Text

Initial Exam 2010-301 Draft SRO Written Exam
	ML110140235
Person / Time
Site:	Watts Bar
Issue date:	12/10/2010
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
References
	50-390/10-301
	Download: ML110140235 (69)
	v • d • e

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

76. 022 AA2.04 076 Given the following events and conditions:

Unit 1 is operating at 100% power when FCV-62-93, Charging Flow Control, malfunctioned to limit charging flow to 65 gpm.

The plant is at normal operating temperature, pressure and level.

Normal letdown is in service at 75 gpm.

Identified leakage is at the Tech Spec Limit.

Unidentified leakage is 0.01 gal/mm.

RCP seal leak-off is 3 gpm per pump.

Which ONE of the following identifies...

(1) If no operator actions are taken, approximately how much time will elapse before annunciator window 92-C PZR LEVEL LO-HTRS OFF & LTDN CLOSED alarms and (2) an effect resulting from the level drop in accordance with Tech Spec Basis?

REFERENCE PROVIDED TIME RESULT A. approx. 82 minutes Cannot assure the DNBR criterion limits for the RCS parameters will be maintained following a reactor trip coincident with a loss of offsite power.

B approx. 82 minutes Cannot ensure the ability to maintain the RCS in a hot pressurized condition with loop subcooling for an extended period following a reactor trip coincident with a loss of offsite power.

C. approx. 115 minutes Cannot assure the DNBR criterion limits for the RCS parameters will be maintained following a reactor trip coincident with a loss of offsite power.

D. approx. 115 minutes Cannot ensure the ability to maintain the RCS in a hot pressurized condition with loop subcooling for an extended period following a reactor trip coincident with a loss of offsite power.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A.

lncorrect Plausible because 82 minutes to reach the alarm (17%) is correct and assuring DNBR criterion limits for RCS parameters can be maintained is wording from the DNB LCO bases not the pressurizer LCO bases.

B.

Correct, It would take approximately 82 minutes to lower the level to the alarm setpoint and the result would challenge the ability to maintain the RCS in a hot pressurized condition with loop subcooling for an extended period following a reactor trip coincident with a loss of offsite power. (See calcuation below)

C.

lncorrect, Plausible because 115 minutes would be calculated if I gpm was used instead of 10 gpm for the identified leakage limit and and assuring DNBR criterion limits for RCS parameters can be maintained is wording from the DNB LCO bases not the pressurizer LCO bases.

D.

lncorrect Plausible because 115 minutes would be calculated if I gpm was used instead of 10 gpm for the identified leakage limit and the result being to challenge the ability to maintain the RCS in a hot pressurized condition with loop subcooling for an extended period following a reactor trip coincident with a loss of offsite power is correct.

PZR level @ 100% power = 60%

Letdown isolates @ 17%

60- 17 = 43% x approx 62 gal/% = 2634 gallons Delta charging & letdown = 10 gpm Identified leakage = 10 gpm Unidentified leakage = 0.Olgpm (negligible)

RCP seal leakoff = 4x3 12 gpm into VCT Total flow out of RCS= 10+10+0+ 12=32gpm 2634/32 = 82.3 minutes If identified leakage rate of I gpm used instead of the correct value of 10 gpm.

Total flow out of RCS= 10+1+0+ 12=23gpm 2634/23 = 114.5 minutes Question Number:

76 Tier:

I Group 1

KIA:

022 AA2.04 Loss of Reactor Coolant Makeup Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup:

How long PZR level can be maintained within limits Importance Rating:

2.9 I 3.8 10 CFR Part 55:

43.5/ 45.13

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 IOCFR55.43.b:

2 K/A Match:

KA is matched because the question requires knowledge of how long the PZR level can be maintained within limits with an ongoing malfunction of Reactor Coolant Makeup and is SRO because it requires knowledge of Tech Spec Bases.

Technical

Reference:

Tech Spec 3.4.9 Bases ARI-8894, Reactor Coolant System, Revision 19 1-Sl-68-32, Reactor Coolant System Water Inventory Balance, Rev 0013 Proposed references 1-Sl-6832, Rev 0013, page 26 to be provided:

Learning Objective:

3-OT-SYSO68C

12. Identify the program setpoints, and describe any automatic actions relative to the pressurizer level program 3-OT-T-50304 2.

Determine the bases for each specification, as applicable, to the RCS.

Cognitive Level:

Higher X

Lower Question Source:

New Modified Bank Bank X

Question History:

SQN Bank question with minor changes and numbers in 2 distractors changed, but stem changes do not meet modification criteria.

Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

77. 025 AA2.03 077 Unit I is in Mode 4 with Train A RHR in service during unit cooldown when the following occurs:

1405 Annunciator 160-C, RX BLDG POCKET SUMP LEVEL HI, alarms.

1407 Annunciator 159-C, RX BLDG F&EQ SUMP LEVEL HI, alarms.

1407 OAC notes pressurizer level is dropping.

Which ONE of the following identifies a correct sequence of action and procedure use when both of the AOls listed below are used in response to the alarms and level loss?

AOl-6, Small Reactor Coolant System Leak AOl-I 4, Loss of RHR Shutdown Cooling A If AOl-6 is entered the AOl will direct raising charging flow; then direct a transition to AOl-i 4.

B.

If AOl-6 is entered the AOl will direct a transition to AOl-i 4 which will then direct the increase in charging flow.

C.

If AOl-14 is entered the AOl will direct raising charging flow; then direct a transition to AOI-6.

D.

If AOl-14 is entered the AOl will direct a transition to AOl-6 which will then direct the increase in charging flow.

DISTRA CTOR ANAL YSIS:

A.

Correct, ARI-160-C states IF RCS leak is suspected THEN REFER TO AOl-6, Small Reactor Coolant System Lealç OR A01-14, Loss Of RHR Shutdown Cooling.

IfAOl-6 is entered, the AOl will direct charging flow to be increased in Section 3.0 step 3 and then direct a transition to A 01-14 in step 4.

B.

Incorrect, Plausible because AOl-6 does direct a transition to A 01-14 and A01-14 does direct the charging flow be raised to maintain RCS level, but it is not the first direction to increase the charging flow.

C.

Incorrect, Plausible because the A0l-6 and AOl-14 are the AOls that are identified in ARI-160-C and both direct increasing charging flow, but AOl-6 would send the crew to AOI-14, not AOl-14 sending the crew to AOI-6.

D.

Incorrect, Plausible because the AOl-6 and AOl-14 are the AOls that are identified in ARI-160-C and ffAOI-14 was entered first, it would direct increasing charging flow, but AOl-14 would not send the crew to A 01-6.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Question Number:

77 Tier:

I Group 1

K/A:

025 AA2.03 Loss of Residual Heat Removal System (RHRS)

Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System:

Increasing reactor building sump level Importance Rating:

3.6 I 3.8 10 CFR Part 55:

43.5 I 45.13 IOCFR55.43.b:

5 KIA Match:

KA is matched because the question requires the ability to interpret actions required for increasing rector building sump levels while on RHR cooling. SRO because it requires the knowledge of procedure flow paths based on the selection of procedures and actions contained within the procedure being used.

Technical

Reference:

1 -47W851 -1 R26 ARI-1 59-1 65, Sumps & COW, Revision 34, windows 1600 and 159C AOI-6, Small Reactor Coolant System Leak, Revision 32 AOI-14, Loss of RHR Shutdown Cooling, Revision 0036 Proposed references None to be provided:

Learning Objective:

3-OT-A0l1400 7.

Demonstrate ability/knowledge of AOl, to correctly:

a.

Recognize Entry conditions.

3-OT-A0l0600 3.

Determine Initial Operator Action(s) for a small RCS leak.

Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRQ NRC Exam as Submitted 7/2/2010

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

78. 027AA2.09 078 The operating crew assumes shift on Unit I with the following conditions:

Unit 1 is operating at 8% power.

Train B SSPS is out of service for surveillance testing.

The Block Valve for pressurizer PORV 340A was closed on the previous shift due to seat leakage on the PORV and the appropriate Tech Spec condition entered.

During the shift the following occurs:

Pressurizer PORV 334 fails open and pressurizer pressure starts dropping.

Operator closes the Block valve for pressurizer PORV 334 because the PORV failed to close when its handswitch was placed to close.

Pressurizer pressure begins to increase after dropping to a low of 1940 psig.

The operating crew stabilizes the unit at 8% power.

The appropriate Tech Spec condition entered due to PORV 334 failure.

Which ONE of the following identifies...

(1) if conditions indicate a failure of SSPS Train A during the transient and (2) the PORV failure that will result in a Tech Spec required shutdown unless the valve operability can be restored within a specificied time period?

(1)

(2)

SSPS Train A failure...

PORV A. did occur.

PORV 334 B. did occur.

PORV 340A C did NOT occur.

PORV 334 D. did NOT occur.

PORV 340A

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANALYSIS:

A.

lncorrect, Plausible because SSPS Train A failure occurring would be correct if the power (turbine (P 13) or reactor (PlO) had been greater than 10% (P7) and the condition of PORV 334 requiring a unit shutdown is correct.

B.

Incorrect, Plausible because SSPS Train A failure occurring would be correct if the power (turbine (P13) or reactor (P10) had been greater than 10% (P7) and the condition of both POR Vs require LCO 3.4.11 conditions to be entered but the requirement for a shutdown would be from the status of PORV 334, not the status of PORV 340.

C.

Correct, With the reactor power at 6%, the low pressurizer pressure trip is not in service so there would not have been a failure of the SSPS Train A. T/S LCO 3.4.11 addresses the condition of the PORVs. Condition A for PORV 340A and Condition B for PORV 334. Condition B requires the PORV to be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or condition D applies and the unit must be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Mode 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D.

Incorrect, Plausible because SSPS Train A failure not occurring is correct and the condition of both PORVs require LCO 3.4.11 conditions to be entered but the requirement for a shutdown would be from the status of PORV 334, not the status of PORV 340.

Question Number:

78 Tier:

I Group 1

KIA:

027 AA2.09 Pressurizer Pressure Control System (PZR PCS) Malfunction Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions:

Reactor power Importance Rating:

3.5 /3.6 10 CFR Part 55:

43.5/45.13 10CFR55A3b:

2 KIA Match:

KA is matched because the question requires the ability to determine if a plant reactor protection system functioned as designed during a pressurizer pressure control system malfunction while the reactor was operating at 8% power (Ability to determine how reactor power applies to the pressure control system failure.) SRO only because if question requires the ability to apply Tech Spec actions (using information below the line)

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Technical

Reference:

Tech Spec 3.4.11, Amendment 55 Tech Spec 3.4.11 Bases, Revision 68 1-47W611-99-2 R12 Proposed references None to be provided:

Learning Objective:

3-OT-SYSO99A 17 Identify the Reactor trips and give setpoints and list logic required for the Reactor trips.

3-OT-T/S0304 4.

Given plant conditions and parameters correctly determine the applicable Limiting Conditions for Operations or Technical Requirements for the various components of the RCS.

Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

79. 056 AG2.4.41 079 Given the following:

Unit us operating at 100% power.

Both 161 kV offsite power supplies are lost due to a storm causing damage to the Watts Bar Hydro switchyard.

All 4 diesel generators restore power to their respective Shutdown Boards.

If neither offsite power supply can be restored within the required time, which ONE of the following identifies...

(1) the required REP classification that must be declared and (2) if the Operations Duty Specialist (ODS) can NOT be reached after the event is declared, how long before the Tennessee Emergency Management Agency (TEMA) would be called directly?

Classification Time A.

NOUE 5 minutes B

NOUE 10 minutes C.

ALERT 5 minutes D.

ALERT 10 minutes

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANALYSIS:

A.

Incorrect, Plausible because an NOUE is the correct classification if each of the DG were supplying power to its respective shutdown board classification and 5 minutes is the normal (should be) time frame for contacting the ODS.

B.

Correct; With both offsite power supplies lost (CSSTs C and D not available) and the DG restore all shutdown boards, the classification is an NOUE and in accordance with EPIP-2, Notification of Unusual Event, IF the ODS CANNOT be contacted within 10 minutes, THEN 1) NOTIFY the Tennessee Emergency Management Agency (TEMA) of the Radiological Emergency Plan activation by calling 9-1-800-262-3300 or 9-1-615-741-0001 or 9-1-800-262-3400.

C.

Incorrect, Plausible because an ALERT would be the correct classification if either Unit I DG failed and 5 minutes is the normal (should be) time frame for contacting the ODS.

D.

Incorrect, Plausible because an ALERT would be the correct classification if either Unit I DG failed and if the ODS can not be reached after 10 minutes, then TEMA being called directly by the SED 10 minutes is correct.

Question Number:

79 Tier:

1 Group 1

K/A:

056 AG2.4.41 Loss of Offsite Power Knowledge of the emergency action level thresholds and classifications.

Importance Rating:

2.9 / 4.6 IOCFRPart55:

41.10/43.5/45.11 IOCFR55.43.b:

5 K/A Match:

KA is matched because the question requires knowledge of conditions the meet emergency action level thresholds during a loss of offsite power event. SROs only both because of the determination of the applicable classification level and because of the knowledge of the time allowed to ensure a proper required notification to the State is made when normal communication methods fail.

Technical

Reference:

EPIP-1, Emergency Classification Matrix, Revision 33 EPIP-2, Notification of Unsual Event, Revision 28 NP-REP Appendix C, Tennessee Valley Authority Nuclear Power Radiological Emergency Plan, Revision 91

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Proposed references None to be provided:

Learning Objective:

3-OT-PCD-048C 1.

Classify emergency events.

26. Understand the critical times associated with Event Declaration Offsite Notification Facility Staffing Printed EPS Report Cognitive Level:

Higher X

Lower Question Source:

New Modified Bank X

Bank Question History:

WBN bank questions 056 G2.4.43 003 and G2.4.42 combined and modified.

Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

80. 065 G.2.4.21 080 Given the following:

A reactor trip occurs on Unit 1 as a result of the loss of all control and essential air.

The operating crew has transitioned to ES-0.1, Reactor Trip Response.

With NO operator action, which ONE of the following identifies Safety Functions that will become less than satisfied (something other than Green)?

A Inventory and Heat Sink B. Inventory and Containment C. Pressurized Thermal Shock and Heat Sink D. Pressurized Thermal Shock and Containment DISTRACTOR ANAL YSIS:

A.

Correct, both the Inventory and the Heat Sink Critical Safety Functions will become less than met if no operator actions are taken. The pressurizer level will rise creating a yellow path due to high level and Heat Sink challenged by low steam generator levels due to the failures in the AFW system. (TDAFW pump LCVs failing closed and while the MDAFW LCVs fall open, the PCVs on the discharge of the pumps falls closed.)

B.

Incorrect, Plausible because the Inventory Critical Safety Functions becoming less than met is correct and many of the valves/dampers associated with containment are air operated devices.

C.

lncorrect, Plausible because concluding that the Pressurized Thermal Shock Critical Safety Function would be less than met would be the result if the affect on AFW were reversed and Heat Sink becoming less than met in correct.

D.

Incorrect Plausible because concluding that the Pressurized Thermal Shock Critical Safety Function would be less than met would be the result if the affect on AFW were reversed and overcooling occurred and many of the valves/dampers associated with containment are air operated devices.

Question Number:

80 Tier:

I Group 1

K/A:

065 G.2.4.21 Loss of Instrument Air

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 K/A:

065 G.2.4.21 Loss of Instrument Air Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Importance Rating:

4.0 I 4.6 IOCFRPart55:

41.7/43.5/45.12 IOCFR55.43b:

5 KIA Match:

The KA is matched and the question is SRO only because the question requires knowledge of the parameters and logic used to assess the status the Critical Safety Functions and to determine the appropriate procedures that would to be directed by the Status Trees to address challenges to the CSFs created by a loss of air event.

Note: At Watts Bar the Instrument Air System and the Station Air System is actually one combined system supplied by the same air compressors.

Technical

Reference:

FR-0, Status Trees, Rev 13 Tl-12.04, Users Guide For Abnormal And Emergency Operating Instructions, Rev. 0008 WOG Background Document F-0 Critical Safety Function Status Trees, Revision 2 Proposed references None to be provided:

Learning Objective:

3-OT-STG-FRI

2. Given a set of plant conditions, use FR-l.1, FR-l.2, FRI.3, and the Critical Safety Function Status Trees, FR-H to correctly implement:Action steps and RNOs.

3-OT-FRH000 1

1. Given a set of plant conditions, use the Heat Sink Status Tree, FR-H to identify and implement the appropriate Function Restoration Instruction.

Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

81. WJE11EG2.L30 081 Given the following:

The crew has transitioned to and is now performing ECA-1.1, Loss of Sump Recirculation, as a result of a reactor trip and safety injection due to a LOCA outside containment.

During performance of ECA-1.1, a step directs the check of Aux Air pressure and if the pressure is less than 75 psig actions are directed to ensure affected train isolation valves 0-FCV-32-82 and 0-FCV-32-85 are positioned correctly.

While performing the ECA, the crew is able to isolate the LOCA and meet SI Termination criteria.

Which ONE of the following identifies...

(1) where the controls for 0-FCV-32-82 and 0-FCV-32-85 are located and (2) the procedure the crew will use to terminate the Safety Injection?

Note:

O-FCV-32-82 Essential Control Air Train A Normal Flow Isolation O-FCV-32-85 Essential Control Air Train B Normal Flow Isolation W

A. On 1-M-15 in the Remain in ECA-1.1 to terminate the SI.

Main Control Room B. On 1-M-15 in the Transition to ES-1.1, SI Termination.

Main Control Room C Elevation 757 in the Remain in ECA-1.1 to terminate the SI.

Auxiliary Building D. Elevation 757 in the Transition to ES-I.1, SI Termination.

Auxiliary Building

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A.

Incorrect, Plausible because there are air system valves on l-M-15 in the main control room that are operated durhg performance of the same step of the ECA and because terminating the SI using ECA-I. I is correct.

B.

Incorrect, Plausible because there are air system valves on I-M-I5 in the main control room that are operated durhg performance of the same step of the ECA and because a transition to ES-I. I is the procedure most often directed to be used to terminate a Safety Injection. It is also the procedure that would have been used to termhate the SI if the LOA outside contahment had been isolated before ECA-I. I was entered.

C.

Correct, The handswitches are located on Elevation 757 h the Auxiliary Building and when SI Termhation criteria is met while performing ECA-I. I, the steps to termhate the SI are contained h EcA-I. I.

D.

Incorrect, Plausible because the handswitches being located on Elevation 757 in the Auxiliary Buildhg is correct and because a transition to ES-I. I is the procedure most often directed to be used to terminate a Safety Injection. It is also the procedure that would have been used to termhate the SI if the LOA outside contahment had been isolated before ECA-I. I was entered.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Question Number:

81 Tier:

1 Group 1

K/A:

WIE11 EG2.1.30 Loss of Emergency Coolant Recirculation Ability to locate and operate components, including local controls.

Importance Rating:

4.4 I 4.0 10 CFR Part 55:

41.7/45.7 10CFR55.43b:

5 KIA Match:

KA is matched because the question requires the ability to locate controls required to be manipulated by the procedure and is SRO because it requires the knowledge of the procedure content and procedure selection needed for terminating a safety injection while performing the Loss of Emergency Coolant Recirculation procedure (ECA-.1.1)

Technical

Reference:

ECA-1.1, Loss of RHR Sump Recirculation, Revision 11 Proposed references None to be provided:

Learning Objective:

3-OT-STG-ECA1

08. Given a set of plant conditions, use procedures ECA-1.1 and 1.2 to identify any required procedure transition.

Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

82. 005 AG2.1.32 082 Given the following:

Unit 1 is operating at 100% power.

Which ONE of the following identifies...

(1) the individual rod position alignment limit specified by Tech Spec LCO 3.1.5, Rod Group Alignment Limits and (2) a Bases for the Tech Spec?

A. (1) The Banks most withdrawn rod must be within 12 steps of the least withdrawn rod.

(2) To limit local heat rate and power peaking factors.

B (1) All rods in the Bank must be within 12 steps of its respective group step counter.

(2) To limit local heat rate and power peaking factors.

C. (1) The Banks most withdrawn rod must be within 12 steps of the least withdrawn rod.

(2) To ensure that a natural flux shape is maintained acceptable for LOCA and loss of flow accidents.

D. (1) All rods in the Bank must be within 12 steps of its respective group step counter.

(2) To ensure that a natural flux shape is maintained acceptable for LOCA and loss of flow accidents.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A.

Incorrect; Plausible because the maximum withdrawn rod being within 12 steps of the minimum withdrawn rod in the group would be correct if the group step counter was inoperable as identified in LCO 3.1.8 and to limit local heat rate and power peaking factors is correct as identified in the Bases for LCO 3.1.5.

B.

Correct; Tech Spec 3.1.5 states All shutdown and control rods shall be OPERABLE, with all individual indicated rod positions within 12 steps of their group step counter demand position. The Bases for the Tech Spec identifies that Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

C.

Incorrect, Plausible because the maximum withdrawn rod being within 12 steps of the minimum withdrawn rod in the group would be correct if the group step counter was inoperable as identified in LCO 3.1.8 and the wording in the distractor is from the bases of LCO 3.2.3 for AFD.

D.

Incorrect, Plausible because all rods being within 12 steps of their group step counter demand position is correct and the wording in the distractor is from the bases of LCO 3.2.3 forAFD.

Question Number:

82 Tier:

1 Group 2

KIA:

005 AG2.1.32 Inoperable/Stuck Control Rod Ability to explain and apply system limits and precautions.

Importance Rating:

3.8 I 4.0 IOCFRPart55:

41.10/43.2/45.12 1

IOCFR55.43.b:

2 K/A Match:

KA is matched because the question requires the application of Tech Spec limits associated with an inoperable control rod and is SRO only because the question requires knowledge of the basis for the Tech Spec that contains the limit.

Technical

Reference:

Tech Spec 3.1.5, Rod Group Alignment Limits Tech Spec 3.1.5, Bases

, Rod Group Alignment Limits Bases, Revision 51 Tech Spec 3.1.8, Rod Position Indication Tech Spec 3.2.3 Bases, Axial Flux Difference (AFD)

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Proposed references None to be provided:

Learning Objective:

3-OT-T/S0301

3. Given plant parameters/conditions, correctly determine the compliance with the LCOs or TRs in the Reactivity Control sections of TIS and TIR manuals.

5. Determinethe bases for the limits placed on control rod positioning and position monitoring equipment (Rod Insertion Limits, Alignment Limits, and Rod Position Indicating Systems).

Cognitive Level:

Higher Lower X

Question Source:

New Modified Bank X

Bank Question History:

WBN bank questions AOl-0200.08 001 and SYSO85A.23 027 modified to include Tech Spec bases.

Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

83. 024 AA2.03 083 Given the following:

The crew has entered AOI-34, Immediate Boration, due to an uncontrolled RCS cooldown following a reactor trip.

The crew initiates action to establish the boration flow using normal boration.

Annunciator window 111 -E BA TO BLENDER FLOW DEVIATION alarms and the deviation can NOT be corrected.

Which ONE of the following identifies...

(1) the effect the condition will have on the boric acid flow and (2) the minimum RCS temperature that the boration is allowed to be terminated prior to verifying adequate Shutdown Margin in accordance with the AOl?

A (1) Automatic isolation of the boric acid flow will occur.

(2) 547°F B. (1) Automatic isolation of the boric acid flow will occur.

(2) 550°F C. (1) Boric acid flow will continue with the alarm condition (2) 547°F D. (1) Boric acid flow will continue with the alarm condition (2) 550°F

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANALYSIS:

A.

Correct, as identified in AR! window 11 1-E, the FCVs that will allow boration flow automatically close if the blender is operated in any mode except MANUAL.

A01-34, Immediate Boration, establlshes boron flow with the switch in BOR. The AOl allows termination of the boration when the RCS temperature is equal to or greater than 547°F and controlled or when hutdown margin is ensured.

B.

Incorrect, Plausible because automatic isolation of the boration flow is correct and 550°F is the setpoint for Lo-Lo Tavg.

C.

Plausible because automatic isolation of the boration flow would not occur if the boration had been established with the blender controls in manual and the AOl allowing termination of the boration when the RCS temperature is equal to or greater than 547°F is correct.

D.

Incorrect, Plausible because automatic isolation of the boration flow would not occur if the boration had been established with the blender controls in manual and and 550°F is the setpoint for Lo-Lo Tavg.

Question Number:

83 Tier:

1 Group 2

K/A:

024AA2.03 Emergency Boration Ability to determine and interpret the following as they apply to the Emergency Boration:

Correlation between boric acid controller setpoint and boric acid flow Importance Rating:

2.9*! 3.0 10 CFR Part 55:

43.5 / 45.13 IOCFR55.43.b:

2 K/A Match:

KA is matched because the question addresses how a deviation between boric acid controller setpoint and boric acid flow will affect an ongoing Immediate Boration and is SRO because it requires knowledge of specific actions in the procedure beyond the overall mitigating strategy and beyond system knowledge Technical

Reference:

AOI-34, Immediate Boration, Rev 23 ARI-1 09-115, CVCS & RHR - RPS & ESF, Revision 16 Window 111-E 1-45W600-62-3 R4

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Proposed references None to be provided:

Learning Objective:

3-OT-SYSO62A 28.

Describe the modes of operation of the CVCS Boron Concentration and Reactor Makeup Control System.

3-CT-AC13400 7.

Demonstrate ability/knowledge of AOl, by:

a. Recognizing Entry conditions.

b. Responding to Actions.

c. Responding to Contingencies (RNO).

d. Responding to Notes/Cautions.

Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

84. 060 AG2.4.3 084 Given the following:

Unit I is at 100% power.

During the last run cycle of the waste gas compressor, a fitting on the pressure instrument tap on the in-service Waste Gas Decay Tank is leaking gas from the tank.

Which ONE of the following describes the radiation monitor that would detect the accidental gaseous radwaste release?

A. The monitor is a PAM instrument and is required by an LCO in Section 3 of Tech Spec.

B. The monitor is a PAM instrument and is used in determining Radiological Emergency Plan classifications.

C. The monitor is NOT a PAM instrument but is required by an LCO in Section 3 of Tech Spec.

D The monitor is NOT a PAM instrument but is used in determining Radiological Emergency Plan classifications.

DISTRA CTOR ANALYSIS:

A.

Incorrect, Plausible because the Aux. Bldg. Vent Noble Gas monitor is measuring a release point from the site (and there are other radiation monitors that are PAM instruments) and other radiation monitors are required by Technical Specification Seótion 3.0 LCOs. Also plausible because waster gas release is normal made via the shield building stack and that monitor is a PAM instrument.

B.

Incorrect, Plausible because the Aux. Bldg. Vent Noble Gas monitor is measuring a release point from the site (and there are other radiation monitors that are PAM instruments) but the monitor being used in EPIP-1, Emergency Plan Classification Flow Chart, Sections 7.1 and 7.4 in determining REP classifications is correct.

Also plausible because waster gas release is normal made via the shield building stack and that monitor is a PAM instrument.

C.

Incorrect, Plausible because the monitor not being a PAM instrument is correct and other radiation monitors are required by Technical Specification Section 3.0 LCOs.

D.

Correct, FSAR Table 7.5-2 identifies that a Deviation was taken to the requirements of Reg Guide 1.97 for Post Accident Monitoring Variables and the Auxiliary Building Vent Monitor is Not a PAM instrument; however the monitor is used is used in EPIP-1, Emergency Plan Classification Flow Chart, Sections 7.1 and 7.4..

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Question Number:

84 Tier:

1 Group 2

K/A:

000060 AG2.4.3 Accidental Gaseous Radwaste Release Ability to identify post accident instrumentation.

Importance Rating:

3.7 I 3.9 10 CFR Part 55:

41.6 /45.4 IOCFR55.43.b:

2, 4, 5 KIA Match:

KA is matched because the question requires the ability to identify a rad monitor that would detect an inadvertent release, determine whether the monitor is post accident instrumentation and is then SRO because the determination of the use of the instrument in the Radiological Emergency Plan implementation must be determined.

Technical

Reference:

WBN FSAR Table 7.5-2, Tech Spec 3.3.3, Post Accident Monitoring Instrumentation, Amendment 72 EPIP-1, Emergency Plan Classification Flow Chart, Revision 33, Sections 7.1 and 7.4 Proposed references None to be provided:

Learning Objective:

3-OT-SYSO9OA

16. Identify where Post Accident monitors are used and read out.

Cognitive Level:

Higher Lower X

Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

85. 061 AA2.01 085 Unit I is in a Refueling outage with the following conditions:

Reactor core has been off-loaded to the Spent Fuel Pit.

The core empty work is in progress.

Spent fuel assembly shuffles in the spent fuel pit is in progress.

The following occurs during the shift:

0900 HS-90-136A2, VENT ISOL RAD MON BLOCK, is placed to the 0-103 position with the handswitch in the PULL-TO-TEST position to support maintenance.

1100

- Annunciatorwindow 184-D SEP 0-RM-102/103 INSTR MALE alarms and the CR0 reports the following indications on 0-M-12: RM-90-1 02, Spent Fuel Pit Area Monitor, has no indicating lights LIT. RM-90-1 03, Spent Fuel Pit Area Monitor, has only the GREEN indicating light LIT Which ONE of the following identifies...

(1) the effect the blocking of the radiation monitor at 0900 has on the fuel shuffles in the Spent fuel Pit and (2) the minimum number of ABGTS trains required to be in service to allow fuel shuffles to continue due to the conditions at 1100?

A. (1) Fuel movement required to be stopped until an ABGTS train is placed in service.

(2) at least onetrain B. (1) Fuel movement required to be stopped until an ABGTS train is placed in service.

(2) Both trains C (1) Fuel movement can continue but an ABGTS train is required to be placed in service within 7 days (2) at least one train D. (1) Fuel movement can continue but an ABGTS train is required to be placed in service within 7 days (2) both trains

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A.

Incorrect, Plausible because other conditions associated with these radiation monitors does require suspension of fuel movement and having at least one train in service at 1400 is required.

B.

Incorrect, Plausible because other conditions associated with these radiation monitors does require suspension of fuel movement and at 1400 LCO 3.3.8 allows placing both trains in service to met the required action but it is not the minimum number required.

C.

Correct, The fuel movement can continue. TS LCO 3.3.8 entry requires one Train ofABGTS to be placed in service in 7 days and at 1400 with both monitors out of service the minimum number of ABGTS trains required to be in service is one. This is in accordance with LCO 3.3.8 Condition B which also requires the applicable conditions of LCO 3.7.12 be entered for one train of ABGTS made inoperable by inoperable actuation instrumentation.

D.

Incorrect, Plausible because fuel movement continuing at 0900 with a required action to place one Train of ABGTS to be placed in service in 7 days is correct and at 1400 LCO 3.3.8 allows placing both trains in service to met the required action but it is not the minimum number required.

Question Number:

85 Tier:

1 Group 2

KIA:

061 AA2.01 Area Radiation Monitoring (ARM) System Alarms Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms:

ARM panel displays Importance Rating:

3.5 I 3.7 10 CFR Part 55:

43.5 / 45.13 IOCFR55.43.b:

2 KIA Match:

KA is matched and the questions is SRO only because the question requires the ability to determine the status of area radiation monitors covered by Technical Specifications and from conditions in the stem determine if Tech Spec covering the radiation monitors or the system the monitors support require entering LCO actions.

Technical

Reference:

ARI-1 80-1 87, Common Radiation Detectors, Revision 31

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Technical

Reference:

ARI-1 80-1 87, Common Radiation Detectors, Revision 31 Window 184-D Tech Spec 3.3.8, ABGTS Actuation Instrumentation Tech Spec 3.3.8, ABGTS Actuation Instrumentation, Bases, Revision 87 Tech Spec 3.7.12, ABGTS Tech Spec B3.7.12, ABGTS Bases, Revision 29,34 Proposed references None to be provided:

Learning Objective:

3-OT-T/S0303

4. Given plant parameters/conditions, correctly determine the OPERABILITY of the various instrumentation systems covered by T/S or T/R.

3-OT-SYS9OA

07. Determine interlocks and/or cause-effect relationships between the Rad Monitoring Systems (ARM & Process) and the areas they monitor.

Include HVAC systems and area isolations.

Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

86. 003 A2.03 086 Given the following:

Unit 1 is in Mode 4 with heatup in progress and the last 2 RCPs have been started.

20 minutes after the start of the RCPs, annunciator window 100-A, RCP STATORJMTR THRUST BRG TEMP HI alarms.

The operator determines the following RCP #4 temperatures and trends:

Motor winding temperatures have risen and are now stable at:

Phase A

- 304°F Phase B

- 303°F Phase C

- 304°F Motor Bearing temperatures have risen and are now stable at:

Upper -186°F Lower -181°F AOl-24, RCP Malfunctions During Pump Operation is implemented.

Which ONE of the following choices completes the statement below?

The AOl entry was required due to the motor (1) temperature and the SRO will direct the pump be removed from service (2)

A.

bearing within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months B.

bearing immediately C.

winding within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months D

winding immediately

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A.

Incorrect, Plausible because the motor bearing temperatures are higher than normal and there are conditions in the AOl that require an RCP to be removed from service within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , B.

lncorrect, Plausible because the motor bearing temperatures are higher than normal and if the temperatures had been higher the AOl would require the pump to be removed immediately.

C.

Incorrect, Plausible because the AOl entry is required due to the high motor winding temperatures and there are conditions in the AOl that require an RCP to be removed from service within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

D.

Correct, the AOl entry was required due to the high motor winding temperatures and with the temperatures greater than 302°F the AOl requires the pump to be removed from service immediately Question Number:

086 Tier:

2 Group 1

K/A:

003 A2.03 Reactor Coolant Pumps Ability to (a) predict the impacts of the following 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 associated with RCP motors, including faulty motors and current, and winding and bearing temperature problems Importance Rating:

2.7 / 3.1 10 CFR Part 55:

41.5 / 43.5/ 45.3 /45/13 IOCFR55.43.b:

5 K/A Match:

KA is matched because the question requires the ability to predict the impact of abnormal RCP motor parameters and then use an attachment within the procedure to identify actions required by the abnormal operating instruction. SRO because the question requires knowledge of the content of the procedure.

Technical

Reference:

ARI-95-1 01, Reactor Coolant Pumps, Revision 31 Window 100-A AOI-24, RCP Malfunctions During Pump Operation, Revision 0029 SOl-68.02, Reactor Coolant Pumps, Revision 0033

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Proposed references None to be provided:

Learning Objective:

3-OT-A012400

09. Identify the parameters listed in AOI-24 that require the RCP to be shutdown

10. Given a set of plant conditions, use AOI-24 to correctly:

b. Identify Required Actions.

11. Given a set of conditions, determine if RCP shutdown is required using AOI-24, Attachment 2.

Cognitive Level:

Higher X

Lower Question Source:

New Modified Bank X

Bank Question History:

SQN bank question 003 A2.03 086 used on 1/2009 exam modified.

Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

87. 005 A2.01 087 Given the following:

Unit 1 is at midloop with RCS level at 718 9 following a refueling outage.

The operating crew is drawing vacuum on the Reactor Coolant System in accordance with GO-b, Reactor Coolant System Drain and Fill Operations, RHR pump lA-A is in service with 2100 gpm flow.

The RHR pump amps and flow begin to fluctuate.

Which ONE of the following identifies...

(1) an action directed by GO-I 0 in response to the pump conditions and (2) the mitigating strategy that will be implemented if the pump conditions cannot be stabilized?

Action Mitigating Strategy A. Raise RCS Level Break vacuum per GO-b 0, then enter AOl-I 4, Loss of RHR Shutdown Cooling.

B. Raise RCS Level Immediately enter AOl-i 4, Loss of RHR Shutdown. Break vacuum as directed by the AOl.

C Lower RHR Flow Break vacuum per GO-I 0, then enter AOl-i 4, Loss of RHR Shutdown Cooling.

D. Lower RHR Flow Immediately enter AOl-i 4, Loss of RHR Shutdown. Break vacuum as directed by the AOl.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A.

Incorrect, Plausible because raising level would be correct if the stated RCS level had been lower and also because breaking vacuum per GO-b, then entering A0l-14, Loss of RHR Shutdown Cooling, is correct, B.

Incorrect, Plausible because raising level would be correct if the stated RCS level had been lower and also because both A01-14 being implemented and vacuum being broken will occur due to the pump conditions but not in the order stated in the choice.

C.

Correct, If RHR Pump cavitation or unstable pump amp/flow readings occur, GO-b directs reducing RHR pump flow to 2000 gpm, raising RCS level to 718 9 and reducing vacuum. Since flow is higher than 2000 gpm, the flow can be reduced to 2000 gpm and if the pump cannot be stabilized then the GO directs the breaking of vacuum prior to the implementation of AOl-14. (see below)

D.

Incorrect, Plausible because reducing RHR flow is correct and A0l-14 will be implemented and also because both A 01-14 being implemented and vacuum being broken will occur due to the pump conditions but not in the order stated in the choice.

GO-b 5.4.3 RCS Vacuum Refill

[42]

IF RHR Pump cavitation or unstable pump amp/flow readings occur, THEN

[42.1] REDUCE RHR pump flow to 2000 gpm.

[42.2] RAISE RCS level to 718 9.

[42.3] ENSURE Vacuum Pump Operator notified to reduce vacuum slowly as required.

[43] IF RHR Pump does NOT stabilize following the performance of Step 5.4.3[42], OR RHR flow is lost for any reason, THEN

[43.1] ENSURE Vacuum Pump Operator notified to break vacuum by performing Section 4.0 Emergency Vacuum Break in APPENDIX AA.

[43.2] GO TO AOl-i 4.

Question Number:

87 Tier:

2 Group 1

KIA:

005 A2.01 Residual Heat Removal System (RHRS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation Importance Rating:

2.7 I 2.9*

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 10 CFR Part 55:

41.5/43.5/45.3/45.13 IOCFR55.43.b:

5 KIA Match:

KA is matched because the question requires prediction of actions directed by the procedure to mitigating the pumps instability and then actions, and procedure selection when the pump cannot be stabilized. SRO because the question requires detailed knowledge of the procedure content beyond the overall mitigating strategy and then when the AOl would be implemented.

Technical

Reference:

GO-b, Reactor Coolant System Drain And Fill Operations, Revision 0042 Proposed references None to be provided:

Learning Objective:

3-OT-GO1 000

4. [Identify the indication(s) of Residual Heat Removal pump cavitation and the checks required per GO-b.

(SOER 88-3, Rec. 3.c)]

5. [Identify the procedure to which the operator is referred if Residual Heat Removal cooling is lost while in during Reduced Inventory/Mid-Loop operations.

(SOER 88-3 & SOER 85-4)]

Cognitive Level:

Higher X

Lower Question Source:

New Modified Bank X

Bank Question History:

WBN bank question AOI1400.07 012 modified and the correct answer location changed from the original question.

Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

88. 013 G2.2.12 088 Given the following conditions:

The plant is at 100% power.

1 -Sl-99-1 0-A, 62 Day Functional Test of SSPS Train A and Reactor Trip Breaker A, is in progress.

During performance of the Surveillance Instruction (SI), which ONE of the following identifies...

(1) a condition where MEW Isolation Actuation handswitches are held in the Reset position to prevent the potential of a main feedwater isolation and (2) the status of the RTA Shunt Trip device if, during an engineering review 15 days after the maximum late date, it is determined that the steps to test the the Reactor Trip Breaker (RTA) shunt trip were not performed?

A. When closing the Reactor Declare the Shunt Trip device Trip Breaker A (RTA) inoperable until the SI steps are performed.

B.

When returning the Input Error Declare the Shunt Trip device Inhibit switch to normal.

inoperable until the SI steps are performed..

C. When closing the Reactor Shunt Trip device remains operable Trip Breaker A (RTA) provided the test is successfully performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D When returning the Input Error Shunt Trip device remains operable Inhibit switch to normal.

provided the test is successfully performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A.

Incorrect, Plausible because a FWI signal can be generated when the Reactor Trip breaker is opened if Tavg is less than 564°F (Reactor trip with Lo Tavg) and the question identifies a condition where a surveillance requirement has not been met.

B.

Incorrect, Plausible because 1-SI-99-10-A requiring the Math Feedwater Isolation Actuation switches to be held when the Input Error Inhibit switch is returned to normal is correct and the question identifies a condition where a surveillance requirement has not been met.

C.

Incorrect, Plausible because a FWI signal can be generated when the Reactor Trip breaker is opened if Tavg is less than 564°F (Reactor trip with Lo Tavg) and the Shunt Trip device remaining operable is correct.

D.

Correc1, 1-Sl-99-10-A, 62 Day Functional Test of SSPS Train A and Reactor Trip Breaker A, requires the Math Feedwater Isolation Actuation switches to be held when the Input Error Inhibit switch is returned to normal and Tech Spec SR 3.0.3 allows the delay in declaring the LCO not met for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as described below.

SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately

+

Question Number:

88 Tier:

2 Group 1

K/A:

013 G2.2.12 Engineered Safety Features Actuation System Knowledge of surveillance procedures.

Importance Rating:

4.1 IOCFRPart55

41.10/45/13 IOCFR55.43.b:

2, 5 K/A Match:

The K/A is matched because the question requires knowledge of Engineered Safety Features Actuation System surveillance

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 KIA Match:

The K/A is matched because the question requires knowledge of Engineered Safety Features Actuation System surveillance procedures. It is SRO only because the questions requires specific knowledge of actions required in the procedure and not just the overall strategy and because it requires knowledge Tech Spec 3.0 Surveillance Requirement (SR) Applicability if it is discovered that a Surveillance was not performed within its specified Frequency, Technical

Reference:

1-Sl-99-10-A, 62 Day Functional Test of SSPS Train A and Reactor Trip Breaker A, Rev. 45 Tech Spec 3.0, Surveillance Requirement (SR)

Applicability, Amendment 42 Proposed references None to be provided:

Learning Objective:

3-OT-SYSO99A

18. Given the condition/status of the Reactor Protection system/component and the appropriate sections of Tech Specs, determine if operability requirements are met and what actions, if any, are required.

Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

89. 059 A2.03 089 With Unit 1 operating at 60% power the following occurs:

A transient on the condensate system resulted in all MFW Reg valves traveling full open.

The SG #2 MEW Reg valve stuck at the full open position and operator efforts to control the valve were unsuccessful.

Without any manual action to initiate a unit trip by the operator, which ONE of the following identifies how the unit will trip and after the trip, the procedure that will ensure proper feedwater isolation on the unit?

A. An automatic reactor trip will cause the main turbine to trip; E-0, Reactor Trip or Safety Injection B. An automatic reactor trip will cause the main turbine to trip; ES-0.1, Reactor Trip Response C. An automatic turbine trip will cause the reactor to trip; E-0, Reactor Trip or Safety Injection D An automatic turbine trip will cause the reactor to trip; ES-0.1, Reactor Trip Response DISTRACTOR ANALYSIS:

A.

Incorrect, Plausible because a reactor trip does cause a turbine trip and E-O, Reactor Trip or Safety Injection does have steps to check FWI but the transition to ES-O. I would be made prior to the step being performed because there is no safety injection.

B.

Incorrect Plausible because a reactor trip does cause a turbine trip and steps to check the FWI being performed during the performance of ES-O. 1, Reactor Trip Response.

C.

Incorrect, Plausible because the turbine trip causing the reactor trip is correct and E-O, Reactor Trip or Safety Injection does have steps to check FWI but the transition to ES-U. I would be made prior to the step being performed because there is no safety injection.

D.

Correct, the steam generator level will rise until a Feedwater Isolation will result in a trip of the main turbine which will cause a reactor trip and the FWI will be ensured during the performance of ES-U. I, Reactor Trip Response.

Question Number:

89

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Tier:

2 Group 1

K/A:

059 A2.03 Main Feedwater (MFW) System Ability to (a) predict the 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:

Overfeeding event Importance Rating:

2.7 3.1*

10 CFR Part 55:

41.5 / 43.5 / 45.3 / 45.13 IOCFR55.43.b:

5 K/A Match:

KA is matched because the question requires the prediction of the impact of an overfeeding event on the feedwater system and the procedure that would be used to verify the event had been mitigated.

SRO because if requires knowledge of the steps in procedures and the selection of the procedure that would verify conditions as described in the question.

Technical

Reference:

ES-0.1 Reactor Trip Response, Revision 0022 E-0, Reactor Trip or Safety Injection, Revision 28 Proposed references None to be provided:

Learning Objective:

3-OT-SYSOO3A

11. Identify the Feedwater Isolation signals

12. List the equipment affected by a Feedwater Isolation Signal Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

90. 078 G2.4.4 090 Given the following:

Unit I is operating at 100% power.

Annunciator windows alarm and are acknowledged by the OAC.

41-F, CONTROL AIR PRESS LO 42-F, SERVICE AIR PCV-33-4 CLOSED 136-B, AUX AIR TR-A PRESS LO CR0 reports...

0-PI-32-104A, AUX AIR A PRESS, at 20 psig and dropping 1-HS-32-80A, AUX AIR TO RX BLDG TR A, green light is lit.

Annunciator window 41-F, CONTROL AIR PRESS LO begins to flash slowly.

Which ONE of the following identifies the procedure set(s) that will be required to be used in response to the event and what action will be required in accordance with Tech Spec 3.7.5, Auxiliary Feedwater (AFW) System?

Note:

AOIs - Abnormal Operating Instructions EOIs - Emergency Operating Instructions Procedures Tech Spec A AOIs, jjy Enter Condition B for one Train Inoperable B. AOIs, gjy Enter Condition C for two Trains Inoperable C. Both AOIs and EOIs Enter Condition B for one Train Inoperable D. Both AOI5 and EOls Enter Condition C for two Trains Inoperable

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A.

Correct, The conditions indicate a loss of the Train A essential air (auxiliary air) system. With the CONTROL AIR PRESSURE LO alarm window flashing slowly, the control air has isolated from the essential air system and is recovering pressure. A 01-10, Loss of ControlAir, contains a section to address the loss of the auxiliary air Train. Contained within this section is the step to evaluate Tech Specs and while the loss of air affects both the MDAFW Train A and the TDAFW Train there is information in the Bases that identifies that the loss of a single train of air (A or B) will not prevent the auxiliary feedwater system from performing its intended safety function and is no more severe than the loss of a single auxiliary feedwater pump. Thus, entering Condition B for one Train inoperable is required.

(see below)

B.

lncorrect Plausible because using only the AOls to respond to the event is correct and because the loss of the air system does affect components in both the MDAFW Train A and the TDAFW flow path.

C.

lncorrect, Plausible because using the AOls and EOIs to respond to the event would be correct if the conditions had identified the problem to be in the control air system instead of the essential air system and because entering Condition B for one Train inoperable is correct.

D.

Incorrect, Plausible because using the AOls and EOls to respond to the event would be correct if the conditions had identified the problem to be in the control air system instead of the essential air system and because the loss of the air system does affect components in both the MDAFW Train A and the TDAFW flow path.

From Tech Spec 3.7.5 Bases Each motor-driven auxiliary feedwater pump (one Train A and one Train B) supplies flow paths to two steam generators. Each flow path contains automatic air-operated level control valves (LCVs). The LCVs have the same train designation as the associated pump and are provided trained air. The turbine driven auxiliary feedwater pump supplies flow paths to all four steam generators. Each of these flow paths contains an automatic air-operated LCV, two of which are designated as Train A, receive A-train air and provide flow to the same steam generators that are supplied by the B-train motor-driven auxiliary feedwater pump. The remaining two LCVs are designated as Train B, receive B-train air, and provide flow to the same steam generators that are supplied by the A-train motor-driven pump. This design provides the required redundancy to ensure that at least two steam generators receive the necessary flow assuming any single failure. It can be seen from the description provided above that the loss of a single train of air (A or B) will not prevent the auxiliary feedwater system from performing its intended safety function and is no more severe than the loss of a single auxiliary feedwater pump. Therefore, the loss of a single train of auxiliary air only affects the capability of a single motor-driven auxiliary feedwater pump because the turbine-driven pump is still capable of providing flow to the two steam generators that are separated from the other motor-driven pump.

Question Number:

90

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Tier:

2 Group 1

K/A:

078 G2.4.4 Instrument Air System (lAS)

Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Importance Rating:

4.5 / 4.7 10 CFR Part 55:

41.10/43.2/45.6 IOCFR55A3.b:

5, 2 K/A Match:

KA is matched because the question requires the ability to recognize the type of failure that has occurred in the air system and realize the procedures required in response to the plant conditions. SRO because the question requires both the selection of procedures and the knowledge of information contained in the Tech Spec Bases.

Technical

Reference:

ARI-1 31-1 37, Miscellaneous, Revision 20 window 136-B ARI-36-42, Heaters, Turb Seal & Air, Revision 17 windows 41-F & 42-F AOl-b, Loss of Control Air, Revision 0039 Tech Spec 3.7.5

, Auxiliary Feedwater (AFW) System Tech Spec B3.7.5

, Auxiliary Feedwater (AFW) System Bases Proposed references None to be provided:

Learning Objective:

3-OT-AOl1000 7.

Given a set of plant conditions, use AOl-b to correctly:

a. Recognize Entry Conditions.

b. Identify Required Actions.

3-OT-SYSOO3B

23. Using plant drawings, determine the effect of a loss of instrument air/control power on the following valves/components:

a. MDAFWP regulating valve (main and bypass)

b. TDAFWP regulating valve

c. AFW pumps 3-OT-T/S0307 3.

Given plant conditions and parameters, correctly determine the OPERABILITY of components associated with different Plant Systems in Section 7 of Technical Specifications Cognitive Level:

Higher X

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Lower Question Source:

New Modified Bank X

Bank Question History:

WBN bank question T1S0307.03 002 modified Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

91. 0341(5.03 091 Given the following:

Unit 1 is in Mode 6 with core reload in progress.

RHR pump lB-B is inoperable due to motor failure.

RHR pump lA-A is scheduled to be removed for 40 minutes to allow testing of the RCS to RHR inlet valves.

In accordance with Tech Spec 3.9.5, Residual Heat Removal(RHR) and Coolant Circulation

- High Water Level, and its Bases, which ONE of the following identifies the operational implication of stopping the Train A RHR pump to allow the valve testing.

A%# Core reload can continue during the time the RHR pump is stopped provided no operations are permitted that would cause a reduction of the RCS Boron concentation.

B. Core reload can continue during the time the RHR pump is stopped provided the RHR pump lA-A is re-aligned for cold leg injection while the pump is off and the testing is in progress.

C. Core reload must be stopped during the time the RHR pump is stopped and no operations are permitted that would cause a reduction of the RCS Boron concentation.

D. Core offload must be stopped during the time the RHR pump is stopped and the RHR pump is required to be re-aligned for cold leg injection while the pump is off and the testing is in progress.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A.

Correct, Tech Specs allows for the RHR pump to be stopped for up to 1-hour in an 8-hour period without requiring the core reload to be stopped but does require that no operation be permitted that would cause a reduction in the RCS boron concentration while the RHR pump is stopped.

B.

lncorrect Plausible because core reload being able to continue is correct and the Bases for LCO 3.9.5 contains wording for the requirement for RHR system alignment while the required RHR pump is out of service for up to 1-hour in an 8-hour period. This wording includes discussion of RHR alignment for injection allowances and RCS to RHR valve testing valve testing requirements but does not connect the two.

C.

Incorrect, Plausible because suspending core reload and any operation that would result in a reduction of RCS boron concentration are required actions if the required RHR loop is not operable except as allowed by the note in the LCO.

D.

lncorrect Plausible because suspending core reload is a required action if the required RHR loop is not operable except as allowed by the note in the LCO and the bases wording contains wording that includes discussion of RHR alignment for injection allowances and RCS to RHR valve testing valve testing requirements but does not connect the two.

An OPERABLE RHR loop includes an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the low end temperature.

The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.

Both RHR pumps may be aligned to the RWST to support continued filling of the refueling cavity or for performance of RHR injection testing. During these modes of operation, the wide range RCS temperature indicators are used to indicate RCS temperature since the RHR temperature elements indicate RWST temperature when RHR pump suction is from the RWST.

The flow path for these modes of operation starts at the RWST and is supplied to the RCS cold legs (or hot legs for hot leg injection testing).

If only one pump is in operation, then hot leg injection testing must be done under the provisions of the NOTE discussed in the following paragraph.

The LCO is modified by a Note that allows the required operating RHR loop to be removed from service for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided no operations are permitted that would cause a reduction of the RCS boron concentration.

Boron concentration reduction is prohibited because uniform concentration distribution cannot be ensured without forced circulation.

This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles and RCS to RHR isolation valve testing.

During this 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, decay heat is removed by natural convection to the large mass of water in the refueling cavity.

Question Number:

91 Tier:

2 Group 2

K/A:

034 K5.03 Fuel Handling Equipment System (FHES)

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 KIA:

034 K5.03 Fuel Handling Equipment System (FHES)

Knowledge of the operational implication of the following concepts as they apply to the Fuel Handling System:

Residual heat removal; decay Importance Rating:

2.2 I 2.7 10 CFR Part 55:

41.5 /45.7 IOCFR55.43.b:

2, 3 K/A Match:

KA is matched because the question requires knowledge of Residual Heat Removal Tech Spec requirements and the associated Bases during movement of fuel.

Technical

Reference:

Tech Spec 3.9.5, Residual Heat Removal (RHR) and Coolant Circulation High Water Level and its Bases Proposed references None to be provided:

Learning Objective:

3-OT-SYSO79A

12. Identify the Tech Specifications! Tech. Requirements relative to Fuel Handling with regard to:

a. Boron concentration

b. Source Range Neutron Monitoring

c. Decay time

d. Water level over the core

e. Communications

f. Definition: Refueling Mode Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

92. 041 A2.03 092 Given the following:

A cooldown is in progress to place Unit 1 in Mode 5 in accordance with GO-6, Unit Shutdown From Hot Standby To Cold Shutdown.

Wire lifts have been performed to allow all 12 steam dump valves to be opened in order to maintain cooldown rate at the desired value.

While in this configuration, the instrument air line supplying pneumatic controller 1-FM-i -3, Steam Dump, is broken off just before it enters the controller.

Which ONE of the following identifies...

(1) how the wire lift allowing the dump valves to be open will be indicated in the Main Control Room, and (2) the action required to restore control of RCS temperature following the loss of air to i-FM-i-I 03?

Note:

TACF - Temporary Alteration Control Form A. (1) TACF (2) Place 1-HS-1-103D, Steam Dump Mode, to TAVG position in accordance with GO-6.

B. (1) TACF (2) Locally operate valves with handwheel as identified in SQl-i.02, Steam Dump System.

C. (1) Caution Order (2) Place i-HS-1-103D, Steam Dump Mode, to TAVG position in accordance with GO-6.

D (1) Caution Order (2) Locally operate valves with handwheel as identified in 501-I.02, Steam Dump System.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANALYSIS:

A.

lncorrecl Plausible because a TACF is used to make temporary changes to the facility (and the change could be a wire lift) and with a loss of instrument air to the controller there are valves in the plant that fall open but the steam dump valves fall to the closed position.

B.

Incorrect, Plausible because a TACF is used to make temporary changes to the facility (and the change could be a wire lift) and using the handwheels on the steam dump valves as describes in a Precaution in SOl-1.02 would allow the cooldown to continue C.

lncorrect, Plausible because the use of a Caution Order to indicate the status of the wire lift is correct and with a loss of instrument air to the controller there are valves in the plant that fall open but the steam dump valves fall to the closed position.

D.

Correct GO-6 Attachment F states in a note that the wire lifts must be tracked until returned normal and Section 3.0 Step [6] direct placing a Caution Order on 1-HS-1-103A, 1-HS-1-103B, AND 1-PIC-1-33 indicating that P-12 interlock is disabled. There is a Precaution in SOl-1. 02 that describes the operation of the handwheel on the steam dump valves and provides direction on using the handwheel. (see below)

Appendix F

- P 12 Interlock Bypass to Support Additional Steam Dump NOTE prior to Section 2.0 Maintaining effective cooldown below 300° F (i.e., as close as practical to our administrative target of 75°F/hr) requires the use of more than three steam dumps. This appendix provides the steps necessary to utilize the option of more than three valves for cooldown. Wires are lifted to defeat the P-12 interlock. Wire lifts must be administratively controlled until returned normal.

Section 3.0

[6] PLACE Caution Order on 1-HS-1-103A, 1-HS-1-103B, AND 1-PIC-1-33 indicating that P-12 interlock is disabled.

SOl-1.02 3.0 PRECAUTIONS AND LIMITATIONS D. The Steam Dump Valve handwheels are reverse action (clockwise to OPEN).

Counter clockwise removes the dog, allowing spring tension to close the dump valves. The handwheel should only be used with the air supply closed and the air pressure equalized across the diaphragm using the valves provided on the positioner.

Question Number:

92 Tier:

2 Group 2

K/A:

041 A2.03 Steam Dump System (SDS)/Turbine Bypass Control Ability to (a) predict the impacts of the following malfunctions or operations

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 KIA:

041 A2.03 Steam Dump System (SDS)JTurbine Bypass Control Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations:

Loss of lAS Importance Rating:

2.8/ 3.1 10 CFR Part 55:

41.5 / 43.5 / 45.3/ 45i3 IOCFR55.43.b:

5 K/A Match:

KA is matched because the question requires the prediction of how a loss of instrument air will affect the steam dump valves and then how the consequences of the failure would be mitigated. SRO only because if requires knowledge of the procedure selected to mitigate the consequence of the failure as well as the procedure required to administratively track the wire lift on the system.

Technical

Reference:

GO-6, Unit Shutdown From Hot Standby To Cold Shutdown, Revision 0044 SQl-i.02, Steam Dump System, Revision 0012 1-47W610-1-3, RiO Proposed references None to be provided:

Learning Objective:

3-OT-SYSOO1 B

05. Given a loss of instrument air/control power, determine the effect on the STEAM DUMP valves.

23. Describe the position in which the Bailey Positioner Supply and Bypass Valve handles must be in for automatic operation of the Steam Dump Valve; for handwheel operation of the Steam Dump Valve.

24. Explain which direction the steam dump valve handwheel must be turned to manually open a valve.

Cognitive Level:

Higher Lower X

Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 93.o79G2.1.20093 Given the following:

The plant is in Mode 6 and performing valve stroke exercising of I -FCV-32-80, Essential Control Air Train A Containment Isolation Valve in accordance with 1-SI-32-901-A, Valve Full Stroke Exercising During Cold Shutdown Control Air (Train A).

The following criteria applies to the valve stroke test:

The first test of the valve measured stroke time was 4.0 seconds The second test of the valve was timed at 3.9 seconds.

Which of the following actions must be taken?

A. The is valve inoperable, but the LCO is Tracking Only until Mode 4 entry.

B. Declare the valve inoperable, and enter LCO 3.6.3 and isolate the penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

C The valve remains operable pending the outcome of an engineering analysis which must be completed within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months of the test.

D. The valve remains operable until the next scehduled performance of the surveillance as long as the cause of the deviation is documented in the test data package.

SDKE 3TRJ

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A.

Incorrect,, Plausible because the valve exceeds the Stroke time acceptance criteria identified in the SI and because the plant is in a mode where LCO 3.6.3 would be tracking only.

B.

Incorrect,, Plausible because declaring the valve inoperable and complying with the actions of LCO 3.6.3 would be correct if the time had exceeded the maximum value and the plant was in a mode where the LCO was applicable.

C.

Correct, As indicated in the SI the valve remains operable pending an engineering evaluation which must be completed within 96 hrs.

D.

Incorrect, Plausible because the stroke times are within the Limiting Value of Full Stroke time identified in the SI and the deviation of the first stroke time is documented in the test package if the second test meets acceptance criteria.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Question Number:

93 Tier:

2 Group 2

K/A:

079 G2.1.20 Station Air Ability to interpret and execute procedure steps.

Importance Rating:

4.6 I 4.6 IOCFRPart55:

41.10/43.4/45.12 IOCFR55.43.b:

2 K/A Match:

The KA is matched because the question requires the ability to interpret and execute procedure steps associated with the performance of a surveillance instruction relative to stroke time testing of station air containment isolation valves when times are not within expected bands.

Technical

Reference:

1-S1-32-901-A, Valve Full Stroke Exercising During Cold Shutdown Control Air (Train A), Revision 8 Tech Spec 3.6.3, Containment Isolation Valves Proposed references None to be provided:

Learning Objective:

3-OT-T/S0306

5. Given plant conditions and parameters, determine applicable Action Conditions, Required Actions, and Completion Times associated with the Containment System.

Cognitive Level:

Higher X

Lower Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

94. G 2.1.21 094 Given the following:

The unit is at 100% power with a degraded condition that has been evaluated by the OPDP-1 1, Operational Decision-Making Issue Evaluation Process.

The ODMI has been revised several times and an on-coming SRO needs to verify trigger points in the 0DM I.

Which ONE of the following identifies how the SRO will verify the trigger points using a controlled copy of the ODMI?

As By looking in the Standing Order book B. By looking in the Business Support Library (BSL)

C. By looking in the Operations Directive Manual D. By looking in the eSOMS narrative log.

DISTRA CTOR ANALYSIS:

A.

Correct, OPDP-1 1, Operations Decision-Making Issue Evaluation Process 3.9 0DM! Implementation directs the Shift Manager to enter the ODMI into the standing Order Book as a Stand-Alone Standing Order to make operators aware of the decision, monitoring requirements, contingencies, etc.

B.

Incorrect, Plausible because the Business Support Library (BSL) to verify a procedure is the controlled version is correct for procedures that are maintained in the Business Support Library.

C.

Incorrect, Plausible because an Operations Directive Manual (0DM) is maintained in the control room and contains the current revision to the ODMs (which are not maintained in BSL)

D.

Incorrect, Plausible because OPDP-1 1, Operations Decision-Making Issue Evaluation Process directs any trigger points reached to be entered into the operating log (as well as the initiation/revision to 0DM!) directs, but the log entry would only identify when trigger points were reached not the trigger points contained within the 0DM!.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Question Number:

94 Tier:

3 Group n/a K/A:

G2.1.21 Conduct of Operations Ability to verify the controlled procedure copy.

Importance Rating:

3*5* / 3.6*

IOCFRPart55:

41.10/45.10/45.13 10CFR55.43b:

3,5 KIA Match:

KA is matched because the question requires knowing how to verify the controlled copy of a procedure being used and is SRO because it requires both the knowledge of the ODMI process which has actions required by an SRO (Shift Manager) relative to the ODMI implementation and how an SRO would perform the verification of trigger points and actions required by the 0DM I.

Technical

Reference:

OPDP-1 1, Operational Decision-Making Issue Evaluation Process, Revision 0000 Proposed references None to be provided:

Learning Objective:

3-OT-SSPO2O2

5. Describe the procedure revision process.

Cognitive Level:

Higher Lower X

Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

95. G2.1.8 095 Given the following:

An ALERT has been declared on Unit 1.

All emergency centers are activated.

Conditions require AUO action to manually isolate 0-ISV-70-700, RCP OIL COOLER CCS RETURN ISOLATION [A4N EL. 710 U-i Penetration room] in accordance with E-0, Reactor Trip Safety Injection, Attachment B3.

In accordance with EPIP-7, Activation and Operation of the Operations Support Center (OSC), which ONE of the following identifies the actions to coordinate and dispatch a team to isolate the valve?

A. OSC Team A stationed in the MCR will be sent to perform the task. The TSC/OSC will be notified that the team has been dispatched and the teams intended location and action.

B. OSC Team A stationed in the MCR will be sent to perform the task.

TSC/OSC notification is not required because the OSC Team A is being tracked as assigned to the MCR.

C After being notified of the task to be performed, the OSC staff will assign a team to perform the task. The team will be briefed, dispatched, and tracked by the OSC.

D. The OSC will be notified of the task and will dispatch a team to the MCR to be briefed prior to being sent to perform the task. The team will be tracked by the OSC.

DISTRA CTOR ANAL YSIS:

A.

Incorrect, Plausible because Team A is assigned to the Shift Manager and stationed in the MCR to be dispatched to perform task as needed but the team is limited to activities in the Control Building and in the Electrical Board Rooms.

B.

Incorrect, Plausible because Team A is assigned to the Shift Manager and stationed in the MCR to be dispatched to perform task as needed but the team is limited to activities in the Control Building and in the Electrical Board Rooms.

C.

CorrecI, In accordance with EPIP-7, the OSC will establish, brieI dispatch and track the performance of the team assigned to perform the task.

D.

Incorrect, Plausible because normally the MCR does perform the briefing for operation activities and OSC being notified and dispatching the team is correct.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Question Number:

95 Tier:

3 Group n/a K/A:

G2.1.8 Conduct of Operations Ability to coordinate personnel activities outside the control room.

Importance Rating:

3.4 /4.1 10 CFR Part 55:

41.10/45.5/45.12/45.13 IOCFR55.43.b:

4, 6 K/A Match:

KA is matched because the question requires knowledge of how to coordinate personnel activities outside the control room during conditions when Radiological Emergency Plan Implementation has resulted in the Emergency Centers being staffed. SRO only because it involves requirements in the Radiological Emergency Plan and specifics of Emergency Plan Implementing procedure.

Technical

Reference:

EPIP-7, Activation and Operation of the Operations Support Center (OSC), Revision 0028 Proposed references None to be provided:

Learning Objective:

3-OT-PCDOO48C

5. Use the WBN Emergency Plan Implementing Procedures (EPIPs).

8. Recognize how AUOs are dispatched and controlled during radiological emergencies.

Cognitive Level:

Higher Lower X

Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

96. G 2.2.43 096 While implementing a General Operating (GO) Instruction, the instruction directs an annunciator to be disabled.

In accordance with OPDP-4, Annunicator Disablement, the US can allow the CRC to disable the alarm A. after a Technical Evaluation and a 50.59 review are completed.

B. without completing either a Technical Evaluation or a 50.59 review.

C after a 50.59 review is completed, but a Technical Evaluation is NOT required.

D. after a Technical Evaluation is completed, but a 50.59 review is NOT required.

DISTRACTOR ANALYSIS:

A.

Incorrect, Plausible because OPDP-4 Appendix A section C does require both a TE and 50.59 review for plant conditions other than those listed in the stem.

B.

Correct In accordance with OPDP-4, Annunciator Disablement, Appendix A Section A, if an approved plant procedure allows an alarm disablement then an additional TE and 50.59 review are not required because the procedure has already been reviewed and approved.

C.

Incorrect, Plausible because OPDP-4 Appendix A section B can require a 50.59 review but not necessarily a TE for plant conditions than those listed in the stem.

D.

Incorrect, Plausible because OPDP-4 Appendix A section B may require a TE and but not necessarily a 50.59 review for plant conditions than those listed in the stem.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Question Number:

96 Tier:

3 Group n/a K/A:

G2.2.43 Equipment Control Knowledge of the process used to track inoperable alarms.

Importance Rating:

3.0 / 3.3 IOCFRPart55:

41.10/43.5/45.13 IOCFR55.43.b:

3 K/A Match:

KA is matched because the question requires knowledge of the process used to track inoperable alarms.

Technical

Reference:

OPDP-4, Annunciator Disablement, Rev 0004 Proposed references None to be provided:

Learning Objective:

3-OT-OPDP-4 1.

Explain the purpose of OPDP-4, Annunciator Disablement.

Cognitive Level:

Higher Lower X

Question Source:

New Modified Bank X

Bank Question History:

SQN question G 2.1.9 used on the 2009 retake exam with minor wording changes and correct answer relocated and distractor locations shuffled.

Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

97. G 2.2.7 097 Which ONE of the choices below completes the two statements in accordance with SPP-8.1, Conduct of Test?

During preparations for a drain down to establish mid-loop conditions in accordance with GO-b, Reactor Coolant System Drain And Fill Operations, the individual assigned responsibility for conducting the management expectations briefing for the test is the (1)

During the time the Unit Supervisor is directing crew actions with the drain down in progress, the individual with overall responsibility for the control of the test is the (2)

Li)

A. Shift Manager Shift Manager B CIPTE Manager Shift Manager C. Shift Manager CIPTE Manager D. CIPTE Manager CIPTE Manager

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A.

Incorrect, Plausible because the Shift Manager does have responsibilities during preparations for the test and the Shift Manager having responsibility for control of the test is correct.

B.

Correct, the CIPTE manager is the individual assigned responsibility for conducting the management expectations briefing for the test but the shift manager retains overall responsibility of the control of the test. (See below)

C.

Incorrect, Plausible because the Shift Manager does have responsibilities during preparations for the test and the CIPTE Manager does have responsibility during the test, but the CIPTE Manager responsibilities do not reduce the Shift Managers responsibility for control of the test.

D.

Incorrect, Plausible because the CIPTE Manager being responsible for the brief is correct and the CIPTE Manager does have responsibility during the test, but his/her responsibilities do not reduce the Shift Managers responsibility for control of the test.

SPP-8. I Section 3.8 E.

At the Test Directors pre-test formal briefing, the Plant Manager or designee for the test shall conduct a briefing for Operations and testing personnel on management expectations for the test utilizing Form SPP-8.1-3.

F.

The Plant Manager or his designee shall determine the need to designate a senior line manager to advise the Shift Manager or Unit Supervisor, who has the authority and experience to exercise continuous responsibility for the oversight of a particular test or evolution. This authority includes control of the pace of the CIPTE and the resolution (or escalation) of problems encountered.

NOTE This is an oversight position and shall not interfere with or reduce the Shift Managers responsibility for control of the test.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Question Number:

97 Tier:

3 Group n/a K/A:

G2.2.7 Equipment Control Knowledge of the process for conducting special or infrequent tests.

Importance Rating:

2.9 / 3.6 IOCFRPart55:

41.10/43.3/45.13 IOCFR55.43.b:

3, 4 K/A Match:

KA is match because the question requires knowledge of the process for conducting special or infrequent tests. SRO because the Unit Supervisor (an SRO) is required to have knowledge of the function of both the Shift Manager and the Manager assigned by the Plant Manager to oversee the test while the test is in progress.

Technical

Reference:

SPP-8.1, Conduct of Testing, Rev. 0006 GO-b, Reactor Coolant System Drain And Fill Operations, Rev 0042 Proposed references None to be provided:

Learning Objective:

3-OT-SPPO8O1

3. Describe the responsibilities of the supervisor, test director, and senior manager assigned to a Complex, Infrequently Performed Test or Evolution (CIPTE).

Cognitive Level:

Higher Lower X

Question Source:

New Modified Bank Bank X

Question History:

WBN bank question G 2.2.7 096 modified by changing format, one half of the question, all choices, and the correct answer relocated.

Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

98. G2.3.15 098 Given the following plant conditions:

The Unit is in Mode 5, with irradiated fuel assembly shuffles being performed in the Spent Fuel Pit.

As part of performing 0-SI-90-5, 92 Day Channel Operational Test Of The General Atomic Main Control Room Intake Radiation Monitor Loop 0-LPR-90-125 Train A, a source check is directed to be performed on the radiation monitor.

Which ONE of the following identifies...

(1) the SRO responsibility for the Tech Spec/ LCD tracking Sheet Entry while the test is being performed and (2) the effect of performing the source check portion of the test?

Note:

CREVS

- Control Room Emergency Ventilation System A. The SRO will make an LCD Tracking Sheet entry that Train A of CREVS is inoperable.

This will result in an automatic actuation of Train A of CREVS.

B. The SRO will make an LCD Tracking Sheet entry that 0-RM-90-125 is inoperable.

This will result in an automatic actuation of Train A of CREVS.

C. The SRO will make an LCD Tracking Sheet entry that Train A of CREVS is inoperable.

The Hi Rad alarm will actuate but CREVS will not be actuated during the source check.

D The SRO will make an LCD Tracking Sheet entry that 0-RM-90-125 is inoperable.

The Hi Rad alarm will actuate but CREVS will not be actuated during the source check.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A.

Incorrect, Plausible because with the radiation monitor being out of service, one of the automatic actuations of the CREVS is prevented from functioning and because if the Hi Rad relay was energized during the source check an auto actuation of the associated train of CREVS would occur if the test did not have the associated handswitch in block.

B.

Incorrect, Plausible because the SRO making the tracking entry that the radiation monitor was inoperable is correct and because if the Hi Rad relay was energized during the source check an auto actuation of the associated train of CREVS would occur if the test did not have the associated handswitch in block.

C.

Incorrect; Plausible because with the radiation monitor being out of service, one of the automatic actuations of the CREVS is prevented from functioning and because the Hi Rad relay being energized during the source check but having its output blocked to prevent the auto actuation of the associated train of CREVS is correct.

D.

Correct, This test requires blocking of the output of RM-90-125, which renders O-RM-9Q-125 inoperable requiring an LCO 3.3.7 entry and the information entered into the LCO tracking log.

The blocking will not prevent the Hi Rad relay from being energized (causing Hi RAD alarm) during the source check but will block the output of the Hi Rad relay preventing the auto actuation of the associated train of CREVS.

Question Number:

98 Tier:

3 Group n/a K/A:

G2.3.15 Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Importance Rating:

2.9 I 3.1 IOCFRPart55:

41.12/43.4/45.9 IOCFR55.43.b:

2, 4 KIA Match:

KA is matched because the question requires knowledge of fixed radiation monitoring systems. it is SRO because in additional to knowledge of the system, the question requires knowledge of the surveillance instruction, the affect of performance of the instruction has on Tech Spec requirements, and the SRO responsibility for tracking the Tech Spec actions entered.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Technical

Reference:

SOl-90.02, Gaseoud Process Radiation Monitors, Revision 0046 Tech Spec LCO 3.3.7, Table 3.3.7-1, Amendment 41 0-S 1-90-5, 92 Day Channel Operational Test Of The General Atomic Main Control Room Intake Radiation Monitor Loop 0-LPR-90-125, Rev 13 Proposed references None to be provided:

Learning Objective:

3-OT-T/S0303

3. Given plant parameters/conditions, correctly determine applicable Limiting Conditions for Operation or Technical Requirement limits for the various instrumentation systems covered by T/S or T/R.

3-OT-OPDP-8

03. Identify the responsibilities of the Unit Supervisor described in OPDP-8, Limiting Condition for Operation Tracking.

Cognitive Level:

Higher X

Lower Question Source:

New Modified Bank X

Bank Question History:

Question modified from question used on WBN 2008 exam by changing the radiation monitor, rewoding the stem by swapping which half of the question was asked first, relocating choices including the correct answer Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010

99. G 2.3.5 099 Given the following:

A High Energy Line Break (HELB) occurs inside Unit 1 containment.

Neither Containment Air Return fan can be started.

Containment High Range Radiation Monitors readings are increasing.

In accordance with EPIP-1, Emergency Plan Implementing Procedures, which ONE of the following identifies...

(1) the Barrier Matrix EAL criterion that will be reached due to the increasing radiation levels on the Containment High Range Radiation monitors and (2) how the failure of the Containment Air Return fans to start affects the accuracy of the monitors?

A. (1) Fuel Clad (2) Monitors indicate lower than actual.

B. (1) Containment (2) Monitors indicate lower than actual.

C (1) Fuel Clad (2) Monitors indicate higher than actual.

D. (1) Containment (2) Monitors indicate higher than actual.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A.

Incorrect Plausible because the Fuel Clad Barrier being the first barrier to be challenged is correct and how the indication is affected by the failure of the containment air return fans is the reverse of what actualily occurs.

B.

Incorrect, Plausible because the Containment Barrier is a barrier challenged by the increasing radiation being detected on the monitors and how the indication is affected by the failure of the containment air return fans is the reverse of what actually occurs.

C.

Correct, the Fuel Clad would be the first barrier to be challenged (setpoints 59r/hr &

74r/hr) as compared to the Containment Barrier (setpoints 86R/hr &104R/hr) and the failure of the Containment Air Return fans will result in the monitor indicating higher than actual level due to the containment temperature remaining elevated D.

Incorrect, Plausible because the containment Barrier is a barrier challenged by the increasing radiation being detected on the monitors and the affect on the Thdication is correct.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Question Number:

99 Tier:

3 Group n/a K/A:

G2.3.5 Radiation Control Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Importance Rating:

2.9 / 2.9 10 CFR Part 55:

41.11 /41.12/43.4/45.9 IOCFR55.43.b:

4 K/A Match:

KA is matched and the question is SRO only because the question requires both the ability to use fixed radiation monitoring displays accordance with Radiological Emergency Plan Implementing Procedures and the knowledge of how an equipment failure would affect the indications on the radiation monitors Technical

Reference:

EPIP-1, Emergency Plan Implementing Procedures, Revision 33 Proposed references None to be provided:

Learning Objective:

3-OT-PCDO48C 1.

Classify emergency events 5.

Use the WBN Emergency Plan Implementing Procedures (EPIPs).

Cognitive Level:

Higher Lower X

Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 100. G2.4.44 100 Which ONE of the following identifies the provisions for making a Protective Action Recommendation (PAR) following the declaration of an event in accordance with the Radiological Emergrency Plan?

General Emerciency Site Area Emercency A.

Required Optional B

Required NOT allowed C.

Optional Optional D.

Optional NOT allowed DISTRACTOR ANALYSIS:

A.

Incorrect Plausible because a Protection Action Recommendations being required at the General Emergency Level is correct and there are conditions that are required at one level and optional at a different level O.e.

Assembly and Accountability).

B.

Correct, Protection Action Recommendations are required at the General Emergency Level and are not allowed during a Site Area Emergency. The EPIP for reporting a General Emergency identifies that a PAR is required and the only option on the SAE form for Protection Action Recommendation is None.

C.

Incorrect, Plausible because there are conditions that have actions that optional at a different level 0. e.

Assembly and Accountability, staffing the emergency centers).

D.

lncorrect Plausible because there are conditions that have actions that optional at a different level (i.e.

Assembly and Accountability, staffing the emergency centers) and not being allowed at the SAE level is correct.

08/2010 Watts Bar SRO NRC Exam as Submitted 7/2/2010 Question Number:

100 Tier:

3 Group n/a KIA:

G2.4.44 Emergency Procedures / Plan Knowledge of emergency plan protective action recommendations.

Importance Rating:

2.4 / 4.4 10 CFR Part 55:

41.10/41.12/43.5 /45.11 IOCFR55.43.b:

4, 6 K/A Match:

KA is matched because the question requires knowledge of emergency plan protective action recommendations.

Technical

Reference:

EPIP-5, General Emergency, Rev 37 EPIP-4, Site Area Emergency, Rev 31 Proposed references None to be provided:

Learning Objective:

3-OT-PCDO48C 5.

Use the WBN Emergency Plan Implementing Procedures (EPIPs).

Cognitive Level:

Higher Lower X

Question Source:

New X

Modified Bank Bank Question History:

New question Comments:

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