ML100410117

Initial Exam 2009-302 Draft SRO Written Exam

ML100410117
Person / Time
Site:	Harris
Issue date:	01/05/2010
From:	NRC/RGN-II
To:
References
IR-09-302
Download: ML100410117 (54)

Text

H~R~1s Dra~T _5 RD HqR~l QUESTIONS REPORT for 2009B NRC SRO QUESTIONS DRAFT

1. Given the following plant conditions:

- The plant was operating at 100% power with 'A' Safety Train Equipment in service

- A fault occurred on Unit Aux Transformer 'A'

- The bus transfer for 6.9KV Aux Bus 'A' and 'D' failed to occur

- The 'A' EDG failed to start

- The crew is performing EPP-004, Reactor Trip Response, and AOP~025, AOP-025, Loss of One Emergency AC Bus (6.9KV) or One Emergency DC Bus (125V)

The following annunciators are locked in:

- ALB-05-8-2, CCW Pump B Disch Header Low Press

- ALB-08-2-1, RCP Seal Water Injection Low Flow Which ONE of the following identifies THE PRIORITY that the condition which caused these annunciators will be addressed lAW AOP-025 AND whether the annunciator is EXPECTED or NOT EXPECTED for the plant conditions?

A. 1. CCW Pump B Disch Header Low Press Expected

2. RCP Seal Water Injection Low Flow NOT Expected B. 1. CCW Pump B Disch Header Low Press NOT Expected

2. RCP Seal Water Injection Low Flow Expected C~ 1. RCP Seal Water Injection Low Flow C,.. Expected

2. CCW Pump B Disch Header Low Press NOT Expected D. 1. RCP Seal Water Injection Low Flow NOT Expected

2. CCW Pump B Disch Header Low Press Expected Tuesday, October 20,200910:29:14 20, 2009 10:29: 14 AM 1

QUESTIONS REPORT for 20098 NRC SRO QUESTIONS Plausibility and Answer Analysis

~' Train in service and )\'

With )\' ~' EDG failing to start, RCP Seal Water Inj Low Flow is expected because the '8' CSIP will not receive an Auto Start. The CCW Pump 8 Discharge Header Low Pressure is NOT expected because the 8 CCW pump should have started on low pressure. AOP-025 addresses the CSIP before addressing the CCWPump.

CCW pump auto-start signals are from sequencer UV or SI (load block 4). The standby pump auto-starts on low discharge pressure of 52 psig sensed on its respective discharge header (PT 649 or 650).

A. Incorrect. Wrong priority and wrong expectation.

B.

8. Incorrect. Wrong priority and right expectation.

C. Correct.

D. Incorrect. Right priorty and wrong expectation.

007 EG2.4.45 007 Reactor Trip 2.4 Emergency Procedures / Plan 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10/ 43.5 / 45.3/ 45.3 / 45.12)

Importance Rating: 4.1 4.3 Technical

Reference:

AOP-025, Rev 25, pg 8 and 9 References to be provided: None Learning Objective: AOP-LP-3.25 Obj. 6 Question origin: NEW Comments: (KIA Match) Applicant must interpret the alarm to determine if expected or NOT expected and then establish the priority in the procedure Tier/ Group: T1 GG11 SRO justification: Must recognize that the plant has tripped on loss of Prioritizing annunciators is an SRO task and requires knowledge of step sequence in AOP-025 Tuesday, October 20, 20,200910:29:14 2009 10:29:14 AM 2

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

2. Given the following plant conditions:

- A Reactor Trip and Safety Injection have occurred due to a Small Break LOCA

- The crew has transitioned to EPP-009, Post LOCA Cooldown and Depressurization The following conditions currently exist:

- CNMT Pressure is 7.8 psig and increasing

- RCS Subcooling is 53°Fand 53°F and increasing

- Pressurizer level is 12% and increasing

- 'A' Train CNMT Phase A will NOT reset

- Nitrogen has been established to CNMT Which ONE of the following identifies the action that will be taken to depressurize the RCS AND the Pressurizer level required to secure the depressurization in accordance with EPP-009?

Action Taken to Depressurize the RCS Required Pressurizer Level A. Open Normal Spray Valves 30%

B. Open ONE Pressurizer PORV 30%

C. Open Normal Spray Valves 40%

D~ Open ONE Pressurizer PORV D¥' 40%

Plausibility and Answer Analysis A. Incorrect. 30% is plausible because it is the SI reinitiation criteria value of EPP-009. Normal Spray is plausible because it is desired and nothing indicates RCPs are not inservice but with ~ Train CNMT Phase A NOT reset, the spray I

valves will not work due to loss of Instrument Air.

B. Incorrect. 30% is plausible because it is the SI reinitiation criteria value of EPP-009. ONE Pressurizer PORV is correct due to the loss of IA to the Spray Valves.

C. Incorrect. 40% is correct. Normal Spray is plausible because it is desired and nothing indicates RCPs are not inservice but with ~ Train CNMT Phase A NOT I

reset, the spray valves will not work due to loss of Instrument Air.

D. Correct. 40% is correct. ONE Pressurizer PORV is correct due to the loss of IA to the Spray Valves.

Tuesday, October 20,200910:29:14 AM 4

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS 009 EA2.11 009 Small Break LOCA EA2 Ability to determine or interpret the following as they apply to a small break LOCA:

(CFR 43.5 / 45.13)

EA2.11 Containment temperature, pressure, and humidity Importance Rating: 3.8 4.1 Technical

Reference:

EPP-009, Rev.14, Page 18 References to be provided: None Learning Objective: EOP-LP-3.5 Obj 2 Question Origin: NEW Comments: (KIA match) Match KIA because candidate must determine that CNMT pressure is sufficient to require Adverse Values during a Small Break LOCA and how that affects the actions of the procedure in affect.

Tier/Group: T1 G1 SRO Justification Requires detailed knowledge the steps contained in EPP-009. 30% is the adverse value for SI reinitiation and 40% is the adverse value for refilling the PZR. The applicant must know the steps in EPP-009 and that the pressurizer is refilled to greater than 40% to provide margine to SI reinitiation.

Tuesday, October 20,200910:29:14 20, 2009 10:29:14 AM 5

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

3. Given the following plant conditions:

- A Reactor Trip and Safety Injection have occurred

- All MSIVs failed to close

- The crew has transitioned to EPP-015, Uncontrolled Depressurization of All Steam Generators from EPP-014, Faulted Steam Generator Isolation

- The RO has just secured one CSIP as part of Safety Injection termination when the BOP reports that 'A' MSIV has been closed The following parameters are observed:

- RCS pressure is 1525 psig and increasing

- Cold Leg temperatures are 472°F and lowering

- 'A' SG pressure is 510 psig and lowering

- 'B' SG pressure is 515 psig and increasing

- 'C' SG pressure is 590 psig and lowering Which ONE of the following identifies the location of the faultAND the action required?

Steam Line Fault Location Action Required A. 'A' Immediately transition to EPP-014 and isolate 'A' SG B~

B!'" 'A' Remain in EPP-015 and terminate Safety Injection C. 'C' Immediately transition to EPP-014 and isolate 'A' SG D. 'C' Remain in EPP-015 and terminate Safety Injection Plausibility and Answer Analysis

~' Steam Line is the location of the break. Its MSIV has been shut but pressure

'A' continues to lower with saturation temperature.

'C' Steam Line MSIV is open and its pressure is ONL Y lowering to saturation temperature. 'C' Steam Line cannot be the location of the break because with 'B' '8' and

'C' MSIVs open, it is connected to the 'B'

'8' Steam Line and its pressure is increasing.

The Caution in EPP-015 prior to SI 81 termination states that EPP-014 transition criteria is not applicable during SI termination.

A. Incorrect. Correct Steam Line but wrong Procedure Action B.

8. Correct. Correct Steam Line and Correct Procedure Action C. Incorrect. Wrong Steam Line and wrong Procedure Action D. Incorrect. Wrong Steam Line but Correct Procedure Action Tuesday, October 20,200910:29:14 AM 7

QUESTIONS REPORT for 20098 NRC SRO QUESTIONS 040 AA2.01 040 Steam Line Rupture AA2. Ability to determine and interpret the following as they apply to the Steam Line Rupture:

(CFR: 43.5/

43.5 / 45.13)

AA2.01 Occurrence and location of a steam line rupture from pressure and flow indications Importance Rating: 4.2 4.7 Technical

Reference:

EPP-015, Rev. 19, Caution on page 21 and foldout item.

References to be provided: None Learning Objective: EOP-LP-3.9, Obj. 9.c Question Origin: NEW Comments: (KIA match) Must diagnose the location of break, using information regarding MSIVs and SG pressures versus RCS temperature.

Tier/Group: T1 G1 SRO Justification Requires knowledge of EPP-015 implementation strategies including when foldout criteria are not applicable and procedural transition prioiritization based on plant conditions (terminate SI or Go to EPP-014).

Tuesday, October 20,200910:29:14 AM 8

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

4. Given the following plant conditions:

- The plant is operating at 89% power

- 'B' MFW Pump tripped

- The crew carried out the actions of AOP-01 0, Feedwater Malfunctions

- The plant has been stabilized at 55% reactor power

- The following annunciators are lit:

- ALB-013-8-2, Bank Low Insertion Limit

- ALB-013-8-3, Bank Low-Low Insertion Limit Which ONE of the following identifies the action required in accordance with Technical Specifications AND the basis for the Rod Insertion Limits (RILs)?

Required Action Basis is to ensure SDM for A'I Restore rods to above RILs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Postulated Steam Line Break B. Immediately initiate Emergency Boration Postulated Steam Line Break C. Restore rods to above RILs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Boron Dilution Accident D. Immediately initiate Emergency Boration Boron Dilution Accident Plausibility and Answer Analysis Tech Spec 3.1.3.6 and bases for 3.1.1.1 A. Correct. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> are allowed to restore Bank 0 D above the insertion limits. In modes 1 and 2, SDM of 1770 pcm is required for the Postulated Steam Line Break.

B. Incorrect. Emergency Boration is required when SDM is not adequate but TS 3.1.3.6 allow 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore. Postulated Steam Line Break is correct.

C. Incorrect. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> are allowed to restore Bank 0D above the insertion limits. Boron Dilution Accident is plausibe because this is Mode 3-5 basis.

D. Incorrect. Emergency Boration is required when SDM is not adequate but TS 3.1.3.6 allow 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore. Boron Dilution Accident is plausibe because this is Mode 3-5 basis.

Tuesday, October 20, 20,200910:29:14 2009 10:29: 14 AM 10

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS 054 AG2.2.40 054 Loss of Main Feedwater (MFW) 2.2 Equipment Control 2.2.40 Ability to apply Technical Specifications for a system.

(CFR: 41.10/

41.10 / 43.2 / 43.5 / 45.3)

Importance Rating: 3.4 4.7 Technical

Reference:

Tech Spec 3.1.3.6 page 3/4 1-21 Tech Spec Bases section 3.1 page B 3/4 3/41-1 1-1 References to be provided: None Learning Objective: Rod Control Student Text, Obj. 15.d Question origin: NEW Comments: HNP does not have a TS associated with MFW or loss of MFW. For this reason the question requires application of the Rod Insertion Limit TS following a partial loss of MFW (1 MFP) and the associated runback.

Tier/Group: T1 G1 SRO justification: Requires knowledge of a TS with actions greater than one hour and bases.

Tuesday, October 20,200910:29:14 20.2009 10:29:14 AM 11

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

5. Given the following plant conditions:

- The crew is performing EPP-001, Loss of AC Power to 1A-SA and 1B-SB Buses

- CNMT pressure is 1.3 psig and increasing

- RCS pressure is 1875 psig and lowering

- Pressurizer level is 18% and lowering

- Core Exit Thermocouples are 56]oF and stable

- Cold Leg temperatures are 535°F and stable

- ALL SG pressures are 940 psig and stable

- ALL SG NR levels are 23% and lowering Which ONE of the following identifies the mechanism by which the RCS is being EPP-001 ?

cooled AND the NEXT action required by EPP-001?

Cooling Mechanism Action Required A. Reflux Boiling Commence SG depressurization B. Reflux Boiling Establish 25% NR level in at least ONE SG C~ Natural Circulation C!' Establish 25% NR level in at least ONE SG D. Natural Circulation Commence SG depressurization Plausibility and Answer Analysis The background document discusses these two mechanisms for cooling to occur (Natural Circulation or Reflux Boiling). With the parameters given, Natural Circulation is occurring. Subcooling is approximately 60°F, therefore reflux boiling is NOT occurring.

A listing of parameters can be found in EPP-004, Rev 18, page 32.

A. Incorrect. Subcooling is approximately 60°F, therefore reflux boiling is NOT occurring. The depressurization should NOT be started until 25% NR level is established in at least ONE SG.

B. Incorrect. Subcooling is approximately 60°F, therefore reflux boiling is NOT occurring. The depressurization should not be started until SG NR level of 25%

has been established in at least ONE SG.

C. Correct. With the parameters given, Natural Circulation is occurring. Establishing 25% NR in at least one SG is required prior to SG depressurization.

D. Incorrect. With the parameters given, Natural Circulation is occurring. The depressurization should not be started until SG NR level of 25% has been established in at least ONE SG.

Tuesday, October 20,200910:29:15 AM 13

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS 055 EA2.02 055 Loss of Offsite and Onsite Power (Station Blackout)

EA2 Ability to determine or interpret the following as they apply to a Station Blackout:

(CFR 43.5/45.13)

EA2.02 ReS RCS core cooling through natural circulation cooling to S/G cooling Importance Rating: 4.4 4.6 Technical

Reference:

EPP-001 rev 31, page 33 ECA 0.0 rev 1C, page 54 (WOG ERG for EPP-001)

References to be provided: None Learning Objective: EOP-LP-3.7, Obj. 6 Question origin: NEW Comments: (KIA Match) Candidate must use plant parameters to determine that Natural Circulation is occurring.

Tier/Group: T1G1 T1 G1 SRO justification: Requires detailed knowledge of how the procedure is to be implemented, including the RNO for when adequate SG inventory does not exist.

Tuesday, October 20,200910:29:15 20, 2009 10:29: 15 AM 14

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

6. Given the following plant conditions:

- The plant is in Mode 3

- Instrument Bus IDP-1A-SI was de-energized due to a fault on the bus Which ONE of the following identifies the Technical Specification required action for the de-energized Instrument Bus?

Re-energize the Instrument Bus within _-.\,(...:...1

_-.1(,.!..1,L..) _ or be in COLD SHUTDOWN within

)1.----

the following m fgl A't (1 ) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> A'!'

(2) 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> B. (1) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (2) 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. (1)

(1 ) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (2) 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> D. (1 ) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (2) 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Plausibility and Answer Analysis 3.8.3. 1 Action b allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to reenergize an instrument bus or be in Hot Tech Spec 3.B.3.

STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> A. Correct answer. The plant is already in Mode 3 so the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> can not be credited B. Incorrect but plausible because this would be correct if the plant was in Mode 1 or 2.

C. Incorrect but plausibe because 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is the allowed restoration time for Action C.

D. Incorrect but plausible because 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is the allowed restoration time for Action C and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> would be correct if the plant was in Mode 1 or 2.

Tuesday, October 20,200910:29:15 AM 16

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS 057 AG2.2.38 057 Loss of Vital AC Electrical Instrument Bus 2.2 Equipment Control 2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7/41.10/43.1 /45.13)

Importance Rating: 3.6 4.5 Technical

Reference:

Tech Spec 3.8.3.1 action b; page 3/48-17 3/4 8-17 References to be provided: None Learning Objective: Student Text 120V UPS, Obj.15 Question Origin: NEW Comments: Original KA provided by NRC was 057 AG2.2.4.

Replacement KA is 057 AG2.2.38 (KIA match) This question requires knowledge of the tech spec actions for the instrument bus power supply and the application of those limitations to the current plant mode.

Tier/Group: T1G1 SRO SRO Justification Requires an in-depth knowledge of applying Tech Specs actions that are significantly greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Tuesday, October 20,200910:29:15 20,2009 10:29:15 AM 17

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

7. Given the following plant conditions:

- The plant is operating at 2% power in accordance with GP-004, Reactor Startup (Mode 3 to Mode 2)

- Both Intermediate Range Nuclear Instruments (Nls) have been declared inoperable lAW Technical Specifications and AP-617, Reportability Determination and Notification, which ONE of the following choices completes the below statement?

Restore at least one Intermediate Range to an OPERABLE status within (1) OR initiate a reactor shutdown and notify the NRC of the shutdown initiation within the next (2)

A'! (1 ) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> A'I (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. (1 ) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. (1 ) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> D. (1 ) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Plausibility and Answer Analysis TS 3.0.3 is applicable therefore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed by technical specifications prior to initiating a shutdown. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is plausible from specifications for instrument buses or DC buses.

Shutdown initiation is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is plausible because Violation of Operating License Conditions is a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report A. Correct. Right restoration time and right report time.

B. Incorrect Right restoration time but Wrong report time.

C. Incorrect. Wrong restoration time and Wrong report time.

D. Incorrect. Wrong restoration time but right report time.

Tuesday, October 20,2009 20,200910:29:15 10:29:15 AM 19

QUESTIONS REPORT for 2009B 20098 NRC SRO QUESTIONS 033 AG2.4.30 033 Loss of Intermediate Range Nuclear Instrumentation 2.4 Emergency Procedures / Plan 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

(CFR: 41.10/43.5/45.11) 41.10/ 43.5 / 45.11)

Importance Rating: 2.7 4.1 Technical

Reference:

AP-617 rev 26, page 13 Tech Spec 3.3.1 References to be provided: None Learning Objective: Student Text Nuclear Instruments, Obj. 12 PP-LP-2.17, Obj 5 Question Origin: NEW Comments: (KIA match) This question requires the candidate to apply tech spec requirements to the loss of both IR Ni's and also to apply Reportability requirements for that failure.

Tier/Group: T1 G2 SRO Justification Requires an in-depth knowledge of the instrumentation Tech Specs and the NRC reportability requirements of AP-617 which is an SRO responsibility.

Tuesday, October 20,200910:29:15 AM 20

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

8. Given the following plant conditions:

- The plant is operating at 100% power

- RM-01 CR-'3561 CSA, Containment Ventilation Isolation (CVI) Radiation Monitor failed 3 days ago and OWP-RM-02 has been implemented andOWP-RM-02

- The BOP is performing the Technical Specification required channel check on the remaining operable CVI Radiation Monitors Which ONE of the following identifies the normal indication for CVI Radiation Monitors and the Technical Specification required action if another CVI Radiation Monitor fails its channel check?

Normal Indication Required Action A'I 110 mRem/hour A'! CLOSE the CNMT Purge Makeup and Exhaust Isolation Valves B. 110 mRem/hour Restore ONE monitor within 7 days or submit a Special Report C. 1 Rem/hour Restore ONE monitor within 7 days or submit a Special Report D. 1 Rem/hour CLOSE the CNMT Purge Makeup and Exhaust Isolation Valves Plausibility and Answer Analysis A. Correct. 110 mRem/hour mRemihour is the normal reading for the RM-3561s, CNMT Ventilation Isolation Rad Monitors at 100% power. Per TS table 3.3-6 action 27-27 -

CLOSE the CNMT Purge Makeup and Exhaust Isolation Valves is the action for when two RM-3561 s, CNMT Ventilation Isolation Rad Monitors are inoperable.

B.

8. Incorrect. 110 mRem/hour mRemihour is the normal reading for the RM-3561s, CNMT Ventilation Isolation Rad Monitors at 100% power. Second part is plausible because the 7 days or special report per TS 3.3.3.6 action c pertains to another Containment radiation monitor - the High Range Radiation Monitors.

C. Incorrect. 1 Remlhour Rem/hour is the normal indication for RM-01CR-3589SA and S8, CNMT Post Accident High Range Rad Monitors. This monitor has an idling current that maintains the reading at 1Remlhour Rem/hour in low radiation fields. Second part is plausible because the 7 days or special report per TS 3.3.3.6 action c pertains to another Containment radiation monitor - the High Range Radiation Monitors.

D. Incorrect. 1 Remlhour Rem/hour is the normal indication for RM-01CR-3589SA and S8, CNMT Post Accident High Range Rad Monitors. This monitor has an idling Rem/hour in low radiation fields. Per TS current that maintains the reading at 1Remlhour table 3.3-6 action 27 - CLOSE the CNMT Purge Makeup and Exhaust Isolation Valves is plausible because this is the action for when two RM-3561s, CNMT Ventilation Isolation Rad Monitors are inoperable.

Tuesday, October 20,200910:29:15 AM 22

QUESTIONS REPORT for 20098 NRC SRO QUESTIONS 061 AA2.02 061 Area Radiation Monitoring (ARM) System Alarms AA2. Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms:

(CFR: 43.5/45.13) 43.5 / 45.13)

AA2.02 Normal radiation intensity for each ARM system channel Importance Rating: 2.9 3.2 Technical

Reference:

OWP-RM-02 Rev. 29 page 10, Tech Spec 3.3.3.1 Table 3.3-6 action 27 References to be provided: None Learning Objective: Student Text Radiation Monitoring System, Obj. 11 Question Origin: NEW Comments: (KIA match) Candidate must recall the normal reading for RM-01 CR-3561's and apply appropriate actions to meet Tech Spec requirements.

Tier/Group: T1 G2 SRO Justification Requires knowledge of a Tech Spec action that is greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OWP-RM-D2 OWP-RM-02 Sbeet1af4 Sheet 1 of4 EIR Number: _ _ __

W/ONumber:

WID Number: -

_- __ - _-

1. OWP-RM-02 Clearance Number: _ _ __

---

2 System: Radiation, and Explosive Gas Monitoting E1ffuent, aM Radimion, Emuent,

3. Component Contm!1illlm Containment VenlilaDon Ventilation l:sotatioo Radiation Monitors

{Area} RadiatiiOn isol:atlon (Area)

4. Scope: Maintenance on Conl:ainmem V~ IsotaIion (Area)

RadlaDoo MOriErs: (cii"de one)

Rml1CR-3561 .* RM=01CR-3561DSB,

5. Applicable Requiremems: 3.3.2, 3.3.3.1 (MOOE14, 51, 3.R!} (MODE 6) 5.

IOWP-RM it} of91 Page 100f911 I Tuesday, October 20,200910:29:15 20, 2009 10:29:15 AM 23

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

9. Given the following plant conditions:

- The crew is performing EPP-006, Natural Circulation Cooldown with Steam Void in Vessel with RVLlS

- The CRS has directed the RO to place LTOPSLTOPS in service prior to further cooldown In accordance with Technical Specifications, (1) is the lowest allowed RCS Temperature prior to placing LTOPS in service. The capacity of LTOPS protects the RCS in the event of (2) starting and injecting into a solid RCS.

A. (1)(1 ) 351°F (2) ONE CSIP B. (1 ) 351°F (1)

(2) BOTH CSIPs C. (1 ) 326°F (2) BOTH CSIPs (1 ) 326°F D!' (1)

(2) ONE CSIP Plausibility and Answer Analysis L TOPS is required to be OPERABLE with RCS temperature less than or equal to 325°F. Therefore, 326°F is the lowest allowed temperature. Bases for L TOPS on page B 3/4 4-14 states "the start of a charging/safety injection pump".

A. Incorrect. 350°F is plausible because less than 350°F represent the change to Mode 4. ONL ONLY Y ONE CSIP is correct B. Incorrect. 350°F is plausible because less than 350°F represent the change to Mode 4.

C. Incorrect. 326°F is correct but BOTH CSIPs is not.

D. Correct. 326°F is correct and ONLY ONL Y ONE CSIP is correct.

Tuesday, October 20,200910:29:15 20, 2009 10:29:15 AM 26

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS WE10 EA2.2 E10 Natural Circulation with Steam Void in Vessel with/without RVLlS EA2. Ability to determine and interpret the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLlS)

(CFR: 43.5/ 45.13)

EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Importance Rating: 3.4 3.9 Technical

Reference:

EPP-006 rev 11, page 16 Tech Spec 3.4.9.4 and Bases (page 3/4 4-40 and B 3/4 4-14)

References to be provided: None Learning Objective: Student Text Pressurizer Pressurizer Pressur Control, Obj.12 Question Origin: NEW Comments: (KIA match) EPP-006 is in progress and the candidate must ensure compliance with EPP-006 and Tech Specs.

Tier/Group: T1 G2 SRO Justification Requires knowledge of the Tech Spec Bases for LTOPS.

LTOPS.

Tuesday, October 20,200910:29:15 AM 27

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

10. Given the following plant conditions:

- The plant is operating at 100% power

- A 12 gpm tube leak exists in the 'B' SG

- AOP-016, Excessive Primary Plant Leakage, is in progress.

- ERFIS has just failed and NIT is investigating The following occurs at 1216:

- ALL MCB annunciation fails

- The 'A' RCP trips resulting in a Reactor Trip

- One Safety on the 'A' SG opens on the Reactor Trip and fails to reseat

- The GFFD GFFO reading is 2.0EO CPM The time is now 1232 and the following conditions exist:

- The GFFD GFFO is reading 2.5E6 CPM

- Safety Injection and Main Steam Line Isolation are actuated

- Phase A Isolation valves are properly positioned Which ONE of the following is the classification for this event in accordance with PEP-110, Emergency Classification?

(References Provided)

A. Alert EAL 2-1-2 B. Site Area Emergency EAL 2-1-3 C. Alert EAL 6-1-2 D~ Site Area Emergency EAL 6-1-3 0,.-

Plausibility and Answer Analysis A. Incorrect. Plausible because the fuel is breached B. Incorrect. Plausible because the fuel is breached and CNMT could be determined to be breached if the applicant fails to recognize that the SG tube leak and the stuck open safety valve have occurred on different SGs. Primary to Secondary leak of> 10 gpm and the affected SG Safety valves is not shut results in CNMT breached. 2 FPBs breached or jeopardized is EAL 2-1-3 C. Incorrect. Plausible because if ERFIS data availability or significant transient in progress was answered incorrectly, the EAL classification would be an ALERT 6-1-2.

D. Correct. With a loss of> 75% of MCB annunciators, a loss of ERFIS, a significant transient and the annunciators lost for> 15 minutes the EAL classification is Site Area Emergency 6-1-3.

Tuesday, October 20,200910:29:15 20,2009 10:29:15 AM 29

QUESTIONS REPORT for 20098 NRC SRO QUESTIONS 003 G2.4.41 003 Reactor Coolant Pump System (RCPS) 2.4 Emergency Procedures / Plan 2.4.41 Knowledge of the emergency action level thresholds and classifications.

(CFR: 41.10 41.10// 43.5 / 45.11) 45.11 )

Importance Rating: 2.9 4.6 Technical

Reference:

EAL Flowpath, Rev. 09-01 EP-EAL, Rev 7, Page 54 References to be provided: EAL Flowpath, Rev. 06-01 Learning Objective: EOP-SIM-3.00, Obj. 6 Question Origin: NEW Comments: (KIA match) The Trip of the Rep RCP has initiated a transient with loss of annunciators that must be properly classified.

Tier/Group: T2G1 SRO Justification Event classification is only a requirement for Senior Reactor Operators Operators..

• ~ 21 Tuesday, October 20, 20,200910:29:15 2009 10:29:15 AM 30

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

11. Given the following plant conditions:

- The Main Control Room has been evacuated due to toxic gas

- AOP-004, Remote Shutdown, is in progress

- RCS temperature is 330°F

- 'A' CCW Pump is under clearance

,;- 'B' CCW Pump is in service

- 'B' Train RHR was placed in service for shutdown COOling cooling but has tripped

- The crew is preparing to place 'A' Train RHR in service _

- 1CC-167, CCW From RHR HX B-SB, is OPEN

- 1CC-147, CCW From RHR HXA-SB, HX A-SB, is CLOSED Which ONE of the following identifies the required sequence for the actions to align CCW to the 'A' RHR HX AND the resultant operational effect of this sequence?

At! CLOSE 1CC-167 at the Auxiliary Transfer Panel-SB then OPEN 1CC-147 locally A't This sequence prevents damage to the ONLY running CCW pump B. CLOSE 1CC-167 1CC-167 at the Auxiliary Transfer Panel-SB then OPEN 1CC-147 1CC-14 7 locally This sequence minimizes pressure transients on relief valves in the CCW system C. OPEN 1 CC-147 locally then CLOSE 1 1CC-147 CC-167 at the Auxiliary Transfer Panel-SB 1CC-167 This sequence prevents damage to the ONLY running CCW pump D. OPEN 1CC-147 locally then CLOSE 1CC-167 at the Auxiliary Transfer Panel-SB This sequence minimizes pressure transients on relief valves in the CCW system Tuesday, October 20,200910:29:15 AM 31

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS Plausibility and Answer Analysis Aligning both RHR HXs throught the Non-Essential Supply and Return Valves with ONL Y one CCW pump in service will result in runout conditions and damage to the CCW Pump. Therefore, 1CC-167 must be closed before 1CC-147 is opened to prevent damage.

Pressure transients have occurred at Harris resulting in relief valves lifting and P&Ls have been added to OP-14S requiring supply valves to be shut prior to shutting return valves. This supports the plausibility of minimizing pressure transients A. Correct. Right sequence to prevent runoutldamage B. Incorrect answer. Right Sequence but wrong reason.

C. Incorrect. Wrong sequence but right reason D. Incorrect. Wrong sequence and wrong reason Tuesday, October 20,200910:29:15 20, 2009 10:29: 15 AM 32

QUESTIONS REPORT for 2009B 20098 NRC SRO QUESTIONS 005 G2.4.34 005 Residual Heat Removal System (RHRS)

( 2.4 Emergency Procedures / Plan i2.4.34 Knowledge of RO tasks performed outside the main control room during an 2.4.34 emergency and the resultant operational effects.

41.10/43.5/45.13)

(CFR: 41.10 / 43.5 / 45.13)

\

Importance Rating: 4.2 4.1 Technical

Reference:

AOP-004 rev 46, page 81 and 106 OP-145 rev 57 page 10 P&L 27 References to be provided: None Learning Objective: AOP-LP-3.4 Obj. 7 Question Origin: NEW Comments: (KIA match) Question requires the applicant to have a designated for the RO outside the knowledge of tasks deSignated MCR during remote shutdown (AOP-004) including the operational implication or the reason for the proper sequence.

Tier/Group: T2G1 SRO Justification AOP-004 provides specific directions for establishing '8'

'B' RHR cooling because this is the preferred training.

Train RHR*cooling ONLY a note is provided on page 81 about allowance for

'A' RHR train to be used. Additionally a failure has occurred the candidate must use knowledge of the procedure and P&Ls of OP-145 to determine the success path.

Tuesday, October 20, 2009 10:29:16 AM 33

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

12. Given the following plant conditions:

- The plant is operating at 100% power

- Excess Letdown is in service in preparation for removing Normal Letdown from service The following occur:

- Control Bank '0' 'D' begins to step in

- Tavg is 591°F and lowering

- Reactor Power is 100.05%

Which ONE of the following identifies the CCW malfunction that has resulted in these.

these conditions AND the action required?

1CC-337 (TCV-144) Letdown TCV A. Instrument Air has been lost to 1CC-337 Enter AOP-003, Malfunction of Reactor Makeup Control and isolate the Normal Letdown Heat Exchanger .

B~ Instrument Air has been lost to 1CC-337 Bl'" 1CC-337 (TCV-144) Letdown TCV Use OP-131.01, Main Turbine, section 5.3, Power Corrections to lower power Exchang~r C. A leak has occurred in the Excess Letdown Heat Exchanger Use OP-131.01, Main Turbine, section 5.3, Power Corrections to lower power D. A leak Exce~s Letdown Heat Exchanger leak has occurred in the Excess Enter AOP-014 and Isolate the Excess Letdown Heat Exchanger Tuesday, October 20,200910:29:16 20, 2009 10:29:16 AM 35

QUESTIONS REPORT for 20098 NRC SRO QUESTIONS Plausibility and Answer Analysis A. Incorrect answer. Malfunction is correct but action is incorrect. AOP-003 is plausible because a dilution has occured and the makeup system is utilized for RCS boron control.

B. Correct. With plant conditions, a dilution has occurred as a result a low temp in the demins when 1CC-337 failed open. OP-131.01, Main Turbine, section 5.3, Power Corrections must be used to lower power.

C. Incorrect. Wrong malfunction. This is plausibe as Excess UD has just been placed in service. But with present plant pressure RCS would leak to CCW.

CCW And no dilution would occur. Correct Action to lower power.

D. Incorrect. Wrong malfunction. This is plausibe as Excess UD has just been placed in service. But with present plant pressure RCS would leak to CCWCCw. And no dilution would occur. AOP-014 is plausible because it addresses the excess UDHX.

Tuesday, October 20,200910:29:16 20, 2009 10:29:16 AM 36

QUESTIONS REPORT for 2009B 20098 NRC SRO QUESTIONS OOB 008 A2.05 OOB 008 Component Cooling Water System (CCWS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

45.3 / 45.13)

(CFR: 41.5 / 43.5 / 45.3/

A2.05 Effect of loss of instrument and control air on the position of the CCW valves that are air operated Importance Rating: 3.3* 3.5 Technical

Reference:

OP-107 pp 57 Rev 79 OP-131.03 pp 13 Rev 31 DWG 5-G-OB22 5-G-0822 References to be provided: None Learning Objective: AOP-LP-3.17, Obj. 4 Question Origin: NEW Comments: (KIA match) The question requires knowledge of the position of the component cooling water valve on a loss of instrument air, its effect on the system, and appropriate response Tier/Group: T2G1 SRO Justification Requires evaluation of plant conditions and must make a determination of the correct procedural implementation.

Tuesday, October 20,200910:29:16 20,2009 10:29:16 AM 37

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

13. Given the following plant conditions:

- The Reactor was tripped on simultaneous trip of BOTH Main Feedwater Pumps

- All AFW was subsequently lost

- FRP-H.1, Response to Loss of Secondary Heat Sink, has been initiated

- The crew has initiated RCS Bleed and Feed

- Core Exit Thermocouples are at an average of 595°F and slowly decreasing

- Main Feedwater Pump 'A' has just been re-started

- All Steam Generators are at approximately 10% Wide Range Range.

Based on current plant conditions, feed (1) Steam Generator(s) at the (2) rate.

A'! (1)

(1 ) ONE (2) minimum controllable B. (1) ALL (2) minimum controllable C. (1) ALL (2) maximum D. (1)

(1 ) ONE (2) maximum Plausibility and Answer Analysis A. Correct. This is the process when ICT's are stable or decreasing.

B. Incorrect. Plausible since all al/ SGs are below the required level and feeding at miniumum flow is the required condition with ICT's rising.

C. Incorrect. Plausible since all al/ SGs are below the required level and Attachment 1 does provide conditions for feeding SGs at the maximum rate when feedwater capability is restored.

D. Incorrect. The first part is correct. Second part is plausible since Attachment 1 does provide conditions for feeding SGs at the maximum rate when feedwater capability is restored.

Tuesday, October 20,200910:29:16 AM 39

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS 059 A2.04 059 Main Feedwater (MFW) System A2 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:

(CFR: 41.5/ 43.5 / 45.3 / 45.13)

A2.04 Feeding a dry S/G Importance Rating: 2.9* 3.4*

Technical

Reference:

FRP-H.1, Attachment 1, pp. 49 Rev 23 References to be provided: None Learning Objective: EOP-LP-3.11 Obj. 1 Question Origin: BANK Comments: Applicant has to evaluate SGs and identify dry SG conditions are met and determine appropriate feed rate.

Tier/Group: T2G1 SRO Justification Required knowledge of specific guidance on restoration of Feed Flow after RCS bleed and feed per an attachment in a Function Restoration procedure.

Tuesday, October 20, 2009 10:29:16 AM 40

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

14. Given the following plant conditions:

- The plant is operating at 100% power

- The 'B' EDG is under clearance for scheduled maintenance

- At 1135, the Outside Operator reports that BOTH Starting Air Compressors for is 210° the 'A' EDG have just failed and the pressure in BOTH Starting Air Receivers 21 psig and lowering

- At 1210, the Outside Operator reports that the Low Pressure Starting Air annunciator has been received and pressure in BOTH Starting Air Receivers for the 'A' EDG is 189 psig and lowering Which ONE of the following identifies the time AND Technical Specifications requirement for the above conditions?

A. By 1235, action must be initiated to place the plant in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B. By 1335, one EDG must be restored or be in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> C. By 1310, action must be initiated to place the plant in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> D~ By 1410, one EDG must be restored or be in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Plausibilty and Answer Analysis This question is evaluating when does the }\ ~' EDG become inoperable and what is the I

action required for two inoperable EDGs. The ~' EDG becomes inoperable at 1410, The}\ I when starting air pressure is no longer greater than or equal to 190 psig. The TS action is to restore one within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The plausibile distractors dis tractors are based on both starting air compressors failing and TS 3.0.3 but an action exists for two inoperable EDGs.

A. Incorrect. Plausible because the time is based on starting air compressor failure action based on TS 3.0.3.

B. Incorrect. Plausible because the action is correct but the answer is using the inoperability wrong in operability time C. Incorrect. Plausible because the answer has the correct inoperability in operability time but the action is based on TS 3.0.3 D. Correct. Inoperability based on air pressure less than 190 psig and two hours to restore one.

Tuesday, October 20,200910:29:16 20,2009 10:29:16 AM 42

QUESTIONS REPORT for 20098 NRC SRO QUESTIONS 064 G2.2.37 064 Emergency Diesel Generators (ED/G) 2.2 Equipment Control 2.2.37 Ability to determine operability and/or availability of safety related equipment.

(CFR: 41.7/43.5/45.12)

Importance Rating: 3.6 4.6 Technical

Reference:

Tech Specs 3/4.8.1.2 pg 3/4 8-3, amend 78 References to be provided: None Learning Objective: Student Text Diesel Engine Obj. 15 Question origin: NEW Comments: KA matched by having applicant determinewhen the 'A' EDG becomes inoperable and detrmine required action.

Tier/Group: T2G1 SRO justification: Requires knowledge of the Surveillance and of a Tech Spec Action that is greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and knowledge that the EDG Air Compressors are not safety related.

fLp;;nU&&19i1!

fLECnmA! srsrDti WiER SYSltllS A,f,;.mw A,C, mCES QPbBATING OPt:MIING LIlilUill t'f",WH"f'Y,fliJ FOR OPEAATION Lll'lnl~ COOOITION OPERATION

{(;O,,!Jtifl~d}:

ACTION (CQot~J'I~d):

ACTHII cL (1, With two of the reqairoo r~ired offstte ltC, sources inoperable:

offsUe Itt, L Restere Restore cne {)f1I2 offsite .

~. ~,~.tjle~st HOT

~~

SHIJ.~ W'Ithtn~'l~'nH~~ the 2,

J, J. tiC>> of one offsite A,C, A.1:. s in fr":

tim? r

~ of A.C. sooree.

e. hie;::;niHI,. rFMII,1.1,l'¥!ll df~~l ~~F~tQr;~:~~~r:ab.;

L Perform Surveillance ~l~td' l),g,.Ll,l,;} \'tith'ln 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> aOO OI'!Ce per 8 hoors therea.ftel": ill'!

3,*.

3 Tuesday, October 20,200910

29:16 AM 43

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

15. Given the following plant conditions:

- The plant is operating at 100% power

- SFPs 'A', 'B', and 'c' 'C' are ALL ALL interconnected through the canals

- Fuel movement is occurring in the Fuel Handling Building. Building

- Chemistry reports that the SFP boron concentration is 995 ppm Pool boron concentration is maintained within limits to ensure Keff in the pools is) maintained less than or equal to ill ' .

With the above conditions, Technical Specifications requires immediately suspending movement of @  ;-"->~-~

(1 ) 0.95

. A. (1)

(2) fuel assemblies and loads over the pools B~ (1)

B:' (1 ) 0.95 (2) fuel assemblies, ONLY ONLY C. (1) ~

---

~.

(2) fuel assemblies and loads over the pools D. (1) 0.99

~

~.

(2) fuel assemblies ONLY Plausibility and Answer Analysis A. Incorrect. Plausible because the first part of the answer is correct (0.95 Kef(

Keff is correct). Loads over the pools is plausible because this is the action listed for water level NOT within limits.

B. Correct. TS 3.7. 14 requires 2000 ppm boron and the basis is to mainitain Keff Kef(

less than or equal to 0.95. With the TS not satisfied it requires suspension of movement of fuel assemblies.

C. Incorrect. Plausible because 0.99 Kef( Keff is the reactor requirement when in Modes 3 through 5. Loads over the pools is plausible because this is the action listed for water level NOT within limits.

D. Incorrect. Plausible because 0.99 Kef( Keff is the reactor requirement when in Modes 3 through 5. The second part of the answer is correct action - suspend movement of fuel assemblies.

Tuesday, October 20, 20,200910:29:16 2009 10:29:16 AM 44

QUESTIONS REPORT for 20098 2009B NRC SRO QUESTIONS 033 A2.01 033A2.01 033 Spent Fuel Pool Cooling System (SFPCS)

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5/

41.5 / 43.5 / 45.3 / 45.13)

A2.01 Inadequate SOM .

Importance Rating: 3.0 3.5 Technical

Reference:

T.S. 3.7.14 and bases (Pages 3/4 7-31 and 83/47-5)

B3/4 7-5)

References to be provided: None Learning Objective: Student Text Fuel Handling and Storage, Obj. 11 Question Origin: NEW Comments: (KIA match) Requires evaluating Tech Spec action for inadequate boron concentration and undertsanding of the basis for T.S. required boron concentration.

Tier/Group: T2G2 SRO Justification Bases for boron Requires knowledge of Tech Spec 8ases concentration in the Spent Fuel Pool and required actions based on low concentration results.

Tuesday, October 20, 2009 10:29:16 10:29: 16 AM 45

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

16. Given the following plant condition:

- The Outdoor Tank Area Drain Transfer Pump Monitor (REM-01 MD-3530) has been declared inoperable

- The tank area must be pumped out with the monitor inoperable lAW the ODCM, which ONE of the following choices completes the statement below?

Releases may continue from this pathway provided that (1) . If the monitor is not restored to operable status within (2) the next Radioactive Effluent Release Report will include an explanation of why the monitor was not restored in a timely manner.

A. (1) once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples are analyzed for radioactivity at a LLD (2) 7 days B~ (1) once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples are analyzed for radioactivity at a LLD (2) 30 days C. (1) samples, release rate calcs, and the valve line-up are Independently Verified (2) 7 days D. (1) samples, release rate calcs, and the valve line-up are Independently Verified (2) 30 days Tuesday, October 20,200910:29:16 AM 47

QUESTIONS REPORT for 20098 2009B NRC SRO QUESTIONS Plausibility and Answer Analysis A. Incorrect. 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples is plausible because this is the correct action required when the Tank Area Drain Pump Rad Monitor is inoperable. 7 Days is plausible because WRGM have a 7 Days or report action.

B. Correct. 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples is the action required when the Tank Area Drain Pump Rad Monitor is inoperable. If the monitor is not restored to OPERABLE status within 30 days, the requirement is to initiate a CR and an explanation in the next Radioactive Effluent Release Report is required pursuant to ODCM, Appendix F, Section F.2 of why this inoperability was not corrected in a timely manner.

C. Incorrect. Plausible because independently verifying samples, calculations, and line-ups is required for other release path monitors (see TL & HS discharge monitors in OWP-RM-10). 7 Days is plausible because WRGM have a 7 Days or report action.

D. Incorrect. Plausible because independently verifying samples, calculations, and line-ups is required for other release path monitors (see TL & HS discharge monitors in OWP-RM-10). If the monitor is not restored to OPERABLE status within 30 days, the requirement is to initiate a CR and an explanation in the next Radioactive Effluent Release Report is required pursuant to ODCM, Appendix F, Section F.2 of why this inoperability was not corrected in a timely manner.

Tuesday, October 20, 20,200910:29:16 2009 10:29: 16 AM 48

QUESTIONS REPORT for 2009B 20098 NRC SRO QUESTIONS 068 G2.1.7 068 Liquid Radwaste System (LRS) 2.1 Conduct of Operations 2.1 .7 Ability to evaluate plant performance and make operational judgments based on 2.1.7 operating characteristics, reactor behavior, and instrument interpretation.

41.5/43.5/45.12/45.13)

(CFR: 41.5 / 43.5 / 45.12 / 45.13)

Importance Rating: 4.4 4.7 Technical

Reference:

OWP-RM, Rev. 29, page 34 and 35, ODCM D-3 0-3 & D-4 (PDF pages 173 and 174), Rev 9 References to be provided: None Learning Objective: Student Text, Liquid Waste Processing Obj. 9

. Question Origin: NEW Comments: (KIA match) Operation knowledge of continued release with an inoperable radiation monitor.

Tier/Group: T2G2 SRO Justification SRO knowledge of ODCM Actions that are greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Tuesday, October 20,200910:29:16 AM 49

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

17. Given the following plant conditions:

- The crew is lining up to vent the PRT in accordance with OP-1 00, Reactor Coolant OP-100, System section 8.2

- 1RC-141 and 1RC-144, N2 TO PRT P RT valves have been opened The following occur:

- An inadvertent Phase A Isolation Signal is received

- The RO reports that 1RC-141 remains open Which ONE of the following identifies the location of the controls for the N2 TO PRT RC-141 ?

valves AND the action required by Technical Specifications for 1RC-141?

Location Required Action Required A. AEP-1 Verify 1RC-144 shut then remove fuses for 1RC-144 B. AEP-1 ONLV, its fuses do NOT need to be removed Verify 1 RC-144 shut ONLY, C. MCB Verify 1RC-144 shut ONLY, ON LV, its fuses do NOT need to be removed D~

D!' MCB Verify 1RC-144 shut then remove fuses for 1RC-144 Plausibility and Answer Analysis A. Incorrect. AEP-1 is plausible as the location because the RCDT Isolation are located on AEP-1 and frequently operated by control room operators. The TS action is correct.

a.

B. Incorrect. AEP-1 is plausible as the location because the RCDT Isolation are located on AEP-1 and frequently operated by control room operators. This action will isolate the failed open valve/penetration but TS requires the valve to be deactivated..

deactivated.

C. Incorrect. MCa is the correct location. This action will isolate the failed open valve/penetration but TS requires the valve to be deactivated.

D. Correct. MCaMCB is the correct location. The TS action is correct.

Tuesday, October 20,200910:29:16 20, 2009 10:29:16 AM 51

QUESTIONS REPORT 2009B NRC SRO QUESTIONS for 20098 071 G2.1.31 071 Waste Gas Disposal System (WGDS) 2.1 Conduct of Operations 2.1.31 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

(CFR: 41.10/ 45.12)

Importance Rating: 4.6 4.3 Technical

Reference:

Tech Spec 3.6.3 page 3/4 6-14 3/46-14 PLP-106 rev 46, pages 18 and 32 References to be provided: None Learning Objective: Student Text Containment Isolation System Obj. 11 Question Origin: NEW Comments: (KIA match) Candidate must identify the location of the controls for venting the PRT to the WG System Tier/Group: T2G2 SRO Justification Requires application of Tech Spec 3.6.3 which is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> action 20,200910:29:16 Tuesday, October 20, 2009 10:29: 16 AM *52 52

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

18. Given the following plant conditions:

- The Plant is in Mode 6

- Refueling Cavity Level is at 23' 6"

- 'A' train of RHR is in service for Shutdown Cooling

- 'B' RHR Pump is under clearance for scheduled maintenance lAW Technical Specifications, the MINIMUM RHR flowrate for the above conditions is (1) AND the purpose of one RHR Pump being in operation is to ensure that sufficient cooling capacity is available to maintain the RCS below (2) ?

A. (1 ) 2500 gpm (2) 200°F B~ (1)

(1 ) 2500 gpm (2) 140°F C. (1 ) 900 gpm (2) 200°F D. (1 ) 900 gpm (2) 140°F Plausibility and Answer Analysis Tech Specs requires 2500 gpm when above the flange. 900 gpm is plausible because 900 gpm is the requirement when below the flange. Tech Specs basis states one in operation to maintain below 140°F (Mode 6). 200°F is plausible because this is Mode 5 and where steam production begins.

A. Incorrect. Right flowrate but wrong purpose.

B. Correct. Right flowrate and right purpose.

C. Incorrect. Wrong flowrate and wrong purpose.

D. Incorrect. Wrong flowrate but right purpose.

Tuesday, October 20,200910:29:16 AM 54

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS G2.1.28 2.1 Conduct of Operations 2.1.28 2.1 .28 Knowledge of the purpose and function of major system components and controls.

(CFR: 41.7)

Importance Rating: 4.1 4.1 Technical

Reference:

Tech Spec 3.9.8.2, page 3/49-10 3/4 9-10 Tech Spec Bases, page B 3/49-2 References to be provided: None Learning Objective: Student Text RHR System Obj. 11 Question Origin: NEW Comments: Meets KIA by requiring the applicant to know the purpose and function (basis) of one RHR pump being in operation in Mode 6 Tier/Group: T3 SRO Justification Requires knowledge of Tech Specs Surveillance requirement for RHR flow in Mode 6.

Tuesday, October 20,200910:29:17 20,2009 10:29:17 AM 55

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

19. Given the following plant conditions:

- The Plant is in Mode 6

- GP-009, Refueling Cavity Fill, Refueling and Drain of the Refueling Cavity, is in progress

- 'A' RHR pump is in service to provide core cooling during refueling operations

- 'B' RHR pump is operable and in standby The Refueling Team has requested that the 'A' RHR pump be secured temporarily lAW Technical Specifications for the above conditions, which ONE of the following completes the statement below?

The operating RHR loop may be secured for a maximum of up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per (1) to perform (2)

A. (1 ) 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period (1)

(2) Reactor Vessel foreign object search and retrieval operations B. (1 ) 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period (1)

(2) Core Alterations in the vicinity of the Reactor Vessel hot legs C~ (1)

C,.. (1 ) 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period (2) Core Alterations in the vicinity of the Reactor Vessel hot legs D. (1)(1 ) 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period (2) Reactor Vessel foreign object search and retrieval operations Tuesday, October 20,200910:29:17 20, 2009 10:29:17 AM 57

QUESTIONS REPORT for 20098 2009B NRC SRO QUESTIONS Plausibility and Answer Analysis A. Incorrect. The first part of the answer is corrrect. The second part is incorrect but plausible because reducing flow in the core during search and retrieval of foreign objects in the vessel seems viable but this is not a condition that is allowed by Technical Specifications. In addition FHP-041, Reactor Vessel Foreign Object Search & Retrieval Operations does not have steps to secure any running RHR pump during search & retrieval operations.

B. Incorrect. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period is plausible because other specifications in section 3.9 use 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Specs 3.9.2, 3.9.8, 3.9.11.

3.9. 11. The second part is correct.

C. Correct. Per GP-009 rev 49, Precaution & Limitation 1, page 10, the operating loop may be secured for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in any 2-hour period for Core Alterations.

Per TS 3.9.8.1 with one additional RHR loop operable the RHR pump may be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during the performance of Core Alterations and core loading verification in the vicinty of the reactor vessel hot legs.

D. Incorrect. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period is plausible because other specifications in section 3.9 use 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Specs 3.9.2, 3.9.8, 3.9.11.

3.9. 11. The second part is incorrect but plausible because reducing flow in the core during search and retrieval of foreign objects in the vessel seems viable but this is not a condition that is allowed by Technical Specifications. In addition FHP-041, Reactor Vessel Foreign Object Search & Retrieval Operations does not have steps to secure any running RHR pump during search & retrieval operations.

Tuesday, October 20, 2009 10:29:17 AM 58

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

20. While performing OST-1 026, Reactor Coolant System Leakage Evaluation, Computer OST-1026, Calculation, Daily Interval, Modes 1-2-3-4, unidentified leakage exceeded the control chart 3 sigma point. The STA has prepared a Troubleshooting Control Form which was evaluated to be 'High Risk'.

What is the MINIMUM level of approval required outside the Max/Safe/Gen time frame?

A. Troubleshooting Lead B. Control Room Supervisor C~ Shift Manager C,",

D. Plant General Manager Plausibility and Answer Analysis A. Incorrect. Plausible because the Troubleshooting Lead does not have approval authority of this troubleshooting 'Risk' activity (STA in this case initiated the form but it could have been the Shift Manager preparing the TCF and would therefore have had the authority to approve this troubleshooting activity) .

. B. Incorrect. Plausible because the CRS would have approval if the risk was a 'No Risk' or 'Low Risk' activity.

c.

C. . Correct. During High Risk activities the SM is the minimum level for approval per AP-929.

D. Incorrect. Plausible because the PGM could approve this troubleshooting but is not the minimum level for approval per AP-929.

Tuesday, October 20,2009 10:29:17 AM 61

QUESTIONS REPORT 20098 NRC SRO QUESTIONS for 2009B G2.2.20 2.2 Equipment Control 2.2.20 Knowledge of the process for managing troubleshooting activities.

(CFR: 41.10/

41.10 / 43.5 / 45.13)

Importance Rating: 2.6 3.8 Technical

Reference:

AP-929 rev 14, pages 13/14 References to be provided: None Learning Objective: PP-LP-3.6 Obj. 11

.Question Origin: NEW Comments: (KIA (K/A match) The question requires the applicant to have knowledge of who can provide athoritiy for managing troubleshooting activities.

Tier/Group: T3 SRO Justification Requires the knowledge of the process for carrying out troubleshooting activities. SRO only because RO's are not required to have an indepth knowledge of this process.

3. I~ng Organi2anoos Implementing Organizations Responsible RespORiS1b!e Superintendent Supermtendlenf or Manager:: Manager:

a.

3. leads 1he the coordinaoon cOiClfdinafion and implementmon implementaoon of of MEDIUM MErnUM:RlSK RISK or HIGH msKRISK troobleshooting pml1lS.

trooblesllootD'lg plans.

D.

b. Ensures adequate amtJoIs controls are in place to rnintmize minimize challenges to the plant or ensures contingency ;:nans plans are developed and ready.

c.

C. Notifies the Work Week Manager of all formal Notmes trou~ plan WIO"s format troubleshooting W/O's for schedUle sd1edule addition addruoo per ADM-NBGC-01il4.

ADM-NGGC-0104.

~

d. AP~I~P~UM RiSK andH£GH1AiSK~eshoooog~es; "u re;""c,f,,,,,,,/tl; ,ecce" W'.0.<"""N""W""u~"" I

.cc-~"'.UU'WM'' W 'W"vc"c,p <MM......x" C'Cc"~, " .. - u"

e. Addffionalfy. the Implementing Org.miZaOOns Manager is responstble for assigning, in cooperaian with lie Manager - Shift Operatioos, the Tl for HIGH RISK troubleshooting activities.

AP-929 Rev. 14 Page 13 of 36 4.0 RESPONSIBIUTES RESPONSIBILITIES (amtinued)

(OOfIUnued)

4. Plant P1alllt General Genera! Manager:

<l.

fl. 'Within the MaxfSafeiGen Within MaxlSafeJGen penod, period, ~ authorizes the ~ingactMties troobleshooting activmes for Medium and High RisK. RiSk.

AP-929 Rev. 14 Page 14 of 36 Tuesday, October 20,200910:29:17 AM 62

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

21. Given the following plant conditions:

- An RCS heatup is in progress

- RCS Tavg is 342°F 342°'1=

- 1 1A-SA A-SA Emergency Diesel Generator is declared INOPERABLE Which ONE of the following identifies the current OPERATrONAL OPERATrONAL MODE AND the Technical Specification applicability regarding Mode changes?

The Unit is in (1) . Ascension to a higher Mode (2) be performed.

A. (1 ) Mode 3 (2) may B. (1 ) Mode 3 (2) may NOT C. (1 ) Mode 4 (2) may D~ (1 ) Mode 4 D!'

(2) may NOT Plausibility and Answer Anlaysis A. Incorrect. Hot Standby is Mode 3, which is >350 degrees F. Additionally, Mode change is plausible because some tech specs indicate TS 3.0.4 3.004 is not applicable.

In this instance, 3.0.4 3.004 does apply, and even though action requirements are met, the LCO does not have an indefinite time requirement as defined by TS section 3.0. .

B. Incorrect. Mode is incorrect as described in Distractor Dis tractor A. (2) Mode change to next higher mode is not permitted.

C. Incorrect. Correct Mode is identified, but incorrect Mode change action is applied.

Plausible as in Distractor A, and reference indicates situations where Mode change may be performed.

D. Correct. Correct Mode is identified as noted above. (2) Mode change to next higher mode is not permitted.

Tuesday, October 20,200910:29:17 AM 63

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS G2.2.35 2.2 Equipment Control 2.2.35 Ability to determine Technical Specification Mode of Operation.

(CFR: 41.7 41.7/41.10/43.2/45.13)

/ 41 .10 / 43.2 / 45.13)

Importance Rating: 3.6 4.5 Technical

Reference:

TS Table 1-2, TS 3.0.4 References to be provided: None Learning Objective: Student Text Diesel Engines Obj. 15 Question Origin: BANK Comments: (KIA match) The question requires the candidate to identify which Mode the plant is operating in based on RCS Temperature Tier/Group: T3 SRO Justification Requires knowledge of applying two Tech Specs, 3.8.1 and 3.0.4 to come to the correct answer. SROs provide approval for entering operational Modes.

Tuesday, October 20,200910:29:17 AM 64

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

22. The MAXIMUM dose allowed for life saving missions during a declared emergency is (1) REM TEDE and must be authorized by the (2)

A. (1)

(1 ) 10 (2) Site Emergency Coordinator B. (1)

(1 ) 10 (2) Radiological Control Director C. (1)(1 ) 25 (2) Radiological Control Director D¥'

D~ (1) 25 (2) Site Emergency Coordinator Plausibility and Answer Anlaysis A. Incorrect. 10 Rem TEDE is plausible because this is the limit for protecting valuable equipment but life saving and the limit is 25 Rem TEDE. Site Emergency Coordinator (SEC) is correct.

B. Incorrect. 10 Rem TEDE is plausible because this is the limit for protecting valuable equipment but life saving and the limit is 25 Rem TEDE. The Radiological Control Director (RCD) is plausible because this individual works in the TSC with the SEC.

C. Incorrect. 25 Rem TEDE is correct. The Radiological Control Director (RCD) is plausible because this individual works in the TSC with the SEC.

D. Correct. 25 Rem TEDE is correct. Site Emergency Coordinator (SEC) is correct.

Tuesday, October 20, 200910:29:17 20,2009 10:29:17 AM 66

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS G2.3.4 2.3 Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12 / 43.4 / 45.10)

Importance Rating: 3.2 3.7 Technical

Reference:

PEP-330, Att. 1, Rev 9, Page 17 References to be provided: None Learning Objective: Task: 345013H602 Question Origin: NEW Comments: (KIA match) Knowledge of limits during an emergency.

PEP-330 attachment 1 Tier/Group: T3 SRO Justification Shift Manager fills the role of the SEC until relieved the the SEC-TSC.

SEC-TSG. SRO must be knowledgable in the event the Shift Manager is unable to report to the MCR during an event.

Tuesday, October 20,200910:29:17 AM 67

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

23. Given the following plant conditions:

- The plant was operating at 100% power

- EDG 1B-SB is out of service The following events have occurred:

- A Loss of Offsite Power

- The TDAFP failed to start

- The 'A' MDAFW pump tripped 2 minutes after starting

- The crew transitioned to FRP-H.1, Loss Of Secondary Heat Sink, based on a CSFST RED Path

- Subsequently, EDG 1A-SA output breaker trips on a bus fault Which ONE of the following describes the actions that will be taken AND how AOP-025, Loss of One Emergency AC Bus (6.9KV) or One Emergency DC Bus (125V) will be utilized?

A'!

A'I (1) Immediately transition to EPP-001, EPP-001 , Loss Of All AC Power to 1A-SA and 1B-SB Buses.

(2) AOP-025 may not be used until at least one train of AC power is restored.

B. (1) Remain in FRP-H.1 until directed to return to procedure in effect, and then transition to EPP-001.

(2) AOP-025 may be used concurrently with FRP-H.1 ONLY if referring to the AOP does NOT result in delaying accident mitigation.

C. (1) Immediately transition to EPP-001, Loss Of All AC Power to 1 1A-SA A-SA and 1B-SB Buses.

1B-SB (2) AOP-025 may be used concurrently with EPP-001 as necessary under all conditions of EOP use.

D. (1) Remain in FRP-H.1 until directed to return to procedure in effect, and then transition to EPP-001.

(2) AOP-025 may be used concurrently with FRP-H.1 as necessary under all conditions of EOP use.

Tuesday, October 20, 20,200910:29:18 2009 10:29: 18 AM 69

QUESTIONS REPORT 2009B NRC SRO QUESTIONS for 20098 Plausibility and Answer Analysis A. Correct. EPP-001 is a direct entry procedure and FRPs assume atleast 1 safety EPP-OO 1. The EOP User's Guide states train in energized. Therefore must enter EPP-001.

"Based on purpose and construction of EOP-EPP-001, concurrent implementation of AOPs is inappropriate; at least until one train ofAC emergency power is restored.""

restored.

B. Incorrect. Plausible since Red path still exists, but with no power available, crew must transition to EPP-001 to restore at least one vital bus. Second part is correct.

C. Incorrect. First part is correct. Second part is plausible since AOPs may be used concurrently with EOPs but only when use of the AOP will not impede efforts to implement the EOP.

D. Incorrect. First part is plausible since Red path still exists, but with no power EPP-OO 1 to restore at least one vital bus.

available, crew must transition to EPP-001 Second part is plausible since AOPs may be used concurrently with EOPs but only when use of the AOP will not impede efforts to implement the EOP.

Tuesday, October 20,200910:29:18 20, 200910:29:18 AM 70

QUESTIONS REPORT for 20098 NRC SRO QUESTIONS G2.4.16 2.4 Emergency Procedures / Plan 2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

41.10/43.5/45.13)

(CFR: 41.10 / 43.5 / 45.13)

Importance Rating: 3.5 4.4 Technical

Reference:

EOP Users Guide section 5.1.4 pp 13 and 5.1.7 pp 16 Rev 26 References to be provided: None Learning Objective: EOP-LP-3.07 Obj. 4 EOP-LP-3.19 Obj. 1b Question Origin: NEW Comments: (KIA match) The question involves Emergency procedure hierarchy implementation and the cooridination of the procedure being used and the priorty of implementation Tier/Group: T3 SRO Justification Requires indepth knowledge of procedures to determine that you should not transistion to another higher order emergency procedure under certain plant conditions.

Tuesday, October 20,200910:29:18 20, 2009 10:29:18 AM 71

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS

24. Given the following plant conditions:

- The Plant is in Mode 3

- The 'A' MDAFW Pump is under clearance for motor replacement

- A loss of DP-1 B-SB occurs

- The crew enters AOP-025, Loss of One Emergency AC Bus (6.9KV) or One Emergency DC Bus (125V)

Which ONE of the following describes the operation of the TDAFW Pump if a start signal occurs and the action required by Technical Specifications as a result of the plant conditions?

The TDAFW Pump will: Technical Specifications Action A. start and continue to run place the plant in Mode 4 in 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> B. start and continue to run maintain Mode 3 C. start and trip on overspeed place the plant in Mode 4 in 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> D!' overspeed D'!" start and trip on overs peed maintain Mode 3 Plausibility and Answer Analysis A. ~' Train DC had been lost this would be the Incorrect. This is plausible because if 'A' response but with the loss of 'B' Train, control power is lost and an overspeed trip will occur. The Tech Spec action listed is from 3.0.3 and is plausible because all three AFW pumps are inoperable but AFW has specific actions for all three inoperable which suspends required mode changes.

8.

B. ~' Train DC had been lost this would be the Incorrect. This is plausible because if 'A' response but with the loss of '8' Train, control power is lost and an overspeed trip will occur. The Tech Spec action is correct.

C. Incorrect. There is a NOTE in the AOP-025 which addresses the TDAFW Overspeed Trip. The Tech Spec action listed is from 3.0.3 and is plausible because all three AFW pumps are inoperable but AFW has specific actions for all three inoperable which suspends required mode changes.

D. Correct. There is a NOTE in the AOP-025 which addresses the TDAFW Overspeed Trip. The Tech Spec action is correct.

Tuesday, October 20, 2009 10:29:18 AM 73

QUESTIONS REPORT for 2009B NRC SRO QUESTIONS G2.4.20 2.4 Emergency Procedures / Plan 0 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

(CFR: 41.10/ 43.5 / 45.13)

Importance Rating: 3.8 4.3 Technical

Reference:

AOP-025 Note on page 45, Rev. 25 Tech Specs 3.7.1.2 pg 3/4 7-4 (PDF page 291) Amend 93 References to be provided: None Learning Objective: AOP-LP-3.25, Obj 3d Question Origin: BANK 2009a SRO NRC Exam Question #5 Comments: (KIA Match) Matches KA because a Note in AOP-025 alerts operator that TDAFW will start and trip on overspeed in this situation.

Tier/Group: T3 SRO Justification Requires application of notes associated with application of Tech Spec action items.

Tuesday, October 20,200910:29:18 AM 74

QUESTIONS REPORT for 20098 2009B NRC SRO QUESTIONS

25. Given the following plant conditions:

- A Loss of Offsite Power has occurred

- The crew transitioned to EPP-004, Reactor Trip Response

- The MSIVs are closed

- During the event, the 'A' SG level increased to 98% but is presently 83%

Which ONE of the following pressures on 'A' SG would meet the YELLOW path condition for FRP-H.2, Response to Steam Generator Overpressure, AND which SG PORV(s) would be used for steam dumping?

'A' SG Pressure SG PORV(s)

A. 1170 psig 'A' ONLY 8.

B. 1170 psig '8' and

'B' 'c' ONLY ONLY C. 1230 psig 'A' ONLY D~

D'" 1230 psig '8' and

'B' 'c' ONL ONLYY Plausibility and Answer Analysis A. Incorrect. 1170 psig is plausible because it would require entry into FRP-HA, Response to Loss of Normal Steam Release Capability. Dumping steam from ~ I I SG PORV is plausible because this action would be taken had the ~ SG NOT I

been overfilled.

B. Incorrect. 1170 psig is plausible because it would require entry into FRP-HA, Response to Loss of Normal Steam Release Capability. Dump steam from the '8' 'B' and 'c' SG PORVs is correct.

C. pSig is correct. Dumping steam from ~

Incorrect. 1230 psig ~'I SG PORV is plausible because this action would be taken had the ~ I SG NOT been overfilled.

I D. Correct. 1230 psig is correct. Dump steam from the '8' 'B' and 'c' SG PORVs is correct.

Tuesday, October 20,200910:29:18 AM 76

QUESTIONS REPORT 2009B NRC SRO QUESTIONS for 20098 WE13 EG2.1.7 E13 Steam Generator Overpressure 2.1 Conduct of Operations .

2.1 .7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5/

41.5 / 43.5 / 45.12 / 45.13)

Importance Rating: 4.4 4.7 Technical

Reference:

FRP-H.2, Rev. 8, Pages 4 and 5 References to be provided: None Learning Objective: EOP-LP-3.11 Obj. 4.a Question Origin: NEW Comments: (KIA match) Candidate must evaluate the 'Affected SG' SG 1 and determine that it has been overfilled and therefore steam should NOT be released from it.

Tier/Group: T1 G2 SRO Justification Requires detailed knowledge of Caution and Action in a Yellow path FRP.

CAUTION

~ h~~relt>AIJR £~1~Y sa ~it.~~i'l<:;;";i:~:1"''Adllt.O g~..'it;.. r 5' f30",lpri?)r ~'lY'>.O\'ertill e'rIaiuatioo *.

6.

8. Coottnue tioDtlnlle Ttl I'Ianually OR .!.<<X:al1y

'1"0 Manually ~al11 Dump 8 lluap Stea. Affee~ed scH.s)

From Affe~ted t _ Jmm SG($}

USin~

U8in~ Ally The lJ'Ql1O'dIlJp Of 'l.'hll!

Any (if Follo.iRa:

a.

!l. sa SG PO]!,VIJ POlVs Page 3j of 17 HEAT SINK CSF-3 EOP-CSFST REV.

RtY 9 Tuesday, October 20,200910:29:18 20, 2009 10:29:18 AM 77