ML081370256
| ML081370256 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/08/2008 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| References | |
| 50-259/08-301 50-259/08-301 | |
| Download: ML081370256 (87) | |
See also: IR 05000259/2008301
Text
3-EOI APPENDIX-IS
Rev. 2
Paqe 2 of 4
( 3. (continued from previous page)
b. IF Main Condenser is desired drain path,
THEN OPEN 3-FCV-74-62, RHR MAIN CNDR FLUSH VALVE.
c. IF Radwaste is desired drain path,
THEN PERFORM the following:
1) ESTABLISH communications with Radwaste.
2) OPEN 3-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE.
d. NOTIFY personnel in Unit 3 RB, El 519 ft, Torus Area
to start RHR Drain Pump 3A(3B) .
e. THROTTLE 3-FCV-74-10S, RHR DR PUMP 3A/B DISCH HDR
VALVE, as necessary.
4. WHEN Suppression Pool level reaches -5.5 in.,
THEN SECURE RHR Drain System as follows:
a. DISPATCH personnel to STOP the Drain System as
follows (Unit 3 RB, El 519 ft, Torus Area) :
1) STOP RHR Drain Pump 3A(3B).
2) CLOSE the following valves:
- 3-SHV-074-0564A(B), RHR DR PUMP A(B) SEAL WTR SPLY
3) CLOSE and LOCK 3-SHV-074-0765A(B) , RHR DR PUMP
A(B) DISCH.
b. CLOSE 3-FCV-74-10S, RHR DR PUMP 3A/B DISCH HDR
VALVE.
c. VERIFY CLOSED 3-FCV-74-62, RHR MAIN CNDR FLUSH
VALVE.
d. VERIFY CLOSED 3-FCV-74-63, RHR RADWASTE SYS FLUSH
VALVE.
e. WHEN ... Suppression Pool level can be maintained
between -1 in. and -5.5 in.,
THEN ... EXIT this procedure.
3-EOI APPENDIX-18
Rev. 2
Pa e 3 of 4
( 5. IF ..... Directed by SRO to Emergency Makeup to the
Suppression Pool from Standby Coolant,
THEN ... CONTINUE in this procedure at Step 9.
6. IF Directed by SRO to add water to suppression pool,
THEN MAKEUP water to Suppression Pool as follows:
a. VERIFY OPEN 3-FCV-73-40, HPCI CST SUCTION VALVE.
b. OPEN 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE.
c. IF ..... HPCI is NOT available for Suppression Pool
makeup,
THEN ... MAKEUP water to Suppression Pool using RCIC
as follows:
1) VERIFY OPEN 3-FCV-71-19, RCIC CST SUCTION
VALVE.
2) OPEN 3-FCV-71-34, RCIC PUMP MIN FLOW
VALVE.
d. IF ..... 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE,
CANNOT be opened from control room,
THEN ... DISPATCH personnel to 250V DC RMOV Board 3B,
Compartment 50, to perform the following:
1) PLACE 3-XS-071-0034, RCTC PUMP MIN FLOW
VALVE EMER TRANS SWITCH, to EMERG.
2) OPEN 3-FCV-71-34, RCIC PUMP MIN FLOW
VALVE.
7. WHEN Suppression Pool level reaches -5.5 in.,
THEN VERIFY CLOSED the following valves:
- 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE
- 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE.
8. DISPATCH personnel to 250V DC RMOV Board 3B,
Compartment 50, to VERIFY 3-XS-071-0034, RCIC PUMP MIN
FLOW VALVE EMER TRANS SWITCH, in NORMAL.
55. RO 295031G2.4.6 OOl/C/A/TlGl/Cl//295031G2.4.6//RO/SRO/NO
Given the following plant conditions:
- Unit 2 was operating at 98% power when an automatic scram occurred due to a Group I
isolation .
- All control rods fully insert as reactor water level immediately drops below Level 2.
- The Recirc pumps trip.
- HPCI automatically initiates but immediately isolates due to a blown inner turbine exhaust
rupture diaphragm.
- RCIC had been tagged out of service previously to repair an oil leak.
- All other systems are operable.
- EOI-1, RPV Control, is entered.
- Pressure control was established with SRVs .
The remaining high pressure injection systems are unable to maintain reactor water level which is
currently at -150 inches and lowering .
Which ONE of the following contingency procedures would be appropriate to execute?
A. to! C1, Alternate Level Control
B. C2, Emergency RPV Depressurization
C. C4, RPV Flooding
D. C5, Level/Power Control
KIA Statement:
295031 Reactor Low Water Level / 2
2.4.6 - Emergency Procedures / Plan Knowledge symptom based EOP mitigation strategies
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the appropriate Emergency Procedure used to mitigate a low reactor water
level condition .
References: 2-EOI-1, EOIPM Sections O-V-C and O-V-G
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following:
1. Whether the given conditions are indicative of a 1055 of HP injection .
2. Based on Item #1 above , which EOI Contingency is appropriate to mitigate that condition.
A is correct.
B is incorrect. This is plausible since ED will eventually become necessary following the initial actions of
EOI-C1. However , additional actions are required before EOI-C2 is appropriate.
C is incorrect. This is plausible since OW temperature may be high enough following ED to create a
condition where RPV level instruments become unavailable. However, additional actions are required
before EOI-C4 is appropriate.
D is incorrect. This is plausible since the only given condition which contradicts the use of EOI-C5 is the
current rod pattern. However, with all rods inserted, EOI-C5 is not appropriate.
(
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56. RO 295037EK2.l 1 OOl/C/A/TlGI/RMCS//295037EK2.Il//RO/SROINO
A hydraulic ATWS has occurred on Unit 2 and the Unit Operator is inserting control rods in accordance
with the EOI appendices 10, 1F, & 2.
( With these plant conditions ...
A. >I all insert blocks are bypassed.
B. rod drift indication is received as soon as rod motion begins.
C. stabilizing valves open to provide increased drive pressure.
D. all RMCS timer functions are bypassed except for the settle bus.
KIA Statement:
295037 SCRAM Cond ition Present and Power Above APRM Downscale or Unknown 11
EK2.11 - Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR
POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following : RMCS : Plant-Specific
KIA Justification: This question satisfies the KIA statement by requiring the cand idate to use spec ific
plant cond itions to determine the status of the RMCS while executing procedures to mitigate an ATWS
condition .
References: 2-EOI Appendicies 1D, 1F, and 2
Level of Knowledge Justification: This question is rated as CIA due to the requ irement to assemble ,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its mean ing to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the following :
1. What affect the actions performed by EOI Append ix implementation have on the RMCS system.
2. What affect the RMCS manipulations required by implementation of the EOI Appendicis have on plant
indications.
A is correct.
8 is incorrect. This is plausible since a Rod Drift indication will occur for each inserted control rod,
however the indication does not occur until the rod is fully inserted and the CRD NOTCH OVERRIDE
switch is released .
C is incorrect. This is plausible because CRD stabilizing valves DO have an effect on drive wate r
pressure, but the efffect is to prevent oscilations while mov ing control rods , NOT increase pressure.
D is incorrect. This is plausible since RMCS timers are bypassed by using the CRD NOTCH
OVERRIDE switch in accordance with EOI Appendix 1D. However, the Settle Bus timer is also bypassed.
(
2-EOI APPENDIX-1D
...
Rev. 6
Page 1 of 3
2-EOI APPENDIX-1D
(
INSERT CONTROL RODS USING REACTOR MANUAL CONTROL
SYSTEM
LOCATION: Unit 2 Control Room, Panel 9-5
ATTACHMENTS: 1. Tools and Equipment
2. Core Position Map
NOTE: This EOI Appendix may be executed concurrently with
EOI Appendix 1A or IB at SRO's discretion when time
and manpower permit.
1. VERIFY at least one CRD pump in service.
NOTE: Closing 2-85-586, CHARGING WATER ISOL, valve may
reduce the effectiveness of EOI Appendix 1A or lB.
2. IF Reactor Scram or ARI CANNOT be reset,
THEN DISPATCH personnel to close 2-SHV-85-586,
CHARGING WATER SHUTOFF (RB NE, El 565 ft).
3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.
4. BYPASS Rod Worth Mi n i mi z e r .
5. REFER TO Attachment 2 and INSERT control rods in the
area of highest power as follows:
a. SELECT control rod.
b. PLACE CRD NOTCH OVERRIDE switch in EMERG ROD IN
position UNTIL control rod is NOT moving inward.
c. REPEAT Steps 5.a and 5.b for each control rod to be
inserted.
NOTE: A ladder may be required to perform the following
step. REFER TO Tools and Equipment, Attachment 1.
IF necessary, an alternate ladder is available at ,
the HCU Modules, EAST and West banks. It is stored
by the CRD Charging Cart.
6. WHEN ... NO further control rod movement is possible or
desired,
THEN ... DISPATCH personnel to verify open 2-SHV-85-586,
CHARGING WATER SHUTOFF (RB NE, El 565 ft).
END OF TEXT
2-EOI APPENDIX-IF
Rev. 4
(
...
2-EOI APPENDIX-1F
Page I of 7
LOCATION: Unit 2 Control Room
ATTACHMENTS: 1. Tools and Equipment
2. Panel 2-9-15, Rear
3. Panel 2-9-17, Rear
1. VERIFY Reactor Scram and ARI reset.
a. IF ...*. ARI CANNOT be reset,
THEN .*. EXECUTE EOI Appendix 2 concurrently with
Step 1.b of this procedure.
b. IF Reactor Scram CANNOT be reset,
THEN DISPATCH personnel to Unit 2 Auxiliary
Instrument Room to defeat ALL RPS logic
trips as follows:
1) REFER to Attachment 1 and OBTAIN four 3-ft banana
jack jumpers from EOI Equipment Storage Box.
2) REFER to Attachment 2 and JUMPER the following
relay terminals in Panel 2-9-15, Rear:
a) Relay 5A-K10A (DQ) Terminal 2 to Relay
5A-K12E (ED) Terminal 4, Bay 1.
b) Relay 5A-K10C (AT) Terminal 2 to Relay
5A-K12G (BH) Terminal 4, Bay 3.
3) REFER to Attachment 3 and JUMPER the following
relay terminals in Panel 2-9-17, Rear:
a) Relay 5A-K10B (DQ) Terminal 2 to Relay
5A-K12F (ED) Terminal 4, Bay 1.
b) Relay 5A-K10D (AT) Terminal 2 to Relay
5A-K12H (BH) Terminal 4, Bay 3.
2. WHEN RPS Logic has been defeated,
THEN RESET Reactor Scram.
3. VERIFY OPEN Scram Discharge Volume vent and drain valves.
57. RO 295038EKl.Ol OOlIMEMlTlGlINEWI1295038EKl.Ol//RO/SRO/RWM
Given the following plant conditions:
- Unit 2 has experienced a LOCA with a loss of Primary Containment.
(
- You have volunteered for a team dispatched from the OSC to enter the Reactor Building and
attempt to energize 20 480v RMOV board.
- Due to environmental and radiological conditions present in the Reactor BUilding, Radcon
provides you with a Sodium Chloride and Potassium Iodine tablet during the prejob briefing.
Which ONE of the following describes the benefit of ingesting Potassium Iodine prior to the Reactor
Building entry?
A. It will reduce the risk of dehydration and heat stress .
B. It will reduce the absorption of radioactive Iodine by the lungs.
C~ It will reduce the absorption of radioactive Iodine by the thyroid.
D. It will reduce the absorption of radioactive Potassium in the blood stream.
KIA Statement:
295038 High Off-site Release Rate I 9
EK1.01 - Knowledge of the operational implications of the following concepts as they apply to HIGH
OFF-SITE RELEASE RATE : Biological effects of radioisotope ingestion
KIA Justification: This question satisfies the KIA statement by requiring the candidate to correctly
identify the pathway and adverse effect of iodine ingestion.
Reference: EPIP -14 Revision 18, page 4
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
A is incorrect. The sodium chloride tablets would be used for this purpose. It is plausible if the
candidate is unsure of the purpose of KI tablets.
B is incorrect. Only the thyroid is the organ at risk, but it is plausible if the candidate assumes that
airborne ingestion is limited to absorption by the lungs.
C is correct.
D is incorrect. Iodine is the element that is absorbed. Potassium becomes a plausible answer due to
recent media coverage regarding health risks related to low potassium levels in the blood stream.
( BROWNS FERRY RADIOLOGICAL CONTROL PROCEDURES
EPIP-14
3.6 Issuing Potassium Iodide (KI)
3.6.1 If the TSC RP Manager has reason to believe that a person's projected
cumulative dose to the thyroid from inhalation of radioactive iodine might exceed
10 rems (see Appendix A), the exposed person should be started immediately on
a dose regimen of KI. This decision shall be immediately communicated to the
SED.
3.6.1.1 If the TSC is not staffed or the RP Manager position has not been filled,
then the senior onsite RP Supervisor has the authority to issue KI
utilizing the bases described in step 3.6.1.
3.6.1.2 The initial dose of KI should be not delayed since thyroid blockage
requires 30 to 60 minutes. Anyone authorized to initiate KI shall be
familiar with the Food and Drug Administration (FDA) patient package
insert and be sure that each recipient is similarly informed .
3.6.1.3 Prior to issuing KI to an individual , the person should be asked if he/she
is allergic to iodine. If the person indicates a possible sensitivity to iodine
they should not be issued KI.
3.6.2 KI is stored in the plant RP supply cage and the REP Van instrument kits.
3.6.3 RP normally will not dispense a container or package of KI to TVA Personnel
involved in activities to support a radiological emergency. RP will however
dispense a single individual dose of KI to team members dispatched from the
OSC.
3.6.4 Follow the dosage outlined on the FDA patient package insert (Appendix B). A
copy of the FDA approved patient package insert shall accompany the issuance
of KI. If KI is distributed in individual doses then verbal instructions of the
significant information on the patient package insert by a knowledgeable
individual is sufficient.
3.6.5 Complete the KI Issue Report (Appendix C) or document on an RWP time sheet
as appropriate for issuance of KI. If the RWP time sheet is used to document
distribution of the KI, note the time of KI distribution on the back of the time
sheet.
PAGE 4 OF 9 REVISION 0018
58 . RO 600000AA 1.08 00 l/MEM/Tl G lIRSWl1600000AA 1.081IRO/SRO/I1120107 RMS
Which ONE of the following describes the appropriate fire extinguishing agent for the specific class of
fire?
(
A. Water used on Class "B" fires .
B." Low pressure CO 2 used on Class "C" fires.
C. Dry Chemical (PKP) used on Class "c" fires.
D. Aqueous Film Forming Foam (AFFF) used on Class "A" fires.
KIA Statement:
600000 Plant Fire On-site I 8
AA 1.08 - Ability to operate and I or monitor the following as they apply to PLANT FIRE ON SITE: Fire
fighting equipment used on each class of fire
KIA Justification: This question satisfies the KIA statement by requiring the candidate to identify the
correct fire fighting agent for a specific class of fire.
References: TVA Safety Manual
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information .
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the following :
1. Which flammable material is of concern based on Fire Class A, Band C.
2. Which extinguishing agent is appropriate for each class of fire.
3. Which extinguishing agent is inappropriate for a given class of fire.
A is incorrect. Class "B" fires are flammable liquids. Using water could cause serious damage by
allowing the liquid to splatter and spread.
B is correct.
C is incorrect. Dry chemical agents are extremely corrosive to electrical components and insulation
typical of Class "B" electrical fires .
D is incorrect. AFFF is designed as a flooding and diluting agent for Class "B" flammable liquid fires.
Application on a Class "A" fire is not effective in extinguishing flammable materials such as wood and
paper.
(
59. RO 295009AK2.01 OOl/C/A/TlG2/PR.INSTRl13/295009AK2.0l/9619/RO/SRO/ll/20/07 RMS
Given the follow ing Unit 1 plant cond itions:
- Due to mult iple high pressure injection system failures , 1-EOI-C1, Alternate Level Control has
been entered.
- Core Spray Pumps 1Band 1D are running and lined up for injection .
- Drywell Temperature is 240 OF and rising slowly .
Which ONE of the following conditions describes the appropriate point where Emergency
Depressurization may be performed in accordance with 1-EOI-C1, Alternate Level Control?
Post Accident Flooding Range level instrument 3-L1-3-S2 is reading _ _inches with reactor pressure at
___psig.
REFERENCES PROVIDED
KIA Statement:
29S009 Low Reactor Water Level I 2
AK2 .01 - Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the follow ing:
Reactor water level indication
KIA Justification: Th is question satisfies the KIA statement by requiring the candidate to use specific
plant condit ions to determine actual reactor water level under conditions of low reactor water level.
References: 1-EOI-C1 Flowchart, PIP-9S-64 Rev 12
Level of Knowledge Justification: This question is rated as CIA due to the requ irement to assemble,
sort , and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: 1-EOI-C1 Flowchart, PIP-95-64 Rev 12.
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. Recognize the requirement that RPV level must be less than -162 inches before Emergency
Depressurization is appropriate.
2. Recognize that the indicated RPV level must be corrected for pressure using PIP-95-64.
3. Recognize that two or more injection systems must be lined up with pumps running to meet the
requirement to Emergency Depressurize.
4. Recognize that only one RHR pump is required to qualify as an injection subsystem since each RHR
pump is rated for 100% capacity.
NOTE: Each distractor is plausible because the conditions specified are possible given the current plant
conditions.
A is correct.
B is incorrect. Level is 5 inches too high or pressure is 100 psig too high.
C is incorrect. Level is -4 inches too high or pressure is 240 psig too high.
D is incorrect. Level is -4 inches too high or pressure is 100 psig too high.
(
3-LI-3-52 & 62 CORRECTION CURVES
-150"
TAF -162" - -162" =TAF (RED LINE )
-185" =MSCRWL (GREEN LINE)
-175" -200" =MZIRWL (BLUE LINE)
....J
-215" =TWO-THIRDS CORE HEIGHT (BLACK UNE)
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...... -215"
-268"
o 100 200 300 400 500 600 700 800 900 1000 1100
REACTOR PRESSURE (PSIG) PIP-95-64
REV. 12
OPL 171.003
Revis ion 17
Page 22 of 54
INSTRUCTOR NOTES
Since no trips or alarms are
associated with this range , this
level signal is not directed
through the Analog Trip System.
(d) One MCR indicator on Panel 9-3
monitors this range of level
indication .
(4) Post-accident Flood Range
(a) -268" to +32" range covering
active core area and overlapping
the lower portion of the Normal
Control Range .
(b) Referenced to instrument zero
(c) Intended for use only under
accident cond itions with reactor
at 0 psig and recirculation pumps
tripped .
(d) Variable leg tap is from diffuser of
jet pumps 1 and 6 (or 11 and 16).
(e) Per Safety Analysis on water
level instruments the conclusion Injecting with RHR
is that the accident range L1-3-52 and 62
instruments adequately indicate (Accident Range)
water level--provided they are Technical Support
corrected for off-calibration letter dated 9/13/95
conditions of RPV pressure (See LP Folder)
utilizing the operator aid on Panel Use Conservative
9-3 for level correction . Decision Making
Obj. V.8.15.
(f) An interlock associated with this Obj. V.8 .11.
range will prevent using the RHR
System for containment de
pressurization when it is needed
to flood the core region .
(g) The -68" to -168" portion of this
range is recorded in the MCR on Unit 3 Recorder
2-L1-3-62 Recorder and two displays a scale of
indicators monitor the full range +32" to -268"
of these instruments.
E MINATION
REFERENCE
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3-LI-3-52 & 62 CORRECTION CURVES
-150"
TAF -162" - -162" =TAF (RED LINE )
-185" =MSCRWL (GREEN LINE )
-175" -200" =MZIRWL (B LUE LINE )
-215" =TWO-THIRDS CORE HEIGHT (BLACK LINE)
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-250" ~
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-268"
o 100 200 300 400 500 600 700 800 900 1000 1100
REACTOR PRESSURE (PSIG) PIP-95-64
REV. 12
I
60. RO 295012G2.2.22 00 l/C/A/Tl G2/64/12/2950 12G2.2.22/IRO/SRO/0606S NEW6/28/2007
Given the following plant conditions:
- You are the oncoming Unit 3 Unit Supervisor.
- During turnover the onshift Unit Supervisor informs you that 2 Drywell Coolers had been
secured during his shift while performing ground isolation on 3C 480v RMOV board.
- Drywell Average Temperature is 152°F and stable.
Which ONE of the following describes the appropriate condition and required action?
A'! Exceeded 3-SR-2, Instrument Checks and Observations, Drywell temperature limit. Address Tech
Spec section 3.6.
B. Exceeded the normal operating Drywell temperature limit. Drywell temperature must be logged
hourly until below the limit.
C. Exceeded the normal operating Drywell temperature limit. Restore Drywell average air temperature
below the limit in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> .
D. Exceeded 3-EOI-2, Primary Containment Control entry condition . Enter and execute 3-EOI-2,
Primary Containment Control.
KIA Statement:
295012 High Drywell Temperature 15
2.2.22 - Equipment Control Knowledge of limiting conditions for operations and safety limits
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine that Technical Specification limits have been exceeded .
References: Unit 3 Tech Specs Section 3.6.1.4
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: 1-EOI-C1 Flowchart, PIP-95-64 Rev 12.
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. The appropriate entry condition for U3 Tech Spec Section 3.6.1.4.
2. The appropriate entry condition for 3-EOI-2, Primary Containment Control.
3. The appropriate action based on the given condition.
A is correct.
B is incorrect. This is plausible because the Tech Spec Iimt was exceeded, however the required action
is to restore the Orywell Temperature within the limit in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> . There is no requirement for hourly logging
of OW temperature.
C is incorrect. This is plausible because the Tech Spec limt was exceeded, however the required action
is to restore the Orywell Temperature within the limit in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> . The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit is based on performing
the surveillance on Orywell Temperature.
o is incorrect. This is plausible because the entry condition for 3-EOI-2 is only 8 of above the given
temperature. However, the entry condition has not been met and OW temperature was reported as
"stable".
Drywell Air Temperature
3.6.1.4
3.6 CONTAINMENT SYSTEMS
3.6.1.4 Drywell Air Temperature
LCO 3.6.1.4 Drywell average air temperature shall be s 150°F.
APPLICABILITY: MODES 1, 2, and 3.
ACTIONS
CONDITION REQUIRED ACTION COMPLETION
TIME
A. Drywell average air A.1 Restore drywell average 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
temperature not within air temperature to within
limit. limit.
B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Completion
Time not met. AND
B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
BFN-UNIT 3 3.6-17 Amendment No. 212
Drywell Air Temperature
3.6.1.4
(
SURVEILLANCE REQUIREMENTS
SURVEILLANCE FREQUENCY
SR 3.6.1.4.1 Verify drywell average air temperature is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
within limit.
( BFN-UNIT 3 3.6-18 Amendment No. 212
61 . RO 2950 15AK l. 02 OOlIMEMlTlG2IBASISI1295015AKl.02///lll21107 RMS
EOI-1 flowchart path RC/Q directs the operator to inhibit the ADS auto blowdown function once Standby
Liquid Control injection has begun.
( Which ONE of the following describes why ADS is inhibited under these conditions?
A. ADS actuation would impose a severe pressure and temperature transient on the reactor vessel.
B. The operator can control pressure better than an automatic system like ADS.
C. otI Severe core damage from a large power excursion could result , if low pressure systems
automatically injected on depressurization.
D. If only steam driven high pressure injection systems are available an ADS actuation could lead to a
loss of adequate core cooling .
KIA Statement:
295015 Incomplete SCRAM I 1
AK1.02 - Knowledge of the operational implications of the following concepts as they apply to
INCOMPLETE SCRAM : (CFR 41.8 to 41.10) Cooldown effects on reactor power
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the effect of a significant cooldown when an incomplete scram has
occurred .
References:
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the follow ing:
1. The basis for inhibiting ADS unde r the spec ific cond itions of boron injection.
NOTE: Each of the three distractors are plausible based on their relationsh ip to the bases for inhibiting
ADS under circumstances OTHER than boron injection. Specifically, Alternate RPV Level Control
actions . Refer to the attached excerpt from EOIPM Section O-V-G.
A is incorrect. This appl ies whenever ADS actuates , but is only the precursor to the issue related to
boron injection.
B is incorrect. This statement is true , but is not addressed in the basis for boron injection.
C is correct.
D is incorrect. This statement applies particularly to a low RPV level condition.
I EOI PROGRAM MANUAL
SECTION O-V-C
EOI-1, RPV CONTROL BASES
( STEP: RC/Q-14 and RC/Q-15
EJlECUlE RC/Q- , 2 "",0 RC/Q-22 CONCVRRENllY
Re/Q-'1
BORON INJ IS REQUIRED
(£01- I. RC/P-8)
L
Re/Q-13
'--- -r-e- ----' L
Rc/Q-14
INHlBIl NlS
L- ---, -.JL
RC/Q-15
VERIFY RWCU SYSTEI.I ISOLATION
L
RC/Q-1S
WHILE EXECUTING THE FOLLOWING STEPS:
1£ Ii:!lli
SLC tNlK LVl DROPS TO c....3>. tRIP THE SLC PUMPS
~~-
RC/Q-I7 _ _- . . .- - - - -. . . L
034
SECTION o-v-c PAGE 116 OF 127 REVISION 1
EOI-1, RPV CONTROL BASES EOI PROGRAM MANUAL
SECTiON o-v-c
- 1 DISCUSSION: STEP RC/Q-14 and RC/Q-15 1
The RC/Q-14 action step directs the operator to manually initiate the SLC System. Because this step is prioritized
with the miniature before decision step RClQ-12 symbol, this action should be performed before suppression pool
temperature reaches <A.64>, Boron Injection Initiation Temperature. EOI Appendix 3A provides step-by-step
guidance for manual initiation of the SLC System. Boron in solution absorbs neutrons, providing negative
reactivity to achieve reactor subcriticality, since the reactor is not yet subcritical on control rod insertion alone.
The RC/Q-15 action step directs the operator to defeat automatic ADS function by placing the ADS inhibit
switches in the inhibit position. Because this step is prioritized with the miniature before decision step RC/Q-12
symbol, this action should be performed before suppression pool temperature reaches.<A.64>, Boron Injection
Initiation Temperature.
ADS initiation may result in the injection of large amounts of relatively cold, unborated water from low pressure
injection systems. With the reactor still critical or subcritica1 on boron , the positive reactivity addition due to boron
dilution and temperature reduction from injection of cold water may result in a reactor power excursion large
enough to cause substantial core damage. Defeating ADS is, therefore, appropriate whenever boron injection is
required. If emergency depressurization of the RPV is subsequently required, explicit direction is provided in the
appropriate EOL Therefore, the ability to maintain automatic initiation capability of ADS is not required.
( * REVISION 1 PAGE 117 OF 127 SECTION O-V-C
EOI PROGRAM MANUAL C1, ALTERNATE LEVEL CONTROL BASES .
SECTION O-V-G
STEP: Cl-l
I fOl-l
RPv CbNTROL
RC/L-12
~~>::;~;F(./'~~;:* *~'CAut(O~-: * ...
1 .. ,
., ." fH: :
~ #1
t:
AMBiENT TEMP MAY AffECT RPv WATER LVl
INDICATION AND TREND i
. ," ' ' ~ . , ....., ..
~ " .- . " ....'. "; ' .' ~' . ~
L
INHIBIT ADS
I IL
Cl **1
WH ILE EXECUTING TH IS PROCEDURE:
lE .I1::f.£.N
ALL CONTROL RODS ARE l\lOI
-.. ' .- -
ExIT THIS PROCE~RE AND
INSERTED TOPc¥: BFYOND ENTER C5. LEVEL POWER CONTROL
POSITION <A. 0> ,
RPV WATfR L~ CMINQI [XIT THIS PROCEDURf AND
BE DETERMINE , ENTER C4, RPV flOODING
RPV WATER LV!. IS RISING. EXIT THIS PROCEDURE AND
LNT£R £01-1, RPV CONTROL,
AT STEP RC/L-l
L
CI-2
SECTION O-V-G PAGE 8 OF 50 REVISION 0
cr, ALTERNATE LEVEL CONTROL BASES EOI PROGRAM MANUAL .
SECTION O-V-G .
1 DISCUSSION: STEP Cl-l 1
This action step directs the operatorto defeat automatic ADS function. An ADS actuationwith
the RPVat pressure imposesa severethermaltransienton the RPV and may significantly
complicate efforts to restore and maintainRPV water level as specified in this procedure.
.
Because ADSinitiation logic receiveslimitedinputsignals,a variety ofplant conditions mayexist
whereautomatic depressurization of the RPV is not appropriate. In certain cases (e.g., RCIC
available but LPCI/CS injectionvalves closed and controlpower for their operationnot available)
ADS actuation may directly lead to loss of adequate core cooling and core damage, conditions
that mightotherwisehave been avoided. Further,conditions assumed in the design of ADS
actuation logic (e.g., no operator action for ten minutes) do not exist when actions specified in
this procedure are being carried out. .
Finally, an operatorcan draw on much more plant information than is availableto ADS logic
(e.g., equipmentout ofservice for maintenance, operating experiencewith certain systems,
probability of restoration of offsite power, etc.) and thus can betterjudge, based on logic specified
in this procedure, when and how to depressurize the RPV. For all of these reasons, it is
appropriate to prevent automatic initiation ofADS as specified.
- REVISION 0 PAGE 9 OF 50 SECTION O-V-G
62. RO 295020AK3 .08 OOl/MEMlEOI/BASIS//295020AK3.08///l1/21/07 RMS
Unit-2 was at 100% rated power when a spurious Group I isolation occurred. The pressure transient
caused a small LOCA to occur inside the drywell.
EOI-2, section PC/P requires certain actions before and after reaching 12 psig Suppression Chamber
pressure .
Which of the following is the reason that 12 psig in the Suppression Chamber was selected?
A. Drywell sprays must be initiated prior to this pressure to prevent opening the Suppression Chamber
to Reactor Building vacuum breakers and de-inerting the containment.
B. Above this pressure indicates that almost all of the nitrogen and other non-condensible gases in the
drywell have been transferred to the torus so initiating Drywell Sprays will not result in containment
failure .
C.oI Above this pressure indicates that almost all of the nitrogen and other non-condensible gases in the
drywell have been transferred to the torus and chugging is possible .
D. Above this pressure indicates that almost all of the nitrogen and other non-condensible gases in the
torus have been transferred to the drywell air space and Suppression Chamber Sprays will be
ineffective.
KIA Statement:
295020 Inadvertent Cont. Isolation I 5 & 7
AK3.08 - Knowledge of the reasons for the following responses as they apply to INADVERTENT
CONTAINMENT ISOLATION: Suppression chamber pressure response
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the effect on Suppression Chamber pressure due to an inadvertent
containment isolation and the basis for that response .
References: EOIPM Section O-V-D
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information.
0610 NRC Exam
REFERENCE PROVIDED: None
j
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. The basis for the Pressure Supression Pressure Limit of 12 psig Suppression Chamber pressure.
A is incorrect. This is plausible because initiation of OW sprays at high SC pressure could reduce
pressure low enough to open the SC to RB vacuum breakers. However, this is part of the bases for the
Orywell Spray Initiation Pressure Limit Curve #5.
B is incorrect. This is plausible because initiating SC sprays with high temperature non-condensible
gases in the SC will result in evaporative cooling and a rapid pressure drop. However, the SC to OW
vacuum relief system is capable of compensating for this pressure drop. This is also part of the bases for
the Orywell Spray Initiation Pressure Limit Curve #5 .
C is correct.
D is incorrect. This is plausible if the LOCA occurred inside the Suppression Chamber and NOT the
Orywell as given in the stem .
EOI PROGRAM MANUAL EOI-2, PRIMARY CONTAINMENT CONTROL BASES
SECTION O-V-O
(
STEP: PC/P-6 .
NO L
YES L
o
INITIATE SUPPR CHMBR SPRAYS USII\C w..x RHR PUMPS WI
REQI)RED TO ASSURE ADEOUA.TE CORE COOLING BY" COO11NlJ()J$
IN! (N)PX 17C)
L
PC/P-5
PC/P-6
SUPPR CHI.4BR PRESS EXCEEDS <A.65>
CONTINUE IN THIS PROCEDURE
L
NO L
TO CURVE 5
(Refer to EOI Program "'onuot "-
Section IV, Appendi. A. Curves
and Tables Used in the EOls)
NO L
TO PC/P-ll
(
i
SECTION O-V-O PAGE 42 OF 244 REVISION 0 *
EOI-2. PRIMARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL
SECTION o-v-n
1 DISCUSSION: STEP PCIP-6 1
This contingent action step requires the operator to wait until the stated condition has been met
before continuing in EOI-2. Performance of subsequent actions in this section ofEOI-2 will not
be performed until suppression chamber pressure exceeds Suppression Chamber Spray Initiation
Pressure.
Engineering calculations have determined that ifsuppression chamber pressure exceeds <A.65>,
Suppression Chamber Spray Initiation Pressure, there is no assurance that chugging will be
prevented at downcomer openings of the drywell vents. This value is rounded -off in the EOI to
use the closest, most conservative value that can be accurately determined on available
instrumentation.
Suppression Chamber Spray Initiation Pressure is defined to be the lowest suppression chamber
pressure that can occur when 95% of noncondensables in the drywell have been transferred to
airspace ofthe suppression chamber. Scale model tests have demonstrated that chugging will not
occur so long as the drywell atmosphere contains at least 1% noncondensables. To prevent the
occurrence of conditions under which chugging may happen, Suppression Chamber Spray
Initiation Pressure is conservatively defined by specifying 5% noncondensables.
- Chugging is the cyclic condensation of steam at downcomer openings ofthe drywell vents.
Chugging occurs when steam bubbles collapse at the exit of downcomers. The rush ofwater that
fills the void (some of which is drawn up into the downcomer pipe) induces a severe stress at the
junction of the downcomer and vent header. Repeated application ofthis stress can cause these
joints to experience fatigue failure (cracks), thereby creating a pathway that bypasses the pressure
suppression function ofprimary containment. Subsequent steam that discharges through
downcomers would then exit through the fatigued cracks, and directly pressurize suppression
chamber air space, rather than discharging to and condensing in the suppression pool.
Although operation of suppression chamber sprays by itselfwill not prevent chugging, the
requirement to wait to initiate drywell sprays until reaching Suppression Chamber Spray Initiation
Pressure assures that suppression chamber spray operation is attempted before operation of
drywell sprays. Therefore, actions to initiate drywell sprays need to be directed only if suppression
chamber sprays were unable to reduce primary containment pressure or they could not be
initiated.
- REVISION 0 PAGE 43 OF 244 SECTION O-V-O
EOI-2, PRIMARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL
SECTION O-V-O
- I DISCUSSION: STEP PCIP-8 ~
This decision step has the operator evaluatethe present status of drywell pressure and drywell
temperature to determine if conditionsare favorable for drywell spray operation.
Drywell spray operationreduces drywellpressureand temperature through the combinedeffects
of evaporative and convective cooling.Duringevaporative cooling, water spray undergoesa
changeof state, liquid to vapor, whereas convective coolinginvolves no change of state.
Evaporative cooling occurs when water is sprayedinto a superheatedatmosphere. Water at the
surfaceof each droplet is heated and flashesto steam, absorbingheat energy from the drywell
atmosphere until the atmosphere reaches saturated conditions. In the drywell, with a typical
drywell sprayflowrate, the evaporativecoolingprocessresults in an immediate,rapid, large
reduction in pressure. This pressure reductionoccurs at a rate much faster than can be
compensated for by the primary containment vacuumrelief system. Unrestrictedoperationof
drywell sprayscould cause an excessivenegativedifferential pressure to occur between the
drywell and suppressionchamber, large enoughto causea loss of primary containmentintegrity.
Convective cooling occurs when water is sprayedinto a saturated atmosphere. Sprayed water
droplets absorbheat from the surrounding atmosphere through convectiveheat transfer (sensible
heat from the atmosphere is transferredto the water droplets). This effect reduces drywell
ambienttemperature and pressure until equilibrium conditionsare established.The convective
coolingprocess occurs at a rate much slowerthan the evaporative cooling process. An operator
can effectively control the magnitudeofa containment temperature/pressure reduction from
convective coolingby terminatingoperationof drywellsprays.
Considering the pressure drop concernsdescribed above,engineeringcalculationshave
determined that primary containmentintegrityis assuredwhen drywell sprays are operated in the
safearea ofDrywell Spray Initiation Limit Curve(Curve5). DrywellSpray InitiationLimit is
defined to be the highest drywell temperature at whichinitiationofdrywell sprays will not result
in an evaporative cooling pressure drop to beloweither: 1) drywell-below-suppression chamber
differential pressurecapability,or 2) high drywell pressurescram setpoint,
If drywell temperature and pressure are within the safe area of Curve 5, the operator continuesat
Step PCIP-9.
Ifdrywell temperature and pressure are not withinthe safe area of Curve 5, then drywell spray
operation is not permitted, and the operator is directed to Step PCIP-l1 .
- REVISION 0 PAGE 47 OF 244 SECTION O-V-D
63 . RO 295032EAl.OI OOl/C/A/TlGI/E0I-3//295032EAl.Ol//RO/SR0/1l/20/07 RMS
Given the follow ing plant cond itions :
- Unit 2 experienced a MSL break from full power.
c' * Both inboard and outboard MSIVs on the "B" steam line fail to isolate however, the reactor
scrams and all rods insert.
- Steam Leak Detection panel 9-21 indications are as follows :
- 2-TI-1 -60A 320°F
- 2-TI-1-60B 323°F
- 2-TI-1-60C 33rF
- 2-TI-1-60D 318°F
- No other temperature indications are alarming at this time .
Which ONE of the following describes the appropriate operator actions and the reasons for those
actions?
REFERENCE PROVIDED
A. Emergency depressurize the reactor due to two EOI-3 areas being above Max Safe .
B. Rapidly depressurize the reactor due to one EOI-3 areas above Max Safe and one area
approaching Max Safe .
c.'; Enter 2-EOI-1, RPV Control and initiate a Reactor Scram due to one EOI-3 area being above Max
Safe.
D. Enter 2-GOI -100-12A, Unit Shutdown and commence a normal shutdown and cooldown due to a
primary system dscharging outside Primary Containment.
KIA Statement:
295032 High Secondary Containment Area Temperature / 5
EA1.01 - Ability to operate and/or monitor the follow ing as they apply to HIGH SECONDARY
CONTAINMENT AREA TEMPERATURE : Area temperature monitoring system
KIA Justification: Th is question satisfies the KiA statement by requiring the candidate to use spec ific
plant cond itions to determine the required actions which result from high secondary conta inment
temperatures as indicated by Area Temperature Monitoring instrumentation.
References: 2-EOI -3 Flowchart, EOIPM Sect ion O-V-E
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. Th is requ ires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
(
REFERENCE PROVIDED: 2-EOI-3 Flowchart
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. Which area(s) are above or approaching Max Safe
2. Based on Item #1 above, determine the appropriate action and the basis for that action .
A is incorrect. This is plausible because all four temperatures provided are greater than 3150F as
indicated on Table 3. However, only one indicator applies to an EOI 3 area, therefore only ONE area is
above Max Safe.
B is incorrect. This is plausible because one area is above Max Safe and given conditions indicate an
un-isolable leak exists which implies conditions are degrading. However, with no other temerature
indications in alarm, anticipating the requirement to Emergency Depressurize is NOT ppropriate.
C is correct.
D is incorrect. This is plausible because all four temperatures provided are greater than 3150F as
indicated on Table 3. However, only one indicator applies to an EOI 3 area, therefore only ONE area is
above Max Safe. In addition, this step is only addressed if Emergency Depressurization will not reduce the
discharge into Secondary Containment. In this case, it would .
(
EOI-3, SECONDARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL
SECTION O-V-E
- ~-- DISCUSSION: ENTRY CONDITIONS: EOI-3
I
Entry conditions for this procedure are symptomatic of conditions which, if not corrected, could degrade into an
emergency. Adverse affects on equipment operability and conditions that directly challenge secondary containment
integrity were specifically considered in the selection of these entry conditions. Following is a description of each
entry condition:
Area temperature above the maximum normal operating value of Table 3
A secondary containment area temperature above the maximum normal operating value of Table 3, Secondary
Containment Area Temperature, is an indication that steam from a primary system may be discharging into
secondary containment. As temperatures continue to increase, continued operability of equipment needed to carry
out EOI actions may be compromised. High area temperatures also present a danger to personnel since access to
secondary containment may be required by actions specified by EOls.
Maximum normal operating temperature is defined to be the highest value of a secondary containment area
temperature expected to occur during normal plant operating conditions with all directly associated support and
control systems functioning properly.
Differential pressure at or above <A.38> inches of water
High secondary containment differential pressure is indicative of a potential loss of secondary containment
structural integrity, and could result in uncontrolled release of radioactivity to the environment.
- Reactor Zone Ventilation exhaust radiation level above <A.39>
High Reactor Zone Ventilation exhaust radiation levels may indicate that radioactivity is being released to the
environment when the system should have automatically isolated.
Refuel Zone Ventilation exhaust radiation level above <A.40>
High Refuel Zone Ventilation exhaust radiation levels may indicate radioactivity is being released to the
environment when the system should have automatically isolated.
Floor drain sump water levcl above <A.41>
A secondary containment floor drain sump water level above maximum normal operating level is an indication that
steam or water may be discharging into secondary containment.
Maximum normal operating floor drain sump water level is defined to be the highest value of secondary
containment floor drain sump water level expected to occur during normal plant operating conditions with all
directly associated support and control systems functioning properly.
Area watcr level above <A.42>
Secondary containment area water level above maximum normal operating level is an indication that steam or water
may be discharging into secondary containment.
Maximum normal operating secondary containment area water level is defined to be the highest value of secondary
containment area water level expected to occur during normal plant operating conditions with all directly associated
support and control systems functioning properly.
REVISION 1 PAGE 9 OF 73 SECTION O-V-E
.... .... . . .... _._ - - - - - - - - - - - - - - - - - - - - - - - - - - - - _ ....
EOI-3, SECONDARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL
SECTION O-V-E
(
el DISCUSSION: scrr-6 and SCrr-7 1
Step SCrr-6 is a before decision step that has the operator evaluate current and future efforts to lower secondary
containment area temperatures, in relation to the current value and trend of secondary containment area
temperatures, to determ ine if a reactor scram is necessary. The before decision step requires that this determination
and subsequent actions be performed before any secondary containment area temperature reaches its respective
maximum safe operating temperature value provided in Table 3.
Maximum safe operating temperature is defined to be the highest temperature at which neither: I) equipment
necessary for the safe shutdown of the plant will fail, nor 2) personnel access necessary for safe shutdown ofthe
plant will be prevented. The maximum safe operating temperature value for all secondary containment areas is
provided in Table 3, Secondary Containment Area Temperature.
This step is reached only when additional actions have been required to reverse an increasing secondary
containment area temperature trend . If all secondary containment area temperatures can be maintained below their
respective maximum safe operating values, the operator returns to Step SCrr*I . Ifit is determined that all
secondary containment area temperatures cannot be maintained below their respective maximum safe operating
values, the operator continues at Step scrr-7.
Step SCrr-7 is an enter and execute concurrently step that requires the operator to enter
EOI-I, RPV Control, at Step RC-I , and to perform the actions concurrently with this procedure. Because this step
is prioritized with the miniature before decision step symbol relating to SCrr-6, this action should be performed
before any secondary conta inment area temperature reaches its respective maximum safe operating value .
- Initiation of reactor scram (Step RC-I) before any secondary containment area temperature reaches its respective
maximum safe operating value may halt the increase in secondary containment area temperature(s), since the RPV
is the only significant source of heat, other than a fire, that could cause secondary containment area temperatures to
exceed their respective maximum safe operating values.
REVISION 1 PAGE 27 OF 73 SECTION O-V-E
/"""'\
TABLE 3
SECONDARY CNTMT AREA TEMP
PANEL~ PANEL9-21 MAX MAX POTENTi6,L
AREA AlARM WINDOW TEMP EleMENT NORMA\. SI.,FE ISOlATION
(UNLESS NOTED) (UNLESSNOTED) VALUEof VALUEof SOURCES
RHRSYS I PUMPS XA-6!>-36-4 74-95.1., A\..I.,RMED 160 FCV-7447,48
RHRSYS II PUMPS X6,-65-3E-4 74-958 ALARMED 210 FC\L1441,48
HPCI ROQM X"I-65-3F-l0 73-65.1., ALI.,RMED 270 FCV-73-2, 3,44, 81
CS SVS I PUMPS
XA-.55-30-10 11-41A ALARMED 190 FCV-1'-2, 3. 39
RCICRooM
CSSYSIlPUMPS X4-.55-3E-29 7lH>9B (pANEL~) ALARMED 150 NONE
X6,-.55-3D-10 71041S,C, D AlARMED 200 FC\L1'-2,3
TO? OFTORUS X6,-55-3F-10 73-55B.C,D Al.6,RMED 240 FCV-13-2, 3, 81
XA-65-3E-4 74-9SG, H AlARMED 240 FCV-1404i.48
1-00.4 (PANEL9-3) Al.A,RMED 315 MSIVa
STEAMTUNNEL(RB) XA-.5!:--3D-24
FCV-71-2, 3, FCV-'S9-1, 2, 12
QWACCESS X4-55-3E-4 74-95E Al.6,RMED 170 FC\L74047.48
R8 EL565W X4-55-.5B-32 (PANEL9-5) 69-S3SA.B,C.D ALARMED
170 FCV~-l, 2, 12
(RWCUPIPE TRENCH) XI.,-.55-.5B-33 (PANEL9-5) {AUXINST ROOM} ALARMED
RWCUH. X. ROOM X4-5'".>-3D-t1 G9-29F,G, H AL4.RMED 220 FCV~-1. 2, 12
RWCUPUMPA X4.-.55-3D-17 G9-29D ALARMED 215 FCV-&.:!-1.2. 12
~"'CUPUMP8 XA-55-3D-17 G9-29E ALARMED 215 FCV.(;'9-1, 2.12
RBE L593 X4-.55-3E-4 74-95C, D Al.6,RMED 195 FCV-14047,48
RBEL621 XI.,-.55-3E-4 74-95F ALl.,RMED 155 FCV-43-13, 14
...J
.J
...J
I:l
...J !l! ~
.
~
I~
~i
~~
.... lZ ~
.. ~
i'
S~
Ii
.. Il.!
I
SIC.
U
all
~~ I: )o\!.l.
~~ ! J
I.*
m ~I ~
i
~
'"
..
WHILE EXECUTING THE FOLLOWING STEPS:
IF THEN
EMERGENCY RPV DEPRESSURIZATION
IS ANTICIPATED
AND
RAPIDLY DEPRESSURIZE THE RPV
WITH THE MAIN TURB BYPASS VLVs
)-- THE REACTOR WILL REMAIN SUBCRITICA,L
WITHOUT BORON UNDER ALL CONDITIONS
(SEE NOTE)
IRRESPECTIVE OF COOLDOWN RATE
L
RCIP-3
E MINATION *
REFERENCE
PROVIDED TO
( CANDIDATE
C")
-ow* J j. ~
il"u
!
hl~i
3
I =
I
1 ,L.a~.....,..,..
il'
~I
W-l-\++-t+tT1
SIU': "'H.++-t+t
-fa
(If)
lill *I H .. " !
[n '*11
II ' i
- 1' . .i
l it! II!! !l II
.! Ii! . i :1* .
illili II ! iit 11 I I
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II~ I il ill Ii I
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1I I I II 'II
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-
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ow
64. RO 295033EA2.01 OOl/C/A/TlG2/SC/R//295033EA2.0l//RO/SR0/1l/20/07 RMS
Given the following plant conditions:
- Unit-2 is at 100% rated power.
- A RWCU drain line cracked and is spilling into the Reactor Building .
- Area Radiation Monitors in the Reactor Building read as follows:
Reactor Building Elevation 593 1100 mR/hr
Reactor Building Elevation 565 West 800 mR/hr
Reactor Building Elevation 565 East 850 mR/hr
Reactor Building Elevation 565 Northeast 1050 mR/hr
All other Reactor Building areas NOT ALARMED
RWCU has been successfully isolated
Which ONE of the following describes the required action that MUST be directed by the Unit Supervisor
and/or Shift Manager?
REFERENCE PROVIDED
A. Enter 2-EOI-1, RPV Control and initiate a Reactor Scram due to two EOI-3 areas being above Max
Safe.
B.oI Enter 2-GOI-100-12A, Unit Shutdown and commence a normal shutdown and cooldown due to two
areas above Max Safe with the source of the leak isolated .
C. Scram the reactor and Emergency depressurize the reactor due to two EOI-3 areas being above
Max Safe.
D. Rapidly depressurize the reactor with Bypass Valves due to Emergency Depressurization being
anticipated.
KIA Statement:
295033 High Secondarv Containment Area Radiation Levels / 9
EA2.01 - Ability to determine and/or interpret the following as they apply to HIGH SECONDARY
CONTAINMENT AREA RADIATION LEVELS: Area radiation levels
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the required actions which result from high secondary containment radiation
levels as indicated by Area Radiation Monitoring instrumentation.
References: 2-EOI-3 Flowchart
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: 2-EOI-3 Flowchart
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. Which area(s) are above or approachin,g Max Safe
2. Based on Item #1 above, determine the appropriate action .
A is incorrect. This is plausible because this action requires at least one area greater than Max Safe.
However, this is not appropriate since the source of the leak is isolated.
B is correct.
C is incorrect. This is plausible because this action requires two areas greater than Max Safe. However,
this is not appropriate since the source of the leak is isolated .
D is incorrect. This is plausible because this action requires at least one area greater than Max Safe and
another area approaching Max Safe . However, this is not appropriate since the source of the leak is
isolated .
TABLE 4
- SECONDARY CNTMTAREA RADIATION
APPUCABLE MAX MAX P01ENTIAL
AREA RADIATION NORMAL SAFE ISOLATION
INDICATORS VALUEMRlHR VALUEMR/HR SOURCES
RHR SYS J PUMPS 9o-2!iA 1000 FCV-74-41* .1a
ALARMED
RHRSYSII PUMPS 9O-2M 1000 FCV-74-41* .1a
ALARMED
HPClROOM 9O-2AA 1000 FC~~2.3.44.81
AlARMED
CSSYS I PUMPS
9O-26A ALAfUAED 1000 FCV-71-2.3.~
RClCROOM
CS SYSIIPUMPS <X>-27A ALARMED 1000 NONE
FCV-13-2. 3. 81
TQPOFTORUS ~29A ALARMED 1000 FCV-74-47.48
GENERAL AREA
FCV-71-Z,3
RBEI.$5W ~20A FCV~1.2.12
ALARMED 1000
SDVVENfS & DRAINS
~21A SDVVENfS 8; DRAINS
RBEl.56SE ALARMED 1000
RBEI.!05NE ~~A AlARMED 1000 NONE
~22A ALARMED 100.000 TIP 8All VALVE
TIPROOM
R8E1.593 ~13A.14A ALARMED 1@ FCV-74-47.48
R8 El.G21 9O-9A ALARMED 1000 FCV-43-13,14
RECIRC tAG SETS 00-4A ALARMED 1000 NONE
REFUEl. FLOOR 00-1A. 2A. 3A AI.A.QMED 1000 NONE
,,I'
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AADlATJONlVlslN20R MOREAAEAS
AAEA80VE MAXSAFE{rABLE4} .~ _._ -.__ _._- - \ \ ....
ll1EN CONnNUE
L
SClR-7
EMERGENCY RPV DEPRESSURIZAnON IS REQUIRED
{EOI-1,RClN; C1-2; C1-21;C5-1)
SCJR-8
.,
~
WHILE EXECUTING THE FOLLOWING STEPS:
IF THEN
EMERGENCY RPV DEPRESSURIZATION
IS ANTICIPATED
AND RAPIDLY DEPRESSURIZE THE RPV
WITH THE MAIN TURB BYPASS VLVs
. THE REACTOR WILL REMAIN SUBCRITICAL IRRESPECTIVE OF COOLDOWN RATE
WITHOUT BORON UNDER ALL CONDITIONS
(SEE NOTE)
L
E MINATION
REFERENCE
PROVIDED TO
( CANDIDATE
(If)
- iI
J~
( o
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J
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ill II _
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f
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65 . RO 295035EA2.02 OOl/C/A/EOI/EOI-3/S85/29503 5EA2 .02//RO/SRO/
Given the following plant conditions:
- Unit 2 is at 100% power.
- During the backwash of a RWCU demineralizer the backwash receiv ing tank ruptured .
- The RWCU system has been isolated.
- Secondary containment conditions are as follows :
- All Reactor and Refuel Zone radiation monitors trip on high radiation.
- ONLY SGT train "C" can be started.
- It is operating at 10000 scfm and taking suct ion on the refuel and reactor zones .
- Refuel zone pressure : -0.12 inches of water.
- Reactor zone pressure : +0.02 inches of water.
- AREA RADIATION LEVELS
RB EL 565 W, 565 E, 565 NE: 250 mr/hr
RB EL 593 upscale
RB EL 621 upscale
Wh ich ONE of the following describes the required action and the type of radioactive release in progress?
REFERENCE PROVIDED
A. Initiate a shutdown per 2-GOI-100-12A. Elevated radiation release .
B~ Initiate a shutdown per 2-GOI-100-12A Ground level radiation release .
C. Scram the reactor, emergency depressurize the RPV. Elevated radiation release .
D. Scram the reactor, emergency depressurize the RPV. Ground level radiation release .
KIA Statement:
295035 Secondary Containment High Differential Pressure
EA2.02 - Ability to determine and/or interpret the follow ing as they apply to SECONDARY
CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Off-site release rate: Plant-Specific
KIA Justification: This question satisfies the KIA statement by requiring the cand idate to correctly
identify the type of off-site release and required actions due to high differential pressure in the secondary
containment.
References: 2-EOI-3 Flowchart
Level of Knowledge Justification: This quest ion is rated as CIA due to the requirement to assemble ,
sort, and integrate the parts of the question to solve a problem . This requires mentally using this
knowledge and its meaning to resolve the problem .
0610 NRC Exam
REFERENCE PROVIDED: 2-EOI-3 flowchart
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the following :
1. Which area(s) are above or approaching Max Safe
2. Based on Item #1 above, determine the appropriate action .
3. Whether plant conditions indicate an elevated or ground level release .
NOTE: EOI-3 steps SC/R-8 and SC/R-9 apply, requiring shutdown per 2-GOI-100-12A because 2 or
more areas are above max safe rad levels but a primary system is not discharging to the RB. Insufficient
RB to atmosphere dp (greater than -0.25 inches of water) indicates loss of secondary containment
integrity. The positive reactor zone pressure is causing an unmonitored and uncontrolled ground level
release of radioactive contaminants.
A is incorrect. The release from the Reactor Building is not elevated . This is plausible because the
required actions are correct except the differential pressure results in a ground level release .
B is correct.
C is incorrect. Conditions do not warrant a scram at this point. In addition, the release from the Reactor
Building is not elevated . This is plausible if the candidate fails to recognize that a primary system is not
discharging to the Reactor Building .
D is incorrect. Conditions do not warrant a scram at this point. This is plausible if the candidate fails to
recognize that a primary system is not discharging to the Reactor Building.
TABLE 4
-
SECONDARY CNTMT AREA RADIATION
APPLICABLE MAX MAX POTENTIAL
AREA RADIATION NORMAL SAFE ISOLATION
INDICATORS VALUEMR/HR VALUEMRIHR SOURCES
RHR SYS I PUMPS 9o-~ 1000 FCV-74-47.48
ALARMED
RHR SYS II PU}.1PS gO-2M 1000 FCV-14-47.48
ALARt.1ED
HPCIROOM 9O-2AA ALARMED 1000 FCV-73-2. ~. 44. 81
CSSYS I PUMPS
9O-2GA AlARMED 1000 FCV-11 -2.~ . 39
RCICROOM
CS SYS II PUMPS 9O-27A ALARMED 1000 NONE
FCV-73-2.3, 81
TOP OF TORUS
9O-29A AlARMED 1000 FCV-74-41. 48
GENERAL AREA
FCV-11-2 ,3
RBEl&S5W 9O-2OA 1000 FCV~1 .2.12
AlARt.IED
SOVVENrS &. DRAINS
RB El!i65 E 90-21A SOVVENrS 8. DRAINS
AlARMED 1000
RBEl!i65 NE 9O-23A AlARMED 1000 NONE
TIP ROOM 9O-22A AlARMED 100.000 nPBAtt VALVE
RBEl593 90-13A., 14A AlARMED 100:> FCV-14-47.48
RBEl621 9O-9A AlARMED 100:> FCV-43-13.14
RECIRC MG SETS 9iJ.4A AlARMED 1000 NONE
REFUel A.OOR 9O-1A. 2A, 3A AlARMED 1 (XX) NONE
t~
ISOIoATe&.SYSTEMS THATARE DISCHARGING
INTOTHEAREAEXCEPT SYSTEMSREQUIRED TO;
- 8EOPEAAlEO 8YEO's
QR
.. SUPPRESS A FIRE
L
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VllHEN RADIATION LVls IN :2OR MOREAREAS
AREABOVE MAXSAFE{TABLE4) ._ _---_._--_ _._ _ \ \
D:!.Sti CONl1NUE
L
SCJR-7
EMERGENCY RPV DEPRESSURlZAll0N IS REQUIRED
(EOI-1, RClP-4: C1-2; C1-21; C5-1)
L
SCIR-8
E MINATION
REFERENCE .
PROVIDED TO
CANDIDATE
(
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66. RO GENERIC 2.1.33 OOl/C/A/T3///GENERIC 2.1.33//RO/SRO/ll/25/07 RMS
Which ONE of the following describes the protective function(s) required to be Operable for the specified
mode and/or condition?
A. Starting up in Mode 2 with IRM's on range 1 to 2:
APRM Hi (setdown - 15%).
B. Starting up in Mode 2 with APRM downscales clear:
APRM Hi (setdown - 15%)
APRM Hi (120%)
Mode switch - Shutdown position
RWM.
C'!" Shutting down in Mode 2 with IRM's on range 1 to 2:
Manual Scram push buttons
RWM.
D. Shutting down in Mode 2 with average SRM readings at -5x101 cps :
APRM Hi (setdown - 15%)
OPRM upscale trip
RWM.
KIA Statement:
Conduct of Operations
2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for
technical specifications
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine when entry into Technical Specifications is required .
References: Technical Specifications
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledqe and its meaning to predict the correct outcome.
0610 NRC Exam
(
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the following:
1. Tech Spec applicab ility for the listed systems with the given plant condition.
NOTE: The distractors are all plaus ibe since only one system or function is incorrect in each distractor .
A is incorrect. The RBM is not required until >27% rated power.
S is incorrect. The APRM Hi (120%) is not required until Mode 1
C is correct.
D is incorrect. the OPRMs are not required until Mode 1 and >25% rated power.
(
Control Rod Block Instrumentation
3.3.2.1
Table 3.3.2.1-1 (page 1 of 1)
Control Rod Block Instrumentation
APPLICABLE
MODES OR
FUNCTION OTHER REQUIRED SURVEILLANCE ALLOWABLE
SPECIFIED CHANNELS REQUIREMENTS VALUE
CONDITIONS
1. Rod Block Monitor
a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 (e)
b. Intermediate Power Range - Upscale (b) 2 SR 3.3.2.1.1 (e)
c. High Power Range - Upscale (f),(g) 2 SR 3.3.2.1.1 (e)
d. Inop (g),(h) 2 SR 3.3.2.1.1 NA
e. Downscale (g),(h) 2 SR 3.3.2.1.1 (i)
2. Rod Worth Minimizer 1(c),2(c) SR 3.3.2.1.2 NA
3. Reactor Mode Switch - Shutdown Position (d) 2 SR 3.3.2.1.6 NA
(a) THERMAL POWER ~ 27% and s 62% RTP and MCPR less than the value specified in the COLR.
(b) THERMAL POWER> 62% and s 82% RTP and MCPR less than the value specified in the COLR.
(c) With THERMAL POWER s 10% RTP .
(d) Reactor mode switch in the shutdown position.
(e) Less than or equal to the Allowable Value specified in the COLR.
(f) THERMAL POWER> 82% and < 90% RTP and MCPR less than the value specified in the COLR.
(g) THERMAL POWER ~ 90% RTP and MCPR less than the value specified in the COLR.
(h) THERMAL POWER ~ 27% and < 90% RTP and MCPR less than the value specified in the COLR.
(i) Greater than or equal to the Allowable Value specified in the COLR.
( BFN-UNIT 1 3.3-20 Amendment No. 2-M,262
September 27, 2006
RPS Instrumentation
3.3.1 .1
Table 3.3.1.1-1 (page 1 of 3)
Reactor Protection System Instrumentation
APPLICABLE CONDITIONS
MODES OR REQUIRED REFERENCED
FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE
SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE
CONDITIONS SYSTEM ACTION 0 .1
1. Intermediate Range Monitors
a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 s 120/125
SR 3.3.1.1.3 divisions of full
SR 3.3.1.1.5 scale
SR 3.3.1.1.14
5(a) 3 H SR 3.3.1.1.1 ~ 120/125
SR 3.3 .1.1.4 divisions of full
SR 3.3 .1.1.9 scale
SR 3.3.1 .1.14
b. Inop 2 3 G SR 3.3.1.1.3 NA
SR 3.3.1.1.14
5(a) 3 H SR 3.3.1.1.4 NA
SR 3.3.1.1 .14
2. Average Power Range Monitors
a. Neutron Flux * High , 2 3(b) G SR 3.3.1.1.1 ~ 15% RTP
Setdown SR 3.3.1.1.6
SR 3.3.1 .1.7
SR 3.3.1.1.13
SR 3.3.1.1 .16
b. Flow Biased Simulated 3(b) F SR 3.3.1.1.1 s 0.66 W
Therma l Power - High SR 3.3.1.1.2 + 66% RTP
SR 3.3 .1.1.7 and s 120%
SR 3.3.1.1.13 RTP(c)
SR 3.3.1.1.16
c. Neutron Flux - High 3(b) F SR 3.3 .1.1.1 ~ 120% RTP
SR 3.3 .1.1.2
SR 3.3 .1.1.7
SR 3.3.1.1.13
SR 3.3.1 .1.16
(continued)
(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.
(b) Each APRM channel provides inputs to both trip systems.
(c) [0.66 W + 66% - 0.66 t> W] RTP when reset for single loop operation per LCO 3.4 .1, "Recirculation Loops Operating :
BFN-UNIT 1 3.3-6 Amendment No. 236 , 262, 269
March 06, 2007
RPS Instrumentation
3.3.1 .1
(
Table 3.3.1.1-1 (page 2 of 3)
Reactor Protection System Instrumentation
APPLICABLE CONDITIONS
MODES OR REQUIRED REFERENCED
FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE
SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE
CONDITIONS SYSTEM ACT ION D.1
2. Average Power Range
Monitors (continued)
d. Inop 1,2 G SR 3.3.1.1.16 NA
e. 2-0ut-Of-4 Voter 1,2 G SR 3.3.1.1.1 NA
SR 3.3.1.1.14
SR 3.3.1.1.16
f. OPRM Upscale SR 3.3.1.1.1 NA
SR 3.3.1.1.13
SR 3.3.1.1.16
SR 3.3.1.1.17
3. Reactor Vessel Steam Dome 1,2 2 G SR 3.3.1.1.1 :S 1090 psig
Pressure - High(d) SR 3.3.1.1.8
SR 3.3.1.1.10
SR 3.3.1.1.14
4. Reactor Vessel Water Level - 1,2 2 G SR 3.3.1.1.1 ~ 528 inches
Low, LeveI3(d) SR 3.3.1.1.8 above vessel
SR 3.3.1.1.13 zero
SR 3.3.1.1.14
5. Main Steam Isolation Valve - 8 F SR 3.3.1.1.8 :S 10% closed
Closure SR 3.3.1.1.13
SR 3.3.1.1.14
6. Drywell Pressure - High 1,2 2 G SR 3.3.1.1.8 :s 2.5 psig
SR 3.3.1.1.13
SR 3.3.1.1.14
Water Level - High
a. Resistance Temperature 1,2 2 G SR 3.3.1.1.8 s 50 gallons
Detector SR 3.3.1.1.13
SR 3.3.1.1.14
5(a) 2 H SR 3.3.1.1.8 s 50 gallons
SR 3.3.1.1.13
SR 3.3.1.1.14
(continued)
(a) 'Mth any control rod withdrawn from a core cell containing one or more fuel assemb lies.
(b) Each APRM channe l provides inputs to both trip systems .
(d) During instrument calibrat ions , if the As Found channe l setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found
band as defined by its assodated Surveillance Requirement procedure, then there shall be an initial determ ination to ensure confidence that the channel
can perform as required before retuming the channe l to service in accordance with the Surveillance . If the As Found instrument channel setpoint is not
conservative with respect to the Allowable Value , the channel shall be dedared inoperable.
Prior to retum ing a channel to service , the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the
setpoint; otherw ise, the channel shall be dedared inoperable.
The nominal Trip Setpoint shall be spedfied on design output documentat ion which is incorporated by reference in the Updated Final Safety Analysis
Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a
listing of the setpoint design output documentation shall be spedfied in Chapter 7 of the Updated Final Safety Analysis Report.
BFN-UNIT 1 3.3-7 Amendment No. 269
234, 262, 259, 257,258, 266
March 06, 2007
67 . RO GENERIC 2.1.16 00 IIMEM/T3//B17/G2.1.16///
~ Wh iCh ONE of the following announcements is an INAPPROPRIATE use of the Plant Paging System in
I accordance with OPDP-1, Conduct of Operations ?
(
A. There is a fire in the Unit-2 Shutdown Board Room. I repeat. There is a fire in the Unit-2 Shutdown
Board Room.
B. Operations will be starting the 2 Alpha RHR pump.
C. Shift Manager dial 2391. Shift Manager dial 2391
D." This is a drill. All personnel evacuate the Unit 2 Reactor Building due to high radiation .
KIA Statement:
Conduct of Operations
2.1.16 Ability to operate plant phone, paging system, and two-way radio
KIA Justification: This question satisfies the KiA statement by requiring the candidate to demonstrate
specific knowledge of the use of the Plant Paging System while communicating with plant personnel.
References: OPDP-1
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information .
0610 NRC Exam
REFERENCE PROVID ED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determ ine the following:
1. The requirements associated with making Page Announcements per OPDP-1 .
2. Whether the announcement meets those requirements
A is incorrect. This is plausible because it is an expected announcement during a fire.
B is incorrect. This is plausible since the page is not repeated. However, repeating pages for a normal
operation is not required.
C is incorrect. This is plausible since the page is repeated . However, there may not be a requirement t
repeat the announcement, but it is not an inappropriate action.
D is correct. The line "This is a drill" is required at the beginning and END of each commun ication during
drills or exercises. In addition, an announcement of such urgency should be repeated .
TVAN Standard Conduct of Operations OPDP-1
Department Rev.OOOS
Procedure Page 55 of 103
Appendix I
(Page 3 of 5)
Communications
b. Use equipment noun names and/or identificat ion (10) numbers to describe a
component.
,
c. The use of sign language is undesired but maybe used when verbal
communications is not practical.
d. Take time when reporting abnormal conditions . Speak deliberately, distinctly
and calmly . Identify yourself and watch station or your location. Describe the
nature and severity of the problem. State the location of the problem if
appropriate. Keep the communication line open if possible or until directed
otherwise.
e. The completion of directed actions should be reported to the governing station,
normally the control room.
f. Require other plant personnel (including contractors) conducting operational
communication to do so in accordance with this procedure.
g. If there is any doubt concerning any portion of the communication or task
assigned, resolve it before taking any action.
h. When making announcements for drills or exercises begin and end the
announcement with "This is a Drill."
4. Emergency Communications Systems
When personnel are working in areas where the public address (PA) system or emergency
signals cannot be heard, alternate methods for alerting these persons should be devised.
Flashing lights, personal pagers that vibrate and can be felt, and persons dedicated to
notifications are examples of alternate methods.
5. PA System
a. Use of the plant PA system shall be limited to ensure it retains its effectiveness
in contacting plant personnel. Excessive use of the PA system should be
avoided. Plant telephones and other point-to-point communications channels
should be used in lieu of the PA system whenever practical.
b. The announcement of planned starting or stopping large equipment should be
made to alert personnel working in that area.
c. The plant PA system may be used in abnormal or emergency conditions, to
announce change of plant status, or give notification of major plant events either
in progress or anticipated.
TVAN Standard Conduct of Operations OPDP-1
Department Rev.OOOa
( Procedure Page 56 of 103
Appendix I
(Page 4 of 5)
Communications
d. When using the plant PA system:
(1) Speak slowly and deliberately in a normal tone of voice.
(2) When announcements of abnormal or emergency conditions are made,
they shall be made at least twice.
(3) When making announcements for drills or exercises begin and end the
announcement with "This is a Drill."
6. Plant Telephones
When using Plant telephones:
a. Identify yourself and watch station.
b. When trying to make contact with the main Control Room, if the message is of a
routine nature, the sender should hang up when the main Control Room fails to
answer after the fifth ring to avoid unnecessary Control Room noise. The
phone shall be allowed to ring until answered if the information is important to
Operations.
c. During times when the DO NOT DISTRUB (DND) function has been used by
MCR personnel, follow the directions on the recording as appropriate.
d. When making announcements for drills or exercises begin and end the
announcement with "This is a Drill."
7. Radio/phone Communication
Radio/phone usage shall not be allowed in areas where electronic interference with
plant equipment may result.
a. When making announcements for drills or exercises, begin and end the
announcement with "This is a Drill."
b. Sender should identify themselves by watch station.
c. Three way communications should be used.
d. Clear concise language should be used since radio/phone contact does not
have the advantage of face to face communication.
68 . RO GENERIC 2.1.18 OOl/MEM/T3/12.11/GENERIC 2.1.18//RO/SRO/l l/27/07 RMS
Which ONE of the following is an INEFFECTIVE use of the phonetic alphabet in accordance with
OPDP-1 , Conduct of Operations?
A. Place Gulf IRM in Bypass per 1-01-92-Bravo.
B. Start 2-Alpha RHR pump per 3-01-74.
C." Place Romeo-Papa-Sierra 2-Alpha on Alternate per 2-0scar-lndia-99.
D. Transfer 2-Alpha 480 volt shutdown board to Alternate.
KIA Statement:
Conduct of Operations
2.1.18 Ability to make accurate, clear and concise logs, records, status boards , and reports
KIA Justification: This question satisfies the KIA statement by requiring the candidate to demonstrate
knowledge of the requirements related to verbal communications or reports during shift operations.
References: OPDP-1
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the requirements for use of the
phonetic alphabet and apply that knowledge to the given communications.
NOTE: Each distractor is plausible because they all contain at least one use of th phonetic alphabet.
A is incorrect. This communication is appropriate.
B is incorrect. This communication is appropriate .
C is correct. The use of the phonetic alphabet for common acronyms, such as RPS, is not required and
could reduce the effectiveness of the communication .
D is incorrect. This commun ication is appropriate.
TVAN Standard Conduct of Operations OPDP-1
Department Rev. 0008
Procedure Page 54 of 103
Appendix I
(Page 2 of 5)
Communications
b. The receiver repeats back the message to the sender. The repeat back can be
verbatim or functional. In many cases a functional repeat back best
communicates the receivers understanding of the message. This can be done in
several ways to accomplish the desired goals. For example the sender might
say, "Bob, report RCS pressure and trend." The receiver could respond in either
of two ways.
(1) The receiver could respond with, "Report RCS pressure and trend. RCS
pressure is 2250 psig and stable."
Or
(2) The receiver could respond with, "RCS pressure is 2250 psig and stable."
c. The sender verbally acknowledges that the receiver correctly understood the
message. The verbal acknowledgement can be simple such as, That is
correct". If the sender has requested and received information then the sender
shall provide either verbatim or functional repeat back to demonstrate his
understanding of the receiver's message. For the example above the sender
could respond with, "I understand 2250 and stable."
2. Phonetic Alphabet
The phonetic alphabet is a tool to improve communications. In general, operations
communication should use the phonetic alphabet except when well established acronyms
describe the subject. If use of phonetic alphabet will reduce effectiveness of
communications then it should not be used. The following are examples of when the
phonetic alphabet should not be used:
a. It is not desirable to use Romeo-Charlie-Sierra to describe the RCS (Reactor
Coolant System).
b. If a procedural step is written using acronyms, it may be read and ordered as
such.
c. If a component tag or label is written using acronyms then the acronyms may be
used.
3. General Standards
a. All communications shall be clear, concise, and precise. All operational
communications shall be conducted in a formal and professional manner. In all
/ communications, the sender and intended receiver should be readily identifiable.
69. RO GENERIC 2.2.13 00I/MEM/T3/10.2/7/18/GENERIC 2.2.13/3.6/3 .8/RO/SR0/11/26/07 RMS
Which ONE of the following describes the requirements when placing a clearance on air operated
valves?
A. An air operated valve that fails closed on loss of air SHALL NOT be considered closed for blocking
purposes unless it is held closed with a gagging device.
B. An air operated valve that fails open on loss of air SHALL NOT be used for blocking purposes.
C.'; An air operated valve that fails open on loss of air, will be held closed with a gagging device that is
tagged as a clearance boundary.
D. An air operated valve that fails 'as-is' on loss of air SHALL NOT be used for blocking purposes until
it is verified closed and a gagging device installed .
KIA Statement:
Equipment Control
2.2.13 Knowledge of tagging and clearance procedures
KIA Justification: This question satisfies the KIA statement by requiring the candidate to demonstrate
knowledge of the Clearance and Tagging requirements.
References: spp 10.2
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the requirements for Clearance
and tagging procedure. SPP 10-2 and apply that knowledge to the given conditions.
A is incorrect. This is plausible since a locking device would ensure the valve does not open , however
SPP 10-2 requires the air supply to actuate the valve be mechanically or electrically isolated.
B is incorrect. This is plausible since using a "Fail-Open" valve presents a difficult problem. however
SPP 10-2 provides specific guidance to allow their use as a clearance boundry .
C is correct.
D is incorrect. This is plausible because in most cases it is true. However, SPP 10-2 provides specific
guidance and controls to allow using them as a clearance boundry under condition that the clearance be
considered "working on energized equipment".
NPG Standard Clearance Procedure to SPP-10.2
Programs and Safely Control Energy Rev. 0010
( Processes Page 50 of 66
Appendix E
(Page 1 of 2)
Special Requirements for Mechanical Clearances
1.0 REQUIREMENTS
A. An air-operated valve that fails open on a loss of air is not be considered closed for
. blocking purposes unless it is held closed with an installed jacking device or device
used to secure the valve in the required position. A clearance tag will be issued and
attached to the jacking or other device.
B. An air-operated valve that fails closed must have its air supply electrically or
mechanically isolated, depressurized, and the valve visually checked-to-be-c1osed by
local or remote indication. The air supply energy-isolating devices must be tagged.
C. An air-operated valve that fails "as is" shall be closed and mechanically restrained. Its
air supply should be electrically or mechanically isolated, depressurized, and the valve
visually checked to be closed by local or remote indication. The air supply energy-
isolating devices and mechanical restraint must be tagged.
D. In cases where it is not possible to physically secure an air operated valve that fails
"as-is" in the closed position, the valve will be tagged closed by applying closing air to
the valve diaphragm by the use of the solenoid valve air overrides and tagging both the
hand-switch in the closed position and the solenoid valve air overrides. Prior to
allowing work to begin, the equipment will be drained and de-pressurized to ensure the
boundary valves are holding. This condition will be noted in the remarks section of the
clearance sheet to inform PAE/Authorized Employee(s) that pressurized air is required
to ensure the valve remains closed. This work is considered "working on energized
equipment" and must be approved by the management official in charge .
E. Pressure controlled valves, relief valves, and check valves will not be used as isolation
boundary valves under normal conditions. Where such a valve does not have an
external means of physical restraint, the work is considered "working on energized
equipment" and must be approved by the management official in charge .
F. The following instructions govern the use of freeze plugs
1. The clearance should be in place, but not issued, before establishing the freeze
plug.
2. The need for the freeze plug should be identified on the Remarks Section of the
clearance sheet. The freeze plug should not be listed as a device held on the
clearance sheet. The establishment and maintenance of the freeze plug shall be
in accordance with approved procedures or work documents.
3. The freeze plug must be attended by qualified personnel to ensure that it is
maintained intact until all work is complete and the proper Post Maintenance
Tests (PMTs) are performed.
4. If the clearance must be released to allow performance of a PMT, the equipment
( must be retagged before allowing the freeze plug to thaw. This will prevent
migration of a portion of the plug.
70. RO GENERIC 2.2.33 00l/CIA/SYS/RWM//G2.2.33/RO 2.5//10122/07
Given the following plant conditions:
- A reactor startup is in progress
( -
-
Reactor Power: 3%
RWM latched into Group 8 (12 control rods)
- Group 9 rods are the same rods as Group 8.
- Sequence Control: ON
- Group 8 Limits: 08-12
- Group 9 Limits : 12-16
Which ONE of following describes when the RWM will automatically latch up to Group 9?
A...; all rods in group 8 have been withdrawn to the group 8 withdraw limit and a rod in group 9 has been
selected .
B. all rods EXCEPT 3 in group 8 are withdrawn to the group withdraw limit, and a rod in group 9 is
selected.
C. all rods EXCEPT 1 in group 8 are withdrawn to the group withdraw limit and a rod in group 9 has to
be selected and moved.
D. the last rod in group 8 is withdrawn to the group 8 withdraw limit and the in-sequence rod in group 9
has NOT been selected .
KIA Statement:
Equipment Control
2.2.33 Knowledge of control rod programming.
KIA Justification: This question satisfies the KIA statement by requiring the candidate to recognize and
apply limitations on control rod programming enforced by the Rod Worth Minimizer program.
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble ,
sort, and integrate the parts of the question to solve a problem. This requires mentally using this
knowledge and its meaning to resolve the problem.
0610 NRC Exam
REFERENCE: Lesson Plan OPL 171.024 Rev. 13 pages 13 - 15
Plausibility Analysis:
( Answer A is the correct answer.
Answer B is incorrect. This is plausible because the RWM normally allows three insert errors without
generating a rod block, however it will not latch up to a higher group under this condition because the
three rods are more than one notch from the withdraw limit.
Answer C is incorrect. The selected control rod does not have to be moved to latch to Group 9. This is
plausible because the RWM will latch to the highest group with one rod past the insert limit if the RWM is
latching to a group from an unknown condition . Since this is a known condition, the RWM will latch to
Group 9 without moving the selected rod.
Answer D is incorrect. The RWM will not latch to the next group until the correct rod is selected in Group
9 because Sequence Control is ON. This is plausible if Sequence Control is OFF . Under that condition,
the RWM only looks for rods within the Group and not within a specific sequence. With Sequence Control
in OFF, the RWM will latch up to Group 9 as soon as the last rod reaches the withdraw limit.
OPL171.024
Revision 13
Page 13 of 53
( INSTRUCTOF1 NOTES
(8) Upon demand by the operator via
the Scan/Latch request function.
(9) Following correction of Insert or
Withdraw Errors. .
d. The latched group is the highest group
which can be achieved without producing
an active insert block condition .
(1) The RWM system will latch to the
highest group in the sequence with :
(a) At least one rod withdrawn
past the group insert limit and
(b) No other groups below have
three insert errors
(2) Example: Relatch at an intermediate
power level
(a) Assume that RWM has been
out of service and rods have
been moved out of sequence.
The following rod distribution
exists:
(1) All rods in Group 1 thru These 3 rods would
7 are at their withdraw cause an insert
limit, except rods 30- block if GP 8 were
35, 38-43 and 38-27 latched.
(GP. 7) which are at
position 02.
(2) All rods in Groups 8
and above are at their
insert limit (04) except
which is at position 06.
(3) No rod is selected
OPL 171.024
Revision 13
Page 14 of 53
( INSTRUCTOR NOTES
(b) After returning the RWM to
service:
(1) Group 7 will be the
latched group
(2) Rod 30-03 will be
displayed as a withdraw
error.
(3) The withdraw block status
indicators will indicate a
withdrawal block condition
on the RWM system
displays and RWM switch
panel.
(4) No other control rod may
be inserted or withdrawn
until the withdraw error
rod from Group 8 (30-03)
is corrected . It can only
be inserted.
(c) The proper way to correct the NOTE: Upon select
out of sequence condition is to of rod 30-03, an
insert the withdraw error rod (30- RWM system
03) to position "04". message will be
generated indicating
This removes the withdraw error;
a target position of
leaves group 7 as the latched
notch "04" for this
group, and removes the
withdraw block indications on the
switch panel.
11. Automatic Latching Up/Down Obj . V.B.10
a. The automatic latching process depends on
whether or not RWM Sequence Control is ON
or OFF . Sequence Control is normally
selected (ON) and enforces a specific order to
pull rods within a latched group.
b. When operating below the LPSP with NOTE: Latching
sequence controlOre", latching to the next within Transistion
higher or next lower rod group is done Zone will be
internally by the RWM program only after a discussed later.
rod in the next group is selected.
OPL 171.024
Revision 13
Page 15 of 53
INSTRUCTOR NOTES
(1) The program will latch down (latch the NOTE: Will latch
next lower group) when all the rods in down if insert errors
the presently latched group have been in GP is lower than
inserted to the group insert limit and a latch GP.
rod in the next group is selected.
(2) The program will latch up (latch the Will latch up
next higher group) upon selection of a provided that the
rod within the next higher group number of insert
provided that only 2 insert errors or less errors produced will
result from within the current latched not give an insert
group and/or any lower groups. block.
c. When sequence control is NOT selected,
(OFF) , latching automatically occurs based on
rod movement within repeating BPWS banked
groups (ex: 2/3/4/5/6 and 7/8/9/10/11/12).
(1) For example, if the rods in a group (GP.
4) are the same rods as in the next
. higher group (GP. 5), the RWM will
NOT latch up based solely upon control
rod selection. Latch up to Group 5 will
automatically occur when any of the
rods Group 4 are moved to a position
defined for Group 5 provided that <3
insert errors would result.
(2) If the rods in a group (GP. 5) are the With rods at both a
same rods as the next lower group GP 4 and GP 5
(GP. 4), the RWM will not latch down defined position, the
based solely upon control rod selection. latched GP after a
Latch down to the next lower RWM movement will be
group will generally occur in this case the GP moved into .
based upon movement of any of the
rods within the group to a position
defined for the next lower RWM group.
(3) If the next rod group is NOT repeating ,
then latching occurs when the next rod
is selected.
(
71. RO GENERIC 2.3.10 OOl/C/A/GFES/GENERIC/C/A/G2.3.l0/BF0030l/2.9/3.3/GEN 2.3
Given the following conditions at a work site.
Airborne activity: 3DAC
( Radiation level: 40 mr/hr
Radiation level with shielding: 10 mr/hr
Time to place shielding : 15 minutes
Time to conduct task with respirator: 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
Time to conduct task without respirator: 30 minutes
Assume the following :
- the airborne dose with a respirator will be zero.
- a dose rate of 40 mr/hr will be received while placing the shielding .
- all tasks will be performed by one worker.
- shielding can be placed in 15 minutes with or without a respirator.
Which ONE of the following would result in the lowest whole body dose?
A. Place the shielding while wearing a respirator and conduct the task with a respirator.
B. ~ Place the shielding while wearing a respirator and conduct the task without a respirator.
C. Conduct the task with a respirator and without shielding.
D. Conduct the task without a respirator or shielding .
KIA Statement:
Radiation Control
2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel
exposure.
KIA Justification: This question satisfies the KIA statement by requiring the candidate to calculate the
expected exposure for a job and determine the correct precautions and radiological controls required to
minimize exposure.
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. Th is requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
Plausibility Analysis:
This question requires the candidate to calculate the exposure recevied for each of the four options in the
distractors. Although this question does not specifically contain incorrect but plausible possibilities, it is
( based entirely on the type of decision which must be made while performing duties as a Licensed
Operator. Using the calculation below, the candidate must correctly perform the analysis and apply
ALARA principles to select the correct answer.
Calculations required:
3 DAC x 2.5 mr/DAC X 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> =3.75 mr
a. 10 mr placing shielding, 10 mr conducting task, zero airborne = 20 mr
b. 10 mr placing shielding, 5 mr conducting task, 3.75 mr airborne =18.75 mr (lowest dose = Correct)
c. 40 mr conducting task, zero airborne =40 mr
d. 20 mr conducting task, 3.75 mr airborne = 23.75 mr
(
72. RO GENERIC 2.3.9 OOl/C/A/T3/PR.CMPTR//RO GENERIC 2.3.9//RO/SRO/ll/27/07 RMS
Unit 2 reactor shutdown is in progress and primary containment de-inerting has been authorized .
Which ONE of the following is the basis for NOT allowing both 2-FCV-64-19 (SUPPR CHBR ATM SPLY
INBD ISOLATION VLV) and 2-FCV-64-18 (DRYWELL ATM SUPPLY INBD ISOLATION VLV) to be open
simultaneously during the performance of this evolution?
A. To prevent the high flow rate from damaging the non-hardened ventilation ducts .
B. To prevent creating a high dP between the primary containment and the Reactor Building.
C." To prevent the possibility of overpressurizing the primary containment during a LOCA.
D. To prevent release of the drywell atmosphere through an unmonitored ventilation flow path.
KIA Statement:
Radiation Control
2.3.9 Knowledge of the process for performing a containment purge .
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions and times to correctly determine the process for performing a containment purge.
References : 2-01-64, Rev.106, section 8.1
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome .
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibilitv Analysis:
In order to answer this question correctly the candidate must determine the requirements for de-inerting
the Primary Containment and their bases .
A is incorrect. This is plausible becaue high flowrates would result from both valves being open,
however the vent ducts are designed to accomodate such flowrates.
B is incorrect. This is plausible because the de-inerting lineup raises the dP between the Drywell and
Reactor Building, however the rise is relatively insignificant and well within the design limits .
C is correct.
D is incorrect. This is plausible since the vent path is unmonitored, however, having both valves open
simultaneously provides no additional path for a release.
(
BFN Primary Containment System 2-01-64
Unit2 Rev. 0106
( Page 40 of 194
8.0 INFREQUENT OPERATIONS
8.1 Purging the Drywell and Suppression Chamber with Primary
Containment Purge Filter Fan
NOTES
1) TOE 970823 identified a potential for a bypass flow path to exist between the Drywell
and Suppression Chamber when purging the Drywell and Suppression Chamber at the
same time (both FCV-64-18 and 64-19 opened concurrently). Should a design basis
LOCA occur with these two valves opened at the same time with the Reactor NOT in
Cold Shutdown (Mode 4 or 5), a potential exists for overpressurizing primary
containment due to the pressure suppression function being bypassed. Therefore,
when Primary Containment purging is required with the Reactor NOT in Cold
Shutdown (Mode 4 or 5), the Suppression Chamber and the Drywell are purged
separately.
2) This section is used when purging both the Drywell and Suppression Chamber
concurrently with the Reactor in Cold Shutdown (Mode 4 or 5).
3) When the Reactor is NOT in Cold Shutdown (Mode 4 or 5), the Suppression Chamber
and the Drywell are purged separately.
[1] REVIEW all Precautions and Limitations in Section 3.0. o
[2] VERIFY all Prestartup/Standby Readiness requirements in
Section 4.0 are satisfied. o
[3] VERIFY the following initial conditions are satisfied:
- Drywell vented to less than 0.25 psig.
REFER TO Section 6.1. 0
- HzOz analyzers are in service REFER TO 2-01-76 0
- Suppression Chamber vented to less than 0.25 psig.
REFER TO Section 6.2. 0
- Reactor Zone Fans in operation with Reactor Zone Supply
and Exhaust Fan in fast speed. REFER TO 2-01-308. 0
[4] REQUEST Chemistry to obtain a Drywell sample. REFER TO
2-SI-4.8.8.2-6. o
[5] IF sample is within limits of 2-SI-4.8.8.2-6, THEN
(
NOTIFY Shift Manager. o
73. RO GENERIC 2.4.47 OOl/C/A/T3/C4/6/G2.4.471IRO/SRO/IO/25107 RMS
Given the following plant conditions:
- Reactor pressure is being maintained at 50 psig.
- Temperature near the water level instrument run in the drywell is 220°F .
- The Shutdown Vessel Flooding Range Instrument ( L1-3-55) is reading +35".
Which ONE of the following describes the highest Drywell Run Temperature at which the L1-3-55 reading
(+35") is considered valid?
REFERENCE PROVIDED
KIA Statement:
Emergency Procedures IPlan
2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate
control room reference material
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the correct reactor water level under emergency conditions .
References: 2-EOI-3 Flowchart
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome . This requires mentally using this
knowledge and its meaning to predict the correct outcome .
0610 NRC Exam
REFERENCE PROVIDED: 2-EOI-1 flowchart
Plausibility Analysis:
In order to answer this question correctly the candidate must use EOI Caution #1 to determine operable
RPV water level instruments.
A is incorrect. This is plausible since 200°F is a valid indication, however the question calls for the
HIGHEST temperature.
B is correct.
C is incorrect. This is plausible if the candidate interpolates the Caution #1 table, however this is not
permissible .
D is incorrect. This is plausible if the candidate interpolates the Caution #1 table, however this is not
permissible.
~
THE MINIMUM INDICATED LVLASSOCIATED WITHTHEHIGHeST MAXOWOR SCRUNTEMP.
- IFOWTEMPS. OR SCAREATEMPS (TABLE 6). ASAPPLICABLE, AREOUTSIDE THESAFE REGION OFCURVE 8.
THEASSOCIATED INSTRUMENT MA.Y BE UNR8..IABlEDUE TO BOIUNG INTHERUN.
MINIMUM MAX DW RUN TEMP MAXSC
INSTRUMENT I RANGE I INDICATED (FROM XR-64-50 RUN TEMP
LVL OR TI-64-52AB) (FROM TABLE 6
ON SCALE N/A 8ElOW150
-145 N/A 151 TO 200
L1-3-S8A, B I EMERGENCY
-140 N/A 201 TO 250
-155 TO +60
-130 N/A 251 TO 300
-120 N/A 301 TO 350
L1-3-53 ON SCALE N/A 8ELOW 150
L1-3-60 -s N/A 151T0200
NORMAL
L1-3-206 +15 N/A 201 TO 250
OTO +60
L1*3-253 +20 N/A 251 TO 300
L1-3-20aA. 8 , C, D +30 N/A 301 TO 350
POST
L1*3-52
L1-J.62A I ACCIDENT
-268 TO +32
I ON SCALE N/A NfA
+10 8ELOW100 N/A
+15 100 TO 150 N/A
SHUTDOWN I +20 151 T0200 NfA
Ll-3-55 I FLOODUP I +30 201 T0250 N/A
OTO +400 I +40 NfA
251 T0300
+50 301 T0350 N/A
+65 35110400 N/A
E MINATION
REFERENCE
PROVIDED TO
( CANDIDATE
o
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f<I.OCIUN<PI.Nff
74. RO GENERIC 2.4 .15 00I/MEM/T3///GENERIC 2.4.15//RO/SR0/11/27/07 RMS
Given the following plant conditions:
- Unit-2 has scrammed and multiple control rods have failed to insert.
(
- The Unit Supervisor has entered EOI-1, RPV Control , and C-5, Level/Power Control.
- You have been designated to assist the crew by performing EOI Appendicies as they are
assigned .
Which ONE of the following precludes the use of a hand held radio to communicate with Control Room
personnel?
A. EOI Appendix 2 in the 2A Electrical Board Room.
B~ EOI Appendix 1C in the U-2 Aux Instrument Room.
C. EOI Appendix 16H at the 2C 250V RMOV Board.
D. EOI Appendix 1B in the Reactor Building 565 elevation.
KIA Statement:
Emergency Procedures IPlan
2.4.15 Knowledge of communications procedures associated with EOP implementation
KIA Justification: This question satisfies the KIA statement by requiring the candidate to demonstrate
knowledge of communication requirements that apply during execution of Emergency Operating
Instructions .
References: OPDP-1
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine which of the given locations
violates the requirements of OPDP-1, Conduct of Operation.
A is incorrect. This is plausible bacuse of the safety related equiment powered from 2A Electric Board
Room, however radio communication is authorized.
B is correct.
C is incorrect. This is plausible because of the safety related equipment fed from 2C 250V RMOV
Board, however radio communication is authorized.
D is incorrect. This is plausible because of the proximity of the RPV level instrumentation, however radio
( communication is authorized.
EOI PROGRAM MANUAL USER'S GUIDE FOR EMERGENCY OPERATING INSTRUCTIONS
SECTION O-VIII-A _
( L. Exiting the EOIs
The operators remain in the EOIs until either directed out by the EOI or when the
SMIUS concludes that an emergency condition no longer exists. Exit from EOI-l and
associated contingency procedures always requires SMIUS determination, since these
procedures have no explicit exit to other plant procedures except from RC/Q to AOI-
100-1. Appendix 100-1 should be reviewed prior to EOI exit to determine , restore, and
document abnormal alterations that were established during EOI execution.
After exiting the EOIs the operator surveys the present plant conditions to ensure no
reason for re-entry to the EOIs exist.
During EOI execution , a SAMG ENTRY IS REQUIRED condition may arise. Entry
into and execution of Severe Accident Management Guidelines (SAMGs) are the
responsibility of the SED in the TSC. Significant time may be required to man the TSC
with the appropriate SAM Team members and tum over plant conditions between the
control room and the TSC. The control room staffterminate execution of ALL EOI
flowcharts ONLY when the SED declares that the SAM Team has assumed command
and control. EOI appendices may continue in use as directed by the SAMGs.
During the time between the development ofthe SAMG ENTRY IS REQUIRED
condition and the time of assumption of command and control by the TSC, the control
room staff shall continue use of available EOI guidance to mitigate the event.
Development ofa SAMG ENTRY IS REQUIRED condition always requires entry into
the SAMGs when the TSC SAM Team assumes command and control, even ifplant
conditions subsequently develop which seem to no longer satisfy a requirement to enter
3.5 Duties of the Control Room Team Members While Executing EOIs
The specific duties ofthe Control Room Team Members are outlined in Conduct of Operations.
3.6 Shift Communications During Execution ofEOIs
The methodology associated with communications during execution of the EOIs is outlined in
Conduct of Operations
3.7 Use ofInstrumentation and ICS/Safety Parameter Display System (SPDS)
Various instruments in the control room are qualified for Post Accident Monitoring . These
instruments are identified with black labels. During the performance of the EOIs, these
instruments are required to be utilized as much as practical. For parameters that have multiple
readouts in the control room, the operator should observe as many of the multiple readouts as
practical for a verification ofthe values being observed.
Most instruments in the control room are provided with what may be considered standard scale
divisions (increments of 1,5, 10, etc.), although there are some that may be considered off-
normal (increments of2, 3, 4, etc.), Some pressure instruments may read out in PSIA rather than
the more common value ofPSIG.
The operator is required to remain aware of these possible differences when reading the values
from the instruments. For pressure instruments, the pressure should be called out in values of
PSIA or PSIG, as applicable. When the operator reading the flowchart asks for the value of a
pressure parameter, it should be assumed that the value be given as PSIG unless he/she solicits
( the value in PSIA.
SECTION O-VIlI-A PAGE 48 OF 52 REVISION 4
TVAN Standard Conduct of Operations OPDP-1
Department Rev.OOOa
( Procedure Page 56 of 103
Appendix I
(Page 4 of 5)
Communications
d. When using the plant PA system:
(1) Speak slowly and deliberately in a normal tone of voice.
(2) When announcements of abnormal or emergency conditions are made,
they shall be made at least twice.
(3) When making announcements for drills or exercises begin and end the
announcement with "This is a Drill."
6. Plant Telephones
When using Plant telephones:
a. Identify yourself and watch station.
b. When trying to make contact with the main Control Room, if the message is of a
routine nature, the sender should hang up when the main Control Room fails to
answer after the fifth ring to avoid unnecessary Control Room noise. The
phone shall be allowed to ring until answered if the information is important to
Operations.
c. During times when the DO NOT DISTRUB (DND) function has been used by
MCR personnel, follow the directions on the recording as appropriate.
d. When making announcements for drills or exercises begin and end the
announcement with "This is a Drill."
7. Radio/phone Communication
Radio/phone usage shall not be allowed in areas where electronic interference with
plant equipment may result.
a. When making announcements for drills or exercises, begin and end the
announcement with "This is a Drill."
b. Sender should identify themselves by watch station.
c. Three way communications should be used.
d. Clear concise language should be used since radio/phone contact does not
have the advantage of face to face communication.
75. RO GENERIC 2.4.8 OOIIMEM/T3///GENERIC 2.4.8//RO/SRO/ll/27/07 RMS
Which ONE of the following describes the use of Event Based procedures during Symptom Based
Emergency Operating Instructions (EOI) execut ion?
( Event Based procedures are _
A. NOT used during Symptom Based EOI execution .
B. ALWAYS used if equipment or plant status require their implementation.
C~ used ONLY if they do not interfere with EOI implementation.
D. used ONLY if specifically directed by an EOI flowchart step.
KIA Statement:
Emergency Procedures /Plan
2.4.8 Knowledge of how the event-based emergency/abnormal operating procedures are used in
conjunction with the symptom-based EOPs
KIA Justification: This question satisfies the KIA statement by requiring the candidate to demonstrate
knowledge of procedure hiearchy during execution of Emergency Operating Instructions.
References: EOIPM Section O-VI II-A
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must deterine the rules for using Event Based
procedures during EOI execution.
A is incorrect. This is plaus ible based on the contradiction often found between Event based and
Symptom based guidance. However, their use is permitted under controls circumstances.
8 is incorrect. This is plaus ible because no specific Event Based procedure is expressly prohibited from
use, however if a conflict exists between the Event based procedure and the EO I, the EOI takes
precedence.
C is correct.
D is incorrect. This is plausible because several EOI steps direct actions in accordance with Event
Based procedures, however it is not a prerequisite to their use.
EOI PROGRAM MANUAL USER'S GUIDE FOR EMERGENCY OPERATING INSTRUCTIONS
SECTION O-VIII-A
I. EOI Flowchart Use With Other Plant Procedures
The EOls are entered, based upon specific conditions symptomatic of emergencies, or
conditions that could degrade into emergencies. Therefore the operator actions, provided
within the EOls, allow the operator to mitigate the consequences of a broad range of
accidents and multiple equipment failures.
Other procedures, such as AOIs, ARPs, EPIPs, etc., have event specific entry conditions
and may be used to supplement EOls . In some instances the EOIs will direct the
operators to the unit operating procedures (Ols, OOIs, and AOIs) for completion of
specific tasks. Usually, the EOIs direct the operators to specific EOI Appendices . The
Appendices are specific task related procedures written to satisfy directives given within
the EOIs.
Actions that contradict any direction given by the EOls, or reduce the effectiveness of
any directions given by the EOIs, WILL NOT be implemented for any reason.
The exception to this rule are the SSls and AOI-IOO-2. The conditions which cause
entry into the SSls are such that the reliability of the information systems required to
execute the EOIs are no longer at a confidence level that would make the EOIs effective.
Any time that the operators must leave the control room, as directed by AOI-lOO-2, the
EOIs shall be exited and AOI-lOO-2 shall be used to shut down and cool down the
reactor. The EOls are not designed, or written, to support their use outside of the main
control room.
Conditions may arise under Station Blackout (SBO) conditions in which the rate ofRPV
cooldown is reduced, or alternate Heat Capacity Temperature Limit or Pressure
Suppression Pressure curves are appropriate to avoid an unnecessary emergency
depressurization, in order to maintain RCIC injection capability. The TSC staff or an
associated abnormal operating instruction may recommend use ofthese alternate curves,
which have been calculated as part ofEOIPM section 2- or 3-VI-F and -H. These
alternate curves meet the assumptions used within the EOls.
It is recognized that during execution ofthe EOIs the control room will receive
assistance from various support groups. This is especially the case under conditions in
the EPIPs that result in the Technical Support Center (TSC) being staffed . For example,
the TSC may make recommendations regarding when it is best to vent primary
containment, based upon present or predicted meteorological conditions . This would not
contradict the directions provided by the EOIs, but help to meet the intent of minimizing
radiological releases to the general public.
J. Execution ofEOI Appendixes
The EOIs rely heavily upon the EOI Appendices to implement EPO and PSTO actions
and tasks that are too involved to outline on the flowchart procedure. These tasks
include the defeating of various interlocks and logic systems. The steps within the
Appendices involve the removing of fuses, placing jumpers across terminals , and
placing boots on relay contacts, as well as some of the more common functions such as
opening and closing valves and operation of systems to support the EOI flowchart
procedure steps .
(
SECTION O-VIII-A PAGE 46 OF 52 REVISION 4