Text

Initial Initial Exam Retake 2009-302 Draft SRO Written Exam
	ML092881149
Person / Time
Site:	Sequoyah
Issue date:	10/15/2009
From:	; Operator Licensing and Human Performance Branch
To:	;
References
	50-327/09-302, 50-328/09-302
	Download: ML092881149 (69)
	v • d • e

QUESTIONS REPORT for 2009 SRO Retake Exam 009 EA2.11 076 Given the following:

- A SGTR on SG #4 caused an automatic safety injection (SI) on Unit 2.

E-3, "Steam Generator Tube Rupture" is being implemented.

- SI has just been terminated.

Subsequently, the crew observes the following:

- SG #1, #2 and #3 levels and pressures are stable.

- SG #4 level and pressure are lowering in an uncontrolled manner.

- Pressurizer level cannot be maintained.

- Containment pressure, temperature and humidity are rising.

Which ONE of the following identifies the required actions?

A. Manually actuate Safety Injection, go to E-O, "Reactor Trip or Safety Injection."

B. Manually establish ECCS flow, go to E-2, "Faulted Steam Generator Isolation. "

C~ Manually establish ECCS flow, go to ECA-3.1, "SGTR and LOCA -

Subcooled Recovery."

l\

D. Manually actuate Safety Injection, go to ECA-3.2, "SGTR and LOCA -

Saturated Recovery."

QUESTIONS REPORT for 2009 SRO Retake Exam DIS TRA GTOR ANAL YSIS:

DISTRAGTOR A. Incorrect, the FOP for E-3 has the crew establish EGGS flow by manual pump start verses actuation of SI. This is plausible as it is common in other EOP procedures to have the SI signal actuated on deteriorating conditions. The correct procedure transition after establishing EGGS flow is to EGA-3.1, SGTR and LOGA - Subcooled Recovery.

This is plausible as in other EOP procedures an SI actuation will return the crew to E-O.

B. Incorrect, The first part of the distracter is plausible because it is correct. The second part is plausible because E-2 is a viable transition from E-3 via the FOP if one of the previous intact SGs became faulted. SG #4 is experiencing lowering level and pressure in the subsequent portion of the stem which is consistent with a SBLOGA and a fault on it. The faultedlruptured SG will also require entry into EGA-3.

key is that a faulted/ruptured EGA-3.1. 1.

G. GORREGT, The indications given in the stem are indicative that SI reinitiation criteria will be met based on PZR level criteria. The FOP for E-3 has the crew establish EGGS flow by manual pump start verses actuation of SI. The correct procedure transition after establishing EGGS flow is to ECA-3.1 , SGTR and LOCA - Subcooled Recovery.

D. Incorrect, the FOP for E-3 has the crew establish EGGS flow by manual pump start verses actuation of SI. This plausible as it is common in other EOP procedures to have the SI signal actuated on deterioration conditions. The second part of the distracter is plausible as EGA-3.2 may be a viable recovery procedure, but only after entering EGA-3. 1 does it become an option.

(

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 76 Tier 1 Group 1 KIA 009 EA2.11 Ability to determine or interpret the following as they apply to a small break LOCA: Containment temperature, pressure and humidity Importance Rating: 3.8/4.1 Technical

Reference:

E-3, Steam Generator Tube Rupture, Rev 17 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271.E-3 B.5 Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

Question Source:

Bank# _ _ __

Modified Bank # X_ _

New _ __

(

Question History: SQN bank question 038 EA2.07 078 modified Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis __X_

10 CFR Part 55 Content: ( 43.5 / 45.13 )

10CFR55.43.b (5)

Comments: SQN bank question 038 EA2.07 078 modified Changed stem to include indications for a small break LOCA and require knowledge of SI actuation criteria and method.

A distracter was changed to the correct answer.

Correct answer relocated

QUESTIONS REPORT for 2009 SRO Retake Exam 015 AA2.11 077 Given the following:

- A LOCA has occurred on Unit 2.

Due to equipment failures, a transition to FR-C.1, "lnadequ9te "Inadequa,te Core Cooling" was required.

- The Core Exit temperature is 1205°F and rising.

- Containment Pressure is 3.6 psig.

- All S/Gs are Intact with levels of ...

of...

1 #2 #3 #4 9% NR 17% NR 26% NR 35% NR

- The crew is currently depressurizing the Intact S/Gs to atmospheric pressure.

Which ONE of the following identifies both:

(1) the number of RCPs that will be started in accordance with FR-C.1 and 0

(2) the minimum number of thermocouples reading above 12000 F required to

make the transition to SACRG-1" SACRG-1, Severe Accident Control Room Guideline Initial Response?"

( RCPs running Thermocouples A':"

Art 2 At least 5 total B. 2 At least 1 in each traid train' I C. 3 At least 5 total

\

D. 3 At least 1 in each train

QUESTIONS REPORT for 2009 SRO Retake Exam DISTRACTOR DIS TRA CTOR ANAL YSIS:

A. CORRECT, The conditions would result is 2 of the RCPs being started. Only loops #3 and #4 have the required minimum SG level to allow starting of the RCP in the loop. If the RCS temperature remained above 120(J'F 12000F on at least 5 core exit thermocouples, then a transition to SACRG-1 would be made.

B. Incorrect, 2 RCPs (Loops #3 and #4) would be started but the transition the SACRG-1 requires at least 5 core exit thermocouples (not 3) to be above 120(J'F.

12000F.

Plausible because starting 2 RCPs to provide temporary cooling for the core is correct and because using redundant instrument loops for conformation is applied in other conditions (such as fire detection systems).

C. Incorrect, only 2 RCPs (Loops #3 and #4) would be started because the SG levels are not above the minimum required in the other 2 but the transition to SACRG-1 does requires at least 5 core exit thermocouples to be above 120(J'F.

12000F. Plausible because starting the RCPS can provide temporary cooling for the core and the #2 loop SG has level above the normal narrow range minimum but the adverse level setpoint must be used and because using redundant instrument loops for conformation is applied in other conditions (such as fire detection systems).

D. Incorrect, only 2 RCPs (Loops #3 and #4) would be started because the SG levels are not above the minimum required in the other 2 but the transition to SACRG-1 requires at least 5 core exit thermocouples (not 3) to be above 120(J'F.

12000F. Plausible because starting the RCPS can provide temporary cooling for the core and the #2 loop SG has level above the normal narrow range minimum but the adverse level setpoint must be used and because using redundant instrument loops for conformation is applied in other conditions (such as fire detection systems).

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 77 Tier 1 Group 1 KIA 015AA2.11 Reactor Coolant Pump (RCP) Malfunctions Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

When to jog RCPs during ICC Importance Rating: 3.4* I/ 3.8*

Technical

Reference:

FR-C.1, Inadequate Core Cooling, Rev. 12 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 FR-C.1 B.3 Summarize the mitigating strategy for the failure that initiated entry-intoFR~C.1 entry into FR-C.1

. Question Source:

Bank# _ __ __ _ __

Modified Bank # X._ __

New _ __

( Question History: Commanche Peak question 015 AA 2.11 modified Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X__

110 0 CFR Part 55 Content: ( 43.5 I/ 45.13 )

10CFR55.43.b (5)

Comments: Commanche Peak SRO Exam 2007 question 015 AA 2.11 modified.

QUESTIONS REPORT for 2009 SRO Retake Exam 038 E2.4.20 078 Given the following:

Unit 1 experiences a Safety Injection due to a steam generator tube rupture.

- All Reactor Coolant Pumps were removed from service due to loss of support systems.

RCS cooldown at maximum rate to target incore temperature is in progress in accordance with E-3, "Steam Generator Tube Rupture."

- The STA reports that 1-FR-0, "Unit 1 Status Trees" indicates a RED path to FR-P.1, "Pressurized Thermal Shock," on the ruptured loop.

Which ONE of the following identifies the required action due to the FR-P.1 RED path?

Remain in E-3 until.

until .....

A. the cooldown is completed, then transition to FR-P.1 only if the RED path still exists.

B. the cooldown is completed, then transition to FR-P.1 even if the RED path no longer exists.

C!" the safety injection is terminated, then transition to FR-P.1 only if the RED C!'

( path still exists.

D. the safety injection inJE3ction is terminated, then transition to FR-P.1 even if the RED path no longer exists.

(

QUESTIONS REPORT for 2009 SRO Retake Exam DIS TRACTOR ANAL YSIS:

A. Incorrect, The transition would not be made until after the SI was terminated not after the cooldown was terminated. Plausible because remaining is E-3 to complete the cooldown is a major action whose completion is necessary in stopping any release.

B. Incorrect, The transition would not be made until after the SI was terminated and then only if the condition still existed. Plausible because remaining is E-3 is correct and completion of the cooldown is a major action in the procedure that has added to the stresses on the vessel that need to be addressed.

C. CORRECT, If cooling down while on Natural circulation, reverse flow can occur in the loop and can cause the SI flow to change. This can result in and indicated cold leg temperature in the stagnant ruptured SG loop below the value required for PTS. The transition to FR-P.1 FR-P. 1 would not be made until after the SI was terminated and then only if the condition still existed.

D. Incorrect, Transitioning to FR-P.1 after the SI is terminated is correct but only if the RED path conditions still exist the transition would only be made after the SI was terminated. Plausible because remaining is E-3 until the SI is terminated is correct and completion of the cooldown is a major action in the procedure that has added to the stresses on the vessel that need to be addressed.

(

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 78 Tier 1 Group 1 KIA 038 E2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

Importance Rating: 3.8/4.3 KIA match: SQN has rewritten the Caution from the WOG E-3 procedure as a step in the procedure as explained in the site background document included with this question. Thus, the question is testing the

'Knowledge of the operational implications of EOP cautions' as addressed in the KIA statement.

Technical

Reference:

E-3, Steam Generator Tube Rupture, Rev 17 EPM-3-E-3, Basis Document for E-3 Steam Generator Tube Rupture, Rev 7 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 E.3 B.4 Describe the bases for all limits, notes, cautions and steps of

( E-3.

Question Source:

Bank#

Bank X _ __

_X'--__

Modified Bank # _ ____ __ __

New _ __

Question History: WBN bank question Question Cognitive Level:

Memory or fundamental knowledge _ __

Com prehension or Analysis _ X _

Comprehension 10 CFR Part 55 Content: 41.10 I 43.5 I 45.13 )

((41.10/43.5/45.13 10CFR55.43.b (5)

Comments:

QUESTIONS REPORT for 2009 SRO Retake Exam 040 AA2.02 079 Given the following:

Unit 1 is operating at 67% power steady state conditions with Rod Control in Manual.

- A transient occurs resulting in the following:

Reactor power at 68% and increasing.

RCS pressure at 2225 psig and slowly decreasing.

- Auctioneered high Tavg at 564°F and decreasing.

- Turbine power at 66% and slowly decreasing.

Generator output at 785 MWe and slowly decreasing.

Which ONE of the following identifies the procedure the crew should implement and actions required?

A. AOP-C.02, "Uncontrolled RCS Boron Concentration Changes" and trip the reactor because of the current temperature difference between Tavg- Tref.

B. AOP-C.02, "Uncontrolled RCS Boron Concentration Changes." A reactor trip not currently required, but will be if reactor power rises greater than 3%

above turbine power.

C. AOP-S.05, "Steam Line or Feedwater Line Break/Leak" and trip the reactor

( because of the current temperature difference between Tavg- Tref.

D~ AOP-S.05, "Steam Line or Feedwater Line Break/Leak." A reactor trip not D!'

currently required, but will be if reactor power rises greater than 3% above turbine power.

QUESTIONS REPORT for 2009 SRO Retake Exam DIS TRACTOR TRA C TOR ANAL YSIS:

A. Incorrect, all of the conditions in the stem would occur from an uncontrolled change is RCS boron concentration (either a boration or a dilution) but not all in the direction

. provided for either one. All except the reactor power change would be applicable for a boration. Plausible because the parameters identified are parameters that would be affected during an uncontrolled change in RCS boron concentration and AOP-C.02 has the direction to trip the reactor if the difference in Tavg- Tref cannot be maintained less than EJ> Ef F.

B. Incorrect, all of the conditions in the stem would occur from an uncontrolled change is RCS boron concentration (either a boration or a dilution) but not all in the direction provided for either one. All except the reactor power change would be applicable for a boration. Plausible because the parameters identified are parameters that would be affected during an uncontrolled change in RCS boron concentration and AOP-C.02 has the direction to trip the reactor based on the magnitude of reactor AOP-G.02 power.

C. Incorrect, AOP-S.05 is the appropriate procedure to be entered but a reactor trip is not required for the current Tavg- Tref difference. The procedure directs a reactor Ef F and the current difference is trip if the difference cannot be maintained less than EJ>F

=

EfF. (Turbine at 66% Tavg. of 567.46. So, 564-567.46 -3.46) less than EJ>F. =

Plausible because the AOP A OP to be entered is correct and the AOP A OP has the direction EfF.

to trip the reactor if the difference in Tavg- Tref cannot be maintained less than EJ>F.

( D. CORRECT, The conditions in the stem are consistent with a steam line leak thus entering AOP-S.05 is the appropriate procedure. The AOP provides conditions that would require the reactor to be tripped and turbine power greater than 3% above reactor power is a condition that would require the reactor to be tripped. Currently the difference is only 2%.

(

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 79 I,

Tier 1 Group 1 KIA 040 AA2.02 Steam Line Rupture Ability to determine and interpret the following as they apply to the Steam Line Rupture:

Conditions requiring a reactor trip Importance Rating: 4.6/4.7 Technical

Reference:

AOP-S.05, Steam or Feedwater Leak, Rev. 7 AOP-C-.02, Uncontrolled RCS Boron Concentration Changes, Rev. 6 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271AOP-S.05 B 2.d & B.b Describe the AOP-S.05 entry conditions.

d. Describe the plant parameters that may indicate a Steam Line or Feedwater Line Break/Leak.

Given a set of initial plant conditions, use AOP-S.05 to correctly:

( b. Identify correct actions; Question Source:

Bank#

Bank # -_ _ __

Modified Bank # _X. __

X- -_-

New - - -

Question History: SQN bank question AOP-S.05-B.2.A 001 modified.

Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X__ __

10 CFR Part 55 Content:

10 (43.5/45.13

( 43.5 / 45.13 )

10CFR55.43.b (5)

Comments: SQN bank question AOP-S.05-B.2.A 001 modified.

Stem and distracters modified, Correct answer relocated, Mitigating strategy element for reactor trip included

(,

QUESTIONS REPORT for 2009 SRO Retake Exam 057 AG 2.2.44 080 Given the following:

Unit 1 is at 38% RTP.

PZR Pressure Channel IV (1-PT-68-322) has failed HIGH

- All actions of AOP-1.04, "Pressurizer Instrument Malfunction," have been taken and all required bistables have been placed in the required Tech Spec position.

Subsequently, an automatic Reactor trip occurs and the OATC reports the following conditions:

Reactor First Out "Pressurizer Low Pressure Reactor Trip" annunciator flashing.

- Steam Dumps closed with Tavg at 550°F.

- CCP suction swapped to RWST with VCT level at 38%.

1-FI-63-93A, Charging Flow, indicates '0' flow.

MDAFW pump B failed to automatically start.

- Safety Injection is not actuated.

Which ONE of the following correctly identifies ...

(1) if the SRO should direct the initiation of a Safety Injection, and AOP-P.03, (2) which section of AOP-P .03, "Loss of Unit 1 Vital Instrument Power Board"

( should be implemented in conjunction with the applicable emergency procedures?

Safety injection should ... AOP-P.03 A. be initiated. Section 2.1, Loss of 120v AC Vital Instrument Power Board 1-1.

B. be initiated. Section 2.2, Loss of 120v AC Vital Instrument Power Board 1-11.

C. NOT be initiated. Section 2.1, Loss of 120v AC Vital Instrument Power Board 1-1.

D~ NOT be initiated.

D!' Section 2.2, Loss of 120v AC Vital Instrument Power Board 1-11.

QUESTIONS REPORT for 2009 SRO Retake Exam TRA CTOR ANAL YSIS:

DIS TRACTOR A. Incorrect, Safety injection should not be directed for the condition as it is not required and Section 2. 1 is not the correct procedure section to be implemented.

Plausible because if the 2 instruments out of service and failed had been any combination other than a Channel IV instrument, the Safety Injection would have been designed to occur automatically and the conditions in the stem are either the same or similar to conditions requiring implementation of Section 2.1. The reactor trip, steam dumps, CCP suction, and safety injection) would be the same for a Channell failure and a Channell failure will also affect but in a different way the charging flow and AFW system.

B. Incorrect, Safety Injection should not be directed for the condition as it is not required but the conditions in the stem are associated with a Channel II failure and Section 2.2 is the correct procedure section to be implemented. Plausible because if the 2 instruments out of service and failed had been any combination other than a Channel IV instrument, the Safety Injection would have been designed to occur automatically and implementing Section 2.2 of the procedure is correct and implementing section 2.2 is correct.

C. Incorrect, The Safety Injection should not be directed, Channel IV is out of service with its bistables tripped, but this instrument does not input to the 2 out of 3 Safety Injection logic and Section 2.1 is the incorrect section of the procedure to be implemented. Plausible because not initiating Safety Injection is correct and the conditions in the stem are either the same or similar to conditions requiring implementation of Section 2.1. The reactor trip, steam dumps, CCP suction, and safety injection) would be the same for a Channell failure and a Channell failure will also affect but in a different way the charging flow and AFW system.

D. CORRECT, the loss of 120v Vital Instrument Power Board 1-11 results in the control room indications identified in the stem. The reactor trip occurs because the board loss causes a pressureinstfument second pressurizer pressure instrument to fail. This combined with the instrument already out of service makes the 2 2 out of 4410giCfor logic for the reactor trip. The Safety Injection logic is a 2 out of 3 logic but the instrument out of o(service

'service is not one of the 3 inputs to the logic, logic, therefore a Safety Injection should not have occurred and is not required. If the 2 instruments out of service and failed had been any combination other than a Channel IV instrument, the Safety Injection would have occurred automatically. The procedure section to be implemented to address the failure is Section 2.2.

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 80

(

Tier 1 Group 1 KIA 057 AG 2.2.44 Loss of Vital AC Instrument Bus Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Importance Rating: 4.2/4.4 Technical

Reference:

AOP-P.03, Loss of Vital Instrument Power Board, Rev 21 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271.AOP-P.03 & & 04 B.2 and 4 Describe the AOP-P.03 & & -P.04 entry conditions

a. Describe the setpoints, interlocks and automatic actions associated with AOP-P.03 & & P.04 entry conditions.

Upon entry into AOP-P.03 & & P.04, diagnose the applicable condition and transition to the appropriate procedural section for response.

Question Source:

(( Bank# - __ -_ -_-

Modified Bank # _-

X- _- _

New _ __

Question History: SQN question AOP-P.03-B.9 001 modified Question Cognitive Level:

Memory or fundamental knowledge _ _~___

Comprehension or Analysis _X _ __ _

10 CFR Part 55 Content: ( 41.5/43.5/45.12 )

10CFR55.43.b (5)

Comments: Stem conditions changed, 2nd part of question changed, distracter modified, correct answer relocated.

QUESTIONS REPORT for 2009 SRO Retake Exam 062 AG 2.1.7 081 Given the following:

Unit 2 was operating at 100% RTP when a loss of all ERCW occurred.

The operating crew implements AOP-M.01, "Loss of Essential Raw Cooling Water" in response to the event.

Eight minutes after entering the AOP, the STA is monitoring Status Trees and reports the Intermediate Range Monitors indicate a positive Startup Rate.

Which ONE of the following identifies the c9 ~F~t p!e'ct use of procedures?

A. The Status Tree condition requires a transition to FR-S.1, "Nuclear Power Generation / ATWS."

B~ Continue in AOP-M.01, Status Tree monitoring is required for information ONL Yeven4hough ONLY-even though an ORANGE path exists.

C. The Status Tree condition requires FR-S.1, "Nuclear Power Generation /

ATWS" to be implemented in parallel with AOP-M.01.

D. Acknowledge the YELLOW path, continue in AOP-M.01, -and instruct STA that Status Tree monitoring is NOT applicable while performing AOP-M.01.

(

DISTRACTOR DIS TRA CTOR ANAL YSIS:YS/S:

A. Incorrect, while performing AOP-M01 in response to to a loss of all ERCW, the status trees are monitored for information only, not to make transitions. Plausible because the Status Tree indicates an orange path and orange paths normally require immediate transition to the applicable Function Restoration Procedure.

B. CORRECT, AOP-M01 section for loss of all ERCW ERCWidentifies identifies that the EOPs, except for ECA-O.O, are not applicable when AOP-M01 is being used in response to a loss of all ERCW. There is a step that directs the Status trees to be monitored for information only and that the SM and TSC should be notified of any red or orange path conditions. The STA would be monitoring the Status trees and if the IRM indicated a Positive startup rate, then an orange path would exist but no transition would be made.

C. Incorrect, while performing AOP-M AOP-M01 01 in response to a loss of all ERCW, the status trees are monitored for information only. Plausible because the Status Tree indicates an orange path and orange paths normally require immediate transition and there are times that Emergency procedures and AOP are used in parallel.

D. Incorrect, the path is not a Yellow path but while continuing in AOP-M01is correct, Status Tree monitoring in the AOP is also required. Plausible because if the Source range were energized there is a decision point on the status tree for IRM startup rate that could result in a yellow path and also because Status Trees are not normally monitored in AOPs.

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 81 Tier 1 Group 1 KIA 062AG2.1.7 062 AG 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, anqancj instrument interpretation.

Importance Rating: 4.4 / 4.7 Technical

Reference:

2-FR-0, Unit 2 Status Trees, Rev 1 AOP-M.01, Loss of Essential Raw Cooling Water, Rev 20 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271AOP-M.01 B.8.b Given a set of plant conditions use AOP-M.01 to correctly:

b. Identify required actions Question Source:

Bank# _ __ __ __ __

Modified Bank # - _- __ -_-

New -X- -

Question History: New question

((

Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X__

10 CFR Part 55 Content: ( 41.5 //43.5 43.5 / 45.12 / 45.13 )

10CFR55.43.b (5)

Comments: New question

QUESTIONS REPORT for 2009 SRO Retake Exam 060 AA2.05 082 Given the following:

(

Both Units in service at 100% power.

- Waste Gas Decay Tank 'B' is the inservice tank, contains high activity gas, and is currently at 92 psig.

- Waste Gas Decay Tank 'G' is being released.

If the Waste Gas Decay Tank 'B' safety relief valve opens, which ONE of the following identifies both ...

(1) the radiation monitor that would detect the inadvertent release from tank 'B' and

<;.letected?

(2) how the releases would be affected due to the radiation being Qetected?

A. (1) O-RE-90-118, 0-RE-90-118, Waste Gas Radiation Monitor; (2) the offsite release from both tanks would be stopped.

B:o" (1) 0-RE-90-118, Waste Gas Radiation Monitor; B!'"

the offsite release from tank 'G' would be stopped, but additional (2) the-manual action would be required to stop the offsite release from tank 'B.'

c. (1) 1-RE-90-1 01, Unit 1 Auxiliary Building Stack Radiation Monitor; (2) the ABI signal generated would terminate the offsite releases from

( both tanks.

D. (1) 1-RE-90-1 01, Unit 1 Auxiliary Building Stack Radiation Monitor; (2) the ABI signal generated would terminate the offsite release from tank 'B' but manual action would be required to stop the offsite release from tank 'G.'

QUESTIONS REPORT for 2009 SRO Retake Exam DIS TRACTOR ANAL YSIS:

(

A. Incorrect, O-RE-90-11B O-RE-90-118 will detect the high radiation but the automatic isolation resulting from detecting the radiation would only isolate the release from tank 'G'.

The release from tank 'B' would continue. Plausible because the high radiation

'G'is signal isolating tank 'G' is correct and the applicant conclude the relief valve line also passes through the normal release automatic isolation valve.

B. CORRECT, 0-RE-90-11B O-RE-90-118 will detect the high radiation and isolate the in progress release from tank 'G' but the relief valve line enters downstream of the isolation valve and would continue to be released to the shield building stack until manually isolated.

C. Incorrect, The release from neither the tank being released nor the tank with the relief valve open would be detected by the Auxiliary Building Stack Radiation Monitor,O-RM-90-101.

Monitor, 0-RM-90-101. Plausible because if the relief valve had been leaking out O-RM-90-101 would have detected the the bonnet into the Auxiliary Building, then 0-RM-90-101 release and caused an ABI which would have stopped the offsite release of the Tank 'B' and because the ABGTS fan being used for the release would have a reduced flow that could result in the normal release line automatically closing to stop the release from tank 'G. '

D. Incorrect, The release from neither the tank being released nor the tank with the relief valve open would be detected by the Auxiliary Building Stack Radiation Monitor,0-RM-90-101.

Monitor,O-RM-90-101. Plausible because if the relief valve had been leaking

( out the bonnet into the Auxiliary Building, then O-RM-90-101 0-RM-90-101 would have detected the release and caused an ABI which would have stopped the offsite release of the Tank 'B' and if the reduced flow interlock was not applied, the release from tank 'G' would continue.

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 82 Tier 1 Group 2 KIA 060 AA2.05 Accidental Gaseous-Waste Release Ability to determine and interpret the following as they apply to the Accidental Gaseous Radwaste:

That the automatic safety actions have occurred as a result of a high ARM system signal Importance Rating: 3.7/4.2 Technical

Reference:

1,2-47W611-77-4 R10 1,2-47W830-4 R45 Proposed references to be provided to applicants during examination: None Learning Objective: OPT200.GRW B.5.d Describe the operation of the GRW system:

d. How a component failure will affect system operation.

Question Source:

Bank# _ _ _ ___

_X,-_ _

Modified Bank # _X'---_

( New _ __

Question History: SQN bank question GRW-B.4 003 modified.

Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _ X _

CFR Part 55 Content: ( 43.5 1/ 45.13 )

10CFR55.43.b (4)

Comments: SQN bank question GRW-B.4 003 modified.

QUESTIONS REPORT for 2009 SRO Retake Exam 061 AG2.4.46 083 Given the following:

Both Units at 100% power.

A dry cask storage campaign is in progressprogresswith with removal of the HI-TRAC from the Cask Loading Area occurring.

The following annunciators alarm:

RA-90-1A, "AUX BLDG AREA RAD MON HIGH RAD" alarms.

RA-90-102A, "FUEL POOL RAD MONITOR HI RAD" alarms.

RA-90-103A, "FUEL POOL RAD MONITOR HI RAD" alarms.

The CRO reports that 0-RA-90-102A and 0-RA-90-103A are blocked.

The alarms received are_ _..>...( 1!...J.}_ _ _ with the evolution in progress

->-(1:...J.}

and _ _ _ _ _ _ _ _~(2=)_~------------------

~(2~}_~------------------

A"! (1) consistent A:I (2) the removal of the HI-TRAC may continue B. (1) consistent (2) the removal of the HI-TRAC must be stopped due to no ABGTS train operable

(

C. (1) inconsistent (2) the removal of the HI-TRAC may continue D. (1) inconsistent (2) the removal of the HI-TRAC must be stopped due to no ABGTS train operable

QUESTIONS REPORT for 2009 SRO Retake Exam DISTRACTOR DIS TRA CTOR ANAL YSIS:

A. CORRECT, The annunciation of Spent Fuel Pit Rad Monitors is expected and are blocked due to this expectation to prevent an ABI. The evolution of lifting the HI-TRAC may continue as allowed by procedures.

B. Incorrect, The first part of this distracter is plausible as it is correct, the second part is plausible as the ABGTS trains are inop, however the basis document for 3/4.9.12 considers movement of a Cask loaded spent fuel assemblies as outside this spec and movement may continue.

C. Incorrect, The first part of this distracter is plausible as radiation alarms are generally unexpected and not consistent in most evolutions. The second part of the distracter is plausible as it is correct.

D. Incorrect, the first part of this distracter is plausible as radiation alarms are generally unexpected and not consistent in most evolutions. The second part is plausible as the ABGTS trains are inop, however the basis document for 3/4.9.12 considers movement of a Cask loaded spent fuel assemblies as outside this spec and movement may continue.

(

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 83

((

Tier 1 Group 2 KIA 061 AG2.4.46 Area Radiation Monitoring System Alarms Ability to verify that the alarms are consistent with the plant conditions Importance Rating: 4.2 14.2 4.2/4.2 Technical

Reference:

Technical Specifications 3/4.9.12 ABGTS Amendment

263/301and 3/4.9.12 Basis.

0-SO-90-5, Area Radiation Monitors, R7.

SQN-DCS-200.2, SQN-MPC-Loading and Transport Operations, R0010.

0-AR-M12-B (B-5), 0-RA-90-103A, Fuel Pool RAD Monitor HI RAD, R29.

Proposed references to be provided to applicants during examination: None Learning Objective: OPT200.DryCask B.5.c & B.6.a Describe the operation of the DCS system: Alarms and alarm response Describe the administrative controls and limits for the DCS system: State Tech SpecslTRM LCOs that govern the DCS

(

Question Source:

Bank# _ _ _ ___

Modified Bank # _ _ __

New _ _X_

Question History: New Question Cognitive Level:

Memory or fundamental knowledge _ X _

Comprehension or Analysis _ __

10 CFR Part 55 Content: (41.10/43.5/45.3/45.12 )

10CFR55.43.b (2& &4 )

Comments:

QUESTIONS REPORT for 2009 SRO Retake Exam W/E13 EA2.01 084 WIE13 Given the following:

The crew is performing FR-H.2, "Steam Generator Overpressure", for an overpressure condition on SG #2.

When the step is addressed to check affected S/G(s) NR level it is noted that the SG #2 level is indicating 86% narrow range.

Which ONE of the following identifies the correct crew actions as a result of the SG level indicating 86%?

A. Continue in FR-H.2, steam release may continue until NR level indicates 100%.

8. Continue in FR-H.2, but do not initiate any steam release until TSC B.

evaluation is complete.

C. Transition to FR-H.3, "Steam Generator High Level"; Steam release may continue until NR level indicates 100%.

D~ Transition to FR-H.3, "Steam Generator High Level"; but do not initiate any D!'

steam release until TSC evaluation is complete.

( DISTRACTOR DIS TRA CTOR ANAL YSIS:

A. Incorrect, The step RNO directs a transition to FR-H.3. However candidate may correctly conclude that FR-H. 3 is lower in priority on the FR-H status tree and not recall the transition. Plausible that steam could be released since even when NR level indicates 100%, there is still significant volume before the steam generator fills with water.

B. Incorrect, The step RNO directs a transition to FR-H.3. Plausible since FR-H.2, if continued, also prohibits the release of steam with a high level (> 84%) condition until after a TSC evaluation is complete. Transition to FR-H.3 is directed from FR-H. 2 at Step 3.

FR-H.2 C. Incorrect, The RNO for the step directs the transition to FR-H.3 and the release of steam is restricted until a TSC evaluation is complete if the level exceeds 84%,

therefore with the level at 86%, the release will be restricted. Plausible to release steam since even when narrow range SG level indicates 100%, there is still significant volume before the steam generator fills with water. ;'

D. CORRECT. The RNO for the step directs the transition to FR-H.3 and FR-H.3 restricts the release of steam until a TSC evaluation is complete.

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 84

(

Tier 1 Group 2 KIA W/E13 EA2. 1 Steam Generator Over-pressure Ability to determine and interpret the following as they apply to the (Steam Generator Overpressure)

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Importance Rating: 2.9 13.4 Technical

Reference:

FR-H.2, Steam Generator Overpressure, Rev 6.

FR-H.3, Steam Generator High Level, Rev 9.

Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 FR.H.2 B.4 &.5 Describe the bases for all limits, notes, cautions, and steps of FR-H.2.

Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

(

Question Source:

Bank#

Bank # X-_-_-- _

Modified Bank # -_- __ -- _

New - - -

Question History: WBN bank question W/E13 EA2.1 (used on WBN 2008 exam)

Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X __

10 CFR Part 55 Content: (43.5 I 45.13) 10CFR55.43.b (5)

Comments: WBN bank question W/E13 EA2.1 (used on WBN 2008 exam)

Correct answer relocated, distracter locations changed.

(

\

QUESTIONS REPORT for 2009 SRO Retake Exam W/E15 G 2.4.30 085 W!E15 Given the following:

- A large break LOCA occurs of Unit 11..

During performance of the emergency instructions the crew transitioned to and completed FR-Z.1, 'High Containment Pressure" and is now currently performing E-1, "Loss of Reactor or Secondary Coolant."

- The Shift Manager has determined the required Emergency Plan declaration.

- The STA monitoring the Status Trees reports the following current containment conditions; Pressure has lowered to 2.6 psid.

Lower Containment Radiation 86 RlHR.

Upper Containment Radiation 42 RlHR.

Containment Sump Level 73%.

Which ONE of the following identifies if a transition to a Containment Functional Restoration Procedure is required and the maximum time allowed to notify the NRC of the REP declaration?

Transition to a Containment Time required for Function Restoration Procedure ... NRC notification A'I A '!I is required. 1-hour notification

(

B. is required. 4-hour notification C. is NOT required. 1-hour notification D. is NOT required. 4-hour notification

QUESTIONS REPORT for 2009 SRO Retake Exam DISTRACTOR ANAL YSIS:

CORREC~transition is required because the conditions indicate an ORANGE A. CORREC-C;:Prtransition path to F4?z.f,t.:.!Containment ffi;[/.f,t.:.!Containment Flooding, " due to the containment sump level not being I~I~s than 68%. The required NRC notification is a 1-hour report.

B. Incorreci{Kttc)sition IncorrectrKtrc)sition is required because the conditions indicate an ORANGE path tyFR-Zq, t(/FR-Zq, 'l!!Containment

'l.!!Containment Flooding, " due to the containment sump level not being le$s1han 68%. The NRC notification is a 1-hour report, not a 4-hour report.

being less1han PlauWie because the transition being required is correct and the reactor trip is a PlauW!e 4-hour immediate report.

Incorre94-1+fi8~sition is required because the conditions indicate an ORANGE C. Incorre<;t,Afra?Jsition path tel FR-Z:8. 'Containment Flooding, " due to the containment sump level not FR-Zjl"Containment

~m3n 68% but the 1-hour report is required. Plausible because the being ~han applicant could mis-apply the containment sump level required for an orange path or determine since FR-Z. FR-Z 1 had been previously completed a Containment FRP transition would not be required and the 1-hour immediate report requirement is correct.

Incorre~9' D. Incorre=9, Qnsition Ynsition is required because the conditions indicate an ORANGE path tdto FR- ~Containment Flooding, " due to the containment sump level not being Y.V. shan 68% and the NRC notification is a 1-hour report, not an 4-hour report. Plausible because the applicant could mis-apply the containment sump level required for an orange path or determine since FR-Z. FR-Z 1 had been previously

( completed a Containment FRP transition would not be required and the reactor trip is a 4-hour immediate report.

(

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 85 Tier 1 Group 2 KIA W/E15 G 2.4.30 E15 Containment Flooding 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.

Importance Rating: 2.7/4.1 Technical

Reference:

1-FR-0, Unit 1 Status Trees, Rev 1 SPP-3.5, Regulatory Reporting Requirements, Rev 21 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 FR-Z.2 B.6.a &.c Given a set of initial plant conditions use FR-O to correctly identify the:

a. Identify required actions

c. requirements when a RED or ORANGE path is diagnosed.

OPL271 SPP-3.5 B.3.d &.e For a given condition, determine the regulatory reporting requirements using appropriate reference material.

( c. State the criteria requiring four-hour notification of the NRC.

d. State the criteria requiring eight-hour notification of the NRC.

Question Source:

Bank# _ __ __ __ __

Modified Bank # - __ - -_-

New _X__

Question History: New question Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X__

10 CFR Part 55 Content: ( 41.10 / 43.5 / 45.11 )

41.10/43.5/45.11 10CFR55.43.b (5)

Comments: New question

QUESTIONS REPORT for 2009 SRO Retake Exam 004 G.2.4.44 086 Given the following:

Unit 1 has experienced an RCS leak.

- The leak has not been identified but all secondary radiation levels are normal.

- The crew has implemented AOP-R05, "RCS Leak and Leak Source Identification."

- The operating crew raised charging flow from 87 gpm to 103 gpm.

PZR level is now stable.

Which ONE of the following is required by Technical Specification LCO 3.4.6.2,"RCS Operational Leakage," and if the current rate of leakage requires a Radiological Emergency Plan (REP) declaration?

A. Reduce leakage to less than the limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or enter LCO 3.0.3; No, leakage is less than the classification threshold value.

B. Reduce leakage to less than the limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or enter LCO 3.0.3; Yes, leakage exceeds the classification threshold value.

C. Reduce leakage to less than the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or initiate a plant shutdown; No, leakage is less than the classification threshold value.

( D~ Reduce leakage to less than the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or initiate a plant shutdown; Yes, leakage exceeds the classification threshold value.

QUESTIONS REPORT for 2009 SRO Retake Exam DIS TRA CTOR ANAL YSIS:

DISTRACTOR TS 3.4.6.2 does not require the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action and LCO 3.0.3 entry. The A. Incorrect, TS leakage is not less than the threshold value for unidentified leakage (10 gpm), an NOUE classification is required. Plausible because entering LCO 3.0.3 is a condition that is required for some conditions different than as stated in the question stem and if the leakage had been identified leakage, it would have been below the 25 gpm NOUE threshold value and no declaration would be required.

B. Incorrect, TS 3.4.6.2 does not require the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action and LCO 3.0.3 entry but a declaration is required because the threshold value for unidentified leakage (10 gpm) has been exceeded. Plausible because entering LCO 3.0.3 3. O. 3 is a condition that is required for some conditions different than as stated in the question stem and a declaration being required is correct.

C. Incorrect, TS 3.4.6.2 does require the leakage to be reduced to less than the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or the unit to be placed in HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The leakage is not less than the threshold value for unidentified leakage (10 gpm), an NOUE classification is required. Plausible because the Tech Spec action is correct and if the leakage had been identified leakage, it would have been below the 25 gpm NOUE threshold value and no declaration would be required..

D. CORRECT, TS 3.4.6.2 requires the leakage to be reduced to less than the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or the unit is to be placed in HOT STANDBY within the following 6

( hours. A declaration is required because the threshold value for unidentified leakage (10 gpm) has been exceeded. This meets the Potential Loss of the RCS Barrier in EPIP-1 requiring an ALERT declaration.

(

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 86 Tier Tier 3 3 ? Ulf KIA KIA OO~4 004G.

L'L' t ch~~iJ~I;nd Chemi al and Volume Control System Knowledge of emergency plan protective action recommendations.

Importance Rating: 2.4 / 4.4 2.4/4.4 Technical

Reference:

Technical Specification 3.4.6.2, Amendment No. 322 EPIP-1, Emergency Plan Classification Matrix, Rev. 41 AOP-R05, RCS Leak and Leak source Identification, Rev. 14 NP-REP Appendix B, Tennessee Valley Authority Nuclear Power Radiological Emergency Plan, Rev 89 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 REP B.3 Classify Events using the appropriate procedures.

OPL217AOP-R05 OPL217AOP-R05 B.9 Describe the Tech Spec and TRM actions applicable during the performance of AOP-R05.

Question Source:

( Bank# _ _ __

Modified Bank # _ _ __

New _X__

Question History: New question Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X__

10 CFR Part 55 Content: 41.10 / 41.12 / 43.5 / 45.11 )

( 41.10/41.12/43.5/45.11 10CFR55.43.b ( 2,5 )

Comments: New question

(

QUESTIONS REPORT for 2009 SRO Retake Exam 005 A2.04 087 Given the following conditions:

RCS temperature is 178°F with the RCS vented.

RHR Train 'A' is in service.

0-GO-13, "Reactor Coolant Systems Fill and Drain Operations !8r-aJ19- Hr-aJA-Down" Section 5.1.1, "To Partial Drain Conditions" is in progress.

Down" 1-Ll-68-321, "RCS PZR Level-Cold Cal," indicates 10%.

Both trains of Mansell indicate level at Elevation 715'9".

Subsequently, the following indications are observed:

RCS temperature is 183°F and rising.

RHR Pump1 A-A flow is 0 GPM.

Pump1A-A RHR Pump1A-A discharge pressure is 180 psig.

RHR Pump1 A-A current is stable at approximately 10 amps.

Pump1A-A 1-Ll-68-321 indicates 11 %.

Both trains of Mansell indicate level at Elevation 716'2' and rising.

The crew enters AOP-R.03, "RHR System Malfunction."

Which ONE of the following describes (1) the event in progress, and (2) the section of the procedure that will be implemented?

Event in progress Procedure section entry required

(

A. RHR Pump sheared shaft Section 2.3, RHR pump(s) failure or trip.

B. RHR Pump sheared shaft Ses!i(m2,J~~HR malfunctions due Seg!iQn2,J-l-.RHR to(!<:>\tV_vyat~r to(low

"--~~~"' ___

water"",)evel

>~._.C_~'" _"_.~

~,tevel during reduced inventory or mid-loop operations.

C~ RHR system valve failure C!' Section 2.3, RHR pump(s) failure or trip.

D. RHR system valve failure Sectlon Section 2.1, RHR malfunctions due to low water level during reduced inventory or mid-loop operations.

QUESTIONS REPORT for 2009 SRO Retake Exam TRACTOR DIS TRAC TOR ANAL YSIS:

A. Incorrect, If the pump shaft were sheared the pressure would not indicate 300 psig. Plausible because the conditions except for pressure would exist for a sheared shaft and the procedure section to be used is correct.

B. Incorrect, If the pump shaft were sheared the pressure would not indicate 300 psig. Plausible because the conditions except for pressure would exist for a sheared shaft and a drain down has been initiated with the level reduced from normal.

C. CORRECT, Pump is operating against closed valve, RCS is heating up, and resulting in small volume change and the level is above the reduced inventory range.

D. Incorrect, Conditions do indicate a valve problem, but the level is above the level (eI699/. Plausible because the conditions are for entering reduced inventory (eI699').

correct for a valve problem and a drain down has been initiated with the level reduced from normal.

(

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 87

(

Tier 2 Group 1 KIA 005 A2.04 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:

RHR valve malfunction Importance Rating: 2.9/2.9 Technical

Reference:

AOP-R03, RHR System Malfunction, Rev 21 GO-13, Reactor Coolant Systems Fill and Drain Operations, Rev 18 Proposed references to be provided toapplicants to applicants during examination: None Learning Objective: OPL271AOP-R03 Upon entry into AOP-R03, diagnose the applicable condition and transition to the appropriate procedural section for response.

( Question Source:

Bank # X"_-_ --

Modified Bank # - _- __ -_-

New _ __

Question History: Question from Ginna bank Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X__

10 CFR Part 55 Content: ( 41.5 / 43.5 / 45.3 / 45.13 )

10CFR55.43.b (5)

Comments: Ginna bank question changed to make applicable to SQN

(

QUESTIONS REPORT for 2009 SRO Retake Exam 012 A2.05 088 Given the following:

Unit 1 at 100% power.

- An Eagle 21 malfunction has occurred that caused several annunciators to alarm and unexpected unexpectedJeactor reactor trip bi-stables to be LIT.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the event, MIG is prepared to initiate work on the failure.

In accordance with AOP-1.11, "Eagle 21 Malfunction," which ONE of the following identifies ...

(1) the type of failure that has occurred and (2) the strategy used in responding to the event as related to the attempt to reset the (2)the system and the hard tripping of the associated bi-stables?

(1 ) (2)

Type Failure Strategy Used A. Test Setpoint Bistables in accordance with applicable AOP Trip Bistablesin Processor (TSP) before attempting System Reset.

B. Test Setpoint Attempt System Reset in accordance with applicahle Processor (TSP) Maintenance PI before tripping Bistables.

( C. Loop Control Trip Bistables in accordance with applicable AOP Processor (LCP) before attempting System Reset.

D~ Loop Control Attempt System Reset in accordance with applicable Processor (LCP) Maintenance PI before tripping Bistables.

QUESTIONS REPORT for 2009 SRO Retake Exam DIS TRACTOR ANAL YSIS:

DISTRACTOR

(

A. Incorrect, the failure is not on the TSP (it is on the LCP) and the hard bi-stables should not be tripped prior to an attempted system reset. Plausible because the failure type would be on the TSP if no additional bi-stables were lit and because if there was not time enough to make the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months TIS to have bi-stables tripped the bi-stables would be tripped prior to an attempted system reset.

B. Incorrect, the failure is not on the TSP (it is on the LCP) but an attempt of a System Reset should be completed prior to tripping the hard bi-stables. Plausible because the failure type would be on the TSP if no additional bi-stables were lit and because the attempted system reset should be performed prior to tripping the bi-stables.

C. Incorrect, the failure is on the LCP but the hard bi-stables should not be tripped prior to an attempted system reset. Plausible because the failure type is correct and because if there was not time enough to make the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months TIS to have bi-stables tripped the bi-stables would be tripped prior to an attempted system reset.

D. CORRECT, If bi-stables are LIT due to the failure, the failure is on the LCP and AOP-I. 11 and the AOPs for instrument failure response have notes and cautions AOP-I.11 stating that the a System reset should be attempted prior to any bi-stables being hard tripped.

(

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 88

(

Tier 2 Group 1 KIA 012 A2.05 Reactor Protection System Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Faulty or erratic operation of detectors and function generators Importance Rating: 3.1** 1 3.1 / 3.2*

Technical

Reference:

AOP-I-11, Eagle 21 Malfunction, Rev 9 Proposed references to be provided to applicants during examination: None Learning Objective: OPT200.EAGLE21 B.5.c & .d Describe the operation of the Eagle 21 system:

c. alarms and alarm response

d. How a component failure will affect system operation.

Question Source:

Bank# _ _ __

Modified Bank # _X. X._ __

( New _ __

Question History: SQN bank question AOP-1.11-B.1 003 modified Question Cognitive Level:

Memory or fundamental knowledge _ X _

Comprehension or Analysis _ __

10 10 CFR Part 55 Content: ((41.5/43.5/45.3/45.5 41.5/43.5/45.3/45.5 )

10CFR55.43.b ((5 5)

Comments: SQN bank question AOP-1.11-B.1 003 modified

(

QUESTIONS REPORT for 2009 SRO Retake Exam 059 A2.06 089 Given the following:

(

- Unit 1 has experienced a Safety Injection due to a steam line break inside containment.

- MSIVs closed automatically.

- Subsequently, all AFW has been lost.

- Containment pressure has peaked at 1.2 'psig and is lowering.

- WR S/G levels as follows:

S/G 1 S/G22 S/G S/G3 S/G 3 S/G4 23%WR 27%WR 25% WR 5%WR

- FR-H.1, "Loss of Secondary Heat Sink" is being implemented with the crew attempting to restore MFW flow to at least one S/G in accordance with EA-2-2, "Establishing Secondary Heat Sink Using Main Feedwater or Condensate System."

Which ONE of the following identifies ...

(1) the mitigating strategy to be used, (1) and MIG support would be required to block auto SI signals?

(2) if MIG

" (1 ) Establish RCS feed and bleed per FR-H.1 A. (1) FR-H.1..

( (2) MIG support is required for blocking auto SI signals.

(1 ) Establish RCS feed and bleed per FR-H.1.

B. (1)

(2) No MIG support is required for blocking auto SI signals.

(1 ) Continue to establish feed flow from the condensate system per EA-2-2.

C. (1)

(2) MIG support is required for blocking auto SI signals.

D~ (1) Continue to establish feed flow from the condensate system per EA-2-2.

(2) No MIG support is required for blocking auto SI signals.

QUESTIONS REPORT for 2009 SRO Retake Exam DIS TRA CTOR ANALYSIS:

DISTRACTOR ANAL YSIS:

((

A. Incorrect, RCS Bleed and feed is not required as levels in at least 3 S/Gs are>

20% wide range with containment not adverse (2.81 psig). With containment pressure only reaching 1.2 psig (verses 1.5 psig) the auto high containment pressure SI is not active. This SI is not blocked from the MCR boards (La Steam and La Pressurizer Pressure SIs Sis are blocked from the MCR) and requires MIG support. In this case blocking the signal for high containment pressure SI is not required. The first part of this distracter is plausible as the examinee will need to understand that the containment is not adverse and S/G levels are above RCS feed and bleed criteria. The second part is plausible as the examinee will have to understand that the auto containment high pressure SI did not occur. MIG is required to assist in blocking that signal if it did.

B. Incorrect, RCS Bleed and feed is not required as levels in at least 3 S/Gs are>

20% wide range with containment not adverse (2.81 psig). With containment pressure only reaching 1.2 psig (verses 1.5 psig) the auto high containment pressure SI is not active. This SI is not blocked from the MCR boards (La Steam and La Pressurizer Pressure SIs Sis are blocked from the MCR) and requires MIG support. In this case blocking the signal for high containment pressure SI is not required. The first part of this distracter is plausible as the examinee will need to understand that the containment is not adverse and S/G levels are above RCS feed and bleed criteria. The second part is plausible because it is correct.

C. Incorrect, with the MSIVs closed the option to use MFW to initiate feed to the S/Gs

( is not possible and the next procedural feed source is using condensate. With containment pressure only reaching 1.2 psig (verses 1.5 psig) the auto high containment pressure SI is not active. This SI is not blocked from the MCR boards (La Stea,m and La Pressurizer Pressure Sis SIs are blocked from the MCR) and requires MIG support. In this case blocking the signal for high containment pressure SI is not required. The first part of the distracter is plausible because it is correct. The second part is plausible as the examinee will have to understand that the auto containment high pressure SI did not occur. MIG is required to assist in blocking that signal if it did.

D. CORRECT, with the MSIVs closed the option to use MFW to initiate feed to the S/Gs is not possible and the next procedural feed source is using condensate.

With containment pressure only reaching 1.2 psig (verses 1.5 psig) the auto high containment pressure SI is not active. This SI is not blocked from the MCR boards (La Steam and La Pressurizer Pressure Sis SIs are blocked from the MCR) and requires MIG support. In this case blocking the signal for high containment pressure SI is not required.

(,

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 89 Tier 2 Group 1 KIA 059 A2.06 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:

Loss of steam flow to the MFW system Importance Rating: 2.7/2.9 2.7 I 2.9 Technical

Reference:

FR-H.1, Loss of Secondary Heat Sink, Rev 17 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 FR-H.1 B.5.a & .b & B.6.e & i Describe the conditions and reason for transitions within this procedure and transitions to other procedures Given an initial set of plant conditions use FR-H.1 to correctly:

a. identify required actions

b. respond to contingencies OPL200.RPS B.4.eBA.e & i Describe the following characteristics of each major

( component in the Reactor Protection and Engineered Safety Actuation Systems:

e. Component operation

i. Protective features (including setpoints)

Question Source:

Bank# _ _ __

Modified Bank # - _- _- _- _- _

New _X'--__

X._ __

Question History: New question Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X, __

_Xc--_

10 CFR Part 55 Content: 41.5 I 43.5 I 45.3 I 45.13 )

((41.5/43.5/45.3/45.13 10CFR55A3.b 10CFR55.43.b (5)

Comments:

QUESTIONS REPORT for 2009 SRO Retake Exam 064 G 2.4.8 090 Given the following:

Both units at 100% power with all system normal.

- The 6.9kV Unit Board 2B trips due to electrical fault resulting in a reactor trip.

The Diesel Generators (DG) start but DG 2A-A fails to connect to the board.

Unit 2 operating crew is ready to transition from E-O, "Reactor Trip or Safety Injection."

Which ONE of the following identifies both ...

(1) the action required when Diesel Generator (DG) 2A-A ERCW cooling is checked and (2) the correct procedure implementation by the Unit Supervisor in accordance with EPM-4, "User's Guide?"

Note: ES-0.1, "Reactor Trip Response" AOP-P.06, "Loss of Unit 2 Electrical Shutdown Board" (1 ) (2)

DG 2A-A would ... Procedure Implementation A"! be stopped because it is A'! Procedure reader will normally running without ERCW cooling. implement ES-0.1 ES-0.1,, while handing off performance of AOP-P .06 to another crew member.

B. be stopped because it is Procedure reader will normally running without ERCW cooling. AOP-P .06, while handing implement AOP-P.06, off performance of ES-0.1 to another crew member.

C. have ERCW being supplied Procedure reader will normally from the normal source. implement ES-0.1 ES-0.1,, while handing off performance of AOP-P .06 to AOP-P.06 another crew member.

D. have ERCW being supplied Procedure reader will normally from the normal source. implement AOP-P.06, while handing off performance of ES-0.1 to another crew member.

(

QUESTIONS REPORT for 2009 SRO Retake Exam DIS TRA CTOR ANAL YSIS:

(

A. CORRECT, DG 2A-A would be stopped because while there is water available in the normal supply, the motor operated valves in both the normal and alternate supply would be closed and the boards supplying power de-energized and in accordance with EPM-4, AOPs can be used with EOPs. The AOP would be implemented by a person other than the procedure reader in order that the procedure reader would stay focused on the EOP until the SM directs otherwise.

B. Incorrect, DG 2A-A would be stopped because of no available ERCW source and the procedure reader would not hand ES-O. 1 off to other crew member. The US ES-O. 1 and hand the AOP would continue in ES-O.1 A OP off. Plausible because stopping the DG due to no ERCW is correct and handing ES-O.1ES-O. 1 off to focus on a AOP occurred at the plant in the past and resulted in lack of focus on stabilizing the plant and also because it can be done but only with SM direction.

C. Incorrect, the DG 2A-A normal ERCW supply header would have pressure from the 1 1AA ERCW header but the supply valve for the DG could not be opened due to the board supplying the motor being de-energized and the procedure reader would continue in ES-O.1 while handing off the AOP. Plausible because the 1A ERCW is available and the MOV normally opens when the DG starts and continuing in ES-O.1 while handing off the AOP is correct.

D. Incorrect, the DG 2A-A normal ERCW supply header would have pressure from the 11A A ERCW header but the supply valve for the DG could not be opened due to

( the board supplying the motor being de-energized and the procedure reader would not implement the AOP while handing off ES-O.1. Plausible because the 1A ERCW is available and the MOV normally opens when the DG starts and handing ES-O.1 off to focus on an AOP occurred at the plant in the past and resulted in lack of focus on stabilizing the plant and also because it can be done but only with SM direction.

I

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 90 Tier 2 Group 1 KIA 064 G 2.4.8 Emergency Diesel Generator (ED/G) System Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Importance Rating: 3.8/4.5 3.8 / 4.5 Technical

Reference:

1,2-15E500-1, R28 EPM-4, User's Guide, Rev 20 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 EPM-4 B.8 Given plant operating conditions determine if AOP entry conditions have been met and state the resultant appropriate operator actions for those conditions.

OPT200.DG B.4.c Describe the following items for each major component in the Diesel Generator system:

c. Support equipment and systems.

( Question Source:

Bank# _ _ _ __ ___

Modified Bank # _ __ __ __ __

New _X__

Question History: New question Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X__ __

10 CFR Part 55 Content: ( 41.10/43.5/45.13 41.10 / 43.5 / 45.13 )

10CFR55.43.b (5)

Comments: New question

(

QUESTIONS REPORT for 2009 SRO Retake Exam 028 G2.1.23 091 During the performance of ES-1.2, "Post LOCA Cooldown and Depressurization," which ONE of the following is the minimum containment hydrogen concentration, which if exceeded, would require the SRO to notify the Technical Support Center (TSC), and why?

Hydrogen Concentration Reason A. 3.0% To evaluate and determine a hydrogen recovery strategy.

B. 3.0% To determine maximum allowable vessel venting time.

C~ 6.0% To evaluate and determine a hydrogen recovery strategy.

D. 6.0% To determine maximum allowable vessel venting time.

DISTRACTOR ANAL YSIS:

A. Incorrect. Plausible because the 3.0% is the concentration used to determine "maximum allowable vessel venting time" in FR-1.3, "Voids in the Reactor Vessel,"

but the TSC is notified to calculate the time if hydrogen concentration is less than 3%, not to determine further recovery actions.

(

B. Incorrect. Plausible because the 3.0% is the concentration used to determine "maximum allowable vessel venting time" in FR-1.3, "Voids in the Reactor Vessel,"

but the TSC is notified to calculate the time if hydrogen concentration is less than 3%.

C. CORRECT. If containment hydrogen concentration reaches or exceeds 6%, the SRO is directed to notify the TSC for additional guidance, per ES-1.2.

D. Incorrect. Plausible because the 6.0% is correct but it is not to determine "maximum allowable vessel venting time."

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 91 Tier 2 Group 2 KIA 028 G2.1.23 Hydrogen Recombiner and Purge Control System (HRPS)

Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Importance Rating: 4.3/4.4 Technical

Reference:

ES-1.2, Post LOCA Cooldown and Depressurization, Rev. 17 FR-1.3, Voids in the Reactor Vessel, Rev 11 EPM-3-ES-1.2, Basis Document for ES-1.2 Post LOCA Cooldown and Depressurization, Rev 5 Proposed references to be provided to applicants during examination: None Learning Objective: OPT200.CGCS B.4.d,.e & & .i Describe the following characteristics of each major component in the Combustible Gas Control system:

d. Normal operating Parameters

e. Component operation

i. Protective Features (including setpoints)

(' Question Source:

Bank # X._ __

X Modified Bank # _ _ __

New _ __

Question History: WBN question Question Cognitive Level:

Memory or fundamental knowledge _X_

Comprehension or Analysis _ __

10 CFR Part 55 Content: ( 41.10/

41.10 / 43.5 / 45.2 / 45.6 )

10CFR55.43.b (5)

Comments: WBN question

(

QUESTIONS REPORT for 2009 SRO Retake Exam 068 A2.04 092 Given the following:

- A planned release of the Cask Decontamination Collector Tank (CDCT) was started at 1300 and terminated at 1415.

- Subsequently, it was determined that counts were higher than expected and that the radiation monitor RM-90-122 had failed to automatically isolate the release.

- The release count rate was determined to have been 1.75 E+06 cpm for the entire release.

Which ONE of the following identifies (1) the Radiological Emergency Plan EAL Classification met due to the release and (2) the action required?

Reference Provided (1 )

(1) (2)

EAL Criteria met Action Required A. Notification of Unusual Event (NOUE). Declare the event.

B~ Notification of Unusual Event (NOUE). Report the event but do NOT Declare.

( *because the condition C. ALERT because Declare the event.

existed for greater than 15 minutes.

D. ALERT because the condition Report the event but do existed for greater than 15 minutes. NOT Declare.

QUESTIONS REPORT for 2009 SRO Retake Exam DIS TRA TRACTOR CTOR ANAL YSIS:

A. Incorrect, The NOUE declaration criteria was met because the release duration was more than 60 minutes but the declarations would not be made because the conditions have been resolved. Plausible because the criteria for NOUE classification did exist and would have been declared if identified at the time of the release.

B. CORRECT, The NOUE declaration criteria was met because the release duration was more than 60 minutes and the condition would be reported but the declaration would not be made because the conditions have been resolved.

C. Incorrect, Condition existing for greater than 15 minutes does not elevate the classification to an ALERT and the event would not be declared. Plausible because there are conditions where an NOUE would become an ALERT if time limits were exceeded (ex. MCR abandonment) and an declaration would have been made if identified at the time of the release. Additionally if the release rate had been higher, an alert condition could have been met and the determination of the top of the scale versus release rate would be required.

D. Incorrect, Condition existing for greater than 15 minutes does not elevate the classification to an ALERT and the event would not be declared. Plausible because there are conditions where an NOUE would become an ALERT if time limits were exceeded (ex. MCR abandonment) and reporting the event but not declaring is correct because the conditions have been resolved. Additionally if the

( release rate had been higher, an alert condition could have been met and the determination of the top of the scale versus release rate would be required.

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 92 Tier 2 Group 2 KIA 068 A2.04 liquid Radwaste System (LRS)

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

Failure of automatic isolation Importance Rating: 3.3 I 3.3 3.3/3.3 Technical

Reference:

EPIP-1, Emergency Plan Classification Matrix, Rev. 41 NP REP Appendix B, Tennessee Valley Authority Nuclear Power Radiological Emergency plan, Rev 89 Proposed references to be provided to applicants during examination:

EPIP-1, Emergency Plan Classification Matrix, pages 43 and 46 Learning Objective: OPL271 REP B. 2

( Classify emergency events using the appropriate procedures.

Question Source:

Bank# _ _ __

Modified Bank # _ _ __

New _ X _

New_X_

Question History: New question Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X__

10 CFR Part 55 Content: ( 41.5 1 I 43.5 1 I 45.3 1 I 45.13 )

10CFR55.43.b (5)

Comments: New question

QUESTIONS REPORT for 2009 SRO Retake Exam 002 A2.02 093 Given the following:

Unit 1 is in Mode 3 with the Reactor Trip Breakers open following a planned shutdown for maintenance.

- The unit is being cooled down in accordance with GO-7, "Unit Cooldown from Hot Standby to Cold Shutdown."

- Current RCS temperature is 465°F and pressurizer pressure 1535 psig.

If a pressurizer PORV inadvertently opened ...

(1) which procedure contains the steps to address the failure and (2) if the PORV could not be closed or isolated, what criteria would be used to determine when a Safety Injection is required to be initiated?

(1 ) (2)

Procedure SI Initiation Criteria A. AOP-R.05, "RCS Leak and RCS subcooling less than 40°F.

Leak Source Identification" B. AOP-R.05, "RCS Leak and Pressurizer level less than 5%.

Leak Source Identification"

(

C'!'

C"!" AOP-1.04, "Pressurizer Instrument RCS subcooling less than 40°F.

and Control Malfunctions" D. AOP-1.04, "Pressurizer Instrument Pressurizer level less than 5%.

and Control Malfunctions"

(

QUESTIONS REPORT for 2009 SRO Retake Exam DISTRACTOR DIS TRA CTOR ANAL YSIS:

A. Incorrect, a PORVopened would be an RCS leak but AOP-R.05 is not the procedure with a section for addressing a failed open PORV, AOP-I.04 does have the section but the initiation of Safety Injection based on lack of subcooling is correct. Plausible because A AOP-R.05 OP-R. 05 is the procedure for an RCS leak and safety cooling less than 40 degrees F is correct.

subcooling injection initiation being required if sub B. Incorrect, a PORVopened would be an RCS leak but AOP-R.05 is not the procedure with a section for addressing a failed open PORV, AOP-I.04 does have the section and the initiation of Safety Injection would be required due to low sub cooling not based on low pressurizer level. Plausible because a failed open subcooling PORV is an RCS leak and pressurizer level less than 5% is criteria for initiating safety injection in the ES-0.2 and ES-0.3 cooldown procedures as well as in the ES-0.11 procedure.

ES-O.

C. CORRECT, AOP-I.04 contains a section for response to a failed open PORV.

Mitigating strategy during performance of the section while in Mode 3 is to initiate a Safety Injection prior to reaching 1870 psig (auto SI setpoint) if the low pressurizer pressure SI is not block or when subcooling is less than 40 degrees F.

The question setup has the pressure below the low pressurizer pressure SI setpoint, set point, thus it must be blocked and would have been during the performance of the GO prior to reaching the pressure stated in the stem.

D. Incorrect, AOP-I.04 contains a section for response to a failed open PORV but the

( initiation of Safety Injection would be requ~~~ow requ~fiiiii..,~ow subcooling not based on low pressurizer level. Plausible because ~ ~correct correct and pressurizer level less than 5% is criteria for initiating safety injection in the ES-0.2 and ES-0.3 cooldown procedures as well as in the ES-O. 1 procedure.

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 93 Tier 2 Group 2 KIA 002 A2.02 Reactor Coolant System (RCS)

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

Loss of coolant pressure Importance Rating: 4.2 / 4.4 4.2/4.4 Technical

Reference:

AOP -R.05, RCS Leak and Leak Source Identification, Rev. 14 AOP -1.04, Pressurizer Instrument and Control Malfunctions, Rev.9 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271AOP-1.04 B.5 8.5 Describe the mitigating strategy for the failure that initiated entry into AOP-1.04.

Question Source:

Bank# _ _ __

Modified Bank 8ank # - __ -_ -_-

New _X X__

Question History: New Question Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X__ __

10 10 CFR Part 55 Content: 41.5 /43.5 /45.3 / 45.5 )

( 41.5/43.5/45.3/45.5 10CFR55.43.b (5)

Comments: New Question

(

QUESTIONS REPORT for 2009 SRO Retake Exam G 2.1.9 094 While implementing a System Operating Instruction (SO), the CRO informs the US that the procedure directs an alarm to be disabled.

In accordance with OPDP-4, "Alarm Disablement," the US can allow the CRO to disable the alarm _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

A'! without completing either a Technical Evaluation or a 50.59 review.

A'I B. only after a Technical Evaluation and a 50.59 review are completed.

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

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

DISTRACTOR DIS TRA CTOR ANAL YSIS:

A. Correct, per OPDP-4, Appendix A section A, if an approved plant procedure allows an alarm disablement then a TE and 50.59 review are not required.

B. Incorrect, per OPDP-4, Appendix A section A, if an approved plant procedure allows an

( alarm disablement then a TE and 50.59 review are not required. This distracter is 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.

C. Incorrect, per OPDP-4, Appendix A section A, if an approved plant procedure allows an alarm disablement then a TE and 50.59 review are not required. This distracter is plausible because OPDP-4 Appendix A section B may require a TE and but not necessarily a 50.59 review for other plant conditions than those listed in the stem.

D. Incorrect, per OPDP-4, Appendix A section A, if an approved plant procedure allows an alarm disablement then a TE and 50.59 review are not required. This distracter is plausible because OPDP-4 Appendix A section B may require a 50.59 review and but not necessarily a TE for other plant conditions than those listed in the stem.

(

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 94 Tier 3 Group 1 KIA G 2.1.9 Conduct of operations Ability to direct personnel activities inside the control room Importance Rating: 2.9/4.5 Technical

Reference:

OPDP-4, Annunciator Disablement, Rev 4 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271.0PDP-4 B.3, B.7 & & B.8 Describe the responsibilities of the operating crew during disabling an Alarm.

Describe when a Technical Evaluation is required for a disabled Alarm.

Describe when a 10CFR50.59 Review is required for a disabled Alarm.

Question Source:

Bank# _ _ __

( Modified Bank # _ _,X X._ _

New _ __

Question History: SQN bank question 055 G2.4.43 091 (used on Audit Exam in 2008) modified.

Question Cognitive Level:

Memory or fundamental knowledge _X _X,_ _

Comprehension or Analysis _ __

10 CFR Part 55 Content: ((41.10/45.5/45.12/45.13 41.10/45.5/45.12/ 45.13 )

10CFR55.43.b (3)

Comments: SQN bank question 055 G2.4.43 091 (used on Audit Exam in 2008) modified.

Modified by re-writing stem for different set of initial conditions to meet KIA and provide a different correct answer.

Made closed reference question by eliminating the "Reference Provided."

Reworded distractors and correct answer.

Correct answer in different location.

QUESTIONS REPORT for 2009 SRO Retake Exam G 2.2.12095 Which ONE of the following identifies ...

(1) the maximum extension time allowed by Technical Specifications for the completion of a daily Surveillance Requirement (SR) and (2) if it was discovered the SR was not completed within the maximum allowable extension time, the longest the requirement to declare the equipment inoperable can be delayed without completing a risk assessment?

(1 ) (2)

Max extension time wlo Risk Assessment Delay time w/o A. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B!'"

8:0" 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months D. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DISTRACTOR DIS TRACTOR ANAL YSIS:

( A. Incorrect, The extension time is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months but the delay time is not up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months without completing a Risk Assessment must (it is up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). Plausible because the extension time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is correct and the delay time being 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is equal to the allowed extension time and can be mistaken.

B. CORRECT, A surveillance requirement must be completed within its specified frequency plus 25% extension (6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for a 24 surveillance requirement) but if it was discovered that the surveillance requirement was not completion within the specified time plus the extension time, the decision to declare the equipment inoperable could be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the delay is to be past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a Risk Assessment must be completed.

C. Incorrect, The extension time is not 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (it is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) but the delay time being up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is correct. Plausible because the time allowed for a missed surveillance is up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and can be confused with the extension time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Also, the delay time being 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is equal to the allowed extension time and can be mistaken.

D. Incorrect, The extension time is not 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (it is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) and the delay time is not up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months without completing a Risk Assessment (it is up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Plausible because the time allowed for a missed surveillance is up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and can be confused with the extension time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Also, the delay time being 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is correct.

(

\

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 95 Tier 3 KIA G 2.2.12 Knowledge of surveillance procedures.

Importance Rating: 3.7 / 4.1 Technical

Reference:

Technical Specification Section 3.0/4.0, Amendment No. 322 Proposed references to be provided to applicants during examination: None Learning Objective: OPT200.TS-lntro BA Explain the purpose of Tech Spec Surveillance Requirements.

Question Source:

Bank#

Bank # -_ _ __

Modified Bank # - _- __ -_-

New -X- -

Question History: New question

( Question Cognitive Level:

Memory or fundamental knowledge _X __

Comprehension or Analysis _ __

10 CFR Part 55 Content: ( 41.10 //45.13 45.13 )

10CFR55A3.b (2)

Comments: New question

(

QUESTIONS REPORT for 2009 SRO Retake Exam G 2.2.17096 2.2.17 096 Given the following:

- The Shift Manager determines that immediate action is needed to begin work on an emergent maintenance activity in parallel with the planning of the associated Work Order.

- The maintenance activity will result in an individual receiving an annual dose greater than the Administrative Dose Limit but still within the 10CFR20 limit.

Which ONE of the following identifies the Work Order Priority Code(s) required to allow work to be performed in parallel with the planning process and the lowest level of management who can authorize the individual to exceed the Administrative Limit?

Work Order Priority Code Management Authorization required _

A. Priority 1 only RADCON Shift Supervisor B~

B:' Priority 1 only Radiation Protection Manager C. Priority 1 or 2 RADCON Shift Supervisor

( D. Priority 1 or 2 Radiation Protection Manager

(

QUESTIONS REPORT for 2009 SRO Retake Exam DISTRACTOR DIS TRA CTOR ANAL YSIS:

A. Incorrect, Shift Manager assigning a Priority 1 code to the WO is correct but Site Radiation Protection Manager (not the RADCON Shift Supervisor) authorization is required to exceed the Administrative Dose Limit. Plausible because assigning a Priority 1 code to the WO is correct and the RADCON Shift Supervisor has functions related to work controlled by an RWP.

B. CORRECT, In accordance with SPP-6. 1, if the Shift Manager assigns a Priority 1 code to the WO, the work and planning can be performed in a parallel and in accordance with RCI-03, the Site Radiation Protection Manager authorization is required to exceed the Administrative Dose Limit.

C. Incorrect, A WO coded Priority 2 is an immediate attention situation. It must be planned and worked continuously (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day) until completion. (Not planned and worked in parallel). Plausible because a WO assigned a Priority 2 causes immediate action to get the WO planned and worked until completion and the RADCON Shift Supervisor has functions related to work controlled by an RWP.

D. Incorrect, A WO coded Priority 2 is an immediate attention situation. It must be planned and worked continuously (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day) until completion. (Not planned and worked in parallel). Plausible because a WO assigned a Priority 2 causes immediate action to get the WO planned and worked until completion and the Site Radiation Protection Manager authorization is required to exceed

( the Administrative Dose Limit.

(

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 96 Tier 3 KIA G 2.2.17 G2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

Importance Rating: 2.6/ 3.8 Technical

Reference:

SPP-7.1, On Line Work Management, Rev. 0013 SPP-6.1, Work Order Process Initiation, Rev 0006 RCI-03, Personnel Monitoring, Rev 48 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 SPP-7.1 B.8 Describe the definitions of the Work Order priority codes.

OPL271 C260 B.5 8.5 List TVA Administrative Dose Levels.

Question Source:

Bank# _ __ __ __ __

Modified Bank # _ _ __

( ----

New -X- -

Question History: New question Question Cognitive Level:

Memory or fundamental knowledge _X __

Comprehension or Analysis _ __

10 CFR Part 55 Content: 41.10/43.5/45.13

( 41.10 /43.5 / 45.13 )

10CFR55.43.b (4)

Comments:

(

QUESTIONS REPORT for 2009 SRO Retake Exam G 2.3.11 097 Given the following:

(

- E-3, "Steam Generator Tube Rupture," has been completed due to a tube rupture on #3 Steam Generator.

~\

- The crew his ready to transition to a post-SGTR cooldown procedure.

If the goal is to use the recovery procedure that will MOST limit contamination offsite radiological releases, which ONE of the following identifies ...

spread and oftsite (1) the procedure that should transitioned to, and (2) after the transition is made, which procedure is required to be implemented if #1 Steam Generator pressure starts dropping in an uncontrolled manner?

)! (,,\

,,\

A'! (1) ES-3.1, "Post-SGTR Cooldown Using Backfill" A"!

(2) E-2, "Faulted Generator Isolation" B. (1) ES-3.1, "Post-SGTR Cooldown Using Backfill" (2) ECA-3.1, "SGTR and LOCA" C. (1) ES-3.2, "Post-SGTR Cooldown Using Blowdown" (2) E-2, "Faulted Generator Isolation" D. (1) ES-3.2, "Post-SGTR Cooldown Using Blowdown"

( (2) ECA-3.1, "SGTR and LOCA"

QUESTIONS REPORT for 2009 SRO Retake Exam TRA C TOR ANAL YSIS:

DIS TRACTOR A. CORRECT, ES-3.1 is the cooldown method that will provide for the least spread of contamination and the minimum offsite dose. If the #1 steam generator pressure starts dropping in an uncontrolled manner after ES-3. 1 is implemented the Fold-Out Page directs the transition to E-2.

B. Incorrect, ES-3.1 is the cooldown method that will provide for the least spread of contamination and the minimum off offsite site dose, but the transition required if #1 steam generator pressure starts dropping in an uncontrolled manner after ES-3. 1 is implemented is not to ECA-3.1 (it is to E-2). Plausible because ES-3.1 is the cooldown procedure to be used and ECA-3. 1 is the procedure directed to be entered from the Fold Out Page for other conditions that would be present if the pressure on the ruptured steam generator had been dropping.

C. Incorrect, ES-3.2 is not the procedure to be entered to provide for the least spread of contamination and the minimum offsite dose (ES-3. 1 is) but the transition to E-2 if the #1 steam generator pressure starts dropping in an uncontrolled manner after ES-3. 1 is implemented is correct. Plausible because ES-3.2 is a cooldown ES-3.1 procedure that could be used and the transition to E-2 is correct.

D. Incorrect, ES-3.2 is not the procedure to be entered to provide for the least spread of contamination and the minimum offsite dose (ES-3.1 is) and the transition required if #1 steam generator pressure starts dropping in an uncontrolled manner after ES-3.1 is implemented is not to ECA-3.1 (it is to E-2). Plausible because

( ES-3.2 is a cooldown procedure that could be used and ECA-3.1 ECA-3. 1 is the procedure directed to be entered from the Fold Out Page for other conditions that would be present if the pressure on the ruptured steam generator had been dropping.

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 97

(

Tier 3 KIA G 2.3.11 Ability to control radiation releases.

Importance Rating: 3.8 / 4.3 3.8/4.3 Technical

Reference:

E-3, Steam Generator Tube Rupture, Rev. 17 ES-3.1, Post-SGTR Cooldown Using Backfill, Rev. 10 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 ES-3.1 B.1 & & .5 State the purpose/goal of this ES-3.1.

Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

Question Source:

Bank# _ _ __

Modified Bank # _-

X- _- _

New - _ __

Question History: Question modified from Byron Plant question Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X _ ___

10 CFR Part 55 Content: ( 41.11 /43.4 / 45.10 )

(41.11/43.4/45.10) 10CFR55.43.b (5)

Comments: Question modified from Byron Plant question

(

QUESTIONS REPORT for 2009 SRO Retake Exam G 2.3.14 098

( With Unit 1 in Mode 1, compliance with Technical Specification 3.4.8, Reactor

, Coolant System Specific Activity ensures that the 2-hour dose at the site boundary will not exceed a small fraction of (1) limits following a SGTR in conjunction with an assumed steady state steam generator tube leakage rate of _ .......(=2)______

_->.(=2)L...-_

ill ~

ill A. 10CFR20 1.0 gpm total B. 10CFR20 0.1 gpm per SG C~

C¥' 10CFR100 1.0 gpm total D. 10CFR100 0.1 gpm per SG DISTRACTOR DIS TRACTOR ANAL YSIS:

A. Incorrect, the limit is based on not exceeding a fraction of the Part 100 limits, not the Part 20 limits but the assumed steady state primary-to-secondary steam generator leakage rate is 1.0 gpm per steam generator. Plausible because Part 20 identifies radiological limits and the assumed steady state primary-to-secondary steam generator leakage rate

( is 1.0 gpm as identified in Technical Specification 3/4.4.8 bases.

B. Incorrect, the limits is based on not exceeding a fraction of the Part 100 limit, not the Part 20 limits and the assumed steady state primary-to-secondary steam generator leakage rate is 1.0 gpm, not 0.1 gpm per steam generator. Plausible because Part 20 identifies radiological limits and the O. 1 gpm leakage per steam generator is the normal operational leakage limit accounted for the safety limits as identified in the bases for Technical Specification 3/4.4.6.2, RCS Operational Leakage.

C. CORRECT, the limit being based on not exceeding a fraction of the Part 100 limits and the assumed steady state primary-to-secondary steam generator leakage rate being 1.0 gpm is identified in Technical Specification 3/4.4.8 bases.

D. Incorrect, the limit is based on not exceeding a fraction of the Part 100 limits, however, the assumed steady state primary-to-secondary steam generator leakage rate is 1.0 gpm, not O. 1 gpm per steam generator. Plausible because not exceeding a fraction of the Part 100 limits is correct and the O. 0.11 gpm leakage per steam generator is the normal operational leakage limit accounted for the safety limits as identified in the bases for Technical Specification 3/4.4.6.2, RCS Operational Leakage.

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 98 Tier 3 KIA KiA GG2.3.14 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Importance Rating: 3.4 /3.8 3.41 3.8 Technical

Reference:

Tech Spec 3/4.4.8 Basis, Amendment No. 322 Proposed references to be provided to applicants during examination: None Learning Objective: OPT200.RCS B.6.a Describe the administrative controls and limits for the RCS as explained in this lesson:

a. State the Tech SpeclTRM LCOS that govern the RCS.

Question Source:

Bank#_X Bank# X._ __

Modified Bank # _ _ __

New _ __

( Question History: AOP-R.06-8.9 001 SQN bank question RCS-B.6 003 and AOP-R.06-B.9 Question Cognitive Level:

Memory or fundamental knowledge _X__

Comprehension or Analysis _ __

10 CFR Part 55 Content: /43.4 / 45.10 )

( 41.12 143.4/45.10 10CFR55.43.b ( 2,4) Requires the candidate to know the Facility operating limitations in the technical specifications and their bases and be knowledgeable of the radiation hazards that may arise during normal, abnormal and emergency conditions.

Comments: SQN bank questions RCS-B.6 003 and AOP-R.06-B.9 001 with some modification in wording

QUESTIONS REPORT for 2009 SRO Retake Exam G 2.4.16 099 Given the following:

- A Station Blackout has occurred.

Unit 1 is performing ECA-O.O, "Loss of All AC Power." and have depressurized the SGs.

RCS Subcooling is Oaf.

O°F.

1A-A Emergency Diesel Generator was started and is supplying its associated bus.

Subsequently,

- The crew has reached the last step of ECA-O.O and is preparing to transition to the appropriate recovery procedure.

- A RED Path exists on the Heat Sink CSF Status Tree.

Which ONE of the following identifies the required recovery strategy?

A. Transition to ECA-0.1, "Loss of All AC Power Recovery Without SI Required",

and enter FR-H.1, "Response to Loss of Secondary Heat Sink" when allowed by ECA-0.1.

B~ Transition to ECA-0.2, Loss of All AC Power Recovery With SI Required",

B:I and enter FR-H.1, "Response to Loss of Secondary Heat Sink" when allowed

( by ECA-0.2.

C. Transition to FR-H.1, "Response to Loss of Secondary Heat Sink" upon exit from ECA-O.O. Perform ECA-0.1, "Loss of All AC Power Recovery Without SI Required", when FR-H.1 is complete.

D. Transition to FR-H.1, "Response to Loss of Secondary Heat Sink" upon exit from ECA-O.O. Perform ECA-0.2, "Loss of All AC Power Recovery With SI Required", when FR-H.1 is complete.

QUESTIONS REPORT for 2009 SRO Retake Exam DIS TRA C TOR ANAL YSIS:

DISTRACTOR

(

A. Incorrect, ECA-G.1 ECA-O.1 s not correct because ECCS flow is required with the subcoofing subcooling identified but not entering FR-H. 1 until identified in procedure would be correct. Plausible because with more subcoofing sub cooling the transition to ECA-G.

ECA-O. 1 would be correct and the resumption of Status Tree implementation when addressed by procedure is correct.

B. CORRECT, ECCS flow is required because subcoofingsubcooling is insufficient, thus ECA-G.2 ECA-O.2 is the correct transition from ECA-G.

ECA-O. G.O. Status tree are monitored for information only until power is restored because the FRPs are written assuming a Train of shutdown power is available. The power will be restored in ECA-G.2 ECA-O.2 and a step will address the resumption of Status Tree implementation.

C. Incorrect, CSFST are only monitored but not addressed until allowed in appropriate recovery procedure. Performance of ECA-O.1ECA-G. 1 is also incorrect due to subcooling. Plausible because a Status Tree RED path is normally the lack of subcoofing.

implemented immediately and ECA-G.ECA-O. 1 would be the correct recovery procedure if subcoofing subcooling had been higher.

D. Incorrect, CSFST are only monitored but not addressed until allowed in appropriate recovery procedure. Performance of ECA-G.2ECA-O.2 is incorrect due to the lack of subcooling but it would be implemented prior to performance FR-H. 1.

Plausible because a Status Tree RED path is normally implemented immediately and ECA-O.2 ECA-G.2 is the correct recovery procedure.

((

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 99

(

Tier 3 KIA G2.4.16 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.

Importance Rating: 3.5 / 4.4 Technical

Reference:

ECA-O.O, Loss of All AC Power, Rev 22 ECA-0.2, Recovery From loss of All AC Power With SI Required, Rev 9 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 ECA-O.O B.3 Summarize the mitigating strategy for the failure that initiated entry into ECA-O.O.

OPL271 ECA-0.2 8.3 B.3 Summarize the mitigating strategy for the failure that initiated entry into ECA-0.2.

Question Source:

( Bank#_X_ __

Modified Bank # _ _ __

New _ __

Question History: SQN Bank question written for 2007 Audit exam.

Question Cognitive Level:

Memory or fundamental knowledge _ __

Comprehension or Analysis _X _ ___

10 CFR Part 55 Content: ( 41.10/43.5/45.13 )

(41.10/43.5/45.13 10CFR55.43.b (5)

Comments: Minor format and wording changes in stem and choices.

Relocated correct answer

(

QUESTIONS REPORT for 2009 SRO Retake Exam G 2.4.38 100 In accordance with the Emergency Plan Implementing Procedures (EPIPs),

which ONE of the following identifies a function of the Site Emergency Director (SED) that can be delegated and who the SED can delegate to perform the function?

Function to be delegated Can be delegated to ...

A. Emergency Doses that exceed CECC director occupational dose limits B. Emergency Doses that exceed Site Vice President occupational dose limits Ct Protective Action CECC director Recommendations D. Protective Action Site Vice President Recommendations DIS TRA CTOR ANAL YSIS:

A. Incorrect, the SED can delegate Protective Action Recommendations (PARs) to the CECC Director but not the authorization of emergency dose limits. Plausible

( because the authorization of emergency dose limits is an SED function and because Protective Action Recommendations (PARs) can be delegated to the CECC director.

B. Incorrect, the SED can delegate Protective Action Recommendations (PARs) to the CECC Director but not the authorization of emergency dose limits. Plausible because the authorization of emergency dose limits is an SED function and because the Site VP is a position in the Technical support Center (TSC) and normally is the highest ranking person on the site.

C. CORRECT, In accordance with EPIP-6, the SED makes Protective Action Recommendations (PARs) to the state and the responsibility cannot be delegated except to the CECC director.

D. Incorrect, the SED can delegate Protective Action Recommendations (PARs) but not to anyone except the CECC director. Plausible because the Site VP is a position in the Technical support Center (TSC) and normally is the highest ranking person on the site.

(

\

QUESTIONS REPORT for 2009 SRO Retake Exam Question No. 100 Tier 3 KIA G 2.4.38 Emergency Procedures IPlan /Plan Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

Importance Rating: 2.4 14.4

/ 4.4 Technical

Reference:

EPIP-6, Activation and support of the Technical Support Center, Rev 44 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271.REP B.5.a State the duties and responsibilities of the Site Emergency Director (SED).

a. State the duties and the responsibilities the SED cannot delegate.

Question Source:

Bank#

Bank # - __ -_ -_-

Modified Bank # X' _-_ -_-

( New _ __

Question History: SQN question EPIP-15 003 modified Question Cognitive Level:

Memory or fundamental knowledge _X __

Comprehension or Analysis _ __

10 CFR Part 55 Content Content: ( 41.10 I 43.5 I 45.11 )

41.10/43.5/45.11 10CFR55.43.b (4)

Comments: SQN question EPIP-15 003 modified and a made 2 part question.

Changed to ask for a function that could be delegated versus a function that could not be delegated and the identification of who could be delegated to perform the function.

Correct answer in different location.

BANK INFORMATION REPORT for 2009 SRO Retake Exam

<Title MCS 25 Multiple choice: single BANK 6 BANK MOD 9 NEW 10 10 BYRON 1 COMMANCHEPEAK COMMANCHE PEAK 1 GINNA 1 SQN 9 WBN 3

~UK~6 ~"'(G;ri' nitt",il

'11~ . ~<:J1\,< ";j\y:K'!0'C'XA;'

HIGHER 17 LOWER 8 25

.. <. '\0.,<<<

fC,~t,g.Q.:'

SRO 25 SEQUOYAH 25 112009 RETAKE 25 NO 25 Saturday, July 11, 20092:42:42 2009 2:42:42 PM 1

ML092881149

Contents

Text

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Navigation menu

ML092881149

Text

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Navigation menu

Search