decreases to 96 psig.
Which of the following is correct?
A. Loss of Power to an IA dryer has occurred.
- 6. Standby Air Compressor has picked up C. Plant Air to I/A X-Conn, 1-IA-2061-CV has opened D. Both dryers are in service due to low IA pressure Answer: A Answer Explanation:
A. Correct per Alarm Manual for Window K-26 B. Incorrect, STBY compressor starts Q 93 PSlG Per AOP7D section 1II.C notes C. Incorrect, This CV opens 8 88 psig per AOP 7B section Ill C. notes D. Incorrect, Pressure is 96 not 93 psig IQuestDon 54 Info Multiple Choice Active No Authorized for practice? No Points: 1.oo r.
RO Importance: ~ ~
SRO Importance: 3.2 38073A301 NEW Question: ComprehensivdAnalysis References : 1C-13 Alarm Manual page 48-50 Compressed Air Lesson Plan- LOI-019-1-2 Slides 62, 63 078 Instrument A3.01 -Ability to monitor 3.1 Air automatic operation of the IAS, including: Air pressure OPERATIONS Page: 106 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 55 ID: Q50365 Points: 1.00 Which of the following system configurations must exist to allow resetting CIS from the Control Room?
A. 1-CC-3832, CC CNTMT Supply Vlv HS must be in Shut and 1-IA-2080 MOV OVERRIDE HS must be in Override B. 1-CC-3832, CC CNTMT Supply Vlv HS must be in Shut and 1-IA-2080 MOV must be in shut C. 1-PA-1040,Plant Air CNTMT Isolation Vlv HS must be in Shut and 1-IA-2080 MOV OVERRIDE HS must be in Normal D. 1-CPA-l4lO-CV, CNTMT Purge Supply Vlv HS must be in Shut, and CNTMT Purge EXH Fan must be in OFF Answer: B Answer Explanation:
EOP Attachment 4 provides a list of the components that actuate on CIS, and provides an asterisk for those components that must have their handswitch in the Post Accident position to enable resetting CIS from the control room.
A Incorrect, 1-IA-2080 MOV override ( 1-HS-208A) must be in normal - Is plausible since it does get a CIS signal to close, the H/S is normally in normal. Candidate needs to recall the normal position of the handswitch. He could have a misconception that this H/S must be in normal position to reset since it is in normal.
- 6. Correct per EOP Attach 4 C. Incorrect, 1- PA-1040 is an administratively controlled valve., PA-1040 is for plant air containment isolation, and is administratively controlled. Candidate could have the misconception that since IA CV has a handswitch for overriding then the PA 1040 should have one too. Candidate has to recall the unique arrangement of plant air to containment in that PA-1040 is inside containment and if it is open then 1044 must be shut.
D. Incorrect, While 1- CPA-1410 CV receives a CIS signal, its status does not effect reset from control room, purge exhaust fan receives a CRS not CIS per LD 58. Candidate could have the misconception that this CV (Containment Purge SUPP) H/S needs to be in shut in order to reset CIS since it gets a CIS signal.
OPERATIONS Page: 107 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 55 info I Question TvDe: I MultiDle Choice I Status: I Active KA Number: 35103K406 Comments: New Question : FundamentaVMemory I I EOP Attachment 4 page 1 &2, LD-58 103 K4.06 -Knowledge of containment 3.
Containment system design feature(s) and/or 1 interlock(s) which provide for the following: Containment isolation svstem OPERATIONS Page: 108 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 56 ID: Q50366 Points: 1.00 Given the following:
Unit is operating at 100% power.
An event occurs.
You are standing next to 1C-15 and the CRS directs you to manually trip the Reactor You press two (2) Reactor Trip PBs on 1C-15 Which ONE of the following describes the effect on the Reactor Trip Switchgear?
Reactor Trip Switchgear Breakers....
A. 1, 2, 5, and 6 OPEN; reactor trip occurs B. 3, 4,7, and 8 OPEN; reactor trip occurs C. 1,4 5, and 8 OPEN, reactor trip occurs D. 2,3, 6 and 7 OPEN; reactor trip occurs Answer: D Answer Explanation:
A. 1, 2, 5, and 6 OPEN; reactor trip occurs. Incorrect combination & panel B. 3, 4, 7, and 8 OPEN; reactor trip occurs. Incorrect combination & panel C. 1, 4 5, and 8 OPEN; reactor trip occurs. Correct combination, wrong panel D. 2, 3,6 and 7 OPEN; reactor trip occurs. Correct for K 2 & 3 from 1C-15 OPERATIONS Page: 109 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 56 Info Question Type:
Status:
Always select on test? No Authorized for Dractice? No Points: 1.oo Time to Complete:
System ID:
User-Defined ID:
Cross Reference Number:
ToDic:
RO Importance:
SRO Importance: 4.1 KA Number: 31001K614 Comments: NEW Question: Fundamental/Memory Reference : RPS Lesson Plan LOI-058-1-2 (RPS part 1) slides 50 , 73, 76, 77, 78 Drive of a loss or malfunction on the following CRDS components:
Location and interpretation of OPERATIONS Page: 110 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 57 ID: Q50368 Points: 1.00 Which one of the following describes PZR level response on 1-LI-l03?
A. When drawing a bubble in Mode 5, 1-LI-l03 will read higher than 1-LIC-ll OX and 1-LIC-11OY.
B. When drawing a bubble in Mode 5, 1-LI-l03 will read lower than 1-LIC-11OX and 1-LIC-11OY.
C. At NOP/NOT, 1-LI-l03 will read lower than 1-LIC-11OX and 1-LIC-ll OY.
D. At NOP/NOT, 1-LI-l03 will read the same as 1-LIC-ll OX and 1-LIC-11OY.
Answer: B Answer Explanation:
When drawing a bubble in Mode 5, 1-LI-l03 will read lower than 1-LIC-11OX and 1-LICl 1OY, AND when in Mode 1 at NOP/NOT, 1LI-103 will read approximately the same as 1-LIC-11OX and 1-LICll OY - this is correct because 1-LIC-ll OX and 1-LICl1OY are calibrated for NOP/NOT conditions and as the pressurizer cools the density gets greater causing the DP ( between reference leg and pressurizer) to lessen, given a higher indicated level. LI-103 is calibrated from 1TE-101 and PT-105A and will indicate correctly below 514 psi and 470 F.
OPERATIONS Page: 11 1 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 57 Info Question Type: Multiple Choice Status: Active Always select on test? No I Authorized for practice? No Points: 1.oo Time to ComDlete: 4 Difficulty:
System ID:
User-Defined ID:
Cross Reference Number:
Recall the feature of the PLCS that provides for indication Topic:
of accurate level when RCS is cold RO Importance: 2.9 SRO Importance: 3.2 KA Number: 3201 1K407 Comments: New Question: FundamentaVMemory References RCS instrumentation Lesson Plan LOI-064A2-1 Slide 43 GFES Lesson Plan Sensors & detectors LOI-301-15-3 Slide 70 OPERAT1ONS Page: 112 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 58 ID:Q50369 Points: 1.00 Which NI power indication is lost on a loss of 120VAC bus 2Y02?
A. 1C43 channel B aux excore wide range.
B. Unit 2 Rx Reg Channel X.
C. 1C15 channel B linear range.
D. 2C43 channel B aux excore wide range.
Answer: A Answer Explanation:
A. Correct per ES-013 BS 2Y02 B. Incorrect, Channel Y is lost C. Incorrect, 1C15B Channel B linear range is not lost D. Incorrect, This would occur on loss of 1YO1 Question 58 Info Question Type:
Status: Active No No 1.oo 2
1.oo 50369 Q50369 CRO-57-1-5-05 NI power indication lost due to loss of 120VAC bus 2Y02 3.3
- a. I 37015K201 BANK Question (CRO-57-1-5-05):Fundamental/Memory
References:
AOP-7J basis page 16, ES-013 BKR 2402 Drawing 61022E following: NIS channels, components, and OPERATIONS Page: 113 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 59 ID: Q50370 Points: 1.00 Unit one was operating at 100% power when an event occurred. The plant was tripped and EOP-0 was implemented. The appropriate optimal recovery procedure has now been implemented.
The following indications are noted:
0 SPDS alarm on 1(2)C06 0 CSF3 (Core and RCS Heat Removal) block turns red for "CET High" Which of the following is the Minimum conditions required to cause these indications?
A. 2 of 4 CETs in a quadrant exceeded 625"F B. 1 CET in 2 of 4 quadrants exceeded 625"F C. 2 of 4 CETs in a quadrant exceeded 650"F D. 1 CET in 2 of 4 quadrants exceeded 650"F Answer: C Answer Explanation:
page 2 provides a quadrant display of four highest CETs in each quadrant
- a. Yellow alarms for > 625°F
- b. Red alarms for > 650°F CSF3 red 2 alarm if 2/4CETs in a quadrant exceed 650°F
- a. SDPS alarm on 1(2)C03
- b. CSF3 block turns red for "CET High" OPERATIONS Page: 114of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM MultiDle Choice e Alwa s select on test?
Authorized for ractice?
Points:
Time to Comdete:
Active No No 3
1.oo 2.00 50370 Q50370 3.2 I KA Number: 37017K101 I Comments: New Question: FundamentaVMemory Reference :SPDS screen last page of Core and RCS heat removal K1.O1 -Knowledge of the physical Temperature connections and/or cause effect relationships between the ITM system and the following systems:
Plant commter Learning Objective SPDS Lesson Plan 3.0 Identify the color coding scheme of the SPDS critical safety function boxes and parameters and determine when safety function parameters are invalid or at a reduced logic.
OPERATIONS Page: 115of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 60 ID: Q50371 Points: 1.00 Each Containment Iodine Removal Unit (IRU) is capacity with each unit being efficient for removing Iodine. As humidity level approaches 99%, filter efficiency is A. So%, go%, - 50%
B. loo%, 99%, - 90%
C. loo%, 90%, - 50%
D. So%, 99%, - 90%
Answer: D Answer Explanation:
D. Correct - (1))Each IRU is 50% capacity, with each unit being 99% efficient for removing Iodine, (2) as humidity level approaches 99%, filter efficiency is 90% - Correct per EOP-5 basis A. Incorrect, wrong efficiency Q 99% humidity B. Incorrect, Wrong capacity C. Incorrect, Wrong capacity and wrong efficiency OPERATIONS Page: 116 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Multiple Choice Active Nn 2
2.00 50371 Q50371 associated with the Containment Iodine Removal RO Importance: 3.1 SRO Importance: 3.4 KA Number: 35027K501 Comments: New question : Fundamental/Memory Reference : EOP-5 basis Page 66 027 K5.01 -Knowledge of the 3 Containment operational implications of the Iodine following concepts as they apply to I Removal the CIRS: Puroose of charcoal filters
- OPERATIONS Page: 117 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 61 ID: (250373 Points: 1.00 Unit-1 is in EOP-1 with feedwater controls in automatic mode (feedwater regulating bypass valves are controlling level) when RCP feeder breaker, 252-1201, trips. Assume no other operator action is taken. Which of the following secondary plant parameters observed 25 minutes after the RCP trips indicate a loss of RCS flow is occurring?
A. Lowering steam flow and feed flow with rising S/G pressures B. Rising steam flow and feed flow with lowering S/G pressures C. Rising steam flow and feed flow with rising S/G pressures D. Lowering steam flow and feed flow with lowering S/G pressures Answer: C Answer Explanation:
Rising steam flow and feed flow with rising S/G pressures--is correct, Tave will increase, causing ADVs to open. This will cause steam flow and feed flow to rise. S/G pressures will rise as Thot increases.
Distractors are possible combinations of secondary plant parameters.
OPERATIONS Page: 118 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 3uestion 61 Info 2uestion Type: Multiole Choice Status: Active 4lwavs select on test? No 4uthorized for practice? No
'oints: 1.oo ~
rime to Complete: 5 Iifficulty: 3.00 System ID:
Jser-Defined ID:
Zross Reference Number:
Determine the core and plant parameters response to a ropic:
Loss ot Flow Accident.
3 0 Importance:
SRO Importance:
(A Number: 34035AT 02 Zomments: Bank question used 7/2002 NRC exam :
Comprehensive/Analysis
References:
EOP-2 basis page 16,29 -30 Loss of Flow Indicators in control room prevent exceeding design limits) associated with operating the SIGS OPERATIONS Page: 119 of 146 11June2008
0 0 1 :SW!Od PLCOSO :ai 29 NVX3 OU 3UN 8002 d d N 3 3 A3Y U3MSNV NOIlVNIWVX3
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 62 Info Question Type: MultiDle Choice I Status: Active Always select on test? No ~~~
Authorized for Dractice? No Points: 1.oo Time to Complete: 4 I Difficulty: 3.00 System ID:
User-Defined ID:
Cross Reference Number:
Topic:
RO ImDortance:
SRO Importance:
KA Number: 34041K301 Comments: Modified from Bank question 302:
ComprehensivdAnalysis
References:
AOP-7K basis page 3, 7 041 Steam K3.01- Knowledge of the effect 3.
Dumpmurbine that a loss or malfunction of the 2 Bypass SDS will have on the following: S/G
- Control OPERATIONS Page: 121 of 146 11June2008
I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 63 ID: Q50376 Points: 1.00 Unit 1 is operating at 100% power with Condensate Pumps 11,12 and 13 running when 12 Condensate Pump trips.
What effect will this have on the secondary and what initial steps should be taken to mitigate the consequences?
A. Reduced feed flow to the S/Gs and lowering levels will result. Bias Feed Pump speed as required to restore S/G levels.
B. Lower Feed Pump suction pressure will result. Verify a Condensate Booster pump automatically starts.
C. Cavitation and increased impeller wear will occur on the Condensate Pumps.
Reduce power to maintain Condensate header flow less than 8,000 GPM.
D. Lower Condensate header pressure will exist. Place Hotwell Level Control in manual and bypass the Condensate Demins and Precoat Filters.
Answer: D Answer Explanation:
Reduced feed flow to the S/Gs and lowering levels will result. Bias feed pump speed as required to maintain S/G levels-is incorrect, S/G levels should be maintained by the feed pumps and the feed reg valves automatically.
Lower feed pump suction pressure will result. Verify a condensate booster pump automatically starts.--is incorrect, suction pressure will lower, but not to the point where the standby CBP starts.
Cavitation and increased impeller wear will occur on the Condensate pumps. Reduce power to maintain condensate header flow less than 8,000 GPM.-- is incorrect, power reduction is not required on loss of 1 condensate pump from 100% power.
This is plausible since with only two condensate pumps running the remaining pumps will be operating closer to runout conditions which means the required NPSH has increased bringing them closer to cavitation conditions. However, a power reduction is not required by procedure other measures are taken to ensure adequate suction pressure for the feed pumps, including bypassing precoats and demins. IF you have less than two condensate pumps then a reduction in power is directed by procedure Lower condensate header pressure will exist. Place hotwell level control in manual and bypass condensate demineralizers and precoat filters.--is correct per indications and actions in AOP-3G.
OPERATIONS Page: 122 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Multiple Choice Active No No 1.oo 2
3.00 50376 Evaluate effects of a loss of a Condensate Pump 40 244 Bank Question Used 7/2002 NRC exam :
ComprehensivdAnalysis
References:
AOP-3G basis pages 9-11 conditions for emergency and OPERATIONS Page: 123 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 64 ID: Q50377R Points: 1.00 Unit 1 is operating at 100% power when a Gaseous Waste Discharge (0-RE-2191) high rad alarm is received with NO gaseous release in progress. Which of the following is NOT a probable cause of this alarm?
A. Spent Resin Metering Tank Venting
- 6. Leakage from VCT gas space through H2 lines.
C. U-1 CVCS Ion Exchangers venting D. U-2 CVCS ton Exchangers venting Answer: 6 Answer Explanation:
Leakage from VCT gas space through H2 lines is correct per Alarm manual response for 1C-22 D-1.1 and A-5.1, and AOPGC possible sources of Gaseous activity. Leakage from VCT gas space through H2 lines would be seen by BAST RM ( 2-RI-7010)
Radiation monitor Spent Resin Metering Tank Venting is incorrect; this is a probable source as listed by 1C-22, D-1.1 and AOP 6C pages 6-14 Ion Exchangers venting is incorrect ; this is a probable source as listed by 1C-22, D-1.1 and AOP 6C pages 6-14 VCT Relief Valve lifting, is incorrect; this is a probable source as listed by 1C-22, D-1.1 and AOP 6C pages 6-14 OPERATIONS Page: 124 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 64 info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to Comdete: 1 2.00 Svstem ID: 50573 User-Defined ID: Q50377R Cross Reference Number:
Topic: Most likely cause of Gaseous waste rad alarm RO Importance: 3.3 SRO Importance: 3.5 KA Number: 39071A409 Comments: New Question : FundamentaVMemory
References:
Alarm manual for 1 C22H Page 102-103, 147-149 , and AOP 6C pages 6-14 Gaseous Waste Discharge High Rad Alarm probable causes I 071 Waste Gas I A4.09- Ability to manually I 3.3 Disposal operate and/or monitor in the control room: Waste gas release rad monitors OPERATIONS Page: 125 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 65 ID: Q50378R Points: 1.00 A smoke detector for the Unit-1 45' Switch Gear Room malfunctions, causing an alarm.
Which one of the following describes the effect on the system and the appropriate response?
A. "FIRE PROT PANEL 1C24B" alarm actuates and after a time delay, Halon system discharges. Reset the fire panel 1FP430 and immediately restore ventilation in the Switchgear room.
B. "FIRE PROT PANEL 1C24B" alarm actuates and immediately results in Halon system discharge. Reset the fire panel 1FP430, do not restore ventilation in the Switchgear room until authorized by the CRS.
C. "FIRE SYS" alarm actuates and after a time delay, Halon system discharges.
Reset the fire panel 1FP430, do not restore ventilation until authorized by the CRS.
D. "FIRE SYS" alarm actuates but does not result in Halon system discharge.
Reset the fire panel 1FP430 Answer: D Answer Explanation:
A. "FIRE PROT PANEL 1C24B" alarm actuates and after a time delay, Halon system discharges. Reset the fire panel 1FP430 and immediately restore ventilation in the Switchgear room. - Incorrect per plant design and procedures B. "FIRE PROT PANEL 1C24B" alarm actuates and immediately results in Halon system discharge. Reset the fire panel 1FP430, do not restore ventilation in the Switchgear room until authorized by the CRS. - Incorrect per plant design and procedures C. "FIRE SYS" alarm actuates and after a time delay, Halon system discharges. Reset the fire panel 1FP430, do not restore ventilation until authorized by the CRS.- Incorrect per plant design and procedures.
D. "FIRE SYS" alarm actuates but does not result in Halon system discharge. Reset the fire panel 1FP430 - Correct per plant design and procedures OPERATIONS Page: 126 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM I Question 65 Info I Question TvDe: MultiDle Choice Status: Active No No 1.00 3
2.00 50378 ~~
Q50378R FIRE PROTECTION 002 What is the effect on the system for a 45' SGR Smoke Detector malfunction?
I RO Imoortance:
ISRO Importance:
38086A203 Modified from 28827 : Fundamental/Memory Reference : 124B Alarm Manual page 88 & 89 Fire System Lesson Plan 086 Fire A2.03 - Ability to (a) predict the Protection impacts of the following malfunctions or operations on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Inadvertent actuation of the FPS due to circuit failure or welding I I OPERATIONS Page: 127 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 66 I D (250381R Points: 1.00 In the event of a Control Room evacuation, what means is provided for control of the Saltwater Pumps?
A. Remote override transfer valves in the SRW pump room.
B. Control switches on 1C43 for 11 & 12 SW pumps C. Localhemote keyswitches at the pump breakers.
D. LOCI sequencer override pushbuttons at the ESFAS cabinets.
Answer: C Answer Explanation:
A. Incorrect, these valves operate other valve not electrical equipment B. Incorrect, Per AOP9A the IocaVremote key switches are used for this purpose. 1C43 is the remote shutdown panel C. Correct D. Incorrect, AOP 9A assumes no LOCI OPERATIONS Page: 128 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Westion 66 Info SIuestion Type: Multiple Choice status: Active 4lwavs select on test? NO 4uthorized for practice? No
'oints: 1.oo rime to Complete: 2 Iifficulty: 1.oo Svstem ID: 50381 Jser-Defined ID: Q50381R 3oss Reference Number: CRO-113-2-5-15 Topic:
provided for control of the Saltwater Pumps 3 0 Importance: 3.9 SRO Importance: 3.4
.(A Number: 2130 Zomments: Bank Question : FundamentaVMemory References Salt Water Lesson Plan LOI-012-1-2 slides 50 -
52 Basis: Control Room Evacuation Means Provided for Control of SW Pumps
References:
KA1: KA2:
2.1. Ability to locate and operate 4.4 30 components, including local controls.
OPERATIONS Page: 129 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 67 ID: Q50383 Points: 1.00 Using Provided
References:
Unit 2 is operating at 100 % power with 23 CCW Pump 00s. The following sequence of events occur:
0 1205 - 21 CCW Pump declared INOPERABLE due to a failed surveillance.
0 1232 - 22 CCW Pump also declared INOPERABLE due to the results of a common cause failure analysis.
0 1259 - Plant Shutdown to Mode 3 commenced.
0 1324 - 21 CCW Pump returned to OPERABLE status.
0 1343 - 22 CCW Pump returned to OPERABLE status.
Which ONE of the following describes the Technical Specification requirements for operation of the plant?
Plant conditions...
A. allow the plant shutdown to be terminated no earlier than 1324.
B. allow the plant shutdown to be terminated no earlier than 1332 C. require that the Shutdown to Mode 3 is completed by 1832 D. require that the Shutdown to Mode 3 is completed by 1932 Answer: A Answer Explanation:
A. allow the plant shutdown to be terminated no earlier than 1324. - Correct 8 13:24 we are back on original 72 hr clock from 1st CCW pump being 00s.
B. allow the plant shutdown to be terminated no earlier than 1332.- Incorrect- Distractor uses number 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after 3.03 is entered C. require that the Shutdown to Mode 3 is completed by 1832. - Incorrect- This would be only 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after 3.03 is entered D. require that the Shutdown to Mode 3 is completed by 1932.- Incorrect - Would be 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> after 3.03 if one pump not returned to service.
OPERATIONS Page: 130 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 67 Info Question Type: MultiDle Choice I Status: Active I No 1 3.00 2132 New question : ComprehensivdAnalysis Reference- Technical specifications 1-3.1 through 1-3.13 and 3.7.5 -2 through 3.7.5-2 2.1, Ability to explain and apply system limits 3.8 32 and mecautions LOlT Learning Objective for ROs and SROs Given plant conditions and/or plant parameters related to CCW system operations and Technical OPERATIONS Page: 131 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 68 ID: Q50385R Points: 1.00 Unit 2 has just completed a refueling outage and is conducting PSTP3," Escalation to Power Test Procedure", to test at the power plateau of 85% power. At 80% it was determined that Frt is greater than the full power value of T.S 3.2.3. While reviewing the data a transient occurs and power rises to 90% and is stabilized. Which of the following is required?
A. Reduce Thermal Power to less than or equal to 85% within 1hour B. Reduce Thermal Power to less than or equal to 85% within 15 minutes C. Reduce Thermal Power to less than or equal to 80% within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> D. Reduce Thermal Power to less than or equal to 80% within 15 minutes Answer: B Answer Explanation:
A. Incorrect, the power level is correct but the time to reduce is wrong B. Reduce Thermal Power to less than or equal to 85% within 15 Mins - Correct per T.S.
3.1.8 C. Incorrect, the time to reduce power and the power level are wrong D. Incorrect, the power level to reduce to is incorrect, Question 68 Info I Question TvDe: I Multide Choice I 1 Status: I Active I Always select on test? No Authorized for practice? No Points: 1.oo Time to ComDlete: 4 SRO Importance: 4.1 KA Number: 2222 Comments: NEW Question : Comprehensive/Analysis I Reference Technical Specification 3.1.8 and PST-3 page 6 2.2. Knowledge of limiting conditions for 4.0 22 operations and safety limits OPERATIONS Page: 132 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 69 ID: Q50384R Points: 1.00 Unit-2 is in Hot Standby and the latest leakage reports are:
0.8 gpm - RCS drain valve weld leakage 1.8 gpm - leakage past check valves from the RCS to the SI system 72 gpd - primary-to-secondaryleakage (57.6 gpd 21 S/G, 14.4 gpd 22 S/G) 2.7 gpm - total leakage.
Which of the following Technical Specification leakage limits are exceeded?
A. Primary to Secondary leakage AND Unidentified leakage.
B. Pressure Boundary leakage AND Unidentified leakage.
C. Pressure Boundary leakage ONLY.
D. Unidentified leakage ONLY,.
Answer: C Answer Explanation:
Per Technical Specifications 3.4.1 3 You can Have NO Pressure Boundary Leakage 1 GPM unidentified Leakage 10 GPM identified leakage 100 GPD per S/G A. Primary to Secondary leakage AND Unidentified leakage.-- is incorrect because primary to secondary leakage is less than T.S. and pressure boundary is > T.S.
B. Pressure Boundary leakage AND Unidentified leakage.-- is incorrect because Unidentified leakage is < 1 gpm. I2.7-(1.8+.8=.05)=.05 gpm}
C. Pressure Boundary leakage ONLY.-- Is correct, pressure boundary leakage exists associated with the RCS drain valve weld crack D. Unidentified leakage ONLY.-- is incorrect is incorrect because Unidentified leakage is
< 1 gpm. (2.7-(1.8+.8=.05)=.05 gpm}
OPERATIONS Page: 133 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question Twe: MultiDle Choice I Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to Complete:
Difficulty:
System ID:
User-Defined ID:
Cross Reference Number:
Topic: RCS leakaae TS RO ImDortance:
SRO Importance: 4.6 KA Number: 2242 Comments: Modified Bank Question : Comprehensive/Analysis
References:
Technical Specification 3.4.1 3-1 through 3.4-13-3 2.2. Ability to recognize system parameters that 3.9 42 are entry-level conditions for Technical Specifications.
OPERATIONS Page: 134 of 146 llJune2008
I EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 70 ID: (250387 Points: 1.00 Unit 2 is performing a startup after a refueling outage. Power is currently 20% and the Turbine is in "Hold". The CRO places Steam Generator Blowdown in service at 100 GPM/SG per plant Chemistry recommendations.
Which of the following describes the immediate plant response to this evolution? (Assume no additional operator actions)
A. Reactor power increases, letdown flow increases, feedwater flow increases.
- 6. Reactor power decreases, letdown flow increases, feedwater flow decreases.
C. Reactor power increases, letdown flow decreases, feedwater flow decreases.
D. Reactor power decreases, letdown flow decreases, feedwater flow increases.
Answer: D Answer Explanation:
Reactor power increases, letdown flow increases, feedwater flow increases.
Incorrect, after refueling a + MTC exists, so increased blowdown flow will cause a lowering in power.
Reactor power decreases, letdown flow increases, feedwater flow decreases.
Incorrect, feedwater flow will increase due to the blowdown flow increasing.
Reactor power increases, letdown flow decreases, feedwater flow decreases.
Incorrect, both temperature and power will decrease, Reactor power decreases, letdown flow decreases, feedwater flow increases.
Correct, + MTC at BOC which means power lowers with decreasing temperature.
OPERATIONS Page: 135 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 70 Info Question TvDe: MultiDle Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to Complete: 4 Difficulty: 4.00 System ID: 50387 User-Defined ID: Q50387 Cross Reference Number:
Topic: Effects of SG Blowdown on Dlant Darameters RO ImDortance:
SRO Importance: 4.4 KA Number: 2244 Comments: New Question: Comprehensive/Analysis References : Blow Down System Lesson plan LOI-83-1-0 Slides 93, 96, 97 01-8A Rev 38 page 18 2.2. Ability to interpret control room indications to 4.2 44 verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
OPERATIONS Page: 136 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 71 ID: Q50388 Points: 1.00 Given the following:
The current month is April 2008 A CCNPP employee worked an outage at Ginna NPP in February 2008.
The dose received at GINNA was 750 millirem.
His TEDE radiation exposure for the year is 1550 millirem.
The remainder of his dose was received at CCNPP.
Which ONE of the following describes the MAXIMUM additional dose the employee may receive prior to exceeding his TEDE alara dose goal for the year?
A. 700 millirem B. 1450 millirem C. 1700 millirem D. 2450 millirem Answer: B Answer Explanation:
1450 millirem = correct 1700 millirem = incorrect, adds dose from Ginna 2450 millirem = incorrect, based on max admin limit of 4000 mr 700 millirem = incorrect, adds Ginna and based on max admin limit of 3000 mr OPERATIONS Page: 137 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Multiple Choice Status: Active No I No 1.oo 3
1.oo Cross Reference Number:
KA Number:
Comments:
2.3. Knowledge of radiation exposure limits under 3.2 4 normal or emergency conditions.
OPERATIONS Page: 138 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 72 ID: Q50389 Points: 1.00 For entry into a LOCKED HIGH RADIATION AREA, which of the following correctly describes ALL requirements that must be met?
A. A Radiation Work Permit, work approved by GS-RP, Radiation Protection approval for access, Pre Job Brief, dosimetry, pre-entry verification B. A Radiation Work Permit, work approved by GS-RP, Radiation Protection approval for access, Pre Job Brief, dosimetry, hand held survey instrument C. A Radiation Work Permit, continuously indicating dose rate meter, Radiation Protection approval for access, Pre Job Brief, dosimetry, pre-entry verification.
D. A Radiation Work Permit, continuously indicating dose rate meter, Radiation Protection approval for access, Pre Job Brief, dosimetry, hand held survey instrument Answer: C Answer Explanation:
A Special Work Permit, continuously indicating dose rate meter, Radiation Protection approval, Pre Job Brief, dosimetry, pre-entry verification = correct per RPl-100.
All others answers are not in accordance with RP-1-100 OPERATIONS Page: 139 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM MultiDle Choice Status: Active I Always select on test?
Authorized for practice? No Points: 1.oo Time to Complete:
Difficulty: 2.00 System ID: 50389 User-Defined ID: Q50389 Recall radiological safety principles, such as locked high Topic:
radiation area RO Importance: 3.2 SRO Importance: 3.7 KA Number: 2312 Comments: New Question : Fundamental/Memory RP-1-100 page 31 pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-OPERATIONS Page: 140 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 73 ID: Q50390 Points: 1.00 Unit-1 Waste Processing Ventilation Radiation Monitor (1-RE-5410) is in alarm. All other RMS indications remain normal.
Which of the following is the correct cause?
A. A fuel handling event in the spent fuel pool B. Elevated dose rates in the ECCS pump rooms due to SDC operation C. Leakage from a Waste Gas Compressor D. Excessive packing leakage from a Charging pump Answer: D Answer Explanation:
Excessive packing leakage from a Charging pump--correct per alarm response manual and operator experience A fuel handling event in the spent fuel pool--incorrect, the SFP area monitor and Main Vent RMS indications would also be expected to rise. Is plausible since a fuel handling event could cause this monitor to alarm, however other indicators would expect to rise as well, since the stem says all other indications re normal , you don't have a confirming rad monitor indication to support this event.
Elevated dose rates in the ECCS pump rooms due to SDC operation-incorrect, ECCS pump room RMS would indicate this, WP would only increase if SDC leakage were present.
Leakage from a Waste Gas Compressor--incorrect, Main Vent and Waste Gas Equipment Room area monitors would indicate this location for leakage.
OPERATIONS Page: 141 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM Question 73 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No I Points:
Time to Complete:
Difficulty:
System ID:
User-Defined ID:
Cross Reference Number:
I Topic:
Liquid Waste releases
, 1 RO Im ortance:
I KA Number:
3.4 3.8 2313 Comments: Bank Question Used 7/2002 NRC exam :
CornprehensivdAnalysis
References:
Alarm Response Manual 1C22 pages 52 -56 pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-OPERATIONS Page: 142 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 74 ID: Q50714 Points: 1.00 AFW Auxiliary Status Panel alarm TURB SYS L/U IMPR annunciates in the Control Room. Which of the following is a reason for this?
A. 1-MS-3988 (12 AFW PP Trip Throttle Valve) shut.
B. 1-HS-4070 (11 S/G AFW STM SUPP & BYPASS) in CLOSE position.
C. 1-HS-4071 (12 S/G AFW STM SUPP & BYPASS) in OPEN position.
D. 1-HS-4520 (1 1 SIG Block Valve) in CLOSE position.
Answer: B Answer Explanation:
The TURB SYS LKJ IMPR alarm is received for the following inputs:
Both 1-MS-3986-CVand I-MS-3988-CV being shut (Turbine Stop Throttle valves). One of them is open and the other is shut. Normally 11 AFW pump is aligned for auto initiation with its trip throttle valve open, 12 AFW pump is not aligned and its trip throttle valve is shut.
1I SG AFW STM SUPP & BYPASS , IHS-4070 in the CLOSE position ( need to be in AUTO) 12 SG AFW STM SUPP & BYPASS, 1-HS-4071 in the CLOSE position (need to be in AUTO)
I-HS 3986A in DISABLE position (will cause the Stop throttle valve to not trip from the control room) 1-HS-3988A in DISABLE position (will cause the Stop throttle valve to not trip from the control room)
I-MS-3988 (12 AFW PP Trip Throttle Valve) shut. Incorrect because this valve is normally shut.
Alarm is based on if both 11 and 12 Trip Throttle Valves are shut.
1-HS-4070 (11 S/G AFW STM SUPP & BYPASS) in CLOSE position. Is correct answer per Alarm Manual 1-HS-4071 (12 S/G AFW STM SUPP & BYPASS) in OPEN position. Incorrect because taking this handswitch to close causes alarm, not open (per alarm manual only CLOSE position causes alarm 1-HS-4520 (11 S/G Block Valve) in CLOSE position. Incorrect because taking this HS to CLOSE causes 1 1SG L/U IMPR. not the TURB SYS L/U IMPR OPERATIONS Page: 143 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 2.4. Ability to verify that the alarms are consistent 4.2 46 with the plant conditions OPERATIONS Page: 144 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM 75 ID: (250398 Points: 1.00 At 6am on a Saturday morning a terrorist attack has resulted in significant damage to the Control Room and Auxiliary Building and all members of the control room staff were incapacitated.
Security has secured the terrorist and are attempting to coordinate activities in the Emergency Plan with Operations. In addition to the security personnel, the following onsite personnel are available :
Operations Manager (Inactive License)
Principal Plant Operator (Non-Licensed)
Plant Operator Instructor (Active License)
Turbine Building watch (Non-Licensed)
Outside Watch (Non-Licensed)
Which of the following is correct for the listed conditions concerning coordination of required actions per the ERPIP?
A. The Principal Plant Operator is the operator in charge and will go to the Secondary Fire Brigade Locker and assume command of the site.
B. The Operations Manager is the operator in charge and will go to the Operational Support Center and assume command of the site.
C. The Plant Operator Instructor is the operator in charge and will go to the Secondary Fire Brigade Locker and assume command of the site.
D. The Operations Manager is the operator in charge and will go to the Fire Brigade locker and assume command of the site.
Answer: C Answer Explanation:
The Plant Operator Instructor is the operator in charge and will go to the Secondary Fire Brigade Locker and assume command of the site- Correct - Per ERPIP 3.0 Attachment 27.
The Principal Plant Operator is the operator in charge and will go to the Secondary Fire Brigade Locker and assume command of the site.-- Incorrect, The Principal Plant Operator is the Senior non-licensed watchstander and would be in charge if there were NO licensed operators present, since there is a currently licensed operator present he would be the operator in charge.
The Operations Manager is the operator in charge and will go to the Operational Support Center and assume command of the site.-- Incorrect. The Operations Manager does not have an active license, and by procedure the highest ranking licensed operator would be in charge. Also the Operational Support Center is not the correct place per procedure to set up the command and control for these conditions.
The Plant Operator Instructor is the operator in charge and will go to the Secondary Fire Brigade Locker and assume command of the site.-- Is Correct, since he has an active license he would assume command and control.
OPERATIONS Page: 145 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC RO EXAM The Operations Manager is the operator in charge and will go to the Fire Brigade locker and assume command of the site. Incorrect. The Operations Manager does not have an active license, and by procedure the highest ranking licensed operator would be in charge.
Question Type: Multiple Choice Status: Active Always select on test?. No Authorized for Dractice? No Points: 1.oo Time to Complete:
Difficulty:
50398 User-Defined ID: Q50398 Cross Reference Number:
Topic:
RO Importance:
SRO Importance:
KA Number: 2437 Comments: New Question: ComprehensivdAnalysis References Per ERPlP 3.0 Attachment 27 Note.
2.4. Knowledge of the lines of authority during 37 implementation of the emergency [plan.
OPERATIONS Page: 146 of 146 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 1 ID: Q50652 Points: 1.00 Unit 1 was operating at 100% power when a Reactor Trip occurred. 5 Minutes after the trip the following indications are noted:
All 13.8 KV busses are energized 11 and 14 4KV busses are energized 11 S/G pressure 925 PSlA 12 S/G pressure 920 PSlA 11 S/G Level is - 150 inches and rising slowly 12 S/G Level is - 165 inches and rising slowly RCS Tcold 536 OF Pressurizer Level is 65 inches and slowly lowering Pressurizer pressure is 1900 PSIA and steady 11 charging pump is running Reactor Power is lo-*"
1 CEA indicates stuck out Which of the following are the correct reports and required actions by the CRO and RO?
A. CRO reports: "Taking alternate actions for Core and RCS heat removal Safety Function, manually opening TBV's and initiating AFW flow.
RO reports : "Taking alternate actions for RCS Pressure and Inventory Control Safety Function, starting manually operating pressurizer heaters" B. CRO reports: "Taking alternate actions for Core and RCS heat removal Safety Function, manually opening the TBVs" RO reports : "Taking alternate actions for RCS Pressure and Inventory Control Safety Function, starting 12 Charging pump" C. CRO reports: "Taking alternate actions for Core and RCS Heat Removal Safety Function, opening TBV's and initiating AFW flow.
RO reports : "Taking alternate actions for Reactivity Control Safety Function, borating the RCS D. CRO reports: "Taking alternate actions for Core and RCS heat removal Safety Function, manually opening TBV's. RO reports: "Taking alternate actions for RCS Pressure and Inventory Control Safety Function, manually operating pressurizer heaters".
Answer: B Answer Explanation:
A. CRO reports: "Taking alternate actions for Core and RCS heat removal Safety Function, manually opening TBV's and initiating AFW flow. RO reports: "Taking alternate actions for RCS Pressure and Inventory Control Safety Function, starting manually operating pressurizer heaters". Is not correct, since RCS pressure has stabilized between 1850 and 2300 PSIA, and AFW flow is not required.
OPERAT1ONS Page: 1 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam B. CRO reports: 'Taking alternate actions for Core and RCS heat removal Safety Function, manually opening the TBVs". RO reports: "Taking alternate actions for RCS Pressure and Inventory Control Safety Function, starting 12 Charging pump". Is correct since the TBVs have not restored Tcold and S/G pressure, and pressurizer level is not being restored by the current charging configuration.
C. CRO reports: "Taking alternate actions for Core and RCS Heat Removal Safety Function, opening TBV's and initiating AFW flow. RO reports: "Taking alternate actions for Reactivity Control Safety Function, borating the RCS. Is not correct, only one CEA is stuck out and the reactor is responding to the trip as expected so the reactivity control safety function is being met.
D. CRO reports: "Taking alternate actions for Core and RCS heat removal Safety Function, manually opening TBVs. RO reports: "Taking alternate actions for RCS Pressure and Inventory Control Safety Function, manually operating pressurizer heaters". Is not correct. Pressurizer pressure is stable at 1900 PSlA so the PPCS has restored RCS pressure so no action is required.
Stabilization - or interpret the following as Recovery /1 they apply to a reactor trip:
Proper actions to be taken if the automatic safety functions OPERATIONS Page: 2 of 51 11June2008
I I EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 2 I D Q50402R Points: 1.00 Unit 1 was in mode 1 when a loss of service water pump 12 occurred. 13 service water pump has been started and is operating properly. The cause of the trip of 12 Service water pump is not known. Electrical Maintenance has initiated a troubleshooting plan to determine and repair the cause. This activity is determined to be Nuclear Medium Risk. Which of the following is the responsibility of the on-shift SRO with regard to troubleshooting activities?
A. Ensuring adequate boundaries are identified to control troubleshooting scope, B. Approving troubleshooting plans as defined by risk assessment process.
C. Classifying the troubleshooting approach as routine or complex.
D. Closing out each troubleshooting plan.
Answer: A Answer Explanation:
A. Ensuring adequate boundaries are identified to control troubleshooting scope. -
Correct per CNG-MN-1-1.01
- 6. Approving troubleshooting plans as defined by risk assessment process. Incorrect, this is GS-Shift Operations Responsibility for Nuclear Medium Risk activities.
C. Classifying the troubleshooting approach as routine or complex. - Incorrect, this is Responsible Group Supervisor (Electrical Maintenance) responsibility.
D. Closing out each troubleshooting plan. Incorrect, this is Responsible Group Supervisor (Electrical Maintenance) responsibility OPERATIONS Page: 3 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Question 2 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to Complete: 3 3.00 System ID: 50570 User-Defined ID:
Cross Reference Number:
Topic:
RO Importance: 2.6 SRO Importance: 3.8 KA Number: 2220 Comments: NEW Question : Fundamental/Memory
Reference:
CNG-MN-l.01-1002- Rev 0001 pages 6-10 062 Loss of Nuclear 2.2.20 - Knowledge of the 3.8 Service Water process for managing troubleshooting activities OPERATIONS Page: 4 of 51 11June2008
EXAMINATION ANSWER KEY GCNPP 2008 NRC SRO Exam 3 ID: Q50405 Points: 1.00 Unit 1 was operating at 100% power when an event occurred which required a reactor trip. The following conditions exist:
RCS Pressure 1825 PSlA and slowly rising Containment Pressure is 0 PSlG and steady Tcold is 500°F and rising slowly 1 1 S/G Pressure is 680 PSlG and rising 12 S/G Pressure is 760 PSlG and rising Pressurizer Level is 100 inches and rising RCS Subcooling is 1 13 "F and slowly rising (Nooperator actions were performed except to trip the reactor).
Which of the following actions should be directed by the CRS for these conditions?
A. Operate the TBVs to restore Tcold to between 525°F and 535°F B. Operate the ADVs to restore Tcold to between 525°F and 535°F C. Operate the TBVs to maintain Tcold at current value.
D. Operate the ADVs to maintain Tcold at current value.
Answer: D Answer Explanation:
A. Operate the TBVs to restore Tcold to between 525°F and 535°F.-- Is Incorrect, MSlVs should be shut for this condition so TBVs would not be useful, and based on the indications an excess steam demand event which has been terminated by the SGlS has occurred . Per the current guidance in EOP-0 and EOP-0 Basis document for step 1.4,Tcold should not be restored to the normal EOP-0 band.
B. Operate the ADVs to restore Tcold to between 525°F and 535°F- Is Incorrect-since based on the indications an excess steam demand event which has been terminated by the SGlS has occurred. Per the current guidance Tcold should not be restored to the normal EOP-0 band.
C. Operate the TBVs to maintain Tcold at current value. -- Is Incorrect, MSlVs should be shut for this condition so TBVs would not be useful.
D. Operate the ADVs to maintain Tcold at current value. - Is correct, Per the current guidance in EOP-0 and EOP-0 Basis document for step 1.4,Tcold should not be restored to the normal EOP-0 band.
OPERAT1ONS Page: 5 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Candidates should be able to interpret from the conditions given that an SGIS occurred that isolated the excess steam demand since 11 SG Pressure is currently less than the SGIS setpoint and rising and containment pressure is 0 psig. The steps to stabilize Tcold at current value if a SGIS has actuated was added following an event at CCNPP described in LER 318/2004-01 Reactor Trip and two SIAS actuations. This step was added to ensure that Tcold is stabilized to reduce the effects of an in-surge of relatively cold water into the pressurizer which could compress the steam bubble and raise RCS pressure above the saturation temperature of the pressurizer water volume, which could lead operators to believe that Safety Injection can be reset. In fact when the transient conditions caused by the insurge stabilizes, pressurizer pressure will return to the saturation pressure for the current pressurizer temperature which can lead to an inadvertent SlAS New Question: Comprehensive/Analysis Reference EOP-4 basis document 00040/Steam Line 2.4.6 Knowledge of EOP 4.7 Rupture Excessive mitigation strategies.
Heat Transfer OPERATIONS Page: 6 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 4 ID: Q50407 Points: 1.00 U-1 is operating at 100% power when a loss of #11 125 volt DC bus occurs. Which of the following sets of statements are correct for these conditions?
A. 11A and 11B RCP breakers are tripped locally, 12A and 12B are secured from the Control Room, No feedwater is available, RCS heat input must be reduced B. 11A and 12A RCP breakers are tripped locally, 11B and 126 are secured from the Control Room, No feedwater is available, RCS heat input must be reduced C. 11A and 11B RCP breakers are tripped locally, 12A and 12B are secured from the Control Room, to protect them from possible damage D. 11A and 12A RCP breakers are tripped locally, 11B and 12B are secured from the Control Room, to protect them from possible damage Answer: D Answer Explanation:
A. 11A and 11B RCP breakers are tripped locally, 12A and 12B are secured from the Control Room, No feedwater is available, RCS heat input must be reduced. -- Is Incorrect. 11B and 12B RCPs are tripped from the control room since they still have control power available (1 1A and 12A do not have control power so must be tripped locally), and while it is true that main feedwater is not available due to the loss of service water to the turbine building, auxiliary feedwater is still available via 11 AFW pump, or 13 AFW pump could be started locally.
B. 11A and 12A RCP breakers are tripped locally, 11B and 12B are secured from the Control Room, No feedwater is available, RCS heat input must be reduced. -- Is Incorrect. the correct RCPs are listed for local and remote tripping, however while it is true that main feedwater is not available due to the loss of service water to the turbine building, auxiliary feedwater is still available via 11 AFW pump, or 13 AFW pump could be started locally.
C. 11A and 11B RCP breakers are tripped locally; 12A and 128 are secured from the Control Room, to protect them from possible damage. -- Is Incorrect. 11B and 126 RCPs are tripped from the control room since they still have control power available (
11A and 12A do not have control power and must be tripped locally).
D. 11A and 12A RCP breakers are tripped locally, 11B and 12B are secured from the Control Room, to protect them from possible damage.--Is Correct. per AOP 7J XI.
A1 .c,d. Since the RCPs have no Component Cooling Flow they must be secured to protect them from possible damage.
AOP-7J XI. A.l.c, d.--Unit-1 rev. 19, and basis.
OPERATIONS Page: 7 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam MultiDle Choice Status: Active No Nn 1.oo Time to Comdete: 5 1.oo 50407 Q50407 AOP-7J-02 loss of #11 125 volt DC -trip RCPs 3.7 4.1 I KA Number: 42058AA203 I Comments: Modified from Q 39067: Comprehensive/Analysis Reference AOP-7J XI. page 59-61 and AOP7J Basis page 18 OPERATOR ACTIONS FOR LOSS OF 120 VOLT DC BUS
References:
AOP-7J 000058 Loss of DC AA2.03- Ability to determine 3.
Power16 and interpret the following as 9 they apply to the Loss of DC Power: (CFR: 43.5 /
45.13)DC loads lost; impact on ability to operate and monitor plant systems OPERATIONS Page: 8 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 5 ID: Q50409R Points: 1.00 U-1 is operating at 100% power when an event occurs and the following Alarms are noted:
CONDS PPS DlSCH PRESS LO 0 SGFP(S) SUCT PRESS LO CNDSR HOTWELL LVL The appropriate AOP is entered and actions taken. A Reactor trip occurs and EOP-0 was entered and the proper Optimal recovery procedure is now being implemented when the following indications are noted:
0 MOTOR SYS NO FLOW TURB SYS NO FLOW 0 11 S/G e -380 inches 0 12 S/G - 300 inches 0 CET temperature 562 "F 0 Tcold rises from 550°F to 558°F in 30 seconds Which of the following is correct for the conditions given?
A. Initiate Once-through core cooling because 11 S/G is less than -350 inches B. Initiate Once-through core cooling because both S/G are less than - 300 inches with NO main or aux feed C. Initiate Once-through core cooling because 11 S/G is less than -380 inches D. Initiate Once-through core cooling because Tcold has risen 5°F uncontrollably Answer: D Answer Explanation:
A is incorrect because it is only one S/G. Both must be less that -350".
B is incorrect because level is greater than -350".
C is incorrect because it is not a requirement for initiation of OTCC , this is the point where the S/G is isolated to prevent dryout D. Initiate Once-through core cooling because Tcold has risen 5°F uncontrollably -
correct per step J.2 of EOP Basis: OTCC
References:
EOP-3 rev. 2 step J.2 Initiate OTCC if both S/Gs are less than -350" or Tcold rises uncontrollably 5 degrees F or greater.
OPERATIONS Page: 9 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Question 5 Info Question Tvpe:
Status: Active I Always select on test?
Authorized for practice?
Points:
Time to Complete: 3 I User-Defined ID: Q50409R Cross Reference Number:
Topic: identify the proper [procedure for the conditions and alarms RO Importance: Q C SRO importance: 3.6 KA Number: 2446 Comments: New Question: ComprehensivdAnalysis
References:
EOP-3 pages 22 -24 and EOP-3 basis pages 31 -33 CE/EO6 Loss of Main 2.4.46 - Ability to verify that 4.
Fedwater/4 the alarms are consistent with 2 the plant conditions.
OPERATIONS Page: 10 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 6 ID: (250493 Points: 1.00 Upon receiving a Gaseous Waste Discharge (0-RE-2191) high rad alarm,(l) what automatic action(s) occur, and (2) what immediate follow-up action is required?
A. (1) Shuts waste gas discharge CVs, 0-WGS-2191 and 2192; (2) Reduce the Waste Gas Surge Tank pressure to prevent lifting the Waste Gas Surge Tank relief valve.
B. (1) Shuts waste gas discharge CVs, 0-WGS-2191 and 2192; (2)manually shut waste gas discharge header flow control valve, O-WGS-2191-PCV to prevent lifting the Waste Gas Discharge Relief Valve.
C. (1) Shuts waste gas discharge header flow control valve, O-WGS-2191-PCV; (2) Reduce the Waste Gas Surge Tank pressure to prevent lifting the Waste Gas Surge Tank relief valve.
D. (1) Shuts waste gas discharge header flow control valve, O-WGS-2191-PCV;(2) manually shut waste gas discharge CVs, 0-WGS-2191 and 2192 to prevent lifting the Waste Gas Discharge Relief Valve.
Answer: B Answer Explanation:
Per alarm actions for D-1.1 O-WGS-2191-PCV must be shut as soon as possible to prevent lifting the Waste Gas discharge header relief from going to the waste gas surge tank. This could lead to an inadvertent release out of the stack via the Waste Gas Surge Tank.(ALM Manual 1C-22 Window 0-1.1).
(1) Shuts waste gas discharge CVs, 0-WGS-2191 and 2192; (2) Reduce the Waste Gas Surge Tank pressure to prevent lifting the Waste Gas Surge Tank relief valve. -- Is Incorrect, the actions to reduce Waste Gas Surge Tank pressure are associated with the Main Vent Gaseous monitor alarming ( 1-RI-5414) and only if the pressure in the Surge Tank is high. No indication were given that these conditions exist.
(1) Shuts waste gas discharge CVs, 0-WGS-2191 and 2192; (2)manually shut waste gas discharge header flow control valve, O-WGS-2191-PCV to prevent lifting the Waste Gas Discharge Relief Valve. -- Is correct since if the PCV is not shut the effect of the CVs auto closing is that pressure will build up in the header and possibly lift the Waste Gas Discharge Header Relief valve back to the Waste Gas Surge Tank which could cause an inadvertent release out of the stack.
(1) Shuts waste gas discharge header flow control valve, O-WGS-2191-PCV;(2) Reduce the Waste Gas Surge Tank pressure to prevent lifting the Waste Gas Surge Tank relief valve.-- Is Incorrect. since the auto action does not shut the PCV, also the actions to reduce Waste Gas Surge Tank pressure are associated with the Main Vent Gaseous monitor alarming ( 1-Rl-5414) and only if the pressure in the Waste Gas Surge Tank is high. No indication were given that these conditions exist.
(1) Shuts waste gas discharge header flow control valve, O-WGS-2191-PCV;(2) manually shut waste gas discharge CVs, 0-WGS-2191 and 2192 to prevent lifting the Waste Gas Discharge Relief Valve. Is Incorrect. since the auto action does not shut the PCV OPERATIONS Page: 11 of 51 11 June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam A, C, D are incorrect per alarm response manual actions for 1C22 D-1.1 OPERATIONS Page: 12 of 51 llJune2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Question 6 Info 3uestion Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points:
Time to Complete: 3 Difficulty: 2.00 50493 User-Defined ID: Q50493 LOI-077-1 Upon receiving a Gaseous Waste Discharge (0-RE-2191)
Topic:
high rad alarm, what automatic action occurs an RO Importance:
SRO Importance: 3.4 KA Number: 42060AA204 Comments: Modified from Bank question LOI-077-1:
FundamentaVMemory
References:
ALM manual for 1C22 pages 102-103 Basis: Gaseous Waste Discharge High Rad Alarm Automatic Action i
000060 Accidental AA2.04 - Ability to 3.
Gaseous Radwaste determine and interpret the 4 Release/9 following as they apply to the Accidental Gaseous Radwaste: The effects on the power plant of isolating a given radioactive gas leak WA Match analvsis - The ;/A addresses the effect on the power plant of isolating a given radioactive gas release.
The question focuses on an important aspect of this knowledge for CCNPP. A Gaseous discharge alarm occurred and to prevent additional contamination from the source, the waste gas discharge valves are shut to isolate the source. The effects of isolating the source is that unless the PCV is shut, the waste gas discharge header relief will lift and discharge to the waste gas surge tank which could cause an inadvertent release. The candidate must interpret from the information that when the waste gas discharge CVs go shut to isolate the source, this will affect the waste gas discharge header and pressure will build up unless he shuts the PCV.
OPERATIONS Page: 13 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 7 ID: Q50574 Points: 1.00 You are the CRS during the performance of EOP-0 on U-1. For which of the following radiation monitors in an alarm status would you expect Alternate Actions to be performed by the CRO?
A. Main Steamline B. Main Vent Gaseous C. Containment D. Wide Range Noble Gas Answer: C Answer Explanation:
EOP -0 Step G, for Verifying the radiation levels external to containment lists 4 monitors to check, Wide Range Noble Gas, Condenser Offgas, S/G BD, and Main Vent Gaseous.
of these the only ones that require alternate actions are the , Condenser off-gas or S/G Blowdown, which require securing blowdown. Under Step F for verifying the containment environment safety function, containment radiation monitor alarm is checked and if a valid containment radiation monitor is in alarm, then alternate actions are required to start all available iodine filter fans.
C. Containment is the correct answer A. Main steam line radiation monitor could be picked by a candidate who associates MS rad monitor with S/G blowdown as indication of S/G tube leakage.
B. Main Vent Gaseous could be selected by a candidate since it is checked during EOP-0 for verifying radiation levels external to containment C. Wide Range Noble Gas could be selected since it is checked during EOP-0 for verifying radiation levels external to containment OPERATIONS Page: 14 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam I Status: I Active I Comments:
NEW Question: Fundamentallmemory References EOP actions and basis associated with Containment Radiation Monitors 000061 ARM System 2.2.44 Ability to 4 Alarmsn interpret control room indications to verify the 4 status and operation of a system, and understand how operator actions and directives affect plant and svstem conditions.
OPERATIONS Page: 15 of 51 llJune2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 8 ID: Q50496 Points: 1.00 Which one of the following conditions would require the implementation of EOP-8?
A. Reactivity Control safety function cannot be met in EOP-0 due to no power available to CEA indications.
- 6. Reactor trips on S/G low level, and 11 125VDC bus voltage is reading 110 volts.
C. EOP-5 is implemented and the intermediate Safety Function Acceptance Criteria is not being met.
D. The EOP-0 flowchart recommends implementing both EOP-2 and EOP-6.
Answer: C Answer Explanation:
A. Reactivity Control-- is incorrect, EOP-0 accepts this condition for not meeting Reactivity. Candidate could be confusing this condition with the condition that would exist if you had power to the CEA indications and you could not verify rods are inserted.
B. Reactor trips on S/G low level, and 11 125VDC bus voltage is reading 110 volts- Is Incorrect. If you do not have all DC buses greater than 105 VOLTS then you will have to go to EOP-8. Candidate could have the misconception that all busses need to be at 125 VOLTS to call Vital Auxiliaries not met.
C. Intermediate acceptance criteria not being met.-- Is correct, EOP-8 would be implemented if the Intermediate Safety Function is not met.
D. Flow chart recommends 2 EOPs. Is Incorrect Optimal recovery procedure for EOP-6 includes the effect of a loss of Offsite power (EOP-2). Candidate could have the misconception that with a loss of offsite power and another event you need to go to EOP-8, however EOP-3, 4, 5, and 6 are designed to provide guidance for the event with a loss of offsite power.
References EOP-8 basis , Ill Entry Conditions), EOP-5 basis OPERAT1ONS Page: 16 of 51 llJune2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam
~~
Multiple Choice Status: Active NO No 1.oo
/Time to Complete: 2 Difficulty: 3.00 I Svstem ID:
Q50496 LOI-201-8 Which one of the followina conditions would require the implementation of EOP-8' 3.5 4.0 44E09EA22 Comments: Modified from bank question 25083 :
ComprehensivdAnalysis References :EOP-5 page 7, EOP-0 pages 19-23.
Lesson Pian LOI-201-8-8 Slides 23 -26 I CE/EO9 Functional EA2.2 - Ability to determine and interpret the following as they apply to the (Functional Recovery) (CFR:43.5 /
45.13): Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.
OPERATIONS Page: 17 of 51 11 June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 9 ID: (250432 Points: 1.00 Unit 2 is operating at 100% power when the following are noted:
0 226 RCP Seal TEMP HI PRESS Alarm 0 RCP CBO Temp HI Alarm 0 -
22B Middle Seal Pressure is 1050 PSlA 0 -
226 Upper Seal Pressure is 1040 PSlA 0 VCT Pressure is 50 PSlG 0 22B Seal Bleed-off temperature is 185°F 0 226 Seal bleed-off flow is Zero Which of the following directions should be given to the CRO and RO?
A. Commence an expeditious plant shutdown per OP-3, when the Rx is shutdown secure 22B RCP
- 6. Implement AOP-2A , trend the remaining seals, contact the System Manager immediately C. Trip Unit 2 Reactor. Implement reactivity portion of EOP-0 plaque, then secure 22B RCP, and complete EOP actions D. Trip Unit 2 Reactor, then immediately secure 22B RCP, and then complete reactivity portions of EOP-0 plaque.
Answer: 6 Answer Explanation:
- 6. Implement AOP-2A, contact the System Manager immediately is correct per ALM manual C. Trip the reactor is action for seal bleed off > 200°F A. Expeditious shutdown is for 2 failed seals, the Alm manual does not count the vapor seal as one of the 2 seals, even though with zero flow it appears that the vapor seal has failed.
D. If a trip is required then, the RX is tripped, reactivity is performed then the pump is tripped Reference Alarm Manual 2C06 - Window E-51 OPERATIONS Page: 18 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Multiple Choice I Status: Active Always select on test? NO Authorized for practice? No I Points:
Time to Complete:
Difficulty: 4.m I User-Defined ID:
Cross Reference Number:
Topic:
RO Importance:
SRO Importance:
KA Number: 34003A201 Comments:
References:
Alarm Manual 1C06 pages 96-99 003 Reactor A2.01 - Ability to (a) predict the Coolant Pump impacts of the following
_I malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations::
Problems with RCP seals, I especially rates of seal leak-Off.
I OPERATIONS Page: 19 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 10 ID: Q50404 Points: 1.00 A large break LOCA has occurred on Unit-2 and all RCPs have been tripped. The RO is attempting to verify subcooled natural circulation and reports the following:
Pressurizer Pressure is 150 PSlA being maintained by HPSl & LPSl flow RCS Subcooling based on CETs is 5°F Which one of the following set of conditions is the minimum needed to ensure adequate core cooling?
A. HPSl and LPSl flow appropriate for current RCS pressure AND Thot - 425°F B. HPSl and LPSl flow appropriate for current RCS pressure AND Thot - 405°F C. HPSl and LPSl flow appropriate for current RCS pressure AND Thot - 388°F D. HPSl and LPSl flow appropriate for current RCS pressure AND Thot - 360°F Answer: B Answer Explanation:
Need to recognize that with CETs at 5°F subcooling, subcooled natural circulation is not being met.
Per EOP-5 Block Step IV. N 2, for verifying subcooled natural circulation , if natural circulation subcooling is not being met, then need to ensure no more than 50" superheat to ensure adequate core cooling.
Since RCS pressure is 150 PSlA the minimum conditions for providing at less than 50°F superheat A. 425°F would not provide c 50°F B. HPSl and LPSl flow appropriate for current RCS pressure AND Thot - 405°F - Correct would give c 50°F superheat (Sat temp for 150 PSlA = 358.4"F)
C. 388°F would provide e 50°F but the question asked the minimum conditions to give <
50" superheat D. 360°F would provide e 50°F but the question asked the minimum conditions to give e 50" superheat OPERATIONS Page: 20 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Comprehensive/Analysis
Reference:
EOP-5, page 40,EOP-5 basis pages 49-52
'I 000011 Large EA2.09- Ability to determine or Break LOCN3 interpret the following as they apply to a Large Break LOCA(CFR 43.5 / 45.13) :
Existence of adequate natural circulation OPERATIONS Page: 21 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 11 ID:Q50450 Points: 1.00 U-1 is at 100% power with PT-102B out of service, the affected RPS bistables were bypassed.
During the shift PT-102D began to act erratically.
(a) What would be the effect on RPS immediately PT-102D fails high, and (b) What actions are required with PT-102D operating erratically?
A. (a) The reactor will trip (b) Declare PT-102D inoperable and bypass PT-102D
- 6. (a) The reactor would trip if PT-102 A or C were to fail high.
(b) Declare PT-102D inoperable and place PT-102D in a tripped condition C. (a) The reactor would trip if PT-102 A or C were to fail high (b) Declare PT-102D inoperable and bypass PT-102D D. (a) The reactor will trip (b) Commence an immediate controlled shutdown Answer: B Answer Explanation:
Need to understand that with one channel out of service, the RPS is in a two of three logic, when the second channel is tripped due to the failed sensor ( 1-PT-l02D) you are in a one of two logic to trip. Per technical specification 3.3.B 2 need to have one channel in bypass and one in trip.
A. (a.) The reactor will trip. (b.) Declare PT-102D inoperable and bypass PT-l02D.--ls Incorrect, since with one trip unit bypassed, when the second trip unit trips due to the sensor (1PI-l02D) failing high you still need one more of the remaining two in service bistables to trip to cause a reactor trip. Also, since with one trip unit already bypassed you are not allowed to bypass PT-102D bistables. You cannot have two bistables of the same parameter in bypass at the same time. This is prohibited by procedure and there is only one key available per trip parameter.
B. (a) The reactor would trip if PT-102 A or C were to fail high (b.) Declare PT-102D inoperable and place PT-102D in a tripped condition.-- Is Correct. since with one trip unit bypassed, when the second trip unit trips due to the sensor (1PI-102D) failing high its bistables will trip and you only need one more of the remaining two in service bistables to trip to cause a reactor trip. Technical specifications require that for this condition you place PT-102D in trip.
C. (a.) The reactor would trip if PT-102 A or C were to fail high. (b.) Declare PT-102D inoperable and bypass PT-102D.- Is Incorrect. You cannot have two bistables of the same parameter in bypass at the same time. This is prohibited by procedure and there is only one key available per trip parameter.
OPE RAT1ONS Page: 22 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam D. (a) The reactor will trip. (b) Commence an immediate controlled shutdown.-- Is incorrect, since with one trip unit bypassed, when the second trip unit trips due to the sensor ( 1PI-l02D) failing high you still need one more of the remaining two in service bistables to trip to cause a reactor trip. Technical specifications do not require a shutdown for this condition. Technical Specifications allow you to operate with one channel bypassed and one channel tripped for an indefinite time.
Question 11 info Question TvDe: Multide Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to Comdete: 3 Difficulty: 3.00 Svstem ID: 50450 User-Defined ID: Q50450 Cross Reference Number:
Topic: PT-102D was declared inoperable RO Importance: 3.4 SRO Importance: 3.7 KA Number: 37012A203 Comments: Modified from Q14450 : ComprehensivdAnalysis References : Technical Specification 3.3.1 -1 through 3.3.1 -
5 012 Reactor A.2.01 Ability to (a) predict the impacts I Protection of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulty Bistable operation (CFR: 41.5 / 43.5 / 45.3 /
45293.7) RO - 3,l , SRO -3.6 OPERATIONS Page: 23 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 12 ID: Q50451 Points: 1.00 U-1 is in Mode 6 with Reactor Vessel level at 43. ESFAS sensor cabinet ZF (1C94) is to be shutdown for cleaning and inspection. (a) Whose permission is required to deenergize Sensor cabinet ZF (1C94), (b.) will any action statement entry be required?
A. (a) GS, Shift Operations, (b) No, ESFAS operability is not required.
B. (a) Shift Manager, (b) No, ESFAS operability is not required.
C. (a) Shift Manager, (b) Yes, one UNIT 1 EDG will be inoperable D. (a) GS, Shift Operations, (b.)Yes, one Containment Radiation Sensor module is inoperable.
Answer: 6 Answer Explanation:
With the Rx Vsl level at 43 candidates should recall that fuel movement will NOT be occurring and therefore CRS is not required. 01-34 specifically requires the on shift SM permission to deenergize an ESFAS sensor cabinet. The GS Shift Operations could not grant this permission. Both A & D are plausible if the candidate has the misconception that a GS of Shift operations could grant this permission. Also they are plausible since ESFAS operability is required for Modes 1-3 only, but CRS is required for Mode 6 moving fuel A & D are incorrect per 01-34 appendix A step A.l requires SM approval C. ESFAS operability required in modes 1-3 per TS 3.3.4 and modes 1-4 for TS 3.3.6, DG- loss of voltage start.
B. Correct per 01-34 App a & TS 3.3.4 & 3.3.6 OPERATIONS Page: 24 of 51 11 June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Question 12 Info Question TvDe: Multiple Choice Status: Active Always select on test?
Authorized for practice?
Points:
Time to Complete:
Difficulty:
User-Defined ID:
CRO-63-1-3-54 Topic:
U-1 is in Mode 5 for a maintenance outage. ESFAS sensor cabinet ZF f1 RO Importance: 2.3 SRO Importance:
KA Number: LLL I Comments: Modified from (217459: ComprehensivdAnalysis
References:
Tech Spec 3.3.4-1 through 3.3.4-4, 3.3.6, And 01-34 Appendix A page 1 013 Engineered 2.2.21 Knowledge of pre- 4.1 Safety Features and post-maintenance Actuation operability requirements.
OPERATIONS Page: 25 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 13 ID: Q50453 Points: 1.00 Given a large steam rupture inside containment in each case, (a) which condition would result in the highest containment temperature? and (b.) which equipment would need to be operated to mitigate this consequence?
A. (a.) Inadvertent CIS Channel B actuates, (b.) Open IA-2080-MOV B. (a.) failure of SlAS Channel A to actuate, (b.) start one(1) Containment Spray Pump C. (a.) failure of CSAS Channel B to actuate, (b) Open one(1) Containment Spray cv D. (a.) failure of SGlS Channel A to actuate, (b.) Shut one(1) Main Feed Isolation Valve Answer: B Answer Explanation:
A. Inadvertent CIS channel B actuation-- Is incorrect. This would not appreciably change containment parameters. This is plausible since a candidate could have the misconception that with a loss of IA to containment the Containment Spray valves could fail shut and would need to have air supplied by opening IA-2080 MOV, containment isolation.
B. failure of SlAS Channel A to actuate-is correct. One train of containment coolers would fail to operate, and one train of containment spray would fail to operate (spray pump would not start).
C. Failure of CSAS channel B to actuate-- Is Incorrect. Would result in only a spray train failure (one spray valve would remain shut) but all containment coolers would operate.
D. Failure of SGlS channel A to operate-- Is incorrect, each channel operates all required equipment. Candidate could have the misconception that with only one channel actuating he could still be supplying inventory to the S/G which could blow down to the containment, but either channel will isolate both MSlVs and Feed Isolations.
OPERATIONS Page: 26 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam MultiDle Choice Active No No 1 .oo 5
3.00 50453 Q50453 Effect of ESFAS failure on containment cooling systems 4.3 4.7 32013K203 Modified from Q 20347 : Comprehensive/Analysis Used on 7/2002NRC exam,
References:
LD-58, EOP-5 basis page 32-35 026 A2.03 - Ability to (a) predict the 4.4 Containment impacts of the following Spray malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR:
41.5 143.5 145.3 145.13): Failure of ESF I
OPERATIONS Page: 27 of 51 11June2008
I I EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 14 ID:Q50455 Points: 1.00 Unit 2 is operating at 100% power during the month of August. Conditions are as follows:
0 23 SRW pump is out of service for maintenance 0 Outside Air Temperature ( 102°F) 0 Bay Temperature 82°F Alarm "TURB BLDG SRW HDR PRESS LO" has been intermittently alarming. Which of the following is an appropriate action to take for this alarm?
A. Determine that the alarm be "Blue Dotted" B. Determine that the alarm be "Black Dotted" C. Determine that the alarm be "Yellow Dotted" D. Determine that the alarm be "Red Dotted" Answer: A Answer Explanation:
Per CNG-OP-1.01 and ALM manual 1C-13 K22 possible cause, this alarm should be blue dotted.
Since the alarm is a nuisance alarm a blue dot would indicate that it is locked in.
B. Black dot would not be used since there is no maintenance being performed that causes this alarm.
D. Red dot is for a tagging activity C. Yellow dot is for one or more inputs out of service OPERATIONS Page: 28 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Question14 Info CIuestion Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.oo rime to Comdete: 3 Difficulty:
-,, 3.00 System ID: 50455 User-Defined ID:
Sross Reference Number:
SRO - Given conditions, indications, and/or parameter Topic:
values associated with inoperable SRW alarms, RO Importance: 3.0 SRO Importance: 3.3
.<A Number: 2243 Zomments: NEW Question: Comprehensive/Analysis Reference CNG-OP-1.01-2003 page 5 and Alarm Manual for C-13 page 39 076 Service Water 2.2.43 Knowledge of the 3.3 process used to track inoperable alarms.
OPERATIONS Page: 29 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 15 ID: Q50456 Points: 1.00 Using provided reference:
Unit 1 was operating at 100% power when a large Loss of Coolant Accident (LOCA) occurred.
EOP-5 has been implemented. Hydrogen concentration rose to 5% and the Hydrogen Recombiners were started. CNTMT TEMP prior to the event was 90°F. Two hours have passed since the Hydrogen Recombiners were started and now the following conditions exist:
H2 concentrations is now .8% and rising 11 Recombiner power setting is 50 KW 12 Recombiner is OFF Containment Pressure is 4.5 PSlG Which of the following is the correct action?
A. Set 11 Hydrogen Recombiner power setting to 57 KW B. Set 11 Hydrogen Recombiner power setting to 60 KW C. Set 11 Hydrogen Recombiner power setting to 63 KW D. Set 11 Hydrogen Recombiner power setting to 65 KW Answer: B Answer Explanation:
Per the graph of 01-41A with a Cntmt Press at the CSAS setpoint of 4.25 psig which gives a KW of 60.5 KW. Per the EOP-5 basis document within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of starting the recombiner it should be functioning, and one recombiner is designed to reduce H2 concentration faster than can be produced from a design basis accident, so if set properly then the H2 concentration should be lowering 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the recombiner was started.
The fact that H2 concentration has risen should indicate that the recombiner is not functioning properly.
A. Set 11 Hydrogen Recombiner power setting to 57 KW-- Is Incorrect for the conditions given this setting is to low.
B. Set 11 Hydrogen Recombiner power setting to 60 KW - Is correct. for the conditions given.
C. Set 11 Hydrogen Recombiner power setting to 63 KW-- Is incorrect for the conditions given, this setting is too high.
D. Set 11 Hydrogen Recombiner power setting to 65 KW-- IS incorrect for the conditions given this setting is to high B is correct for the conditions given A, C, D are incorrect since the KW does not match the containment conditions.
OPERATIONS Page: 30 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam auestion 15 Info 3uestion Type: Multiole Choice Status: Active 4lways select on test? No 4uthorized for Dractice? No D o ints 1 .oo rime to Complete: 3 2.00 System ID: 50456 Jser-Defined ID: Q50456 2ross Reference Number:
ropic:
indications associated with the Hydrogen recombiner 3 0 Importance: 3.4 SRO Importance:
.<A Number: 35028A201 2omments: NEW Question: ComprehensivdAnalysis Fieference 01-41 page 5 and Figure 1 -
028 Hydrogen
. - A2.01 -Ability to (a) predict the 3.6 Recombiner impacts of the following
- and Purge malfunctions or operations on the Control H R P S ; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: (CFR:
41.5 143.5 145.3 145.13)
Hydrogen recombiner power setting, determined by using plant data OPERATIONS Page: 31 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 16 ID: (250710 Points: 1.00 Using Provided references:
Given the following conditions:
0 Unit 1 is in Mode 6 0 Fuel Handling is in progress 0 Containment Purge is in operation to reduce noble gas activity below ALARA Dose Goals.
RI-5316A (ContainmentArea Monitor) has just been removed from service due to erratic operation. and has been placed in a tripped condition per Technical Specification 3.3.7A. You are the Operations Work Coordinator (OWC). Instrument Maintenance wants to trouble shoot the failed RI-5316A and has informed you that they will utilize peer checking to ensure they are working on RI-5316A only.
Which of the following is the correct Risk Assessment for this troubleshooting activity?
A. Medium Risk to Nuclear Safety AND Low Risk to Environmental Safety B. High Risk to Nuclear Safety AND Low Risk to Environmental Safety C. Medium Risk to Nuclear Safety, AND Medium Risk to Environmental Safety.
D. High Risk to Nuclear Safety AND Medium Risk to Environmental Safety Answer: C Answer Explanation:
Candidate should conclude based on the information in the stem of the question that this work is emergent work .His knowledge of the arrangement of the RMS panels informs him that channel A and B of RI-5316 are right next to each other and in the same drawer so there is a potential for incorrect performance which could lead to an ESF actuation (CRS) which would secure containment purge. He should then use attachments 11 and 2 of NO-1-1 17 to assess the risk. The note on Attachment 11 tells him that risk assessment applies to the shutdown unit, and step 1 of the attachment directs him to attachment 2 to determine the risk. From page 1 attachment 2 A1 he should determine that for the third bullet incorrect performance could lead to an ESF actuation and therefore leads him to section 2 for high risk assessment. On Page 7 of attachment 2 A l , he should determine based on information in the stem that verification practices are practical so the answers to this section is NO. and should lead him to Medium Risk to nuclear safety.
Environmental Risk is affected due to the potential for terminating a release associated with the containment purge.
Medium Risk to Nuclear Safety AND Low Risk to Environmental Safety -- Is Incorrect because it does not meet the criteria for attachment 2 (page 7) of NO-1-1 17 High Risk to Nuclear Safety AND Low Risk to Environmental Safety -- Is Incorrect because it does not meet the criteria for attachment 2 of NO-1-1 17 is related to electrical issues.(page 8)
High Risk to Nuclear Safety AND Medium Risk to Environmental Safety -- Is incorrect because it screens out as medium Risks , because the use of verification practices is practical.(page 7 item A l )
OPERATIONS Page: 32 of 51 11June2008
I I EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Medium Risk to Nuclear Safety, AND Medium Risk to Environmental Safety. - Is correct because verification practices are practical in this situation and this makes it screen out as Medium Nuclear Risk and since potential for terminating the containment purge, this screens out to Medium Risk for Environmental Safety.
Question16 Info guestion Type:
Status: Active Always select on test?
Authorized for practice?
Points: 1.00 rime to Complete:
Difficulty:
User-Defined ID: UdUI iv CRO-134-1/4.1 SRO - Apply appropriate conservative decision making Topic:
Practices for Area Radiation Monitorina RO ImDortance:
SRO Importance: 4.3 Ol Number: 2139 2omments: New Question: ComprehensivdAnalysis
References:
NO-1-1 17 attachment 2 and 11 Reference Provided to the candidate is NO-1-1 17 attachment 2 and attachment 11 IF APPLICANTS ASK, IT IS OKAY TO TELL THEM WE ARE NOT IN ACTION LEVEL ONE.
OPERATIONS Page: 33 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 17 ID: Q50459 Points: 1.00 Using provided reference(s):
A tornado on site results in a Loss of Offsite Power and damage to 11 CST. Both units were tripped at 1600 on 413. All other systems and equipment operated as expected.
Based on the following CST inventories taken at 1700, approximately how long can U-2 remain in Hot Standby?
11 CST - Punctured by tornado debris, no useable level 12 CST - 29 21 CST - 21.5 A. 0500 on 414 B. 0900 on 415 C. 1100 on 4/6 D. 2300 on 414 Answer: B Answer Explanation:
Modified from Q26566
Reference:
are EOP attachment 9 ( 5 pages)
The available CST volume available to U-2 is -308,000 gallons once calculated using the EOP Attachment (9) for U-2. Looking at the graph for Makeup water required to maintain Hot Standby, a time of -40 hours is determined, or 4/5 at 0900.
Answer A is incorrect since it utilizes the same input parameters but uses the U-1 EOP Attachment (9) to perform the calculation rather than the U-2 Attachments, resulting in a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance.
Answer C is incorrect if, during the calculation on the U-2 Attachment (9), the 12 CST volume is not divided by 2, yielding -66 hours.
Answer D is incorrect if, during the calculation on the U-1 Attachment (9), the 12 CST volume is not divided by 2, yielding -30 hours.
OPERATIONS Page: 34 of 51 llJune2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Question 17 Info Question TvPe: I MultiDle Choice I Status: I Active I Always select on test? 1 No Authorized for Practice? I No Topic: Time to remain in hot standby, 12 CST unavailable RO Importance: 2.8 SRO Importance: 3.1 KA Number: 2125 Comments: ComprehensivdAnalysis 2.1. Ability to interpret reference materials, such as 4.2 25 graphs, curves, tables, etc.
I OPERATIONS Page: 35 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 18 ID: (250462 Points: 1.00 During movement of irradiated fuel, a 30 gprn leak in the refueling cavity hatch seal requires that fuel handling cease. The refueling machine operator has just removed a fuel assembly from the core and is indexed over the upender. The transfer carriage is in the refueling cavity and vertical.
What actions should the FHS direct?
A. Immediately secure the refueling machine and evacuate containment.
B. Send the carriage to the spent fuel pool, place the fuel assembly in any empty core location and shut the transfer tube gate.
C. Place the fuel assembly in the deep pocket, beside the upender and shut the transfer tube gate valve.
D. Place the fuel assembly in the upender carriage, send it to the spent fuel pool and shut the transfer tube gate valve.
Answer: D Answer Explanation:
Immediately secure the refueling machine and evacuate containment. -- Is Incorrect.
Because the fuel assembly is not in a defined " safe location" (Inserted in core, Spent Fuel Pool, Upender, or grappled in either the bottom of the SFP upender area or bottom of the Spent Fuel Pool Upender area), securing the refueling machine is not correct. AOP
-6E specifically directs that the fuel assembly be placed in a safe location prior to securing power to the refueling machine.
Send the carriage to the spent fuel pool, place the fuel assembly in any empty core location and shut the transfer tube gate.-- Is Incorrect, per AOP 6E, since the fuel assembly was fully withdrawn, it must be placed in an approved core or SFP storage location. Since Nuclear Fuel Management has not designated all empty locations as approved this is not correct. If this had stated return to the previous core location this would have been acceptable per AOP 6E.
Place the fuel assembly in the deep pocket, beside the upender and shut the transfer tube gate valve.-- Is Incorrect, per AOP6E , the transfer tube gate valve cannot be shut until the transfer carriage is in the SFP. Since this has not been done this is not an acceptable answer.
Place the fuel assembly in the upender carriage, send it to the spent fuel pool and shut the transfer tube gate valve. - Is correct per AOP 6E. for the conditions stated the fuel assembly must be placed in upender, approved core location, or bottom of the RFP upender area grappled per AOP6E IV A.2.e OPERATIONS Page: 36 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Topic: Action for stopping core alterations RO Importance: 3.0 SRO Importance: 4.1 KA Number: 2136 Comments: Bank question Q14516 : Comprehensive/Analysis I I Reference AOP 6E pages 8-11 2.1. Knowledge of procedures and limitations 4.1 36 involved in core alterations.
OPERATIONS Page: 37 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 19 ID: Q45548R Points: 1.00 Using Provided references:
Unit 1 is in Mode 1. An MO was generated to work a deficiency on 13 AFW pump. The deficiency is a clouded oil sight glass. The pump is still operable. It cannot be worked by FIN and must be planned and scheduled. Operations is concerned about the deficiency and it has operator compensatory actions associated to it, however it requires the13 AFW pump to be declared inoperable and removed from service to perform the work. What priority and schedule should this MO be?
A. 4 -White, Schedule as resources allow with the normal process
- 6. 2- Yellow, Schedule at earliest opportunity within T-3 C. 3 - Orange, schedule at next available system week with 13 week matrix D. 1 - Red, Begin immediately and work around the clock Answer: C Answer Explanation:
C is correct Per CNG-MN-4.01 -1002, attachment 1, Online prioritization matrix , Since this is elective maintenance Orange, schedule at next available system week with 13 week matrix is correct.
A, B, D are incorrect per CNG-MN-4.01 Att 1 and the fact that this maintenance is not corrective maintenance per definition of corrective maintenance. Key is AFW pump is still operable and is stated as such in the stem.
OPERATIONS Page: 38 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Westion 19 Info 3uestion Type:
I Multiple Choice Status: Active Always select on test? No Authorized for Dractice? No
%ink 1.oo rime to Complete:
2.00 System ID:
User-Defined ID:
Zross Reference Number:
Topic:
RO Importance:
SRO Importance:
KA Number: 2218 Comments: NEW Question : Comprehensive/Analysis Reference CNG-MN-4.01 attachment 1.
Note: Attachment one should be provided 2.2. Knowledge of the process for managing 3.9 18 maintenance activities during shutdown
]operations, such as risk assessments, work prioritization, etc.
OPERATIONS Page: 39 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 20 ID: (250655 Points: 1.00 Given the following conditions:
-- SG blowdown tank radiation monitor (RI-4014) is in alarm
-- SG tube rupture is in progress on the effected SG and level is
+30 inches and steady
-- Upon the reactor trip a loss of main condenser vacuum occurred
-- SG level on unaffected SG reached -180 inches in EOP-0 and is presently at -100 inches and slowly returning to 0 inches It is desired to establish SG blowdown to restore the affected SG level to 0 inches. Which of the following is a correct action?
Assume all admin and radiological controls have been satisfied A. Direct the CRO to place the alarm in cutout to reopen the blowdown isolation cvs.
- 6. Direct the CRO to deenergize SG blowdown tank rad monitor, R1-4014, to reopen the blowdown isolation CVs.
C. Direct the CRO to bypass RIC-4014 using the Alarm Bypass Key Lock Switches.
D. Direct the CRO to place the blowdown isolation CVs in RAD TRIP OVERRIDE.
Answer: D Answer Explanation:
Direct the CRO to place the alarm in cutout to reopen the blowdown isolation CVs.- Is Incorrect. The alarm cutout pushbutton on 1C22 will silence the alarm and allow the master RMS alarm to clear but will not allow the CVs to open.
Direct the CRO to deenergize SG blowdown tank rad monitor, R1-4014, to reopen the blowdown isolation CVs.-- Is Incorrect. This would cause the blowdown CVs to go shut.
Direct the CRO to bypass RIC-4014 using the Alarm Bypass Key Lock Switches--
Incorrect RI-4014 does not have Alarm Bypass Key Lock Switches, RIC-4095 (
Blowdown Recovery Monitor) has Alarm Bypass Key Lock Switches.
Direct the CRO to place the blowdown isolation CVs in RAD TRIP OVERRIDE.-- IS Correct, the RAD TRIP OVERRIDE H/S will allow you to open the S/G bottom blowdown CVs and place blowdown in service.
This question requires knowledge of the RMS system for S/G blowdown and blowdown recovery including the auto actions associated with RI-4014 and 4095 and the means provided for controlling S/G level with an RMS alarm.
OPERATIONS Page: 40 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam hestion 20 Info hestion Type: Multiple Choice Status: Active Slways select on test? No Suthorized for practice? NO
'oints: 1.oo rime to Complete: 3 Iifficulty: 3.00 System ID: 50655 Jser-Defined ID: QFi0655 2ross Reference Number: Q24660 Establish/maintain SG blowdown flow when RI-4014 in ropic: alarm 3 0 Importance:
SRO Importance:
W Number: 020170420 2omments: Modified from bank question Q24660 :
Comprehensive/Analysis 11 K;ledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring 1
Basis: SG Blowdown Tank Radiation Monitor in Alarm
References:
- ". 3.4 e ui ment, etc.
References:
01-35 Radiation Monitoring System Section 6.1 1 for RIC-4014 and 6.18 for Blowdown Recovery Radiation Monitor (
l(2)-4095)
EOP -6 Section 0.1 for maintaining the affected S/Level during a SGTR Alarm Manual for !C-22 Window A-4.2 RI -4014 and Window D-5.1 Rl-4095 Drawings 1E-83 Sh.3 and Sh 16 OPERATIONS Page: 41 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 21 ID:(250653 Points: 1.00 Unit 2 experienced a Reactor Trip and EOP-0 has been implemented. After all safety function have been assessed the containment normal sump alarm annunciates. Which actions are correct per NO-1 -201 ?
A. Drain the containment norrnal sump to establish a trend for RCS leak rate, then implement the optimal EOP.
B. Briefly evaluate the Containment environment for degrading conditions then implement the optimal EOP.
C. Implement the optimal EOP and then evaluate containment environment for degrading conditions.
D. Implement the optimal EOP and then drain the containment normal sump to establish a trend for RCS leak rate.
Answer: B Answer Explanation:
Drain the containment normal sump to establish a trend for RCS leak rate, then implement the optimal E0P.-- Is Incorrect, Per NO-1 -201 the containment environment must be evaluated for this alarm prior to implementing the optimal EOP Briefly evaluate the containment environment for degrading conditions then implement the correct optimal E0P.-- Is correct per NO-1 -201 Implement the optimal EOP and then evaluate containment environment.-- Is Incorrect, prior to going to the optimal EOP the containment environment must be evaluated since the alarm came in after the CE safety function was evaluated.
Implement the optimal EOP and then drain the containment normal sump. Is Incorrect, prior to going to the optimal EOP the containment environment must be evaluated since the alarm came in after the CE safety function was evaluated.
OPERATIONS Page: 42 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Question 21 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to Complete: 0 Difficulty: 2.00 System ID:
User-Defined ID:
Cross Reference Number:
Topic:
RO Importance:
SRO Importance:
KA Number:
Comments:
2.4.5 Knowledge of the organization of the 4.3 operating procedures network for normal, abnormal, and emergency evolutions. (CFR:
41.10 / 43.5 / 45.13)
Question 21 Table-Item Links A / 0 Trainina Proqram Licensed Operator Requal Training (LOR)
Operations Procedure References (from Nucleisl EOP EOP-04-1 EXCESS STEAM DEMAND EVENT System Desianations Emergency Operating Procedures (EOPs)
OPERATIONS Page: 43 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 22 ID: Q50470 Points: 1.00 A fault has occurred on P-13000-2, resulting in a U-2 reactor trip. During EOP-0, the 2 8 DG experienced a start failure. Given the following additional plant conditions, what Optimal Recovery Procedure should be implemented?
Boration in progress with no CEA indication RCS pressure 1900 psia and lowering Pzr Level 130 inches and slowly lowering RCS Tcold 518°F and lowering RCS Subcooling is 107 O F 21 SG Pressure 835 psia and steady 22 SG Pressure 800 psia and lowering 21 SG Level (-)lo5 inches and slowly rising 22 SG Level (-)180inches and lowering Containment Environment and Rad Levels External to Containment are met A. EOP-6, Steam Generator Tube Rupture B. EOP-2. Loss of Forced Circulation C. EOP-4. Excess Steam Demand D. EOP-5, Loss of Coolant Accident Answer: C Answer Explanation:
Based on EOP-0 diagnostic flow chart, Core and RCS Heat Removal would not be met, and subcooled would be high, along with 22 S/G pressure lowering and 22 S/G level lowering. This would lead to the conclusion that an excess steam demand event is in progress.
A, B, D incorrect per EOP-0 flowchart A incorrect, Radiation Levels External to Containment Safety Function is met, no RMS trends or alarms so no indication of SGTR B EOP-2, Loss of Forced Circulation is incorrect, based on indications there is more than a loss of power occurring D. incorrect since subcooling is too high to be consistent with a LOCA OPERATIONS Page: 44 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Question22 Info auestion Type:
Status: Active Always select on test? No Authorized for practice? No Points: 1.oo Time to Complete:
Difficulty:
System ID: 50470 User-Defined ID: Q50470 Cross Reference Number:
Topic:
RO Importance:
SRO ImDortance:
KA Number:
Somments: Modified from 245 : ComprehensivdAnalysis References : EOP-0 pages 19-23 Provide diagnostic flow chart assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release OPERATIONS Page: 45 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 23 ID: Q50553 Points: 1.00 For each of the following maintenance activities identify the category of post maintenance testing required
- 1. 11 Condensate Pump coupling replacement
- 2. 13 Charging pump repacking
- 3. 11 Demineralized water transfer pump motor maintenance A. 1. Post Maintenance Functional Test
- 2. Post Maintenance Operability Test
- 3. Rotational Check B. 1. Rotational Check
- 2. Post Maintenance Operability Test
- 3. Post Maintenance Functional Test C. 1. Rotational Check
- 2. Post Maintenance Operability Test
- 3. Rotational Check D. 1. Post Maintenance Functional Test
- 2. Post Maintenance Operability Test
- 3. Post Maintenance Functional Test Answer: A Answer Explanation:
- 1. Post Maintenance Functional Test
- 2. Post Maintenance Operability Test
- 3. Rotational Check correct per MN-1-201 , and NO-1-208 B & C incorrect, per NO-1-208 5.6.B.1 post maintenance functional test may be required for equipment that is trip sensitive and PP is trip sensitive D is incorrect 11 DI xfr pump is not trip sensitive OPERATIONS Page: 46 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam written for WA 2.2.1 4 Reference NO-1-208 section 5.6 page 1, and pages 6, 8-9 2.2. Knowledge of the process for controlling 14 equipment configuration or status.
OPERATIONS Page: 47 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 24 ID: Q20602 Points: 1.00 Unit 2 is in Mode 6 with refueling in progress and Normal Containment Purge in service. The Equipment hatch is installed and the Personnel Airlock (PAL) is open. A momentary loss of power causes the operating Main Exhaust Fan to trip.
(a) What is the effect on containment parameters, (b) What is the correct action?
A. (a) Containment refueling pool level decreases, (b) Continue refueling operations.
B. (a) Containment pressure rises 1 to 2 PSIG, (b) Initiate additional containment cooling C. (a) Containment area radiation monitors (RE-5316-A through -D) indicate higher, (b) Start all available Iodine Filter Units D. (a) Containment refueling pool level increases, (b) Continue refueling operations.
Answer: A Answer Explanation:
A. Containment pressure rises 1 to 2 PSIG--incorrect, containment pressure will change, but experience indicates, the change will be less than .5 PSIG.
B. Area radiation monitors (RE-5316A-D) indicate higher-incorrect, the area monitors would not change if Purge is lost.
C. Refueling pool level increases-Incorrect The Main Exhaust Fan tripping would cause Containment Purge to secure. This would cause.containmentpressure to rise slightly, with the transfer tube gate valve open, refueling pool level will decrease (Not Increase) accordingly due to the differential pressure between the SFP area and containment.(SFP is maintained at a slight negative pressure)
D. Refueling pool level increases--1s Correct. The Main Exhaust Fan tripping would cause Containment Purge to secure. This would cause containment pressure to rise slightly, with the transfer tube gate valve open, refueling pool level will decrease accordingly due to the differential pressure between the SFP area and containment. (SFP is maintained at a slight negative pressure)Continue refueling operations since no loss of RFP level.
per 01-36 general precaution F. The Main Exhaust Fan tripping would cause Containment Purge to secure which would cause a change in the differential pressure between the SFP and the RFP OPERATIONS Page: 48 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Status: Active Always select on test? No Authorized for practice? No 1 Points:
rime to Complete:
User-Defined ID:
Topic:
RO Importance:
SRO ImDortance:
KA Number:
Somments:
Comprehensive/Analysis
References:
01-36, page 5 029 A2.01-Ability to (a) predict the Containment impacts of the following Purge malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR:
41.5 143.5 145.3 I 45.13)Maintenance or other activity taking place inside containment OPERATIONS Page: 49 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam 25 ID: Q20588 Points: 1.00 Unit-2 is operating at 100% power with:
2-LT-11OX, PZR Level transmitter selected 0 2-HS-100-3, PZR Heater Low Level Cutout Switch in the X+Y position If a leak on 2-LT-11OX variable leg developed, what is one of the operator actions that must be directed by the CRS to restore pressurizer control system functions to automatic?
A. Shut Letdown Stop Valves, 2CVC-515 and 516
- 6. Place 2-HS-100-3, PZR Heater Low Level Cutout Switch to channel Y C. Adjust Pressurizer Level Controller, 2-LIC-11OX, setpoint to shut the letdown control valve D. Place Pressurizer Level Controller, 2-LIC-11OY, in AUTO-LOCAL Answer: B Answer Explanation:
Place 2-HS-100-3, PZR Heater Low Level Cutout Switch to channel Y--correct per Alarm Response Manual.
Shut Letdown Stop Valves, 2CVC-515 and 516--incorrect, this may be necessary for a large leak, but would not allow automatic level control.
Adjust Pressurizer Level Controller, 2-LIC-11OX, setpoint to shut the letdown control valve--incorrect, not supported by procedures and would take continuous operator monitoring to maintain level.
Place Pressurizer Level Controller, 2-LIC-11OY, in AUTO-LOCAL--incorrect, the remote setpoint function is still available, the proper action would be to place LIC-110-Y in service OPERATIONS Page: 50 of 51 11June2008
EXAMINATION ANSWER KEY CCNPP 2008 NRC SRO Exam Question 25 Info Question Type: Multiple Choice Status: Active Always select on test?
Authorized for practice? No Points: 1.oo Time to Complete:
Difficulty: 3.00 System ID: 50670 User-Defined ID: Q20588 Cross Reference Number:
Topic: Identify effects and actions for a LT-110 X variable leg leak
~
RO Importance:
SRO Importance:
KA Number: 028AA203 Comments: Bank Question Used 7/2002 NRC exam :
Comprehensive/Analysis
References:
55.41.7. 55.43.5 OPERATIONS Page: 51 of 51 11June2008