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See also: [[see also::IR 05000327/2009301]]


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{{#Wiki_filter:( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
{{#Wiki_filter:}}
76. 008 AG2.4.2 076 Given the following:
Unit 1 at 100% power when a pressurizer
safety valve failed open. The operator manually tripped the reactor and initiated
a safety injection.
While performing
the step to determine
if the RHR spray should be placed in service in accordance
with E-1, "Reactor Trip or Safety Injection", the crew determines
the following:
When pressurizer
pressure dropped to 1280 psig, the safety valve reclosed and pressurizer
pressure started to rise. Containment
pressure rose to 2.6 psig, and began trending down. Pressurizer
level is 100%. RCS subcooling
is 43°F. -All four SG levels at 33% narrow range. Which ONE of the following
identifies
the correct procedure
implementation
and operation
of the RCPs for the above conditions?
A'I Transition
from E-1 to ES-1.1, SI Termination;
The RCPs will have remained running throughout
the event. B. Transition
from E-1 to ES-1.1, SI Termination;
The RCPs would have been shutdown but will be restarted
in ES-1.1, SI Termination.
C. Continue E-1 until a transition
is directed to ES-1.2, Post LOCA Cooldown;
The RCPs will have remained running throughout
the event. D. A transition
will be made to ES-1.2, Post LOCA Cooldown;
Page 1 The RCPs would have been shutdown but will be restarted
in ES-1.2, Post LOCA Cooldown. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRA C TOR ANAL YSIS: Page 2 A. CORRECT, with the safety valve reclosed and the conditions
as identified
in the stem, SI termination
criteria is met. While the crew would be beyond the step in E-1 that first checks for SI termination
and beyond the followup step for checking the criteria, the SI termination
step is a continous
action step and if the criteria is met the transition
is to be made. Subcooling
is greater than he 400F setpoint, pressurizer
level above the 10% setpoint, heat sink is established
and RCS pressure rising meet the entry conditions
for ES-1.1. Containment
pressure did not rise to the automatic
initiation
setpoint of 2.8 psig (Phase B) nor did the RCS pressure drop to the 1250 psig setpoint, so the RCP trip criteria was not met and the pumps remained in service. B. Incorrect, With the conditions
identified
in the stem, the SI termination
criteria is met and a transition
to ES-1.1 is required.
The RCP trip criteria was not met and the pumps would have remained in service throughout
the even. Plausible
because the transition
to ES-1.1 is the correct transition
and if the RCPs had been stopped they would be restarted
in ES-1.1. C. Incorrect, While ES-1.2 would be entered if E-1 was continued, the conditions
identified
in the stem indicate SI termination
criteria is met and a transition
to ES-1.1 is required.
The RCP trip criteria was not met and the pumps would have remained in service through out the event. Plausible
because the transition
to ES-1.2 would be the correct transition
if the SI could not be terminated
and the RCPs remaining
in service through out the event is correct. D. Incorrect, While ES-1.2 would be entered if E-1 was continued, the conditions
identified
in the stem indicate SI termination
criteria is met and a transition
to ES-1. 1 is required.
Because the RCP trip criteria was not met, the pumps would have remained in service through out the event. Plausible
because the transition
to ES-1.2 would be the correct transition
if the SI could not be terminated
and if the RCPs had been stopped they would be restarted
in ES-1.2. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 76 Tier 1 Group 1 KIA 008 AG2.4.2 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) Knowledge
of system set points, interlocks
and automatic
actions associated
with EOP entry conditions.
Importance
Rating: 4.5 1 4.6 Technical
Reference:
E-1, Loss of Reactor Or Secondary
Coolant, Rev 23 ES-1.1, SI Termination, Rev 10 ES-1.2, Post LOCA Cooldown, Rev 17 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 E-1 B.5 Question Source: Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
Bank# ----Modified Bank # ----New X ---Question History: New question for SQN 1/2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X __ 10 CFR Part 55 Content: (41.7/45.7/45.8 ) 10CFR55.43.b ( 5 ) Comments:
New question for SQN 1/2009 exam Page 3 
. ' . .... "'. SQN LOSS OF REACTOR OR SECONDARY
COOLANT FOLDOUT PAGE RCP TRIP CRITERIA IF any of the following
conditions
occurs: * RCS pressure less than 1250 psig AND at least one CCP or SI pump running OR * Phase B isolation, THEN STOP all RCPs. SI REINITIATION
CRITERIA IF any of the following
conditions
occurs: * RCS subcooling
based on core exit TICs less than 40°F OR * Pressurizer
level CANNOT be maintained
greater than 10% [20% ADV], THEN RAISE ECCS flow by performing
one or both of the following
as necessary:
* . ESTABLISH
CCPIT flow USING Appendix C * START CCPs or SI pumps manually.
EVENT DIAGNOSTICS
* IF both trains of shutdown boards de-energized, THEN GO TO ECA-O.O, Loss of All AC Power. * IF any SIG pressure dropping in an uncontrolled
manner or less than 140 psig AND SIG NOT isolated, THEN GO TO E-2, Faulted Steam Generator
Isolation.
* IF any S/G has level rising in uncontrolled
manner or has abnormal radiation, THEN: a. RAISE ECCS flow by performing
one or both of the following
as necessary:
* ESTABLISH
CCPIT flow USING Appendix C * START CCPs or SI pumps manually.
b. GO TO E-3, Steam Generator
Tube Rupture. TANK SWITCHOVER
SETPOINTS
* IF CST level less than 5%, THEN ALIGN AFW suction to ERCW. * IF RWST level less than 27%, THEN GO TO ES-1.3, Transfer to RHR Containment
Sump. Page 1a of 26 E-1 Rev. 23 
( SQN LOSS OF REACTOR OR SECONDARY
COOLANT E .. 1 Rev. 23 I STEP II ACTION/EXPECTED
RESPONSE I I RESPONSE NOT OBTAINED 7. MONITOR SI termination
criteria:
a. RCS subcooling
based on core exit TICs greater than 40°F. b. Secondary
heat sink: * Narrow range level in at least one Intact S/G greater than 10% [25% ADV]. OR * Total feed flow to Intact SIGs greater than 440 gpm. c. RCS pressure STABLE or RISING. d. Pressurizer
level greater than 10% [20% ADV]. e. GO TO ES-1.1, SI Termination.
---.----a. GO TO StepS. b. GO TO Step S. c. GO TO Step S. L lKJ N 01 -fff),IJ!
t'{1 ::a-.. d. ATTEMPT to stabilize
RCS pressure with normal pressurizer
spray. GO TO Step S. Page 10 of 26 
( OPL271 E-1 Revision 2 Page 3 of 85 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: E-1, "Loss of Reactor or Secondary
Coolant" IV. LENGTH OF LESSON/COURSE:
2 hours V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of E-1, "Loss of Reactor or Secondary
Coolant. B. Enabling Objectives
O. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with E-1, "Loss of Reactor or Secondary
Coolant that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of E-1 . 2. Discuss the E-1 entry conditions.
3. Summarize
the mitigating
strategy for the failure that initiated
entry into E-1. 4. Describe the bases for all limits, notes, cautions, and steps of E-1. 5. Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
6. Given a set of initial plant conditions
use E-1 to correctly:
a. Identify required actions b. Respond to Contingencies
c. Observe and Interpret
Cautions and Notes 7. Apply GFE and system response concepts to the performance
of E-1 conditions. 
( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
77. 009 EA2.37 077 Given the following:
-Unit 1 is operating
at 100% power when a loss of offsite power occurs. -The operators
subsequently
initiate Safety Injection
due to a small break LOCA. -Thirty minutes after the Safety Injection, the following
conditions
exist: -E-1, "Loss of Reactor or Secondary
Coolant" is being performed.
-All 4 SG pressures
are approximately
1010 psig and stable. -RCS pressure is 2230 psig and stable. -Thot is approximately
575°F in all 4 loops and lowering slowly. -Core Exit TCs indicate approximately
580°F and slowly rising. -T cold is approximately
560°F in all 4 loops and stable. Based on the above indications, which ONE of the following
identifies
the condition
of the RCS and the procedure
transition
that will be made? A. Natural Circulation
exists and a transition
will be directed to ES-O.2, Natural Circulation
Cooldown as the E-1 procedure
is continued.
B. Natural Circulation
exists and a transition
will be directed to ES-1.2, Post LOCA Cooldown as the E-1 procedure
is continued.
C. Natural Circulation
does NOT exist and a transition
will be directed to ES-0.2, Natural Circulation
Cooldown as the E-1 procedure
is continued. Natural Circulation
does NOT exist and a transition
will be directed to ES-1.2, Post LOCA Cooldown as the E-1 procedure
is continued.
Page 4 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted ( DIS TRA CTOR ANAL YSIS: Page 5 A. Incorrect, Natural Circulation
does not exist. Tcold (5600F) is approximately
10 degrees higher than saturation
temperature
of all 4 SGs (-54ffJF).
Although SG pressures
are at approximately
the Atmospheric
dump valve pressure, they mayor may not be open. The only way to tell if natural circulation
exists is by trending Tcold. The transition
to ES-O.2 from E-1 is not correct. Plausible
because Thot lowering and Tcold stable could exist with natural circulation
and ES-0.2 is a natural circulation
procedure.
B. Incorrect, Natural circulation
does not exist. Tcold (5600F) is approximately
10 degrees higher than saturation
temperature
of all 4 SGs (-54ffJF).
The only way to tell if natural circulation
exists is by trending Tcold. Steam dumps are unavailable
due to the loss of off site power. The transition
to ES-1.2 from E-1 is correct. Plausible
because Thot lowering and Tcold stable could exist with natural circulation
and ES-1.2 is the correct procedure.
C. Incorrect, Natural circulation
does not exist. Tcold (5600F) is approximately
10 degrees higher than saturation
temperature
of all 4 SGs (-54ffJF).
The only way to tell if natural circulation
exists is by trending Tcold. Steam dumps are unavailable
due to the loss of offsite power. The transition
to ES-0.2 from E-1 is not correct. Plausible
because natural circulation
does not exist (Core Exit temperatures
rising) and ES-0.2 is a natural circulation
procedure.
D. CORRECT, Natural Circulation
does not exist. Tcold (5600F) is approximately
10 degrees higher than saturation
temperature
of all 4 SGs (-54ffJF).
Although SG pressures
are at approximately
the Atmospheric
dump valve pressure, they mayor may not be open. The only way to tell if natural circulation
exists is by trending Tcold. The transition
to ES-1-1 is the correct transition
from E-1 for the conditions
stated. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 77 Tier 1 Group 1 KIA 009 EA2.37 Ability to determine
or interpret
the following
as they apply to a small break LOCA: Existence
of adequate natural circulation
Importance
Rating: 4.2 I 4.5 Technical
Reference:
E-1, Loss of Reactor or Secondary
Coolant, Rev 23 Steam Tables EA-68-6, Monitoring
NAtural Circulation
Conditions, Rev 0 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 E-1 B.5 Question Source: Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
Bank# ___ _ Modified Bank # X Question History: ---New ---SQN Bank question ES-O.2-B.6
002 modified to add SRO component
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:
SQN Bank question ES-0.2-B.6
002 modified to add SRO component
SQN Question modified from Salem Unit 1 NRC exam dtd 11/04/2002
Page 6 
SQN LOSS OF REACTOR OR SECONDARY
COOLANT E-1 Rev. 23 I STEP I I ACTION/EXPECTED
RESPONSE 15. c. MONITOR containment
sump level less than 68%. d. NOTIFY TSC to initiate post-accident
sampling as necessary.
e. EVALUATE plant equipment
status USING EA-0-4, Evaluation
of Equipment
Status. 16. DETERMINE
ifRCS cooldown and depressurization
is required:
a. CHECKRCS pressure greater than 300 psig. b. GO TO ES-1.2, Post LOGA Cooldown and Depressurization.
--.. 11----I I RESPONSE NOT OBTAINED c. NOTIFY TSG to evaluate containment
sump level and actions of FR-Z.2, Containment
Flooding.
a. IF RHR injection
flow greater than 1000 gpm, THEN GO TO Step 17. Page 19 of 26 
SQN MONITORING
NATURAL CIRCULATION
CONDITIONS
1,2 4.2 Verification
of Natural Circulation
2. DETERMINE
parameter
trends between monitoring
intervals
and EVALUATE if natural circulation
is occurring.
3. IF natural circulation
NOT verified, THEN NOTIFY ASOS. 4. GO TO Section 4.1, step in effect. END OF TEXT EA-68-6 Rev. 0 Page 4 of 5 D D D 
ES-O.2-B.6
002 Given the following:
QUESTIONS
REPORT for BANK SQN Questions
-Unit 2 is operating
at 100% RTP when a Loss of Off-Site power causes a reactor trip. Ten minutes after the trip, the following
conditions
exist: -SG #1 Pressure1
01 0 psig and stable SG #2 Pressure1005
psig and stable SG #3 Pressure 1015 psig and stable SG #4 Pressure1
01 0 psig and stable RCS Pressure is 2230 psig and stable Thot is approximately
575 of in all 4 loops and lowering slowly Core Exit TCs indicate approximately
580 OF Tcold is approximately
560 OF in all 4 loops and stable Based on the above indications, what is the condition
of the RCS? A. Natural Circulation
exists. S/G PORVs are maintaining
heat removal. B. Natural Circulation
exists. The steam dumps are maintaining
heat removal. C. Natural Circulation
does NOT exist. Heat removal may be established
by opening the steam dumps. Natural Circulation
does NOT exist. Heat removal may be established
by opening the S/G PORVs. Monday, November 10, 2008 8:14:12 AM 1 
OPL271E-1
Revision 2 Page 3 of 85 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: E-1, "Loss of Reactor or Secondary
Coolant" IV. LENGTH OF LESSON/COURSE:
2 hours V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of E-1, "Loss of Reactor or Secondary
Coolant. B. Enabling Objectives
o. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with E-1, "Loss of Reactor or Secondary
Coolant that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of E-1 . 2. Discuss the E-1 entry conditions.
3. Summarize
the mitigating
strategy for the failure that initiated
entry into E-1. 4. Describe the bases for all limits, notes, cautions, and steps of E-1 . 5. Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
6. Given a set of initial plant conditions
use E-1 to correctly:
a. Identify required actions b. Respond to Contingencies
c. Observe and Interpret
Cautions and Notes 7. Apply GFE and system response concepts to the performance
of E-1 conditions. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
78. 011 EA2.14 078 Given the following:
Unit 1 is at 100% power. RHR Pump 1 B-8 is tagged for motor bearing maintenance. -A Safety Injection
occurs due to a LOCA. RHR Pump 1 A-A trips on instantaneous
overcurrent
when it attempts to start. -As the crew transitions
from E-1, Loss of Reactor or Secondary
Coolant, to ECA-1.1, "Loss of RHR Sump Recirculation", the STA reports a RED path to FR-P.1, "Pressurized
Thermal Shock". RCS pressure is currently
230 psig. RWST level is 54% and dropping.
Which ONE of the following
describes
the required operator action? A. The transition
to FR-P.1 should NOT be made due to the ECA-1.1 entry requirement.
B. The transition
to FR-P.1 should NOT be made until transfer to the containment
sump is accomplished.
C. The transition
to FR-P.1 should be made but a transition
back to ECA-1.1 will be directed if SI termination
criteria is NOT met in FR-P.1. The transition
to FR-P.1 should be made but a transition
back to ECA-1.1 will be directed with no RHR pump running. Page 7 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DISTRAGTOR
ANAL YSIS: Page 8 A. Incorrect, The transition
to FR-P.1 is required to be made from EGA-1.1. Plausible
because some other EGAs suspend the implementation
of Status trees
or direct the transition
to FRGs not be made during the EGA performance.
B. Incorrect, The transition
to FR-P.1 is required to be made from EGA-1.1. Plausible
because some other EGAs suspend the implementation
of Status trees
or direct the transition
to FRGs not be made during the EGA performance
until certain conditions
are met and the establishment
of sump recirculation
is a priority for the given conditions.
G. Incorrect, The transition
to FR-P.1 is required to be made and the transition
back to EGA-1.1 is required but not because to the status of SI termination
criteria.
Plausible
because FR-P.1, Step 3 does direct a return to the procedure
in effect prior to reaching the SI Termination
check which would also direct the same transition
back to the instruction
in effect. D. GORREGT, The transition
to FR-P.1 is required to be made and when FR-P.1 is entered, Step 3 will direct a return to the procedure
in effect due to the RGS pressure being less than 300 psig along with both RHR pumps being stopped and sump recirculation
capability
loss. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 78 Tier 1 Group 1 KIA 011 EA2.14 Ability to determine
or interpret
the following
as they apply to a Large Break LOCA: Actions to be taken if limits for PTS are violated Importance
Rating: 3.6* 14.0 Technical
Reference:
ECA-1.1, Loss of RHR Sump Recirculation,Rev
11 FR-P.1, Pressurized
Thermal Shock, Rev 13 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 FR-P.1 B.5 Question Source: Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
Bank# ____ _ Modified Bank # ___ _ New --'X'-'--__ Question History: New question for SQN 1/2009 exam 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:
New question for SQN 1/2009 exam Page 9 
I SQN I PRESSURIZED
THERMAL SHOCK FR-P.1 Rev. 13 I STEP II ACTION/EXPECTED
RESPONSE II RESPONSE NOT OBTAINED 1. MONITOR RWST level greater than 27 %. 2. MONITOR CST level greater than 5%. 3. CHECK RCS pressure greater than 300 psig. IF RHR pumps aligned to RWST, THEN GO TO ES-1.3, Transfer to RHR Containment
Sump. ALIGN AFW suction to ERCW USING EA-3-9, Establishing
Turbine Driven AFW Flow, and EA-3-10, Establishing
Motor Driven AFW Flow. IF, any of the following
conditions
exist: * RHR injection
flow greater than 1000gpm OR * both RHR pumps STOPPED AND sump recire capability
has been lost THEN RETURN TO procedure
and step in effect. Page 3 of 25 
OPL271 FR-P.1 Revision 1 Page 3 of 16 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: FR-P.1, PRESSURIZED
THERMAL SHOCK IV. LENGTH OF LESSON/COURSE:
1 hours V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of the FR-P.1, Pressurized
Thermal Shock. B. Enabling Objectives
o. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
FR-P.1, Pressurized
Thermal Shock, that are rated;::::
2.5 during Initial License Training and;:::: 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of FR-P .1. 2. Discuss the FR-P.1 entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with FR-P.1 entry conditions.
b. Describe the requirements
associated
with FR-P.1 entry conditions.
3. Summarize
the mitigating
strategy for the failure that initiated
entry into FR-P.1. 4. Describe the bases for all limits, notes, cautions, and steps of FR-P.1. 5. Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
6. Given a set of initial plant conditions
use FR-P.1 to correctly:
a. Identify required actions
b. Respond to Contingencies
c. Observe and Interpret
Cautions and Notes 7. Apply GFE and system response concepts to the performance
of FR-P.1 conditions. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
79. 015 AA2.01 079 Given the following:
Unit 1 is at 32% power. -The following
annunciator
windows alarm on 1-M-6: FS-68-48A
REACTOR COOLANT LOOP 3 LOW FLOW RCP BUS UNDERFREQUENCY
1 UNDERVOL TAGE RCP #3 indications
are: -Ammeter -'0' amps. -Green and white lights above the handswitch
are lit. Flow -'0' on all 3 indicators.
-The other RCPs remain in service and the reactor trip breakers remain closed. -The OATC manually trips the reactor as directed by the SRO. Which ONE of the following
identifies
the condition
causing the RCP trip and the earliest NRC notification
required by 1 OCFR50. 72? Condition
causing RCP Trip Notification
Requirement
A. Under voltage 1 hour Under voltage 4 hours C. Under frequency
1 hour D. Under frequency
4 hours Page 10 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRACTOR ANAL YSIS: Page 11 The RCPs have protection
from undervoltage
and underfrequency
via relays. Both conditions
can cause trips of the RCPs and if conditions
and logic is met, the reactor would automatically
trip. A. Incorrect, The conditions
in the stem describe the trip of RCP #3 from an undervoltage
condition
on the 6.9 Kv Unit Board feeding the RCP, however NO automatic
trip would be initiated
at the stated power level (below P8) and the emergency
plan would not be implemented. (No 1 hour notification
would be required).
Plausible
because the condition
causing the RCP trip is correct and if the power level had been greater than P8, an automatic
reactor trip should have occurred .. If the reactor had failed to trip automatically, a 1 hour report would be required due to implementation
of the Emergency
Plan. B. CORRECT, The conditions
in the stem describe the trip of RCP #3 from an undervoltage
condition
on the 6.9 Kv Unit Board feeding the RCP. With the reactor power level below P-8, no automatic
reactor trip would be initiated
from low flow condition
in a single loop. AOP-R.04, Malfunction
of Reactor Coolant Pump, would be the controlling
procedure
and at the stated power level, a manual reactor trip would be directed.
The reactor trip would require a 4 hour notification.
C. Incorrect, The conditions
in the stem describe the a trip of RCP #3 from an undervoltage
condition
on the 6.9 Kv Unit Board feeding the RCP not an underfrequency
trip. The underfrequency
trip of the RCP requires 214 logic to be made in the SSPS which then will trip the reactor and all RCPs. The logic is not met in the stem even though power is above the P7 permissive
(10%), only one RCPs is involved and the logic requires 2 out of 4. NO automatic
trip would be initiated
at the stated power level (below P8) and the emergency
plan would not be implemented. (No 1 hour notification
would be required).
Plausible
because underfrequency
can cause RCP trip with conditions
different
than stated in the stem and following
the underfrequency
initiation
at greater than P7, a failure of the reactor to automatically
trip would require a 1 hour due to the implementation
of the Emergency
Plan. D. Incorrect, The conditions
in the stem describe the trip of RCP #3 from an undervoltage
condition
on the 6.9 Kv Unit Board feeding the RCP not an underfrequency
trip. The underfrequency
trip of the RCP requires 214 logic to be made in the SSPS which then will trip the reactor and all RCPs. The logic is not met in the stem even though power is above the P7 permissive
(10%), only one RCPs is involved and the logic requires 2 out of 4. the 4 hour notification
due to the manual reactor trip is correct. Plausible
because underfrequency
can cause RCP trip with conditions
different
than stated in the stem and the 4 hour notification
due to the manual reactor trip is correct. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 79 Tier 1 Group 1 KIA 015 AA2.01 Ability to determine
and interpret
the following
as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Cause of RCP failure Importance
Rating: 3.0/3.5* Technical
Reference:
1-AR-M6-A
Reactor Protection
and Safeguards
1-XA-55-6A, Rev 15. 1,2-47N763-2
R20 1-47W611-99-6
R2 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPT200.RCP
S.4.i. & 5.c Question Source: Describe the following
items for each major component
in the Reactor Coolant Pump system as described
in this lesson: i. Protective
features (including
setpoints)
Describe the operation
of the RCP system as it relates to the following:
c. Alarms and Alarm Response Sank# -----Modified Sank # ----New _X __ Question History: New question for SQN 1/2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X __ 10 CFR Part 55 Content: ( 43.5 1 45.13 ) 10CFR55.43.b ( 5 ) Comments:
New question for SQN 1/2009 exam Page 12 
( 32 (E-4) Source Setpoint SER 460,467,476,477,478,482,1613, 1614 57Hz and -72% voltage (-5022 volts) Rep BUS UNDERFREQUENCYI
UNDERVOLTAGE
UV and UF relays OneUVand UF Relay on each Rep bus. 1/4 for alarm 2/4 for trip input Probable Causes NOTE 1 NOTE 2 Corrective
Actions References
1. Loss of power to 6.9kVunit
bds. 2. Channel malfunction
or testing. 3. RCP Undervoltage
or Underfrequency
relay failure. Underfrequency
condition
on 2/4 RCPs will trip all four RCPs automatically.
Undervoltage
condition
on individual
Rep buss will block operation
of associated
UF relay. [1] IF no reactor trip occurs, THEN [a] CHECK 1.,.XX-55-6A
Reactor Trip 81 status panel for bistables
that may be tripped. [b] EVALUATE reactor trip criteria with SRO per AOP-R.04, Reactor Coolant Pump Malfunctions.
[c] IF reactor trip should have occurred automatically, ntEN GO TO E-O, Reactor Trip or Safety Injection.
[2] IF RCP trips, THEN GO TO AOP-R.04, Reactor Coo/ant Pump Malfunctions, [3] IF RCP undervoltage
or underfrequency
relay has failed, THEN GO TO AOP-I.10, RCP Undervoltage
or underfrequency
Instrument
Malfunction.
[4] EVALUA TE Technical
Specifications, 45B655-06A-O, 47m10-68, 47W811-99, 45N721-1,45N763-2
SQN Page 37 of 40 1 1-AR-MS-A
Rev. 16 
Source SER46S SER 583 SER592 SER 596 1-Fs.6S-'4BA
.1-FS-68-48S
1-FS-68-48D
1/3 for alann 2J3 for trip input Probable Causes Corrective
Actions References
18 (C-4) Setpoint FS-68-48A
REACTOR COOLANT LOOP 3 LOW FLOW Flow < 90% 1. Pump trip. 2. Flow instrument
malfunction
or testing. 3. Operator stopping pump. [1] IF no reactor trip occurs, THEN [a] CHECK 1-XX-55-6A
Reactor Trip $1 status panel for blstables
that may be tripped. [b) EVALUATE reactor trip criteria with SRO per AOP-R.04, Reactor Coo/ant Pump Malfunctions.
[e] IF reactor trip should have occurred automatically, THEN TRIP the reactor, AND GO TO E-O, Reactor Trfp or Safety Injection.
[d] CHECK ri-FI-68-48Al, ri-FI-68-48BJ, and r1-FI-68-48D]
for channel failure. [e] IF a single flow channel has failed, THEN GO TO AOP-1.03, ReS Flow Instrument
Malfunction.
[2] IF Rep trips, THEN GO TO AOP-R.04, Reactor Coolant Pump Malfunctions. . [3] EVALUATE Technical
Specifications.
45B655-06A-O, 47'M510-68, 47W611-99
I 1 SQN I i-AR-MS-A
Page 23 of 40 Rev. is 
Trip Logic 1 Breaker Trip I----------
l HS "A" in STOP -:P
Transfer Switch in NORMAL ---I I i HS 'C' in STOP
I i I
Breaker TriPi ___ J Transfer Switch ---. inAUX I I I I 1 6.9KV Bus 1 ____ _J I I RCP Mota, DC jn-----[ ! I RCP BUS UF r.D--J (Train A) RCP BUS UF ---(Train B) E04 x. LESSON BODY: J. Purpose/function
of components
COMPONENT
PURPOSE / FUNCTION RCP Trip logic * A RCP may be stopped from its handswitch
or at the switchgear.
The (SD page 33) breaker will trip open on a motor protection
fault (unit board UV, motor overcurrent)
or an underfrequency
trip of either reactor protection
system train. * In addition to providing
a trip to the RCP, an UV trip signal provides an input to the reactor protection
system (RPS). The purpose of the RPS trip is to provide DNB protection
to the reactor by anticipating
a complete loss of flow condition.
* This reactor trip requires an undervoltage
condition
on two RCPs and is blocked below P-7 (10%). OPT200.RCP
Rev.5 Page 28 of 56 
RCP Bus 1A UF -.!r-'\ Test Switch
in Test RCP Bus 1B UF -Test Switch in Test RCP Bus 1C UF :b J Test Switch in Test RCP Bus 10 UF --'!r-'\ Test Switch
in Test Power> P7 _____ _ Under-Frequency
Logic RCP Bus UF (Train A) E04 X. LESSON BODY: J. Purpose/function
of components
COMPONENT
PURPOSE / FUNCTION RCPUF logic * An UF condition
on two of the four RCP buses supplying
the RCPs (SDpage 34) will initiate a reactor trip and trip all four RCP breakers.
The purpose of the UF trip is to provide DNB protection
to the reactor by anticipating
a complete loss of flow condition
in two or more loops. * This reactor trip requires an underfrequency
condition
Hz) on two RCP buses and is blocked below P-7 (10%). * A rapid decrease in electrical
frequency
can decelerate
the RCPs faster than a complete loss of power. This trip, in conjunction
with the rotational
inertia provided by the RCP flywheel, ensures the proper flow coast-down
time is maintained.
OPT200.RCP
Rev. 5 Page 29 of 56 
A B c o E F G H K Z-&#xa3;9LNGt-Z'1
2 3 I 30" :2: l-F'U2-158-IB
}ro ReP SENSOR PANEl. (UV" UF RELAYS) tlOV AC METERIHC POTENTIAL
ON UNITBD1A t 51
I-i1 L--I r-------<i
L...-J DETAIL A2 :11 :'1
'9 fi 1-..1 r-------<i
I-.-J DETAIL 82 DETAIL C2 I 30" 2 ,30'" 2 I 30 ... :2: REACTOR COOLANT PIMI" I
REACTOR COOLANT PIMP :2: REACTOR COOLANT PIAIP J REACTOR COOLANT PUMP .. 4101/ REACTOR BLDG :1
}+-) &.!!:::!!
""-, t , XS-68-8 Alii( ..,5 ".; X9-SB-8+
I:'
__
81 START + AlO XS-68-84 START
o 2 ... 2Nl HS-68-84A
"';
"7 2 ... 2H ef2 1.6 XS-68-XS-88-14 + +!L-A8 XS-68-+ " 88 """"ii5ii:"" 4
82889 t RCl'I !-___ "_"_R*-.JDI2
B8 HOlI. fUl-68-SC
MOIl" "ux COttY PENETRAUOH
ReP lOlL un PU", l_FU:2:_IB_84A.
a+ l-F'U3-88-84
Rep 2 OIL UFT PU"," l-F'U2-88-85A
85 I-FUJ-68-86
ReP JaIL UFT PUMP l-FU2-6B-86A, a6 l-FU3-,a.-86
ReP .. OIL LIfT PUll" l-FU2-68-17A
a7 l-FUJ-&&-17
2 12 11 2A2Y REACTOR BLDG VENT BD WIRE PREFIX 1.\-" V,2 111-8 JE lA-A 782 .... ,-. ,-. M-O CONT srs HS-68-84Aa:C
HS-68-8i5AaC
HS-68-86AaC
HS-III-87Aa.C
XS-68-&4 XS-68-85 XS-68-86 5 5 6 TACF 2-08-008-068
t' )(5-158-1
.. t 'AnI" -""""l. NOR_ TACF 2-08-008-068
2711. 121,1, "'" 622,1, II[ 272B 622B 7 RaOIS '''')0, 20r e y-. OJ .. " 1 TO ilEAC , Sf RCP UVa. (UMITS aliI. I 1121. I 7 
8 tANH-PMl "r 1-11-72 fUSE SlOlN" AYS !IF , TEST SWITCHES J H UF IA HS-68-343
,. IDII12B IA 208118
HS-68-346 " 2011118 IA 307118 HS<-II-347
HS-H-34B " 3071211 13. 409038 HS-68-3+8
40904B 8 IIA FUSE MO. (NOrEO lA/I IB/I WI ,./1 Ttl REAC TRIP LClCJC L'! .. -i _ -344 -81UlI[ I -f--I -PICA' J 2 P)(I. PilI,
10811al
PICA
9 10 * 8111.)[ , MAINTAINED
PTAIC-SEC
l-FU2-S&-IH
2-fU2-S8-IH
l-FU2-68-J1H
2-fU2-I58-J1H
2-FU2-B8-50H
l-FU2-6S-7JH
2-FU2-S8-7JH
11 12 NOTES: L FOR UNIT 2. rUSE UlUDS OlAMGE UNIT OESICtlATION.
EXAMPLE, l-F'U+-611-6A
'-THIS IIJIIIBER
WILL CHANCE TO 2, FOR UNIT 2. BOARDS. 2. RELAY TYPES ARE AS FOLLOWS: A -D.C. RELAY. toE TTPE 1AC77A B -D.C. RELAY. GE: TYPE IACII6.I C -DISTR RELAY.! TYPE KOIO l. 'IRE DESICNATIONS
SHOWN ARE fOR Rep lOlL LIFT PUMP. WIRE DESIGNATIONS
FOR !tel' 2. J 4 + OlL LIrT I'UtrS ARE SHOrtt 011 CONNECTIOH
DIAGft,wS.
REFERENCE
Dltt,'IM!:;:SI
4SN721-f.
-2 -----6800V UNIT eo SINCt,E LINE DtAGRAWS 4SN755 * ___________
480V RUCTOR BLOC '/[1(1, BO SINGLE LINE DIAGRAMS +5"'721 -----------
ISOOV UNIT SO COHNECTIOI4
DIAGRAIoIS
A B C o E
---------480V REACTOR BLOC VE.rtT so CONNECTION
DJAGRAIoIS
F SYNEIOLS, .. --_ EQUIPMENT
LOCATtD OM UNIT CONTROL &I) t -EQUIPMENT
tlJUNTED OM &!IaDV SWnCHGEAf!:
+ " __ EQUll't.IEMT
IIIlUHTED
OM 480V ..:lTIlR ootnRO\. CENTER c---EOUIPt.lENT
LOCATED ON TVA BOP RACKS III. AUX ItGTR ROCN * -_ EQUIPMENT
t.OCATED ON HSSS RACKS IN AUX INSTR ROCIoI 20 TACY 2-OS-oDS-<l68
INC:TACf'2-Q8-0CI1Hl88
REV CHANGE REI" PREPARER OlECK(II SCALE: HONE TURBINE BUILDING UNITS 1 a:. 2 WIRING DIAGRAMS * , 0/1 APPROVED DATE' EXCEPT AS NOTED CATECORY 1 6900V UNIT AUXILIARY
POWER SCHEMATIC
DIAGRAMS SEQuovAH NUCLEAR PLANT TENNESSEE
VALLEY AUTHORITY
ENGIIlEERIHC
API'RO.fAL
.20 G H 4-{Jul( .. /'fLL -..
., 
9-86-L L9MLjoo-L 2 OVERTEMPERATURE
AT A 8 UNDERVOLTAGE (NOTE 2) c D oA-U E FLOW LOOP 1 F ...... G H 2 OVERPOWER
4T 'OVERPOWER
loT REACTOR TR I P SHEET 1 4 5 UNDERFREQUENCY
RCP-BUSSES (NOTE 2) Loo YQ"IOOD FLOW LOOP 2 FLOW LOOP :3 4 iA.-.2 6 
lliillJ FLOW LOOP 4 \ ' 7 8 9 '0' POWER RANGE 10 12 TURBINE TRIP [SS,JED BYI P.G.TRUDEL
PROCAD MAINTAINED
DRAWING THI!
ctlNTMlL DftA'II'INC
[S
BY 11-&#xa3; :!QN CIIIJ UNIT 0II1II1J lS I<<JII PAltr OF THE T'tA I"Rot:AM.!
DATABASE.
lURIHhI&#xa3; TRIP REAeTOA TlUFI NOTEs: 1, FOR
Attn (liENEIiAl-
$EE 3HEET 1. 1 ,2: sECONOs .altD A B ,0 s c D
TIU&#xa3; E :5. O]GITAL II.HD ANALOC LOCIC SYliISOLS
ARE USED ON LOG;IC [)lAGRAIIS
10 FUNCT10NALLY
DEseRlBE THE PROCESS CONTROL. REFER TO THE ASSOCIATED
IIIUNt: SCHEIdATIC
FOR THE ELECTRICAL
CO\oIPONIi:NTS
USED TO UIPLEIdENT
THE CONTROL SCHU.oIE. REFER TO TABLE .... t FDA ADDITIONAL
DIGITAl COWPUTER POINTS (STEAM GENEAATOR
HI-HI LEYEL TURBINE TI'Up gTATus). 6, MEFE" TO TASLE 1.5 SHEET 5 FOR AODITIOfI,&,L
DI0ITAL COMflUTER
PQ1NT$ t6900Y aHVTD01l'N
aQARD eLACKOUT
PCN MI15.!!.A (TY? ALL BACKCIRCLES)
FmI ORIO'[t.JAL
SIGNATURES
IN TITlE BLOCK SEE REVIslQhI
0 \CICIiOFILWi
'0(11'1'.
Z I1CN l,t11.5sIIA
N/A aGR INC:
RD REV &#xa3;eN/DeN CHG DOC DRiIl'N m*u{O APPD DATE
NONE POWERHOUSE
UNIT 1 MECHANICAL
LOGIC DIAGRAM EXCEPT ,s,S NOTED CATEGORY 1 REACTOR PROTECTION
SYSTEM SEQUOYAH NUCLEAR PLANT Q TENNESSEE
VALLEY AUTHORITY
NUIiLEAR ENF;]NEERING:
DmICN ORMTERl
L.8EASLEY
IJE$lQHERl
REVlt'WER: . __
K.R.SPINO
DATE UUT]AL IssuE RO JS!UE flU: ENC"INEERINC" AP,.IIiC'lAL
.. D.), -04, 1 L,BEASLt"'f
Ii 2 ".8.N[8IUlT
S 'W.R.StDLACII</L'IIA
S-z;1 .. iQ 45 w CGO NO:1-47W611-99-6
R2 F 
I. PROGRAM: OPERATOR TRAINING II. COURSE: SYSTEMS TRAINING III. TITLE: REACTOR COOLANT PUMP SYSTEM IV. LENGTH OF LESSON: 4 hour lecture; 2 hour simulator
demonstration;
2 hour self-study/workshop
V. TRAINING OBJECTIVES:
A. Terminal Objective:
OPT200.RCP
Rev.5 Page 3 of 56 Upon completion
of this lesson and others presented, the student should be able to apply the knowledge
to support satisfactory
performance
of the tasks associated
with the Reactor Coolant Pump system in the plant and on the simulator.
B. Enabling Objectives:
O. Demonstrate
an understanding
of NUREG 1122 knowledge's
and abilities
associated
with the Reactor Coolant Pump System that are rated::::
2.5 during Initial License training for the appropriate
license position as identified
in Appendix A. 1. State the purpose/functions
of the Reactor Coolant Pump System as described
in the SQN FSAR. 2. State the design basis of the Reactor Coolant Pump System in accordance
with the SQNFSAR. 3. Explain the purpose/function
of each major component
in the flow path of the Reactor Coolant Pump System as illustrated
on the simplified
system drawing. 4. Describe the following
items for each major component
in the Reactor Coolant Pump System as described
in this lesson: a. Location b. Power supply (include control power as applicable)
c. Support equipment
and systems d. Normal operating
parameters
e. Component
operation
f. Controls g. Interlocks (including
setpoints)
h. Instrumentation
and Indications
i. Protective
features (including
setpoints)
j. Failure modes k. Unit differences
1. Types of accidents
for which the Reactor Coolant Pump System components
are designed m. Location of controls and indications
associated
with the Reactor Coolant Pump System in the control room and auxiliary
control room 
V. TRAINING OBJECTIVES (Cont'd):
B. Enabling Objectives (Cont'd):
5. Describe the operation
of the RCP system as it relates to the following:
a. Precautions
and limitations
b. Major steps performed
while placing the RCP system in service c. Alarms and alarm response d. How a component
failure will affect system operation
e. How a support system failure will affect RCP system operation
f. How a instrument
failure will affect system operation
OPT200.RCP
Rev.5 Page 4 of 56 6. Describe the administrative
controls and limits for the RCP system as explained
in this lesson: a. State Tech Specs/TRM
LCOs that govern the RCPs b. State the::;1 hour action limit TS LCOs c. Given the conditions/status
of the RCP system components
and the appropriate
sections of the Tech Spec, determine
if operability
requirements
are met and what actions are required 7. Discuss related Industry Events: a. SQ961761PER;
RCP#4 above 15mil vibration
alarm b. SOER-82-5;
Reactor Coolant Pump seal failure c. SER 20-86; RCP shaft failure at Crystal River VI. TRAINING AIDS: A. Classroom
Computer and Local Area Network (LAN) Access B. Computer projector
C. Simulator (if available) 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
SO. 022 AG2.1.28 080 Given the following:
Unit 1 is in Mid-Loop Operation
with RHR Train A in service during a refueling
outage with the reactor core reloaded.
SG nozzle dams are installed.
SI pump 1 B-B is tagged and disassembled
for impeller replacement
CCP 1 A-A trips due to motor failure. RCS level begins to drop and the RHR pump 1 A-A is stopped due to cavitation.
-Attempts to open 1-FCV-63-1, RHR Pump Suction from RWST, are unsuccessful.
Core exit thermocouples
have increased
to 20SoF. Which ONE of the following
identifies
the procedure
to be used and how SI pump IA-A will be aligned during performance
of the procedure?
A. AOP-R02, "Shutdown
LOCA". SI pumps will be aligned for Hot Leg injection.
B. AOP-R02, "Shutdown
LOCA". SI pumps will be prevented
from injecting
due to L TOP requirements. AOP-R03, "RHR System Malfunction".
SI pumps will be aligned for Hot Leg injection.
D. AOP-R03, "RHR System Malfunction".
SI pumps will be prevented
from injecting
due to L TOP requirements.
Page 13 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRA CTOR ANAL YSIS: Page 14 A. Incorrect, AOP-R03 is the correct procedure
to be used not AOP-R02. The AOP-R03 does direct the SI pumps to be aligned in Hot Leg injection.
Plausible
because the conditions
in the stem would exist if a shutdown LOCA was occurring
and if AOP-R02 were entered a note at the start of the AOP directs the use of AOP-R03 if the plant is in Mid-Loop operation
and the SI pumps alignment
in Hot Leg injection
is correct. B. Incorrect, AOP-R03 is the correct procedure
to be used not AOP-R02 and the AOP directs the SI pumps to be aligned in Hot Leg injection.
Plausible
because the conditions
in the stem would exist if a shutdown LOCA was occurring
and if A OP-R 02 were entered a note at the start of the A OP directs the use of A OP-R 03 if the plant is in Mid-Loop operation
and the SI pumps normal alignment
with injection
prevented
due to the L TOPS requirement.
C. CORRECT, when the charging pump trips (loss of RCS makeup) the vessel level will start to decrease.
When the RHR pump is stopped due to cavitation, core cooling flow is terminated, the core starts heating up and with 1-FCV-63-1
unable to be opened, the AOP -R03 directs the SI pumps to be aligned for Hot Leg Injection.
A note prior to the step to align the pumps states that core cooling takes priority over the L TOPS requirements
for the SI pumps. The conditions
would be the same if a shutdown LOCA was occurring
and if AOP-R02 was entered there is a note at the start of the A OP directing
the use of A OP-R 03 if the plant is in Mid-Loop operation.
D. Incorrect, AOP-R03 is the correct procedure
to be used but the AOP directs the SI pumps to be aligned in Hot Leg injection.
Plausible
because the procedure
is the correct procedure
to be used and and the SI pumps normal alignment
is with injection
prevented
due to the L TOPS requirements. 
( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 80 Tier 1 Group 1 KIA 022 AG2.1.28 Loss of Reactor Coolant Makeup Knowledge
of the purpose and function of major system components
and controls.
Importance
Rating: 4.1 14.1 Technical
Reference:
AOP-R02, Shutdown LOCA, Rev 10 AOP-R03, RHR System Malfunction, Rev 20 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 AOP-R03 B.5 Question Source: Summarize
the mitigating
strategy for the failure that initiated
entry into AOP-R03. Bank# ___ _ Modified Bank # X ---New ---Question History: SON 1/2009 exam, AOP-R03 006 modified Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X __ 10 CFR Part 55 Content: (41.7) 10CFR55.43.b ( 5 ) Comments:
SON question AOP-R03 006 modified Page 15 
(' SQN RHR SYSTEM MALFUNCTION
AOP-R.03 Rev. 20
.1 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.0 OPERA TOR ACTIONS NOTE: If this procedure
is entered from AOP-P .05 or AOP-P.06 due to loss of shutdown board(s), then Section 2.3 is the applicable
section. 1. DIAGNOSE the failure: GOTO IF ... SECTION PAGE RHR malfunctions
due to low water level 'J 1 J1 during reduced inventory
or mid-loop operations RHR overpressurization
due to high RCS pressure 2.2 29 RHR pump(s) failure or trip 2.3 36 RHR system leak 2.4 41 Failure of RHR due to Loss of CCS 2.5 48 END OF SECTION / Page 3 of 97 ! 
( SQN RHR SYSTEM MALFUNCTION
AOP-R.03 Rev. 20 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions
Due to Low Water Level During Reduced Inventory
or Mid-Loop Ops CAUTION Changes in RCS pressure could result in inaccuracies
in RCS level readings.
1. DETERMINE
whether RHR pumps should be STOPPED: a. CHECK any RHR pump RUNNING. b. MONITOR RCS level c. CHECK RHR flow less than 2000 gpm. d. CHECK RHR pump CAVITATING.
e. STOP RHR pumps and PLACE in PULL-TO-LOCK.
a. GO TO Step 2. b. STOP RHR pumps and GO TO Step 2. c. REDUCE RHR flow to between 1000 gpm and 1500 gpm. [C.6] d. PERFORM the following:
1) RESTORE RCS level to normal band by adjusting
charging and letdown. 2) IF RHRleak is suspected, THEN GO TO Section 2.4, RHR Leak. 3) GO TO appropriate
procedure.
Page 4 of 97 
SQN RHR SYSTEM MALFUNCTION
AOP-R.03 Rev. 20 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions
Due to Low Water Level During Reduced Inventory
or Mid-Loop Ops (continued)
2. CHECK RCS Vacuum Refill NOT in progress.
3. ISOLATE RCS letdown and drain paths: a. CLOSE FCV-62-83, RHR Letdown Flow Control Valve. b. CLOSE CVCS letdown isolation
valves: * FCV-62-69
* FCV-62-70
c. ISOLATE any known RCS drain paths. IF RCS vacuum refill in progress, THEN PERFORM the following:
a. ENSURE vacuum break valve VB OPEN. [Vacuum Refill Skid] b. ENSURE vacuum pump STOPPED. [Vacuum Refill Skid] c. ENSURE charging flow has been raised. d. WHEN RCS is at atmospheric
pressure, Incl''4 PERFORM the following:
1) ENSURE PZR PORVs CLOSED. 2) ENSURE Rx Head Vent FSVs CLOSED. Page 5 of 97 
SON RHR SYSTEM MALFUNCTION
AOP-R.03 Rev. 20 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions
Due to Low Water Level During Reduced Inventory
or Mid-Loop Operations (continued)
CAUTION Containment
will become a harsh environment
due to steam and potential
airborne activity if boiling occurs. [C.S] 4. INITIATE actions to protect personnel
in containment:
a. ANNOUNCE over PA to evacuate
personnel
from containment.
b. NOTIFY RADCON to evacuate
personnel
I fOrTI COflldlf IIIH:Jfll.
c. NOTIFY RADCON to monitor containment
radiation
conditions.
[C.10] 5. INITIATE actions to establish
containment
closure: a. NOTIFY WCC to initiate closure of all containment
penetrations
being tracked in
Containment
Closure Control. b. ENSURE all valves with Containment
Closure tags on MCR bench boards are CLOSED. Page 6 of 97 
SQN RHR SYSTEM MALFUNCTION
AOP-R.03 Rev. 20 I STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions
Due to Low Water Level During Reduced Inventory
or Mid-Loop Operations (continued)
6. EVALUATE the following
Tech Specs for applicability:
* 3.4.1.4, Reactor Coolant System Cold Shutdown * 3.9.8.1, Refueling
Operations
Residual Heat Removal and Coolant Circulation
* 3.9.8.2, Refueling
Operations
Low Water Level 7. EVALUATE EPIP-1, Emergency
Plan Classification
Matrix. Page 7 of 97 
SON RHR SYSTEM MALFUNCTION
AOP-R.03 Rev. 20 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions
Due to Low Water Level During Reduced Inventory
or Mid-Loop Operations (continued)
CAUTION The following
step may result in water spilling in lower containment
if RCS is breached. . NOTE: Opening FCV-63-1 will raise RCS level and flush air from the high point in the RHR suction line using gravity fill from RWST. 8. RAISE RCS level to fill RHR suction line: a. IF any S/G primary side manways =n .*. Q'v'V 1-,', THEN CONTACT RADCON to verify personnel
are clear of S/G platforms.
b. DISPATCH operator to ensure power restored to FCV-63-1 USING Appendix A. c. ENSURE FCV-74-1 and FCV-74-2 OPEN. d. CLOSE FCV-74-3 and FCV-74-21.
b. IF power CANNOT be restored to FCV-63-1, THEN DISPATCH operator with radio to operate FCV-63-1 locally. [AB el. 669 Pipe Chase] (Step continued
on next page) Page 8 of 97 
SQN RHR SYSTEM MALFUNCTION
AOP-R.03 Rev. 20 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions
Due to Low Water Level During Reduced Inventory
or Mid-Loop Operations (continued)
8. e. OPEN FCV-63-1 USING one of the following
methods: * handswitch
in MCR * handswitch
on Rx. MOV Bd * local handwheel
[AB el. 669 pipe chase] f. WHEN FCV-63-1 has been THEN PERFORM the following:
1) CLOSE FCV-63-1.
2) OPEN FCV-74-3 and FCV-74-21.
N<J1
NOTE: ERCW isolation
valves for available
upper and lower compartment
coolers which were closed in step 5 may be reopened if ERCW piping is intact. 9. START available
upper and lower compartment
coolers USING Appendix B. Page 9 of 97 . / -.! 
SQN STEP RHR SYSTEM MALFUNCTION
AOP-R.03 Rev. 20 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions
Due to Low Water Level During Reduced Inventory
or Mid-Loop Operations (continued))
----------------------------
-------The following
step establishes
ECCS flow for feed-and-bleed
cooling if core boiling point is approached.
Restoration
of core cooling takes priority over L TOPS requirements
for SI pumps. 10. MONITOR Core Exit TICs less than 200&deg;F USING available
TICs on Exosensor
display. PERFORM the following:
a. ESTABLISH
SI pump flow to hot legs as follows: 1) ENSURE operator dispatched
to perform App. C, Restoring
Power to SI Pumps and Hot Leg Inj Valves. 2) ENSURE FCV-63-5, SI pump suction from RWST, OPEN. 3) ENSURE SI pump suction valves OPEN: * FCV-63-47, SI pump A suction * FCV-63-48, SI pump B suction 4) ENSURE SI pump cold leg injection
flowpath isolated:
* FCV-63-22
CLOSED OR * FCV-63-152
and FCV-63-153
CLOSED 5) OPEN SI pump hot leg injection
valve for SI pump to be started: * FCV-63-156, SI pump A OR * FCV-63-157, SI pump B 6) ENSURE one SI pump RUNNING. (step continued
on next page) Page 10 of 97 
SQN SHUTDOWN LOCA lAO. P-R.02 Rev. 10 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS NOTE 1: AOP-R.03 should be used for LOCA while in reduced inventory
or midloop. NOTE 2: This procedure
has a foldout page. 1. MONITOR if RHR pumps should be stopped: a. CHECK RHR aligned for shutdown cooling: * FCV-74-1 OPEN *
OPEN. b. CHECK the following:
* Pressurizer
level less than 10% [20% ADV] OR * RGS subcooling
based on core exit T/Cs less than 58&deg;F c. STOP RHR pumps and PLACE in PULL-TO-LOCK.
a. GO TO Step 2. b. IF Pressurizer
level greater than 10% [20% ADV] AND RCS subcooling
greater than 58&deg;F, THEN PERFORM the following:
1) DISPATCH operator to ensure HCV-74-34
RHR Return to RWST CLOSED. [AB el. 690, RHR HX Room B] 2) GO TO Step 2. (step continued
on next page) Page 3 of 90 
SON 1.0 PURPOSE SHUTDOWN LOCA I AOP-R.02 Rev. 10 The procedure
provides the actions necessary
to mitigate the effects of a LOCA which exceeds normal charging capacity during Mode 4 or Mode 5 (with the exception
of leaks during reduced inventory/midloop
operation
resulting
in loss of RHR, which are addressed
in AOP-R.03).
[C.1] Page 2 of 90 
AOP-R.03 006 QUESTIONS
REPORT for BANK SQN Questions
Unit 1 is preparing
for a refueling
outage, the unit is in mode 6 RCS is at Reduced Inventory
level. Indications
of a LOCA are observed.
Which of the following
procedures
is applicable?
A. AOP-R02 Shutdown LOCA B:' AOP-R03 RHR System Malfunction
C. AOP-R05 RCS Leak & Leak Source Identification
D. E-1 Loss of Reactor or Secondary
Coolant 'j!\" incorrect:
Note in AOP-R.02 states that AOP-R.03 should be used for LOCA when in reduced inventory
or midloop. "B" correct: per note in AOP-R.02 "c" incorrect:
Note in AOP states to use AOP-R.03 if in reduced inventory
or mid/oop. "0" incorrect:
Unit in mode 5 Reference:
KIA: 2.4.4 (4.0 -4.3) 41.10/43.2/45.6
034 A 1.02 (2.9 -3.7) 41.5/45.5
OPL273C0611
obj 6 Friday, November 14, 2008 7:44:31 AM 1 
OPL271 AOP-R.03 Revision 2 Page 3 of 39 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-R03, RHR SYSTEM MALFUNCTION
IV. LENGTH OF LESSON/COURSE:
2 hours V. TRAINING OBJECTIVES:
o. 1. 2. 3. 4. 5. 6. A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of AOP-R03, RHR SYSTEM MALFUNCTION.
B. Enabling Objectives
Objectives
Demonstrate
an understanding
of NUREG 1122 knowledge's
and abilities
associated
with RHR SYSTEM MALFUNCTIONs
that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification
Training for the appropriate
position as identified
in Appendix A. State the purpose/goal
of this AOP-R.03.
Describe the AOP-R03 entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with AOP-R03 entry conditions.
b. Describe the ARP requirements
associated
with AOP-R03 entry conditions.
c. Interpret, prioritize, and verify associated
alarms are consistent
with AOP-R.03 entry conditions.
d. Describe the plant parameters
that may indicate an RHR System Malfunction.
Describe the initial operator response to stabilize
the plant upon entry into AOP-R.03.
Upon entry into AOP-R03, diagnose the applicable
condition
and transition
to the appropriate
procedural
section for response.
Summarize
the mitigating
strategy for the failure that initiated
entry into AOP-R03. Describe the bases for all limits, notes, cautions, and steps of AOP-R03. 
7. 8. 9. 10. OPL271 AOP-R.03 Revision 2 Page 4 of 39 Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
Given a set of initial plant conditions
use AOP-R.03 to correctly:
a. Recognize
entry_ conditions.
b. Identify required actions. c. Respond to Contingencies.
d. Observe and Interpret
Cautions and Notes. Describe the Tech Spec and TRM actions applicable
during the performance
of AOP-R.03.
Apply GFE and system response concepts to the abnormal condition
-prior to, during and after the abnormal condition.
/ 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
81 . 077 AG2.4.5 081 Given the following:
-Unit 2 operating
at 100% power with the switchyard
in normal alignment.
-Generator
operating
at 24 Kv and +150 MVAR. -A disturbance
occurs causing main generator
voltage to spike upward and the following
annunciators
alarm: -on 0-XA-55-ECB6-A; "GEN 2 MVARABNORMAL
OR MVAR RELAY FAILURE" "OSCILLOGRAPH
OPERATION
OR FAILURE" -on 0-XA-55-ECB6-B; "CC RELAY TEST OR OPERA TIOi'fV -on 2-XA-55-1A "GEN VOLT REGULATOR
TRIP" -The BLUE light on161kV CONCORD LINE CARRIER RECEIVED is lit. -Reactive power stabilized
at +230 MVARS following
the disturbance
and Concord line PCB in the switchyard
remain closed. 3 Which ONE of the following
statements
describes
additional
required actions? A. Manually open the Concord line PCB. Notify the Southeast
Area Load Dispatcher
to evaluate offsite power status for determining
operability.
B. Manually open the Concord line PCB. Declare both trains of Offsite power inoperable
until the Southeast
Area Load Dispatcher
completes
evaluation
of the status of the offsite power system. C!' Notify the Southeast
Area Load Dispatcher
to evaluate offsite power status for determining
operability.
Within 24 hours, notify the Operations
Duty Specialist
of the time period the unit was without automatic
voltage control. D. Declare both trains of Offsite power inoperable
until the Southeast
Area Load Dispatcher
completes
evaluation
of the status of the offsite power system. Within 24 hours, notify the Operations
Duty Specialist
of the time period the unit was without automatic
voltage control. Page 16 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRACTOR ANAL YSIS: Page 17 A. Incorrect, The Concord Line PCB is not required to be opened with the stated conditions.
The Southeast
Area Load Dispatcher
is required to be notified immediately
to determine
off site power voltage requirements
to determine
if offsite power supplies are operable.
Plausible
because conditions
would require opening the PCB and if the PCBs had opened, the MOD s are required to be opened manually.
Additionally, the dispatcher
notification
to detemine status of the off site power operability
is required.
B. Incorrect, The Concord Line PCB is not required to be opened with the stated conditions
and the Southeast
Area Load Dispatcher
is required to be immediately
notified so that an evaluation
can be performed
but the offsite power operability
status is determined
after the evaluation
is complete.
Plausible
because conditions
would require opening the PCB and if the PCBs had opened, the MOD s are required to be opened manually.
Additionally, the dispatcher
notification
to detemine status of the offsite power is required so that an evaluation
can be completed
to determine
the offsite power operability
status. C. CORRECT, Operation
of the unit without automatic
voltage control requires the Southeast
Area Load Dispatcher
be notified immediately
to determine
offsite power voltage requirements
to determine
if off site power supplies are operable and the Operations
Duty Specialist
is required to be notified of the time interval without automatic
voltage control within 24 hours. D. Incorrect, The Southeast
Area Load Dispatcher
is required to be immediately
notified so that an evaluation
can be performed
but the offsite power operability
status is determined
after the evaluation
is complete.
The Operations
Duty Specialist
is required to be notified of the time interval without automatic
voltage control within 24 hours. Plausible
because the dispatcher
notification
is required so that an evaluation
can be completed
so that the offsite power operability
status can be determined
and the Operations
Duty Specialist
notification
is required within 24 hours. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 81 Tier 1 Group 1 KIA 077 AG2.4.5 Generator
Voltage and Electric Grid Disturbances
Ability to prioritize
and interpret
the significance
of each annunciator
or alarm. Importance
Rating: 4.1 14.3 Technical
Reference:
O-AR-ECB6-A, Electrical
Control Board, Rev 38 2-AR-M1-A, Generator
and Transformers, Rev 33 GOI-6, Apparatus
Operation, Rev. 128 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
No Learning Objective
identified
Question Source: Bank# ___ _ Modified Bank # ___ _ New X ---'------
Question History: New question for Sequoyah 2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X __ 10 CFR Part 55 Content: ( 41.10/43.5/45.3
/45.12 ) 10CFR55.43.b ( 2,5 ) Comments:
New question for Sequoyah 2009 exam Page 18 
Source Setpoint SER 2743 N/A 1. Carrier Test Signal 27 CC RELAY TEST OR OPERATION
(0-6) 2. Carrier Blocking Signal sent or received for any 161kV or 500kV PCB Probable Causes Corrective
Actions References
1. Manual actuation
of carrier test for any 161 kV on 500kV PCB. 2. Carrier blocking signal sent or received for any 161 kV or 500 kV line. (EM relays only for Bradley line) [1] OBSERVE Line Loading (Amps) and PCB position indicating
lights to VERIFY affected PCB has not tripped. [2] IF Line PCB has tripped, THEN OPEN the related MODs. [3] NOTIFY Dispatcher
to EVALUATE IF the line needs to De returned to service. [4] IF no PCB has tripped, THEN RESET the Carrier Signal indicating
light (light and reset button located on apron of ECB or on Static Relay Panels in Relay Room). 55N651,45B655-ECB6-B
SQN O-AR-EC86-8
Page 36 of 44 o Rev. 20 __ . __________ -L ______________ __________ 
( 33 (E-5) Source Setpoint SER 3091: RELAY 274VH OP OR FAIL 220 MVAR increasing
GEN 2 MVAR ABNORMAL OR MVAR RELAY FAILURE SER 3092: RELAY 274VL OP OR FAIL -100 MVAR decreasing
Probable Causes CAUTION NOTE Corrective
Actions References
1. High MVAR on Unit 2 main generator.
2. Low MVAR on Unit 2 main generator.
3. Failure of relay 274VH or 274VL. 4. Failure of supervisory
MW relay 237W, at low power. Unit 2 main generator
must be operated within the limits shown in the generator
capability
curve (TI-28, figure A.14). This alarm is automatically
defeated when Unit 2 output < 240 MWE by relay 237W. [1] CHECK Unit 2 MVARs indicated
on 2-EI-57-8
(2-M-1). [2] CHECK CRT SER point to determine
which relay actuated.
[3] IF Unit 2 MVARs out of limit, THEN [a] ADJUST voltage USING Unit 2 generator
voltage adjust and/or intertie transformer
tap changer to restore MVARs to within limit. [b) REFER to GOI-6 and 15E500 Sheet 3. [c] EVALUATE off-site power sources for Tech Spec LCO 3.8.1.1. [d] IF attempts to restore Unit 2 MVARs to within limit are unsuccessful, THEN CONTACT Dispatcher
for assistance.
[4] IF Unit 2 MVARs are within limits, THEN CONTACT Transmission
Power Supply (TPS) to investigate
suspected
failure of relay 274 VH or 274 VL. [5] IF Unit 2 output < 240MWe, THEN CONTACT Transmission
Power Supply (TPS) to investigate
suspected
failure of supervisory
MW relay 237W. 55N652-1, 1,2-458655-ECBBA-O, 2-45W541, 55N715-1 SQN O-AR-ECB6-A
Page 42 of 45 0 Rev. 38 
Source SER 1545 94RB relay 260 relay Probable Causes Corrective
Actions CAUTION 3 (A-3) Setpoint N/A GEN VOLT REGULATOR
TRIP 1. Loss of 30 Regulating
Potential.
2. Exciter Field Breaker (41) OPEN. 3. Volts-Hertz
Relay Operation.
4. Generator
overexcitation
relay operation.
[1] CONFIRM alarm by verifying
[2-HS-57-20]
Exciter Regulator
Control Switch green light LIT. If the exciter field breaker opens with the generator
tied to grid, rapid and excessive
heating of the stator will occur. If this condition
exists, the generator
should trip by 221 GB relay operation.
[2] IF Exciter Field Breaker is TRIPPED, AND PCB's CLOSED, AND Generator
NOT TRIPPED, THEN PERFORM the following:
[a] IF Reactor Power greater than 50% (P-9), THEN TRIP the Reactor, AND GO TO E-O, Reactor Trip or Safety Injection.
[b] IF Reactor Power is less than 50% (P-9), THEN TRIP the turbine, AND GO TO AOP-S.06, Turbine Trip. [3] PLACE [2-HS-57 -20] Exciter Regulator
Control Switch to OFF position.
[4] IFWindow A-6, GENERATOR
EXCITER FIELD OVERCURRENT
is in alarm, THEN GO TO Window A-6 of this Instruction.
[5] IF Window B-2, GENERATOR
VOLTS PER CYCLE HIGH is in alarm, THEN GO TO Window B-2 of this Instruction. (step continued
on next page) SQN 2 \2-AR-M1-A
Page 6 of 53 Rev. 33 
( CONTINUED
CAUTION Corrective
Actions (Continued)
References . 3 (A-3) GEN VOLT REGULATOR
TRIP Operation
without automatic
voltage regulator
may impact offsite power voltage requirements.
[6] IF Unit 2 main generator
remains in service without automatic
voltage regulator, THEN PERFORM the following:
[a] NOTIFY SELD to evaluate off-site power voltage requirements
with Unit 2 voltage regulator
in MANUAL. [b] MAINTAIN Unit 2 MVARs within limits specified
in GOI-6, Apparatus
Operations, Section E, Turbogenerator
Operations.
[7] CHECK for blown Voltage Regulating
PT fuses with [2-HS-57 -15] Generator
Voltmeter
Selector Switch. [8] NOTIFY Operations
Duty Specialist (ODS) within twenty four (24) hours of any time interval without automatic
voltage control. 45B655-01A-O, 45N573-1, 45N551 SQN 2 !2-AR-M1-A
Page 7 of 53 Rev. 33 
( \ SQN APPARATUS
OPERATIONS
GOI-6 Rev: 128 Page 48 of 174 SECTION E Page 2 of 4 3.0 MVAR LIMITS FOR GENERATOR
STABILITY (REFERENCE
USE) NOTE Operation
of main generator
without automatic
voltage control could impact grid voltage requirements.
SELD should be notified immediately
if automatic
voltage regulator
is lost. Studies show that there is some risk of instability
in the event of a fault at high side of SON 500/161 kV Intertie Transformer
Bank plus a failure of a high side breaker to clear. Backup breakers would then take the entire 500kV bus section out of service. This double-fault
event is not postulated
to occur simultaneously
with a LOCA and is therefore
not a scenario used to determine
nuclear offsite power adequacy.
This is an issue related to grid reliability
only and operating
guidelines
to ensure stability
are included in this document for convenience.
SON Units 1 and 2 must observe generation
limits under certain grid conditions
in order to ensure stability
under the above double-fault
scenario.
Both Units are limited to a Maximum Outgoing Reactive Load of 240 Mvar. This limitation
supports offsite power source qualification.
Transmission
Reliability
Organization's
SON Grid Operating
Guide directs that the Transmission
Operator will notify the SON Generator
Operator of any Mvar Limits recommended.
Grid stabilization
following
the loss of an element depends on the coordination
of multiple changes including
SON reactive loading. Real time information
on the factors affecting
grid stability
is not available
to SON Operators, therefore
the Transmission
Operator will coordinate
the effort. The limits provided in the following
tables are for information
only. 
( SQN APPARATUS
OPERATIONS
GOI-6 Rev: 128 Page 53 of 174 SECTION F Page 3 of 9 3.1 Offsite Power Source Requirements (continued)
D. The plant will coordinate
and communicate
with the SELD for entry into and exit out of the alternate
alignment
so the transmission
operators
will know which criteria to use in monitoring
Sequoyah Nuclear Plant offsite power adequacy.
E. The SON 161 kV normal scheduled
voltage is 165kV, +/-1 kV. The voltage may be increased
up to 168kV if necessary
due to light load conditions.
F. The 161kV switchyard
undervoltage
relay is set to alarm at 164KV to ensure the minimum 161 kV grid level required to maintain a minimum of 6,560 volts at the 6,900 volt shutdown boards for design basis trips. G. The 500kV bus voltage should be maintained
at a level of 525kV. This should be done by means of the Generator
No.1 reactive control coordinated
with the load tap changer for the SON 500/161 kV Intertie Transformer
Bank while adhering to the 161 kV bus voltage schedule.
During emergencies
or abnormal conditions, the 500kV bus voltage may be raised as coordinated
with the power system dispatcher, but it should not exceed 535kV. H. The load tap changer (L TC)on the high side winding of the intertie transformer, 161 KVcapacitor
banks, and reactive output of the units shall be coordinated
to maintain the published
voltage schedule.
[C.3] I. If, for any reason, the voltage schedule cannot be maintained, the Southeast
Area Load Dispatcher (SELD) should be notified as soon as possible to evaluate offsite power operability.
This notification
shall include the time and date of the start of the inability
to maintain the voltage schedule, an explanation
of the problem and the time of anticipated
return to compliance.
J. Operation
of main generator
without automatic
voltage control could impact grid voltage requirements.
The Load Dispatcher
should be notified immediately
if generator
is in service without automatic
voltage regulator.
Also, refer to Section E for Mvar limits. 
29 (E-1) Source Setpoint SER 2683 * Any actuation
signal for Oscillograph
No.1 or No.2 NfA OSCILLOGRAPH
OPERATION
OR FAILURE * Loss of control power to Oscillograph
No.1 or No.2 Probable Causes Corrective
Actions References
1. Fault condition
on a transmission
line. 2. Fault condition
at a remote yard or substation.
3. 500KV or 161 KV bus voltage drop. 4. Unit trip. 5. Loss of control power to either Oscillograph.
6. Oscillograph
sending a fax. NOTE: If alarm asa result of Oscillograph
sending a fax, NO operator action required.
[1] CHECK for other annunciators
lit or breaker.cJisagreement
lights lit. [2] IF Reactor Trip, THEN GO TO E-O, Reactor Trip or Safety Injection.
[3] DISPATCH operator to the relay room to check for any relay targets, AND WRITE down relay targets dropped. [4] NOTIFY Dispatcher
of Oscillograph
operation
and targets noted in step [3]. 45B655-E CB6A-O 55N634-1,2
55N3763-1, 2,3,4 SQN 0 O-AR-ECB6-A
Page 37 of 45 Rev. 38 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
82. 068 AG2.4.7 082 Given the following:
Conditions
require that the Main Control Room (MCR) be abandoned.
While completing
the MCR actions of AOP-C.04, "Shutdown
From Auxiliary
Control Room", the Unit 2 OATC notices that a Safety Injection (SI) has occurred.
-All AOP-C.04 actions in the MCR were completed.
-The crew establishes
control in the Auxiliary
Control Room and determines that
the plant is to be cooled down to Mode 5. Which ONE of the following
identifies
... (1) how the safety injection
termination
will be performed
and (2) the instruments
on 2-L-10 used to trend the RCS cooldown rate if the RCPs remain out of service? A. (1) AOP-C.04 directs the use of ES-1.1, SI Termination, to terminate
the SI. (2) The Thot instruments
on each loop. B. (1) AOP-C.04 directs the use of ES-1.1, SI Termination, to terminate
the SI. (2) The SG Main Steam Pressure instrument
on each loop. C. (1) SI termination
steps are contained
within AOP-C.04.
(2) The Thot instruments
on each loop. D!' (1) SI termination
steps are contained
within AOP-C.04.
(2) The SG Main Steam Pressure instrument
on each loop. Page 19 
( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRACTOR ANAL YSIS: Page 20 A. Incorrect, The SI will be terminated
using AOP-C.04 Appendix F, Terminating
SI flow not by the use of ES-1.1, SI Termination
and with all MCR actions completed, the reactor coolant pumps are shutdown leaving the RCS cooldown to be trended by the SG pressure instruments
which have a saturation
temperature
scale on them. Thot can only be used when the RCPs are in service. Plausible
because procedure
ES-1.1, SI Termination, is the procedure
used for SI termination
in many conditions
and Thot could be used to monitor cooldown if the RCPs were in service. B. Incorrect, The SI will be terminated
using AOP-C.04 Appendix F, Terminating
SI flow not by the use of ES-1.1, SI Termination.
Since all MCR actions have been completed, the reactor coolant pumps are shutdown resulting
in the steam generator
pressure instruments
being used to trend the RCS cooldown to be correct. Plausible
because procedure
ES-1.1, SI Termination, is the procedure
used for SI termination
in many conditions
and the steam generator
pressure instruments
are used to monitor cooldown.
C. Incorrect, The SI will be terminated
using AOP-C.04 Appendix F, Terminating
SI flow but with all MCR actions completed, the reactor coolant pumps are shutdown leaving the RCS cooldown to be trended by the SG pressure instruments
which have a saturation
temperature
scale on them. Thot can only be used when the RCPs are in service. Plausible
because the procedure
used for SI termination
is correct and Thot could be used if the RCPs were in service. D. CORRECT, The SI will be terminated
using AOP-C.04 Appendix F, Terminating
SI flow and with all MCR actions completed, the reactor coolant pumps are shutdown leaving the RCS cooldown to be trended by the SG pressure instruments
which have a saturation
temperature
scale on them. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 82 Tier 1 Group 2 KIA 068 AG2.4.7 Control Room Evacuation
Ability to diagnose and recognize
trends in an accurate and timely manner utilizing
the appropriate
control room reference
material.
Importance
Rating: 4.2/4.2 Technical
Reference:
AOP-C.04, Shutdown From Auxiliary
Control Room, Rev 16 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271-C.04
B.5 & 10.b Question Source: Describe the actions that must be taken prior to abandoning
the main control room, including
a basis for each actions. Describe actions per AOP-C.04, that are required to: b. Cooldown from Aux Control Room. Bank# ___ _ Modified Bank # ___ _ New X ---Question History: New question for SQN 1/2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X __ 10 CFR Part 55 Content: ( 41.10/43.5/45.12 ) 10CFR55.43.b ( 5 ) Comments:
New question for SQN 1/2009 exam Page 21 
Page 1 of 1 file:III:\Sequoyah\Control
Room Photos\2-L-1O\2-L-1O-01.JPG
07/2312008 
( SQN SHUTDOWN FROM AUXILIARY
CONTROL ROOM AOP-C.04 Rev. 16 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 Control Room Abandonment
NOTE EOPs are NOT applicable
when evacuating
MCR. 2. ENSURE reactor TRIPPED. [M-4] 3. ENSURE MSIVs and MSIV bypass valve handswitches
in CLOSE. [M-4] 4. PLACE RCP handswitches
*in STOP/PULL
TO LOCK. [M-5] 5. ENSURE one CCP placed in PULL TO LOCK. 6. IF MCR must be
evacuated
due to life-threatening
conditions, THEN PERFORM the following:
a. EVACUATE MCR on affected unit(s). h. NOTIFY AUOs of MeR evacuation
using radio or PAsystem.
c. GO TO Cautbn prior to Step 11. ._. ___ ._""' ... _ .. _ ....... ___
.... *_.-=_** _'-_______________ . Page 5 of 185 
SQN STEP SHUTDOWN FROM AUXILIARY
CONTROL ROOM AOP-C.04 Rev. 16 ACTION/EXPECTED
RESPONSE I RESPONSE NOT OBTAINED 2.1 Control Room Abandonment (cont'd) NOTE 1 NOTE Accidents
requiring
containment
spray or ECCS operation
are outside scope of this procedure.
If containment
spray or ECCS is running during MCR abandonment
event, it is assumed that these pumps are not needed and should be stopped to prevent RWST depletion
and/or pressurizer
overfill.
Checklist
5 (Unit 1) or 6 (Unit 2) directs local operator to ensure CCPIT valves closed from Rx MOV Boards within 13 minutes. 21. CHECK SI signal NOT actuated:
IF SI signal has actuated, THEN * NO indication
of SI actuation
prior to leaving MCR * NO reports of SI or RHR pump breakers CLOSED. PERFORM Appendix F, Terminating
SI Flow.
Page 13 of 185 
( SQN STEP SHUTDOWN FROM AUXILIARY
CONTROL ROOM AOP-C.04 Rev. 16 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 Control Room Abandonment (cont'd) 24. CHECK unit in Mode 3 (greater than 350&deg;F). IF unit in Mode 4, 5, or 6, THEN NOTE 1 NOTE 2 NOTE 3 GO TO Step 26. S/G pressure indicators
on L-10 are scaled to correlate
S/G pressure to T -sat. S/G T-sat indicates
approximate
RCS T-cold. Atmospheric
relief controllers
should be set for 85% in AUTO to maintain T-cold at approximately
547&deg;F. Local control of S/G #1 and 4 atmospheric
relief valves may be required if essential
air has been lost. 25. CONTROL RCS temperature:
a. ENSURE S/G atmospheric
relief valves maintaining
RCS T-cold at desired value (540-550&deg;F
following
trip). b. GO TO Step 27. Page 18 of 185 a. OPERATE S/G #1 and 4 atmospheric
relief valves locally as necessary:
(60 minutes) 1 ) DISPATCH personnel
to perform Appendix K, Local Control of S/G Atmospheric
Reliefs. 2) PLACE S/G #1 and 4 atmospheric
relief valve controllers
in MANUAL and ADJUST controller
output to zero. 
OPL271 AOP-C.04 Revision 2 Page 3 of 27 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-C.04, SHUTDOWN FROM AUXILIARY
CONTROL ROOM IV. LENGTH OF LESSON/COURSE:
3 hour(s) V. TRAINING OBJECTIVES:
O. 1. 2. 3. 4. 5. 6. 7. 8. A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of AOP-C.04, SHUTDOWN FROM AUXILIARY
CONTROL ROOM. B. Enabling Objectives:
Objectives
Demonstrate
an understanding
of NUREG 1122 knowledge's
and abilities
associated
with Shutdown from the Auxiliary
Control Room that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification
Training for the appropriate
position as identified
in Appendix A. State the purpose/goal
of this AOP-C.04.
Describe the AOP-C.04 entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with AOP-C.04 entry conditions.
b. Describe the ARP requirements
associated
with AOP-C.04 entry conditions.
c. Interpret, prioritize, and verify associated
alarms are consistent
with AOP-C.04 entry conditions.
d. Describe the plant parameters
that may indicate a Shutdown from the Auxiliary
Control Room is required.
Upon entry into AOP-C.04, diagnose the applicable
condition
arid transition
to the appropriate
procedural
section for response.
Summarize
the mitigating
strategy for the failure that initiated
entry into AOP-C.04.
Describe the actions that must be taken before abandoning
the main control room, including
a basis for each action. Explain the staffing requirements
for unit abandonment
per AOP-C.04.
Describe the types of equipment
that are on the various checklists
associated
with AOP-C.04 Describe the actions that may be necessary
if procedure
steps are taken before all checklists
are complete. 
9. 10. 11. 12. 13. 14. 15. Objectives
Describe the bases for the limits, notes, cautions of AOP-C.04.
Describe actions per AOP-C.04, that are required to: a. Maintain Plant in Hot Shutdown b. Cooldown plant form Aux. Control Room c. Return to Main Control Room OPL271 AOP-C.04 Revision 2 Page 4 of 27 Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
Given a set of initial plant conditions
use AOP-C.04 to correctly:
a. Recognize
entry conditions.
b. Identify required actions. c. Respond to Contingencies.
d. Observe and Interpret
Cautions and Notes. Describe the Tech Spec and TRM actions applicable
during the performance
of AOP-C.04.
Discuss the parameters
to be considered
by the SED when making a REP classification
during a control room evacuation.
Apply GFE and system response concepts to the abnormal condition
-prior to, during and after the abnormal condition. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
83. 076 AA2.04 083 Given the following
conditions:
Unit 1 is in Mode 3 with RCS at normal operating
pressure and temperature
awaiting secondary
plant equipment
repair to continue the startup. -At 1300 on 01/25109 RCS Activity was determined
to be 0.38 microcuries/gram
DOSE EQUIVALENT
1-131. -At 1300 on 01/27/09 Chemistry
reports that the RCS Activity has been on a continuous
slow increase and is now 0.43 microcuries/gram
DOSE EQUIVALENT
1-131. Which ONE of the following
identifies
actions that are required by 1900 on 01/27/09 and the bases for the RCS Specific Activity limit? A. Reduce RCS Tavg below 500&deg;F to limit doses at the site boundary in the event of a LOCA in conjunction
with 0.25La leakage from containment. Reduce RCS Tavg below 500&deg;F to limit doses at the site boundary in the event of a SGTR in conjunction
with steady state SG tube leakage of 1 gpm. ( C. Reduce RCS Tavg below 350&deg;F to limit doses at the site boundary in the event of a LOCA in conjunction
with 0.25La leakage from containment.
D. Reduce RCS Tavg below 350&deg;F to limit doses at the site boundary in the event of a SGTR in conjunction
with steady state SG tube leakage of 1 gpm. Page 22 
( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Page 23 DISTRACTOR
ANAL YSIS: A. Incorrect, reducing Tavg below 500&deg;F is correct but the bases is not due to a LOCA with assumed containment
leakage. Plausible
because the action stated is correct and a LOCA with leakage from containment
could cause elevated doses at the site boundary.
B. CORRECT, with the activity above the 0.35 microcurieslgram
limit in the Tech Spec 3.4.8 for 48 continuous
hours, Tavg is required to be reduced to less than 500&deg; F within 6 hours in accordance
with the Tech Spec. The TIS bases states that reducing Tavg below 500&deg;F prevents the release of activity should a steam generator
tube rupture since the saturation
pressure of the primary coolant is below the lift pressure of the atmospheric
steam relief valves. The limit on activity is based on the resulting
2-hour doses at the site boundary not exceeding
a small fraction of the 10 CFR 100 limits following
a SGTR in conjunction
with an assumed steady state SG tube leak of 1 gpm. C. Incorrect, reducing Tavg below 350&deg;F is not correct and the bases is not due to a LOCA with assumed containment
leakage. Plausible
because lowering Tavg to 350&deg;F would mean changing to Mode 4 within the next 6 hours (which is a directed action in many TIS) and a LOCA with leakage from containment
could cause elevated doses at the site boundary.
D. Incorrect, reducing Tavg below 350&deg;F is not correct but the bases is being to limit doses in the event of a SGTR is correct. Plausible
because lowering Tavg to 3500F would mean changing to Mode 4 within the next 6 hours (which is a directed action in many TIS) and the bases is to limit doses at the site boundary during SGTR accident. 
( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 83 Tier 1 Group 2 KIA 076 AA2.04 Ability to determine
and interpret
the following
as they apply to the High Reactor Coolant Activity:
Corrective
actions required for high fission product activity in RCS Importance
Rating: 2.8 1 3.4 Technical
Reference:
Technical
Specifications
3.4.8 and Bases, Amendments
301 and 305. Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271AOP-R06
B.9 Question Source: Describe the Tech Spec and TRM actions applicable
during the performance
of AOP-R06. Bank# ___ _ Modified Bank # X'--__ Question History: New __ _ SQN questions
AOP-R06-B.2
001 and AOP-R06-B.9
001 combined and modified for SQN 1/2009 exam. Question Cognitive
Level: Memory or fundamental
knowledge
_X __ Comprehension
or Analysis __ _ 10 CFR Part 55 Content: ( 43.5 1 45.13 ) 10CFR55.43.b ( 2 ) Comments:
SQN questions
AOP-R06-B.2
002 and 0 AOP-R06-B.9
001 combined and modified.
Page 24 
( REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION
FOR OPERATION
3.4.8 The specific activity of the primary coolant shall be limited to: a. Less than or equal to 0.35 microcuries/gram
DOSE EQUIVALENT
1-131, and b. Less than or equal to 100/E microcuries/gram.
APPLICABILITY:
MODES 1, 2, 3, 4 and 5 ACTION: MODES 1, 2 and 3* a. With the specific activity of the primary coolant greater than 0.35 microcuries/gram
DOSE EQUIVALENT
1-131 for more than 48 hours during one continuous
time interval or exceeding
the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T avg less than 500&deg;F within 6 hours. LCO 3.0.4.c is applicable.
b. With the specific activity of the primary coolant greater than 100jE microcuries/gram, be in aUeast HOT STANDBY with Tavg less than 500&deg;F within 6 hours. MODES 1, 2, 3, 4 and 5 a. With the specific activity of the primary coolant greater than 0.35 microcuries/gram
DOSE EQUIVALENT
1-131 or greater thaif 100jE microcuriesfgram, perform the sampling and analysis requirements
of item 4a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. *With Tavg greater than or equal 500&deg;F. SEQUOYAH -UNIT 1 3/44-19 April 11 , 2005 Amendment
No. 36,117,237,301 
( REACTOR COOLANT SYSTEM BASES I 3/4A8 SPECIFIC ACTIVITY The limitations
on the specific activity of the primary coolant ensure that the resulting
2 hour doses at the site boundary will not exceed an appropriately
small fraction of Part 100 limits following
a steam generator
tube rupture accident in conjunction
with an assumed steady state primary-ta-secondary
steam generator
leakage rate of 1.0 GPM. The values for the limits on specific activity represent
interim limits based upon a parametric
evaluation
by the NRC of typical site locations.
These values are conservative
in that specific site parameters
of the Sequoyah Nuclear Plant site, such as site boundary
and meteorological
conditions, were not considered
in this evaluation
.. The ACTION statement
permitting
POWER OPERATION
to continue for limited time periods with the primary coolant's
specific activity greater than 0.35 microcuriesigram
DOSE EQUIVALENT
1-131, but within the allowable
limit shown on Figure 3.4-1, accommodates
possible iodine spiking phenomenon
which may occur following
changes in THERMAL POWER Operation
with specific activity levels exceeding
0.35 microcuriesigram
DOSE EQUIVALENT
1-131 but within the limits shown on Figure 3.4-1 should be limited to no more than 800 hours per year since the activity levels allpwed by Figure 3.4-1 increase the 2-hour thyroid dose at the site boundary by a factor of up to 20 following
a postulated
steam generator
tube rupture. A Note permits the use of the provisioris
of LCO 3.0.4.c. This allowance
permits entry into the applicable
MODE(S) while relying on the ACTIONS. This allowance
is acceptable
due to the significant
conservatism
incorporated
into the specific.
activity limit, the low probability
of an event which is limiting due to exceeding
this limit, and the ability to restore transient
specific activity excursions
while the plant remains at, or proceeds to power operations
.. , ., Reducing Tavg to less than 500&deg;F prevents the release of activity should a steam generator
tube rupture since the saturation
pressure of the primary coolant is below the lift pres5'ureof
the atmospheric
steam reliefvalves.
The surveillance
requirements
provide adequate assurance
that excessive
specific activity levels in the primary coolant will be detected insufficient
time to take corrective
action. Information
obtained on iodine spiking will be used to assess the parameters
associated
with spiking phenomefla.
A reduction
in frequency
of isotopic analyses following
power changes may be permissible
if justified
by the data obtained.
SEQUOYAH -UNIT 1 . December 28, 2005 Amendment
No. 117, 237. 301 305 
AOP-R.06-B.2
001 QUESTIONS
REPORT for BANK SQN Questions
Given the following
plant conditions:
Unit 1 has been at 90% power for 5 days . -At 1300 on 8/15/03 ReS Activity was determined
to be 0.5 microcuries/gram
DOSE EQUIVALENT
1-131 -At 1330 on 8/17/03 Chemistry
informs you RCS Activity has increased
to 55 microcuries/gram
DOSE EQUIVALENT
1-131. Which ONE (1) of the following
identifies
the required action for this condition?
A'! Immediately
initiate a plant shutdown and reduce RCS T avg below 500 0 F by 1900 on
B. Increase fre-quency
of ReS sampling and analysis for RCS activity to once every 2 hours. C. If RCS activity remains unchanged, at 1300 on 8/17/03 initiate a plant shutdown to HOT STANDBY. D. Reduce power level to 50% of rated and have Chemistry
re-sample
the RCS for specific activity.
KIA: 000076SG08
[2.8/3.5]
Reference:
Tech Spec LCO 3.4.8 Objective:
OPL271 C370, B.2 History: i\lew Level: fI.nalysis
Note: Provide a copy section 3/4.4.8 as an attachment
to the exam. Note: Make SHO only question, developed
replacement
for RO exam -PEH 8/8/97 Wednesday, September
03, 20088:33:05
AM 3 
( AOP-R.06-B.9
00 I QUESTIONS
REPORT for BANK SQN Questions
Which one of the following
describes
the basis for LCO 3.4.8, RCS Specific Activity?
Ensures that the resulting
doses at the site boundary will not exceed a small fraction of 10 CFR 100 limits following
a __ _ A. LOCA in conjunction
with 0.25 La leakage from containment.
B:t SGTR in conjunction
with steady state S/G tube leakage of 1 gpm. C. steam line break in conjunction
with steady state S/G tube leakage of 1 gpm. D. LOCA in conjunction
with 0.25 La leakage from secondary
containment
enclosure
boundary.
Justification:
A. Incorrect.
is the bases for LCO 3.6.1, Primary Containment
Integrity.
Plausible
because this is a steaming path that would affect dose rate at the site boundary.
B. Correct. C. Incorrect.
This is part of the assumptions (MSLB with S/G tube leakage) made for S/G Operational
Leakage. Plausible
because this is a steaming path that would affect dose rate at the site boundary.
D. Incorrect.
*(his is the bases for LCO 3.6.2, Secondary
Containment
Bypass Leakage. Plausible
because this is a steaming path that would affect dose rate at the site boundary.
Notes: KIA: Ref: LP/Obj: History: Level: Est Time: Comment: 076A1<3.05
O?13AK3.06
076(32.2.25
[2.9/3.6]
[3.2/3.8]
[2.5/3.7]
::PL271 AOP-R.06, Obj 9 9/07 -New min Wednesday, September
02 20n8 8:33:05 AM {41.5,41.10}
{41.5,41.10}
{41.5} 4 
( OPL271 AOP-R.06 Revision 0 Page 3 of 15 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-R.06 HIGH RCS ACTIVITY IV. LENGTH OF LESSON/COURSE:
1.0 hour(s) V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of AOP-R.06 HIGH RCS ACTIVITY.
B. Enabling Objectives:
O. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with High RCS Activity that are rated z 2.5 during Initial License Training and z 3.0 during License
Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of AOP-R.06.
2. Discuss the AOP-R.06 entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with AOP-R06 entry conditions.
b. Describe the ARP requirements
associated
with AOP-R06 entry conditions.
c. Interpret, prioritize, and verify associated
alarms are consistent
with AOP-R.06 entry conditions.
d. Describe the Administrative
conditions
that require Turbine Trip/ Reactor trip due to Reactor Coolant Pump Malfunctions.
3. Describe the initial operator response to stabilize
the plant upon entry into AOP-R06. 4. Upon entry into AOP-R06, diagnose the applicable
condition
and transition
to the appropriate
procedural
section for response.
5. Summarize
the mitigating
strategy for the failure that initiated
entry into AOP-R06. 6. Describe the bases for all limits, notes, cautions, and steps of AOP-R.06. 
7. 8. 9. 10. OPL271 AOP-R.06 Revision 0 Page 4 of 15 Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
Given a set of initial plant conditions
use AOP-R.06 to correctly:
a. Recognize
entry conditions
b. Identify required actions c. Respond to Contingencies
d. Observe and Interpret
Cautions and Notes Describe the Tech Spec and TRM actions applicable
during the performance
of AOP-R.06.
Apply GFE and system response concepts to the abnormal condition
-prior to, during and after the abnormal condition 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
84. WIEOl EG2.4.11 084 Given the following:
Unit 2 is operating
at 100% power when a loss of Train A CCS occurs. -The crew enters AOP-M.03, "Loss of Component
Cooling Water", and initiates
a Reactor trip. -As the crew is performing
the Immediate
Operator Actions of E-O, "Reactor Trip or Safety Injection", an automatic
Safety Injection
occurs. -The crew performs E-O to the last step without identifying
a transition.
Which ONE of the following
identifies
both the correct use of the Emergency
Procedure
and the proper crew action relative to the use of the Abnormal Operating
Procedure?
Emergency
Abnormal Operating
Instructions
Procedures
A. Loop back in E-O to AOP-M.03 can NOT re-perform
steps to be implemented
in identify a transition.
in parallel with E-O. B:' Loop back in E-O to AOP-M.03 can re-perform
steps to be implemented
in identify a transition.
parallel with E-O. C. Transition
to ES-O.O, AOP-M.03 can NOT be "Rediagnosis" ,and identify implemented
in parallel the proper transition.
with ES-O.O, "Rediagnosis".
D. Transition
to ES-O.O, AOP-M.03 can be
identify implemented
in parallel the proper transition.
with ES-O.O, "Rediagnosis".
Page 25 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRACTOR ANAL YSIS: Page 26 A. Incorrect, if E-O is performed
to the last step, the step will direct a loop back to step 8 so that a transition
can be identified
either to an accident procedure
or to the procedure
to terminate
the safety injection.
EPM-4, User's Guide does permit AOP performance
while in the EOP network with the conditions
stated in the stem. Plausible
because the loop back in E-O is correct and the EOPs do have priority over the AOPs except under certain conditions.
B. CORRECT, if a transition
is not made while E-O is performed
the last step will direct a loop back to step 8 so that a transition
can be identified
either to an accident procedure
or to the procedure
to terminate
the safety injection.
EPM-4, User's Guide, does permit the AOP performance
while in the EOP network with the conditions
stated in the stem.
C. Incorrect, while ES-O.O, Rediagnosis, can be used to determine
the correct procedure, it is not applicable
until a transition
is made from E-O. EPM-4, User's Guide does permit AOP performance
while in the EOP network with the conditions
stated in the stem. Plausible
because ES-O.O, Rediagnosis, can be used to determine
the correct procedure
under different
conditions
and the EOPs do have priority over the AOPs except under certain conditions.
D. Incorrect, while ES-O.O, Rediagnosis, can be used to determine
the correct procedure, it is not applicable
until a transition
is made from E-O. EPM-4, User's Guide does permit AOP performance
while in the EOP network with the conditions
stated in the stem. Plausible
because ES-O.O, Rediagnosis, can be used to determine
the correct procedure
under different
conditions
and EPM-4, User's Guide, does permit AOP performance
while in the EOP network with the conditions
stated in the stem. 
( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 84 Tier 1 Group 2 KIA WE01 EG2.4.11 Rediagnosis
Knowledge
of abnormal condition
procedures.
Importance
Rating: 4.0/4.2 Technical
Reference:
E-O, Reactor Trip or Safety Injection, Rev 30 EPM-4, User's Guide, Rev 20 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 EPM-4 B.7 & 8 Question Source: Given plant operating
conditions, determine
if EOP entry conditions
have been met and state the resultant
appropriate
immediate
action steps for those conditions.
Given plant operating
conditions, determine
if AOP entry conditions
have been met and state the resultant
appropriate
actions for those conditions.
OPL271 EPM-4 B.2.b Discuss the ES-O.O entry conditions.
Describe the requirements
associated
with ES-O.O entry conditions.
Bank# ___ _ Modified Bank # ___ _ New _X __ Question History: New question for Sequoyah 2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X __ 10 CFR Part 55 Content: (41.10/43.5/45.13 ) 10CFR55.43.b ( 5 ) Comments:
New question for Sequoyah 2009 exam Page 27 
SQN EOI EPM-4 PROGRAM USER'S GUIDE Rev. 20 MANUAL Page 63 of 97 3.11.4 Two-Column
Pr;ocedure
Walkthrough
Demonstration (Continued)
3.11.5 K. Step 7 directs the operator to "CHECK SI termination
criteria." 1. If all of the criteria in Substeps a through d are met, the operator remains in the AER column and transitions
to ES-1.1 to terminate
SI. 2. If any of the criteria in Substeps a through d are NOT met, the operator moves to the RNO column and proceeds to Step 8. L.
directs the operator to "GO TO E-1, Loss of Reactor or Secondary
Coolant." 1. This step ends performance
of procedure
E-2 with a transition
to E-1. 2. The highlighted
word "END" centered after the last step emphasizes
that the action steps for E-2 are complete.
Use of ES-O.O, Rediagnosis
A. ES-O.O, Rediagnosis, is unique among the EOPs in that it has no specific transition
into it. It is entered strictly based on operator judgment and is applicable
only if SI is in progress and E-O has already been performed (diagnostic
steps completed
and!.!:ansition
made to another procedure).
ES-O.O should be used when the operator has any concern that he may not be in the right EOP based on plant conditions.
This is most likely to happen if multiple accidents
occur either simultaneously
or sequentially.
B. Once entered, ES-O.O will either transition
the operator to ECA-2.1, E-1, E-2, or E-3, or will return him to the procedure
and step in effect, depending
on ,diagnostics
done within the procedure.
If ES-O.O determines
that an operator should be in a certain series of procedures (e.g., E-1 or ECA-1 series), and he is, then he simply returns to the procedure
and step in effect. If ES-O.O determines
that an operator should be in a certain series of procedures (e.g., E-3 or ECA-3 series), and he is NOT, then he is sent to either E-1 (if he should be in E-1 or ECA-1 series) or E-3 (if he should be in E-3 or ECA-3 series) to enter the appropriate
series at the beginning
and work his way through the series normally from that point on. 
SQN EOI EPM-4 PROGRAM USER'S GUIDE Rev. 20 MANUAL Page 65 of 97 3.11.7 Use of AOPs Within the EOP Network A. EOPs have priority over AOPs at all times, except when a reactor trip or safety injection
has occurred in conjunction
with an Appendix R fire N.08), Control Room abandonment(AOP-C.04), or Loss of all ERCW capability (AOP-M.01).
B. AOP performance
while in the EOP network is allowable
under the following
two circumstances:
[C.1] 1. AOPperformance
is directed by EOPs in effect. 2. AOP performance
is deemed necessary
by the SM or US to address abnormal plant conditions
NOT directly addressed
by the EOPs but which have a significant
impact on the ability of the EOPs to perform their function (e;g., loss of ERCW, CCS, off-site power, vital instrument
power board, etc.) In this case, the following
guidelines
should be followed:
a. Concurrent
performance
of the EOPs and the AOP should enhance, NOT degrade, the performance
of EOPs in progress .. b. Manpower reSources
are adequate to allow performing
the EOPs and the AOP concurrently.
c.* The AOP should be performed
using the single perfomer method so the procedure
reader remains dedicated
to the EOPs in progress, which are mitigative
in nature. The SM may elect to deviate from this requirement
when in ES-O.1. d. Certain AOPs may be required to be performed
concurrently
with the EOPs in order for the EOPs to function as intended;
for example, loss of CCS, loss of ERCW, loss of air or vital power to equipment
important
to safety--any of these could have a significant
impact on the ability of the EOPs to achieve their goals. e. Upon transition
to ES-O.1, the SM will designate
the mitigating
crew responsibilities
as appropriate, based on the events in progress.
Normally, the procedure
reader and OATC will perform ES-O.1 while the CRO performs the AOP using the single perfomer method. 
-E-O SON REACTOR TRIP OR SAFETY INJECTION
Rev. 30 .. I STEP I I ACTION/EXPECTED
RESPONSE 25.' DETERMINE
if diesel generators
should be stopped: a. VERIFY shutdown boards ENERGIZED
from start busses. b. STOP any unloaded diesel generators
and PLACE in standby USING EA-82-1, Placing DIGs in Standby. 26. GO TO Step 8. END II
______ a. ATTEMPT to restore offsite
power to shutdown boards USING EA-202-1, Restoring
Off-Site Power to 6900 V Shutdown Boards. Page 19 of 21 
( SQN LOSS OF COMPONENT
COOLING WATER AOP-M.03 Rev. 11 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.3 Train A CCS Header Failure CAUTION: During operation, the Containment
Spray Pumps may experience
bearing failure after 10 minutes of a loss of CCS cooling. NOTE 1: When the associated
TRAIN of CCS is out of service the CCPs, SI Pumps, and RHR Pumps are INOPERABLE
for ECCS purposes due to not being able to fulfill it's design function for sump recirculation.
LCOs 3.5.2, 3.5.3, 3.6.2.1,3.7.3
should be evaluated
and appropriately
entered by the SRO. NOTE 2: When CCS is out of service to mechanical
seal HXs ONLY, the affected CCPs, SI Pumps, and RHR Pumps have been evaluated
to be OPERABLE and AVAILABLE.
These pumps can run indefinitely
without CCS cooling water to mechanical
seal HXs (Ref: PER 72528, DCN Q-11452-A, and RIMS B38941123
802). 1. MONITOR REACTOR COOLANT PUMPS MOTOR THRUST BEARING TEMP HIGH annunciator
DARK [M-5B, E-3]. PERFORM the following
on the affected Unit: a. IF affected unit in Mode 1 or 2, THEN PERFORM the following:
1) TRIP reactor. 2) STOP RCPs. 3) GO TO E-O, Reactor Trip or Safety Injection, WHILE continuing
in this procedure.
[C.2] b. ENSURE RCPs are TRIPPED. c. IF in Mode 4, 5, or 6, THEN STABILIZE
RCS temperature
USING RHR shutdown cooling. Page 14 of 64 
PROGRAM: OPL271 EPM-4 Revision 1 Page 3 of 26 OPERATOR TRAINING -LICENSED I. COURSE: LICENSE TRAINING II. LESSON TITLE: III. LENGTH OF LESSON/COURSE:
4-6 hour(s) IV. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of HLC Procedures
training, the participant
shall be able to explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of EMP-4, EOP-E-O, "User's Guide". B. Enabling Objectives:
1. Determinelidentify
the correct procedural
application(s)
based on the operating
procedures
network for normal, abnormal, and emergency
evolutions.
2. Analyze an EOP layout and determine (according
to EPM-4): a. correct procedural
layout application;
b. if the use of terms is correct (e.g.: Faulted Steam Generator, Shall, Lowering, etc per Appx. B); c. correct use of symbols and icons. 3. Define EOP warnings, cautions, and notes and, given an EOP condition, determine
appropriate
usage. 4. Compare and contrast event-based
emergency/abnormal
operating
procedures
used in parallel with the symptom-based
EOPs. 5. Given an example, apply general guidelines, crew roles and responsibilities
for EOP procedural
use and determine:
a. format and use of sequenced
and non-sequenced
sub steps; b. transition
between Action/Expected
Response column and the Response Not Obtained column; c. requirements
for task completion
prior to proceeding
to the next action (and how any exceptions
are identified);
d. requirements
for task completion
still in progress following
transition
to another procedure
or step; e. actions based on fold-out page use; f. actions based on hand-out page use; g. if EOP termination
is appropriate
based on given conditions.
6. Identify post-accident
instrumentation
and determine
if its use is required.
7. Given plant operating
conditions, determine
if EOP entry conditions
have been met and state the resultant
appropriate
immediate
action steps for those conditions. 
OPL271 EPM-4 Revision 1 Page 4 of 26 8. Given plant operating
conditions, determine
if AOP entry conditions
have been met and state the resultant
appropriate
actions for those conditions.
9. Identify general operating
crew responsibilities
during emergency
operations
including
appropriate
implementation
of prudent operator actions. 10. Identify general operating
crew responsibilities
during emergency
operations
including
requirements
for actions outside Technical
Specifications/plant
licensed conditions
(1 OCFR50.54x
application).
11. Given a set of conditions, analyze the EOP/FRP implementation:
a. identify the basis for the implementation;
b. determine
the correct implementation
hierarchy;
c. determine
if Critical Safety Function Status Trees (CFSTs) implementation
is required;
d. identify the status tree colors by priority and summarize
each tree's purpose; e. identify conditions
which will allow a FRP to be exited once it is entered (a RED or ORANGE condition);
f. state the monitoring
frequency
of CFSTs and when this can be relaxed; g. determine
correct coordination
with other support procedures
h. identify conditions
permissible
to terminate
CFSTs monitoring.
12. Given an operational
Situation, analyze a crew brief and determine
if it meets Management
expectations. 
II OPL271 ES-O.O Revision 0 Page 3 of 15 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: ES-O.O, "Rediagnosis" IV. LENGTH OF LESSON/COURSE:
.5 hour(s) V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of this lesson and others presented, the student shall demonstrate
an understanding
of the EOP-ES-O.O, "Rediagnosis" by successfully
completing
a written examination
with a score of 80 percent or greater. B. Enabling Objectives
O. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with ES-O.O, Rediagnosis, that are rated 2 2.5 during Initial License Training and 2 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of ES-O.O. 2. Discuss the ES-O.O entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with ES-O.O entry conditions.
b. Describe the requirements
associated
with ES-O.O entry conditions.
3. Summarize
the mitigating
strategy for the failure that initiated
entry into ES-O.O. 4. Describe the bases for all limits, notes, cautions, and steps of ES-O.O. 5. Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
6. Given a set of initial plant conditions
use ES-O.O to correctly:
a. Identify required actions b. Respond to Contingencies
c. Observe and Interpret
Cautions and Notes 7. Apply GFE and system response concepts to the performance
of ES-O.O conditions.
II 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
85. W/E16 EA2.1 085 Given the following:
Unit 1 is at 100% power. -A LOCA occurred inside containment.
The crew has just implemented
E-1, "Loss of Reactor or Secondary
Coolant".
The STA has completed
the initial performance
of the status trees and reports the highest priority path exists on the CONTAINMENT
status tree. Containment
conditions
are as follows: Pressure is 2.6 psig and lowering.
Upper containment
Rad Monitors read 85 Rlhr. Lower containment
Rad Monitors read 125 Rlhr . Containment
Sump Level is 58%. Based on the above conditions, the Unit Supervisor
... A. is required to IMMEDIATELY
implement
and complete FR-Z.2, "Containment
Flooding", then transition
back to E-1. B. will acknowledge
entry criteria for FR-Z.2, "Containment
Flooding", is met but entry into the FR is optional.
C. is required to IMMEDIATELY
implement
and complete FR-Z.3, "High Containment
Radiation", then transition
back to E-1. will acknowledge
entry criteria for FR-Z.3, "High Containment
Radiation", is met but entry into the FR is optional.
Page 28 
( ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRA CTOR ANAL YSIS: Page 29 A. Incorrect, the level is the sump is below the 68% level required to enter FR-Z. 2, Containment
Flooding but if the level was high enough for entry the transition
would be required.
Plausible
because the containment
sump level is elevated and if the required level was present an Orange would be present and require immediate
transition
to FR-Z.2. B. Incorrect, the level is the sump is below the 68% level required to enter FR-Z. 2, Containment
Flooding but if the level was high enough for entry the transition
would not be optional.
Plausible
because the containment
sump level is elevated and if the required level being met resulted in a yellow path, the entry would be optional.
C. Incorrect, the radiation
in lower containment
is greater than the threshold
level for entering FR-Z.3, High Containment
Radiation, but the challenge
is a Yellow path which allows the performance
of the procedure
to be optional.
Only Red and Orange path challenges
are required to be immediately
implemented.
Plausible
because the entry conditions
are met and if the challenge
had been an Orange path, immediate
transition
would be required.
D. CORRECT, the radiation
in lower containment
is greater than the threshold
level for entering FR-Z.3, High Containment
Radiation, and the challenge
is a yellow path which allows the performance
of the procedure
to be optional. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 85 Tier 1 Group 2 KIA W/E16 EA2.1 High Containment
Radiation
Ability to determine
and interpret
the following
as they apply to the (High Containment
Radiation)
Facility conditions
and selection
of appropriate
procedures
during abnormal and emergency
operations.
Importance
Rating: 2.9 I 3.3 Technical
Reference:
EPM-4, User's Guide, Rev. 20 1-FR-0, Status trees, Rev. 1 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 EPM-4 B.11 Question Source: Given a set of conditions, analyze the EOP/FRP implementation.
OPL271 FR-O B.6 Given a set of initial plant conditions
use FR-O to correctly
identify the: a. Identify required actions. Bank# ____ _ Modified Bank # X, __ _ New ___ _ Question History: Modified from Braidwood
12-2007 SRO exam 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:
Modified from Braidwood
12-2007 SRO exam Page 30 
( I SON EOI PROGRAM MANUAL I USER'S GUIDE EPM-4 Rev. 20 Page 48 of 97 3.10.5 Status Tree Rules of Usage (continued)
6. If any ORANGE challenge
is encountered, the person monitoring
status trees continues
monitoring
until all six status trees have been evaluated.
This is necessary
because a subsequent
RED challenge
has priority over any ORANGE challenge.
If any RED is encountered, then Rule 3.10.5.0.4
applies. Otherwise, once it is determined
that no RED challenges
exist, then the person monitoring
status trees informs the procedure
reader of the highest priority ORANGE challenge.
7. RED or ORANGE challenges
must be addressed
immediately
by implementing
appropriate
FRPs in order of priority and per the rules of usage. When the person monitoring
status trees informs the procedure
reader that a RED or ORANGE challenge
exists, the procedure
reader immediately
suspends the ORP (or lower priority FRP) in progress and implements
the appropriate
FRP, as indicated
at the terminus point of the CSF under challenge.
8. YELLOW challenges
may be addressed
by implementing
appropriate
FRPs if desired, but do not require immediate
operator action. Addressing
YELLOW challenges
is optional since these are usually temporary, off-normal
conditions
that will be restored to normal status by actions already in progress.
In other cases, the YELLOW path might provide an early indication
of a developing
RED or ORANGE condition.
Following
FRP implementation, a YELLOW might indicate a residual normal condition.
When the person monitoring
status trees informs the procedure
reader that a YELLOW challenge
exists, the procedure
reader should evaluate if the YELLOW challenge
FRP should be implemented.
This decision will be based on the following:
* Whether the procedures
in effect will address the challenge
as a matter of course. * Whether the procedures
in effect are more important
at that time based upon available
time and current plant conditions.
* Whether the challenge
is of a nature that it will likely develop into an ORANGE or RED condition
if action is not taken early. 
CONTAINMENT
PRESSURE NO LESS THAN 12.0 PSIG YES CONTAINMENT
PRESSURE LESS THAN 2.8 PSIG CONTAINMENT
F-O.5 SQN 1-FR-O Rev. 1 R GOTO , , FR-Z.1 NO Ii!r'Ji,UI
"'"
:GOTO d FR-Z.1 YES NO Immlli CONTAINMENT
SUMP LEVEL LESS THAN 68% YES UPPER AND LOWER CONTAINMENT
RADIATION
MONITORS LESS THAN 100 RlHR Page 10 of 16
I'b'!lWi:mitI GOTO
FR-Z.2 NO F=======I
y I
YES IIil Iii {oi. il1il CSF SAT 
Braidwood
12-2007 exam Change to LOCA Quest No: RO SRO: TIER: GROUP: Topic No: KA No: RO: SRO: Cog Level: 83 SRO 1 2 00WE14 00WE14EA2.1
3.33.8 High System/Evolution
Name: Category Statement:
High Containment
Pressure Ability to determine
and interpret
the following
as they apply to the High Containment
Pressure:
KA Statement:
Facility conditions
and selection
of appropriate
procedures
during abnormal and emergency
operations
UserID: Topic Line: Question Stem: Given: -Unit 1 was at 100% power. -All systems were normally aligned. -A large steam break occurred inside containment.
-The crew has just implemented
1BwEP-2, FAULTED STEAM GENERATOR
ISOLATION.
-The STA has completed
the initial scan of the status trees and the following
conditions
exist: -An ORANGE path exists on the containment
status tree. -Containment
pressure is 26 psig and lowering.
-The faulted SG has NOT been isolated.
-ALL other CSFs are GREEN. Based on the above conditions, the Unit Supervisor
will direct the crew to ... A IMMEDIATELY
implement
and complete 1BwFR-Z.1, RESPONSE TO HIGH CONTAINMENT
PRESSURE, THEN transition
back to 1BwEP-2. B remain in 1BwEP-2 and continue to monitor containment
pressure.
If containment
pressure begins to rise, THEN implement
1BwFR-Z.1, RESPONSE TO HIGH CONTAINMENT
PRESSURE.
C remain in 1BwEP-2 until the faulted SG is isolated, THEN implement
1BwFR-Z.1, RESPONSE TO HIGH CONTAINMENT
PRESSURE.
D IMMEDIATELY
implement
1BwFR-Z.1, RESPONSE TO HIGH CONTAINMENT
PRESSURE, and remain in 1BwFR-Z.1
until the containment
CSF is restored to GREEN OR YELLOW, THEN transition
back to 1BwEP-2. 
Answer: Task No: S-FR-01S Question Source: Question Difficulty
A Obj No: 7D.FR-00SA
New Low Time: Cross Ref: 10CFRSS.43(b)(S)
1 Reference:
No reference
will be provided to examinee.
ILT lesson plan Il-FR-XL-01, BwFR-Z BwAP 340-1, Use of Procedures
for the Operating
Department
1BwFR-Z.1, Response to High Containment
Pressure Explanation:
Question meets KA. Question requires examinee ability to determine
and interpret
facility conditions
and select appropriate
procedures
during high containment
pressure.
With containment
pressure>
20 psig, an orange path conditions
exist on the containment
status tree. With an orange path present, immediate
transition
is made to 1BwFR-Z.1
to restore the CSF. Once entered, 1BwFR-Z.1
is entirely completed, then transition
is made back to procedure
and step in effect at time of transition
to 1BwFR-Z.1.
A is correct, see explanation
above. B is incorrect, even though containment
pressure is lowering and nearing point at which CSF will change to yellow status, transition
is made to 1BwFR-Z.1.
C is incorrect, 1BwFR-Z.1
will perform faulted SG isolation
sequence.
D is incorrect, 1BwFR-Z.1
is completed
and procedure
is exited even if CSF is not restored.
Date Written: 6/28/2007
Author: Darren Stiles ill App. Ref: none 
OPL271FR-O
Revision 1 Page 3 of 42 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: FR-O, STATUS TREES IV. LENGTH OF LESSON/COURSE:
1 hours V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of FR-O Status Trees. B. Enabling Objectives
O. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with Status Trees that are rated 22.5 during Initial License Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of FR-O. 2. Explain the bases for prioritizing
safety functions
during emergency
operations.
3. Summarize
the mitigating
strategy for the failure that initiated
entry into FR-O 4. Discuss requirements
for monitoring
Status Trees. a. Describe the conditions
when monitoring
is to be initiated . b. Describe the required frequency
for monitoring
the status trees and how the frequency
is determined
c. Describe the conditions
when monitoring
can be terminated.
5. Describe the bases for all decision blocks, limits, notes, cautions, and steps of FR-O. 6. Given a set of initial plant conditions
use FR-O to correctly
identify the : a. Identify required actions b. Observe and Interpret
Cautions and Notes c. Requirements
when a RED or ORANGE Path is diagnosed
7. Apply GFE and system response concepts
to the performance
of FR-O 
PROGRAM: OPL271 EPM-4 Revision 1 Page 3 of 26 OPERATOR TRAINING -LICENSED I. COURSE: LICENSE TRAINING II. LESSON TITLE: III. LENGTH OF LESSON/COURSE:
4-6 hour(s) IV. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of HLC Procedures
training, the participant
shall be able to explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of EMP-4, EOP-E-O, "User's Guide". B. Enabling Objectives:
1. Determine/identify
the correct procedural
application(s)
based on the operating
procedures
network for normal, abnormal, and emergency
evolutions.
2. Analyze an EOP layout and determine (according
to EPM-4): a. correct procedural
layout application;
b. if the use of terms is correct (e.g.: Faulted Steam Generator, Shall, Lowering, etc per Appx. B); c. correct use of symbols and icons. 3. Define EOP warnings, cautions, and notes and, given an EOP condition, determine
appropriate
usage. 4. Compare and contrast event-based
emergency/abnormal
operating
procedures
used in parallel with the symptom-based
EOPs. 5. Given an example, apply general guidelines, crew roles and responsibilities
for EOP procedural
use and determine:
a. format and use of sequenced
and non-sequenced
sub steps; b. transition
between Action/Expected
Response column and the Response Not Obtained column; c. requirements
for task completion
prior to proceeding
to the next action (and how any exceptions
are identified);
d. requirements
for task completion
still in progress following
transition
to another procedure
or step; e. actions based on fold-out page use; f. actions based on hand-out page use; g. if EOP termination
is appropriate
based on given conditions.
6. Identify post-accident
instrumentation
and determine
if its use is required.
7. Given plant operating
conditions, determine
if EOP entry conditions
have been met and state the resultant
appropriate
immediate
action steps for those conditions. 
OPL271 EPM-4 Revision 1 Page 4 of 26 8. Given plant operating
conditions, determine
if AOP entry conditions
have been met and state the resultant
appropriate
actions for those conditions.
9. Identify general operating
crew responsibilities
during emergency
operations
including
appropriate
implementation
of prudent operator actions. 10. Identify general operating
crew responsibilities
during emergency
operations
including
requirements
for actions outside Technical
Specifications/plant
licensed conditions
(1 OCFR50.54x
application).
11. Given a set of conditions, analyze the EOP/FRP implementation:
a. identify the basis for the implementation;
b.
correctimplementatio
rl hierarchy;
c. determine
if Critical Safety Function Status Trees (CFSTs) implementation
is required;
d. identify the status tree colors by priority and summarize
each tree's purpose; e. identify conditions
which will allow a FRP to be exited once it is entered (a RED or ORANGE condition);
f. state the monitoring
frequency
of CFSTs and when this can be relaxed; g. determine
correct coordination
with other support procedures
h. identify conditions
permissible
to terminate
CFSTs monitoring.
12. Given an operational
situation, analyze a crew brief and determine
if it meets Management
expectations. 
( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
86. 003 A2.03 086 Given the following:
Unit 1 in Mode 3 with RCS at normal operating
temperature
and pressure.
RCPs #1, #2 and #3 are running. Following
the start of RCP #4, the pump stabilizes
as follows: -Motor current at 625 amps. -#1 sealleakoff
at 1.1 gpm. -All pump and motor temperatures
stable and within limits. -The operators
implement
AOP-R.04, "Reactor Coolant Pump Malfunctions".
Which ONE of the following
identifies
why AOP-R.04 entry was required and the action directed by the AOP? A'I The AOP was entered due to the high motor current and the AOP will direct removal of the RCP from service. B. The AOP was entered due to the high motor current but the AOP will NOT direct removal of the RCP unless bearing or stator temperatures
are increasing.
C. The AOP was entered due to the low #1 seal leakoff flow and the AOP will direct removal of the RCP from service. D. The AOP was entered due to the low #1 sealleakoff
flow but the AOP will NOT direct removal of the RCP unless lower bearing or seal water temperatures
are increasing.
Page 31 
( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRA CTOR ANAL YSIS: Page 32 A. CORRECT, the AOP entry was required due to the high current flow to the motor and the with the current flow greater than 608 amps, the procedure
directs the RCP to be stopped. B. Incorrect, the AOP entry being required due to the high current flow to the motor is correct but with the current flow greater than 608 amps, the procedure
does not require the bearing or stator temperatures
to be increasing
to direct the RCP to be stopped. Plausible
because the high motor current is why the procedure
entry was required and increasing
temperatures
do result in stopping the RCP with other conditions
in the procedure.
C. Incorrect, the AOP entry was not required due to low #1 sealleakoff
flow as the leakoff is above the low flow alarm, and if the flow was low with the bearing and seal water temperature
stable, the AOP would not require the pump be stopped unless the #1 sealleakoff
flow was further reduced. Plausible
because low seal leakoff flow would require the procedure
to be entered and if the flow were low enough the stopping of the RCP would be required.
D. Incorrect, the AOP entry was not required due to low #1 sealleakoff
flow but if the seal was less than entry conditions, rising temperature
would result in stopping the RCP. Plausible
because low sealleakoff
flow would require the procedure
to be entered and if the lower bearing or seal water temperatures
were increasing, the stopping of the RCP would be required. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 86 Tier 2 Group 1 KIA 003 A2.03 Ability to (a) predict the impacts of the following
malfunctions
or operations
on the RCPS; and (b) based on those predictions, use procedures
to correct, control, or mitigate the consequences
of those malfunctions
or operations:
Problems associated
with RCP motors, including
faulty motors and current, and winding and bearing temperature
problems Importance
Rating: 2.7/3.1 Technical
Reference:
AOP-R.04, Reactor Coolant Pump Malfunctions, Rev. 23 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271AOP-R.04
B.8.a & b Given a set of plant conditions
use AOP-R.04 to correctly:
a. Recognize
entry conditions
b. Identify required actions Question Source: Bank# ___ _ Modified Bank # ____ _ New _X __ Question History: New question for SQN 1/2009 exam 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:
New question for SQN 1/2009 exam Page 33 
 
__ ---'-_A_O_P_-R_._04---1
L Rev. 23 [ STEP I ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS CAUTION: Exceeding
the following
limitations
requires trip of the affected RCP, unless RCP operation
is required by FR-C.1, Inadequate
Core Cooling or FR-C.2, Degraded Core Cooling: * RCP #1 Seal
than 220 psid * Rep #1 Seal Temperature
greater than 225&deg;F * RCP Lower Bearing Temperature
greater than 225&deg;F * RCP Upper Motor Bearing Temperature
greater than 200&deg;F * RCP Lower Motor Bearing Temperature
greater than 200&deg;F * RCP Motor Voltage less than 5940V or greater than 7260V * RCP Motor Amps greater than 608 amps * RCP Vibration
greater than 20 mils on any axis (x and/or y) [C.3] NOTE 1: During plant startup following
seal maintenance, the seal package should seat and operate normally following
24 hours of run time. NOTE 2: RCP trip criteria is also located in Appendix B. This appendix should be referred to throughout
the performance
of this procedure.
1. DIAGNOSE the failure: IF ... Reactor Coolant Pump(s) tripped or shutdown required RCP #1 Seal Leakoff high flow (high flow Alarm) RCP #1 Seal Leakoff low flow (low flow Alarm) RCP #2 Seal Leakoff high flow (high RCP standpipe
level) RCP #3 Seal Leakoff high flow (low RCP standpipe
level) RCP Motor Stator Temperature
High Page 3 of 34 GOTO SECTION 2.1 2.2 2.3 2.4 2.5 2.6 PAGE 4 7 13 18 21 24 /' rr
e 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 Reactor Coolant Pump Tripped or Shutdown Required CAUTION: A rapid drop in level and steam flow on the affected loop S/G may occur when RCP is tripped. 1. CHECKunit
in Mode 1 or 2. GO TO Step 3. NOTE: This procedure
is intended to be performed
concurrently
with E-O, Reactor Trip or Safety Injection.
2. TRIP the reactor, and GO TO E-O, Reactor Trip or Safety Injection, WHILE continuing
in this procedure.
---.---3. STOP and LOCK OUT affected RCP(s). Page 4 of 34 
( ( SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.1 Reactor Coolant Pump Tripped or Shutdown Required (cont'd) CAUTION: If the RCP seal return flow control valve (FCV) is NOT closed within 5 minutes of stopping the RCP with excessive
leakoff, seal damage may occur. [C.2] 4. MONITOR Rep seal leakoff less than 8 gpm per pump: * FR-62-24 [RCP 1 & 2] * FR-62-50 [RCP 3 & 4] 5. PULL TO DEFEAT affected loop f.. T and T-avg: * XS-68-2D(f..
T) * XS-68-2M (T-avg) 6. CHECK RCPs 1 and 2 RUNNING. WHEN the RCP has coasted down (30 sec.), THEN CLOSE affected RCP seal return FCV: [C.2] * FCV-62-9 [RCP 1] * FCV-62-22
[RCP 2] * FCV-62-35
[RCP 3] * FCV-62-48
[RCP 4] CLOSE affected loop's pressurizer
spray valve. Page 5 of 34 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE I* RESPONSE NOT OBTAINED 2.1 Reactor Coolant Pump Tripped or Shutdown Required (cont'd) CAUTION: Restoring
seal water injection
to a hot seal package could result in failure of the RCP seals. NOTE: The plant should be cooled down to reduce heat input into the pump seal package if RCP seal injection
flow has been lost and cannot be restored prior to exceeding
temperature
limits. 7. IF RCP Seal Temperatures
or Bearing Temperatures
are increasing
uncontrolled
due to loss of Seal Injection, THEN EVALUATE initiating
RCS cooldown.
8. EVALUATE EPIP-1, Emergency
Plan Initiating
Conditions
Matrix. 9. EVALUATE the following
Tech Specs for applicability:
* 3.2.5, DNB Parameters
* 3.4.1.1, Reactor Coolant Loops and Coolant Circulation
-Startup and Power Operation
* 3.4.1.2, Reactor Coolant Hot Standby * 3.4.1.3, Reactor Coolant System -Shutdown * 3.4.6.2, RCS Operational
Leakage 10. GO TO appropriate
plant procedure.
---.---END OF SECTION , Page 6 of 34 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.2 RCP #1 Seal Leakoff High Flow * CAUTION: RCP bearing damage may occur if temperature
exceeds 225&deg;F . * CAUTION: If the RCP seal return flow control valve is NOT closed within 5 minutes of stopping the RCP with excessive
leakoff, seal damage may occur. [C.2] 1. MONITOR #1 seal leakoff less than 6 gpm per pump: * FR-62-24 [Rep 1 & 2] * FR*62**50
[Rep 3 & 4] a. MONITOR Rep lower bearing temperature
and seal temperature.
IF Rep lower bearing temperature
OR seal temperature
are rising uncontrolled, THEN GO TO Section 2.1, Rep Tripped or Shutdown Required.
[C.1] [C.2] IF lower bearing temperature
AND seal temperature
indication
are NOT available
for affected Rep, THEN GO TO Section 2.1, Rep Tripped or Shutdown Required.
[C.1] (Step continued
on next page.) Page 7 of 34 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.2 RCP #1 Seal Leakoff High Flow (cont'd) 1. (Continued)
b. CHECK #1 seal leakoff flow: IF #1 seal leakoff flow greater than 8 gpm, THEN PERFORM the following:
1) INITIATE plant shutdown at 2-4% per minute USING AOP-C.03, Rapid Shutdown or Load Reduction.
2) WHEN reactor is tripped, THEN GO TO Section 2.1, RCP Tripped or Shutdown Required.
[C.1] IF #1 seal leakoff flow less than 8 gpm, THEN PERFORM the following:
1) CONTROL RCP seal injection
flow for the affected RCP greater than or equal to 9 gpm. 2) CONTACT Engineering
for recommendations
WHILE continuing
with this procedure. (Step continued
on next page.) Page 8 of 34 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXP.ECTED
RESPONSE 2.2 RCP #1 Seal Leakoff High Flow (cont'd) 1. (Continued) . 2. MONITOR RCP lower bearing and seal water temperatures
less than 225&deg;F. Page 9 of 34 RESPONSE NOT OBTAINED 3) IMPLEMENT
Engineering
recommendations
to address specific RCP seal performance
conditions.
OR COMPLETE normal plant shutdown within 8 hours USING appropriate
plant procedure.
4) WHEN reactor is shutdown or tripped, THEN GO TO Section 2.1, RCP Tripped or Shutdown Required.
[C.1] ____ ::e--IF any of the following
conditions
met: * RCP lower bearing temperature
or seal water temperature
greater than 225&deg;F OR * seal leakoff flow greater than 6 gpm AND lower bearing and seal temp NOT available
for affected RCP THEN GO TO Section 2.1, RCP Tripped or Shutdown Required.
[C.1] __ ..... ::e--
( SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE 2.2 RCP #1 Seal Leakoff High Flow (cont'd) 3. MONITOR RCP #1 seal greater than 220 psid: * POI-62-8A
* POI-62-21A
* POI-62-34A
* PDI-62-47A
4. ENSURE RCP seal water supply flow 6-10 gpm per pump: * FI-62-1A * FI-62-14A
* FI-62-27A
* FI-62-40A
5. CONTACT Engineering
for recommendations
WH I LE continuing
with this procedure.
6. EVALUATE EPIP-1, Emergency
Plan Initiating
Conditions
Matrix. RESPONSE NOT OBTAINED GO TO Section 2.1, RCP Tripped or Shutdown Required.
[C.1] IF seal water supply flow is less than 6 gpm AND CANNOT be restored, THEN ENSURE CCS supply to thermal barriers less than 105&deg;F on TR-70-161
[CCS HX 1A1/1A2 (2A1/2A2)
Outlet Temp] Page 10 of 34 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE 2.2 RCP #1 Seal Leakoff High Flow (cont'd) 7. EVALUATE the following
Tech Specs for applicability:
* 3.2.5, DNB Parameters
* 3.4.1.1, Reactor Coolant Loops and Coolant Circulation
-Startup and Power Operation
* 3.4.1.2, Reactor Coolant System -Hot Standby * 3.4.1.3, Reactor Coolant System -Shutdown * 3.4.6.2, RCS Operational
Leakage RESPONSE NOT OBTAINED CAUTION: Slow and uniform temperature
adjustments (approx. 50&deg;F in one hour) will prevent thermal shock to the seals. 8. CHECK VCT outlet temperature
less than 130&deg;F [TI-62-131].
9. ENSURE VCT pressure between 17 psig and 45 psig [PI-62-122].
ADJUST HIG-62 .. 78A to reduce VCT temperature
to less than 130&deg;F. Page 11 of 34 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE 2.2 Rep #1 Seal Leakoff High Flow (cont'd) 10. CHECK RCPlower bearing and seal water temperature
less than 180&deg;F: 11. GO TO appropriate
plant procedure.
RESPONSE NOT OBTAINED IF any of the following
conditions
met: * affected RCP lower bearing or seal water temperature
greater than 180&deg;F OR * lower bearing and seal water temp indication
NOT available
for affected RCP, THEN GO TO Step 1. END OF SECTION Page 12 of 34 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE 2.3 RCP #1 Seal Leakoff Low Flow 1. CHECK #1 sealleakoff
flow greater than 0.8 gpm per pump: * FR-62-23 [RCP 1 & 2] * FR-62-49 [RCP 3 & 4] 2. CHECK WI seal leakoff flow greater than 0.9 gpm per pump and NOT decreasing:
* FR-62-23 [RCP 1 & 2] * FR-62-49 [RCP 3 & 4] 3. GO TO appropriate
plant procedure.
4. ENSURE RCP seal water supply flow between 6 gpm and 10 gpm per pump: * FI-'62-1A
* FI-62-14A
* FI-62-27A
* FI-62-40A
GO TO Step 4. GO TO Step 4. IF seal water supply flow is less than 6gpm AND CANNOT be restored, THEN ENSURE CCS supply to thermal barriers is less than 1 05&deg;F on TR-70-161.
[CCS HX 1A1/1A2 (2A1/2A2)
Outlet Temp] Page 13 of 34 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE 2.3 RCP #1 Seal Leakoff Low Flow (cont'd) 5. CONTACT Engineering
for recommendations
WHILE continuing
with this procedure.
6. ENSURE VCT pressure between 17 psig and 45 psig [PI-62-122J.
7. CHECK RCP standpipe
level alarms DARK [M-5B, A-2, B-2, C-2, 0-2]. RESPONSE NOT OBTAINED MONITOR the following:
a. RCDT parameters (O-L-2 AB, el. 669) * Level, U-77-1 * Pressure, PI-77-2 * Temperature, TI-77-21 b. Cntmt FI. & Sump Level rate of rise (ICS pt. U0969) Page 14 of 34 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE 2.3 RCP #1 Seal Leakoff Low Flow (cont'd) 8. VERIFY RCP #2 sealleakoff
less than or equal to 0.5 gpm USING Appendix A, RCDT Level Rate-of-Change.
9. MONITOR RCP lower bearing temperature
and seal water temperature
are stable and within limits (less than 225&deg;F). RESPONSE NOT OBTAINED GO TO Section 2.4, RCP #2 Seal Leakoff High Flow. IF any of the following
conditions
met: * affected RCP lower bearing temp or seal water temp rising uncontrolled
OR * affected RCP lower bearing temp or seal water temp greater than 225&deg;F OR * affected RCP lower bearing temp and seal temp indication
NOT available
THEN GO TO Section 2.1, RCP Tripped or Shutdown Required.
[C.1] Page 15 of 34 
SQN REACTOR COOLANT PUMP MALFUNCTIONS
AOP-R.04 Rev. 23 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.3 RCP #1 Seal Leakoff Low Flow (cont'd) CAUTION: If low seal leakoff compensatory
actions are NOT successful, seal failure may result as indicated
bya sudden increase in seal leakoff flow (greater than 8 gpm). NOTE: Plant shutdown may be terminated
if Seal Leakoff flow stabilizes
at greater than 0.8 gpm with pump Lower Bearing temperature
and Seal Water Temperature
remaining
stable (no indications
of seal failure).
10. MONITOR Rep #1 seal leakoff flow greater than 0.8 gpm: * FR-62-23 [Rep 1 & 2] * FR-62-49 [Rep 3 & 4] 11. CHECK #1 sealleakoff
flow greater than 0.9 gpm per pump and NOT
* FR-62-23 [Rep 1 & 2] * FR-62-49 [Rep 3 & 4] INITIATE normal plant shutdown USING appropriate
plant procedures
AND STOP affected Rep within 8 hours. IF Rep #1 seal leakoff flow reverts to high leakage (greater than 8.0 gpm): * FR-62-24 [Rep 1 & 2] * FR-62-50 [Rep 3 & 4] THEN GO TO Section 2.1, Rep Tripped or Shutdown Required.
GO TO Step 1.
..... Page 16 of 34 
[S_Q_N_--, ___
__ Rev. 23 [ STEP I ACTION/EXPECTED
RESPONSE 2.3 RCP #1 Seal Leakoff Low Flow (cont'd) 12. EVALUATE EPIP-1, Emergency
Plan Initiating
Conditions
Matrix. 13. EVALUATE the following
Tech Specs for applicability:
* 3.2.5, DNB Parameters
* 3.4.1.1, Reactor Coolant Loops and Coolant Circulation
-Startup and Power Operation
* 3.4.1.2, Reactor Coolant System -Hot Standby * 3.4.1.3, Reactor Coolant System -Shutdown * 3.4.6.2, RCS Operational
Leakage 14. GO TO appropriate
plant procedure.
---.. ---END OF SECTION Page 17 of 34 RESPONSE NOT OBTAINED 
Source SER 2123 SER 2122 SER 2121 SER2120 Probable Causes Corrective
Actions CONTINUED
3 (A-3) Setpoint Pump 1 FS-62-tO Pump 2 FS-62-23 Pump 3 FS-62-36 Pump 4 FS-62-49 .9 gpm decreasing
.9 gpm decreasing
.9 gpm decreasing
.9 gpm decreasing
FS-62-10 REAC COOL PMPS SEAL LEAKOFF LOW FLOW 1. No.1 seal less than 275 psid. 2. No.1 seal damage. 3. No.2 seal failure. [1] VERIFY Low Leakoff flow condition
on affected RCP(s) with the following
instruments:
Pump Leakoff Instrumentation
RCP 1 1-FR-62-23
RCP 2 1-FR-62-23
RCP3 1-FR-62-49
RCP4 1-FR-62-49
[2] ENSURE No.1 Seal Return Isolation
Valves OPEN Pump Valve RCP 1 1-FCV,..62-9
RCP2 1-FCV-62-22
RCP3 1-FCV-62-35
RCP4 1-FCV-62-48
[3] IF Unit 1 isin Mode 1 or 2, THEN GO TO AOP-R04, Reactor Coolant Pump Malfunctions.
[4] IF Unit 1 is in Mode 3, 4,or 5, THEN PERFORM the following:
[a] VERIFY No.1 Seal 220 psid AND No.1 seal leakoff greater than the minimum value shown in 1-S0-68-2
Appendix D. I SQN 1 1 1-AR':M5-S
Page 60t 45 Rev. 36 
3 A-3 CORRECTIVE
ACTIONS (CONTINUED)
References
FS-62-10 REAC COOL PMPS SEAL LEAKOFF LOW FLOW [b] ENSURE RCP seal water supply flow is between 6 gpm and 10 gpm per pump. [c] IF No.1 Seal ilP OR No.1 sealleakoff
is less than the minimum required values, THEN STOP the affected RCP USING 1-S0-68-2.
[d] ENSURE VCT pressure is between 17 psig and 45 psig. [e] CONTACT Engineering
for assistance.
458655-058-0,478601-62-2,4,7,9,47W610-68-1
SQN 1 Rev. 36 1 1-AR-M5-S
J Page 7 of 45 
OPL271 AOP-R.04 Revision 1 Page 3 of 26 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-R.04, REACTOR COOLANT PUMP MALFUNCTIONS
IV. LENGTH OF LESSON/COURSE:
2 hours V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of AOP-R.04, Reactor Coolant Pump Malfunctions.
B. Enabling Objectives:
O. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with , Reactor Coolant Pump Malfunctions
that are rated 2.5 during Initial License Training and 3.0 during License
Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of AOP-R.04.
2. Discuss the AOP-R.04 entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with AOP-R.04 entry conditions.
b. Describe the ARP requirements
associated
with AOP-R.04 entry conditions.
c. Interpret, prioritize, and verify associated
alarms are consistent
with AOP-R.04 entry conditions.
d. Describe the Administrative
conditions
that require Turbine Trip/ Reactor trip due to Reactor Coolant Pump Malfunctions.
3. Describe the initial operator response to stabilize
the plant upon entry into AOP-R.04. 4. Upon entry into AOP-R.04, diagnose the applicable
condition
and transition
to the appropriate
procedural
section for response.
5. Summarize
the mitigating
strategy for the failure that initiated
entry into AOP-R.04. 6. Describe the bases for all limits, notes, cautions, and steps of AOP-R.04.
7. Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures. 
8. 9. 10. OPL271 AOP-R.04 Revision 1 Page 4 of 26 Given a set of initial plant conditions
use AOP-R.04 to correctly:
a. Recognize
entry conditions
b. Identify required actions c. Respond to Contingencies
d. Observe and Interpret
Cautions and Notes Describe the Tech Spec and TRM actions applicable
during the performance
of AOP-R.04.
Apply GFE and system response concepts to the abnormal condition
-prior to, during and after the abnormal condition 
( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
87. 026 G2.4.20 087 Given the following:
Unit 1 at 100% power with Containment
Spray Pump 1 A-A out of service and tagged. -A LOCA results in a Reactor Trip and Safety Injection.
-After transferring
to the containment
sump the crew observes the amps fluctuating
on Containment
Spray Pump 1 B-B. In response, the crew stops the pump and transitions
to ECA-1.3, "Containment
Sump Blockage".
-The ST A reports a RED path to FR-Z.1, "High Containment
Pressure".
Which ONE of the following
identifies
the condition
that results in the restart of Containment
Spray Pump 1 B-B and what would be the desired flow rate? A. Due to implementing
FR-Z.1 , the pump will be restarted
and the desired flow rate is the DESIGN spray flow due to the challenge
to the Containment
Barrier. B. Due to implementing
FR-Z.1 , the pump will be restarted
and the desired flow rate is the MINIMUM spray flow required to control containment
pressure provided it does not cause the RHR pump to cavitate.
C. ECA-1.3 will restart the spray pump after TSC evaluation
and the desired flow rate is the DESIGN spray flow due to the challenge
to the Containment
Barrier. D!" ECA-1.3 will restart the spray pump after TSC evaluation
and the desired flow rate is the MINIMUM spray flow required to control containment
pressure provided it does not cause the RHR pump to cavitate.
Page 34 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRA CTOR ANAL YSIS: Page 35 A. Incorrect, when ECA-1.3 is entered FRGs are monitored
for information
only and the implementation
of FRGs is suspended.
Transition
to FR-Z 1 will not be made and a note in the ECA Appendix E for throttling
spray flow identifies
the desired flow rate to be the MINIMUM to control containment
pressure without causing cavitation
of the RHR pump not the MAXIMUM without causing cavitation.
Plausible
because FR-Z 1 would restart the spray pump and with a RED path identified
the Containment
pressure is at or above the design pressure and a severe challenge
to the barrier exists and establishing
the maximum flow would cause a greater drop in containment
pressure.
B. Incorrect, when ECA-1.3 is entered FRGs are monitored
for information
only and the implementation
of FRGs is suspended.
Transition
to FR-Z 1 will not be made. A note in the ECA Appendix E for throttling
spray flow identifies
the desired flow rate is the MINIMUM to control containment
pressure without causing cavitation
of the RHR pump. Plausible
because FR-Z 1 would restart the spray pump and establishing
the minimum flow to control containment
pressure without causing cavitation
of the RHR pump is correct for the conditions
in the stem. C. Incorrect, The FRGs are monitored
for information
only so ECA-1.3 is continued
and the ECA directs the TSC evaluation
if containment
pressure is greater than 9.5 psig and the pressure would be with a RED path present (12psig).
A note in the ECA Appendix E for throttling
spray flow identifies
the desired flow rate to be the MINIMUM to control containment
pressure without causing cavitation
of the RHR pump not the MAXIMUM without causing cavitation.
Plausible
because the ECA is the procedure
which restart and control the pump and with a RED path identified
the Containment
pressure is at or above the design pressure and a severe challenge
to the barrier exists and establishing
the maximum flow would cause a greater drop in containment
pressure.
D. CORRECT, A RED path on the Containment
Status Tree occurs when the pressure is greater than 12 psig. The FRGs are monitored
for information
only so ECA-1.3 is continued.
The ECA directs the TSC evaluation
if containment
pressure is greater than 9.5 psig and the pressure would be with a RED path present (12psig).
A note in the ECA Appendix E for throttling
spray flow identifies
the desired flow rate is the MINIMUM to control containment
pressure without causing cavitation
of the RHR pump. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 87 Tier 2 Group 1 KIA 026 G2.4.20 Containment
Spray Knowledge
of the operational
implications
of EOP warnings, cautions, and notes. Ilmportance
Rating: 3.8/4.3 Technical
Reference:
ECA-1.3, Containment
Sump Blockage, Rev 1 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 ECA-1.3 B.4 Question Source: Describe the bases for all limits, notes, cautions, and steps of ECA-1.3 Bank# ___ _ Modified Bank # ___ _ New X, __ Question History: New question for SQN 1/2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X __ 10 CFR Part 55 Content: (41.10/43.5/45.13 ) 10CFR55.43.b ( 5 ) Comments:
New question for SQN 1/2009 exam Page 36 
( SQN CONTAINMENT
SUMP BLOCKAGE I STEP] I ACTION/EXPECTED
RESPONSE 1. SUSPEND FRP implementation
and MONITOR status trees for information
only. II RESPONSE NOT OBTAINED Page 3 of63 ECA-1.3 Rev. 1 
SQN CONTAINMENT
SUMP BLOCKAGE ECA-1.3 Rev. 1 I S'TEP II ACTION/EXPECTED
RESPONSE 9. NOTIFY TSC to determine
optimum ECCSand containment
spray alignment
WHILE continuing
in this procedure.
10. ENSURE makeup water being added to RWST USING EA-63-2, Refilling
the RWST. II RESPONSE NOT OBTAINED CAUTION Re-establishing
Containment
Spray flow may result in RHR pump cavitation.
11. MONITOR containment
pressure:
a. CHECK containment
pressure less than 9.5 psig. a. NOTIFY TSC to evaluate restarting
Containment
Spray USING Appendix E, Throttling
Containment
Spray Flow. Page 14 of63 
( SQN CONTAINMENT
SUMP BLOCKAGE*
ECA-1.3 Rev. 1 Page 1 of 3 NOTE 1 NOTE 2 APPENDIX E THROTTLING
CONTAINMENT
SPRAY FLOW This appendix assumes
containment
spray suction is still aligned to containment
sump as specified
in ES-1.3. Throttling
containment
spray flow is desired to allow controlling
flow to prevent RHR pump cavitation.
Flow rate should be established
as directed by TSC. Desired flow
is the minimum flow rate needed to control containment
pressure without causing RHR cavitation.
1. RESET Containment
Spray Signal. 2. STATION operator in communication
with MeR at breaker for containment
spray discharge
valve (identified
in Step 6). 3. ENSURE Containment
Spray pump recirc valve for train to be started in PULL P-AUTO: * FCV-72-34 (Train A) OR * FCV-72-13 (Train B) 4. ENSURE discharge
valve CLOSED for pump to be started: * FCV-72-39 (Train A) 'OR * FCV-72-2 (Train 8) 5. START one containment
spray pump. Page 61 of 63 o o o o o o o 
( OPL271 ECA-1.3 Revision 0 Page 3 of 22 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: EMERGENCY
OPERATING
PROCEDURE
ECA-1.3, "Containment
Sump Blockage" IV. LENGTH OF LESSON/COURSE:
1 hour(s) V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of ECA-1.3, "Containment
Sump Blockage" B. Enabling Objectives:
B. Enabling Objectives:
o. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with Containment
Sump Blockage that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of ECA-1.3. 2. Discuss the ECA-1 .3 entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with ECA-1.3 entry conditions.
b. Describe the requirements
associated
with ECA-1.3 entry conditions.
3. Summarize
the mitigating
strategy for the failure that initiated
entry into ECA-1.3. 4. Describe the bases for all limits, notes, cautions, and steps of ECA-1.3. 5. Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
6. Given a set of initial plant conditions
use ECA-1.3 to correctly:
a. Identify required actions b. Respond to Contingencies
c. Observe and Interpret
Cautions and Notes 7. Apply GFE and system response concepts to the performance
of ECA-1.3 conditions. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
88. 039 A2.01 088 Given the following:
Unit 1 is at 100% power when a LOCA occurs. -The reactor trips when the containment
pressure rise causes a Safety Injection.
-The EOP network is entered. RCS pressure stabilizes
at 1580 psig. -Containment
pressure rises to 2.4 psig and stabilizes.
-The crew is in the process of terminating
Safety Injection.
-When determining
if..Jhe SI pumps should be stopped the following
is noted: RCS pressure is now 1540 psig and trending down. RCS subcooling
is 43&deg;F. Pressurizer
level is 19% and dropping.
-SG pressures:
#1 -590 psig and dropping.
#2 -600 psig and dropping.
#3 -580 psig and dropping.
#4 -605 psig and dropping.
MSIVs are open. Which ONE of the following
identifies
the status of the MSIVs and the proper crew response to the conditions?
A'I The MSIVs have failed to automatically
close. SI Reinitiation
Criteria does NOT exist, a transition
should be made to E-2, Faulted Steam Generator
Isolation.
B. The MSIVs have failed to automatically
close. SI Reinitiation
Criteria exist, restart the CCP, establish
CCPIT flow and Go To E-1, Loss of Reactor or Secondary
Coolant. C. MSIV automatic
closure Signal would NOT have been initiated.
SI Reinitiation
Criteria does NOT exist, a transition
should be made to E-2, Faulted Steam Generator
Isolation.
D. MSIV automatic
closure signal would NOT have been initiated.
Page 37 SI Reinitiation
Criteria exists, restart the CCP, establish
CCPIT flow and Go to E-1, Loss of Reactor or Secondary
Coolant. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRACTOR ANAL YSIS: Page 38 A. Correct, the MSIVs should have received a signal to automatically
close due to SG pressure but the SI Reinitiation
Criteria not being met is correct and a transition
to E-2 should be made. B. Incorrect, the MSIVs should have received a signal to automatically
close and the SI Reinitiation
Criteria not being met is correct. Plausible
because of the 2 parameters
that cause an MSIV to automatically
close, one (containment
pressure)
is identified
below the setpoint and the other (SG pressure)
is only slightly less than setpoint and if the SI Reinitiation
Criteria were applicable, the actions listed and transition
to E-1 are correct. C. Incorrect, the MSIVs should have received a signal to automatically
close due to pressure in the steam generators.
The transition
to E-2 is correct. Plausible
because of the 2 parameters
that cause an MSIV to automatically
close, one (containment
pressure)
is identified
below the setpoint and the other (SG pressure)
is only slightly less than setpoint and the transition
to E-2 is correct. D. Incorrect, the MSIVs having not received a signal to automatically
close is correct but the SI reinitiation
criteria is not met. Plausible
because of the 2 parameters
that cause an MSIV to automatically
close, one (containment
pressure)
is identified
below the setpoint and the other (SG pressure)
is only slightly less than setpoint and if the SI Reinitiation
Criteria were applicable, the actions listed and transition
to E-1 are correct. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 88 Tier 2 Group 1 KIA 039 A2.01 Ability to (a) predict the impacts of the following
malfunctions
or operations
on the MRSS; and (b) based on predictions, use procedures
to correct, control, or mitigate the consequences
of those malfunctions
or operations:
Flow paths of steam during a LOCA Importance
Rating: 3.1 / 3.2 Technical
Reference:
ES-1.1, SI Termination, Rev 10 1,2-47W611-1-1
R13 TI-28, Attachment
9, Unit 1 and 2 Cycle Data Sheet, Effective
Date 06/28/2007
Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPT200.MS
B.4.e & i Question Source: Describe the following
features for each major component
in the Main Steam System as described
in this lesson. e. Component
operation
i. Protective
features (including
setpoints)
OPL271ES-1.1
B.5 Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
Bank# ___ _ Modified Bank # ___ _ New X --Question History: New question for SQN 1/2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X __ 1 0 CFR Part 55 Content: ( 41.5/43.5/45.3/45.13 ) 10CFR55.43.b ( 5 ) Comments:
New question for SQN 1/2009 exam Page 39 
A 8 c D E F G H L-L-LL9MLv-Z'
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ElLOlfOO\1N (oI\7nl1-'-3
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I-lS-1-4D})T\
,----1 I .sAlliE ASI I F'CV-'-17
\ I NOTE 1 H1GH TEMP IN AV)(
-----j TVRtUNE FlOOI.t is-1-MA c HIGH HIotP IN AUK FEED'II'''TER
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PULL P-AUTO ." CllOsE 1
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FRCt.! STEA'II! CENERATOR
LOOfl ::5 tSAI.lE AS LOOP 1) -TYPICAL CONTROL ADMIN CHAN&#xa3;lc
____ LOCIC
lSSUED BY; P.G. TRUDEL ------------------_.
DR"'I1Ml:"tEIt'lJlnS
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[! IlAINTMH[D
IIY THE: 5QN CAD UNIT AND [5 PlAIU Of THE TVA
()ATAliA!!:,
47I1S"-O-1.2-----------LOGIG
DIAGRA\I IND'EX &-SY'UBOLS 4711810-'-'
,2 * .!l.4--......
DIAGItAU 4711601-1---------------FLO'lf
DIAGRA'III
FOJ1 ORICINAL INFORMATION
IN TITLE SLOCK SEE REVISION 0 MICfiOFIlU
COPY. 13 AD\UN CHANGE I.1JB REV CHANGE FIEF PREPARER CHECKER APPROVED DATE
NONE POWERHOUSE
UNITS 1 &: 2 MECHANICAL
LOGIC DIAGRAM EXCEPT AS NOTED CATEGORY 1 MAIN AND REHEAT STEAM SEQUOYAH NUCLEAR PLANT TENNESSEE
VALLEY AUTHORITY
NUt:LENt ENI;[NEEIIHNC
J A 8 c D E F 
( SQN UNIT 1 & 2 CYCLE DATA SHEET TI-28 Att. 9 Unit 0 {FOR INFORMATION
ONLy} Effective
Date: 06-28-2007
Page 7 of 16 Signal Setpoint Logic Block/Permissive
SAFETY INJECTION
SIGNALS (SIS) 1. Containment
Press Hi 1.54 psid 2/3 PTs None 2. Pressurizer
Press. Low 1870 psig 2/3 PTs Manual Below P-11 3. Steamline
Press Low 600 psig 2/3 PTs on 1/4 loops Manual Below P-11 4. Manual N/A 1/2 HS's SIRESET AFTER SIINITIATION, MUST WAIT FOR 60 SECOND TIMER TO RESET. THEN THE SI RESET PB FOR EACH TRAIN MUST BE ACTUATED.
THIS WILL BLOCK ANY AUTOMATIC
SI ACTUATION
SIGNAL BUT MANUAL SIIS NOT BLOCKED. TO REMOVE AUTO SI BLOCK, THE RX TRIP BREAKERS MUST BE CYCLED TO REMOVE THE P-4 SEAL-IN SIGNAL. CONTAINMENT
ISOLATION
SIGNALS (CIS) Phase A 1. SIS Any Signal 2. Manual Phase AI CVI HS 1/2 CONTAINMENT
ISOLATION
SIGNALS (CIS) Phase B 1. Containment
Press Hi-Hi 2.81 psid 2/4 2. Manual Phase B Handswitch
2/2 CONTAINMENT
VENT ISOLATION
SIGNALS (CVI) 1. RM-90-130
& 131 High Rad Signal 1/2 2. SIS Any Signal 3. Manual Phase B Phase B Handswitch
2/2 4. Manual Phase A Phase A Handswitch
1/2 CONTAINMENT
SPRAY ACTUATION
SIGNALS 1. Containment
Press Hi-Hi 2.81 psid 2/4 2. Manual Phase B Handswitch
2/2 MAIN STEAMLINE
ISOLATION
SIGNALS 1. Containment
Press Hi-Hi 2.81 psid 2/4 PTs 2. Steam line Press Low 600 psig 2/3 PTs on 1/4 loops Manual Below P-11 3. Steamline
Press Negative 100 psig decreasing
in 2/3 PTs on 1/4 loops Enabled only when Rate a 50 second time Steamline
Press SI constant signal blocked. FEEDWATER
ISOLATION
SIGNALS 1. S/G Level Hi-Hi 81% (P-14) 2/3 L Ts on any S/G 2. Rx Trip (P-4) with Lo T ave Rx Trip Bkrs Open 550'F 2/4 loops 3. SIS Any signal 
I SQN SI TERMINATION
I ES-1.1 . ________ ______ __________________________
I STEP II ACTION/EXPECTED
RESPONSE II RESPONSE NOT OBTAINED NOTE RCS pressure may be slowly dropping due to spray bypass flow or slight cooling of the pressurizer;
This should be considered "stable" pressure.
10. DETERMINE
if SI pumps should be stopped: a. CHECK RCS pressure:
* RCS pressure STABLE or RISING * RCS pressure greater than 1500 psig. b. STOP SI pumps, and PLACE in A-AUTO. a. IF NO S/G is Faulted, THEN GO TO ES-1.2, Post LOCA Cooldown and Depressurization.
---.---IF any S/G is Faulted, THEN PERFORM the following:
1) DO NOT CONTINUE this procedure
UNTIL Faulted S/G depressurization
stops OR criteria for stopping SI pumps are satisfied.
2) IF criteria for stopping SI pumps CANNOT be satisfied
after Faulted S/G depressurization
stops, THEN GO TO ES-1.2, Post LOCA Cooldown and Depressurization.
___ . ' b. IF pump(s) CANNOT be stopped in A-AUTO, THEN PLACE *affected
SI pump(s) in PULL TO LOCK. Page 10 of 27 
( SQN SI TERMINATION
FOLDOUT PAGE SIREINITIATION
CRITERIA IF SI has been terminated (CCPIT isolated, 81 pumps stopped, and RHR pumps NOT running in ECCS mode) AND either of the following
conditions
occurs: * RCS subcooling
based on core exit TICs less than 40&deg;F OR * Pressurizer
level CANNOT be maintained
greater than 10% [20% ADV], THEN a. ESTABLISH
ECCS flow by performing
one or both of the following:
* ESTABLISH
CCPIT flow as necessary
USINGAppendix
C * START CCPs or Sipumps manually as necessary.
b. GO TO E-1, Loss of Reactor or Secondary
Coolant. EVENT DIAGNOSTICS
* IF both trains of shutdown boards de-energized, THEN GO TO ECA-O.O, Loss of All AC Power. I ES-1.1 Rev. 10 * IF any S/G pressure dropping in an uncontrolled
manner or less than 140 psig AND SIG NOT isolated, THEN GO TO E-2, Faulted Steam Generator
Isolation.
TANK SWITCHOVER
SETPOINTS
* IF CSTlevelless
than 5%, THEN ALIGN AFW suction to ERCW. * IF RWST level less than 27%, THEN GO TO ES.:.1.3, Transfer to RHR Containment
Sump. Page 1a of 27 
OPL271 ES-1.1 Revision 1 Page 3 of 45 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: ES-1.1, "SI Termination" IV. LENGTH OF LESSON/COURSE:
1 hour(s) V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of ES-1.1, SI Termination.
B. Enabling Objectives:
o. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with SI Termination
that are rated;::::
2.5 during Initial License Training and;:::: 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of ES-1.1. 2. Discuss the ES-1.1 entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with ES-1.1 entry conditions.
b. Describe the requirements
associated
with ES-1.1 entry conditions.
3. Summarize
the mitigating
strategy for the failure that initiated
entry into ES-1.1. 4. Describe the bases for all limits, notes, cautions, and steps of ES-1.1. 5. Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
6. Given a set of initial plant conditions
use ES-1.1 to correctly:
a. Identify required actions b. Respond to Contingencies
c. Observe and Interpret
Cautions and Notes 7. Apply G FE and system response concepts to the performance
of ES-1.1 conditions. 
I. PROGRAM: OPERA TOR TRAINING II. COURSE: SYSTEMS TRAINING III. LESSON TITLE: Main Steam IV. LENGTH OF LESSON: 1 112 HOURS V. TRAINING OBJECTIVES
A. Terminal Objective: (included
in slides/slide
notes) OPT200.MS
Rev. 3 Page 3 of 54 Upon completion
of this lesson and others presented, the student should be able to apply the knowledge
to support satisfactory
performance
of the tasks associated
with the Main Steam System in the plant and on the simulator.
B. Enabling Objectives: (included
in slides/slide
notes) Demonstrate
an understanding
ofNUREG 1122 knowledge's
and abilities
associated
with the Main Steam System that are rated 2.5 during Initial License training for the appropriate
license position as identified
in appendix A. I.State the purpose/functions
of the Main Steam System as described
in the FSAR. 2.State the design basis ofthe Main Steam System in accordance
with the SQNFSAR. 3.Explain
the purpose/function
of each major component
in the flow path of the Main Steam System as illustrated
on the simplified
system drawing. 
( V. TRAINING OBJECTIVES (continued)
4. Describe the following
features for each major component
in the Main Steam System as described
in this lesson. a. Location b. Power supply (include control power as applicable)
c. Support equipment
and systems d. Normal operating
parameters
e. Component
operation
f. Controls g. Interlocks (including
setpoints)
h. Instrumentation
and Indications
1. Protective
features (including
setpoints)
J. Failure modes k. Unit differences
1. Types of accidents
for which the Main Steam components
are designed m. Location of controls and indications
associated
with the Main Steam in the control room and auxiliary
control room. OPT200.MS
Rev. 3 Page 4 of 54 
( Guidance for SRO-only Questions
Rev 0 Figure 2: Screening
for SRO-only linked to 10CFRSS.43(b)(S) (Procedures)
Can question be answered by knowing "systems knowledge", i.e., how the system works, flowpath, Yes 1 logic, location, etc. RO question No 1"1 SJ V $1 <::1(1)17
h J B,.xt P ILC{J:'D
+v pc., I /")17 I
vv' 4 W',/q UJ 11/9111 \;;) (..() \ 4 N \It Can question be answered by knowing immediate
Yes 1 operator actions? '\ RO question >:iNo Can question be answered by knowing entry conditions
for major EOPs? Yes I t RO question I No Does the question involve one or more of the following?
* * Assessing
plant conditions (normal, abnormal, or emergency)
and then prescribing
a procedure
or section of a procedure to mitigate, recover, or with which to proceed Recalling
what strategy or action is written into a plant,)'es
*1 procedure, including
when the strategy or action is required 1 No Question is not linked to 1 OCFR55.43(b)(5)
for SRO-only Page 8 of 19 SRO-only question 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
89. 062 G2 4.18089 Given the following:
Unit 1 is operating
at 100% power Diesel Generator
1 B-B is out of service and is expected to return to service in 2 hours. Subsequently, the following
events occur: -A loss of offsite power occurs. -The reactor trips and the crew enters the emergency
procedures.
-SI is NOT actuated.
-The crew transitions
to FR-H.1, "Loss of Secondary
Heat Sink" due to a RED Path condition
and is performing
the first step. No other Status Tree RED paths are present. -A fault on Shutdown Board 1A-A results in the emergency
supply breaker from Diesel Generator
1A-A tripping due a differential
relay. Which ONE of the following
identifies
the correct action to be taken and the bases for the action? A. Transition
to ECA-O .0, Loss of All AC Power because the ECA will direct actions to establish
heat sink with the TD-AFW pump. B. Remain in FR-H.1 unless a higher priority RED path occurs because the FR will direct actions to establish
heat sink with the TD-AFW pump. Transition
to ECA-O .0, Loss of All AC Power because all other procedures
in the EOP network assume
a minimum of at least one 6.9kV Shutdown Board is available.
D. Remain in FR-H.1 and initiate actions to manually restore 6.9kV Shutdown Board 1 B-B from 2B-B 6.9kV Shutdown Board maintenance
breaker because restoring
heat sink is the highest priority evolution
in progress.
Page 40 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRA GTOR ANAL YSIS: Page 41 A. Incorrect, The transition
to EGA-O.O is the correct action and the EGA will direct actions ensuring the TO AFW pump is in service but the reason is not because the EGA establishes
the heat sink. Plausible
because the transition
is correct and EGA will direct actions to place the TO AFW pump in service and establish
a heat sink. B. Incorrect, Remaining
is FR-H.1 is not correct, a transition
to EGA-O.O is required.
Plausible
because other EGAs do not take precedence
over the FRGs and with a RED path there is a severe challenge
to the Heat Sink function that would be addressed
if the TO AFW pump were in service. G. GORREGT, The transition
to EGA-O.O is the correct action and the reason is because all other procedures, including
FR-H. 1, assume a minimum of at least one train of shutdown power is available.
O. Incorrect, Remaining
is FR-H.1 is not correct, a transition
to EGA-O.O is required.
Plausible
because power could be restored to the 1 B-B board using the maintenance
breaker and restoring
Heat Sink is critical safety function being challenged. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 89 Tier 2 Group 1 KIA 062 G2 4.18 AC Electrical
distribution
Knowledge
of the specific bases for EOPs. Importance
Rating: 3.3/4.0 Technical
Reference:
1,2-15E500-1
R26 EPM-3-ECAOO.0, Basis Document for Loss of All AC Power, Rev 10 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 ECA-O.O BA Question Source: Describe the bases for all limits, notes, cautions and steps of ECA-O.O. Bank# ___ _ Modified Bank # X __ _ New __ _ Question History: Question modified from VC SUMMER 2007 SRO exam Question Cognitive
Level: Memory or fundamental knowledge
__ _ Comprehension
or Analysis _X. __ 10 CFR Part 55 Content: (41.10/43.1/45.13 ) 10CFR55A3.b ( 5 ) Comments:
VC SUMMER 2007 SRO exam question modified Page 42 
( 2007 ill NRC Exam ill SRO 100. Given the following
plant conditions:
* The plant is operating
at 100% power. * EDG "B" is out of service and is expected to return to service in two (2) hours. * Subsequently, the following
events occur: * A loss of offsite power. * The reactor is tripped and the crew enters EOP-l.O, Reactor Trip or Safety Injection.
* 51 is NOT actuated.
* The crew made a transition
to EOP-1S.0, Loss of Secondary
Heat Sink, based on a CSFST RED Path. * EDG "A" output breaker subsequently
trips on a differentia/lockout
on Bus 1DA. Which ONE (1) of the following
describes
the actions that will be taken and its bases? A. Immediately
transition
to EOP-6 .0, Loss OfAlI ESF AC Power. All other procedures
in the ERG network assume both 7.2 KV ESF busses are available.
B. Immediately
transition
to EOP-6 .0, Loss OfAlI ESF AC Power. All other procedures
in the ERG network assume a minimum of ONE (1) 7.2 KV ESF bus is available.
C. Remain in EOP-1S.0 until feed is restored and the RED condition
is cleared, and then transition
to EOP-6.0 , Loss of All ESF AC Power. RED path Function Recovery procedures
must be performed
until the condition
is cleared. D. Remain in EOP-1S.0 until directed to return to procedure
in effect, and then transition
to EOP-6 .0, Loss of All ESF AC Power. RED path Function Recovery procedures
must be finished to completion. 
SON EOI BASIS DOCUMENT FOR ECA .. O.O EPM .. 3 .. ECA .. 0.0 PROGRAM LOSS OF ALL AC POWER Rev. 10 MANUAL Page 8 of 98 EOP Step Number: 1 SUSPEND FRP implementation
and MONITOR status trees for information
only. ERG Step Number: 1, Note 2 of 2 CSF Status Trees should be monitored
for information
only. FRGs should not be implemented.
Purpose: To inform the operator that this guideline
should not be exited to perform any FRP in response to an identified
CSF challenge.
ERG Basis: The guideline
has priority over all FRGs and is written to implicitly
monitor and maintain critical safety functions.
This priority is necessary
since all FRGs are written on the premise that at least one shutdown board is energized.
Knowledge:
Guideline
ECA .. O.O has priority over the FRGs. EOP Basis: Same. Deviation:
Converted
note into a step. Justification:
Converted
note into a step since it contains a hidden action. SON EOP writer's guide disallows
hidden actions in cautions and notes. Since the action is required upon entry to ECA .. O.O, it is made an immediate
action step. Setpoint:
None. 
/ i I l. A B c o E F G H I( 1-00S3S I-Z' I TO 6.9KY co.tIlNBDB
(15&#xa3;500-2)
2 2 :3 :3 4 0) (411N732-1)
! QU!i&sect;l AU?! '9 3M-" 4 5 6 TO 161KV SWl1'CHYARD
HI.:52.NC)3DSg-
UNIT 1624NC)
i -+--'-'":12iZ-)=
TO &.tKV -17":2-'!+-)--+";'" NC
NC 4aOVUNITIiD
UTILlTV BUS
Ugh All '9 aM-" 5 6 
'-___ TO eel COOL TOWiR f * TRAItSFOlNtR
A (45N50I)1 . (15[500*2)
'&#xa5;b ccu.rJN STATION stRVICE -
.1 *
X Y 4/32/40 L C !ii6' '=' NC)414 HC)!!12 NO) d l TURS. BLOC fl .-:t ) r-------I I "-" I I
I 'XFRS i
J 'n'L AUXBO IB1_B HC) Y us'" '0) (45N732-Z1
! gInn All! R!l !,u_A r --------, -, I s:.!. y) NOTES; 1.
3.
INCREASE THE YOND BE 4. X: ONLY ONE BREAI(ER REQUIRED rOR PANELS
* .. e. THESE CABLES ARE NOT URhlINATrn
TO THE DIESEL CENERATOR
EXC1TER SUPI'OAT[D
UMOEIUtEATH
THEIR CABLE TRAY 7. THESE CABLES ARE NOT TERhltNAT[O
TO THE SHUTDOWN BOARDS. THEY ARE COILED UP 1M NAIiiOLE NUtMI[RS lJA. 13B. 14" ANO 14B * .. 9. CABLE LEADS FROM BREAKER 1120SB TO TIWISFROMER
Ct*! HAVE BEtN DISCONNECTED
AND ARE HOT TO BE RECONNECTED
WITHOUT M.E. APPROVAL.
ID. GENERAL CATEGORY 1 KEY DIAGRAM STATION AUX POWER SYSTEM SEQUOYAH NUCLEAR PLANT TENNESSEE
VALLEY AUTHORITY 
OPL271 ECA-O.O Revision 1 Page 3 of 21 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: EMERGENCY
OPERATING
PROCEDURE
ECA-O.O, LOSS OF ALL AC POWER IV. LENGTH OF LESSON/COURSE:
1 hour V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of ECA-O.O, LOSS OF ALL AC POWER. B. Enabling Objectives
O. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with ECA-O.O, LOSS OF ALL AC POWER that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of ECA-O.O. 2. Discuss the ECA-O.O entry conditions.
3. Summarize
the mitigating
strategy for the failure that initiated
entry into ECA-O.O. 4. Describe the bases for all limits, notes, cautions, and steps of ECA-O.O. 5. Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
6. Given a set of initial plant conditions
use ECA-O.O to correctly:
a. Identify required actions b. Respond to Contingencies
c. Observe and Interpret
Cautions and Notes 7. Apply GFE and system response concepts to the performance
of ECA-O.O conditions. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
90. 076 A2.02 090 Given the following
plant conditions:
Both Units operating
at 100% power. ERCW system in normal alignment.
-The following
annunciators
are LIT on Essential
Raw Cooling Water 0-XA-55-27A
panel on 1-M-27: -"UNIT 1 HEADER A PRESSURE LOW". -"UNIT 2 HEADER A PRESSURE LOW". -"PUMP J-A DISCH PRESS LOW". -"PUMP Q-A DISCH PRESS LOW". -The following
annunciator
status on Miscellaneous
1-XA-55-15B
panel on 1-M-15: -"ERCW DECK SUMP PUMP A RUNNING" is LIT. ERCW headers 1A and 2A indicating
LOW flow. -The crew implements
AOP-M.01, "Loss of Essential
Raw Cooling Water". Which ONE of the following
identifies
the correct section of AOP-M.01 to be implemented
for the conditions
and a mitigating
action directed to be taken in response to the conditions?
A. Section 2.7, "Supply Header 1A/2A Failure in the Yard"; Start additional
ERCW pumps to maintain pressure.
B. Section 2.7, "Supply Header 1A12A Failure in the Yard"; Stop and Lockout out all A Train ERCW pumps. C. Section 2.9, "Supply Header A Failure Upstream of Strainer Inlet Valves"; Start additional
ERCW pumps to maintain pressure. Section 2.9, "Supply Header A Failure Upstream of Strainer Inlet Valves"; Stop and Lockout out all A Train ERCW pumps. Page 43 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRACTOR ANAL YSIS: Page 44 A. Incorrect, All conditions
match entry conditions
for Section 2.7 except for sump pump running alarm and the Strainer Dp alarm being lit if the leak were downstream
of the strainer.
While starting additional
pumps might restore pressure,the
mitigating
action is to stop and lock out the A Train pumps to terminate
the leakage for a leak at the location identified
by the stem conditions.
Plausible
because Section 2.7 would be the correct procedure
section and the starting on additional
pumps is a mitigating
action during performance
of Section 2. 7 for a leak downstream
of the strainer.
B. Incorrect, All conditions
match entry conditions
for Section 2.7 except for sump pump running alarm and the Strainer Dp alarm being lit if the leak were downstream
of the strainer.
The A Train pumps being directed to be stopped and locked out is correct for the conditions
stated. Plausible
because Section 2.7 would be the correct procedure
section for a leak downstream
of the strainer and the stopping and locking out of the pumps in correct for the leak identified
in the stem. C. Incorrect, All alarms stated would be lit for a header break upstream of the Strainer.
Header pressure sensors are located is located just upstream of the strainers.
the sump pump running differentiates
the leak upstream from a yard leak (i .e. downstream
of the strainer).
While starting additional
pumps might restore pressure,the
mitigating
action is to stop and lock out the A Train pumps to terminate
the leakage. Plausible
because Section 2.9 is the correct procedure
section and the starting on additional
pumps is a mitigating
action for a leak downstream
of the strainer (Section 2.7). D. CORRECT, All alarms stated would be lit for a header break upstream of the Strainer.
Header pressure sensors are located is located just upstream of the strainers.
the sump pump running differentiates
the leak upstream from a yard leak (i .e. downstream
of the strainer).
Mitigating
action in the AOP section is to stop and lock out all A Train pumps. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 90 Tier 2 Group 1 KIA 076 A2.02 Ability to (a) predict the impacts of the following
malfunctions
or operations
on the SWS; and (b) based on those predictions, use procedures
to correct, control, or mitigate the consequences
of those malfunctions
or operations:
Service water header pressure.
Importance
Rating: 2.7/3.1 Technical
Reference:
AOP-M.01, Loss of Essential
Raw Cooling Water, Rev 19 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 AOP-M.01, BA & 5 Question Source: Upon entry into AOP-M.1 , diagnosis
the applicable
condition
and transition
to the appropriate
procedural
section for response.
Summarize
the mitigating
strategy for the failure that initiated
entry into AOP-M.01 Bank # X __ _ Modified Bank # ----New ---Question History: SQN question 076 A2.02 053 with some modification.
Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X, __ 10 CFR Part 55 Content: ( 41.5/43.5/45/3/45/13)
10CFR55A3.b ( 5 ) Comments:
SQN question 076 A2.02 053 with modification
to correct answer, all distractors, and stem. Used most plausible
2 of original distractors
and added requirement
to identify mitigating
actions. Changed correct answer to D. No significant
modification
to data in the stem. Page 45 
SQN LOSS OF ESSENTIAL
RAW COOLING WATER AOP-M.01 Rev. 19 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS CAUTION: ERCW header rupture in Auxiliary
Building could fill the passive sump in 15 minutes. Prompt action is needed. 1. DIAGNOSE the failure: IF ... GO TO SECTION PAGE E.RCW Pump(s) tripped or failed 2.1 ERCW pump failure 5 High flow ERCW Supply Header 1A 2.2 Supply Hdr 1A Failure 8 to Aux Bldg High flow ERCW Supply Header 1 B 2.3 Supply Hdr 1B Failure 12 to Aux Bldg High flow ERCW Supply Header 2A 2.4 Supply Hdr 2A Failure 16 to Aux Bldg High flow ERCW Supply Header 2B 2.5 Supply Hdr 2B Failure 23 to Aux Bldg Indications
of an ERCW Return Header rupture 2.6 Return Hdr rupture 29 (must be diagnosed
locally since M-27 indications
in Aux Bldg are not affected)
Low flow ERCW Supply Header 1A and 2A, 2.7 Supply Header 1A12A 41 AND Failure in Yard Area STRAINER OIFF PRESS HIGH alarm LIT [M-27 A, C-3 and/or 0-2] Low flow ERCW Supply Header 1 Band 2B, 2.8 Supply Header 18/2B 55 AND Failure in Yard Area STRAINER OIFF PRESS HIGH alarm LIT [M-27 A, C-6 and/or 0-5] (step continued
on next page) Page 3 of 149 
*** Jo -. SQN LOSS OF ESSENTIAL
RAW COOLING WATER -AOp .. M.01 Rev. 19 ( STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS (Continued)
3. (Continued)
IF ... GO TO SECTION PAGE Low flow ERCW supply headers 1A and 2A, AND STRAINER DIFF PRESS alarms DARK [M-27 A, C-3 and 0-2], AND at least one of the following
alarms LIT: * ERCW DECK SUMP LEVEL HI alarm LIT 2.9 Supply Header A 67 [1-M-15B, A-3] Failure Upstream of Strainer Inlet Valves OR * ERCW DECK SUMP PMP RUNNING [1-M-15B, 0-2 or 0-4] OR * MECH EQUIP SUMP LVL HI alarm LIT [1-M-15A, B-6] Low flow ERCW supply headers 1 Band 2B, AND STRAINER DIFF PRESS alarms DARK [M-27 A, C-6 and 0-'5], AND at least one ofthe following
alarms LIT: -. -.. * ERCW DECK SUMP LEVEL HI alarm LIT [1-2.10 Supply Header B 76 M-15B, A-3] Failure Upstream of Strainer Inlet Valves OR * ERCW DECK SUMP PMP RUNNING [1-M-15B, 0-2 or 0-4] OR * MECH EQUIP SUMP LVL HI alarm LIT [1-M-15A, B-6] Loss of flow on ALL ERCW supply headers 2.11 Loss of all ERCW flow 83 in modes 1-4. END OF SECTION Page 4 of 149 
( ( [ LOSS OF ESSENTIAL
RAW COOLING WATER AOP-M.01 Rev. 19 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED '------'---
2.9 ERCW Supply Header A Failure Upstream of ERCW Strainer Inlet Valves CAUTION: During operation, CCP and SI Pumps may experience
bearing failure 10 minutes after loss of ERCW. 1. STOP and LOCK OUT all Train A ERCW Pumps. 2. DISPATCH operators
with radios to perform the following:
* PERFORM Appendix F, Rx MOV Board ERCW Valves [Aux Bldg el. 749', Rx MOV Boards] * PERFORM Appendix G, ERCW MCC Valves [ERCW Pumping Station] o ENSURE all pumping station watertight
doors are CLOSED [ERCW Pumping Station] Page 67 of 149 
SQN LOSS OF ESSENTIAL
RAW COOLING WATER AOP-M.01 Rev. 19 L-________ -L _____________________________ ________________ -L ________ STEP ACTION/EXPECTED
RESP_O_N_S_E_-.l.I_.
___ R---,E_S_P_O_N_S_E_N_O_T_O_B_T_A_IN-'-E_D
__ ----Il ........ 2.7 ERCW Supply Header 1A12A Failure inYard Area CAUTIONS:
* During operation, CCP and SI Pumps may experience
bearing failure 10 minutes after loss of ERCW cooling. NOTE: * Loss of 2A ERCW Supply Header affects both Units' Train A CCS Heat Exchangers.
Isolation
of ruptured Unit and restoration
of ERCW to intact CCS Heat Exchangers
is time critical to prevent tripping both Units. * Engineering
may be able to identify ruptured yard header using yard piping drawings (17W300 series). 1. DISPATCH personnel
to locate failure. 2. DISPATCH operators
with radios to PERFORM the following
appendixes:
* Appendix F, Rx MOV Board ERCW Valves [Aux Bldg, 749' elev, Rx MOV Boards] * Appendix G, ERCW MCC Valves [ERCW Pumping Station] 3. START additional
Train A ERCW Pumps as required to maintain pressure between 78 psig and 124 psig. Page 41 of 149 
( 10 (B-3) Source Setpoint SER 1078 (Unit 1 annunciator
system) 2-PS-67 -493A 50 psig decreasing
UNIT 2 HEADER A PRESSURE LOW Retransmitted
to U-2 SER 2096 (Unit 2 annunciator
system) Probable Causes Corrective
Actions References
1. Unit 2 ERCW "An train pumps tripped. 2. Unit 2 "A" Train ERCW line break. 3. System realignment
increasing
demand excessively.
[1] IF alarm is in conjunction
with any of the following
indications:
* ERCW pump trip/failure
* HIGH header flow * LOW header flow THEN GO TO AOP-M.01, Loss of Essential
Raw Cooling Water. [2] IF system realignment, in progress, is most probable cause for alarm, THEN CORRECT alignment
in accordance
with instruction
in progress, OR START additional
Train A ERCW Pump as required, in accordance
with instruction
in progress.
[3] IF cause not apparent, THEN DISPATCH personnel
to check pumping station and piping for ruptures or cause of alarm. [4] EVALUATE EPIP-1 Emergency
Plan Classification
Matrix. [5] IF Unit 2 "A" train is declared inoperable, THEN CONSULT Technical
Specifications
3.7.4. 45B655-27A-0, 45N655-32, 47B601-55-13
SQN O-AR-M27-A
Page 12 of 39 o Rev. 19 
1 (A-1) Source Setpoint SER 1066 (Unit 1 annunciator
system) 1-PS-67 -493A 50 psig decreasing
UNIT 1 HEADER A PRESSURE LOW Retransmitted
to U-2 SER 2084 (Unit 2 annunciator
system) Probable Causes Corrective
Actions References
1. Unit 1 ERCW "A" train pumps tripped or manually stopped. 2. 1A ERCW line break. 3. System realignment
increasing
demand excessively.
[1] IF alarm is in conjunction
with any of the following
indications:
* ERCW pump trip/failure
* HIGH header flow * LOW header flow THEN GO TO AOP-M.01, Loss of Essential
Raw Cooling Water. ' [2] IF system realignment, in progress, is most probable cause for alarm, THEN CORRECT alignment
in accordance
with instruction
in progress, OR START additional
Train A ERCW Pump as required, in accordance
with instruction
in progress.
[3] IF cause not apparent, THEN DISPATCH personnel
to check pumping station and piping for ruptures or cause of alarm. [4] EVALUATE EPIP-1 Emergency
Plan ClaSSification
Matrix. [5] IF Unit 1 "A" train is declared inoperable, THEN CONSULT Technical
Specifications
3.7.4. 458655-27
A-O, 45N655-32, 478601-55-13
SQN O-AR-M27-A
Page 3 of 39 o Rev.19 
( ( 16 Source Setpoint SER 1075 (Unit 1 annunciator
system) 0-PS-67-461
and 50 psig decreasing
with pump breaker closed PUMP Q-A DISH PRESS 52a on breaker LOW Retransmitted
to U-2 SER 2093 (Unit 2 annunciator
system) Probable Causes Corrective
Actions References
1. ERCW pump Q-A damaged. 2. Train A ERCW line break. 3; Instrument
malfunction.
4. Insufficient
pumps running for system flow demand. [1] ENSURE sufficient
pumps running for system configuration.
[2] VERIFY ERCW pump Q-A running. [3] IF pump is running, THEN DISPATCH operator to determine
problem. [4] IF Q-A ERCW pump is failed, THEN GO TO AOP-M.01, Loss of Essential
Raw Cooling Water. [5] IF pressure low due to ERCW line Break, THEN GO TO AOP-M.01, Loss of Essential
Raw Cooling Water. [6] IF "A" train is declared inoperable, THEN CONSULT Technical
Specification
3.7.4. 45B655-27A-0, 45N655-32, 47B601-55-13 (C-2) SQN O-AR-M27-A
Page 18 of 39 o Rev. 19 
8 Source Setpoint SER 1067 (Unit 1 annunciator
system) 0-PS-67 -433 and 50 psig decreasing
with pump breaker PUMP J-A DISH PRESS LOW 52a on breaker closed Retransmitted
to U-2 SER 2085 (Unit 2 annunciator
system) Probable Causes Corrective
Actions References
1. ERCW pump J-A damaged. 2. Train A ERCW line break. 3. Instrument
malfunction.
4. Insufficient
pumps running for system flow demand. [1] ENSURE sufficient
pumps running for system configuration.
[2] VERIFY ERCW pump J-A running. [3] IF pump is running, THEN DISPATCH operator to determine
problem. [4] IF J-A ERCW pump is failed, THEN GO TO AOP-M.01, Loss of Essential
Raw Cooling Water. [5] IF pressure low due to ERCW line Break, THEN GO TO AOP-M.01, Loss of Essential
Raw Cooling Water. [6] IF "A" train is declared inoperable, THEN CONSULT Technical
Specification
3.7.4. 45B655-27
A-O, 45N655-32, 4 7B60 1.,55-13 (B-1) SQN O-AR-M27-A
Page 10 of 39 o Rev. 19 
076 A2.02 053 QUESTIONS
REPORT for BANK SQN Questions ( Given the following
plant conditions:
Both Units operating
at 100% power. ERCW system in normal alignment.
The following
annunciators
are LIT on Essential
Raw Cooling Water 0-XA-55-27A
panel on 1-M-27: -"UNIT 1 HEADER A PRESSURE LOW". -"UNIT 2 HEADER A PRESSURE LOW". -"PUMP J-A DISCH PRESS LOW". -"PUMP Q-A DISCH PRESS LOW". The following
annunciator
is LIT on Miscellaneous
1-XA-55-15B
panel on 1-M-15: -"ERCW DECK SUMP PUMP B RUNNING".
ERCW headers 1A and 2A indicating
LOW flow. Which ONE of the following
describes
what has occurred in the ERCW system? A. 'A' header pumps have tripped. B. Train A ERCW 480v. board has been deenergized. Header has ruptured upstream of the '2A' strainer.
D. '1A' header has ruptured between the IPS and Auxiliary
Bldg. A. Incorrect
-Low disch pressure alrams on the pumps indicate the pump breakers are closed B. Incorrect
-Strainer alarm would be lit for a clogged strainer.
No sump pump running alarm with high pressure.
C. Correct -All alarms stated would be lit for this accident.
Pressure sensors are located is located just upstream of the strainers.
D. Incorrect
-Strainer Dp alarm would be lit, all other conditions
match except sump pump running alarm. Thursday, July 31,20088:28:19
AM 1 
OPL271 AOP-M.01 Revision 0 Page 3 of 44 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-M.01 LOSS OF ESSENTIAL
RAW COOLING WATER IV. LENGTH OF LESSON/COURSE:
2.0 hour(s) V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of AOP-M.01, LOSS OF ESSENTIAL
RAW COOLING WATER B. Enabling Objectives:
O. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with a Loss of Essential
Raw Cooling Water that are rated:;::
2.5 during Initial License Training and:;:: 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Explain the purpose/goal
of AOP-M.01.
2. Discuss the AOP-M.01 entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with AOP-M.01 entry conditions.
b. Describe the ARP requirements
associated
with AOP-M.01 entry conditions.
c. Interpret, prioritize, and verify associated
alarms are consistent
with AOP-M.01 entry conditions.
d. Describe the Administrative
conditions
that require Turbine Trip/ Reactor trip due to Loss of Essential
Raw Cooling Water. 3. Describe the initial operator response to stabilize
the plant upon entry into AOP-M.01. 4. Upon entry into AOP-M.01 , diagnose the applicable
condition
and transition
to the appropriate
procedural
section for response.
5. Summarize
the mitigating
strategy for the failure that initiated
entry into AOP-M.01. 6. Describe the bases for all limits, notes, cautions, and steps of AOP-M.01. 
( 7. 8. 9. 10. OPL271 AOP-M.01 Revision 0 Page 4 of 44 Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
Given a set of initial plant conditions
use AOP-M.01 to correctly:
a. Recognize
entry conditions
b. Identify required actions c. Respond to Contingencies
d. Observe and Interpret
Cautions and Notes Describe the Tech Spec and TRM actions applicable
during the performance
of AOP-M.01.
Apply GFE and system response concepts to the abnormal condition
-prior to, during and after the abnormal condition 
Guidance for SRO-only Questions
RevO Figure 2: Screening
for SRO-only linked to 10CFRSS.43(b)(S) (Procedures)
Can question be answered by knowing "systems l knowledge", i.e., how the system works, flowpath, Yes RO question logic, location, etc. No Can question be answered by knowing immediate
Yes I \ RO question operator actions? No Can question be answered by knowing entry Yes I -j RO question I conditions
for major EOPs? iNo r Does the question involve one or more of the following?
* Assessing
plant conditions (normal, abnormal,or
emergency)
and then prescribing
a procedure
or section of a procedure
to mitigate, recover, or with which to proceed * Recallingwhat
strategy or action is written into a plant Yes I SRO-only procedure, including
when the strategy or action is required question I No Question is not linked to 10CFR55.43(b)(5)
for SRO-only Page 8 of 19 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
91. 055 G2.4.3 091 Which ONE of the following
correctly
completes
the statement
below? The Unit 1 Condenser
Vacuum Mid Range Radiation
Monitor, 1-RM-90-255, is a monitor ... A. included in Tech Spec LCO 3.3.3.7, "Accident
Monitoring
Instrumentation", and is used in the Radiological
Emergency
Plan (REP) to classify an event based on gaseous effluent release. B. included in Tech Spec LCO 3.3.3.7, "Accident
Monitoring
Instrumentation", and is used in the Radiological
Emergency
Plan (REP) to classify an event based on the fission product barrier matrix. C'!" NOT included in Tech Spec LCO 3.3.3.7, "Accident
Monitoring
Instrumentation", but is used in the Radiological
Emergency
Plan (REP) to classify an event based on gaseous effluent release. D. NOT included in Tech Spec LCO 3.3.3.7, "Accident
Monitoring Instrumentation", but
is used in the Radiological
Emergency
Plan (REP) to classify an event based on the fission product barrier matrix. Page 46 
( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRA C TOR ANAL YSIS: Page 47 A. Incorrect, 1-RM-90-255
is not included in the TIS LCO for Accident Monitoring
but the monitor is included in EPIP-1, Emergency
Plan Classification
Matrix", Table 7-1 under Gaseous Monitors whose output can be used for determining
Emergency
Plan classifications.
Plausible
that it would be included in the TS LCO because the monitor is a Post Accident Monitor and the monitor being included in EPIP-1, Emergency
Plan Classification
Matrix", Table 7-1 under Gaseous Monitors
whose output can be used for determining
Emergency
Plan classifications
is correct. B. Incorrect, 1-RM-90-255
is not included in the TIS LCO for Accident Monitoring
but the monitor is not used to classify an event based on the fission product barrier matrix. Plausible
that it would be included in the TS LCO because the monitor is a Post Accident Monitor and there are radiation
monitors used to make classifications
in the fission product barrier matrix. C. CORRECT, 1-RM-90-255
is a post accident monitor but is not included in the TIS LCO for Accident Monitoring.
The monitor is included in EPIP-1, Emergency
Plan Classification
Matrix", Table 7-1 under Gaseous Monitors whose output can be used for determining
Emergency
Plan classifications.
D. Incorrect, 1-RM-90-255
is a post accident monitor but is not included in the TIS LCO for Accident Monitoring.
while the monitor is included in EPIP-1, Emergency
Plan Classification
Matrix", Table 7-1 under Gaseous Effluent release, the monitor is not used to classify an event based on the fission product barrier matrix. Plausible
because the monitor not being included in LCO for Accident Monitoring
is correct and there are radiation
monitors used to make classifications
in the fission product barrier matrix. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 91 Tier 2 Group 2 KIA 055 G2A.3 Condenser
Air Removal System Ability to identify post-accident
instrumentation.
Importance
Rating: 3.7/3.9 Technical
Reference:
EPIP-1, Emergency
Plan Classification
Matrix, Rev 40 Technical
Specification
LCO 3.3.10 FSAR Section 7.5 Amendment
2 1-AR-M30-A, Post Accident Radiation
Monitoring, Rev 15 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPT200. RM BA.I Question Source: Describe the following
characteristics
of each major component
in the Radiation
Monitoring
system: I. Types of accidents
for which the components
are designed.
OPT200. CONDVAC BA.I Describe the following
characteristics
of each major component
in the CDVAC system: I. Types of accidents
for which the CDVAC components
are designed.
Bank# ____ _ Modified Bank # -----New _X __ Question History: New question for SQN 1/2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
_X __ Comprehension
or Analysis __ _ 10 CFR Part 55 Content: ( 41.6 / 4504 ) 10CFR55A3.b ( 2,4,5 ) Comments:
New question for SQN 1/2009 exam Page 48 
( ( ( Source SER 1745 1-RE-90-255
Probable Causes Corrective
Actions References
Setpoint see 1-RM-90-255
display and/or data base 1. Steam generator
tube rupture. 2. Instrument
Malfunction.
17 (C-1) 1-RA-255A
CONDVAC EXH MID RANGE HIRAD [1] CHECK [1-RM-90-255]
on panel 1..:M-30 to verify activity level. [2] EVALUATE EPIP-1, Emergency
Plan Classification
Matrix; [3] IF high radiation
indicated
in condenser
vacuum exhaust, THEN GO TO AOP-R.01, Steam Generator
Tube Leak. 478601-55-75,47W610-90-4
SQN 1-AR-M30-A
Page 20 of 27 1 Rev. 15 
( [iiQUOY_A_H_L.........
___ E_M_E_R_G_E_N._C_Y_P_L.A
__ N_C_L_A_S_S_1F_IC_A_T_IO_N_M_A_TR_I_X
__ ..l...-.-,-E_P_IP_-1---1
A L L A L L A L L A L L EA8 dose, resulting
from an actual or imminent release of gaseous radioactivity>
1 Rem TEDE or > 5 Rem thyroid CDE for the actual or projected
duration of release. (1 or 2 or 3): 1. A. VALID rad monitor reading exceeds the values under General Emergency
in Table 7-1 for >15 min, unless assessment
within that 15 min confirms that the critenon is not exceeded.
2. Field surveys indicate >1 Rem/hr gamma or an 1-131 concentration
of 3.9E-06 ;ICi/cm' at the EAB (Fig. 7-A) OR 3. Dose assessment
results indicate EAB dose >1 Rem TEDE or >5 Rem thyroid CDE for the actual or projected
duration of the release (Fig. 7-A). EAB dose resulting
from an actual or imminent release of gaseous radioactivity
>100 mrem TEDE or >500 mrem thyroid CDE for actual or projected
duration of release. (1 or 2 or 3): 1. A VALID rari monitor reading> Table 7-1 values under Site I,rea for:> 15 min. unless assessment
within that 15 min confirms that the criterion
is not exceeded.
PR 2 Field surveys indicate >100 mrem/hr gamma or an 1-131 cone of 3.9E-07 pCi/cm' at the EAB (Fig. 7-A). OR .'. Dose assessment
results indicate EAB dose >100 mrem TEDE or >500 mrem thyroid CDE for actual or projected
duration of the release (Fig. I-A). Any UNPLANNED
release of gaseous radioactivity
that exceeds 200 times the aDeM Section 1.2.2.1 Limit for >15 minutes. (1 or 2 or 3 or 4) '1. A VALID rad monitor reading> Table 7-1 values under Alert for >15 minutes, unless assessment
within that 15 minutes confirms that the criterion
is not exceeded.
OR 2. Field surveys indicate >10 mrem/hr gamma at the EAB tor >15 minutes (Fig 7-A). 3. Dose assessment
results indicate EAB dose >10 mrem TEDE for the duration of the release (Fig. 7 -A). OR 4. Sample results exceed 200 times the ODCM limit value for an unmonitored
release of gaseous radioactivity
>15 minutes in duration.
Any UNPLANNED
release of gaseous radioactivity
that exceeds 2 times the aDCM Section 1.2.2.1 Limit for >60 minutes. (1 or 2 or 3 or 4) 1. A VALID rad monitor reading> Table 7-1 values under UE for >60 minutes, unless assessment
within that 60 minutes confirms that the criterion
is not exceeded.
OR 2. Field surveys indicate >0.1 mrem/hr gamma at the EAB for >60 minutes (Fig 7-A) 3. Dose assessment
results indicate EAB dose >0.1 mrem TEDE for the duration of the release (Fig. 7-A). OR 4. Sample results exceed 2 times the ODCM limit value for an unmonitored
release of gaseous radioactivity
>60 minutes in duration A L L A L L Page 43 of4? Not Applicable.
Any UNPLANNED
release of liquid radioactivity
that exceeds 200 times the aDCM Section 1.2.1.1 Limit for >15 minutes. (1 or 2) 1. A VALID rad monitor reading > Table 7-1 values under Alert for >15 minutes, unless assessment
within this time period confirms that the criterion
is not exceeded.
aR 2. Sample results indicate an ECl >200 times the aDCM limit value for an unmonitored
release of liquid radioactivity
>15 minutes in duration Any UNPLANNED
release of liquid radioactivity
to the environment
that exceeds 2 times the aDCM Section 1.2.1.1 Limit for >60 minutes. (1 or 2) 1, A VALID rad monitor reading> Table 7-1 values under UE for >60 minutes, unless assessment
within this time period confinms that the criterion
is not exceeded.
2. Sample results indicate an ECl >2 times the aDCM limit value for an unmonitored
release of liquid radioactivity
>60 minutes in duration.
Revision 40 
( ( SEQUOYAH EMERGENCY
PLAN CLASSIFICATION
MATRIX EPIP-1 TABLE 7-1 EFFLUENT RADIATION
MONITOR EALS NOTE: The monitor values below, if met or exceeded, indicate the need to perform the required assessment
If the assessment
can not be completed
within 15 minutes (60 minutes for UE), the appropriate
emergency
classification
shall be made based on the VALID reading. GASEOUS MONITORS Units (2) UE Alert SAE General Site Total Release Limit fl Cils 4.90E+05 4.90E+07 1.31E+OS 1.31E+09 1-RI-90-400 (EFF LEVEL) -U-1 Shield Bldg pCi/s 4.90E+05 4.90E+07 1.31E+OS 1.31E+09 2-RI-90-400 (EFF LEVEL) -U-2 Shield Bldg pCi/s 4.90E+05 4.90E+07 1.31E+OS 1.31 E+09 O-RM-90-1
01 B -Auxiliary
Bldg cpm 1.03E+05 Offscale l1j Offscale(1)
Offscale(1)
O-RM-90-132B
-Service Bldg cpm 2.62E+06 Offscale (1) Offscale (1) Offscale(1)
1-RI-90-421
thru 424 -U-1 MSL Monitors(2)
pCi/cc 1.71 E-01 1.71E+01 4.5SE+01 4.5SE+02 2-RI-90-421
thru 424 -U-2 MSL Monitors(2)
pCilcc 1.71 E-01 1.71E+01 4.5SE+01 4.5SE+02 1-RM-90-255
or 256 -U-1 eVE mR/h 4.10E+02 4.10E+04 1.09E+05 1.09E+06 2-RM-90-255
or 256 -U-2 eVE mR/h 4.10E+02 4.10E+04 1.09E+05 1.09E+06 RELEASE DURA nON minutes >60 >15 >15 >15 LIQUID MONITORS Units UE Alert Site Area General Site Total Release. Limit pCi/ml 6.50E-03 6.50E-01 N/A N/A RM-90-122
-RadWaste cpm 1.45E+06 Offscale(1)
N/A N/A RM-90-120,121
-S/G Bldn cpm 1.07E+06 Offscale(1) N/A N/A RM-90-225
-Cond Demin cpm 1.90E+06 Offscale (1) N/A N/A RM-90-212
-TB Sump cpm 3.2SE+03 3.2SE+05 N/A N/A RELEASE DURA nON minutes >60 >15 >15 >15 ASSESSMENT
METHODS: * Airbome Dose Assessment
per SON EPIP-13 "Dose Assessment" * ODCM Liquid Release Rate assessment
per SON 0-TI-CEM-030.030.0
* Integrated
Airborne Release Rate assessment
per SON 0-TI-CEM-030.030.0
(1) The calculated
value is outside of the upper range for this detector.
The maximum monitor output which can be read is 1.0E+07 cpm. Releases in excess of monitor capacity should be evaluated
for proper classification
by use of Dose Assessment.
(2) These unit values are based on flow rates through one PORV of 890,000 Ib/hr at 1078.7 psia with 0.25% carry over (0.9975 quality).
Before using these values, ensure a release to the environment
is ongoing, (e.g., PORV). NOTE 1: These EALs are based on the assumption
that an emergency
release is restricted
to one pathway from the plant. In all cases, the total site EAL is the limiting value. Therefore, in the case where there are multiple release paths from the plant, it is the total release EAL (obtained
from ICS and/or SON 0-TI-CEM-030-030, "Manual Calculation
of Plant Gas, Iodine, and Particulate
Release Rates for Offsite Dose Calculation
Manual (ODCM) Compliance")
that will determine
whether an emergency
classification
is warranted.
NOTE 2: In the case when there is no CECC dose assessment
available, the length and relative magnitude
of the release is the key in determining
the classification.
For example, in the case of the NOUE EAL of 2 times the Tech Spec limit, the classification
is based more on the fact that a release above the limit has continued
unabated for more than 60 minutes, than on the projected
offsite dose. NOTE 3: See REP Appendix B for basis information.
Page 46 of 47 Revision 40 
SQN
OF.CONDENSAlE
VACUUM 1-PI-ICC-090-255.0
1 EXHAUST POST ACCIDENT RADIATION
MONITOR Rev. 7 1-R-90-255
Page 6 of 65 2.0 REFERENCES
2.1 Performance
References
None 2.2 Developmental
References
A. Administrative
References
None B. TVA and Vendor drawing 1. 47W610-90-1
2. 45N1620-12
3. 45WI651-16,-17,-20,-21
C. Manufacturer
Manuals 1. SQN-VTM-WI30-130
& VTD-WI30-0150
Vendor Manual For Westronics
Smartview
Data Recorder 2. SQN-VTM-G063-0430, RM-I000 Digital Radiation
Processor
Technical
Manual (Document
04508100-1TM)
3. SQN-VTM-G063-0010,Vendor
Technical
Manual For Radiation
Monitoring
System, Volumes I, II, III, and IV Contract No. 92759 D. Other Developmental
References
1. FSAR Sections:
7.5, 11.4, 12.1.4, 12.2.4 2. Set Point and Scaling Document (SSD) l-R-90-255 
SQN-18 TABLE 7.5-2 (Sheet 2) TABLE OF VARIABLES
FOR POST ACCIDENT MONITORING
Variable Typel Minimum Minimum Redundancy
Description
Cateoorv Ranoe From Ranoe To Reauired Notes Aux Bldg EXH Vent Rad Level -E3 1E-9 1E-4uCi/CC
N/A See Deviation
No. 14 Particulates
& Halogens Remote Analysis Utilizing
Removable
Filter May be Used. Aux Bldg Passive Sump C3 SEE NOTES N/A Low & Hi Level Alarm in MCR (FLR & EQP DRN SMP) LVL AUX Bldg Pressure D2 -0.5 +0.5 Inches WG N/A AUX Cntl Air Sys Pressure D2 0 125 Psig N/A RG1.97 R2 -POWER SUPPLY Boron Injection
Flow D2 0 110% (Design) N/A (Flow in HPI System) 0 864GPM Component
Cooling Sys Surge D3 0 100% N/A Actual Range 0 to 124 Inches Tank Level 0 10,000Gal
Component
Cooling Water Flow D2 0 110% (Design) N/A to ESF Equip 0 5523GPM Component
Cooling Water Temp D2 30 130DEG F N/A See Deviation
No.7 to ESF Equip Condenser (Air Removal Sys) E2 0 110% (Design) N/A Vacuum EXH Flow 0 49.5CFM Condenser (Air Removal Sys) B3 C3 E2 1E-6 1E4 uCi/CC N/A Part of Sec Side RAD Lvi Vacuum EXH RAD Level -Noble Gas Condensate
Storage Tank Water D2 0 367,000 GAL N/A Safety Source is-ERCW Level See ERCW to AFW Valve Position T75-2.doc 
( TABLE 3.3-10 ACCIDENT MONITORING
INSTRUMENTATION
INSTRUMENT
1. Reactor Coolant T HOT (Wide Range) (Instrument
Loops 68-001,-024,-043,-065)
2. Reactor Coolant T COLD (Wide Range) (lnstrumeni
LOops 68-018,-041,-060,-083)
3. Containment
Pressure (Wide Range) (Instrument
Loops 30-310,-311)
4. Containment
Pressure (Narrow Range) (Instrument
Loops 30-044,-045)
5.:..
Water Storage Tank Level ,(Instrument
Loops
-----'\ 6. Reactor Coolant Pressure (Wide Range) (Instrument
Loops 68-062,-066,-069)
7. Pressurizer
Level (Wide Range) (Instrument
Loops 68-320,-335,-339)
8. Steam Line Pressure (Instrument
Loops 1-002A,-002B,-009A,-009B,-
020A, -020B, -027 A, -027B) 9. Steam Generator
Level-(Wide Range) (Instrument
Loops 3-043,-056,-098,-111)
10. Steam Generator
Level-(Narrow Range) (Instrument
Loops 3-039,-042,-052,-055,-094,-
097,-107,-11
0) 11. Auxiliary
Feedwater
a. Flow Rate (Instrument
Loops
b. Valve Position Indication (Instrument
Loops 3-164,-164A,-172,-156, -156A,-173,-148,-148A,-174,-171,-171A,-175)
SEQUOYAH -UNIT 1 TOTAL NO. OF CHANNELS 4(1/RCS Loop) 4(1/RCS Loop) 2 2 2 3 3 21steamline
4(1/steam
generator)
2/steam generator
1/steam generator
3/steam generator
3/43-56 MINIMUM CHANNELS ACTION REQUIRED 4(1/RCS Loop) 1 4(1/RCS Loop) 1 2 1 2 1 2 1 C 3 2 3 2 21steam line 1 4(1/steam
1 generator)
2/steam generato"r
1 1/steam generator
5 3/steam generator
5 July 9, 1992 Amendment
No. 46,114, 149,159 
TABLE 3.3-10 (ContirlUed)
ACCIDENT MONITORING
INSTRUMENTATION , INSTRUMENT
17. Neutron Flux a. Source Range (Instrument
loops 92-5001,-5002)
b. Intermediate
Range (Instrument
loops 92-5003,-5004)
18. ERCW to AFW Valve Position ') a. Motor Driven Pumps (Instrument
loops ,3-116A, -116B, -126A, -126B) b. Turbine Driven Pumps (Instrument
loops
-136B, -179A, -179B) , i 19. Contain!'11ent
Isolation
ValvePositiOri
", (Panels
& TR-B XX-55-6l)
,;: ';'. . ... ..... " TOTAL NO. OF CHANNELS 2 2 1ffrain/Pump
(2 ValveS/Train)
2 Trains " (2 ValveS/Trairi) , ' 1 Naive MINIMUM CHANNELS REQUIRED 2 1ffrain/Pump
(2 ValveslTrain)
2 Trains (2 ValveslTrain)
1 Nalve## #Source P-6'(Block
of Sotirce Range Reacto; Trip) setpoint ##Not required for isolation
valVes that are'dosed
and deactivated.
' " ) ACTION 1 1 1 1 3 SEQUOYAH -UNIT 1 3I43-56b July 9,1992 Amendment
No. 112, 149, 159 
( TABLE 3.3-10 (Continued)
ACCIDENT MONITORING
INSTRUMENTATION
INSTRUMENT
12. Reactor Coolant System Subcooling
Margin Monitor (Instrument
Loops 94-101,-102)
13; Containment
Water Level (Wide Range) (Instrument
Loops 63-178,-179)
14. Incore Thermocouples
a. Core Quadrant (1) b. Core Quadrant (2) c. Core Quadrant (3) d. Core Quadrant (4) 15. Reactor Vessel Level Instrumentation
a. Dynamic Range (Instrument
Loops 68-367,370)
b. Lower Range (Instrument
Loops 68-368,371) c. Upper Range (Instrument
Loops 68-369,372) 16. Containment
Area Radiation
Monitors a. Upper Compartment (Instrument
Lo'ops 90-271,-272)
b. Lower Compartment (Instrument
Loops 90-273,-274)
SEQUOYAH -UNIT 1 TOTAL NO OF CHANNELS 2 2 65 6 2 2 3/43-56a MINIMUM CHANNELS REQUIRED ACTION 2 1 2 1 2(1fTrain)
1 2(1fTrain)
1 2(1fTrain)
1 2(1fTrain)
1 2 1 2 1 2 1 1 4 1 4 October 4, 1995 AmendmentNo.
112, 149,159,213
/ / 
I. PROGRAM: OPERATOR TRAINING II. COURSE: SYSTEMS TRAINING III. TITLE: RADIATION
MONITORING
SYSTEM OPT200.RM
Rev. 2 Page 3 of 166 IV. LENGTH OF LESSON: 4 hour lecture; 1 hour simulator
demonstration;
2 hour study/workshop
v. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of this lesson and others
presented, the student should be able to apply the knowledge
to support satisfactory
performance
of the tasks associated
with the Radiation
Monitoring
System in the plant and on the simulator.
B. Enabling Objectives:
O. Demonstrate
an understanding
of NUREG 1122 knowledge's
and abilities
associated
with the Radiation
Monitoring
System as identified
in Appendix A. 1. State the purpose/functions
of the Radiation
Monitoring
System as described
in the SQN FSAR. 2. State the design basis of the Radiation
Monitoring
System in accordance
with the SQN FSAR. 3. Explain the purpose/function
of each major component
in the flow path of the Radiation
Monitoring
System as illustrated
on a simplified
system drawing. 4. Describe the following
characteristics
of each major component
in the Radiation
Monitoring
System: a. Location b. Power supply (include control power as applicable)
c. Support equipment
and systems d. Normal operating
parameters
e. Component
operation
f. Controls g. Interlocks (including
setpoints)
h. Instrumentation
and Indications
1. Protective
features (including
setpoints)
J. Failure modes k. Unit differences
1. Types of accidents
for which the components
are designed m. Location of controls and indications
in the control room and auxiliary
control room 
V. TRAINING OBJECTIVES (Cont'd):
B. Enabling Objectives (Cont'd):
5. Describe the operation
of the Radiation
Monitoring
System: a. Precautions
and limitations
b. Major steps performed
while placing the system in service c. Alarms and alarm response d. How a component
failure will affect system operation
e. How a support system failure will affect system operation
f. How a instrument
failure will affect system operation
OPT200.RM
Rev. 2 Page 4 of 166 6. Describe the administrative
controls and limits for the Radiation
Monitoring
System: a. State Tech Specs/TRM
LCOs that govern the system. b. State the :::::1 hour action limit TS LCOs c. Given the conditions/status
of the Radiation
Monitoring
System components
and the appropriate
sections of the Tech Spec, determine
if operability
requirements
are met and what actions are required 7. Discuss related Industry Events VI. TRAINING AIDS: A. Classroom
Computer and Local Area Network (LAN) Access B. Computer projector
C. Simulator (if available) 
I. PROGRAM: OPERATOR TRAINING II. COURSE: SYSTEMS TRAINING III. TITLE: CONDENSER
VACUUM IV. LENGTH OF LESSON: 2 hour lecture; 1 hour simulator
demonstration;
1 hour self-study/workshop
V. TRAINING OBJECTIVES:
A. Terminal Objective:
OPT200.CONDVAC
Rev. 1 Page 3 of 31 Upon completion
of this lesson and others presented, the student should be able to apply the knowledge
to support satisfactory
performance
of the tasks associated
with the Condenser
Vacuum (CDVAC) system in the plant and on the simulator.
B. Learning Objectives:
O. Demonstrate
an understanding
ofNUREG 1122 knowledge
and abilities
associated
with the CDVAC system that are rated::::
2.5 during Initial License Training for the appropriate
license position as identified
in Appendix A. 1. State the purpose/functions
of the CDVAC system as described
in the FSAR. 2. State the design basis of the CDVAC system in accordance
with the SQN FSAR. 3. Explain the purpose/function
of each major component
in the flow path of the CDV AC system as illustrated
on a simplified
system drawing. 4. Describe the following
characteristics
of each major component
in the CDV AC system: a. Location b. Power supply (include control power as applicable)
c. Support equipment
and systems d. Normal operating
parameters
e. Component
operation
f. Controls g. Interlocks (including
setpoints)
h. Instrumentation
and Indications
i. Protective
features (including
setpoints)
j. Failure modes k. Unit differences
l. Types of accidents
for which the CDV AC system components
are designed m. Location of controls and indications
associated
with the CDV AC system in the control room and auxiliary
control room 
V. TRAINING OBJECTIVES (Cont'd):
B. Learning Objectives (Cont'd):
5. Describe the operation
of the CDVAC system: a. Precautions
and limitations
OPT200.CONDVAC
Rev. 1 Page 4 of 31 b. Major steps performed
while placing the CDV AC system in service c. Alarms and alarm response d. How a component
failure will affect system operation
e. How a support system failure will affect CDV AC system operation
f. How a instrument
failure will affect system operation
6. Describe the administrative
controls and limits for the CDVAC system: a. State Tech Specs/TRM
LCOs that govern the CDV AC b. State the:::;l hour action limit TS LCOs c. Given the conditions/status
of the CDVAC system components
and the appropriate
sections of the Tech Spec, determine
if operability
requirements
are met and what actions are required 7. Discuss related Industry Events VI. TRAINING AIDS: A. Classroom
Computer and Local Area Network (LAN) Access B. Computer proj ector C. Simulator (if available) 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
92. 071 A2.05 092 Given the following: -A planned release of Waste Gas Tank B is in progress when power is lost to 0-RM-90-118, Waste Gas Effluent Rad Monitor. Which ONE of the following
identifies
(1) the effect the loss of power will have on the release and (2) the requirement
to allow any additional
release of the tank with the radiation
monitor out of service? A'! (1) The release will automatically
terminate;
(2) A new 0-SI-CEM-077-41
0.4, "Waste Gas Decay Tank Release", package would be required for any additional
release of the tank after ODCM actions were met. B. (1) The release will automatically
terminate;
(2) The existing 0-SI-CEM-077-41
0.4, "Waste Gas Decay Tank Release", package could used for any additional
release of the tank after ODCM actions were met. C. (1) An alarm will be generated
and MANUAL termination
of the release will be required;
(2) A new 0-SI-CEM-077-41
0.4, "Waste Gas Decay Tank Release", package would be required for any additional
release of the tank after ODCM actions were met. D. (1) An alarm will be generated
and MANUAL termination
of the release will be required;
Page 49 (2) The existing 0-SI-CEM-077-41
0.4, "Waste Gas Decay Tank Release", package could used for any additional
release of the tank after ODCM actions were met. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRACTOR ANAL YSIS: Page 50 A. CORRECT, the release would be automatically
terminated
due to an instrument
malfunction
and a new package would be required to make any additional
releases from the tank. B. Incorrect, the release would be automatically
terminated
due to an instrument
malfunction
but the current SI package would not be used to make any additional
releases from the tank, a new package would be required.
Plausible
because the release would be automatically
terminated
and releases can be stopped and restarted
using the same package under other conditions.
C. Incorrect, the release would not require manual actions to terminate, it would be automatically
terminated
due to an instrument
malfunction.
A new SI package would be required to make any additional
releases from the tank. Plausible
because some release point radiation
monitor instrument
malfunctions
only alarm and ta new SI isrequiredd
for any addition release from the tank. D. Incorrect, the release would not require manual actions to terminate, it would be automatically
terminated
due to an instrument
malfunction.
The current SI package would not be used to make any additional
releases from the tank, a new package would be required.
Plausible
because some release point radiation
monitor instrument
malfunctions
only alarm and releases can be stopped and restarted
using the same package under other conditions. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 92 Tier 2 Group 2 KIA 071 A2.05 Ability to (a) predict the impacts of the following
malfunctions
or operations
on the Waste Gas Disposal System; and (b) based on those predictions, use procedures
to correct, control, or mitigate the consequences
of those malfunctions
or operations:
Power failure to the ARM and PRM Systems Importance
Rating: 2.5* / 2.6 Technical
Reference:
1 ,2-47W611-77-4
R10 0-AR-M12-B, Common Radiation
Monitor 0-XA-55-12B, Rev 28 0-SO-77-15, Waste Gas Decay Tank Release, Rev. 15 0-SI-CEM-077-410.4,Waste
Gas Decay tank Release, Rev.0014 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
No training objective
identified
Question Source: Bank# ___ _ Modified Bank # X __ _ New ---Question History: WBN bank question Question Cognitive
Level: Memory or fundamental
knowledge
_X __ Comprehension
or Analysis __ _ 10 CFR Part 55 Content: (41.5/43.5/45.3/45.13 ) 10CFR55.43.b ( 5 ) Comments:
Page 51 
( Watts Bar KIA: 000059 AA2.03 Accidental
Liquid RadWaste ReI. Ability to determine
and interpret
the following as
they apply to the Accidental
Liquid Radwaste Release: Failure modes, their symptoms, and the causes of misleading
indications
on a radioactive-liquid
monitor. Question:
Given the following:
* The unit is at 100% power and all equipment
is available.
* A planned Cask Decontamination
Collector
Tank Release is in progress when the following
occurs: o Annunciator
181-C 'WOS RELEASE LINE 0-RM-90-122
INSTR MALF" alarms. o Annunciator
181-A 'WOS RELEASE LINE 0-RM-90-122
L1Q RAO HI" remains dark. Which ONE of the following
failures would cause the above to prevent an accidental
release, and the action required to allow the restart of the release with the radiation
monitor out of service? A. Loss of power to the rate meter. The existing 001 package could be used to resume the release. B. Loss of power to the rate meter. The existing 001 package closed and new 001 package issued prior to resuming the release. C. Loss of signal from the detector.
The existing 001 package could be used to resume the release. O. Loss of signal from the detector.
The existing 001 package closed and new 001 package issued prior to resuming the release DISTRACTOR
ANALYSIS a. Incorrect.
The 'loss of power to the rate meter' would generate both the 181 C and 181 A alarms so the 181A alarm would also be LIT. The existing 001 package could not be used to continue the release. Plausible
because the candidate
may not recall the inputs to the alarms correctly
and there are conditions
identified
in the release instructions
that do allow the release to be continued
using the current 001 package. b. Incorrect.
The 'loss of power to the rate meter' would generate both the 181 C and 181 A alarms so the 181A alarm would also be LIT. A new 001 package would be needed to continue the release. Plausible
because the candidate
may not recall the inputs to the alarms
correctly, but realize a new 001 package is required to continue the release. c. Incorrect.
The 'loss of signal from the detector'
would generate a 181-C alarm but would NOT generate a 181 A alarm as stated in the distractor, however the existing 001 package could not be used to continue the release. Plausible
because the alarm status listed is correct for the listed condition
would cause the termination
of the release and the there are conditions
identified
in the release instructions
that do allow the release to be continued
using the current 001 package. d. CORRECT. The 'loss of signal from the detector'
would generate a 181-C alarm but would NOT generate a 181 A alarm and a new 001 package would be required prior to continuing
the release. 
7 I I I J B I I LJ 9 I LJ DESCRIPTION
SELECTOR SW GAS A GAS COMMON TO ANAL INC VLV ALL TANKS GAS B GAS r-------, ANAL INC VLV I I GAS C GAS I I ANAL INC \lLV 10 TVA NO WEST NO
10368
HB 10.378 PCV-77-11.3S
10388 CONTROL RELAY GOX!" GAXI GDX2A GAXI GDX3A GAXI 11 _I L.I L i i 12 NOTES: 1. FOR GENERAL NOTES AND REFERENCES
SEE SHEET 47"611-77-1.
2. ONE WASTE CAS CCMPRESSOR:
WIl.l. BE RUN CONTINUOUSl.
Y WITH THE OTHER SERVING AS A BACKUP TO BE STARTED WHEN HEADER PRESSURE EXCEEDS 2 PSIG. THE COMPRESSORS
liIl.l. BE Al.TERNATED
FOR UNIFORM WEI\R. I-I-
GOX-H --==:" 9 C 0 E I A I
XS F I
"\.J-I J 77-185 I I G I
H GAS o GAS ANAL INC ,VLV PCV-77-100B
103gB GAS E CAS PCV-77-IOIB
10S28 ANAL INC \lLV GAS F' G"S GDX4A GAXI COX5" GAXt 3. ONE DECAY TANK PRESSURE ISOLATION
VALVE WILL NORMALLY 8E OPEN WITH ANOTHER SELECTED FOR ST .... ND8Y. ALL OTHERS Ii I LL BE CLOSED. -l-. THIS IS II HAND OPERATED NEEDLE VALVE. I-A I-1-8 I-
e X-9A -=--99 LS LVL > SP 77-958/.1., (HI-HI) loS LVL>SP 77-95A/B (HI) Lev Y-a 77-958 VENT 7 I I L __ ___ ...J "NAL ING ,VLV GAS G GAS ANAL tNG VLV r AUTDr---"\
GAS H GAS l ANAL SAMPLING YLV GAS TANK J GAS O?EN HS SAMPlE ANAL Y SAMPLING Vt.. V 77-115 "-OFF PCV-77-1028
PCV-77-1+58
PCV-77-1+6B
?CV-17-1+79
10538 GOX6" 105+8 COX7A loSSE COX8" 10568 GOX9A t-HS-77-1
H -HS-77-113
r-HS-77-100
GAXI GAXI ADMIN CHANCE GAXI GAXI (TYP AU
FLOW> SP 1-FS-JO-1Soj" "'" I L:I-_{}iHIELD
BLDG EXH',c ENT UNIT 2 SELECTED
pAU'-'T"'OC":'J\::::=
_____ -I'i-HS-77-IOI I--i-HS-17-102
r-HS-77-1+5
!--HS-77-1+6
r-H S-77-1+7 SHIELD BLDG EXH. VENT UNtT 2 SELECTED
.FVENT PCV 77-115A X GAS CECAY TANK A PRESS. ISOLATION
VLV (PCV-l036A)
LO PRESSURE N2
158 '\1'11'61
r-17-1. COORD 1'-4 >-t
__ __ I GAS DECAY TANK A X GAS ANAl. YZING SAMPLING VALVE (PCV-l0368) 8 r-U PRESS. < SP
-I-I-D I-\ PS VENT 77-'" OFF I-77-119 ON 77-119 FSV -i
HS \.. S FIC 77-24
f-,.....i>lX::J-----
..... -.l------C>.I:::l---
I-I-I-E I-----t-----------CAS
DECAY TANK 8
_ I-f-----,'-----------GA5
DECAY TANK C
_ I + ---,'-----------GAS
DECAY TANK 0
_ I I-f-----,'-----------GAS
DECAY TANK E
I I-f-----,'-----------OA5
DECAY TANK F
_ I + ---,'-----------OAS
DECAY TANK 0
_ I I-f-----,'-----------OAS
DECAY TA"K H -J RAOlAnDN.
ISOLATlDN
VALVE (RCV-Ol+) LJ.:::""" GAS ANAL m. ) I-e-I-F l-I-
INC: ADMIN CHANGE PER RIMS: 837
001 REV CHANGE REf PREPARER CHECKER APPROVED DATE SCALE: NONE EXCEPT AS NOTED POWERHOUSE
CATECORY 1 UNITS 1 '" 2 MECHANICAL
LOGIC DIAGRAM WASTE DISPOSAL SYSTEM SEQUOYAH NUCLEAR PLANT TENNESSEE
VALLEY AUTHORITY
DESIGN INITIAL tSSUE ---,'-----------GAS
D'CAY TA"K J ---------'1-;::
-.'!:::::::' --I--CRAFTER, CHECKER, J.CDlE L.WALTERS
--------------RO ISSUE f"ER: ENGINEERING
APf>ROVAL
DRAIING MAOE 1 N/A II ceo fROM AC-N/}' AO-R5
DESIGNER, __ __ DATE N/A 7-3-91 I I I 8 I I I I I I I I I 9 I I I I I I I I 10 I I I I CAD MAINTAINED
DRAWING I REVIEWER;
"/A 3. M.R.SEDlAC!K/KRS
CCO NO: 1. 2-47W611-77-4Rl0 
/ ----A-----------8----C --D --E --F --G --H I \ I I 7-LL-L L9A1.L7-Z*
L 1 " 1 ;.1 1 I '-' CONTROLS ,I SIMIL.AR TO
., LCV X 77-10' TO DRAIN 2 I 3 4 I I ,-----------------------------11---
I1NAS STOP
__ __, A \Z!:J 1A to-".21T P-AUTO 5 I DESCRIPTION
'( TANK A ISOLATION
VLV Y TANK 8 ISOLATION
VLV C ION VLV 0 ION VLV E ION VLV F ION YLV G ION VLV H ION VLY J ION VLV TVA NO PCV-77-11SA
F'CV-77-II4A
PCV-77-11JA
PCV-77-IOOA
PCV-77-I01A
PCV-77-I02A
PCV-77-I.SA
PCV-77-t4GA
PCV-77-147A
6 I 1 1 1 1 1 1 1 1 1 1 CONTROL.S
.1 SIMILAR TO I 1'"'1 LCVI'OS eves VCT -UNIT t eves VCT -UKIT 2 eves EVA? -UNIT 1 eves EVAP UNIT 2 was G'A -WDS SRST -S5 VCT UNIT 2_ SS VCT UNIT ,-' eves HT -,OS ReDT -UNIT 2 wos !!.COT UNIT 1 -1 1 1 1 1 1
PRESS. < SF' PS ",j' TO DRAIN r--PS J 0 r--+--->I'--{
r--. N I ( 77-3"/0
77-88 -y ---FSV 77-90
77-89 fCV 77 -90 COMPRESSOR
A _MOISTURE -I 5EPARATlOR
@o -L-r---.
X I SEAL WATER COOLER v J II Q r---
Ib:: * "":'.'."-rt:::
Lev
77-95A OPEN: HS CLOSE a 77-405 x LCV 77-405 77-40.5 S TO DRAIN ,......, PRIMARY WATER SYSTEM TO eves HOLDUP TANK < 47161 I-sa-s. caORD 0-4 PCV 77-92 VEN;-ifLy
t&#xa9;U sv '-' ___ ----------_-------=------
COMPRESSO!!.
____________________
.-r---1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 I 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2 3 4 5 6 
18 (C-4) Source Setpoint SER 768 (Unit 1 annunciator
system) O-RM-90-118
N/A O-RA-90-1188
WDS GAS EFF MON INSTR MALFUNC Probable Causes Corrective
Actions References
1. Instrument
power failure. 2. Instrument
placed in TRIP ADJ position (Except for monitors equipped with RM-1000 modules only). 3. Instrument
downscale
failure or loss of signal. 4. Operate/Calibrate
switch set to calibrate (RM-1000 modules only). . [1] IF gas release in progress, THEN REQUEST Radwaste AUO to verify 0-RCV-90-119
closed. [2] CHECK 0-RM-90-118
on 0-M-12 for possible trouble. [3] IF O-RM-90-118
is inoperable, THEN [a] NOTIFY the Chemistry
Shift Supervisor
to comply with ODCM requirements.
[b] PLACE equipment
off normal or inop tags, identifying
the condition.
[c] WHEN ODCM action is satisfied, THEN RESUME the release using appropriate
procedures.
[4] COMPLY with ODCM, Section 1.1.2 requirements.
[5] INITIATE WO for maintenance, if required.
458655-128-0, 45N657-18, 45N667-1, SON O-AR-M12-B
Page 22 of 40 0 Rev. 28 
WASTE GAS DECAY TANK RELEASE 0-SO-77-15
Rev: 15 Page 11 of 16
6.0 NORMAL OPERATION (Continued)
[23] VERIFY [O-PCV-77-117]
gas release header pressure control valve is maintaining
5.3 psig as indicated
on the HEPA filter inlet pressure gauge on paneI1-L-335
located under the stairway near the WGDT Valve Gallery. [24] RECORD below the time this release was started. Release started hrs. [25] RECORD below, the rate of gas release. Release rate In. H 2 0 Date CAUTION If during the remainder
of this instruction
a malfunction
of 0-RE-90-118
or 0-FCV-77-119
occurs, this release must be stopped. [26] NOTIFY the Unit 1 Operator that a gas release is in progress.
-----NOTE The activity level recorded for RM-90-400
should be for the applicable
Shield Building.
[27] OBTAIN the following
information
from the U-1 UO, AND RECORD the information
below: [a] IF RM-90-400
operable, THEN RECORD Activity Level on RM-90-400
_____ CPM. [b] IF 0-RE-90-118
operable, THEN RECORD Activity Level on 0-RE-90-118
CPM. 
SQN Waste Gas Decay Tank Release 0-SI-CEM-077
-410.4 Unit 0 Rev. 0014 Page 15 of 42 6.2 Pre-Release
Instructions
-Chemistry (continued)
[16] SIGNOFF for item A and either item B or item C and CIRCLE B or C to indicate which one was satisfied.
A. Approval of pre-release
data generated
by this Instruction.
/ Performer
Date B. Verification
that monthly projected
offsite dose limits (ODCM SR 2.2.2.4) have NOT been exceeded, based on most recent performance
of SI-422.1, OR C. Verification, with Operations
support, that selected WGDT has been held a minimum of 60 days and all applicable
requirements
have been met. / Performer
Date [17] TRANSMIT release package to Operations
with authorization
/ / D Time Time to release. D 6.3 Release Instructions
-Operations
[1] REVIEW Steps 6.1 [3] and 6.1 [4] and Steps 6.2[11] monitor data. [2] IF radiation
monitor O-RM-90-118
setpoint change is required, (when setpoint in Step 6.2[11] is greater than the setpoint in Steps 6.1 [3]), THEN REQUEST a setpoint change. 
SQN Waste Gas Decay Tank Release 0-SI-CEM-077
-410.4 Unit 0 Rev. 0014 Page 16 of 42 6.3 Release Instructions
-Operations (continued)
[3] IF a release is to be made outside of normal release hours (0900 -1600), THEN OBTAIN US/SRO justification
and initials in remarks section of Surveillance
Task Sheet. [4] OBTAIN US/SRO approval of pre-release
data generated
by this Instruction
and approval for this release. / US/SRO Date [5] INITIATE release of selected WGDT contents in accordance
with 0-SO-77 -15 at or below the flow rate (Le., pressure drop) recorded on Appendix B, and RECORD release start time and information
requested
in the table in Appendix B for release initiation
and at one-half hour intervals.
NOTE / Time OPERABLE status of 0-FE-77-230
can be determined
by noting deflection
of indicator.
[6] IF [O-FE-77-230J
is INOPERABLE
at initiation
of release, THEN GO TO Step 6.3[8]. [7] IF [O-FE-77-230J
becomes INOPERABLE
during release, THEN PERFORM the following
substeps:
[7.1] STOP release. [7.2] NOTIFY US/SRO. 
SQN Waste Gas Decay Tank Release 0-SI-CEM-077
-410.4 UnitO Rev. 0014 Page 17 of 42 6.3 Release Instructions
-Operations (continued)
[8] IF release is to continue with [0-FE-77-230JINOPERABLE, THEN PERFORM the following
substeps.
[8.1] ENSURE that a test gauge (0 -20 inches of H20 suggested)
is installed
across [0-FE-77-2301.
[8.2] ENSURE serial number, range and calibration
due date of test gauge along with installing
Instrument
Mechanic's
initials are recorded in remarks section of Appendix B. [8.3] ENSURE pressure
readings from test gauge are recorded in place of [0-FE-77-230J
readings on Appendix B. [9] IF [0-RM-90-118J
or [RM-90-400J
alarms, THEN NOTIFY On-shift Chemistry
Personnel
who will contact the Cognizant
Chemist/System
Engg. for further guidance in processing
tank contents.
[10] WHEN release is complete or stopped, THEN RECORD the following
on Appendix B. [10.1] Release stop time [10.2] WGDT psig [10.3] Initials [11] IF radiation
monitor setpoint changes were made (Step 6.3[2]), THEN RETURN the radiation
monitors to their initial setpoints.
[12] NOTIFY the US/SRO and On-shift Chemistry
Personnel
that this release is complete.
[13] REVIEW 0-SO-77-15.
[14] ATIACH 0-SO-77-15
to this release package. [15] TRANSMIT the release package to the Chemistry
Laboratory
for post release evaluation.
o o o 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
93. 072 A2.02 093 Given the following:
Unit 1 operating
at 100% power. Unit 2 in MODE 6 with the core off-load in progress An internal electrical
failure causes the output of Spent Fuel Pit area radiation
monitor 0-RM-90-103
to fail above the HI RAD setpoint.
Which ONE of the following
identifies
how the Auxiliary
Building Isolation (ABI) is affected by the failure and whether the actuation
is required to be reported to the NRC in accordance
with SPP-3.5, "NRC Reporting
Requirements"?
A. Only Train B is initiated;
8-hour notification
required. Only Train B is initiated;
8-hour notification
NOT required.
C. Both Train A & Train B are initiated;
8-hour notification
required.
D. Both Train A & Train B are initiated;
8-hour notification
NOT required.
DISTRACTOR
ANAL YSIS: Page 52 A. Incorrect, Only the Train B ABI will be initiated
from the monitor but no 8-hour notification
would be required because the actuation would
be an invalid actuation.
Plausible
because the initiation
of Train B only is correct and ESF actuations
are normally reportable.
B. CORRECT, Only the Train B ABI will be initiated
from the monitor and actuation
would be an invalid actuation, thus,no 8-hour notification
would be required.
C. Incorrect, Both'trains
will not be initiated, only the Train B will initiate and no 8-hour notification
would be required because the actuation
would be an invalid actuation.
Plausible
because other actuations
do come from multiple sensors through isolation/separation
relays and ESF actuations
are normally reportable.
D. Incorrect, Both trains will not be initiated, only the Train B will initiate and no 8-hour notification
would be required because the actuation
would be an invalid actuation.
Plausible
because other actuations
do come from multiple sensors through isolation/separation
relays and the actuation
not being reportable
is correct. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 93 Tier 2 Group 2 KIA 072 A2.02 Ability to (a) predict the impacts of the following
malfunctions
or operations
on the ARM system-and (b) based on those predictions, use procedures
to correct, control, or mitigate the consequences
of those malfunctions
or operations:
Detector failure. Importance
Rating: 2.8/2.9 Technical
Reference:
0-AR-M12-B, Common Radiation
Monitor 0-XA-55-12B, Rev 28 SSP-3.5, Regulatory
reporting
Requirements, Rev 20 NuReg-1022, Event Reporting
Guidelines
10 CFR 50.72 and 10 CFR 50.73, Rev. 2 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 SPP-3.5 B.2.e. Question Source: For a given condition, determine
the regulatory
reporting
requirements
using appropriate
reference
material.
e. State the criteria requiring
eight hour notification
to the NRC. OPT200.ABVENT
B.5.f Describe the operation
of the AB Vent system as it relates to the following:
f. How a instrument
failure will affect system operation.
Bank# ___ _ Modified Bank # _X __ _ New ---Question History: Modified SQN question SPP-3.5 008 Question Cognitive
Level: Memory or fundamental
knowledge
_X __ Comprehension
or Analysis __ _ 10 CFR Part 55 Content: ( 41.5 / 43.5 / 45.3 /45.13 ) 10CFR55.43.b ( 5 ) Comments:
Modified SQN question SPP-3.5 008 Page 53 
SPP-3.5008
QUESTIONS
REPORT for BANK SQN Questions
During trouble shooting of O-RM-90-1
02, MIG accidently
initiated
an ABI by inducing a voltage transient.
Which one of the following
describes
the license reporting
requirements.
A. A four hour notification
and LER B. A one hour notification
and LER No reporting
required D. A LER This situation
describes
an invalid actuation
of ASI. Since invalid actuation
of ASI is one of the exceptions
to 50.72.b.2.ii, it is not reportable.
KIA: 2.1.10, 2.1.14 Tuesday, July 22, 2008 7:44:52 AM 1 
) paragraphs
as well as under 10 CFR 50.72(b){3)(v)
and 10 CFR 50.73(a)(2)(v) (event or condition
that could have prevented
the fulfillment
of the safety function of .... ). With regard to preplanned
actuations, operation
of a system as part of a planned test or operational
evolution
need not be reported.
Preplanned
actuations
are those which are expected to actually occur due to preplanned
activities
covered by procedures.
Such actuations
are those for which a procedural
step or other appropriate
documentation
indicates
the specific actuation
is actually expected to occur. Control room personnel
are aware of the specific signal generation
before its occurrence
or indication
in the control room. However, if during the test or evolution, the system actuates in a way that is not part of the planned evolution, that actuation
should be reported.
For example, if the normal reactor shutdown procedure
requires that the control rods be inserted by a manual reactor scram, the reactor scram need not be reported.
However, if unanticipated
conditions
develop during the shutdown that cause an automatic
reactor scram, such a reactor scram should be reported.
The fact that the safety analysis assumes that a system will actuate automatically
during an event does not eliminate
the need to report that actuation.
Actuations
that need not be reported are those initiated
for reasons other than to mitigate the consequences
of an event (e.g., at the discretion
of the licensee as part of a planned evolution).
Note that if an operator were to manually scram the reactor in anticipation
of receiving
an automatic
reactor scram, this would be reportable
just as the automatic
scram would be reportable.
Valid ESF-actuations
are those actuations
that result from "valid signals" or from intentional
manual initiation, unless it is part of a preplanned
test. Valid signals are those signals that are initiated
in response to actual plant conditions
or parameters
satisfying
the requirements
for initiation
of the safety function of the system. Note tRis definition
of "'Velid" refjuires
tRet the initietion
signel must be en [SF signel. TRis distinction
eliminetes
ectuetions
They do not include those which are the result of other signals from tRe elessof velid ectuetions.
Invalid actuations
are, by definition, those that do not meet the criteria for being valid. Thus, invalid actuations
include actuations
that are not the result of valid signals and are not intentional
manual actuations.
Except for critical scrams, invalid actuations
are not reportable
by telephone
under &sect; 50.72. In addition.
invalid actuations
are not reportable
under &sect; 50.73 in any of the following
circumstances: (A) The invalid actuation
occurred when the system is already properly removed from service. This means all requirements
of plant procedures
for removing equipment
from service have been met. It includes required clearance
documentation, equipment
and control board tagging, and properly positioned
valves and power supply breakers.
(8) The invalid actuation
occurred after the safety function has already been completed.
An example would be RPS actuation
after the control rods have already been inserted into the core. If an invalid ESF-actuation
reveals a defect in the ESF-system
so the system failed or would fail to perform its intended function,the
event continues
to be reportable
under other requirements
of 10 CFR 50.72 and 50.73. When invalid tsF-actuations
excluded by the conditions
described
49 NUREG-1022, Rev. 2 
') NPG Standard Regulatory
Reporting
Requirements
SPP-3.5 Programs and Rev. 0020 Processes
Page 14 of 64 5.0 DEFINITIONS (continued)
Incident Investigation
-Process conducted
by the NRC for the purpose of accident prevention.
The process includes gathering
and analyzing
information, determining
findings and conclusions, including
the cause(s) of a significant
operational
event; and the disseminating
of the investigation
results for NRC, industry, and public review. Independent
Spent Fuel Storage Installation (ISFSI) -A complex designed and constructed
for the interim storage of spent nuclear fuel and other radioactive
materials
associated
with spent fuel storage. An ISFSI which is located on the site of another facility licensed under &sect;Part 72 or a facility licensed under &sect;Part 50 (e.g., an operating
nuclear power plant) and which shares common utilities
and services with that facility or is physically
connected
with that other facility may still be considered
independent.
Initiation
of Shutdown -Physical act of reducing power or temperature
to change modes. Invalid Actuation (Signal) -Signals that do not meet the criteria for being valid. Invalid actuations
include instances
where instrument
drift, spurious signals, human error or other invalid signals that result in manual or automatic
actuation
of the systems listed in &sect;50.73(a)(2)(iv)(8).
Major Loss of Communication
-Constitutes
the loss of communication
capabilities.
Major Deficiency
-A condition
or circumstance
which under normal operating
conditions, an anticipated
transient, or postulated
design basis accident could contribute
to exceeding
a safety limit or cause an accident. "Major deficiency" also means a condition
or circumstance
which in the event of an accident due to other causes could, considering
an independent
single failure, result in a loss of safety function necessary
to mitigate the consequences
of the accident.
Natural Phenomenon
-Act of nature (e.g., fire, flood, tornado).
News Release -Known items which may be distributed
to the media (UPI, television, radio, newspaper, etc.) and those items identified
to be going on TVA news tape distributed
by the TVA Public Affairs Staff. Noncompliance (Failure To Comply) -A noncompliance
for the purposes of this procedure
means any failure to comply with the Atomic Energy Act of 1954, as amended, or with any applicable
rule or regulation
of the NRC relating to sUbstantial
safety hazards. A noncompliance
may be in operations, engineering, or construction
of the facility or basic component
thereof. Organization
Manager -This is the most senior manager available
who is in the same organization
as the individual
who discovered
the abnormal event. The senior manager is not normally interpreted
to be the plant manager or site vice president.
Preplanned
Sequence -Part of an approved procedure, including
workplans, work request, work orders, surveillance
instructions, general operating
instructions
and system operating
instructions.
Prevented
The Fulfillment
-Failure or possible failure of a safety system to properly complete a safety function. 
NPG Standard Regulatory
Reporting
Requirements
SPP-3.5 Programs and Rev. 0020 Processes
Page 15 of 64 5.0 DEFINITIONS (continued)
Principal
Safety Barrier -Fuel cladding, RCS pressure boundary, or the containment.
Redundant
Equipment
-Equipment, systems, structures
capable of performing
the same intended function within the same Technical
Specification
allowable
values. (In most cases, this means opposite train equipment.)
Safe Shutdown -Mode 3, as defined by the Technical
Specifications.
Safety Function -A component
or structure
designed to actuate upon receiving
the proper signal (ESF or RPS). Significant
Operational
Event -Any radiological, safeguards, or other safety-related
operational
event at an NRC licensed facility that poses an actual or potential
hazard to the public health and safety, property, or the environment.
These events or those that typically
result in a &sect;50.72 immediate
notification. (See Appendix A of this procedure)
A significant
operational
event also may be referred to as "an incident".
Examples of these events include: * Operations
that exceeded, or were not included in the design basis of the facility, * A major deficiency
in design, construction, or operation
having potential
generic safety implications, * A significant
loss of integrity
of the fuel, the primary coolant boundary, or the primary containment
boundary, * A loss of safety function or multiple failures in systems used to mitigate an actual event, * Significant
unexpected
system interactions, * Repetitive
failures or events involving
safety related equipment
or deficiencies
in operation, * Questions
or concerns pertaining
to licensee operational
performance.
Substantial
Safety Hazard -Loss of safety function to the extent that there is a major reduction
in the degree of protection
provided to public health and safety for any facility or activity licensed.
Threat -Physical hazard (e.g., fire, severe radioactive
release).
Unanalyzed
Condition
-Plant Condition
outside the bounds of the initial conditions
as described
in the FSAR accident analysis.
Valid Actuation (Signal) -Valid actuations
are those that result from "valid signals" or from intentional
manual initiation.
Valid signals are those that are initiated
in response to actual plant conditions
or parameters
satisfying
the requirements
for initiation
of the safety function of the system. 
NPG Standard Programs and Processes
Regulatory
Reporting
Requirements
Appendix A (Page 3 of 11) 3.1 Immediate
Notification
-NRC (continued)
SPP-3.S Rev. 0020 Page 19 of 64 3. &sect;50.72(b).(1))
-Any deviation
from the plant's Technical
Specifications
authorized
pursuant to &sect;50.54(x).
C. The following
criteria require 4-hour notification:
1. &sect;50. 72(b )(2)(i) -The initiation
of any nuclear plant shutdown required by the plant's Technical
Specifications.
2. &sect;50.72(b)(2)(iv)(A)
-Any event that results or should have resulted in Emergency
Core Cooling System (ECCS) discharge
into the reactor coolant system as a result of a valid signal except when the actuation
results from and is part of a pre-planned
sequence during testing or reactor operation.
3. &sect;50.72(b )(2)(iv)(8)
-Any event or condition
that results in actuation
of the reactor protection
system (RPS) when the reactor is critical except when the actuation
results from and is part of a pre-planned
sequence during testing or reactor operation.
4. &sect;50.72(b )(2)(xi) -Any event or situation, related to the health and safety of the public or onsite personnel, or protection
of the environment, for which a news release is planned or notification
to other government
agencies has been or will be made. Such an event may include an onsite fatality or inadvertent
release of radioactive
contaminated
materials.
D. The following
criteria require 8-hour notification:
NOTE The non-emergency
events specified
below are only reportable
if they occurred within three years of the date of discovery.
1. &sect;50.72(b )(3)(ii)(A)
-Any event or condition
that results in the condition
of the nuclear power plant, including
its principal
safety barriers, being seriously
degraded.
2. &sect;50.72(b)(3)(ii)(8)
-Any event or condition
that results in the nuclear power plant being in an unanalyzed
condition
that significantly
degrades plant safety. 3. &sect;50. 72(b )(3)(iv)(A)
-Any event or condition
that results in valid actuation
of any of the systems listed in paragraph (b)(3)(iv)(8)
[see list below], except when the actuation
results from and is part of a pre-planned
sequence during testing or reactor operation.
a. Reactor protection
system (RPS) including:
Reactor scram and reactor trip. 
') ) NPG Standard Programs and Processes
Regulatory
Reporting
Requirements
Appendix A (Page 5 of 11) SPP-3.5 Rev. 0020 Page 21 of 64 3.1 Immediate
Notification
-NRC (continued)
NOTE According
to &sect;50.72 (b)(3)(vi)
events covered by &sect;50.72(b)(3)(v)
may include one or more procedural
errors, equipment
failures, and/or discovery
of design, analysis, fabrication, construction, and/or procedural
inadequacies.
However, individual
component
failures need not be reported pursuant this paragraph
if redundant
equipment
in the same system was operable and available
to perform the required safety function.
5. &sect;50.72(b)(3)(xii)
-Any event requiring
the transport
of a radioactively
contaminated
person to an offsite medical facility for treatment.
6. &sect;50.72(b)(3)(xiii)
-Any event that results in a major loss of emergency
assessment
capability, offsite response capability, or offsite communications
capability (e.g., significant
portion of control room indication, emergency
notification
system, or offsite notification system).
E. Follow-up
Notification
(&sect;50.72(c>>
With respect to the telephone
notifications
made under paragraphs (a) and (b) [&sect;50.72 (a) and &sect;50.72 (b), respectively]
of this section [&sect;50.72], in addition to making the required initial notification, during the course of the event: a. Immediately
report (i) any further degradation
in the level of safety of the plant or other worsening
plant conditions
including
those that require the declaration
of the Emergency
Classes, if such a declaration
has not been previously
made; or (1) Any change from one Emergency
Class to another, or (2) A termination
of the Emergency
Class. b. Immediately
report (i) the results of ensuing evaluations
or assessments
of plant conditions, (1) The effectiveness
of response or protective
measures taken, and (2) Information
related to plant behavior that is not understood.
c. Maintain an open, continuous
communication
channel with the NRC Operations
Center upon request by the NRC. 
10 (B-3) Source Setpoint SER 760 (Unit 1 annunciator
system)
02 50 mr/hr > 1 second 0-RA-90-102A
FUEL POOL RAD MONITOR HIGH RAD Retransmitted
to U-2 SER 2243 (Unit 2 annunciator
system) Probable Causes Corrective
Actions 1. High radiation
in spent fuel pit area elevation
734. [1] IF dry cask storage processing
in progr*ess, THEN NOTIFY Cask Supervisor.
[2] CHECK 0-RM-90-1
02 and 0-RM-90-103
on 0-M-12 to verify alarm. NOTE 0-RM-90-102
will be blocked and is expected to alarm during portions of Dry Cask Storage loading/unloading
activities.
Under this condition, remaining
steps are N/A. [3] VERIFY the following:
a. Auxiliary
Building General Supply and Exhaust and Fuel Handling exhaust isolate (A-Train)
(1-M-9). b. Auxiliary
Building Gas Treatment
System starts (1-M-9). [4] IF high radiation
alarm valid, THEN [a] ANNOUNCE "High Radiation
at spent Fuel Pool Area" over PA system. [b] NOTIFY SM. [c] NOTIFY RADCON. [5] IF B-Train ABI has not actuated from a valid High Radiation
condition, THEN INITIATE manually B-Train Auxiliary
Building Ventilation
Isolation
via [1-HS-30-101B]
or [2-HS-30-101BJ (M-6). [6] IF fUel handling in the Spent Fuel Pit is in progress, THEN REFER TO AOP-M.04, Refueling
Malfunctions.
[7] REFER TO AOP-M.06, Loss of Spent Fuel Cooling. [8] IF Auxiliary
Building Ventilation
Isolation
resulted from an invalid ABI signal, THEN, REFER to 0-SO-30-10
Auxiliary
Building Ventilation
Systems to recover from ABI. (Continued
on next page) SQN 0-AR-M12-B
Page 12 of 40 o . Rev. 28 
Corrective
Actions (Continued)
References
10 (B-3) 0-RA-90-102A
FUEL POOL RAD MONITOR HIGH RAD [9] EVALUATE EPIP-1, Emergency
Plan Classification
Matrix. [10] EVALUATE Technical
Specifications
3.3.3.1 and 3.9.12. [11] INITIATE Corrective
Actions. [12] WHEN conditions
return to normal, THEN RETURN Auxiliary
8uilding Ventilation
System to normal in accordance
with 0-SO-30-10, Auxiliary
Building Ventilation
Systems. . 458655-12B-O,47W610-90-1
SQN O-AR-M12-B
Page 13 of 40 0 Rev. 28 
11 (B-4) Source Setpoint SER 761 (Unit 1 annunciator
system) O-RM-90-102
N/A O-RA-90-102B
FUEL POOL RAD MONITOR INSTR MALFUNC Retransmitted
to U-2 SER 2244 (Unit 2 annunciator
system) Probable Causes Corrective
Actions References
1. Instrument
downside ratemeter
trip. 2. Instrument
loss of power. 3. Instrument
placed in TRIP ADJ position.
[1] CHECK O-RM-90-102
on O-M-12 to attempt to determine
problem. [2] NOTIFY RADCON. [3] DISPATCH personnel
to check O-RM-90-1
02 locally to determine
problem. [4] EVALUATE Technical
Specification
3.3.3.1 and 3.9.12. [5] INITIATE WO for maintenance, if required.
458655-12B-0, 47W610-90-1
SQN 0 Page 14 of 40 O-AR-M12-8
Rev. 28 
I. PROGRAM: OPERATOR TRAINING OPT200.ABVENT
Rev. 1 Page 3 of94 II. COURSE: SYSTEMS TRAINING III. TITLE: AUXILIARY
BUILDING VENTILATION
SYSTEM IV. LENGTH OF LESSON: Initial License Training:
3 hour lecture; 1 hour simulator
demonstration;
Ihour self-study/workshop
V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of this lesson and others presented, the student should be able to apply the knowledge
to support satisfactory
performance
of the tasks associated
with the Auxiliary
Building Ventilation
systems (AB Vent) in the plant and on the simulator.
B. Enabling Objectives:
O. Demonstrate
an understanding
ofNUREG 1122 knowledge's
and abilities
associated
with the Auxiliary
Building Ventilation
system that are rated 2.5 during Initial License training for the appropriate
license position as identified
in Appendix A. 1. State the purpose/functions
of the Auxiliary
Building Ventilation
system as described
in the SQN FSAR. 2. State the design basis of the Auxiliary
Building Ventilation
system in accordance
with the SQN FSAR. 3. Explain the purpose/function
of each major component
in the flow path of the Auxiliary
Building Ventilation
system as illustrated
on the simplified
system drawing. 4. Describe the following
items for each major component
in the Auxiliary
Building Ventilation
system as described
in this lesson: a. Location b. Power supply (include control power as applicable)
c. Support equipment
and systems d. Normal operating
parameters
e. Component
operation
f. Controls g. Interlocks (including
setpoints)
h. Instrumentation
and Indications
i. Protective
features (including
setpoints)
j. Failure modes k. Unit differences
1. Types of accidents
for which the Auxiliary
Building Ventilation
system components
are designed m. Location of controls and indications
associated
with the Auxiliary
Building Ventilation
system in the control room and auxiliary
control room 
V. TRAINING OBJECTIVES (Cont'd):
B. Enabling Objectives (Cont'd):
5. Describe the operation
of the AB Vent system as it relates to the following:
a. Precautions
and limitations
b. Major steps performed
while placing the AB Vent system in service c. Alarms and alarm response d. How a component
failure will affect system operation
e. How a support system failure will affect AB Vent system operation
f. How a instrument
failure will affect system operation
6. Describe the administrative
controls and limits for the AB Vent system. 7. State Tech Specs/TRM
LCOs that govern the AB Vent system a. State the hour action limit TS LCOs OPT200.ABVENT
Rev. 1 Page 4 of94 b. Given the conditions/status
of the AB Vent system components
and the appropriate
sections of the Tech Spec, determine
if operability
requirements
are met and what actions are required 8. Discuss related Industry Events: Event Title: Hot Particle Discovered
on Auxiliary
Building Roof. INPO Event # 327-940131-1
VI. TRAINING AIDS: A. Computer.
B. Computer Display Projector
& Controls.
C. Local Area Network (LAN) Access. D. Simulator (if available) 
OPL271 SPP-3.5 Revision 1 Page 3 of 21 ) I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: REPORTING
REQUIREMENTS
IV. LENGTH OF LESSON/COURSE:
4 hour(s) V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the actions necessary
to comply with regulatory
and plant reporting
requirements.
B. Enabling Objectives
o. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with Regulatory
and Plant Reporting
Requirements
that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Perform a plant response assessment
using the 0-TI-QXX-000-001.0," Event Critique, Post Trip Report, and Equipment
Root Cause,". a. State the responsibilities
of each control room crew member. [C.1] b. Explain the process or Conduct a plant response assessment.
2. For a given condition, determine
the regulatory
reporting
requirements
using appropriate
reference
material.
a. List the tools available
to the operator for determining
regulatory
reporting
requirements.
b. Define the key terms used to determine
regulatory
reporting
requirements.
c. State the criteria requiring
one hour notification
of the NRC. d. State the criteria requiring
four hour notification
of the NRC. e. State the criteria requiring
eight hour notification
of the NRC. f. State the criteria requiring
24 hour notification
of the NRC. g. State the criteria requiring
2 day notification
of the NRC. h. State the criteria requiring
a written report or LER to the NRC. i. State the criteria allowing a telephone
notification
to be made in lieu of a written LER to the NRC. 3. For a given condition, determine
plant management
reporting
requirements
using SPP-3.5. 4. Complete a PER reportability
determination
per SPP-3.1. ---------I i i I 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
94. G 2.1.25 094 Given the following:
Unit 1 at 100% power with Boric Acid Tank (BAT) A aligned. RWST Boric Acid concentration
is 2575 ppm. Which ONE of the following
identifies
BAT A Volume and Boric Acid Concentration
that will meet Operability
requirements
and when the maximum expected boration capability
requirement
occurs in accordance
with the Technical
Requirement
Bases? REFERENCE
PROVIDED A. 9600 gallons at 6500 ppm; Near End of Life peak Xenon conditions.
B. 9600 gallons at 6500 ppm; Near Beginning
of Life peak Xenon conditions. 9200 gallons at 6775 ppm; Near End of Life peak Xenon conditions.
D. 9200 gallons at 6775 ppm; Near Beginning
of Life peak Xenon conditions.
Page 54 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRA CTOR ANAL YSIS: Page 55 A. Incorrect, BA T A volume of 9600 gallons at 6500 ppm with an RWST boron concentration
of 2575 ppm is in the unacceptable
region of TRM Figure 3.1.2.6 for BA T Tank Limits. Near the of life peak xenon is when the maximum expected boration capability
requirement
occurs as identified
in Technical
Requirement
bases. Plausible
because the BA T volume, BA T boron concentration, and the RWST boron concentration
are near the limit and require interpolation
when using the graph. Also, the higher level in the tank is not the correct answer and the limiting conditions
occurring
at near EOL is correct. B. Incorrect, BA T A volume of 9600 gallons at 6500 ppm with an RWST boron concentration
of 2575 ppm is in the unacceptable
region of TRM Figure 3.1.2.6 for BAT Tank Limits, and the Beginning
of life peak xenon is not when the maximum expected boration capability
requirement
occurs. Technical
Requirement
bases identifies
the requirement
to be at near EOL with peak Xenon conditions.
Plausible
because the BA T volume, BA T boron concentration, and the RWST boron concentration
are near the limit and require interpolation
when using the graph. Also, the higher level in the tank is not the correct answer and for other circumstances
the limiting conditions
occur at the beginning
of life. C. CORRECT, BA T A volume of 9200 gallons at 6775 ppm with an RWST boron concentration
of 2575 ppm is in the acceptable
region of TRM Figure 3.1.2.6 for BAT Tank Limits and the bases for the Technical
Requirement
states that the maximum expected boration capability
requirement
occurs at near EOL with peak Xenon conditions.
D. Incorrect, BA T A volume of 9200 gallons at 6775 ppm with an RWST boron concentration
of 2575 ppm is in the acceptable
region of TRM Figure 3.1.2.6 for BAT Tank Limits, but the Beginning
of life peak xenon is not when the maximum expected boration capability
requirement
occurs. Technical
Requirement
bases identifies
the requirement
to be at near EOL with peak Xenon conditions.
Plausible
because the combination
of BA T volume and boron concentration
place the tank in an acceptable
region of the graph and other limiting conditions
to occur at the beginning
of life. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 94 Tier 3 KIA G 2.1.25 Ability to interpret
reference
materials, such as graphs, curves, tables, etc. Importance
Rating: 3.9/4.2 Technical Reference: Technical
Requirements
Bases 3.1.2.6 Proposed references
to be provided to applicants
during examination:
TRM Figure 3.1.2.6, Boric Acid Tank Limits Based on RWST Boron Concentration
Learning Objective:
OPT200.TRM
B.4 Explain the TRM bases for each LCO (KIA 2.2.6) Question Source: Bank# ___ _ Modified Bank # ___ _ New _X __ Question History: New question for Sequoyah 2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X __ 10 CFR Part 55 Content: (41.10/43.5/45.12 ) 10CFR55.43.b ( 2) Facility operating
limitations
in the technical
specifications
and their bases. Comments:
New question for Sequoyah 2009 exam Page 56 
". .-*** " ----...
...... -...... "--..... -...
....... _ ..........
_._..,,,.
__ " **
****
J" __ * ) REACTIVITY
CONTROL SYSTEMS BORATED WATER SOURCES -OPERATING
LIMITING CONDITION
FOR OPERATION
TR 3.1.2.6* As a minimum, the following
borated water source(s)
shall be OPERABLE as required by TR 3.1.2.2: a. A boric acid storage system with: 1. A contained
volume of borated water in accordance
with Figure 3.1.2.6, 2. A boron concentration
in accordance
with Figure 3.1.2.6, and 3. A minimum solution temperature
of 63&deg;F. b. The refueling
water storage tank with: 1. A contained
borated water volume of between 370,000 and 375,000 gallons, 2. Between 2500 and 2700 ppm of boron, 3. A minimum solution temperature
of 60&deg;F, and 4. A maximum solution temperature
of 105&deg;F. APPLICABILITY:
MODES 1, 2, and 3. ACTION: a. With the boric acid storage system inoperable
and being used as one of the above required borated water sources, restore the storage system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent
to at least 1 % delta klk at 200&deg;F; restore the boric acid storage system to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 30 hours. b. With the refueling
water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following
30 hours. SEQUOYAH -UNITS 1 AND 2 TECHNICAL
REQUIREMENTS
3/41-8 January 4, 2001 Revision Nos. 13 
p TRM FIGURE 3.1.2.6 (Units 1 & 2) BORIC ACID TANK LIMITS BASED ON RWST BORON CONCENTRATION \ l.f') (
11000 'I. 1 ..
I I I
i{
I 1 10000 I I \ "'l. "" i ........ '" "' / \ ' I' I U)
I
J q\9 vJ \ J
... JAKNST=2700ppmE
J I I 9500 I I i
I I ,I 3
!' "\" "i:
'\ \ t -'{vo I 0 ! I' i"" .... ........ , > 9000 I I ,,' 'I \ I I I i ..... I I, Z I'! I ' .......::!
I )\ I I ! I ' ,I r-... I ........ I &sect; \ I \ I I I \',
I 8500' I, , '. '" I U I I I I, I I ,I ( 6990 ppm (Maximun)
J,r-...., I I '\ Ii! ! \' I
w i'-i ! ! !'"", .-8000 I I " I ; I I i \ \ 1
ppm
I \ I ! I I I I \ I I \ \ I I \ \ 7500 I, I, , ,I * I , \ I I I i I I I If REGION OF UNACCEPTABLE
OPERATION
i 7000 t I \ r l i I \ I \ II I i Ii I [i I i 1'1 I I --'-----_
.. _------. -. Indicated
values indude 1140 gal unusab e volume and 800 gal f r i
6000 6100 6200 6300 6400 6500 6600 6700 I '800 6900 7000 7100 BORIC ACID TANK CONCENTRATION
-BORON RWST Concentration l) ry '( -++-2600 PPM ---2650 PPM -'-2700 PPM 1 r -+-2500 PPM ---2550 PPM september
26. 2003 SEQUOYAH -UNITS 1 AND 2 TECHNICAL
REQUIREMENTS
3/41-10 Revision Nos. 13.26.27 
TRM FIGURE 3.1.2.6 (Units 1 &2) BORIC ACID TANK LIMITS ' .. "'. BASED ON RWST BORON CONCENTRATION
11000 1 1 I I I ! 1 I ! I I I
OF
OPERATION'
I I I I 10500, I "-I
-'-J500 B I I I I I I I I.! I I I I I J RWST = 2550 ppm B 1 I 10000 C/) I ...... " vi ' . I I K }t RWST = 2600 ppm B ) I i " ....... 1 I I f' lix VI ' I ' I , I RWST = 2650 ppm B J " I I z 0 ...J ...J << 9500 C> w :::> ...J 0 9000 > , z << l-e u << 8500 u 0 co e w I-8000 << 5:2 e z I 1 l' K ;Q'
= 2700 ppm E I ,! I I , i'.. 0 I I I I .... 1 1 t I I I , 0 I I , ,! I I I I I , I I "'I I 1 I 1 I i I i ! ..... 1 ....... 1 I 1 I I I I I i'l" I'tl I I I I I I I I I I , I [ 6990 ppm (Maximum)
J, I I I, I I I i I i 1 I ! I ! i I I 11 6120 ppm (Minimu r) I I I I I 1 ! I I ! 7500 I I I I I I I I I I I I I ! I I I I I i I I I I ! I REGION OF UNACCEPTABLE
OPERATION
I I 7000 I I I I I I I I I I I ! I , ! : I I I I I I I I i I I ! -Indicated
values indude 1140 gal unusable volume and 800 gal for instrument
error. 1>-1 I I I I I I I I I I I 6500 6000 6100 6200 6300 6400 6500 6600 6700 6800 6900 7000 7100 BORIC ACID TANK CONCENTRATION
-PPM BORON RWST Concentration
-+-2550 PPM -+E-'-2600
PPM -2650 PPM TECHNICAL
REQUIREMENTS
3/41-10 
REACTIVITY
CONTROL SYSTEMS ') BASES ) TRB 3/4.1.2 BORA TION SYSTEMS The boron injection
system ensures that negative reactivity
control is available
during each mode of facility operation.
The components
required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency
power supply from OPERABLE diesel generators.
With the RCS average temperature
above 350&deg;F, a minimum of two boron injection
flow paths are required to ensure single functional
capability
in the event an assumed failure renders one of the flow paths inoperable.
The boration capability
of either flow path is sufficient
to provide a SHUTDOWN MARGIN from expected operating
conditions
of 1.6% delta klk after xenon decay and cooldown to 200&deg;F. The maximum expected boration capability
requirement
occurs at near EOl from full power peak xenon conditions
and requires borated water from a boric acid tank in accordance
with Figure 3.1.2.6, and additional
makeup from either: (1) the common boric acid tank and/or batching, or (2) a minimum of 26,000 gallons of 2500 ppm borated water from the refueling
water storage tank. With the refueling
water storage tank as the only borated water source, a minimum of 57,000 gallons of 2500 ppm borated water is required.
The boric acid tanks, pumps, valves, and piping contairi a boric acid solution concentration
of between 3.5% and 4.0% by weight. To ensure that the boric acid remains in solution, the air temperature
is monitored
in strategic
locations.
By ensuring the air temperature
remains at 63&deg;F or above, a 5&deg;F margin is provided to ensure the boron will not precipitate
out. To provide operational
flexibility, if the area temperature
should fall below the required value, the solution temperature (as determined
by the pipe or tank wall temperature)
will be monitored
at an increased
frequency
to compensate
for the lack of solution temperature
alarm in the main control room. With the ReS temperature
below 350&deg;F, one injection
system is acceptable
without single failure consideration
on the basis of the stable reactivity
condition
of the reactor and the additional
restrictions
prohibiting
CORE ALTERATIONS
and operations
involving
positive reactivity
additions
that could result in loss of required SDM (Modes 40r 5) or boron concentration (Mode 6) in the event the single injection
system becomes inoperable.
Suspending
positive reactivity
additions
that could result in failure to meet minimum SDM or boron concentration
limit is required to assure continued
safe operation.
Introduction
of coolant inventory
must be from sources that haVe a boron concentration
greater than or equal to that required in the RCS for minimum SDM or refueling
boron concentration.
This may result in an overall reduction
in RCS boron concentration
but provides acceptable
margin to maintaining
subcritical
operation.
Introduction
of temperature
changes including
temperature
increases
when operating
with a positive MTC must also be evaluated
to ensure they do not result in a loss of required SDM. The boron capability
required below 350&deg;F, is suffiCient
to provide a SHUTDOWN MARGIN of 1.6% delta klk after xenon decayand cool down from 350"F to 200&deg;, and a SHUTDOWN MARGIN of 1% delta klk after xenon decay andcooldown
from 200&deg;F to 140&deg;F. This condition
requires either 6400 gallons of 6120 ppm borated water from the boric acid storage tanks or 13,400 gallons of 2500 ppm borated water from the refueling
water storage tank. The contained
water volume limits include allowance
for water not available
because of discharge
line location and other physical characteristics.
The 6400 gallon limit in the boric acid tank for Modes 4, 5, and 6 is based on 4,431 gallons required for shutdown margin, 1,140 gallons for the unusable volume in the heel of the tank, 800 gallons for instrument
error, and an additional
29 gallons due to rounding up. The 55,000 gallon limit in the refueling
water storage tank for modes 4, 5,and6 is based upon 22,182 SEQUOYAH -UNITS 1 AND 2 TECHNICAL
REQUIREMENTS
B 3/41-2 October 19, 2005 Revision Nos. 13,25,35,36 
REACTIVITY
CONTROL SYSTEMS ') BASES )\ ) , ') gallons that is undetectable
due to lower tap location, 19,197 gallons for instrument
error, 13,400 gallons required for shutdown margin, and an additional
221 gallons due to rounding up. The limits on contained
water volume and boron concentration
of the RWST also ensure a pH value of between 7.5 and 9.5 for the solution recirculated
within containment
after a LOCA. This pH band minimizes
the evolution
of iodine and minimizes
the effect of chloride and caustic stress corrosion
on mechanical
systems and components.
The OPERABILITY
of one boron injection
system during REFUELING
ensures that this system is available
for reaCtivity
control while in MODE 6. SEQUOYAH -UNITS 1 AND 2 TECHNICAL
REQUIREMENTS
B 3/41-3 October 19,2005 Revision Nos. 13 
I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: TECHNICAL
REQUIREMENTS
MANUAL IV. LENGTH OF LESSON/COURSE:
2 hour(s) V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of this lesson and others presented, the student shall demonstrate
an understanding
of the purpose and background
of the Technical
Requirements
Manual by successfully
completing
a written examination
with a score of 80%. B. Enabling Objectives:
Slide 2 In order to accomplish
these objectives, the student shall be able to successfully:
1. Determine
plant mode of operation
from memory. (KIA 2.1.22) 2. Apply TRM action statements
of greater than one hour given a copy of the TRM (KIA 2.1.12, 2.2.22) 3. Explain the process for making changes to the TRM (KIA 2.2.6) 4. Explain the TRM bases for each LCO. (KIA 2.2.25) 5. Apply less than one hour LCO actions using a copy of the TRM. (KIA 2.1.11) 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
95. G 2.2.21 095 Given the following:
-During the Operations
review of a Work
the SRO determines
the return to operability (RTO) Post Maintenance
Test (PMT) needs revision.
-The WO did not require an Independent
Qualified
Review (IQR), a 10CFR50.59
review, or 10CFR72.48
review. In addition to documenting
the reason for the revision, which ONE of the following
identifies
both ... (1) the requirement
to revise the RTO test without routing the WO back through the review process and (2) when this type PMT revision would require Shift Manager approval?
A'! (1) Revision signed by 2 SROs. (2) when a generation
risk critical component
is involved.
B. (1) Revision signed by 2 SROs. (2) when a configuration
change to a component
on the unit is required to perform the PMT. C. (1) Revision signed by the Work Week Manager and an SRO. (2) when a generation
risk critical component
is involved.
D. (1) Revision signed by the Work Week Manager and an SRO. (2) when a configuration
change to a component
on the unit is required to perform the PMT. Page 57 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRACTOR ANAL YSIS: Page 58 A. CORRECT, SPP-6.3 allows revisions
to RTO tests to be made by 2 SROs without routing back through the review cycle for the conditions
identified
in the stem of the question and does require Shift Manager approval if a generation
critical component
is involved.
B. Incorrect, SPP-6.3 allows revisions
to RTO tests to be made by 2 SROs without routing back through the review cycle for the conditions
identified
in the stem of the question but does not require Shift Manager approval because a configuration
change on the unit would be involved.
Plausible
because 2 SRO being required to make the revision is correct and verification
of configuration
changes are identified
in the SPP as needing independent
verification.
C. Incorrect, SPP-6.3 allows revisions
to RTO tests to be made by 2 SROs, not an SRO and the Work Week Manager but Shift Manager approval if a generation
critical component
is involved is correct. Plausible
because the Work Week Manager has other functions
and requiring
Shift Manager approval if a generation
critical component
is involved is correct. D. Incorrect, SPP-6.3 allows revisions
to RTO tests to be made by 2 SROs, not an SRO and the Work Week Manager and Shift Manager approval because a configuration
change on the unit would be involved is also not correct. Plausible
because the Work Week Manager has other functions
and verification
of configuration
changes are identified
in the SPP as needing independent
verification. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 95 Tier 3 KIA G 2.2.21 Knowledge
of pre-and post-maintenance
operability
requirements.
Importance
Rating: 2.9/4.1 Technical
Reference:
SPP-6.3, Pre-/ Post-Maintenance
Testing, Rev 2 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
No training objective
identified
Question Source: Bank # ----Modified Bank # ----:-_____ --New X ---Question History: New question for Sequoyah 2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
_X __ Comprehension
or Analysis __ _ 10 CFR Part 55 Content: (41.10/43.2 ) 10CFR55.43.b ( 2,3 ) Comments:
New question for Sequoyah 2009 exam Page 59 
TVAN STANDARD PROGRAMS AND PROCESSES
PRE-/POST -MAINTENANCE
TESTING SPP-6.3 Rev. 2 Page 9 of 16 3. A formal instruction
prepared for the WO PMT receives an IQR, 10 CFR 50.59 and/or 10 CFR 72.48 review in accordance
with SPP-2.2, "Administration
of Site Technical
Procedures." System Engineer F. Review WO PMTs (Maintenance
Testing and RTO Testing) when requested
for TS, ASME Section XI, 10 CFR 50 Appendix J, or other complex activities
to ensure that they are correct and concur with the PMT. Operations
Shift Manager/SRO
Designee G. Perform initial reviews of WO in accordance
with the criteria specified
in SPP-7.1, "Work Control Process." H. Review/revise
RTO tests as necessary
to ensure TS operability
and surveillance
requirements
are met without imposing any adverse affects on the system or equipment.
Contact System Engineering
if assistance
is needed. Changes to RTO tests are made in accordance
with 3.4.N below or by revision to the work order. I. Approve the RTO tests. J. Ensure that the PMT satisfies
all TS requirements.
K. Include status control requirements
in the WO package. Specify the correct systems and component
alignment
required for the PMT performance
and the configuration
required to restore systems and components
to ensure correct operating
or standby mode following
the completion
of the PMT. L. Ensure that the PMTs are performed
at the appropriate
system operating
conditions
or plant modes. M. Review WO scope changes and revise RTO tests as necessary
in accordance
with 3.4.N below or by revision to the work order. N. Revise RTO tests when warranted.
Revisions
to RTO tests may be made by two SROs without routing back through the review cycle provided that 1) the reason is documented, 2) both SROs sign the revision and 3) IQR, 10 CFR 50.59 and/or 10 CFR 72.48 review was not required previously.
This type revision requires Shift Manager approval if the PMT involves a generation
risk critical component
as classified
in MEL. O. Waive PMT when warranted.
A PMT requirement
may be waived provided that the Shift Manager authorizes
the waiver, the reason is documented (e.g., plant conditions
prevent functional
testing of a logic circuit), and affected configuration
changes are verified by independent
verification. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
96. G 2.2.6 096 Given the following:
Unit 1 is in MODE 5 during a refueling
outage. It is determined
that 1-S0-63-1, "Cold Leg Injection
Accumulators" needs an Urgent "Minor/Editorial
Change" revision to support system alignment
changes due to system modification.
Which ONE of the following
statements
identifies
the need for an Indepenent
Qualified
Reviewer (IQR) and if the Unit SRO can be the approval authority
for the procedure
revision?
A'! An IQR is required.
The Unit SRO may sign as the approval authority.
B. An IQR is required.
The Unit SRO may NOT sign as the approval authority.
C. An IQR is NOT required.
The Unit SRO may sign as the approval authority.
D. An IQR is NOT required.
The Unit SRO may NOT sign as the approval authority.
DIS TRA CTOR ANAL YSIS: Page 60 A. CORRECT, SPP-2.2 requires an IQR review for minor editorial
changes for Quality Related procedures
and 1-S0-63-1
is a Quality Related procedure
the SPP also provides for the SRO on the Unit being the approval authority.
B. Incorrect, IQR review for minor editorial
changes for Quality Related procedures
is required and SPP-2.2 allows for the SRO on the Unit being the approval authority.
Plausible
because the IQR review is required for minor/editorial
changes to quality related procedures
and the SRO on the unit is not the approval authority
for normal procedure
revisions.
C. Incorrect, SPP-2.2 requires IQR review for minor editorial
changes for Quality Related procedures
and allows for the SRO on the Unit being the approval authority.
Plausible
because the IQR review is not required for minor/editorial
changes to non-quality
related procedures
and the SPP allows for the SRO on the unit being the approval authority.
D. Incorrect, IQR review for minor editorial
changes for Quality Related procedures
is required and allows for the SRO on the Unit being the approval authority.
Plausible
because the IQR review is not required for minor/editorial
changes to non-quality
related procedures
and the SRO of the unit is not the approval authority
for normal procedure
revisions. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 96 Tier 3 KIA G 2.2.6 Knowledge
of the process for making changes to procedures.
Importance
Rating: 3.0/3.6 Technical
Reference:
SPP-2.2, Administration
of Site Technical
Procedures, Rev 15 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271.SPP-2.2
B.5 Describe the procedure
revision process. Question Source: Bank# ___ _ Modified Bank # _X, ___ _ New ---Question History: Question modified from Summer 2006 exam Question Cognitive
Level: Memory or fundamental
knowledge
_X __ Comprehension
or Analysis __ _ 10 CFR Part 55 Content: ( 41.10/43.3/45.13 ) 10CFR55.43.b
(3,5) Comments:
Question modified from Summer 2006 exam Page 61 
) ,\ NPG Standard Administration
of Site Technical
SPP-2.2 Programs and Procedures
Rev. 0015 Processes
Page 20 of 42 3.4.4 Comments, Approval & Implementation
A. Reviewers
should provide comments in BSL, or Form SPP-2.2-2, "Site Technical
Procedure
Review, Comment, and Concurrence
Form", or by other appropriate
means. B. Preparers
shall resolve comments and escalate unresolved
comments to appropriate
management.
C. The responsible
organization
obtains approval from the appropriate
approval authority.
The approval also ensures, as applicable, the reviewer is independent
of the preparer and qualified
to perform the review. Approval by telecon with the manager/supervisor
who is the designated
approval authority
is acceptable
for hard copy changes. D. If required, PORC shall review and concur with the procedure.
The PORC Chairman recommends
to the Plant Manager his approval.
Upon concurrence, the Plant Manager shall approve the procedure.
NOTES 1) PORC review is required for all procedures
identified
as a CIPTE and all revisions
to those procedures.
2) Minor/editorial
revisions
do not require PORC review. E. Following
approval, the responsible
manager establishes
the effective
date taking into consideration
any work in progress or parallel changes, training, etc., identified
and transmits
the procedure
package to MS for distribution
and EDM archival.
The effective
date is normally three working days after receipt by MS. NOTE The 10 CFR 50.59 and 10 CFR 72.48 documents
will be archived in EDM as standalone
documents.
F. Revisions
which will cause Operations
to change the normal configuration
of a component (i.e., changes to normal valve alignment
checklists, switch position checklists, etc.), should be communicated
to the affected Shift Manager or Unit Supervisor
prior to the effective
date. 3.5 Minor/Editorial
Changes A. Minor changes, such as inconsequential
editorial
corrections
that do not affect the outcome, results, functions, processes, responsibilities, and requirements
of the performance
of procedure
or instructions, require review by an lOR (quality-related
procedures
only) and approval by the appropriate
approval authority.
Minor changes do not require 10 CFR 50.59 review, 10 CFR 72.48 review, or PORC review. Minor changes shall not change the intent of the procedure
or alter the technical
sequence of procedural
steps. 
NPG Standard Administration of
Site Technical
SPP-2.2 Programs and Procedures
Rev. 0015 Processes
Page 21 of 42 3.5 Minor/Editorial
Changes (continued)
B. Procedure
changes that meet the following
criteria are considered
minor changes: 1. correction
of punctuation, style changes 2. redundant
or insignificant
word or title changes 3. correction
of typographical
errors including
capitalization
4. annotation
of critical steps, 5. correction
of reference
errors 6. omitted symbols that
do not alter results 7. incorrect
units of measure due to editorial
error 8. misplaced
decimals that are neither setpoint values nor tolerances
g. page number discrepancies
10. missing sign-offs, signatures, or date lines 11. corrections
to attachment
identifiers
12. corrections
to titles of plant organizations, position titles, departmenUsection/unit
names when there is no change in authority, responsibility, or reporting
relationships.
13. corrections
to addresses, telephone
numbers, or computer application
names 14. corrections
to or additions
of equipment
nomenclature
or locations
in procedures
to be consistent
with approved drawings, documents, labels, or procedure
content. 15. addition of or changes to equipment
unique identifier
information (UNID) in procedures
consistent
with design output documents
and which do not alter what component
is operated 16. corrections
to or clarification
of a note or precaution
which does not alter
the method of accomplishing
a task 17. changes which are purely administrative
and non-technical
in nature which do not change the intent or outcome of an activity (e.g. adding a step requiring
a log entry, a plant announcement, informational
notifications, or initiation
of a PER) C. A BSL System Administrator
or Sponsor may make the following
changes: 1. Organizational
changes. 2. Reference
changes, e.g., MMI is superseded
by MCI-0-000-TRB001. 
NPG Standard Administration
of Site Technical
SPP-2.2 Programs and Procedures
Rev. 0015 Processes
Page 22 of 42 3.5 Minor/Editorial
Changes (continued)
3. Misspelled
words. 4. Language, grammar, syntax corrections.
D. The revision description
will describe the reason for change. E. The approval authority
will obtain the lOR for quality-related
procedures
and approve the change. F. The procedure
package is transmitted
to MS for distribution
and EDM archival.
3.6 Urgent Procedure
Changes Urgent changes to procedures
are revisions
which are deemed necessary
by plant management
to maintain plant safety, operability
or critical schedules
and inadequate
time exists to make a normal revision using BSL. Urgent changes may be handwritten
and require the following:
A. Tracking Numbers from BSL. B. Affected organization/Cross
Discipline
Review (see Sections 3.4.2D, 3.4.3C and 3.4.3D for review determination/requirements)
when other organizations
are affected by the change and obtain necessary
signatures
on the PCF. C. Technical
review by an lOR. D. A 10 CFR 50.59 screening
review, if required, in accordance
with SPP-9.4, "10 CFR 50.59 Evaluations
of Changes, Tests, and Experiments" or 10 CFR 72.48 Screening
Review if required, in accordance
with SPP-9.9, "10 CFR 72.48 Evaluations
of Changes, Tests, and Experiments
for Independent
Spent Fuel Storage Installation." E. Approval by the PORC, if required, and the approval authority.
A licensed SRO on the unit affected may sign as the approval authority
for minor/editorial
changes. F. The preparer shall obtain applicable
reviews and approval documented
on PCF. The preparer shall submit a hardcopy to the affected unit control room, if filed in the control room. The preparer shall forward the original to MS. The procedure
may then be used to perform work. G. The preparer shall ensure the change is processed
and provided to MS for distribution.
The preparer shall submit a copy to the sponsor by the next working day. H. The responsible
organization
should ensure the handwritten
change is incorporated
and processed
into BSL within 14 days for changes initiated
during normal plant operation
or within 30 days following
an outage. 
NPG Standard Administration
of Site Technical
SPP-2.2 Programs and Procedures
Rev. 0015 Processes
Page 23 of 42 3.7 One-Time-Only
Procedures
and Revisions
One-Time-Only (OTO) procedures, and one-time revisions
to existing procedures, can be developed
and approved for use where existing procedures
do not adequately
address the activity due to unusual plant conditions.
Such procedures/revisions
are intended to be used only once and will automatically
expire following
completion
of the use of the procedure.
A. OTO procedures/revisions
shall be reviewed and approved as specified
in Sections 3.4, 3.5, through 3.6 of this procedure.
The preparer shall indicate that the revision is a OTO revision on the PCF. The tracking number on the PCF should be N/A'd. B. The performer
shall ensure that the revision is inserted/incorporated
into the controlled
copy of the procedure
being used for performance
of the activity.
OTO procedures/revisions
will not be distributed
to other controlled
manuals. C. The completed
procedure, PCF, and all other supporting
documents
are transmitted
to MS to be retained as a record. D. OTO procedures/revisions do
not require cancellation.
Following
completion
of performance, the active revision existing before the OTO revision was made will be the active version of the procedure.
3.8 Administrative
Hold A. Responsible
organizations
are required to place procedures
on "administrative
hold" when a revision is not feasible.
The cover sheet of the procedure
should indicate administrative
hold. The responsible
procedure
sponsor shall forward the PCF (or audit trail) and the cover sheet to MS for placement
on Administrative
Hold. B. MS shall not issue working copies of procedures
on administrative
hold. C. Employees
shall not use procedures
on administrative
hold to perform work. Only the Plant Manager/Duty
Plant Manager can authorize
issuance of a working copy of a document that is on administrative
hold. D. All reasonable
measures shall be exhausted
to release (if applicable)
the procedure
from administrative
hold through the procedure
revision process. The Plant Manager should carefully
evaluate the situation
before releasing
the procedure
for work. E. When the reason for the hold has been resolved, the sponsor shall initiate a procedure
revision (Section 3.4) to request release from administrative
hold. Include the reason for administrative
hold removal in the revision log. F. The following
procedures, due to their usage should have deficiencies
corrected
immediately, and should not be placed on administrative
hold: 1. Abnormal Operating
Instructions.
2. Emergency
Operating
Instructions. (E, ECA, FRG, ES). 3. Emergency
Preparedness
Implementing
Procedures. 
QUESTIONS
REPORT for summer 10-4-06 II G2.2.6001
The RO is attempting
to vent the PRT in accordance
with SOP-101, Reactor Coolant System, when he notes that a valve not identified
in the SOP needs to be open to complete venting. The CRS determines
that a restricted
change is required.
Which ONE (1) of the following
is the MINIMUM review/approval
necessary
for this procedure
change? a. Manager, Operations.
b. Duty Shift Supervisor.
c ..... Qualified
Reviewer AND the Duty Shift Supervisor.
d. Procedure
Group Supervisor
AND the Plant Manager. Requires examinee to determine
that a restricted
change is classified
as a temporary
change and then applies this to requirements
in SAP-139. SAP-139 requires a Qualified
Reviewer and the Duty Shift Supervisor
as minimum review/approval
for approval of a Operations
restricted
procedure
change. Other Discipline
Supervisors
may approve changes within their discipline.
Knowledge
of the process for making changes in procedures
as described
in the safety analysis report. Question Number: Tier 3 Group 2 Importance
Rating: 3.3 SR021 Technical
Reference:
SAP-139 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
4378 Question Source: Bank Question History: VCS Bank Question Cognitive
Level: Lower 10 CFR Part 55 Content: 41 Comments:
NRC Comment: Question appears to match KIA, and is SRO knowledge.
SAT BANK 1 OCFR55.43(b)
is met because the applicant
must understand
the approval process for changes made to facility procedures.
Friday. July 18. 2008 2:43:38 PM 1 
OPL271 SPP-2.2 Revision 1 Page 3 of 31 ) I. PROGRAM: OPERATOR TRAINING ) ) II. COURSE: LICENSE TRAINING III. LESSON TITLE: SPP-2.2, ADMINISTRATION
OF SITE TECHNICAL
PROCEDURES
IV. LENGTH OF LESSON/COURSE:
1 Hour V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, an understanding
of SPP-2.2 "Administration
of Site Technical
Procedures" and OPDP-1 Attachment
F "Plant Operating
Procedures." B. Enabling Objectives:
o. Demonstrate
an understanding
of NUREG 1122 Knowledge's
and Abilities
associated
with Rules of Procedure
Use that are rated;;:::
2.5 during Initial License Training and;;::: 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A 1. Discuss the purpose of SPP-2.2. 2. Discuss management
philosophy
for procedure
use and adherence.
3. List the four levels of use for technical
procedures
including
examples of each level. 4 Give examples when a procedure
step may be marked not applicable (N/A). 5 Describe the procedure
revision process. 6 Briefly define a minor/editorial
change. 7 Describe the conditions
under which personnel
may take reasonable
action within the scope of their training that departs from procedure.
8 Explain how to obtain and verify controlled
procedure
copy. i 
QUESTIONS
REPORT for additional
Questions
G 2.3.13 197 iven the following:
Unit 2 was shutdown for refueling
on 01/20109 at 2400. Today is 01/25/09 at 0200 and the Unit is in Mode 5. Lower cavity fill has not been started Transfer Tube Wafer Valve, 2-78-610 is required to be open for repair on the transfer tube and transfer system. In accordance
with 0-GO-9, "Refueling
Operations", which ONE of the following
identifies
two conditions
where either one of the two could be used to minimize the potential
spread of airborne contamination?
A. Containment
Equipment
Hatch ... open Containment
Purge System ... in service and aligned to upper containment
B:t open stopped. C. D. held closed by a minimum of 4 bolts. held closed by a minimum of 4 bolts. in service and aligned to upper containment.
stopped. DISTRACTOR
ANAL YSIS: A. Incorrect, containment
equipment
hatch can be opened but the containment
prurge being in service would not be a method to reduce the pressure.
B. CORRECT, In accordance
with O-GO-9 Precaution
S, two of the four methods to prevent the potential
spread of airborne radiation
due to prevent excessive
air flow through the transfer tube due to a differential
pressure between the containment
and the aux building is for the Equipment
Hatch to be open and for the containment
purge being stopped. C. Incorrect, containment
hatch being closed is not one of the method to identified
to prevent the potential
spread of airborne radiation.
D. Incorrect, containment
hatch being closed is not one of the method to identified
to prevent the potential
spread of airborne radiation.
Monday, December 15, 2008 9:38:45 AM 13 
Question No. 97 Tier 3 KIA G 2.3.13 QUESTIONS
REPORT for additional
Questions
Knowledge
of radiological
safety procedures
pertaining
to licensed operator duties, such as response to radiation
monitor alarms, containment
entry requirements, fuel handling responsibilities, access to locked high-radiation
areas, aligning filters, etc. Importance
Rating: 3.4 /3.8 Technical
Reference:
0-GO-9, Refueling
Operations, Rev 34 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 C259 B.2 Question Source: Describe the basis of each precaution, limitation
and prerequisite
in this procedure
Bank # ----Modified Bank # ----New X ---Question History: New question for Sequoyah 2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
_X __ Comprehension
or Analysis __ _ 10 CFR Part 55 Content: ( 41.12 /43.4 / 45.9 / 45.10 ) 10CFR55.43.b ( 7 ) Comments:
New question for Sequoyah 2009 exam Monday, December 15, 2008 9:38:45 AM 14 
O-GO-9 SQN REFUELING
PROCEDURE
Rev: 34 1&2 Page 9 of 69 3.1 PRECAUTIONS (Continued)
S. Due to relative air pressure differences
between the Containment
and Auxiliary
Buildings
which causes excessive
air flow through the transfer tube, if the Transfer Tube Wafer Valve 1,2-78-610
is open, at least one of four conditions
should exist to minimize the potential
spread of airborne contamination:
Reference
TS 3.9.4. 1. The Containment
equipment
hatch stays open. This is not permitted
during movement of recently irradiated
fuel within containment (less than 100 hours since shutdown).
2. A Containment
airlock is breached open (both doors) (not permitted
during fuel movement within the containment
unless one train of ABGTS is operable and one door of each breach airlock is capable of closure).
3. The lower reactor cavity has water filled to a level above the transfer tube elevation.
4. Containment
ventilation
purge supply/exhaust
are stopped and measured differential
pressure is approximately
0.25 psid or less between Computer Points P1001A to P1002A. T. The use of the 1 ton Jib Crane in Unit One Containment
is limited to Modes 6 and defueled conditions.
The 1 ton Jib Crane will not be used during fuel handling operations
or when there is the potential
for interference
with other cranes operation
in the vicinity.
3.2 LIMITATIONS
A. Do not allow reactor vessel head to come into contact with the refueling
water. B. The fuel shall not be moved in the core unless the water visibility
is adequate to allow the operator to see the top nozzle of the seated fuel assembly for core unload, and lower core plate pin holes for core reload. C. Pressurizer
manway must be open with airflow unobstructed
whenever reactor head is in place and S/G U-tubes drained or pressurizer
level off scale low. This ensures that adequate RCS vent exists to allow gravity fill from the RWST on a SBO event without natural circulation
capability.
This requirement
does not apply when closing RCS in preparation
for RCS vacuum refill (0-GO-13 section 5.3.4) 
REFUELING
OPERA nONS ) 3/4.9.4 CONTAINMENT
BUILDING PENETRATION?
LIMITING CONDITION
FOR OPERATION
3.9.4 The containment
building penetrations
shall be in the following
status: a. The equipment
door closed and held inpJace by a minimum of four bolts, b. A minimum of one door in each airlock is closed, and both doors of both containment
personnel
airlocks may be open if: 1. One personnel
airlock door in each airlock is capable of closure, and 2. One train of the Auxiliary
Building Gas Treatment
System is OPERABLE in accordance
with Technical
Specification
3.9.12, and c. Each penetration*
providing
direct access from the containment
atmosphere
to the outside atmosphere
shall be either: 1. Closed by an isolation
valve, blind flange, manual valve, or equivalent, or 2. Be capable of being closed by an OPERABLE automatic
Containment
Ventilation
isolation
valve. APPLlCABIL
TY: 3.9.4.a. Containment
Building Equipment
Door -During movement of recently irradiated
fuel within the containment
3.9.4.b. and c. Containment
Building Airlock Doors and Penetrations
-During movement of irradiated
fuel within the containment
ACTION: n 1. With the requirements
of the above specification
not satisfied
for the containment
building equipment
door, immediately
suspend all operations
involving
movement of recently irradiated
fuel in the containment
building.
The provisions
of Specification
3.0.3 are not applicable.
2. With the requirements
of the above specification
not satisfied
for containment
airlock doors or penetrations, immediately
suspend all operations
involving
movement of irradiated
fuel in the containment
building.
The provisions
of Specification
3.0.3 are not applicable.
SURVEILLANCE
REQUIREMENTS
4.9.4 Each of the above required containment
building penetrations
shall be determined
to be either in its required condition
or capable of being closed by an OPERABLE automatic
Containment
Ventilation
isolation
valve once per 7 days during movement of irradiated
fuel in the containment
building by: * a. Verifying
the penetrations
are in their required condition, or b. Testing the Containment
Ventilation
isolation
valves per the applicable
portions of Specification
4.6.3.2. Penetration
flow path(s) providing
direct access from the containment
atmosphere
that transverse
and terminate
in the Auxiliary
Building Secondary
Containment
Enclosure
may be unisolated under
administrative
controls.
SEQUOYAH -UNIT 1 3/49-4 October 28, 2003 Amendment
No. 12, 209, 249, 260, 288 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
97. G 2.3.13 097 Given the following:
The Unit 1 Elevation
690' pipe chase is locked in accordanceAith
its normal radiological
posting. Conditions
require Operations
to make an emergency
e pipe chase. Which ONE of the following
identifies
how the room is the "key control" for unlocking
the pipe chase? A. Very High Radiation
Area; Both Rad Ops and the Shift Manager have Jio ntro I of keys to the lock. B. Very High Radiation
Area; ONLY Rad Ops has a key, the Shift &#xa5;Bnager would contact Rad Ops to open the lock. C!" Locked High Radiation
Area; Both Rad Ops and the Shift
have control of keys to the lock. D. LockedHigh
Radiation
ONLY Rad Ops has a kj!y, the Shift Manager would contact Rad Ops to open the lock. Page 62 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DISTRACTOR
ANAL YSIS: Page 63 A. Incorrect, The 690 pipe chase on both units is not a Very High Radiation
Area (VHRA) but a key being located in the MCR under the administrative
control of the Shift Manager is correct. Plausible
because VHRA do exist and the Shift Manager does have access to a key located in the MCR. B. Incorrect, The 690 pipe chase on both units is not a Very High Radiation
Area (VHRA) and Rad Ops does not have not only key. Plausible
because if the area had been a VHRA then only Rad Ops would have a key. C. CORRECT, The 690 pipe chase on both units is a Locked High Radiation
Area (LHRA). RCI-29 identifies
the area as such and designates
2 key locations.
One in the Rad Ops Lab and the other in the MCR under the administrative
control of the Shift Manager. D. Incorrect, The 690 pipe chase on both units is a Locked High Radiation
Area (LHRA). Rad Ops is not the only location of a key. Another key is located in the MCR under the administrative
control of the Shift Manager. Plausible
because if the area is a locked high radiation
area and for all other LHRA only Rad Ops has a key. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 97 Tier 3 KIA G2.3.13 Knowledge
of radiological
safety procedures
pertaining
to licensed operator duties, such as response to radiation
monitor alarms, containment
entry requirements, fuel handling responsibilities, access to locked high-radiation
areas, aligning filters, etc. Importance
Rating: 3.4 /3.8 Technical
Reference:
RCI-29, Control of Radiation
Protection
Keys, Rev 7 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 C259 B.1.g Question Source: Identify the requirements
for entering and working in the following
areas: g. Locked High Radiation
Area Bank# ___ _ Modified Bank # ----New X ---Question History: New question for Sequoyah 2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
_X __ Comprehension
or Analysis __ _ 10 CFR Part 55 Content: ( 41.12 / 43.4 / 45.9 / 45.10 ) 10CFR55.43.b ( 4 ) Comments:
New question for Sequoyah 2009 exam Page 64 
TENNESSEE
VALLEY AUTHORITY
SEQUOYAH NUCLEAR PLANT RADIOLOGICAL
CONTROL INSTRUCTION
RCI-29 CONTROL OF RADIATION
PROTECTION
KEYS Revision 7 QUALITY RELATED PREPARED BY: Terry F. Johnston RESPONSIBLE
ORGANIZATION:
Radiation
Protection
APPROVED BY: James S. McCamy EFFECTIVE
DATE: 09/12/2007
VALIDATION
DATE: NIA LEVEL OF USE: INFORMATION
ONLY REVISION DESCRIPTION:
This revision is generated
to provide that Operations
shall maintain LHRA keys for the Unit 1 and Unit 2 EI. 690' Pipe Chases in a Main Control Room key box under the administrative
control of the Shift Manager. 
) ) ) RCI-29 SQN Control of Radiation
Protection
Keys Revision 7 Page 3 of 6 6.0 REQUIREMENTS
Note Use of the term 'keys' in the text of this Instruction
is understood
to refer to LHRA and VHRA keys interchangeably.
unless specifically
noted otherwise.
6.1 General A. Rad Ops will maintain positive control of LHRA and VHRA keys. B. Rad Ops shall initiate a Work Order (WO), as necessary, for the installation
or removal of a LHRA tumbler lock. C. Rad Ops shall install or remove LHRA or VHRA security locks. 6.2 Key Controls [C.2] [C.3] [C.4] A. Keys controlled
by this Instruction
shall be maintained
in locked key boxes in the EI. 690' Rad Ops Lab in the following
configuration:
1. There is a designated
LHRA key box that contains keys for tumbler locks and security locks that are in current use for active LHRAs only. This key box is kept locked at all times. It is opened each shift for the performance
of a key inventory
and is opened when issuing and returning
LHRA keys. 2. There is a designated
VHRA key box that contains keys for security locks that are in current use for active VHRAs only. This key box is kept locked at all times. Additionally, this key box has a wire tamper-proof
security seal installed.
It is only opened to issue or return VHRA keys. B. The Operations
section shall maintain one LHRA key each for the Unit 1 and Unit 2 EI. 690' Pipe Chases. 1. These LHRA keys shall be maintained
in the MCR in a locked break-glass-to-access
key box, under the administrative
control of the Shift Manager. 2. These LHRA keys are for emergency
use only and Operations
shall immediately
notify the Rad Ops Lab if these keys are used. 3. Rad Ops shall maintain the key to the Ops LHRA key box in the Rad Ops Lab LHRA key box. 4. If during the course of normal duties the Operations
LHRA key box is found to be open/unlocked, or the keys are not present in the box, Rad Ops shall be notified immediately.
C. The on-duty Rad Ops Shift Supervisor, or designee, is responsible
for maintaining
control of the keys to open the respective
LHRA and VHRA key boxes, and access to the key boxes themselves
and the keys stored in the key boxes, for their respective
shift. D. At the start of each shift, the Rad Ops Shift Supervisor, or designee, will verify that the keys to the LHRA and VHRA key boxes are present, that the LHRA key box is inventoried
and that all LHRA keys are accounted
for, to include the key to the Ops LHRA key box, and that the VHRA key box is locked and the seal is intact. Performance
of this verification
and inventory
of the LHRANHRA key boxes shall be noted in the Radiation
Operations
Log. PER #84532 E. If any key cannot be accounted
for during shift inventory, or the seal on the VHRA key box is not intact, the Rad Ops Manager shall be notified, and immediate
actions taken to locate and secure the missing key. 
) RCI-29 SQN Control of Radiation
Protection
Keys Revision 7 Page 2 of 6 ------------------------------_
.. _------1.0 PURPOSE The purpose of this Instruction
is to provide guidelines
for controlling
Radiation
Protection
keys. 2.0 SCOPE This Instruction
establishes
the requirements
for maintaining
positive control of the Radiation
Protection
keys utilized to access Locked High Radiation
Areas and Very High Radiation
Areas. 3.0 REFERENCES
1 OCFR19, Notices, Instructions, and Reports to Workers: Inspection
and Investigations
10CFR20, Standards
for Protection
Against Radiation
Reg Guide 8.38, Control of Access to High and Very High Radiation
Areas in Nuclear Power Plants SPP-5.1, Radiological
Controls RCDP-1, Conduct of Radiological
Controls SON Technical
Specifications, Unit One and Unit Two 4.0 DEFINITIONS/ABBREVIATIONS
Absorbed Dose -The energy imparted by ionizing radiation
per unit mass of irradiated
material.
Accessible
Area -Any area that can reasonably
be occupied by a major portion of the whole body of an individual (as defined in 10CFR20).
ANSI-Qualified
Personnel
-Rad Protection
personnel
assigned to SON, permanent
or temporary, who meet ANSI qualifications.
LHRA and VHRA Keys -Keys used to lock/unlock
LHRA tumbler locks and security locks and/or VHRA security locks. LHRA and VHRA Security Lock -A padlock in use to secure a LHRA (where a tumbler lock cannot be used) and a VHRA. LHRA Tumbler Set -Lock tumblers installed
exclusively
in doors leading directly into a LHRA. Locked High Radiation
Area (LHRA) -Any area, accessible
to individuals, in which radiation
levels from radiation
sources external to the body could result in an individual
receiving
a dose equivalent
in excess of 1.0 rem (1,000 mrem) in one hour at 30 centimeters
from the radiation
source or from any surface that the radiation
penetrates.
Positive Control -Control of the keys utilized by Rad Ops to access a LHRA or VHRA, in accordance
with the requirements
of this Instruction.
Very High Radiation
Area (VHRA) -Any area, accessible
to individuals, in which radiation
levels from radiation
sources external to the body could result in an individual
receiving
an absorbed dose in excess of 500 rads in one hour at one meter from a radiation
source or from any surface that the radiation
penetrates.
5.0 RESPONSIBILITIES
Responsibilities
are defined in Section 6.0. 
TVA 40385 [6-2003] Page 2 of 2 1 PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: RADIOLOGICAL
PO STINGS AND SIGNS IV. LENGTH OF LESSON/COURSE:
1-2 hour(s) V. TRAINING OBJECTIVES:
A. Tenninal Objective:
OPL271C259
Revision 8 Page 3 of 18 Upon completion
of this lesson and others presented, the student shall demonstrate
an understanding
of Radiological
Postings and Signs by successfully
completing
a written examination
as defined by program procedures.
B. Enabling Objectives:
1. Define and identifY the requirements
for entering/working
in the following
areas: a. Unrestricted
Area b. Restricted
Area c. Radiologically
Controlled
Area d. Radioactive
Material/Radioactive
Material Storage Areas e. Radiation
Area f. High Radiation
Area g. Locked High Radiation
Area h. Very High Radiation
Area i. Contamination
Area j. High Contamination
Area k. Airborne Radioactivity
Area. 2. Identify the criteria for utilizing
Hot Spot Labels/Tags
and Radioactive
Material Tags. The following
list contains knowledge
and ability statements (KlAs) from The Knowledge
and Abilities
Catalog for Nuclear Power Plant Operators:
Pressurized
Water Reactors (PWR) NUREG-1122, Revision 2 that are applicable
to the Initial Licensed Candidate
training program. As such, questioning
in these areas will be included on any testing in preparation
of, or included in obtaining
either RO or SRO NRC license. lOCFR55 IMPORTANCE
KIA # KIA Statement
Sect. Link(s) RO/SRO G 2.3.1 Knowledge
of 10CFR20 and related facility radiation
control 41.12/43.4.
2.6/3.0 requirements.
45.9/45.10
G2.3.2 Knowledge
of facility ALARA program. 41.12/43.4
/ 2.5/2.9 45.9/45.10
G 2.3.4 Knowledge
of radiation
exposure limits and contamination
control, 43.4 / 45.10 2.5/3.1 ) including
pennissible
levels in excess of those authorized.
IG 2.3.5 Knowledge
of use and function of personnel
monitoring
equipment.
41.11 /45.9 2.3/2.5 G 2.3.7 Knowledge
of the process for preparing
a radiation
work pennit. 41.10/45.12
2.0/3.3 G 2.3.10 Ability to perfonn procedures
to reduce excessive
levels of radiation
43.4 / 45.10 2.9/3.3 and guard against personnel
exposure. 
\ LESSON BODY 3) A radiation
monitoring
device that continuously
transmits
dose rate and cumulative
dose information
to a remote reciever monitored
by radiation
protection;
or, 4) A self-reading
dosimeter
and be under the surveillance
while in the area of an individual
qualified
and equiped with
a radiation
monitoning
device that continuously
displays radiation
dose rates, or under the surveillance
while in the area by means of closed circuit television
and the means to communicate.
F. Locked High Radiation
Areas (LHRA) ) 1. Definition:
An accessible
area to individuals
in which radiation
levels could result in an individual
receiving
a dose equivalent
in excess of 1,000 mrem in 1 hour at 30 centimeters
from the radiation
source or from any surface that the radiation
penetrates.
2. Explanation:
a. The Shift Manager (SM) shall be notified when a LHRA is established
or removed. b. LHRA keys shall be maintained
under the administrative
control of the SM, Rad Ops Manager, or their designees.
c. Each LHRA shall be posted with a conspicuous
sign or signs bearing the standard radiation
symbol and the words "Caution -Locked High Radiation
Area," or "Danger -Locked High Radiation
Area." d. Entry shall be established
by the use of a RWP. In addition, each individual
or group entering such an area shall possess: 1) A radiation
monitoring
devise which has an appropriate
alarm setting capability, continuously
integrates
the radiation
dose rate in the area and alarms when the device's dose alarm setpoint is reached; or, 2) A radiation
monitoring
device that continuously
transmits
dose rate and collective
dose information
to a remote recorder monitored
by Radiation
Protection
personnel
with the means to communicate
with and control every individual
in the area; or, 3) A self-reading
dosimeter
and be under closed circuit television
surveillance
by a qualified
Radiation
Protection
Technician
and equipped with a monitoring
device that continuously
displays radiation
dose rates in the area; or, 4) A self-reading
dosimeter
and be under closed circuit television
surveillance
by a qualified
Radiation
Protection
Technician
with the means to communicate
with and control every individual
in the area. OPL271C259
Revision 8 Page 10 of 18 INSTRUCTOR
NOTES Obj.l.g 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
') 98. G 2.3.14098 . Given the following
conditions:
Both Units operating
at 100% power. 0600 Due to extremely
heavy rainfall, RSO/KEOC issues a Stage I flood warning. Which one of the following
identifies
the time the Stage 1 flood mode actions are required to be completed, and if Stage II actions are required, how the Tritiated
Drain Collector
Tank would be prepared to prevent a possible release of radioactivity?
Time Tritiated
Drain Collector
Tank A. 1600 Pressurized
to greater than 23 psig. 1600 Filled with water. C. 2300 Pressurized
to greater than 23 psig .. D. 2300 Filled with water. DIS TRA CTOR ANAL YSIS: Page 65 After entering AOP-N.03, "Flooding" Stage I preparations
must be implemented
and completed
within the following
1 0 hours and if Stage /I actions are required they must be completed
within the following
17 hours. A. Incorrect, The time requirement
is correct. However, Stage /I actions do require the Tritiated
Drain Collector
Tank to be filled with water to prevent the tank from floating away and becoming a radiation
hazard. Plausible
because the time requirement
is correct and pressurizing
the tank to greater than 23 psig is correct for other tanks during Stage /I preparations.
B. CORRECT, The time requirement
is correct and the requirement
is to fill Tritiated
Drain Collector
Tank to be filled with water to prevent the tank from floating away and becoming a radiation
hazard. C. Incorrect, The time requirement
is not correct and pressurizing
the tank to greater than 23 psig is not correct. Plausible
because the time identified
is 17 hours which is the time required to complete Phase /I actions after Phase /I is initiate and pressurizing the
tank to greater than 23 psig is correct for other tanks during Stage /I preparations.
D. Incorrect, The time requirement
is not correct but filling the tank with water is correct. Plausible
because the time identified
is 17 hours which is the time required to complete Phase /I actions after Phase /I is initiate and filling the tank with water is correct. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
') Question No. 98 Tier 3 KIA G 2.3.14 Knowledge
of radiation
or contamination
hazards that may arise during normal, abnormal, or emergency
conditions
or activities.
Importance
Rating: 3.41 3.8 Technical
Reference:
AOP-N.03, Flooding, Rev 28 TRM 4.7.6, Flood Protection
Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271AOP-N.03
B.4,5, & 8b Question Source: 4. Upon entry into AOP-N.03, diagnose the applicable
condition
and transition
to the appropriate
section for response.
5. Describe the bases for all limits, notes cautions and steps of AOPN.03. 8.b Given a set of initial plant conditions
use to correctly (b) Identify required actions. Bank # _-:--__ Modified Bank # _X, __ _ New __ _ Question History: Sequoyah Question AOP-N.03-B.4
002 modified Question Cognitive
Level: Memory or fundamental
knowledge
__ _ Comprehension
or Analysis _X __ 10 CFR Part 55 Content: ( 41.12/43.4/45.10 ) 10CFR55.43.b
(2,4) Requires the candidate
to know the facility Technical
Requirements, implementation
of AOP sections and be knowledgeable
of provisions
to prevent radiation
hazards that may arise during normal, abnormal and emergency
conditions.
Comments:
Sequoyah Question AOP-N.03-B.4
002 modified Page 66 
PLANT SYSTEMS TR 3/4.7.6 FLOOD PROTECTION
LIMITING CONDITION
FOR OPERATION
TR 3.7.6 The flood protection
plan shall be ready for implementation
to maintain the plant in a safe condition.
APPLICABILITY:
When one or more of the following
conditions
exist: a. early warning of major flood-producing
rainfall conditions
in the east Tennessee
watershed, b. an early warning that a criticalcornbination
of flood and/or higher than normal Summer pool levels plus possible darn failures or other darn safety emergencies
mayor have developed, c. or warnings that flood elevation
is predicted
to exceed plant grade (Stages I and II) . ACTION: a. With a Stage I flood warning issued initiate and complete within 10 hours the Stage I flood protection
procedure
which shall include being in at least HOT STANDBY within 6 hours, with a SHUTDOWN MARGIN of at least 5% delta k/k and less than or equal to 350&deg;F within the following
4 hours. If within 10 hours following
the issuance of a Stage I flood warning communications
between the TVA Water I Management
River Scheduling (RS) and the Sequoyah Nuclear Plant cannot be verified, initiate and complete the Stage II flood protection
procedure
within the following
17 hours. With a Stage II flood warning issued initiate the Stage II flood protection
plan in time to ensure completion
before the predicted
flooding of the site. Initiation
shall be no later than 17 hours prior to the predicted
arrival time of the initial critical flood level (703 ft msl winter and summer). Completion
of any actions are not required if warnings are retracted
by RS. b. After an early warning is issued, verify communications
between TVA RS and the Sequoyah Nuclear Plant within 5 hours or initiate and complete the Stage I flood protection
plan within the following
10 hours. If communications
have not been established
upon completion
of the Stage I flood protection
plan initiate and complete the Stage II flood protection
plan within the following
17 hours.' Completion
of any actions are not required if corrmunications
are verified.
SEQUOYAH -UNITS 1 AND 2 TECHNICAL
REQUIREMENTS
3/4 7-4 January 20, 2006 Revision Nos. 5, 14 
PLANT SYSTEMS TR 3/4.7.6 FLOOD PROTECTION
SURVEILLANCE
REQUIREMENTS
TR 4.7.6.1 This requirement
deleted. TR 4.7.6.2 Communications
between Sequoyah Nuclear Plant and TVA RS shall be maintained
on 3 hour intervals.
If not maintained, within 5 hours from previous contact initiate Stage I Flood Plan and continue with Stages I and II until contact is re-established
and RS confirms that the flood plans are not required.
SEQUOYAH -UNITS 1 AND 2 TECHNICAL
REQUIREMENTS
3/4 7-5 January 20, 2006 Revision Nos. 5, 14 
) SQN FLOODING APPENDIXC
CVCS AND WDS TANK FILLING INSTRUCTIONS
NOTE 1 . This appendix provides instructions
for filling the partially
filled and possibly radioactive
tanks located below Maximum Probable Flood level of 723.1' elev. Performance
of this procedure
minimizes
NOTE 2 the possibility
of the tanks collapsing
or breaking loose and possible release of radioactivity.
Steps 1 through 22 may be performed
out of sequence.
1. PLACE Reactor Building Floor and Equipment
Drain Sumps IN SERVICE to Tritiated
Drain Collector
Tank USING 0-SO-77-10.
2. FILL pressurizer
relief tank USING 1,2-S0-68-5.
3. NOTIFY Maintenance
to connect 2 inch passive failure discharge
connection
on Auxiliary
Building Floor and Equipment
Drain Sump Pumps discharge
USING O-FP-MXX-OOO..:OOB.O.
4. OPEN the following
valves to fill Floor Drain Collector
Tank: * 0-77-915 *
NOTE The time required for filling the Tritiated
Drain Collector
Tank is approximately
4 hours. 5. OPEN valve 0-77-914 and FILL Tritiated
Drain Collector
Tank. 6. ALIGN Laundry and Hot Shower Tank Pump to Waste Condensate
Tanks A, B,and C USING 0-SO-77-5.
7. Al,.lGN Laundry and Hot Shower Tanks A and B and Chemical Drain Tank. Page 143 of 215 AOP-N.03 Rev. 28 Page 1 of 4 
AOP-N.03-BA
002 QUESTIONS
REPORT for BANK SQN Questions
Which ONE of the following
is the correct time allowed in AOP-N.03, Flooding, to complete any Stage I procedure
section? A'!I 10 hours. B. 17 hours. C. 27 hours. D. 40 hours. Justification:
A. Correct -Correct as stated in NOTE prior to step 1 of AOP-N.03 B. Incorrect
-as it defines the 14 hours required to complete Stage 2 plus 3 hours margin. C. Incorrect-
as it defines the total time to complete Stage 1 + 2 plus 3 hours margin. D. Incorrect-
not defined by AOP-N.03.
Monday, November 24, 2008 7:32:55 AM 1 
OPL271 AOP-N.03 Revision 1 Page 3 of 40 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-N.03, FLOODING IV. LENGTH OF LESSON/COURSE:
2 hours V. TRAINING OBJECTIVES:
O. 1. 2. 3. 4. S. 6. A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of AOP-N.03, FLOODING.
B. Enabling Objectives
Objectives
Demonstrate
an understanding
of NUREG 1122 knowledge's
and abilities
associated
with a plant Flood that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification
Training for the appropriate
position as identified
in Appendix A State the purpose/goal
of this AOP-N.03.
Describe AOP-N.03 entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with AOP-N.03 entry conditions.
b. Describe the ARP requirements
associated
with AOP-N.03 entry conditions.
c. Interpret, prioritize, and verify associated
alarms are consistent
with AOP-N.03 entry conditions.
d. Describe the plant parameters
that may indicate a plant Flood. Describe the initial operator response to stabilize
the plant upon entry into AOP-N.03.
Upon entry into AOP-N.03, diagnose the applicable
condition
and transition
to the appropriate
procedural
section for response.
Summarize
the mitigating
strategy for the failure that initiated
entry into AOP-N.03.
Describe the bases for all limits, notes, cautions, and steps of AOP-N.03. 
I OPL271 AOP-N.03 Revision 1 Page 4 of 40 7. Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures.
8. Given a set of initial plant conditions
use AOP-N.03 to correctly:
a. Recognize
entry conditions.
b. Identify required actions. c. Respond to Contingencies.
d. Observe and Interpret
Cautions and Notes. 9. Describe the Tech Spec and TRM actions applicable
during the
performance
of AOP-N.03.
10. Apply GFE and system response concepts to the abnormal condition
-prior to, during and after the abnormal condition.
OBJECTIVES
TO BE COVERED IN THESE SEQUOY AH OPERATOR TRAINING PROGRAMS OBJECTIVE
I NONLICENSED
LICENSE TRAINING NO. RO SRO REQUAUSPECIAL
OPERATORS
O. X X 1. X X 2. X X 3. X X 4. X X 5. X X 6. X X 7. X X 8. X X 9. X X 10. X X Selected objectives
to be covered in: PowerPoint
presentation
to be used: Sequoyah Operator Training Manager / Date Sequoyah Operations
Manager / Date 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
99. G 2.4.27099
Given the following:
Both units in service at 100% power. -The plant is operating
with minimum operations
staffing. -A fire develops at the ERCW pumping station. -AOP-N.01, "Plant Fires", is implemented.
-The Shift Manager declares an ALERT emergency
and initiates
Assembly and Accountability.
Fire Ops actions to extinquish
the fire continue.
Which ONE of the following
identifies
... (1) the direction
the AUOs are to be given and (2) if the Shift Manager later decides to implement
AOP-N.08, "Appendix
R Fire Safe Shutdown", must both units enter the AOP or can the decision to enter the AOP be made separately
for each unit? A'! (1) Direct all AUOs to report to the Main Control Room. (2) Decision to enter AOP-N.08 can be made separately
on each unit. B. (1) Direct all AUOs to report to the Main Control Room. (2) Decision to enter AOP-N.08 must be made on both units at the same time. C. (1) Direct 2 AUOs to report to the OSC and the rest to report to the Main Control Room. (2) Decision to enter AOP-N.08 can be made separately
on each unit. D. (1) Direct 2 AUOs to report to the OSC and the rest to report to the Main Control Room. Page 67 (2) Decision to enter AOP-N.08 must be made on both units at the same time. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRA CTOR ANAL YSIS: Page 68 A. CORRECT, with the fire a the ERCW pumping station still burning and the staffing at minimum (8 AVOs), all AVOs will report to the main control room because in accordance
with AOP-N.01, the requirement
to assemble AVOs to assign and brief the actions of AOP-N.08 take priority over staffing the OSC and Level II Fire Brigade. Separate entry into AOP-N.08 is allowed in accordance
with AOP-N.01 because the criteria requiring
both units to enter the AOP at the same time are not met. The fire is not in one of the four limiting areas that require simultaneous
entry by both units .. B. Incorrect, with the fire a the ERCW pumping station still burning and the staffing at minimum (8 AVOs), all AVOs will report to the main control room because in accordance
with AOP-N.01, the requirement
to assemble AVOs to assign and brief the actions of AOP-N.08 take priority over staffing the OSC and Level II Fire Brigade. However the AOP-N.08 entry does not met the criteria requiring
both units to enter at the same time. Plausible
because the directions
to the AVOs is correct and a fire in a different
location could cause the units to enter the AOP at the same time. C. Incorrect, with the fire a the ERCW pumping station still burning and the staffing at minimum (8 AVOs), 2 AVOs would not be sent to the OSC, all would report to the main control room in accordance
with AOP-N.01 because the requirement
to assemble A VOs to assign and brief the actions of A OP-N. 08 takes priority over staffing the OSC and Level II Fire Brigade. The AOP-N.08 can be made separately
because entry does not met the criteria requiring
both units to enter at the same time. Plausible
because the directions
to send 2 AVOs to the OSC is the normal protocol during an emergency
event and the units entry into the AOP being allowed at separate times in correct. D. Incorrect, with the fire a the ERCW pumping station still burning and the staffing at minimum (8 AVOs), 2 AVOs would not be sent to the OSC, all would report to the main control room in accordance
with AOP-N.01 because the requirement
to assemble AVOs to assign and brief the actions of AOP-N.08 takes priority over staffing the OSC and Level II Fire Brigade. Also, the AOP-N.08 entry does not met the criteria requiring
both units to enter at the same time. Plausible
because the directions
to send 2 AVOs to the OSC is the normal protocol during an emergency
event and a fire in a different
location could cause the units to enter the AOP at the same time. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 99 Tier 3 KIA G 2.4.27 Knowledge
of "fire in the plant" procedures.
Importance
Rating: 3.4 / 3.9 Technical
Reference:
AOP-N.01, Plant Fires, Rev 26 AOP-N.08, Appendix R fire Safe Shutdown, Rev 5 OPDP-1, Conduct of Operations, Rev 0010 Lesson Plan OPL271 REP, Rev 1 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271AOP-N.01
8.5 Question Source: Describe the bases for all limits, notes, cautions, and steps of AOP-N.01.
Bank# ___ _ Modified Bank # ---:--:--__ New _X __ Question History: New question for Sequoyah 2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
_X __ Comprehension
or Analysis __ _ 10 CFR Part 55 Content: ( 41.10/43.5/45.13 ) 10CFR55.43.b ( 5 ) Comments:
New question for Sequoyah 2009 exam Page 69 
) SQN APPENDIX R FIRE SAFE SHUTDOWN AOP-N.08 Rev. 5 STEP ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 2.0 GENERIC ACTIONS NOTE AUOs assigned to Level II fire brigade, OSC,and monitoring
fire pumps should be recalled for safe shutdown actions. 4. NOTIFY all available
AUOs to perform the following:
a. REPORT to MCR area immediately.
b. OBTAIN radio and SCBA from MCR. 5. PLACE all RCP handswitches
in STOP/PULL
TO LOCK. [M-5] 6. ENSURE MSIV and MSIV bypass valve handswitches
in CLOSE position.
[M-4] 7. MONITOR fire* location based upon reports from Fire Brigade Leader or Incident Commander.
Page 4 of 1380 
SQN PLANT FIRES AOP-N.01 Rev. 26 STEP I ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED NOTE 1: To ensure manual actions in AOP-N.08 can be completed
within the required time, AUOs must be ready to be dispatched
when AOP-N.08 is entered. The requirement
to assemble AUOs in this step takes priority over staffing the OSC and Level II Fire Brigade. NOTE 2: Step 17 (two pages) should be handed off to CRO, if applicable.
17. IF fire is still burning AND fire is in any of the following
locations:
* Aux Building * Additional
Equipment
Bldg * Reactor Bldg (Containment
or Annulus) * ERCW Pumping Station THEN PERFORM the fOllowing:
a. NOTIFY at least eight (8) AUOs to report to Control Room. b. IF fire is in Aux Building, THEN ENSURE the following
CCS pumps RUNNING: [O-M-27B]
* CCS pump 1 A-A * CCS pump 1 B-B * CCS pump 2A-A * CCS pump 2B-B (step continued
on next page) Page 12 of 84 
SQN PLANT FIRES AOP-N.01 Rev. 26 STEP I ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED NOTE 1 NOTE 2 NOTE 3 NOTE 4 The decision to implement
AOP-N.08 or AOP-C.04 is an SRO judgment based upon the severity of fire and its potential
effect on plant equipment.
BOTH UNITS must enter AOP-N.08 at same time if ALL of the following
conditions
are met: * fire is in one of four limiting Aux Bldg areas (el. 690 General Area, el. 714 General Area, 6.9KV Shutdown Board Rm A, or 6.9KV Shutdown Bd Rm B) * both units in Mode 1-3 * criteria for AOP-N.08 entry in Step 19 is met on either unit In fire areas other than the four limiting Aux Bldg areas, the decision to enter AOP-N.08 is made separately
on each unit. AOP-N.08 is NOT applicable
for a total loss of &#xa3;!l ERCW capability.
AOP-M.01 provides required actions. 19. MONITOR magnitude
of fire and potential
to impact control of unit: a. CONSULT Incident Commander
or Fire Brigade Leader. b. MONITOR MCR indications
and controls for equipment
failures or spurious operation. (step continued
on next page) Page 16 of 84 
X. LESSON BODY: 2. Minimum TSC staffing * SED * RP Manager * Operations
Manager or Operations
Communicator
* Technical
Assessment
Manager, Technical
Assessment
Team Leader, or Reactor Engineer * Mechanical
Engineer * Electrical
Engineer 3. Transfer of SED role from the SM to the TSC * Use EPIP-6, Appendix B G. EPI P-7, Activation
and Operation
of the Operations
Support Center (OSC) 1. This procedure
provides the guidance for activating
and operating
the OSC. The OSC is required to be activated
at the ALERT classification
and above. The SED directs mitigating
actions and determines
OSC priorities.
The OSC Manager oversees OSC activities.
2. Minimum OSC staffing * OSC Manager * Mechanical
Maintenance
Group (1) * Electrical
Maintenance
Group (1) * Instrument
Maintenance
Group (1) 3. AUO teams responding
to procedure
driven activities (EOPs, EAs, AOPs, etc.) are under the direction
of the SM but tracked by the OSC. 4. RP and Operations
should normally have personnel
on each response team. H. EPIP-8, Personnel
Accountability
and Evacuation
1. This procedure
provides the method for accounting
for all personnel
and visitors in the protected
area within 30 min. OPL271 REP Revision 1 Page 23 of 32 INSTRUCTOR
NOTES Refer to EPIP-6, Appendix B for discussion.
SM -Ensure you keep OSC aware of AUO usage for tracking purposes.
Assembly should not be initiated
if assembly would present a danger to employees. 
x. LESSON BODY: 2. Operations
Personnel
a. The protocol requires two AUOs to go to the OSC and the remaining
AUOs go to the Main Control Room. The SM may send the AUOs to a waiting area within the protected
ventilation
area of the Control Building (such as the SM Office) if noise or congestion
become a problem in the MCR. b. When the OSC Ops Advisor arrives in the OSC, he and the SM will collectively
manage the resources
to have the fastest response to problems yet still OPL271 REP Revision 1 Page 24 of 32 INSTRUCTOR
NOTES provide a check of radiological
conditions
prior to I Obj S.b dispatching
an AUO. Dispatching
AUOs into plant areas needs input by RP as to appropriate
protective
equipment.
This is best done by the OSC Ops Advisor once he/she arrives. The OSC Ops Advisor is expected to relieve the SM of some of the administrative
burden of tracking field personnel
so he/she can focus on the higher priority of operational
safety. c. The Shutdown Board Rooms and Control Building are spaces with protected
ventilation
similar to the MCR. Until advised by RP that these areas are becoming unsafe, they should be considered
part of the MCR for the purpose of dispatching
AUOs. This means that AUOs may be dispatched
into these areas for tasks without being considered
a "team" (much like sending an AUO behind the panels in the MCR to perform a task). As always, communications
must be maintained
at all times with AUOs while out on assignments
and the TSC should be informed of AUO activities.
d. This protocol allows dispatching
AUOs by telephone
briefing and/or radio briefing as well as transfer of oversight
of individual
AUOs between the OSC SRO and SM as needed. This provides the best utilization
of personnel
while still protecting
the AUOs from radiological
hazards. e. AUOs assigned to the OSC will report directly to the OSC when required.
f. The SM is responsible
to account for MCR personnel. 
NPG Standard Conduct of Operations
OPDP-1 Department
Rev. 0010 Procedure
Page 47 of 62 Attachment
1 (Page 2 of 2) Shift Staffing WBN WBN Mode 1-4 Mode 5& 6 SON BFN Shift Manager (SRO) 1 1 1 Unit Supervisor (SRO) 1 1 3 3 Unit Operator (RO) 2 1 4 6 Non-licensed (AUO) 5 3 8 8 STA 1 1 1 SON The SM, a US or the WCC may be the STA and one US will be the Incident Commander.
The STA need not be licensed.
Two active-licensed
SROs are required for Unit Supervisor
positions
and a third active licensed SRO is required as Shift Manager. BFN One of the US can be the STA. The Incident Commander
position may be filled by an additional
qualified
person. Notification
of Absences * Operations
personnel (except Fire Operations)
unable to report for shift duty shall, althe earliest possible time and no later than 2 hours before the scheduled
time, inform the SM/US of the situation.
The SM/US shall make necessary
arrangements
for obtaining
a replacement.
* Fire Operations
personnel
unable to report for shift duty shall, at the earliest possible time and no later than 2 hours before the scheduled
time, inform the Fire Operations
Foreman of the situation.
The Fire Operations
Foreman shall make necessary
arrangements
for obtaining
a replacement. 
OPL271 AOP-N.01 Revision 1 Page 3 of 15 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-N.01, PLANT FIRES IV. LENGTH OF LESSON/COURSE:
1 hour V. TRAINING OBJECTIVES:
O. 1. 2. 3. 4. 5. 6. A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of AOP-N.01, PLANT FIRES. B. Enabling Objectives
Objectives
Demonstrate
an understanding
of NUREG 1122 knowledge's
and abilities
associated
with Plant Fires that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification
Training for the appropriate
position as identified
in Appendix A. State the purpose/goal
of this AOP-N.01.
Describe the AOP-N.01 entry conditions.
a. Describe the setpoints, interlocks, and automatic
actions associated
with AOP-N.01 entry conditions.
b. Describe the ARP requirements
associated
with AOP-N.01 entry conditions.
c. Interpret, prioritize, and verify associated
alarms are consistent
with AOP-N.01 entry conditions.
d. Describe the plant parameters
that may indicate a Plant Fire. Describe the initial operator response to stabilize
the plant upon entry into AOP-N.01.
Summarize
the mitigating
strategy for the failure that initiated
entry into AOP-N.01.
Describe the bases for all limits, notes, cautions, and steps of AOP-N.01.
Describe the conditions
and reason for transitions
within this procedure
and transitions
to other procedures. 
I I OPL271 AOP-N.01 Revision 1 Page 4 of 15 7. Given a set of initial plant conditions
use AOP-N.01 to correctly:
a. Recognize
entry conditions.
b. Identify required actions. c. Respond to Contingencies.
d. Observe and Interpret
Cautions and Notes. 8. Describe the Tech Spec and TRM actions applicable
during the performance
of AOP-N.01.
9. Apply GFE and system response concepts to the abnormal condition
-prior to, during and after the abnormal condition.
-OBJECTIVES
TO BE COVERED IN THESE SEQUOYAH OPERATOR TRAINING PROGRAMS OBJECTIVE
I NON LICENSED LICENSE TRAINING NO. RO SRO REQUAUSPECIAL
OPERATORS
O. X X 1. X X 2. X X 3. X X 4. X X 5. X X 6. X X 7. X X 8. X X 9. X X 10. X X NOTE: The following
approval is required for License Requalification
and special training only: Selected objectives
to be covered in: PowerPoint
presentation
to be used: Sequoyah Operator Training Manager / Date Sequoyah Operations
Manager / Date 
REP-B.1.D
001 Given the following:
QUESTIONS
REPORT for BANK SQN Questions
-Unit 1 has experienced
a LOCA and loss of 1 A-A 6.9 kV SO Bd. -You are the OATC. -The SM just announced
an ALERT and sounded assembly.
How will the AUOs respond to the siren going off? The two AUOs assigned to the OSC will report to the OSC, the remaining
AUOs will report to the SM. 8. The two AUOs assigned to the SM will report to the MCR, the remaining
AUOs will report to the OSC. C. ALL AUOs will immediately
report to the SM until the OSC Operations
Advisor SRO is ready to assume control of the AUOs. O. ALL AUOs will immediately
report to the OSC Operations
Advisor SRO until the SM is ready to assume control of the AUOs. Justification:
A. Correct. Refer to EPIP-7, Appendix D for stage
1, "Declaration
of the Emergency. " B. Incorrect.
Two AVOs are assigned to support the MSS in the OSC until staffed; all other AVOs are under SM control. C. Incorrect.
Two AVOs are assigned to support the MSS in the OSC until staffed; all other AVOs are under SM control. After OSC Operations
Advisor arrives, the SM may assign additional
AVOs not in the field to the OSC Operations
Advisor. D. Incorrect.
Only two AVOs are assigned to support the MSS in the OSC until staffed; all other AVOs are under SM control. Monday, November 24, 2008 7:46:32 AM 1 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted , 100. G 2.4.28 100 Given the following:
Both units trip from power operation
due a loss of off site power and sabotage is suspected.
Nuclear Security reports the following:
-Information
has been received that a specific credible insider threat exists associated
with the loss of offsite power. and -Concerns exist for the health and safety of any oncoming emergency
responders.
The Shift Manager determines
the need to staff the emergency
centers, makes REP declaration
and activates
Assembly and Accountability.
The Two-person
line of sight rule has been implemented.
The Operating
crew needs to send personnel
to the DG building due to an alarm occurring
on DG 1A-A. Which ONE of the following
describes
the selection
the Shift Manager will make when activating
the Emergency
Paging System (EPS) and the required measures to be taken when sending
personnel
to the DG building.
A. STAGING AREA will be selected on the EPS; Both individuals
sent must be qualified
for the task to be performed. STAGING AREA will be selected on the EPS; Only one of the individuals
sent must be qualified
for the task to be performed.
C. EMERGENCY
will be selected on the EPS; Both individuals
sent must be qualified
for the task to be performed.
D. EMERGENCY
will be selected on the EPS; Only one of the individuals
sent must be qualified
for the task to be performed.
Page 70 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
DIS TRACTOR ANAL YSIS: A. Incorrect, In accordance
with the Emergency
Plan, if concerns exist for the health and safety of oncoming responders, the "Staging Area" choice is to be selected when initiating
the staffing of the Emergency
Centers but when the two-man rule is implemented, the individuals
do not have to have the same qualifications.
Plausible
because choosing the "Staging Area" when initiating
the staffing of the Emergency
Centers is correct and typically
when personnel
are sent to perform a task both would be qualified.
Page 71 B. CORRECT, In accordance
with the Emergency
Plan, if concerns exist for the health and safety of oncoming responders, the "Staging Area" choice is to be selected when initiating
the staffing of the Emergency
Centers and the individual
do not have to posses the same qualification
when the two-man rule is implemented.
C. Incorrect, the "Staging Area" choice is to be selected when initiating
the staffing of the Emergency
Centers (not the the "Emergency''), when concerns exist for the health and safety of oncoming responders, and the individual
do not have to posses the same qualification
when the two-man rule is implemented.
Plausible
because the the "Emergency" choice would be correct with different
conditions
when inititiating
the staffing of the Emergency
Centers and typically
when personnel
are sent to perform a task both would be qualified.
D. Incorrect, the "Staging Area" choice is to be selected when initiating
the staffing of the Emergency
Centers (not the the "Emergency''), when concerns exist for the health and safety of oncoming responders, and the individual
do not have to posses the same qualification
when the two-man rule is implemented.
Plausible
because the the "Emergency" choice would be correct with different
conditions
when inititiating
the staffing of the Emergency
Centers and the requirement
for only one of the individuals
to be qualified
is correct. 
Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted
Question No. 100 Tier 3 KIA G 2.4.28 Knowledge
of procedures
relating to a security event (non-safeguards
information).
Importance
Rating: 3.2/4.1 Technical
Reference:
SPP-1.3, Access Authorization
and Nuclear Security, Rev 0011 EPIP-8, Personnel
Accountability
and Evacuation, Rev 17 EPIP-3, Alert, Rev 30 Proposed references
to be provided to applicants
during examination:
None Learning Objective:
OPL271 REP B. 5.b. Question Source: State the duties of the Site Emergency
Director (SED) . b. State the conditions
under which the SED may order relocation
from one assembly point to another. Bank# ___ _ Modified Bank # ___ _ New X '----Question History: New question for SON 1/2009 exam Question Cognitive
Level: Memory or fundamental
knowledge
_X __ Comprehension
or Analysis __ _ 10 CFR Part 55 Content: ( 41.10/43.5/45.13 ) 10CFR55.43.b ( 5 ) Comments:
New question for SON 1/2009 exam Page 72 
NPG Standard Access Authorization
and Nuclear Security SPP-1.3 Programs and Rev. 0011 Processes
Page 32 of 51 3.9.9 Maintenance (continued)
B. Repair/replace
failed security devices and components
in the minimum time necessary.
3.9.10 Unauthorized
Materials
Firearms, explosives, incendiary
devices, alcoholic
beverages
and illegal drugs are prohibited
on nuclear plant sites. 3.9.11 Credible Insider Threat When Nuclear Security determines
that a specific credible insider threat exists, Nuclear Security shall request Shift Manager/Site
Emergency
Director to implement
a two-person (line of sight) rule for personnel
in vital areas, unless unusual circumstances
exists where emergency
actions by a single individual
are required to ensure nuclear safety. (See also Appendix A, paragraph
4.4 and EPIP-8, Personnel
Accountability
and Evacuation).
3.9.12 Hostage/Duress
Situation
Any employee who is coerced, influenced
or pressured
in any way to initiate or be party to an act that presents an unsafe situation
at any of the TVA NPG Sites will immediately
contact Nuclear Security or Corporate
Nuclear Security, as appropriate, and provide as much information
as possible.
Nuclear Security or Corporate
Nuclear Security (CNS), as appropriate, will contact the appropriate
agency to respond, (see also Appendix A). 3.9.13 CameraNideo
Requirements
Individuals
are prohibited
from taking pictures, videos, etc of security equipment, security posts, or other security areas/items
without prior authorization
by the SSM or designee.
3.10 Safeguards
Events (SGE) 3.10.1 Reporting
SGEs Individuals
that discover an actual or suspected
SGE are responsible
for the immediate
reporting
of that event to Security.
A. If the actual or suspected
SGE occurs offsite, contact the Manager, CNS or SSM immediately, in person, or by telephone (whichever
is faster). B. If the actual or suspected
SGE occurs onsite, contact the SSM immediately, in person, or by telephone (whichever
is faster). NOTE This immediate
notification
is necessary
for the timely implementation
of contingency
plans and reporting
requirements.
C. Provide as much detail regarding
the incident as possible and, if requested, complete statement
describing
the event in as much detail as possible (for example: who, what, when, where, and why or how, if known). Upon completion
submit the statement
to Security. 
NPG Standard Programs and Processes Access Authorization
and Nuclear Security Appendix A (Page 2 of 2) SPP-1.3 Rev. 0011 Page 42 of 51 Guidelines
for Initial Actions by Plant Personnel
of Incidents
Involving
Suspected
Tampering
or Sabotage 4.2 Shift Manager (continued)
B. Shift Manager evaluates
appropriate
AOI or AOP entry for Security Event Response.
C. Shift Manager evaluates
EPIP-1, Emergency
Plan Classification
Matrix. 4.3 Employee If any act of coercion, influence
or pressure is committed
with intent to initiate an act of tampering
or sabotage, then Notify Nuclear Security.
4.4 Credible Insider Threat This action is triggered
when a credible insider threat specific to a facility exists. Once triggered, implementation
of the 2-man rule will be as expeditious
as resources
permit recognizing
that additional
personnel
may need to be called to the particular
site. The two persons do not have to possess similar skills or knowledge, but must remain in visual contact with each other unless personnel
or plant safety would be adversely
impacted.
5.0 SUBSEQUENT
ACTIONS 5.1 Employee Retain any relevant information
and provide to Nuclear Security to aid in investigation.
5.2 Security Supervisor
If an event as described
in Sections 4.1, 4.2, 4.3 through Section 4.4 above is reported, then A. Nuclear Security evaluate need to contact TVA Police Criminal Investigation
Division for assistance, and B. Nuclear Security evaluate in accordance
with this instruction
and NSDP-1, "Safeguard
Event Reporting
Guidelines." 
) ) SEQUOYAH PERSONNEL
ACCOUNTABILITY
AND EVACUATION
EPIP-8 9. 10. 11. APPENDIXD
Page 3 of 3 NUCLEAR SECURITY -ASSEMBLY AND ACCOUNTABILITY
ACTIONS REPORT the results of accountability
to the SM/SED within 30 minutes after the assembly and accountability
sirens have sounded. Accountability
is considered
complete when all personnel
have been accounted
for or are known by nameifnot
accounted
for. Unaccounted
Individuals
IF ... Individuals
remain unaccounted
45 minutes following
the activation
of the assembly and accountability
sirens, THEN ... REQUESTpermission
from the SM/SED to form search teams to locate the missing individual(s), AND NOTIFY RP and request they accompany
all search teams. Implement
the Two Person (Line of Sight) Rule and make a Public Address Announcement
WHEN ... Assembly and Accountability
have been completed, AND Nuclear Security has determined
that the Two Person (Line of Sight) Rule is required.
THEN ... REQUEST permission
from the SM/SED to make the following
Public Address Announcement: (Accountability
area PA is accessible
at x4800 from selected TSC and MCR phones) "Attention
all personnel.
A credible insider threat exists. Effective
immediately, all personnel
entering the Vital Areas must observe the 2-person rule. This rule requires that all persons in a vital area must remain in visual contact with another person unless personnel
or plant safety would be adversely
impacted.
This does not require that the two persons possess similar skills or knowledge.
I repeat. The 2-person rule is being. implemented
immediately." (REPEAT) PAGE 19 OF 27 D D Initials Time REVISION 17 
tSEQUOYAH
ALERT EPIP-3 3.1 ALERT DECLARATION
BY THE MAIN CONTROL ROOM (Continued)
[3] ACTIVATE Emergency
Paging System (EPS) as follows: [a] IF EPS has already been activated, THEN GO TO Step 4. [b] IF ongoing onsite Security events may present risk to the emergency
responders, THEN CONSULT with Security to determine
if site access is dangerous
to the life and health of emergency
responders.
[c] IF ongoing events makes site access dangerous
to the life and health of emergency
responders
THEN SELECT STAGING AREA button on the terminal INSTEAD of the EMERGENCY
button. [d] ACTIVATE EPS using touch screen terminal.
IF EPS fails to activate, THEN continue with Step 4. [4] COMPLETE Appendix B (TVA Initial Notification
for Alert). o o o o o NOTE: ODS should be notified within 5 minutes after declaration
of the event. ) [5] NOTIFY ODS. Initial Time ODS: Ringdown Line or 5-751-1700
or 5-751-2495
or 9-785-1700
[a] IF EPS failed to activate fromSQN when attempted, THEN DIRECT ODS to activate SON EPS. [b] IF ODS is also unable to activate EPS, THEN continue with step [5] [b]. [c] READ completed
Appendix B to ODS. [d] FAX completed
Appendix B toODS. 5-751-8620 (Fax) [e] MONITOR for confirmation
call from ODS that State/Local
notifications
complete:
RECORD time State notified.
o D o D Notification
Time FORWARD COMPLETED
PROCEDURE
TO EMERGENCY
PREPAREDNESS
MANAGER PAGE 5 of 15 REVISION 30 
\ 1 OPL271 REP Revision 1 Page 3 of 32 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: NP RADIOLOGICAL
EMERGENCY
PLAN AND SEQUOYAH EMERGENCY
PLAN IMPLEMENTING
PROCEDURES
IV. LENGTH OF LESSON/COURSE:
8 hours (Hot License Class), 2 -4 hours (LOR) V. TRAINING OBJECTIVES:
A. Terminal Objective:
Upon completion
of License Training, the participant
shall be able to demonstrate
or explain, using classroom
evaluations
and/or simulator
scenarios, the requirements
of the Radiological
Emergency
Plan (REP). B. Enabling Objectives:
O. Demonstrate
an understanding
of NUREG 1122 Knowledge
and Abilities
associated
with Radiological
Emergency
Plan that are rated;:::
2.5 during Initial License Training and;::: 3.0 during License Operator Requalification
Training for the appropriate
license position as identified
in Appendix A. 1. Discuss the Radiological
Emergency
Plan a. Discuss the regulatory
bases for the REP b. State the purpose of the REP. c. Define and state the purposes of a(n) NOUE, Alert, Site Area Emergency, and General Emergency
d. State the purpose and major job functions
of the Technical
Support Center (TSC), the Operations
Support Center (OSC), the Central Emergency
Control Center (CECC) and give the location of each. e. Describe the role the state and federal agencies play during an event f. Describe the process of authorizing
Emergency
Radiological
Exposures
in accordance
with EPIP-15. g. State the conditions
under which onsite personnel
would be administered
potassium
iodide (KI). h. Describe Chemistry
and Radiation
Protection
tasks during emergency
operations.
i. Discuss the termination
of a declared Radiological
Emergency
in accordance
with EPIP-16. 2. Determine
the required notifications
based upon the event, including
time requirements.
3. Classify emergency
events using appropriate
procedures.
------_._-
4. 5. 6. OPL271 REP Revision 1 Page 4 of 32 Determine
protective
action recommendations
using appropriate
procedures.
State the duties
and responsibilities
of the Site Emergency
Director (SED). a. State the duties and responsibilities
the SED may not delegate b. State the conditions
under which the SED may order relocation
from one assembly point to another. Discuss medical emergency
response per EPIP-10.
}}

Latest revision as of 12:52, 14 January 2025

2009 Initial Examination 05000327/2009301 and 05000328/2009/301, Draft SRO Written Examination
ML090710653
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Issue date: 02/06/2009
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