ML14154A373
| ML14154A373 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/02/2014 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| Shared Package | |
| ML14155A059 | List: |
| References | |
| 50-259/OL-14, 50-260/OL-14, 50-296/OL-14 | |
| Download: ML14154A373 (171) | |
Text
QUESTION 76 With Unit 2 operating at 100% Reactor Power, a Normal Supply Breaker has tripped open AND suffered damage due to arcing. In addition to RHR AND Core Spray logic alarms, the following significant alarms result:
HPCI LOGIC POWER FAILURE, (2-9-8A, Window 3)
HPCI 120 VAC POWER FAILURE, (2-9-8A, Window 7)
ADS BLOWDOWN POWER FAILURE, (2-9-3 C, Window 32)
Which ONE of the following completes both statements below?
The 250 VDC RMOV Board (1) has been lost.
After manually transferring the 250 VDC RMOV Board to the Alternate Source, the Board is considered (2) in accordance with Tech Spec 3.8.7, Distribution Systems Operating.
A. (1)2A (2) inoperable B. (1)2A (2) OPERABLE C. (l)2B (2) inoperable D. (1)2B (2) OPERABLE ANSWER: A
Level:
1 Group#
1 Examination Outline Cross-Reference KIA#
295004 G2. 1.7 Importance Rating I
295004 Partial or Complete Loss of D.C. Power 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Explanation: A CORRECT: All of the alarms mentioned above, with the exception of one, occur for either board power loss. The reason that the RHR and Core Spray alarms were written generically is because one will be Loop I and the other would be Loop II. The HPCI 120 VAC failure is the only unique alarm, as they are listed here. HPCI is a DIV II System, but has an ALPHA Board Power Supply this is counterintuitive and often confused. The ARP will have the operators manually transfer the board to its alternate power supply. The Tech Spec Bases for 3.8.7 discusses the fact that the board is considered inoperable, even with power restored; because of single failure considerations.
B-Incorrect. First part correct as detailed in A above. Second part is incorrect per the Tech Spec Bases for 3.8.7 (as discussed above) the board is considered inoperable due to single failure considerations C-
. Incorrect. First part incorrect
- The presence of the HPCI 120 VAC POWER FAILURE alarm is the designator that this is the Alpha versus the Bravo board. Second part is correct (for either board).
D-Incorrect.. First part incorrect
- The presence ofthe HPCI 120 VAC POWER FAILURE alarm is the designator that this is the Alpha versus the Bravo board. Second part is incorrect per the Tech Spec Bases for 3.8.7 (as discussed above) the board is considered inoperable due to single failure considerations.
Technical Reference(s): 2-ARP-9-8C window 4 and 11, 2-ARP-9-3C window 32, 2-ARP-9-3F window 3and 7, U2 Tech Spec 3.8.7 page 3.8-87a Proposed references to be provided to applicants during examination: None Learning Objective (As available):
Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: BFN 1006 #76 Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis :
X 10 CFR Part 55 Content:
55.43 (2) Facility operating limitations in the technical specifications and their bases.
BFN Panel 9-8 2-ARP-9-8C Unit 2 2-XA-55-8C Rev. 0015 Page 7 of 46 SensorJTri Point:
250V REACTOR 72N-BA Normal supply overcurrent.
MOVBD 2A 72E..BA Alternate supply overcurrent.
2-EA-57-94 27EX Normal supply undervoltage.
27B MOV bd undervoltage (7sec TDDO) r (Page 1 ofl)
Sensor 250V RMOV Bd 2A, El 621, R-14 Q-LINE, Shutdown Bd Rm C Location:
Probable A. Loss of normal supply (250V Battery Bd 2, Pnl 3, Bkr 302).
Cause:
B. Overcurrent on normal or alternate supply to the board.
C. Fuse failure.
D. Sensor malfunction.
Automatic None.
Action:
Operator A. VERIFY conditions of alarm:
Action:
Loss of HPCI indicating lights on Panel 2-9-3.
D Loss of backup scram valve lights on Panel 2-9-5.
D B.
DISPATCH Personnel to MOV board to check for abnormal conditions: undervoltage, breaker tripped, etc.
D NOTE wc If the mechanism resets (hear click and feel resistance when pushing in), this indicates that an overcurrent condition tripped the breaker.
C.
IF Normal or Alternate feeder breaker tripped, THEN Manually DEPRESS mechanical trip/reset mechanism on breaker face to reset Bell Alarm lockout device.[NER.c n-B-92-069J C
D. VERIFY Bkr 302 closed at Battery Bd Room 2, Panel 3, El 593.
C E.
REFER TO TS Section 3.8.7.
C F.
REFER TO 0-Ol-57D to re-energize or transfer the board.
C G. REFER TO appropriate 01 for recovery or realignment of equipment.
C
References:
45N6201 I 245E7121 45N7147 TS Section 3.8.7.
BFN Panel 9-8 2-ARP9-8C Unit 2 2-XA-55-8C Rev. 0015 Page 15 of 46 Sensor/Trip Point:
250V REACTOR 72N-BA Normal supply overcurrent MOVBD 2B 72EBA Alternate supply overcurrent 2-EA-57-100 27EX Normal supply undervoltage 27B Rx 2B MOV bd underioltage (7sec TDDO)
(Page 1 of 1)
Sensor 250V RMOV Bd 2B, El 593, R-14 Q-LINE. Shutdown Bd Rm B Location:
Probable A.
Loss of normal supply (250V Battery Bd 3, Pnl 3, Bkr 303).
Cause:
B. Overcurrent on normal or alternate supply to the board.
C. Fuse failure.
D. Sensor malfunction.
Automatic None Action:
Operator A. VERIFY alarm by checking:
D Action:
Loss of HPCI and RHR indicating lights (Panel 2-9-3).
LI Loss of backup scram valve lights (Panel 2-9-5).
LI B.
DISPATCH Personnel to MOV board to check for abnormal conditions: undervoltage, breaker tripped, etc.
LI NOTE
[H!Cj If the mechanism resets (hear click and feel resistance when pushing in), this indicates that an overcurrent condition tripped the breaker.
C.
IF Normal or Alternate feeder breaker tripped, THEN Manually depress mechanical trip/reset mechanism on breaker face to reset Bell Alarm lockout device.
[NER!C ll-B-92-069J LI D. VERIFY bkr 303 closed at Battery Bd Room 3, Panel 3, El 593.
LI E.
REFER TO 0-Ol-57D to re-energize or transfer the board.
LI F.
REFER TO TS Section 3.8.7.
LI G. REFER TO appropriate 01 for recovery or realignment of equipment.
References:
45N620i I 2-45E712-2 45N7147 TS Section 3.8.7.
BFN Panel 9-3 2-ARP-9-3C Unit 2 2-XA-55-3C Rev. 0022 Page 39 of 42 Sensor/Trip Point:
Relay 2E-K40 Panel 9-33 De-energized Relay 2E-KIA Panel 9-30 De-energized Relay 2E-KIB Panel 9-33 De-energized Relay 2E-K12 Panel 9-30 De-energized Relay 2E-K32 Panel 2-25-32 De-energized (Page 1 of 1)
Relay 2E-K33 Panel 2-25-32 De-energized Relay 2E-K37 Panel 2-25-32 De-energized Relay 2E-K38 Panel 2-25-32 De-energized Panel 2-25-32 Panel 2-9-30 and 2-9-33 Sensor Backup Control Center Aux Instrument Rm Location:
El 621, R-13 Q-LINE El 593 Probable A. Cleared Fuse(s)
Cause:
B.
Loss of 250V DC power supply to panels.
C. Auto Xfer of Logic Bus B Power Supplies.
Automatic Main steam auto relief valves, PCV-1 -22 and 30, auto transfer power supply from Action:
250V DC Rx Mov Bd A to 250 DC Rx Mov Bd B (PCV-1-22) & Bd C (PCV-1-30) on loss of normal power supply.
Logic Bus B transferred to Alternate supply (250V Rmov Bd 2A) upon Loss of Normal Supply (250V RMOV Sd 2B) or fuse failure.
Operator A. VERIFY power is available to PCV-1-22 and -30.
D Actio,v B.
IF annunciator HPCI LOGIC POWER FAILURE, XA-55-3F
Window 3, is in alarm, this is indicative of loss of power to 250V DC Rx Mov Bd 2N2B. DISPATCH personnel to check 250V DC Rx Mov Sd 2A Breaker I 1A2 and 250V DC Rx MOV Sd 2B Breaker IB1.
C C. DISPATCH personnel(s) to check:
1.
Logic Bus A a.
250V DC Rx Mov Sd 2B, Breaker IFI.
C b.
Fuses 2-FU2-00I-2E-K3 in Panel 9-30.
C 2.
Logic Bus B a.
250V DC Rx Mov Sd 2A, Breaker 9A1.
C b.
Fuses 2-FUI-001-2E1K22A and 2-FUI-001-2E/K22B on Panel 9-33.
C c.
Fuses 2-FU2-00I-2E-K13 in Panel 9-30:
C d.
Fuses 2-FU2-1-2E-KI IA and 2-FU2-I-2E-K11) on Panel 9-33 (GG Block).
D. REFER TO Tech Spec Section 3.5.1.
C E.
REFER TO TRM 3.3.3.4.
C
References:
245N6202 2-45E7121, -2 and -3 GE 730E929 -1, -2 and 3 Technical Specifications 3.5.1 TRM 3.3.3.4.
BFN Panel 9-3 2-ARP-9-3F Unit 2 2-XA-55-3F Rev 0033 Page 6 of 39 Sensor/Trip Point:
Relay 23A-K39 (Bus A)
Loss of 250V DC P0 FAILURE Relay 23A-K44 (Bus B)
Control Power Relay 23A-K44B (Bus B) r (Page 1 ofi)
Sensor Panel 9-32, Bus A Panel 9-39, Bus B Location:
Aux lnstr Rm, El 593 Aux lnstr Rm, El 593 Probable A. Cleared fuse(s).
Cause:
B.
Lass of 250V DC power supply to panels.
Autom1A.
Logic Bus A failure renders Channel A trip and automatic isolation logic mop.
Action HPCI continues to function.
B.
Logic Bus B failure renders Channel B trip, automatic initiation, and automatic isolation logics mop.
If HPCI is in service the HPCI TURBINE STOP VALVE, 2-FCV-73-1 8, closes. HPCI becomes inoperable.
Operator A.
DETERMINE which logic bus has failed, REFER TO automatic Action:
action section.
D B.
DISPATCH personnel to verify source of power failure:
1.
LogicBusA C
a.
Fuses 2-FU2-073-23A-K36 (23A-F1 9) and 2-FU2-073-23A-K36 (23A-F20), Panel 9-32.
0 b.
Power supply 250V DC Rx May Bd 2B, Breaker IBI.
C 2.
Logic Bus B C
a.
Fuses 2-FU2-073-0039A and 2-FU2-073-0039B, Panel 9-39.
C b.
Power supply 250V DC RMOV Bd 2A, Breaker 11D1.
C C. REFER TO Tech Spec 3.5.1, 3.5.2, 3.3.5.1, 3.3.6.1, and TRM 3.3.3.4.
C
References:
2-45E620-1 GE 730E928-2-3 and -4.
Technical Specifications 3.3.5.1, TRM 3.3.3.4 3.3.6.1,3.5.1,3.5.2,
BFN Panel 9-3 2-ARP-9-3F Unit 2 2-XA-55-3F Rev..0033 Page 10 of 39 Sensor/Trip Point:
Relay 23A-K50 Loss of the 120 VAC from DIV II ECCS HPCI 120 VAC ATU inverter and Loss of power to the POWER FAILURE HPCI Flow IND Controller (2-FIC-73-33) r (Page 1 of 1)
Sensor Panel 9-19 Location:
Aux lnstr Rm, El 593 Probable A.
Blown fuses, Fuse 2-FU2-073-0033C. Panel 2-9-82 AA1 & AA2, Cause:
B.
DIV II ECCS ATU inverter failure.
C. Loss of 250V DC power supply to DIV II ECCS ATU inverter (RMOV BD 2A compt hAl).
Automatic A.
HPCI controller loses power. HPCI becomes inoperable.
Action:
B.
If HPC1 is in service, the HPCI Turbine Stop Valve, 2-FCV-73-1 8, closes. HPCI controller loses power. HPCI becomes inoperable.
C. 2-Pl-064-67B will lose povier and become mop.
Operator A.
DISPATCH personnel to CHECK the following:
El Action:
Fuses 2-FU2-073-0033C, Panel 2-9-82, AM & AA2.
El DIV II ECCS ATU inverter.
El DIV II ECCS ATU inverter breaker, RMOV BD 2A, compt HAl.
El B. 2-Pl-064-67B will lose power and become mop.
REFER TO TECH SPEC 3.3.3.1, Table 3,3.3.1-1, TRM 3.3.5.
El C. REFER TO: Tech Spec 3.5.1. Tech Spec 3.3.3.1, Table 3.3.3.1-1.
El
References:
2-45E620-1 GE 730E928-2 and -4 Technical Specifications 3.5.1, and 3.3.3.1 Technical Specifications Bases 3.3.3.1, TRM 3.3.5
Distribution Systems - Operating 83.8.7 BASES LCO When 480 V Shutdown Board 28 is aligned to the alternate (continued) supply 4.16 1W Shutdown Board C, a LOCNLOOP with a failure of the Shutdown Board D Battery would disable the normal supply 4.16 kV Shutdown Board D, and would also prevent the 480 V Shutdown Board 28 from load shedding its 480 V loads which would overload the alternate supply Diesel Generator D.
This would result in the loss of diesel generators C and D, associated 4.16 kV shutdown boards and RHRSW pumps.
Therefore, the restrictions on the associated drawings shall be adhered to whenever 480 V Shutdown Board 28 is on its alternate supply.
The Unit 2480 V RMOV boards 2A and 28 have an alternate power supply from the other 480 V shutdown board. Interlocks prevent paralleling normal and alternate feeder breakers. The boards are considered inoperable when powered from their alternate feeder breakers because a single failure of the power source would affect both divisions, The Unit 2250 V DC RMOV boards 2A, 28, and 2C have alternate power supplies from another 250 V Unit DC board.
Interlocks prevent paralleling normal and alternate feeder breakers. The boards are considered inoperable when powered from their alternate feeder breakers because a single failure of the power source could affect both divisions depending on the board alignment.
If a 4.16 kV or 480 V shutdown board is aligned to its alternate 250 V DC control power source a single failure of the alternate power source could affect both ECCS divisions and common equipment needed to support the other units depending on the board alignment. Therefore, the restrictions on the associated drawings shall be adhered to whenever a 4.16 kV or 480 V shutdown board is on its alternate control power supply.
HLT 081011006 Written Exam 76.
295004 AA2.01 With Unit 2 operating at 100% Reactor Power, a Normal Supply Breaker has tripped open AND suffered damage due to arcing. In addition to RHR AND Core Spray logic alarms, th following significant alarms result:
HPCI LOGIC POWER FAILURE, (2-9-8A, Window 7)
HPCI 120 VAC POWER FAILURE, (2-9-8A, Window 7)
ADS BLOWDOWN POWER FAILURE, (2-9-3C, Window 32)
Which ONE of the following completes the statements?
The 250 VDC RMOV Board _(1)_ has been lost.
After manually transferring the 250 VDC RMOV Bd, the Board is considered _(2)_ in accordance with Tech Spec 3.8.7, Distribution Systems Operating.
A.
(1) 2A (2) inoperable B.
(1) 2B (2) inoperable C.
(1) 2A (2) operable D.
(1) 2B (2) operable
Clarification Guidance for SRO-only Questions RevI (0311112010>
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing I
Yes hour TSITRM Action?
RO question Can question be answered solely by knowing the Yes LCOITRM information listed above-the-line7 RO question Can question be answered solely by knowing the Yes TS Safety Limits?
RO question Does the question involve one or more of the following for TS, TRM, or ODCM2 Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
Application of generic LCO requirements (LCO 301 thru 101 and SR 401 thru 404)
Yes SRO-only Knowledge of TS bases that is required to analyze TS question required actions and terminology No j Question might not be linked to 10 CFR 55A3(b)(2) for SRO-only
QUESTION 77 A reactor scram has occurred on Unit 1.
In accordance with I -AOl-100-1, Reactor Scram, the RO ATC reports the following:
Mode Switch is in Shutdown 12 Rods are at position 24 o
RPV water level is (-) 9 inches and slowly recovering Reactor Pressure is 955 psig and steady MSIV are open ATWS Actions are Complete and APRMs are downscale Which ONE of the following completes both statements below?
When the RO ATC reports ATWS Actions are Complete and APRMs are downscale this means that BOTH channels of ART are initiated and the (I)
Assuming ART caused NO rod motion, the Unit Supervisor shall (2)
A.
(1) Reactor Recirc Pumps are tripped (2) remain in the RC/L leg of EOI-1, RPV CONTROL B.
(1) Reactor Recirc Pumps are tripped (2) exit the RC/L leg of EOI-l, RPV CONTROL and enter C-5, LEVEL/POWER CONTROL C.
(I) Reactor Recirc Pumps are at minimum speed (2) remain in the RC/L leg of EOI-I, RPV CONTROL D.
(1) Reactor Recirc Pumps are at minimum speed (2) exit the RC/L leg of EOI-1, RPV CONTROL and enter C-5, LEVEL/POWER CONTROL ANSWER: D
Level:
1 Group#
I Examination Outline Cross-Reference KIA#
295006 G2.1.20 Importance Rating 4.6 295006 SCRAM 2.1.20 Ability to interpret and execute procedure steps.
Explanation: D CORRECT: When the RO ATC reports ATWS Actions are Complete this means that BOTH channels of ART are initiated and Reactor Recirc Pumps are at minimum speed. Given the conditions in the stem, i.e. it has NOT been determined that the reactor will remain shutdown under all conditions without boron, RC/L should be exited and C5, Level/Power Control should be entered.
A-Incorrect. First part incorrect Plausible because step RC/Q-3 directs tripping the Recirc pumps under these conditions, however ATWS Actions are Complete in AOl-100-1, Reactor Scram, ONLY means that Recirc Pumps have been run back to minimum speed. Second part is incorrect, plausible if the candidate does not recognize that plant conditions meet override RC/L-3 in EOI-1, RPV CONTROL, directing the operator to exit the RC/L leg of EOI-1, RPV CONTROL and enter C-5, LEVEL/POWER CONTROL.
B-Incorrect. First part incorrect Plausible because step RC/Q-3 directs tripping the Recirc pumps under these conditions, however ATWS Actions are Complete in AOI-100-1, Reactor Scram, ONLY means that Recirc Pumps have been run back to minimum speed. Second part is correct.
C-Incorrect. First part correct. Second part is incorrect, plausible if the candidate does not recognize that plant conditions meet override RC/L-3 in EOI-1, RPV CONTROL, directing the operator to exit the RC/L leg of EOI-1, RPV CONTROL and enter C-5, LEVEL/POWER CONTROL.
Technical Reference(s): 1-EOI-1, RPV CONTROL, 1-AOI-100-l, Reactor Scram, 0DM 4.20 Proposed references to be provided to applicants during examination: None Learning Objective (As available):
Question Source:
Bank:
Modified Bank:
New:
X Question History:
Previous NRC: None Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis X
10 CFR Part 55 Content:
55.43 (5), assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
BFN Reactor Scram 1-AOl-I 00-1 Unit I Rev. 0015 Page 6 of 76 4.2 Subsequent Actions NOTES 1)
The steps in this section are written in general order of importance for most anticipated events: however, they are not required to be performed in order, but as required to maintain stable conditions. Once a step is entered, all associated substeps are required to be completed in order, except those in Step 4.2[33](Return to Service).
Steps which are not applicable for this scram should be N/Ad.
2)
For Scram Response logic to initiate, all of the following conditions must be met:
Scram Response Logic is not inhibited (amber light at SCRAM RESPONSE INHIBIT/RESET switch, 1-HS-46-5 on Panel 1-9-5, is extinguished).
REACTOR WATER LEVEL CONTROL PDS, 1-LIC-46-5 on Panel 1-9-5, is in AUTO and at least one individual RFPT Speed Control PDS in AUTO.
Either RPS A or B Backup Scram channel activates.
Reactor Level (narrow range) falls below 0 inches within 60 seconds of first Backup Scram channel activating.
3)
If Programmed Scram Response is initiated, the logic is reset by ANY of the following conditions:
1.
Placing REACTOR WATER LEVEL CONTROL PDS, 1-LIC-46-5 on Panel 1-9-5 in MANUAL.
2.
Reactor level (narrow range) exceeding level setpoint.
3.
Five minutes expire from the time the Scram Response logic was activated.
4.
Depressing SCRAM RESPONSE INHIBIT/RESET Switch, 1-HS-46-5, on Panel 1-9-5.
[1]
ANNOUNCE Reactor SCRAM over PA system.
D
[2]
IF all control rods CAN NOT be verified fully inserted, THEN PERFORM the following (otherwise NIA):
[2.1]
INITIATE ARI by arming and depressing BOTH of the following:
ARI MANUAL INITIATE, 1-HS-68-119A D
ARI MANUAL INITIATE, 1-HS-68-1 198 D
[2.2]
VER1FY the Reactor Recirc Pumps (if running) at minimum speed at Panel 1-9-4.
D
[2.3]
REPORT ATWS Actions ComIete and power level.
C
BFN Operations Strategies for Successful Transient BFN-ODM-4.20 Directive Manual Mitigation Rev. 0001 Page 13 of 15 4.7.3 RPV Control (EOI-1) (continued)
C.
Power Leg of flowchart If the determination is made that the reactor is subcritical by the use of nuclear instrumentation, then the subsequent actions of AOI-100-1 should be directed, If control rods remain withdrawn from the core and EOl appendices have been directed that will insert the control rods prior to the determination of subcriticality, then the appendices should continue to be used until all control rods are fully inserted into the core.
D.
ATWS Actions It is the expectation that IF all control rods CANNOT be verified fully inserted, the OATC should actuate both channels of ARI, run both Recirc pumps to minimum speed and report ATWS actions are complete and Reactor Power is as per AOl-100-1 Hard Card.
During an ATWS, the US should not exit RC/L and enter C-5, LEVEL/POWER CONTROL, until OATC ATWS actions per the AOI-100-1 Hard Card have been completed and Reactor Power has been reported to the US.
When EOI-1, Step RC/Q-9 is reached, IF core oscillations are observed, THEN INITIATE SLC.
When EOI-1. Step RC/Q-1O is reached, IF reactor power is greater than APRM downscale, THEN INITIATE SLC.
Alternate Lvl Control Cl-2. Cl-2 I.
CAUTION
.1 GE-i L
Ambient temp may affect RPV water M indication and trend MONITOR and CONTROL RPV water wl RGL-l VERIFY each as required:
PCIS isolations (Groups 1, 2 and 3)
- RCIC L
CL-2 L
0--
WHILE EXECUTING THE FOLLOWING STEPS:
IF THEN It has NOT been determined that EXIT ROt and the reactor wll reman subormoal ENTER 05, LeveliPower Control without boron under all conditions RPV water lvi CANNOT be EXIT ROt and determined ENTER C4, RPV FIoodin PC water M CANNOT be maintained below lOft ft STOP in nto the RPV from souroes OR external to the PC NOT required tor adequate care coaling Suppr chmbr press CANNOT be maintained below 55 psg L
xlQ MONITOR and CONTROL reactor power L
V WHILE EXECUTING THE FOLLOWING STEPS:
VERIFY reactor mode switch, in SHUTDOWN L
oc-t INITIATE API RCO-4 L
tnppinLpum NO L RCQ-S YEOL NOL TRIP Pedro pumps JL IF ThEN
- 1. STOP boron ir unless required by other procedures The reactor wril rematn subcrmcal without boron under at ocndh,ions
- 2. EXIT RCIO and ENTER AO[-IaO-1, Reactor Scram The reactor is subcritical AND EXIT RC1Q and ENTER AOlt:OO1 Reactor Scram NO boron has been iniecied L
VERIFY Pedro runback pump speed 4D rpm or tess L
EXECUTE RCiC-i, RCQ-i I and RciQ-1 P conourrently
Figure 2: Screening for SRO-only linked to 10 CFR 5&43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, Le. how the system works, Yes RO uestion flowpath, loc. component location?
q Can the question be answered solely by knowing immediate operator actions?
Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?
Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitve strategy of a procedure?
Does the question require one or more of the following?
Q Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed 0
Knowledge of when to implement attachments and appendices including how to coordinate these items with procedure steps Yes SRO-only 0
Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normaL, abnomml, and emergency procedures No Question might not be linked to 10 CFR 55A3(b)(5) for SRO-only
QUESTION 78 Given the following conditions for Unit 2:
Reactor has scrammed due to a rapid loss of control air header pressure.
Multiple rods remain out Suppression pool temperature is 1150 F Control Air header pressure is 25 PSIG and lowering RPV water level was lowered to (-) 55 inches Which ONE of the following completes the statements below?
MSlVstatusis (1)
In order to re-establish the condenser as a heat sink, the MSIV should be re-opened using (2)
NOTE: 2-AOI-32-2, Loss of Control Air 2-EOI APPENDIX 8B, Reopening MSIVs/Bypass Valve Operation A. (I) all MSIVs CLOSED (2) 2-AOI-32-2 ONLY B. (1) all MSIVs CLOSED (2) 2-EOI-APPENDIX 8B and 2-AOI-32-2 C. (1) inboard MSIVs OPEN, outboard MSIVs CLOSED (2) 2-AOI-32-2 ONLY D. (1) inboard MSIVs OPEN, outboard MSIVs CLOSED (2) 2-EOI-APPENDIX 8B and 2-AOI-32-2 ANSWER: D
Level:
I 1
Group#
I i Examination Outline Cross-Reference KIA#
295019 AA2.02 Importance Rating 3.7 295019 Partial or Complete Loss of Instrument AA2.02 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Status of safety-related instrument air system loads (see AK2.1
- AK2.19)
Explanation: D CORRECT: First Part-CORRECT: On a rapid loss of control air, the outboard MSIVs close at 45 psig.
. Second Part-CORRECT: In order to open MSIVs, 2-EOI-1 directs the use of 2-EOI APPENDIX 8B. However, this appendix does not contain info to align DW control air to the outboard MISVs. 2-AOI-32-2 directs aligning drywell control air to open the Outboard MSIVs. Given the plant conditions in the stem, reopening the MSIVs is appropriate. The candidate must interpret the listed conditions and recognize that the normal air supply to the MSIVs has been lost and that DW air must be aligned to accomplish the goal of the applicable EOI. Thus, a detailed knowledge of procedures needed to open the MSIVs under these conditions is needed.
A-Incorrect: Part (I) Incorrect-It is plausible that all MSIVs would be closed on either a misconception of isolation on level or a loss of control air.. (2) Incorrect-See A.
B-Incorrect: Part (1) Incorrect-See C. (2) CORRECT-See B.
C-Incorrect: Part (1) Correct see A. Part (2) Incorrect-Although, 2-AOI-32-2 directs aligning drywell control air to open the Outboard MSIVs, 2-EOI-1 directs the use of Appendix B to open the MSIVs when re-establishing the Main Condenser as a heat sink.
Technical Reference(s): 2-EOI-l;2-EOI APPENDIX 8B; 2-AOI-32-2 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OPL 171.054 Obj. B.8 Question Source:
Bank:
Modified Bank:
New:
X Question History:
Previous NRC:
Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis :
X 10 CFR Part 55 Content:
55.43 (5), assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations
Figure 2: Screening for SRO-oniy linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing y5 knowledge Le.. how the system works Yes flowpath, logic, component location?
RO question Can the question be answered solely by knowing immediate o rator actions?
Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters RO question that require ct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mjLçve strategy of a procedure?
Does the question require one or more of the following?
0 Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed 0
Knowledge of when to implement attachments and appendices including how to coordinate these items with procedure steps Yes SRO-only Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy. implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only
WHILE EXECUTING THE FOLLOWING STEPS:
W THEN lESS cScEgl rsLJsE DT R0P C14r IESpasM CANtTtE LOWER prrt nri na aas aea rpasi M Ii aas!s sa Ciras 3 rES3IKto sxced I0FJtlr SuØr M NOTbE rrairtasa MAINTAIN PJ E5E LI tithe as 01 alre 4 ea of C1r1e 4 10 WerFLIr IlSR1 TaIe}
WH LE EXECUTING THE FOllOWING STEPSI WH1LE EXECUTING THE FOLLOWNO STEPS:
P 114tH PERIFOthstIlasr1o of1lle d4teraEr a
Blon LIJ s tIES r&
- OPEN f4GVa (AFX E TheirrnnESis 3atie 0l POtS Wa RPJ LtPPX8E
- .1ESTORE iLatLI as Th&e liES beRn no tiIo3brln 3 OECRSENY PPXEE Eypans G 6 sn Ee osaE P05 Wa LIPJ Ll anlI Hli OW PIeRS. APPX*!F LIestire f?a LIE URnitlaLon)
- 6 HPCI cCwilar abcea WOF L
AUGMENT LIPS? FfReR as tisssary StJl NY of 105 lbLoSing OEF10EOSUATKP4SYSTEN SGR C5L SIRS S E5.S S IF REIN STEkI RELSF tLtRE PRESS LC rrlrcStSr ofA-SS-3-lS
- EESIe ThEW PLACES USIeVO RSsitESIfl CLOSE:elsrC PLACE LISRV a4 SIRE X-l-Z2lS S&-1SIT ZI EI 0ETs iSIe 110 010 SES 05755105 t5SIie 115 T551505R lIP LIES Stf 105SI1ES05 110 sRSz 110 RE 0ps05ei P0 do-d R00U 1505 55sF 15 SIlts RPV lIE L
I?
OWOQI1IILI atIlextileR lR L
-r L
STAEIIJZE LIFV R5S& beto.v IIJ7L psIg 5511 ifle 10n 1111151 b)p3esi11s (kZPX 11El L
CAUflON
- 3 10s6 &pr ctmorpcees nay 11F SOlO I
BFN Loss of control Air 2-AOl-32-2 Unit 2 Rev. 0032 Page 9 sf25 42 Subsequent Actions (continued)
[7.2.11 ESTABLISH lube oil temperature between 80F and 100F using the following TCV BYPASS VALVE(s):
For A RFP use 2-24-626A or 3-24-627A 1]
For B REP use 2-24-6268 or 3-24-627B El For C REP use 2-24-626C or 3-24-627C U
[731 CLOSE EHC fluid cooler TCV isolation valves 2-24-592 or 2-24-593, THEN ESTABLISH fluid temperature between 85F and 125SF on Ti-47-59 using TCV BYPASS VALVE 2-24-590 or 2-24-591.
El
[8]
VERIFY dryweil control air system is being supplied by either the Nitrogen System or CAD system.
U NOTE DRYWELL control air can be valved into control air lines for outboard MSIVs.
[9]
IF Unit Supervisor determines outboard MS1Vs need to be opened in order to establish the main condenser as a heat sink, THEN PERFORM Attachment 2. (Otherwise NJA)
U
BFN Reopening MSIVsIBypass Valve 2-EOl Appendix-BB Unit 2 Operation Rev. 0006 Paqe 3 of B 1.0 INSTRUCTIONS LOCATION:
Unit 2 Control Room ATTACHMENTS None
[1]
IF pressure control with bypass valves is desired and MSIVs are open. THEN PROCEED to step 1.O[i 3]
C
[2]
VERIFY ALL MSIV control switches in CLOSE positiom C
[3]
RESET PCIS logic (Panel 2-9-4).
C
[4]
DEPRESS the following pushbuttons to trip RFPTs (Panel 2-9-6):
2-HS-3-125A, RFPT2A TRIP C
2-HS-3-151A, RFPT2B TRIP C
24-lS-3-176A, RFPT 2C TRIR C
BFN Reopening MSlVsIBypass Valve 2-EOI Appendix-SB Unit 2 Operation Rev. 0008 Page 6 of S 1.0 INSTRUCTIONS (continued)
[5j VERIFY CLOSED the following drain valves (Panel 2-9-3>:
NOTE To prevent auto opening OT2FCV158, handswitch 2-HS-i-58A must be held in the CLOSE position until main turbine speed decreases to below 1701) RPM, 2-FCV-1-58, UPSTREAM MSL DRAIN TO CONDENSER U
2-FCV-1-59, DOWNSTREAM MSL DRAIN TO CcNDENSER, U
[6)
VERIFY CLOSED the following drain valves (Panel 2-9-7):
2-FcV-6-100, STOP VALVE 1 BEFORE SEAT DR VLV U
2-FCV-6-10i, STOP VALVE 2 BEFORE SEAT DR VLV U
2-FcV-6-i 02, STOP VALVE 3 BEFORE SEAT DR VLV U
2-FcV-6-I 03, STOP VALVE 4 BEFORE SEAT DR VLV U
7j VERIFY CLOSED the following drain valves (Panel 2-9-6):
2-FCV-6-122, RFPT 2A HP STOP VLV ABOVE SEAT DR U
2-FCV-6-127, RFPT 2B HP STOP VLV ABOVE SEAT DR U
2-FCV-6-132, RFPT 2C HP STOP VLV ABOVE SEAT DR U
f8)
OPEN the following outboard MSIV5 (Panel 2-9-3):
2-FCV-1-15, MSIV LINE A OUTBOARD U
2-FCV-1-27, MSIV LINE B OUTBOARD U
2-FV-i-38, MSIV LINE C OUTBOARD U
2-FV-1-52, MSIV LINE D OUTBOAD.
U
QUESTION79 Unit I was at 100% power when one Safety Relief Valve failed open and was unable to be closed. The Reactor Mode Switch was placed in SHUTDOWN.
The following conditions exist:
- Reactor power 8% and lowering
- Reactor pressure 900 psig and stable
- Suppression pool temperature 182 °F and rising
- Suppression pooi level 16.0 ft and slowly rising Which ONE of the following completes the statement below?
The Unit Supervisor should direct REFERENCE PROVIDED A.
lowering reactor pressure within the limits of the cool down rate, in accordance with 1 -EOI-1, RPV Control B.
lowering reactor pressure irrespective of cool down rate as required, in accordance with l-EOl-1, RPV Control C.
lowering reactor pressure by anticipating Emergency Depressurization in accordance with 1-EQ1-1, RPV Control D.
immediate Emergency Depressurization of the reactor in accordance with I -C-2, Emergency RPV Depressurization Answer: B
Level:
1 Group#
1 Examination Outline Cross-Reference K1A#
295025 EA2.03 Importance Rating 4.1 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE:
Suppression pool temperature Explanation: Answer B CORRECT: lowering reactor pressure irrespective of cooldown rate is correct for given conditions A
incorrect plausible. In that lowering reactor pressure is correct but the requirement is OK to exceed cooldown rate limits C incorrect plausible ED is likely but cannot anticipate ED in an ATWS D
incorrect plausible with margin to HCL and RPV pressure at 900 psig ED is not correct for current conditions.
Technical Reference(s): 2-EOI-land 2-EOI-2 Flowchart Proposed references to be provided to applicants during examination: Heat Capacity Temperature Limit Curve 3 Learning Objective (As available): OPL171.202 V.B.15 Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: Perry NRC 2009 #3 SRO Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis :
X 10 CFR Part 55 Content:
55.43 b(5), assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Ir p1 tep and lvi CANNOT be WRPrsto mtn
- mawitaaned in a safe area of suppr p1 temp and lvi in a safe area a Curve 3 at the existing RPV press of Curve 3 (ok to exceed 100°F/hr cooldown rate) r H EN
.ppr p1 tamp and lvi N:OT be aofcuwe3 SPfl I
1(
EMERGENCY RPVDEPRESS1JRIZATION IS. REQUIRED 1EO-i, RCP4 Cl-i. Cl-2D; 05-12, 05-14}
L Curve 3 Heat Capacity Temp Limit 260 250 240 210-H
200 c
190 0.z(
IDU 170 160 150 itS 12 ACTION REQUIRED if above curve for existing RPV press Suppr P1 Lvi (if)
Figure 2: Screening for SRO-only linked to 10 CFR 5&43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing ytni knowledge [e, flow the system works Yes flowpath, logic, componert location?
RO question Can the question be answered solely by knowing I
immediate operator actions?
Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?
Can the question be answered solely by knowing the puipose, overall sequence of events, or Yes RO question yflI 2
itigaive strategy of a procedure?
Does the question require one or more of the following?
Q Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed 0
Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SROOnly 0
Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only
QUESTION 79 1108 audit high miss 42.9% got it right Which ONE of the following completes the statements below in accordance with 2-EOl-2, Primary Containment Control and 2-EOl-1, RPV Control?
The SRO _(1)_ direct exceeding the Tech Spec cooldown limit in order to maintain within the SAFE region of the Heat Capacity Temperature Limit (HCTL).
If the plant enters the UNSAFE region of the Heat Capacity Temperature Limit (HCTL), then the SRO at this time _(2) allowed to reduce pressure AND return to the SAFE region to avoid an unnecessary Emergency Depressurization.
A.
(1)may (2) is B.
(1)may (2) isNOT C.
(1) may NOT (2) is D.
(1) may NOT (2) sNOT ANSWER: D A
INCORRECT: Part 1 correct See explanation B. Part 2 incorrect
See Explanation C B
CORRECT: Part 1 correct Reactor Pressure may be reduced as required to maintain within the safe area of HCTL, even in an ATWS.
Part 2 correct Pressure reduction is allowed to prevent from entering the UNSAFE region, but once there the plant is not allowed to restore to the SAFE region except by Emergency Depressurization.
C INCORRECT: Part I incorrect Plausible in that the SRO may believe with an ATWS in progress a cooldown may not exceed TS limits or may not be initiated. Part 2 incorrect Plausible in that some conditions requiring Emergency Depressurization lAW EOIs allow for the parameter to be restored and maintained.
D INCORRECT: Part I incorrect See explanation C. Part 2 correct
See Explanation B.
QUESTION 80 Given the following Unit I conditions:
A Loss of off-site power has occurred OATC reports 26 control rods are at position 02 and the remaining rods are at position 00 e
Reactor Power is unknown RPV pressure is being maintained between 800 to 1000 psig RPV level is (-) 20 inches and steady Suppression Pool temperature is 102° F and steady Which ONE of the following completes both statements below?
In accordance with Emergency Operating Instructions, the RPV Level Band is (1) inches.
(2) required to be initiated in accordance with 1-EOI-1.
A. (1)(+)51 to(-) 180 (2) is B. (1)(+)51 to(-) 180 (2) is NOT C. (l)(-)5Oto(-) 180 (2) is D. (1)(-)5Oto(-) 180 (2) is NOT ANSWER: D
Level:
1 Group#
1 Examination Outline Cross-Reference 1<1A#
295037 EA2.05 Importance Rating I
Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following: Control rod position Explanation: D CORRECT: No RPS power is available which eliminates the use of APRMs for power level determination. Rod Position indication is available; as it is powered from DIG backed Unit Preferred or Batteries, and based on the conditions stated in the stem, sufficient power is available to use the RPIS function. Subcritical under all conditions is sufficient to determine the power and level control functions of the EOIs. First part: The reactor is NOT subcritical under all conditions. Therefore RPV water level is lowered in step C5-l0 to (-)50 inches. A water level band of(-)50 to (-) 180 inches would be correct. SLC would NOT be required because Suppression Pool Temperature is 102° F and steady.
A-Incorrect. First Part: Incorrect. Plausible because this is the normal control band for RPV water level in a subcritical reactor. Second Part: Incorrect. Plausible because thus is still in the RCIQ leg of EOI-1.
However suppression Pool temperature is less than 110° F and steady.
B-Incorrect. First Part: Incorrect. Plausible because this is the normal control band for RPV water level in a subcritical reactor. Second Part: Correct.
C-Incorrect. First Part: Correct. Second part: Incorrect. Plausible because thus is still in the RCIQ leg of EOI-1. However suppression Pool temperature is less than 110°F and steady.
Technical Reference(s): I -EOI-1, 1 -EOI-C5 Proposed references to be provided to applicants during examination: None Learning Objective (As available):
Question Source:
Bank:
Modified Bank: X New:
Question History:
Previous NRC: BFN 0610 (2006) #80 Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis:
X 10 CFR Part 55 Content:
55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
BFN EOI-1, RPV CONTROL BASES EOIPM SECTION 0-V-C Unit 0 Rev. 0002 Page 21 of 125 1.0 EOI-1, RPV CONTROL BASES (continued)
I DISCUSSION: RCIP-3 I
Positive confirmation that the reactor will remain subcritical under all conditions is best obtained by determining that no control rod is withdrawn beyond the Maximum Subcritical Banked Withdrawal Position (MSBWP, **A70**) The MSBWP is the greatest banked rod position at which the reactor will remain shutdown under all conditions. Refer to EQ 1PM Section O-lI-ZB for discussion of the MSBWP.
1.0 EOI-1, RPV CONTROL BASES (continued)
I DISCUSSION: RCIP-3(contd)
I Other criteria may also be employed to determine that the reactor is shutdown.
Possibilities are listed in Note 1. Note 1 identifies the bounding control rod positions that ensure the reactor will remain subcritical without boron under all conditions when Reactor Engineering is not available to support this determination. This instruction requires a positive determination, not only that the reactor is subcritical, but that it will remain subcritical, without reliance upon boron, under worst-case cold shutdown conditions. The phrase without boron does not imply that the condition cannot be met if boron has been injected, but that credit cannot be taken for the negative reactivity contributed by the boron. Control rod insertion alone must provide the necessary shutdown margin.
NOTE
The reactor will remain subcritical without boron under all conditions when:
Any 19 controt rods are at position 02 with all other control rods ruDy inserted OR All control rods except one are inserted to or beyond position 00 OR Determined by Reactor Engineering
TSC stafl may recommend an alternate curve br Station Blackout per O-AOl-57-IA
=
L WHLE EXECUTING THE FOLLOWING STEPS:
IF THEN it has NOT been determined that EXIT RC/L and the reactor will remain subcritical ENTER C5, Level/Power Control without boron under alt conditions RP\\/ water lvi CANNOT be EXIT RC(L and determined ENTER C4, RPV Flooding PC water lvi CANNOT be maintained below 105 ft STOP inj into the RPV from sources OR external to the PC NOT required for adequate core cooling Suppr chmbr press CANNOT be maintained below 55 psig APX tr CN FW 4
H a
LPC a
Cs L
L L
Figure 2: Screening for SRO-only linked to 10 CFR 5&43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledg&, Le., how the system works, Yes flowpath, logic, component location?
RO question Can the question be answered solely by knowing I
immediate operator actions?
Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?
Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question yallitive strategy of a procedure?
Does the question require one or more of the following?
Q Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Knowledge of when to implement attachments and appendices including how to coordinate these items with procedure steps Yes SRO-only 0
Knowledge of diagnostic steps and decision points in the q1sti0n EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, andfor coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-orily
0610 SRO Final Examination
- 80. SRO 29503 7EA2 05 002 Given the following Unit 1 conditions:
A Loss of off-site power has occurred.
OATC reports 26 control rods are at position 02 and the remaining rods are at position 00.
APRM indications are NOT available.
RPV pressure is being maintained between 800 to 1000 psig.
RPV level is (-)20 inches and steady.
Suppression Pool temperature is 102 OF and steady.
Which ONE of the following describes the appropriate actions to be performed in accordance with Emergency Operating Instructions?
Level band is (+)2 to (÷)51 inches.
SLC should NOT be initiated.
RPV cooldown is permitted.
B. Level band is (-)50 to (-)100 inches.
SLC should be initiated.
RPV cooldown is NOT permitted.
C. Level band is (+/-)2 to (+/-)51 inches.
SLC should be initiated.
RPV cooldown is NOT permitted.
D. Level band is (-)50 to (-)100 inches.
SLC should NOT be initiated.
RPV cooldown is permitted.
QUESTION 81 The following conditions exist for Unit 1:
o Mode3 Reactor Pressure 75 psig Core Spray Loop 2 is INOPERABLE RHR Loop 2 is operating in Shutdown Cooling Which ONE of the following subsystems, if inoperable, would require an Hourly Fire Watch?
[REFERENCE PROVIDEDI A. Preaction System for HPCI, 1-26-37 B.
Fire Detection for RCIC on Panel 1-LPNL-25-545 C.
Fire Detection for RHR on Panel 1-LPNL-25-545 D.
Unit I Auxiliary Instrument Room C02 System ANSWER:
D
Level:
1 Group#
1 Examination Outline Cross-Reference KIA#
600000 AA2.1 5 Importance Rating I
Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE: Requirements for establishing a fire watch Explanation: D CORRECT: Unit I Aux Instrument Room is required to be Operable and with the C02 system inoperable it is required to establish an hourly fire watch.
A-Incorrect. Plausible in that this would be correct if HPCI was OPERABLE.
B-Incorrect. Plausible in that this would be correct if RCIC was OPERABLE.
C-Incorrect. Plausible because this would require a continuous Fire Watch.
Technical Reference(s): Fire Protection Report Volume I Proposed references to be provided to applicants during examination: Fire Protection Report Volume 1 Learning Objective (As available):
Question Source:
Bank:
Modified Bank:
X New:
Question History:
Previous NRC: Perry 2009 SRO #4 Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis:
X 10 CFR Part 55 Content:
55.43 (1) Conditions and limitations in the facility license.
Fire Protection Report Volume I 3.i1D 002 SYSTEMS 002 The low pressure CO aystena protecting
- 2. ach of the required C2 syataas shall.
the followino areas shall be onaaAnt be d onstrated OPAaLa whenever e ipnent pr ected by the no syatanis is required to be Onasta.
a.
Lnit 2 and 2 D:ieaei Oenerator a.
At least weekly by ve:r+/-fylnq the CO Rooms, AuxiLtary Ocard Roors, Fuel storage tank level to he qreater Cransfer Puop R000s than B.t tons for Units 1 and 2 and 2 taos for Jait I and nressure to
- b. unit a D.ieasl Oenerator noons, be reater than 27F pain, and Auxiliary Ooard Scorns, and Fuel Transfer Punp 50005
- b. At least once per 15 nonths by verifyi.n:
- c. computer Scone 1,2, and 3 Pt control 2uilding 1.
The systen,
.includin associated vent i la tiom system ii. re danpers
- d. Auxiliary Inatrurrent Scoms 2,
and fire door release and 3
Tnechanisns, actuates nanually and automatically upon receipt 2.
WIth one or more of the above co.
of a sinulated actuation atonal, systems inoperable, wIthin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish an hourly fire watch patrol.
2.
Flow from each nozzle durinc, a NO1E:
A delay of fifteen %5 Puff Test.
ninutem for the start of rounds is acceptable during shift turnover, to facilitate proper turnover with Olninal plant inpact.
- 2. With one or more of the above required local control panels inoperable, within one hour establish a continuous fire watch for those areas specifically identified in Table 9.311.A; for other areas listed in Table 3llA, establish an hourly rovinn fire watch.
a The fire detection systems heat and smoke detectors for all protected areas shall be OPERABLE
- b. If recuiremneut P311.A.2a cannot be met, a patrolling fire watch will be established unless noted othervise in Table 31iA to ensure tac each protected fire zone or area with inoperable detectors is checked at intervals no greater than once each hour.
NOTE delay of fifteen IS) minutes for the start of rounds is acceptable during shift turnover, to facilitate oroper turnover with minimal plant impact.
g3ilA PIrE DETECTION 41iA PINE DETECTIGI INSUERTION INSTPUETATION 1.
As a minicn.n, the fire 1.
Each of the required fire detection detection local control panel instruments shall he demonstrated shown in Table P3iLA shall be OPERABLE at least annually by OPERABLE whenever ecu_poent erfonuance of a CHANNEL FUNCTIONAL troected b1 the fire de-ectsc-i ET Instrument is required to be S
OPERABLE 2.
The supervised circuits associated with alarms of each of the above required fire detection instruments shall be demonstrated OPEBA3LE semiannually.
3 The nonsupervised circuits associated with alarms of each of the above required instruments shall be Demonstrated OPERABLE at least monthly
- wEan one or more or toe raoulreo rite alan oartela are inaoeraote wlthln one hour eatabtlali a continuous tire aaron rot the area(e( or coverage aeaooiated alto: the at-plloaola alan panel tel The roflowang Ia a liar or Ire parela/alevarlona/ayate-zta requirIng a ointlnuote tire aaron when looperatle ltFilt2t--ii5 (III lla/t4L net. tea, etI, cad/i -ac-la 2-lENt-tt-ill (El tl-e/l4t, liii; 2-LFEt-2t--b (EL l2(
and/or 2-tPt(t-2l-t47 (Ct e21, tie)/a-ae-77 3-LPEL--Zit--hit (EL iea(
tEtlattiE (hi, tat); 2idtit25-h47 (EL eal,eaetfa-2e77 anPlllr-ll-iaallntake Puopiog Elation Cravat too ttil/d-2c-72E The ooottououe tire watch will oot he etatl.oned Iii one locatIon, rut wIll rove conllnuoualy throughout the area(a( normally protected ny toe alan-panel (a) each nour. The coot bonus tIre watch ahal I not leave the epeott led area Ia I wIthout a proper teller Oependl ng or the capabIlity or the tIre watch to oooiate the oatrol or the depboyrerit araa(ei within the allotted tIre rraee, ooe rite watch may be reapnatbl a tor eultlple par.ela/eyarems located wlthln or.a or core unit (al 0! the reactor sot Idlog. The tire watch tor the lotake Pumping eratton nay not nave reapona-lollitlea that would require leaving the rotate Punolng atatlon aa part or the natrol.
- NOTE, to Ooqueraatory eeaauree are reauloed It a alngte detector Ir a tire area/tone Ia loaperable, unleea the deteotor that Ia dereoulnad ironerable ia the only detector In that rote/area to-a only detector ot a orosa-- toned or 2 out or a logic actuatIon eyateo.
a door detector It the detector oeiaalna Inoparaale ad daya or longer, oorpenaatorj neasurea need to be established per aeotbon C.
li.a.a.o or evalitared by tlte EngIneerIng Oh-a oaae-oy--oaaa basIs (tatereooa a.aai.
tend reran inn tonal area drnteered/
detector (sailding EL) hand Eaoirmant jne Puncoann I.
Reantar
- a:e l--ranr--aa--tae ant aauke actuate Preaet;un tyanen a.
aea:tor
- era i-taar-ae-aee ccl!
aeat/tneke anaanntation at.
aeaatar
- ate/cal irlab-ta--liaa bet teaks Actuate tresetioa tyaren
- -i.
aeantan-- ace I -rtlir-la-tal camerat area teaks actuate treantira Oyanara
- a.
aaaator eat I-trilL-ta-ate aer.eral hrea brake datuata treaetian Ityaten at.
Reactor cli
--ttat-Za-a-ili oaneral area brake Actuate traactioa tyaten 7.
Rear.tar
-- ebb r--rrbn.--aa--a-aa aenenal area (truth ride) teaks actuate tresatientyatan asneral Area tmorth tidal brake anaanatatier.
- 1. The spray and sprinkler systems in Table 9311. B shall be OP3LE whenever eruipoent protected by the spray and/or sprinkler sy ems is reonired to be CPER3LE gprink.ler and/cr spray eystens are considered
- pebl if their water zupoiy is unavailable.
- 2. With one cr more of the above required spray end/or sprinkler systems innmerabie. within one hour establish a continuous fire watch for those areas specificatly identified in Table 3lL3r for other areas listed in Table 3llB establish a roving hourly fire watch patrol.
- a. Icr sprinkler and/or spray systems, the associated sprinkler/spray nozzles for all protected areas shall be OPEBABLE.
- b. If requirement 3.LLC.2.a cannot be mat, a rovIng hourly fire watch patrol will he estab.lishedur.less noted c.chenise in Table.3.ll.E),
to ensure that each protected area with inoperable sprinkler/spray nozzles is checked at intervals no greater than once each hour.
NOTB:
A delay of fifteen (IEI minutes for the start of rounds is acceptable during shift turnover to facilitate proper turnover wtth minimal lant impact.
L Each of the required spraY and systems in Table B.3. lB shall be demonstrated OPEEABLE:
a.
intentionally left blank.
b.
Ar least yearly by cycling each testable valve in the flow path Through at least one complete cycle of full travel.
c.
At least once per lB months (1)
By performing a system functional test which includes simulated automatic actuation of the system, verifying that the automatic valves in the flow path actuate to their correct positions on a fire alarm test signal.
By a visual InspectIon of the lion-air supervised spray and sprinkler headers to verify their integrity.
(3)
By a visual inspection of each srinkler or water spray nc.zzles spray area to verify that the spray pattern is not obstructed.
- d. At least once per 3 years, by performing an air flow test through each open head spray header and verifying that each open head spray and sprinkler nozzle is unobstructed.
g3l1C S?PAY ND/Ofl SPRINKZEfl
.4i1C SPRAY AND/On SPP1NXLE SYSTEMS SYSTEMS
ISRC EXAM - 2009 QUESTION SRO 4 Which one of the following conditions requires an Hourly Fire Watch Patrol?
Reference Provided: PAP-1910 Fire Protection Program Body & Attachment #3 A.
RCIC Pump Room Wet-Pipe Sprinkler will not deliver water.
B.
Heat Detection for Reactor Recirculation Pinup B is out of service.
C.
Unit 1 Division 1 Cable Spreading Pre-Action Spray System will not deliver water.
D.
General area smoke detectors in Containment are fiuiictional but the detection system will not transmit an alarm to SAS.
Clarification Guidance for SRO-only Questions RevI (0311112010)
Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D1.c];
A.
Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
Reporting requirements when the maximum licensed thermal power output is exceeded.
Q Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
Processes for TS and FSAR changes.
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
QUESTION 82 The following conditions exist for Unit 3 after the Reactor scrammed due to a LOCA:
Reactor water level is (-) 165 inches and lowering RHR pump 3A is lined up and injecting to the RPV as the only available injection source O-AOI-57-1 B, Grid Instability, has been entered RHR pump 3A amps are in the Red band Which ONE of the following completes both statements below?
RHR pump 3A (F) remain running.
The HIGHEST given water level at which Emergency Depressurization is required is (2) inches.
A. (1)can (2) (-) 180 B.
(l)can (2) (-) 195 C.
(1)CANNOT (2) (-) 180 D.
(1)CANNOT (2) (-) 195 ANSWER:
A
Level:
I Group#
1 Examination Outline Cross-Reference KIA#
700000 G2.4.9 Importance Rating 4.2 700000 Generator Voltage and Electric Grid Disturbances G2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.
Explanation: A CORRECT: Part 1: CORRECT-RHR pump 3A can remain running as it is required to maintain adequate core cooling. Part 2: CORRECT-Emergency Depressurization would be required prior to
(-)180 inches lAW 3-C-i, Alternate Level Control.
B-Incorrect: Part 1: CORRECT-See A. Part 2: Incorrect-This is plausible as Emergency Depressurization is required at (-) 195 inches JAW 3-C-I, Alternate Level Control, while in steam cooling.
C-Incorrect: Part 1: Incorrect-This is plausible as pumps not required for adequate core cooling that are operating in the red band should be secured lAW 0-AOI-57-IE, Grid Instability. Part 2: CORRECT-See A.
D-Incorrect: Part 1: Incorrect-See C. Part 2: Incorrect-See B. In addition if RHR pump A was secured that would place the plant in steam cooling in which (-) 195 inches would be the correct answer.
Technical Reference(s): 3-C-i, Alternate Level Control; 0-AOI-57-iE, Grid Instability Proposed references to be provided to applicants during examination: None Learning Objective (As available): OPL17I.205 Rev 10 ILT Obj. 2 Question Source:
Bank:
Modified Bank:
New:
X Question History:
Previous NRC: None Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis:
X 10 CFR Part 55 Content:
55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
O-AOI-57-IE Rev 11 BFN Gild nstabiity Q-AOi47iE Uriit Rev. O11 Page 10 f IS 4.2 Subsequent Action fcnünued
[7A]
IF system irztes are hgl due io abnmaI grid conditions. THEN ADJUST fiow per pmpraie 01 (within allowed lirnits to lower i9owrate as desired..
D
[75 IF system pump amps are hç:h due to artomiI grid conditions, THEN ADJUST flow per appropriate 01 to lower pump amperage ratee to alowed Firns Cersire less ttian red ban cr control room ammeters) per the apprcprate 01 D
NOTE Pumps that are operating in e conoI room ammeter red band shci1d be secured unless required to assure adequale core coding as dected by the E01s.
[7 6]
IF system pump fowrates andbr amps canrot be restored wthio normaI baids, ThEN EALLJATE the riced to secure the aff1cd pumpisyste!n.
D
[7.7]
IF Dlesel Generators are suppIng the 4 kV Shutdon Boards, T:HEP PERFORM applicable sections of O-A01-57-IA in parafiei wilh thks insrudon.
D
3-C-I ALTERNATE LEVEL CONTROL Rev 12 L
I ETh pcQC I acc F
L.
L L
L
Figure 2: Screening for SRO-oniy linked to 10 CFR 5543(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge ie. how the system works Yes flowpath, logic, component location?
RO question Can the question be answered solely by knowing I
immediate operator actions?
Yes RO question I
Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?
Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitive strategy of a procedure?
Does the question require one or more of the following?
Q Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Knowledge of when to implement attachments and appendices including how to coordinate these items with procedure steps Yes SRO-only Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, andfor coordination of plant normal, abnormal, and emergency procedures No j Question might not be linked to 10 CFR 55.43(b:l(5) for SRO-only
QUESTION 83 During a Unit 2 startup, the following conditions exist:
REACTOR MODE SWITCH in STARTUP/STANDBY Reactor pressure is 855 psig Control rod 22-11 is at position 00, its nitrogen accumulator has a cracked weld and is isolated for repair.
The operating Control Rod Drive (CRD) pump trips and the following conditions are noted:
CRD Charging Water Header Pressure 900 psig CRD ACCUM PRESS LOW/LEVEL HIGH (2-9-5A Window 29) annunciator is received for the following rods:
Rod Position Accumulator Pressure 18-27 00 900 psig 3 8-23 48 900 psig Which ONE of the following are the correct actions to take in accordance with Technical Specification 3.1.5, Control Rod Scram Accumulators?
A.
Declare BOTH control rods inoperable immediately AND immediately place the REACTOR MODE SWITCH in SHUTDOWN.
B.
Fully insert control rod 38-23 and declare BOTH control rods inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If the first action is NOT met, immediately place the REACTOR MODE SWITCH in SHUTDOWN.
C.
Declare ONLY control rod 3 8-23 inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If any other accumulator becomes inoperable, immediately place the REACTOR MODE SWITCH in SHUTDOWN.
D.
Declare control rod 3 8-23 slow within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND if charging header pressure CANNOT be restored to at least 940 psig within 20 minutes, place the REACTOR MODE SWITCH in SHUTDOWN.
ANSWER:
B
Level:
I Group#
2 Examination Outline Cross-Reference K/A#
295022 AA2.01 Importance Rating 3.6 Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS : Accumulator pressure Explanation: B CORRECT: For the given condition this is action C, and if C.1 cannot be completed within one hour, then action D requires the mode switch in shutdown.
A-Incorrect. Plausible in that both Control Rods are inoperable but there is no immediate requirement to declare inoperable. In addition the requirement to place the mode switch in shutdown is not immediate unless certain conditions are NOT met within one hour.
C-Incorrect. Plausible in that this is part of Condition B and part of Condition C and part of Condition D.
D-Incorrect. Plausible in that this is the correct answer for Condition B with Reactor Pressure greater than 900 psig.
Technical Reference(s):
Unit 2 Tech Spec 3.1.5 Proposed references to be provided to applicants during examination: Unit 2 Tech Spec 3.1.5 Learning Objective (As available):
Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: Perry 2004 #84 Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis :
X 10 CFR Part 55 Content:
55.41 (2) Facility operating limitations in the technical specifications and their bases.
Control Rod Scram Accumulators 3.1.5 3.1 REACTiVITY CONTROL SYSTEMS 3.1.5 Control Rod Scram Accumulators LCO 3.1.5 Each control rod scram accumulator shall be OPERABLE.
APPLICABILrrY:
MODES I and 2.
ACTIONS Separate Condition entry is allowed for each control rod scram accumulator.
CONDITION REQUI RED ACTION COMPLETION TIME A. One control rod scram A.1
-NOTE------
accumulator inoperable Only applicable if the with reactor steam dome associated control rod pressure 900 psig.
scram time was within the limits of Table 3.1.4-1 during the last scram time Surveillance.
Declare the associated 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> control rod scram time sIow.
OR A.2 Declare the associated 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> control rod inoperable.
(continued)
Control Rod Scram Accumulators ACTIONS (continued) 3.1.5 CONDITION REQUIRED ACTION COMPLETION TIME
- 8. Two or more control rod B.1 Restore charging water 20 minutes from scram accumulators header pressure to 940 discovery of inoperable with reactor psig.
Condition 8 steam dome pressure concurrent with 900 psig.
charging water header pressure
<940 psig AND 8.2.1
NOTE Only applicable if the associated control rod scram time was within the limits of Table 3.1.4-1 during the last scram time Surveillance.
Declare the associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> control rod scram time slow.
t OR 5.2.2 Declare the associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> control rod inoperable.
(continued)
Control Rod Scram Accumulators 3.1.5 One or more control rod scram accumulators inoperable with reactor steam dome pressure
<900 psig.
Ci Verify all control rods associated with inoperable accumulators are fully inserted.
AND Immediately upon discovery of charging water header pressure
<940 psig D. Required Action and associated Completion Time of Required Action B.1 or C.1 not met.
C.2 Declare the associated control rod inoperable.
D.1
NOTE-Not applicable if all inoperable control rod scram accumulators are associated with fully inserted control rods.
Place the reactor mode switch in the shutdown position.
ACTIONS (continued)
C.
CONDITION REQUI RED ACTION COMPLETION TIME 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Immediately
Clarification Guidance for SRO-only Questions Rev 1(03/1112010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing I
Yes hour TSITRM Action?
RO question Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line?
RU question Can question be answered solely by knowing the Yes TS Safety Lims?
RD question Does the question involve one or more of the following for TS, TRM, or ODCM2 Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
Application of generic LCD requirements (LCD 101 thru 101 and SR 4.0.1 thRi 40.4)
Yes SRO-only O Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CER 55A3(b)(2) for SRO-only
Perry 2004 #84 SENIOR REACTOR OPERATOR Page 7 QUESTION 084 During a plant startup, the following conditions exist:
REACTOR MODE SWITCH in STARTUP/STANDBY Reactor pressure is 855 psig.
Control rod 22-11 is at position 00, its nitrogen accumulator has a cracked weld and is isolated for repair.
The operating Control Rod Drive (CR0) pump trips, CR0 Charging Header Pressure indicates 50 psig, and the CRD HCU LEVEL HI/PRESS LO annunciator is received forthe following rods:
Rod Position Accumulator Pressure 18-27 00 1500 psig 38-23 48 1500 psig Which ONE of the following should you direct the control room operators to do?
a.
Dedare both CR0 accumulators INOPERABLE and have the Supervising Operator place the REACTOR MODE SWITCH to SHUTDOWN.
b.
Declare control rod 38-23 accumulator INOPERABLE; insert and isolate control rod 38-23 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or place the REACTOR MODE SWITCH to SHUTDOWN.
c.
If charging header pressure CANNOT be restored to at least 1600 psig within 20 minutes, place the REACTOR MODE SWITCH to SHUTDOWN. Both control rods are still OPERABLE.
d.
Declare control rod 18-27 and 38-23 iNOPERABLE. Monitor accumulator status.
If any other accumulator becomes INOPERABLE, immediatety place the REACTOR MODE SWITCH to SHUTDOWN.
ANSWER:
B
QUESTION 84 Unit I has experienced a LOCA and the following containment parameters exist:
o All Control Rods fully inserted MSIVs are Open Drywell Pressure is 23.4 psig and rising Suppression Chamber Pressure is 22 psig and lowering slowly Hydrogen concentration in the Drywell is 2.9%
Suppression Pool Level is 15 feet Emergency Depressurization in progress Reactor Water Level is (-) 170 inches and rising Which ONE of the following completes the statements below?
The required procedure to vent primary containment is (1)
Vent under these conditions (2)
A.
(1) 1-EOIAPPENDIX-12, Primary Containment Venting (2) irrespective of offsite radioactive release rates B.
(1) 1-EOIAPPENDIX-12, Primary Containment Venting (2) ONLY if offsite radioactive release rates can be maintained below ODCM limits C.
(1) l-EOIAPPENDIX-15, RPV Venting for Primary Containment Flooding (2) irrespective of offsite radioactive release rates D.
(1) 1 -EOIAPPENDIX-15, RPV Venting for Primary Containment Flooding (2) ONLY if offsite radioactive release rates can be maintained below ODCM limits ANSWER: B
Level:
1 Group#
2 Examination Outline Cross-Reference KIA#
295010 G2.4.20 Importance Rating I
295010 High Drywell Pressure 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.
Explanation: B CORRECT: Part 1-CORRECT: 1-EOJ-2, requires PC vented when Hydrogen is detected at greater than 2.4% lAW l-EOl APPENDIX-12, Primary Containment Venting. Part 2-CORRECT-Venting is to be stopped if offsite radioactive release rate reach ODCM limits A-Incorrect. Part 1: correct Second Part-Incorrect:Plausible because if venting lAW appendix 13 and / or Appendix 15 Off site release rates limits may be exceeded.
C-Incorrect. Part 1-Incorrect, plausible in that this would be correct if the SRO remained in the Primary Containment Flood flow path if RPV water level did not recover above minus 180 inches.
Part 2-Incorrect for appendix 15 venting stack release rates would be maintained within table 7 limits and the off Gas Release rate limits may be exceeded D-Incorrect. Part 1-Incorrect: See C.Part 2-Correct: See B.
The Current Venting procedures are rarely used in training due to the conditions needed to meet the requirements to vent.
Technical Reference(s): I -EOI-2; 1-EOI-C-1, I -EOIAPPENDIX-1 2; 1 -EO1 APPENDIX-IS Proposed references to be provided to applicants during examination: None Learning Objective (As available):
Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: BFN 1006 NRC Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis :
X 10 CFR Part 55 Content:
55.41 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
EO1 Appendix-12 PRIMARY CONTAINMENT VENTING CAUTION Stack release rates exceeding 1.4 x 107 pCils, or 0-SI-4.8.B.1.a.1 release fraction above 1.0 will result in ODCM release limits being exceeded.
ADJUST 1-FIC-84-19, PATH B VENT FLOW CONT, or 1-FIC-84-20, PATH A VENT FLOW CONT, as applicable, to maintain ALL of the following:
Stable flow as indicated on controller, AND U tPA-84-21, VENT PRESS TO SGT HIGH, alarm light extinguished, AND U Release rates as determined below:
iii.
IF Venting for ANY other reason than items i or ii above, THEN MAINTAIN release rates below U Stack release rate of 1.4 x 107 pCi/s AND U 0-Sl-4.8.B.1.a.1 release fraction of 1.
- lip,
\\ PCH,/
\\ / --LU-.
1s-J qr
-f LE EXECJTI.G 1iE OLLOW bJG STEPS:
VERIFY Ha!02 anyzerln seM fAPPX i WHEN HjIsetei in PC 24%c rzntrz ir riaxs EO1-2 iF Th!$
rJtu s&sm NOTIFY Cl Lab ar1r IrperaIe
.arI:&Jc nrHardO SECURE PC reais CDCM lnhI p
C.
Fi2 Is NO ereie i PC SECURE PCt NOT rsireiI by
{Z4 ether Eiseps IF crsze r1adFty rsIsase rae s epeIe rsiai tei or*cp Inits THEN VENT ar PURGE PC as 1cia t
NOTIFY CnLaforoTsIIe raaMty reeaae 2
VEIU PC fAP t2
.3.
W PC ar be vsrte ThEN PURGE PC wllti nItrogen rvaksp (APPX 14A)
L L
EOl Appendix-15 RPV VENTING FOR PRIMARY CONTAINMENT FLOODING CAUTION Off-Gas Release Rate Limits may be exceeded.
Ci-26 Table 7 Max Post LOCA Stack Release Rates tine After Core Noble Gases Uncovery Chr)
{pCc) 0 0.5 3.12 1O 0.52 1.8x 1O 22 8.36x10 824 3.25 a l0 2448 5.38x1W 48-96 3.98s1 96120 3,10sl0 120240 1.97s 10 240480 2.8910 480720 4.53 a 10 VENT the RPV. MAINTAIN offsite radiosctivity release rates within Table 7 iirnit APPX 15)
Ci38
[liKEN PC water lvi reaches ft MAINTAIN PC watered beween 90 ft and 1 Q ft with the foilowing mi sources talcing suction from sources external to the PC ONLY as required:
l:NJ SOURCE APPX IN PREc ONOS SA 480 psig cR0 58 1640 peg LPCI 320 psip
Question 97 1006 Exam Unit 3 was operating at 100% Reactor Power, when a coolant leak in the Drywell caused a Reactor Scram. The following conditions are noted:
ALL Control Rods fully inserted Drywell Pressure is 23.4 psig and lowering slowly Suppression Chamber Pressure is 22 psig and lowering slowly Suppression Pool Level is 15 feet MSIVs are OPEN Reactor has been Emergency Depressurized Reactor Water Level lowered to (-) 180 inches and is now (-) 170 inches and rising Given these conditions, which ONE of the following completes the statement?
In accordance with the EOls, venting the Primary Containment is required to be performed using A.
3-EOl APPENDIX-i 2, Primary Containment Venting, irrespective of radioactive release rates B.
3-EOl-APPENDIX-i 5,RPV Venting for Primary Containment Flooding, irrespective of radioactive release rates C.
3-EOl APPENDIX-12, Primary Containment Venting, ONLY if radioactive release rates can be maintained below ODCM limits D.
3-EOl-APPENDIX-i5,RPV Venting for Primary Containment Flooding, ONLY if radioactive release rates can be maintained below ODCM limits
Figure 2: Screening for SRO-only tinked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge ie flow the system works Yes flowpath, logic, component location?
RO question Can the question be answered solely by knowing immediate operator actions?
Yes RO quesfion qNo Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?
Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mie strategy of a procedure?
Does the question require one or more of the following?
Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SROonly Q
Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55A3(b)(5) for SRO-only
QUESTION 85 Unit I was operating at 100% Reactor Power when a Reactor Scram occurred.
Following the Scram a primary system began discharging into Secondary Containment, with the following alarms and indications:
RX BLDG AREA RADIATION HIGH, (l-9-3A, Window 22)
RX BLDG, TURB BLDG, RF ZONE EXH RADIATION HIGH, (l-9-3A, Window 4)
Elevation 565 East ARM meter indicating off-scale high Elevation 565 Northeast ARM meter indicating 600 mr/hr and stable Which ONE of the following completes the statement?
In accordance with I -EOI-3, Secondary Containment Control, the crew is required to enter (I)
AND a potential isolation source is (2)
A. I-EOI-I, RPV Control B.
I-EOI-1, RPV Control C. 0-EOI-4, Radioactivity Release Control D. 0-EOI-4, Radioactivity Release Control SDV vents and drains RWCU suction and return isolation valves SDV vents and drains RWCU suction and return isolation valves (1)
(2)
Answer: A
Level:
1 Group#
2 K/A#
295033EA2.03 Importance Rating 4.2 Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Cause of high area radiation Explanation: A CORRECT: First Part: Correct, EOI-I is correct to enter. Step SC/R-2 is answered YES, 1050 mr/hr is> MAX SAFE. Second Part: Correct, SDV valves are correct source choice, with rad levels elevated on both the east / west sides.
B Incorrect First Part: Correct, EOI-l is correct to enter, but with info given on Hi rads on both east /
west, rules out RWCU (69-1, 2 12). They are ONLY applicable to the west side alarm. Step SC/R-2 is answered YES, 1050 mr/hr is> MAX SAFE. Second Part: Incorrect, The 565 elevation Northeast has no possible isolation sources listed in EOI-3 table 4.
C Incorrect First Part: Incorrect, ONLY one MAX SAFE has been exceeded. EOI-4 is entered based on off-site dose. No indications are given as to indications of exceeding any. Second Part:
Correct, SDV valves are the correct source choice.
D Incorrect First Part: Incorrect, Given conditions indicate that there is ONLY one source> MAX SAFE, EOI-4 is entered based on off-site dose. No indications are given as to indications of exceeding any. Second Part: Incorrect, RWCU (69-1, 2 12) are NOT a possible leakage source based on given alarm locations, only on west side (no reference to rad alarms given in stem).
Technical Reference(s)
I -EOI-1, 1 -EOI-3 Proposed references to be provided to applicants during examination: None Learning Objective (As available):
Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: BFN 0801 #85 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis:
X 10 CFR Part 55 Content:
55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
BFN Panel 93 1-ARP-9-3A Unit I XA-55-3A Rev. 0042 Page 33 of 52 Sensor/Trip Point:
RX BLDG AREA RADIA11ON Rl-90-4A 10 MR/HR Rl-90-24A 10 MR/HR 1-RA-SO-1D Rl-90-8A 10 MR/HR R90-25A 100 MR/HR Rl-90-9A 10 MR/HR Rl-90-26A 30 MR/HR Rl-90-13A 10 MR/HR Rl-90-27A 30 MR/HR (Page 1 of 2)
Rl-90-14A 10 MR/HR Rl-90-28A 60 MR/HR Rl-90-20A 40 MR/HR R[-90-29A 110 MR/HR Rl-90-21A 80 MR/HR Rl-90-22A 1500 MR/HR Rl-90-23A 10 MR/HR R1-90-23A 10 MR/HR SENSOR 1-RE-090-0004 MG Set Area Rx Bldg. El 639, R-5 S-Line LOCATION:
1-RE-090-0008 Main Control Room, Rx Bldg. El 617, R-7 P-Line 1-RE-090-0009 Clean-up System, Rx Bldg. El 621, R-6 T-Line 1-RE-090-0013 North Clean-up Sys, Rx Bldg. El 593, R-6 P-Line 1-RE-090-0014 South Clean-up Sys, Rx Bldg. El 593, R-6 S-Line 1-RE-090-0020 CRD-HCU West, Rx Bldg. El 565, R-2 R-Line 1-RE-090-0021 CRD-HCU East, Rx Bldg. El 565, R-6 R-Line 1-RE-090-0022 Tip Room, Rx Bldg. El 565, R-5 P-Line 1-RE-090-0023 Tip Drive, Rx Bldg. El 565, R-5 P-Line 1-RE-090-0024 HPCI Room, RX Bldg. El 519, R-1 U-Line 1-RE-090-0025 RHR West, Rx Bldg. El 519, R-2 U-Line 1-RE-090-0026 Core Spray-RCIC, Rx Bldg. El 519, R-3 U-Line 1-RE-090-0027 Core Spray, Rx Bldg. El 519, R-6 U-Line 1-RE-090-0028 RHR East, Rx Bldg. El 519, R-6 U-Line 1-RE-090-0029 Suppression Pool, Rx Bldg. El 519, R-5 U-Line Probable A. Radiation levels have risen above alarm point.
Cause:
B.
Dry Cask Storage activities in progress (activities could affect rad levels sensed by 1-RE-090-0004, 1-RE-090-0009, 1-RE-090-0014, 1-RE-090-0021).
NOTE Due to location of the Rad Monitor in relation to the test line in the HPCI quad, the HPCI Room rad alarm may be received when the HPCI flow test is in progress.
C. HPCI Flow Rate Surveillance in progress.
Continued on Next Page
Table 4 Secondary Cntmt Area Radiation Applicable Max Max Potential Area Radiation Normal Safe Isolation Indicators Value mRItw Value mR/hr Sources RHR sys I pumps90-25A Alarmed 1000 FCV-74-47. 48 RHR sys II pumps90-28A Alarmed 1000 FCV-74-47. 48 HPCI room 9D-24A Alarmed 1000 FCV-73-2, 3, 44, 81 CS sys I pumps90-26A Alarmed 1000 FCV-71-2, 3, 39 RCIC room CSsys II pumps90-27A Alarmed 1000 None FCV-73-2, 3, 81 Top of torus90-29A Alarmed 1000 FCV-74-47, 48 General area FCV-71-2, 3 RB el 565 W 90-20A Alarmed 1000 FCV-69-1, 2, 12 SDV vents & drains RB el 565 E 90-21 A Alarmed 1000 SOy vents & drains RB el 565 NE 90-23A Alarmed 1000 None TiP room 90-22A Alarmed 100,000 TIP ball valve RB el 593 90-13A, 14A Alarmed 1000 FCV-74-47, 48 RB el 621 90-9A Alarmed 1000 FCV-43-13, 14 Recirc MG sets 90-4A Alarmed 1000 None Refuel floor 90-IA, 2A 3A Alarmed 1000 None
1 L
ISOLATE all systems that are distharging into the area EXCEPT systems required:
For damage coned!
OR To be operated by ECIs SC!R-3 GCt-4 L
L L
Figure 2: Screening for SRO-only linked to 10 CFR 5&43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e. how the system works Yes flowpath, logic, component location?
RO question Can the question be answered solely by knowing immediate operator actions?
Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question eqiiire entry to major EOPs?
Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RD question llitiative strategy of a procedure?
Does the question require one or more of the following?
C Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed 3
Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
)
Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to to CFR 55.43(b)(5) for SRO-only
BFN 0801 #85 4
S Examination Outline Cross-reference Level RO R
0 295033 High Secondary Containment Area 4
Radiation Levels I 9 Tier #
- /
EA2 03 (10 CFR 55 43 5 - SRO Only) 2 Ability to determine andlor interpret the following Group #
as they apply to HIGH SECONDARY 295033EA2.
CONTAINMENT AREA RADIATION LEVELS:
KIA #
03 o
Cause of high area radiation Importance 4
Rating 2
Proposed Question: #
85 Unit I was operating at 100% Reactor Power when an inadvertent Reactor Scram occurred.
Following the Scram a primary system began discharging into Secondary Containment, with the following alarms and indications:
RX BLDG AREA RADIATION HIGH, (1-9-3A, Window 22)
RX BLDG, TURB BLDG, RF ZONE EXH RADIATION HIGH, (1-9-3A, Window 4)
Elevation 565 East ARM meter indicating off-scale high Elevation 565 Northeast ARM meter indicating 600 mrlhr and stable Which ONE of the following completes the statement?
In accordance with I -EOI-3, Secondary Containment Control, the crew is required to enter _(1)_ AND a potential isolation source is _(2)_?
_(1)_
_(2)_
A.
I -EOI-1, RPV Control, FCV 69-1, 2, 12.
B.
0-EOI-4, Radioactivity Release Control, FCV 69-1, 2, 12.
C.
I -EOI-1, RPV Control, SDV vents and drains.
D.
0-EOI-4, Radioactivity Release Control, SDV vents and drains.
QUESTION 86 Unit 2 was shutdown due to RHR Loop II being inoperable for 6 days.
RHR Loop II is tagged out for maintenance on the 2-FCV-74-66, Outboard LPCI Injection Valve RHR Loop I is in Shutdown Cooling with RHR pump 2A Which ONE of the following completes both statements below?
In accordance with Technical Specification 3.5.2, ECCS Shutdown, for these plant conditions, (1) low pressure ECCS injectionlspray subsystem(s) is(are) required to be Operable.
While in Shutdown Cooling, RHR pump 2A (2) be considered an Operable ECCS subsystem.
A. (l)one (2) can B.
(1)one (2) CANNOT C. (l)two (2) can D. (l)two (2) CANNOT ANSWER:
C
Level:
2 Group#
1 Examination Outline Cross-Reference KIA#
203000 G2.2.38 Importance Rating j 4.5 203000 Residual Heat Removal /Low Pressure Coolant Injection: Injection Mode (Plant Specific)
Knowledge of conditions and limitations in the facility license.
Explanation: C CORRECT: 3.5.2 BASES requires 2 pumps. It also provides direction on the ability to align (manual / remote) and still maintain operability. Additionally 90 psig is less than the permissive for SDC, therefore it is permitted per TSs.
A-lncorrect. First Part: Incorrect. Plausible because the candidate may be aware of that one subsystem can be in SDC in TS 3.5.2, and therefore may conclude that only one subsystem is required for LPCI. Second Part: Correct.
B-Incorrect. First Part: Incorrect. Plausible because the candidate may be aware of that one subsystem can be in SDC in TS 3.5.2, and therefore may conclude that only one subsystem is required for LPCI. Second Part: Incorrect.
D-Incorrect. First Part: Correct. Second Part: Incorrect. Plausible because 3.5.2 BASES provides direction on the ability to align (manual / remote) and still maintain operability.
Technical Reference(s): TS 3.5.1, TS 3.5.2 Proposed references to be provided to applicants during examination: None Learning Objective (As available):
Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: None Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis :
X 10 CFR Part 55 Content:
55.41 (2) Facility operating limitations in the TS and their bases.
ECCS - Shutdown 3.5.2 3.5 EMERGENCY CORE COOUNG SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.2 ECCS
- Shutdown LCD 3.5.2 Two low pressure ECCS injection/spray subsystems shall be OPERABLE.
APPLICABILITY:
MODE 4, MODE 5, except with the spent fuel storage pool gates removed and water level 22 ft over the top of the reactor pressure vessel flange.
- Shutdown B 35.2 BASES (continued)
LCO Two low pressure ECCS injectionlspray subsystems are required to be OPERABLE. The low pressure ECCS injection/spray subsystems include CS subsystems and LPCI subsystems. Each CS subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool to the reactor pressure vessel (RPV). Each LPCI subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool to the RPV.
In MODES 4 and 5, the LPCI crosstie valve is not required to be closed. The necessary portions of the Emergency Equipment Cooling Water System are also required to provide adequate cooling to each required ECCS subsystem.
An LPCI subsystem may be aligned for decay heat removal and considered OPERABLE for the ECCS function, if it can be manually realigned (remote or local) to the LPCI mode and is not otherwise inoperable. Because of low pressure and low temperature conditions in MODES 4 and 5, sufficient time will be available to manually align and initiate LPCI subsystem operation to provide core cooling prior to postulated fuel uncovery.
Clarification Guidance for SRO-only Questions Rev 1 (03111/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing 1
hour TSITRM Action?
Can question be answered solely by knowing the LCO/TRM information listed ab:ove-the_line?
Can question be answered solely by knowing the TS Safety Limits?
Noj Yes RO question Yes RO question Yes RO question Does the question involve one or more of the following for TS, TRM, or QDCM?
Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
Application of generic LCO requirements (LCO 30 I thru 301 and SR 40:l thru 404)
O Knowledge of TS bases that is required to analyze TS required actions and terminology No SROonly question Question might not be linked to 10 CFR 5543(h)(2) for SRO-only
QUESTION 87 Unit 3 is at 40% power with power ascension in progress with the following conditions:
Core flow is 55 Mlbm/hr APRM #3 has failed downscale and is bypassed Subsequently, Recirc Loop A flow transmitter input to APRM #1 fails UPSCALE Which ONE of the following (1) describes the plant response and (2) the required operator action(s)?
(REFERENCE PROVIDED]
A.
(1) Rod Block ONLY (2) Enter LCO 3.3.1.1 Condition A ONLY.
B.
(1) Rod Block ONLY (2) Enter LCO 3.3.1.1 Conditions A and B.
C.
(1) Rod block and an Upscale trip input to ALL 4 voters (2) Enter LCO 3.3.1.1 Condition A ONLY.
D.
(1) Rod block and an Upscale trip input to ALL 4 voters (2) Enter LCO 3.3.1.1 Conditions A and B.
ANSWER:
A
Level:
2 Group#
1 Examination Outline Cross-Reference K/A#
215005 A2.05 Importance Rating f 3.6 Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONTTORILOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of recirculation flow signal Explanation: A CORRECT: Per T.S. table 3.3.1.1-1, 3 APRMs are required. With one less than the required number an APRM must be restored within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or the channel placed in a tripped condition. A rod block will occur due to >10% difference in total recirc flow signals from the APRMs. APRM #1 summer will have an output> 75%, Other APRM flow summers will indicate approximately 50% at this core flow.
B-Incorrect. First Part: Correct. Incorrect. An APRM Upscale trip will NOT be generated on HIGH recirc flow. The Recirc flow transmitter failing upscale will cause the flow biased scram setpoint to increase.
Second Part: Incorrect. LCO 3.3.1.1 B does not apply because it does not affect both trip systems.
C-Incorrect. First Part: Incorrect. An APRM Upscale trip will NOT be generated on HIGH recirc flow.
The Recirc flow transmitter failing upscale will cause the flow biased scram setpoint to increase. Second Part: Correct.
D-Incorrect. First Part: Incorrect. An APRM Upscale trip will NOT be generated on HIGH recirc flow. The Recirc flow transmitter failing upscale will cause the flow biased scram setpoint to increase. Second Part: Incorrect. LCO 3.3.1.1 B does not apply because it does not affect both trip systems.
Technical Reference(s): TS LCO 3.3.1.1, OPL17I.007, 3-ARP-9-5A window 7 Proposed references to be provided to applicants during examination: Unit 3 TS 3.3.1.1 Learning Objective (As available):
Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: Nine Mile 2 2010 #88 Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (2) Facility operating limitations in the TS and their bases.
RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.
APPLICABILITY:
According to Table 3.3.1.1-1.
ACTIONS NOTE Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> channels inoperable.
OR A.2 NOTE Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.
Place associated trip 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> system in trip.
(continued)
Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.
One or more Functions with one or more required channels inoperable in both trip systems.
One or more Functions with RPS trip capability not maintained.
RPS Instrumentation 3.3.1.1 ACTIONS (continued)
B.
NOTE B.1 CONDITION REQUIRED ACTION COMPLETION
- TIME Place channel in one trip system in trip.
OR B.2 Place one trip system in trip.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 6 hours c.
0.1 Restore RPS trip capability.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> D. Required Action and Di Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, or Table 3.3.1.1-1 for the C not met.
channel.
E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Action 0.1 and POWER to < 30% RTP.
referenced in Table 3.3.1.1-1.
F As required by Required F.1 Be in MODE 2.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action 0.1 and referenced in Table 3.3.11-i (continued)
Table 3.3.11-1 (page 1 of3)
Reactor Protection System Instrumentation RPS Instrumentation 331.1 APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION Dl I
Intermediate Range Monitors a.
NeulronFlux
- High 2
3 G
SR 3.3.1.11 120!125 SR 3.31.1.3 divisionsoffull SR 33.11.5 scale SR 3.3.1.1.6 SR 33.1.1.9 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.
mop 2
3 13 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.
NeutronFlux
- 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 I
3 (b)
F SR 3.3.1.1.1 0.66W Thermal Power
- High SR 3.3.1.1.2
+ 66% RTP SR 3.3.1.1,7 andI20%
SR 3.3.1.1.13 RTP(C>
SR 3.3.1.1.16 c.
Neutron Flux
- High 1
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 (contmued)
(a>
With any control rod withdrawn from a care cell containing one or more fuel assemblies.
(b)
Each APRM channel provides inputs to both trip systems.
(c)
[.136 W + 66%
-.66 i W] RTP when reset For single loop operation per LCO 3.4.1. Recirculation LOOpS Operating.
OPLI7I.007 2.
Average Power Range Monitor (APRM)/Rod Block Monitor (REM)
System.
a.
The Recirculation Loop flow elements (68-5, 68-81) provide the flow signals to the APRM for flow biased setpoint ILT Objective 19b determination.
LOR Objective 9c b.
The APRM flow signals are then sent to the Rod Block Monitor, where they are evaluated for differences in flow between APRM drawers (5% disparity will result in flow compar& alarm on panel drawer).
c.
Failure of one of the two flow transmitters at high power will result in a reduction in the measured flow, with possible rod block or scram signals resulting.
BFN Panel 9-5 3-ARP-9-5A Unit 3 3-XA-55-5A Rev. 0043 Page 11 of 47 Sensor/Trip Point:
CONTROL ROD Relays:
3A-K1 Nuclear Instrumentation WITHDRAWAL 3A-K2 Refuel Equipment in Use BLOCK High Level in Scram Discharge Volume FT Scram Discharge Volume High Water Level Bypass (Page 1 of 2)
Rx Mode Switch in SHUTDOWN PRNM (ANY APRM OPRM or RBM)
Sensor Panel 3-9-28 Location:
Elevation 593 Aux lnstr Room Probable A. One or more sensors at or above set point.
Cause:
B.
Malfunction of sensor.
C. Control rod drop accident.
Automatic Rod withdrawal block.
Action:
Operator A.
DETERMINE initiating condition from corresponding rod withdrawal Action:
block alarm(s) and REFER TO operator action for alarm(s).
El B.
IF alarm due to inadvertent criticality during incore fuel movements, THEN REFER TO 3-AC1-79-2.
El C.
IF alarm is from a control rod drop, THEN REFER TO 3-AC1-85-1.
El D.
IF NO corresponding alarm exists. THEN 1.
AT ICS console. DETERMINE if there is a refuel rod block by selecting single Point menu. Single Value display, and typing C602, Return.
El 2.
IF rod block was from Refuel Floor, THEN CALL Refuel Floor Operator to have dummy plug (Refuel floor between cavity and pool, south side) checked and check jumpers in U-3 Aux Inst. Rm, Panel 3-9-28 Bay 3, if Installed per 3-Ol-85 Section 8.34.
El 3.
WHEN IRM switches are below Range 3 with REACTOR MODE SWITCH NOT in RUN, THEN CHECK SRM detectors NOT FULL IN.
El 4.
WHEN REACTOR MODE SWITCH is in START-UP position.
THEN CHECK IRM detectors NOT FULL IN.
El Continued on Next Page
Nine Mile 22010 #88 Nine Mile Point Unit 2 2010 NRC SRO Written Examination Facility:
Nine Mile Point Unit 2 Vendor:
GE Exam Date:
2010 Exam Type:
S Examination Outline Cross-reference:
Level SRO Tier#
2 Group#
1 KIA#
215005 A2.05 Importance Rating 3.6 Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of recirculation flow signal Question:
SRO #88 The plant is at 40% power with a power ascension in progress. Core flow is 55 MIbm/hr.
APRM #3 has failed downscale and is bypassed.
With these initial conditions, the Recirc Loop A flow transmitter inputting to APRM #1 fails UPSCALE.
Which one of the following describes the plant response and required operator action(s)?
A.
Rod block and a % scram on RPS K.
Enter LCO 3.3.1.1 A. No additional action is required for RPS due to existing V2 scram.
B.
Rod Block ONLY.
Enter LCO 3.3.1.1 A.
If an inoperable APRM channel is not restored within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, insert an APRM upscale trip.
C.
Rod Block ONLY.
Enter LCO 3.3.1.1 C. If an inoperable APRM channel is not restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, insert an APRM upscale trip.
0.
Rod block and a Y2 scram on RPS A.
Enter LCO 3.3.1.1 B. No additional action is required for RPS due to existing % scram.
Answer:
B
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO..only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing the Yes LCOITRM information listed above-the-line?
RO question Yes j
SRO-only question Can question be answered solely by knowing I
Yes hour TS/TRM Action?
RO question Can question be answered solely by knowing the TS Safety Limits?
Yes RO question Does the question involve one or more of the following for TS, TRM, or ODCM?
6 Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
Application of generic LCO requirements (LCO 3O 1 thru 101 and SR 401 thru 404)
O Knowledge of TS bases that is required to analyze TS required actions and terminology No Question might not be linked to 10 CFR 55A3(b)(2) for SRQ-only
QUESTION 88 Given the following plant conditions:
Unit I is starting up currently in Mode 2 Unit 2 is operating in Mode I o
Unit 3 is in Mode 5 with fuel movement in progress Diesel Generator 3ED is out of service A Unit Operator reports that while attempting to start SGT Fan B that the fan will NOT start.
Which ONE of the following completes the statement below?
In accordance with Tech Spec 3.6.4.3, Standby Gas Treatment System,
[REFERENCE PROVIDED]
A. fuel movement can continue on Unit 3. Unit I AND 2 operation is permitted for 7 days B. fuel movement can continue on Unit 3. Enter LCO 3.0.3 on Unit I AND 2 in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. Unit 2 must be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Unit 3 fuel movement AND Unit I startup can continue for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. IMMEDIATELY suspend fuel movement on Unit 3. IMMEDIATELY enter LCO 3.0.3 on Unit I AND 2 Answer: B
Level:
RD SRO Tier#
2 Group#
1 Examination Outline Cross-Reference KIA#
261 000G2.2.4 Importance Rating 3.6 261000 SGTS G2.2.4 (multi-unit license) Ability to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility.
Explanation: CORRECT B With 3ED DIG inoperable, when B SBGT blower fails to start, the redundant component requirement comes into effect. Therefore LCO 3.8.1.B.3 allows 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to declare the C train of SBGT Inoperable and then enter 3.0.3.
A Incorrect: Two SBGT trains are Inoperable, and requires entry into TS 3.0.3. Plausible because this is the action for one SBGT Inoperable TS 3.6.4.3 A.1.
C Incorrect. Units 1 & 2 have 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to enter TS 3.0.3. This is plausible if LCO 3.8.1.B.3 completion time (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to declare the C train of SBGT Inoperable) is missed and TS 3.0.3 is entered immediately.
D Incorrect. Suspension of fuel handling on Unit 3 is not required. This is plausible if on Units I and 2 the LCO 3.8.1.B.3 completion time (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to declare the C train of SBGT Inoperable) is missed and TS 3.0.3 is entered immediately.
Technical Reference(s): 0-01-65, Unit 1/2/3 TS 3.6.4.3 Proposed references to be provided to applicants during examination: Unit 1/2/3 TS 3.6.4.3 & 3.8.1 (NO BASES)
Learning Objective (As available):
Question Source:
Bank:
Modified Bank:
X New:
Question History:
Previous NRC: BFN 1006 #88 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis: X 10 CFR Part 55 Content:
55.43 b (2) Facility operating limitations in the technical specifications and their bases.
SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Three SGT subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3.
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable, to OPERABLE status.
B. Required Action and 8.1 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 of Condition A not met in MODE 1,2, or 3.
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 /> (continued>
BFN-UNIT 1 3.6-51 Amendment No. 234, 251 September 27, 2004
SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Three SGT subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2. and 3.
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUiRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable, to OPERABLE status.
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 of Condition A not met in MODEl, 2, or3.
8.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
BFNUNIT 2 3.6-51 Amendment No. 23 290 September 27, 2004
3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Three SGT subsystems shall be OPERABLE.
SGT System 3.6.4.3 APPLICABILITY:
MODES 1. 2, and 3, During operations with a potential for draining the reactor vessel (0PDRVs )
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable, to OPERABLE status.
- 8. 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 of Condition A not J1 met in MODE 1. 2, or 3.
8.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 3.6-51 Amendment No. 242, 249 September 27, 2004 (continued)
BFN-UNIT 3
SGT System 3.6.4.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Place two OPERAELE Immediately associated Completion SGT subsystems in Time of Condition A not operation.
met during OPDRVs.
OR C.2 Initiate action to suspend Immediately OPDRVs.
D. Two or three SGT Di Enter LCO 3.0.3.
Immediately subsystems inoperable in MODE 1, 2, or 3.
(continued)
AC Sources - Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources
- Operating LCO 3,8.1 The following AC electrical power sources shall be OPERABLE:
- a. Two qualified circuits between the :oftsite transmission network and the onsite Class I E AC Electrical Power Distribution System; b.
Unit 3 diesel generators (DOs) with Iwo divisions of 480 V load shed logic and common accIdent signal logic OPERABLE; and c.
Unit I and 2 00(5) capable of supplying the Unit I and 2 4.16 kV shutdown board(s) required by LCO 3.8.7, TM Dishibution Systems
- Operating.
APPLIIDABILITY:
MODES 1,2, and 3.
ACTIONS LCO 3.0.4.b is not applicable to DOs.
CONDITION REQUIRED ACTION COMPLETION TIME A. One required ofisite A.1 Verity power availability 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> circuit inoperable.
from the remaining OPERABLE oftsite AND transmission networt Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter AND (continued)
BFN-UNIT 3 3,8-1 Amendment No. ZI2T 244 December 1, 2003
AC Sources
- Operating 38.1 ACTIONS CONDITON REQUIRED ACTION COMPLETION TIME A. (contInued)
A.2 Declare required 24 flours from feature(s) with no offsite discovery of no power available offsite power to inoperable when the one shutdown redundant required board concurrent feature(s) are inoperable, with inoperability of redundant required feature(s)
AND A.3 Restore required offsite 7 days circuit to OPERABLE status.
PJJD 21 days from discovery of failure to meet LCO B.
One required Unit 3 DG B. 1 Verify power availability 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable, from the offsite transmission network.
AND Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter AND (continued)
BFN-UNIT 3 3.8-2 Amndment No. 266
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued)
B.2 Evaluate availability of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> both temporary diesel generators (TOGs).
AND AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter B.3 Declare required 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from feature(s), supported by discovery of the inoperable Unit 3 0G.
Condition B inoperable when the concurrent with redundant required inoperability of feature(s) are inoperable, redundant required feature(s)
AND B.4.1 Determine OPERABLE 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Unit 3 DG(s) are not inoperable due to common cause failure.
OR B.4.2 Perform SR 3.81.1 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE Unit 3 OG(s).
AND (continued)
BFN-.UNIT 3 3.8-3 Amendment No. 266
Clarification Guidance for SRO-only Questions Rev 1 (0311112010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing I
Yes hour TS/TRM Action?
RO question Can question be answered solely by knowing the Yes LCOITRM information Fisted above-the-line?
RO question
Can question be answered solely by knowing the Yes TS Safety Limits?
RO question Does the question involve one or more of the following for TS, TRM, or ODCM?
Application of Required Actions <Section 3) and Surveillance Requirements <Section 4) in accordance with rules of application requirements (Section 1)
Application of generic LCO requirements (LCO 301 thru 3M7 and SR 4.O 1 thru 4O4)
Yes SRO-only Knowledge of TS bases that is required to analyze IS question required actions and terminology No Question might not be linked to 10 CFR 5543(b)(2) for SRO-only
BFN 1006 #88 Examination Outline Cross-reference:
261000 SGTS 02.2.4 (IOCFR 55.43.5
- SRO Only)
(multi-unit license) Ability to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility.
C.
(1)
Unit3 (2) the OPDRV must be suspended immediately D.
(1)
Unit3 (2)
SGTS A must be restored to operable within 7 days Proposed Answer:
B A
INCORRECT: Part I correct See Explanation B. Part 2 incorrect See Explanation C.
SRO 2
L Level Tier#
Group#
K/A#
Importance Rating 261000G2.2.4 3.6 Proposed Question: # 88 Given the following plant conditions:
Unit I AND Unit 2 are at 100% Reactor Power Unit 3 is in Mode 5 with an Operation with Potential to Drain the Vessel (OPDRV) in progress The Unit 3 Unit Supervisor directs starting of ALL Standby Gas Treatment Subsystems (SGTS).
Which ONE of the following completes the statements?
The PREFERRED location in accordance with 0-01-65, Standby Gas Treatments System, is to start SGTS from the (1)
Control Room.
SGTS A trips following manual start. In accordance with Tech Spec 3.6.4.3, Standby Gas Treatment System, _(2).
[REFERENCE PROVIDED]
A.
(I)
Unit I ORUnit2 (2) the OPDRV must be suspended immediately B.
(1)
Unit I OR Unit 2 (2)
SGTS A must be restored to operable within 7 days Explanation (Optional):
B CORRECT: Part 1 correct
- Per 0-01-65, Standby Gas Treatments System, Precaution and Limitation
- Although all three trains of the SGT System can be started from the Unit 3 Control Room, it is recommended that the trains be started from the Units I and 2 Control Rooms due to the availability of instrumentation and shutdown capability. Part 2 correct
Although TS 3.6.4.3 is not applicable for current conditions on Unit 3, it is applicable for Units I and 2. In accordance with Unit I and 2 TS 3.6.4.3 Condition A, SGTS A must be restored within 7 days.
C INCORRECT: Part I incorrect Plausible in that Unit 3 controls are available to start SGTS in the Unit 3 Control Room. Part 2 incorrect
Plausible in that an OPDRV is in progress on Unit 3. However, it is not required immediately.
D INCORRECT: Part I incorrect See Explanation C. Part 2 correct See Explanation B.
QUESTION 89 Unit 3 is at 100% power and the following condition exists:
Unit Battery 3 is on a float charge when a loss of ventilation to Battery Room 3 occurs.
Unit 3 Battery Room Temperature is 79° F and rising.
Which ONE of the following completes both statements below?
The associated battery is (I)
The Unit Supervisor should direct use of temporary ventilation in accordance with (2)
A. (1) OPERABLE (2) 0-O1-30F, Common and DG Building Ventilation B. (1) OPERABLE (2) 0-01-3 1, Control Bay and Off-Gas Treatment Building Air Conditioning System C. (1) INOPERABLE BUT Available (2) 0-0I-30F, Common and DO Building Ventilation D. (1) INOPERABLE BUT Available (2) 0-01-3 1, Control Bay and Off-Gas Treatment Building Air Conditioning System ANSWER:
B
Level:
2 Group#
I Examination Outline Cross-Reference K/A#
263000 A2.02 Importance Rating 2.9 Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of ventilation during charging Explanation: B CORRECT: The batteries are Operable until additional criteria are reached (low volts, Hi temps). 0-01-31 gives reference to temporary ventilation setup.
A-Incorrect. First Part: Correct. Second Part: Incorrect. Plausible because 0-Ol-30F contains the procedure sections for controlling battery Room 3EB ventilation.
C-Incorrect. First Part: Incorrect. The batteries are Operable until additional criteria re reached (low volts, Hi temps). Plausible because it makes sense that a loss of ventilation lineup would require LCO entry.
Second Part: Incorrect. Plausible because 0-OI-30F contains the procedure sections for controlling battery Room 3EB ventilation.
D-Incorrect. First Part: Incorrect. The batteries are Operable until additional criteria are reached (low volts, Hi temps). Plausible because it makes sense that a loss ofventilation lineup would require LCO entry.
Second Part: Correct.
Technical Reference(s): OPLI7I.037, 0-01-31, TRM 3.7.6 Proposed references to be provided to applicants during examination: None Learning Objective (As available):
Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: None Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis :
X 10 CFR Part 55 Content:
55.41 (2) Facility operating limitations in the TS and their bases.
Electric Board Room AC System B 3.7.6 TR 3.7 PLANT SYSTEMS TR 3.7.6 Electric Board Room Air Conditioning (AC) System BASES BACKGROUND The Unit 3 Electric Board Room AC System provides temperature control for the two Unit 3 Electric Board Rooms on elevations 593 and 621 of the reactor building for both normal operation as well as accidents and plant transients. The Unit 3 Electric Board Room AC System consists of two redundant subsystems that provide cooling of recirculated electric board room air. A subsystem consists of an air handling unit, a condensing unit, ductwork, dampers, piping, and instrumentation and controls to provide for electric board room temperature control. Heat removed by the Electric Board Room AC System is transferred to the environment via the Emergency Equipment Cooling Water System.
A single subsystem provides the required temperature control to maintain the reactor building 593 and 621 elevation electric board rooms temperature within acceptable limits for operation of equipment and for uninterrupted safe occupancy under all plant conditions. The design conditions for the electric board room environment are 104°F and 80% relative humidity. Each subsystem is capable of maintaining the electric board rooms temperature at or below 104°F during abnormal or accident conditions. Alternate methods of cooling the Unit 3 Electric Board Rooms are available. These include, but are not limited to, the use of fans in the doors to the electric board rooms coupled with the Unit 1/2 and Unit 3 Control Room AC Systems and the Relay Room AC Systems. The Electric Board Room AC System operation in maintaining the electric board rooms temperature is discussed in the FSAR, Section 10.12 (Ref.1).
[CO Two redundant subsystems of the Unit 3 Electric Board Room AC System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. Total system failure could result in the equipment operating temperature exceeding limits unless alternate methods of cooling are implemented. Equipment in the room is declared inoperable whenever the temperature in the room exceeds 104°F.
BFN Control Bay and Off-Gas Treatment 0-Ol-31 Unit 0 Building Air Conditioning System Rev. 0142 Page 171 of 285 8.15 Temporary Ventilation for Electrical Equipment Rooms NOTES 1)
This instruction contains the temperature requirements and methods for monitoring and supplying temporary ventilation due to the loss or malfunction of HVAC system to one or more of the following rooms:
UI/U2 Main Control Room U3 Main Control Room Ui, U2, U3 Auxiliary Instrument Rooms Ui, U2, U3 125V/250V Battery and Battery Board Rooms a
UI, U2. U3 Unit Preferred MG Set Rooms Ui, U2, U3 RPS MG Set Rooms U3 SHUTDOWN BOARD ROOMs 3EA, 3EB, 3EC, 3ED Ui, U2 250V Shutdown Battery Rooms a
Relay Room Cable Spreading Room 2)
MSI-O-000-PR0005, ELECTRICAL EQUIPMENT ROOM EMERGENCY VENTILATION FOLLOWING AN APPENDIX R FIRE EVENT contains additional guidance for compensatory measures to ensure operability of electrical equipment following an Appendix R fire.
3)
Temporary portable fans and generators with necessary equipment for placement are located in the T-Warehouse. Row 6
[1]
IF ventilation is lost to Battery Room(s), THEN (Otherwise NA)
PERFORM (explosive gas) monitoring as follows:
C
[1.1]
MONITOR for H2 buildup daily, THEN RECORD on Illustration 2, Battery Room Atmosphere Readings, (detector(s) can be obtained from Fire Operations Personnel).
C
7.
Battery Room Ventilation Systems NLO/NLQRObJ 8
- a. Purpose FUND The various battery room ventilation systems provide adequate room ventHation to prevent an explosive atmosphere due to hydrogen buildup from the batteries.
b.
The Unit Battery Rooms 1, 2 and 3 and the Communications Battery Room are supplied air through the door ventilators. Air is exhausted with Battery and Board Room Exhaust Fans 1A and 18 (Battery Room 1 and 2, and communications battery room), and Unit 3 Battery and Board Room Exhaust Fans 3A and 3B (Battery Room 3). Plant/Station Battery Rooms are supplied air via an HVAC unit located outside the rooms to maintain an optimum temperature between 70 and 80 degrees F. A small exhaust fan is located in the ceiling with an off and on switch located on the wall (speed is variable). The purpose of the exhaust fan is to keep hydrogen concentration below 2%. With the exhaust fan off it will take over 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to reach the design limit of 2% hydrogen. Upon loss of the exhaust fan, a system abnormal will alarm in the control room. The ceiling also has vent pipes to exhaust the flow of air. Battery Room 4 also required the installation of a new bypass damper with the existing ventilation fan to maintain hydrogen concentration below the design limit.
Electric Board Room AC System TR 3.7.6 TR 3.7 PLANT SYSTEMS TR 3.7.6 Electric Board Room Air Conditioning (AC) System LCO 3.7.6 Two Unit 3 Electric Board Room AC subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1,2, 3, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One Unit 3 electric A.1 Restore Unit 3 electric 30 days board room AC board room AC subsystem inoperable, subsystem to OPERABLE status.
(continued)
Electric Board Room AC System TR 3.7.6 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B.
Two Unit 3 electric 3.1 Initiate action to restore Immediately board room AC one Unit 3 electric board subsystems room AC subsystem to inoperable.
OPERABLE status.
AND 3.2 Place an alternate 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> method of cooling in operation.
AND B.3 Restore one electric 7 days board room AC subsystem to OPERABLE status.
C.
Required Action and C.1 Declare the electrical 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion equipment in the electric Time of Condition A or board room inoperable.
B not met in MODE 1, 2, or3.
(continued)
Nine Mile 1 2008 #90 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:
2 Group#
1 KJA#
263000 A2.02 Importance Rating 2.9 (K&A Statement) Ability to la) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of ventilation during charging Proposed Question:
SRO 90 The plant is operating at 100% power when Annunciator A3-2-3, LOSS OF VENTILATION BATTERY ROOM 11 OR 12, alarms. The alarm is caused by a failure of the Turbine Building Ventilation to Battery Room 11.
Over the next hour the following events occur:
Hydrogen levels in Battery Room 11 exceed 2.0%
Maintenance recommends securing the battery charger for Battery 11 Which one of the following actions is required?
Secure the in-service battery charger (161A or 161B) and...
A.
line up MG 167 as an alternate battery charger, TS entry is not required.
B.
line up Battery Board 12 to supply Valve Board 11, TS entry is not required.
C.
return a battery charger to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or then enter TS 3.1.5.b which requires RPV pressure <110 psig within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
D.
return a battery charger to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or then enter TS 3.0.1 which requires placing the Mode Switch to SHUTDOWN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Proposed Answer:
C.
Clarification Guidance for SRO-only Questions Rev 1(03111/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs>
Can question be answered solely by knowing I
Yes hour TSITRM Action?
RD question Can question be answered solely by knowing the Yes LCOiTRM information listed above-the-line?
RU question Can question be answered solely by knowing the Yes TS Safety Limits?
RD question Does the question involve one or more of the following for TS, TRM, or 00CM?
Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
Application of generic LCD requirements (LCD 30 1 thru 3O7 and SR 401 thru 404)
SRO-only 0 Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55A3(b)(2) for SRO-only
QUESTION 90 Unit I is Mode 5 with vessel reassembly in progress to the point of installing the Dryer. RHR Pump IA is operating in Shutdown Cooling.
RBCCW Pump I B has tripped RBCCW PUMP SUCTION HDR TEMP has increased to 106° F Spare RBCCW Pump is UNAVAILABLE RWCU System AND the Fuel Pool Cooling System have been shutdown as directed by 1-AOI-70-1, Loss of Reactor Building Closed Cooling Water.
NOTE: 1-01-70, Reactor Building Closed Cooling Water 1-01-74, Residual Heat Removal System Which ONE of the following completes the statements below?
The Fuel Pool temperature limit in TRM 3.9.2 is (1)
In accordance with l-AOI-78-I, Fuel Pool Cleanup System Failure, the Unit Supervisor would direct (2)
A. (1) 125°F (2) placing EECW in Service to the RBCCW Heat Exchangers, lAW 1-01-70 B. (1)125°F (2) initiation of Supplemental Fuel Pool Cooling with RHR Drain Pump B, lAW 1-01-74 C. (1)150°F (2) placing EECW in Service to the RBCCW Heat Exchangers, lAW 1-01-70 D. (1)150°F (2) initiation of Supplemental Fuel Pool Cooling with RHR Drain Pump B, JAW 1-01-74 Answer: B
Level:
2 Group#
1 Examination Outline Cross-Reference KIA#
400000A2.01 Importance Rating I
400000 Component Cooling Water. Ability to (a) predict the impacts of the following on the CCWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of CCW Pump Explanation: B CORRECT: Actions to correct for a loss of fuel pooi cooling, based on the above conditions are contained within 1-AOI-78-1, Fuel Pool Cleanup System Failure, Subsequent Action
[3.7] directs placing RHR in Supplemental Fuel Pool Cooling mode per 1-01-74, as necessary to maintain Fuel Pool temperature less than 125° F. At this time Shutdown Cooling would not be a viable option to cool the spent fuel pool. This is because the fuel pool gates are installed while the reactor cavity is drained for vessel assembly.
A Incorrect: First Part: Correct. Second Part: Incorrect. Plausible in that 1-01-70 will correct fuel pool temperatures for a different type of failure.
C Incorrect. First Part: Incorrect. Plausible because 150°F is the TRM limit. Second Part: Incorrect.
Plausible in that 1-01-70 will correct fuel pooi temperatures for a different type of failure.
D Incorrect. First Part: Incorrect. Plausible because 150°F is the TRM limit. Second Part: Correct.
Technical Reference(s): 1-AOl-70-1, l-A0I-78-1 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OPLI7I.074 V.B.1 Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: BFN 1006 #90 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis : X 10 CFR Part 55 Content:
55.43 b(S) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations
BEN Loss of Reactor Building Closed 1-AOl-b-I Unit I Coaling Water Rev. 0013 Pa :B of 13 4.2 Subsequent Actions continued NOTE Opening RBCCW TCV Bypass valves may cause an EEW pump start due to ow pressure.
J93 IF RBCCW PUMP SUCTION I-IDR TEMP. 1-TIS-70-3 (Panel 1-9), cannot be maintained below 105F, THEN PERFORM the following (otherwise NM):
SLOWLY OPEN bypass valves around RBCCW TCVs..
D e
PLACE Spare RBCCW Heat Exchanger in service in accordance with 1-01-70.
[1 VERIFY RWCU System removed from service.
U NOTE It may be necessary to place the RHR System in the Supplemental Fuel Pool Cooling Mode to maintain Fuel Pool temperature below 150F (Tech. Specs. 3.10.C.2) (TRM 3.9.2).
SHUT DOWN Fuel Pool Cooling Syslern n accordance with 1-01-78.
U 110]
IF RBCCW PUMP SUCTIOF I-ICR TEMP, I -T1S-70-3 iVanel 1-9-4), cannot be maintained below 11 0F. THEN PERFORM the following (otherwise NM):
flOl]
IF core flow is above 60%, 11-lEN REDUCE core flow to between 50-60% (otherwise NM).
U 110.2]
MANUALLY SCRAM the reactor and PLACE Mode Switch to SHUTDOWN.. (REFER TO 1-AOl-I 00-1)
BFN Fuel Poof Cleanup System Failure 1-AOl78-1 Unit I Rev. 0020 Page 14 of 22 42 Subsequent Actions (continued)
[36)
ESTIMATE the time for the Fuel Pool temperature to rise to 125F. I 5OF and 2OOF, using the HeatUp Rates as provided in Attachment I, Table 1 at least once per shift until Fuel Pool Cooling is restored I]
3.7j PLACE RHR Supplemental: Fuel Pool Cooling mode in operation and Refer to 1OlJ4, as necessary to maintain Fuel Pool temperature less than 125°F, as indicated; an RHRIFUEL POOL CLG TEMPERATURE recorder, i-TR-74BO on Panel 921 El
Figure 2: Screening for SRO-only linked to 10 CFR 5&43(b)(5)
(Assessment and selection of procedures)
Can the question be answered so/ely by knowing systems knowledge, i.e. how the system works Yes flowpath, logic, component location?
P0 question Can the question be answered solely by knowing I
immediate operator actions?
Yes P0 question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters es RO question that require direct entry to major EOPs?
Can the q:uestiol-i be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?
Does the question require one or more of the following?
0 Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed 0
Knowledge of when to implement attachments and appendices includina how to coordinate these items with procedure steps Yes SRO-only Knowledge of diagnostic steps and decision points in the question FOPs that involve transitions to event specific sub procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55A3(b)(5) for SRO-only
BFN 1006 #90 Examination Outline Cross-reference:
Level 400000 Component Cooling Water A2.01 (10CFR 55.43.5
- SRO Only)
Ability to (a) predict the impacts of the following on the CCWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Loss of CCW Pump L Proposed Question: # 90 Unit 1 is Mode 5 with vessel reassembly in progress to the point of installing the Dryer. RHR Pump 1A is operating in Shutdown Cooling.
RBCCW Pump I B has tripped RBCCW PUMP SUCTION HDR TEMP has increased to 106°F Spare RBCCW Pump is UISIAVAILABLE RWCU System AND the Fuel Pool Cooling System have been shutdown as directed by 1-AOI-70-1, Loss of Reactor Building Closed Cooling Water.
NOTE: 1-01-70, Reactor Building Closed Cooling Water 1-01-74, Residual Heat Removal System Which ONE of the following completes the statement?
In order to maintain Fuel Pool temperature below the MAXIMUM allowed temperature of (l)_,as established in 1-AOI-78-1, Fuel Pool Cleanup System Failure, the Unit Supervisor is required to enter
_(2)_.
A.
(1) 125°F (2) Section 8.14 Initiation of Supplemental Fuel Pool Cooling with RHR Drain Pump B per 1-01-74 B.
(1) 125°F (2) Section 8.8 Placing EECW in Service to the RBCCW Heat Exchangers per 1-01-70 C.
(1)150°F (2) Section 8.14 Initiation of Supplemental Fuel Pool Cooling with RHR Drain Pump B per 1-01-74 D. (1) 150°F (2) Section 8.8 Placing EECW in Service to the RBCCW Heat Exchangers per 1-01-70 Tier#
Group #
K/A#
Importance Rating SR 0
7 400000A2.0 I 3.4 Proposed Answer:
A
QUESTION 91 Unit I was operating at 75% power when the Reactor Recirc Loop B flow transmitter, I -FT-68-8 IC, input to APRM 3 fails downscale.
Which ONE of the following completes both statements below?
The effect on Rod Block Monitor (RBM) A will be a (1)
The Required Action (if any) in accordance with Control Rod Block Instrumentation Technical Specification 3.3.2.1, if any, is (2)
[REFERENCE PROVIDEDI A.
(1) Flow Compare (inverse video) alarm ONLY (2) no action B.
(1) Flow Compare (inverse video) alarm ONLY (2) restore RBM channel A to OPERABLE status in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C.
(I) Flow Compare (inverse video) alarm AND a Control Rod Block (2) no action D. (I) Flow Compare (inverse video) alarm AND a Control Rod Block (2) restore RBM channel A to OPERABLE status in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Answer: A
Level:
2 Group#
1 ExaminaUon Outline Cross-Reference KIA#
215002 A2.02 Importance Rating I
Ability to (a) predict the impacts of the following on the ROD BLOCK MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss or reduction in recirculation system flow (flow comparator): BWR-3,4,5 Explanation: A CORRECT: The only effect on Rod Block Monitor (RBM) A will be a Flow Compare (inverse video) alarm. Since RBM A primary APRM is APRM I (which gets its Reactor Recirc flow input from FT-68-8 1 A), the RBM remains unaffected and therefore Operable.
B Incorrect. First Part: Correct. Second Part: Incorrect. Plausible because this would be the correct required Action per TS 3.3.2.1 if RBM A was Inoperable.
C Incorrect: First Part: Incorrect. Plausible if the candidate confuses the Reactor Recirc flow inputs to the RBM. Second Part: Correct.
D Incorrect. : First Part: Incorrect. Plausible ifthe candidate confuses the Reactor Recirc flow inputs to the RBM. Second Part: Incorrect. Plausible because this would be the correct required Action per TS 3.3.2.1 if RBM A was Inoperable.
Technical Reference(s): OPLI7I.148, TS 3.3.2.1 Proposed references to be provided to applicants during examination: TS 3.3.2.1 Learning Objective (As available):
Question Source:
Bank:
Modified Bank:
New:
X Question History:
Previous NRC: None Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis : X 10 CFR Part 55 Content:
55.43 b (2) Facility operating limitations in the technical specifications and their bases.
OPL1 71.148 Revision 12 Page 24 of 106 INSTRUCTOR NOTES (4)
The STP is also used by the RBM for reference APRM power level.
g.
An APRM downscale condition (at< 5%)
Obj. V.B.13.a also produces a rod block when the reactor Obj V.D.7a mode switch is in RUN.
h.
Recirculation Flow Monitor Obj. V.B.10, V.D.6 (1>
Each flow monitor channel consists of two flow inputs used to calculate a Total Recirculation Flow, one from Recirculation Loop A and one from Recirculation Loop B.
(2)
Each APRM receives the inputs from Displays on two (4 to 20 mA) differential pressure Input Status (AP) transmitters used to measure display the recirculation loop flows.
(3)
Each RBM receives the four Total Recirculation Flow values from the APRM channels to determine the status of the flow compare alarm.
(4)
The Recirculation Flow Monitor Function provides the following alarm functions for each Total Recirculation Flow level:
- Flow Upscale Alarm (generated by Inputs to flow upscale the APRM) alarm APRM flow alarm is when flow is
- Flow Compare Alarm (generated by
>107 % or upscale, the RBM, display alarm only).
generates rod block.
(5)
The table below shows the Attention to Detail relationship between the flow Right unit/train/comp monitors and each APRM channel.
Loo A
Loo B
Discuss effects if p
p Flow transmitters fail APRM I FT-68-5A FT-68-81A APRM 2 FT-68-5B FT-68-81 B APRM 3 FT-68-5C FT-68-81C APRM 4 FT-68-5D FT-68-81 D
One transmitter fails FT-68-82C If the B recirc loop signal fails Refer to flow for APRM 3, a flow compare transmitter alarm is initiated by the RBM.
assignment failure Since for APRM 3, it is only Human seeing 1/2 of the total reciro Performance
- what flow signal, this exceeds the other instruments flow compare setpoint.
are available for Also the indicator on panel 9-use for backup 4 is lost indication?
The APRM channel 3 flow biased alarm is also reduced by this reduction in recirculation flow.
Other channels are not affected.
OPL171.148 Revision 12 Page 45 of 106 INSTRUCTOR NOTE 2.
General Description a.
RBM consists of two redundant channels for monitoring of reactor power in the immediate vicinity of a control rod selected for movement.
(1)
b.
RBM is active above 25% Simulated Obj. V.B.22,d Thermal Power as determined by the reference APRM for the associated RBM channel.
(1)
A RBM receives Simulated Thermal Obj. V.C.6.b Power (STP) input from APRM #1 with alternate APRM being APRM #3 and second alternate channel being APRM #4.
(a)
Alternate APRM is automatically selected when associated primary APRM is bypassed (2)
B RBM receives Simulated Thermal Obj. V.B.22.b Power (STP) input from APRM #2 with alternate APRM being APRM #4 and second alternate channel being APRM #3.
(a)
Alternate APRM is automatically selected when associated primary APRM is bypassed (3)
Both RBM channels are bypassed when APRM reference power is less than 25% STP.
(4)
Both RBM channels are also bypassed when a peripheral (edge) rod is selected.
(5)
RBM must have an internal control rod selected to be active.
Control Rod Block Instrumentation 13.2.1 3.3 INSTRUMENTATION 3.3.2.1 Control Rod Block Instrumentation LCO 3.3.2.1 The control rod block instrumentation for each Function in Table 3.3.2.1-1 shall be OPERABLE.
APPLlCABILITY According to Table 3.3.2.1-1.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One rod block monitor A.1 Restore RBM channel to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (RBM) channel OPERABLE status.
B. Required Action and B.1 Place one RBM channel 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> associated Completion in trip.
Time of Condition A not met.
OR Two RBM channels inoperable.
C. Rod worth minimizer C.1 Suspend control rod Immediately (RWM) inoperable during movement except by reactor startup.
OR (continued)
(continued) analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in References 4, 5, 6, 7, and 12. The standard BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, Rod Pattern ControL
Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.11 (page 1 of 1)
Control Rod Block Instrumentation V
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)
SR 3.3.2.1.4 SR 3.3.2.1.8 b.
Intermediate Power Range - Upscale (b) 2 SR 3.3.2.1.1 (e)
SR 3.3.2.1.4 SR 3.3.2.1.8 c.
High Power Range - Upscale (t).(g) 2 SR 3.3.2.1.1 (e)
SR 3.3.2.1.4 SR 3.3.2.1.8 d.
Inop (g).(h) 2 SR 3.3.2.1.1 NA a.
Downscale (g).(h) 2 SR 3.3.2.1.1 (i)
SR 3.3.2.1.4 2.
(c) 2 (c) 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.7 3.
Reactor Mode Switch Shutdown Position (d) 2 SR 3.3.2.1.6 NA (a)
THERMAL POWER 27% and 62% RTP and MCPR less than the value specified in the COLR.
(b)
THERMAL POWER> 62% and 82% RTP arid MCPR less than the value specified in the COLR.
(c)
With THERMAL POWER 10% RTP, except during the reactor shutdown process if the coupling of each withdrawn control rod has been confirmed.
(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.
Clarification Guidance for SRO-only Questions RevI (0311112010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing 1
Yes hour TSITRM Action?
RO question Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line?
RO question Does the question involve one or more of the following for TS, TRM. or 00CM?
0 Application of Required Actions (Section 3) and Surveillance Requirements (Section 4> in accordance with rules of application requirements (Section 1)
Application of generic LcO requirements (LCO 3.0.1 thru 3.01 and SR 4.0:1 thru 404)
Yes SRO-only Knowledge of TS bases that is required to analyze TS question required actions and terminology No j Question might not be linked to 10 CER 5543(b)(2) for SRO-only
QUESTION 92 Unit 2 Mode Switch is in Startup/Hot Standby, Reactor Power is 8%, and preparations are being made to place the Mode Switch in Run.
Subsequently:
A failure in the Main Turbine Bypass Valve control causes Reactor Pressure to rise Attempts to manually control Main Turbine Bypass Valves are unsuccessful The Reactor is manually scrammed Following the scram:
All rods are in The Mode Switch has been placed in Shutdown Reactor Pressure has stabilized at 1060 psig Which ONE of the following completes both statements below?
For these conditions, the required Technical Specification 3.4.10, Reactor Steam Dome Pressure, actions are (1)
The fl Immediate NRC Notification required is a (2) report.
[REFERENCE PROVIDED]
A.
(1) no longer applicable (2) 4-Hour B.
(1) no longer applicable (2) 8-Hour C.
(1) still required to be completed (2) 4-Hour D.
(1) still required to be completed (2) 8-Hour Correct Answer: A
Level:
1 Group#
I KIA#
241000 G2.4.30 Importance Rating 4.1 241000 Reactor/Turbine Pressure Regulating System:
2.4.30 Knowledge of which events related to system operation/status should be reported to outside agencies.
Explanation: A CORRECT: TS 3.4.10 requires Rx Pressure to be less than 1050 psig when Rx is in Mode I and 2, TS 3.7.5 requires Turbine Bypass System to be Operable when Rx power is >25%. In the conditions given, Rx is shutdown in Mode 3, no TS Action required. Therefore the Tech Spec actions are no longer applicable. RPS actuation requires a 4-Hour notification when the reactor is critical.
B Incorrect First Part: Correct. Second Part: Incorrect, Plausible because RPS actuation requires 8-Hour Notification, unless Rx is critical when actuated, then a 4-Hour notification is required.
C Incorrect First Part: Incorrect, Plausible because TS 3.4.10 requires Rx Pressure to be less than 1050 psig when Rx is in Mode I and 2, and TS 3.7.5 requires Turbine Bypass System to be Operable when Rx power is >25%. In the conditions given, Rx is shutdown in Mode 3, no TS Action required. Second Part: Correct.
D Incorrect First Part: Incorrect, TS 3.4.10 requires Rx Pressure to be less than 1050 psig when Rx is in Mode 1 and 2, TS 3.7.5 requires Turbine Bypass System to be Operable when Rx power is
>25%. In the conditions given, Rx is shutdown in Mode 3, no TS Action required. Second Part: Incorrect, Plausible because RPS actuation requires 8-Hour Notification, unless Rx is critical when actuated, then a 4-Hour notification is required.
Technical Reference(s)
NPG-SPP-03.5, Tech Spec 3.4.10, Tech Spec 3.7.5 Proposed references to be provided to applicants during examination: NPG-SPP-03.5, Appendix A Learning Objective (As available):
Question Source:
Bank:
Modified Bank:
New: X Question History:
Previous NRC: None Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis:
X 10 CFR Part 55 Content:
55.43 (2) Facility operating limitations in the technical specifications and their bases.
Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.410 Reactor Steam Dome Pressure LCO 3.410 The reactor steam dome pressure shall be 1050 psig, APPLICABILITY:
MODES I and 2.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A. I Restore reactor steam 15 minutes pressure not within limit, dome pressure to within limit.
B. Required Action and B.I 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.
31 PLANT SYSTEMS 31.5 Main Turbine Bypass System
[00 3.7.5 The Main Turbine Bypass System shall be OPERABLE The following limits are made applicable:
- a. [CO 3.2.1, AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)° limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and b.
LCO 3.2.2 MINlMUM CRITICAL POWER RATIO CPR) limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and c.
LCO 3.2.3, LINEAR HEAT GENERATION RATE (LHGR),
limits for an inoperable Main Turbine Bypass System, as specified in the COLR.
APPLICABILITY:
THERMAL POWER 25% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the [CO Al Satisfy the requirements 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> not met.
of the LCO.
B. Required Action and B,1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to < 25% RTP.
Time not met BFN-UNIT 2 3.7-17 Amendment No.454-287 December 30, 2003
NPG Standard Regulatory Reporting Requirements NPGSPP-O3.5 Programs and Rev. 0008 Processes Page 22 ot99 Appendix A (Page 4 of 15)
Reporting of Events or conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification NRC {:contjnu) d.
The actual or attempted introduction of contraband into a protected area, material access area, vital area, or transport.
C.
The following criteria require 4-hour notification:
1.
§50.72(b)(2)(i)
- The initiation of any nuclear plant shutdown required by the plants Technical Specifications.
2.
§5O.72(b){2)(iv(A) - Any event that results or should have resulted in Emergency Core Cooling System (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-plarined sequence during testing or reactor operation.
3.
§50.72(bX2)(iv)(B)
- Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is criticai except when The actuation results from and is part of a pre-planned sequence during testing or reactor operation.
NOTES 1)
NPG-SPP-05,14 provides additional instructions regarding addressing and informally communicating events to outside agencies involving radiological spills and leaks.
2)
Routine or day-to-day communications between WA organizations and state agencies typically do not constitute a formal notification to other government agencies that would require a report in accordance with §50.72(b)(2)(xi).
4.
§50.72(b)(2)(xi)
- Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactive contaminated materials.
D.
The following criteria require 8-hour notification:
NOTE With the excepflon of Events or Conditions that Could Have Prevented Fulfillment of a Safety Function, ENS notifications are required far any event that occurred within three years of discovery, even if the event was not ongoing at the time of discovery.
NPG Standard Regulatory Reporting Requirements NPG4PP435 Programs and Rev. 0008 Processes Pane 23 of 99 Appendix A (Page 5 of 15)
Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification NRC (continued)
§50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety banters, being seriously degraded.
2.
§50.72(b)(3)(ii)tB)
- Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.
3.
§50. 72(bM3)(iVXA)
- Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(SXiv)( ) [see list below], except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.
a.
The systems to which the requirements of paragraph §50.72(bX3)(iv)(A) apply are:
NOTE Actuation of the RPS when the reactor is critical is also reportable under §50.72(b)(2)(iv)(B) above.
(1)
Reactor protection system (P PS) including: reactor scram or reactor (2)
General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVS).
Clarification Guidance for SRO-only Questions RevI (0311112010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing 1
Yes hour TSITRM Action?
RO queston Can question be answered solely by knowing the Yes LCOJTRM information listed above-the-line?
RO question Can question be answered solely by knowing the Yes TS Safety Limits?
RO question Does the question involve one or more of the following for TS, TRM, or ODCM?
0 Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
Application of generic LCO requirements (LCO 301 thru 301 and SR 401 thru 40.4)
Yes SRO-only Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55A3b)(2) for SRO-only
QUESTION 93 Unit 2 is operating at 100% power with Offgas System Isolation Valve, 2-FCV-66-28, mechanically restrained open.
SUBSEQUENTLY, a fuel leak results in the following:
OG POST TRTMT RAD MONITOR HI-HI-HI/INOP (2-9-4C,Window 35) is received 2-RM-90-265A, OG POST TRTMT RAD MONITOR, indicates 6.4 x I cps and rising 2-RM-90-266A, OG POST TRTMT RAD MONITOR, indicates 7.2 x I cps and rising Stack Noble Gas (WRGERMS RM-90-306) indicates 2.95 X i09 llCi/sec and rising Which ONE of the following completes both statements below?
The correct actions to mitigate these conditions are to (1)
In accordance with EPIP-l, Emergency Classification Procedure, prior to the emergency declaration based on WRGERMS indication, an assessment of the release rate by another method must be performed, provided the assessment can be accomplished within (2)
[REFERENCE PROVIDEDI A.
(1) enter 2-AOI-66-2, Offgas Post Treatment Radiation Hi Hi Hi ONLY (2) 15 minutes B.
(I) enter 2-AOI-66-2, Offgas Post Treatment Radiation Hi Hi Hi ONLY (2) 60 minutes C.
(1) enter 2-AOI-66-2, Offgas Post Treatment Radiation Hi Hi Hi AND 0-EOI-4, Radioactivity Release Control (2) 15 minutes D.
(1) enter 2-AOI-66-2, Offgas Post Treatment Radiation Hi Hi Hi AND 0-EOI-4, Radioactivity Release Control (2) 60 minutes Answer: C
Level:
2 Group#
2 Examination Outline Cross-Reference KIA#
271 000A2.09 Importance Rating 2.8 Ability to (a) predict the impacts of the following on the OFFGAS SYSTEM ; anu (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve closures Explanation: CORRECT C First Part: The Offgas Post Treatment Hi Hi Hi is valid based on the radiation monitor indications. Therefore, the correct actions to mitigate these conditions are to enter 2-AOI-66-2, Offgas Post Treatment Radiation Hi Hi Hi. 2-AOl-66-2 will direct the closure of Offgas System Isolation Valve, 2-FCV-66-28. 0-EOI-4, Radioactivity Release Control is entered at an Emergency Plan declaration of Alert or higher. In this case entry would be required. Second Part: Prior to declaring the ALERT, EPIP-1, Emergency Classification Procedure states that an assessment of the release rate by another method must be performed, provided the assessment can be accomplished within 15 minutes.
A Incorrect: First Part: Incorrect. 0-EOI-4, Radioactivity Release Control is entered at an Emergency Plan declaration of Alert or higher. In this case entry would be required. Second Part: Correct.
B Incorrect. First Part: Incorrect. 0-EOl-4, Radioactivity Release Control is entered at an Emergency Plan declaration of Alert or higher. In this case entry would be required. Second Part: incorrect. Plausible because Prior to declaring the NOUE, EPIP-1, Emergency Classification Procedure states that an assessment of the release rate by another method must be performed, provided the assessment can be accomplished within 60 minutes.
D Incorrect. First Part: Correct. Second Part: incorrect. Plausible because Prior to declaring the NOUE, EPIP-1, Emergency Classification Procedure states that an assessment of the release rate by another method must be performed, provided the assessment can be accomplished within 60 minutes.
Technical Reference(s): EPIP-1, 2-AOI-66-2, 2-ARP-9-4C Proposed references to be provided to applicants during examination: EPIP-1 Section 4 Learning Objective (As available):
Question Source:
Bank:
Modified Bank:
New: X Question History:
Previous NRC: None Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis: X 10 CFR Part 55 Content:
55.43 b(S) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations
BFN Panel 9-4 2-ARP-9-4C Unit 2 2-XA-55-4C Rev. 0031 fg4 pf 44 Sensor/Trip Point:
OG POST TRTMT 2-RM-90-265A 6.2 x 1 5 cps 2-RM-90-266A 6.2 x I 0 eps 2-RA-90-265C (Page 1 of 1)
Sensor 2-RE-90-265 Panel 2-25-94 Off-Gas Building Location:
2-RE-90-266 Elevation 538.5 Probable A.
Resin trap failure (RWCU or Condensate demins).
Cause:
B.
Fuel damage.
Automatic OFFGAS SYSTEM ISOLATION VALVE 2-FCV-66-28 closes after a 5 second time Action:
delay Operator A. VERIFY alarm condition on the following D
Action:
OFFGAS RADIATION recorder, 2-RR-90-266 on Panel 2-9-2 0
OG POST-TREATMENT CHAN A RAD MON RTMR radiation monitor, 2-RM-90-266A on Panel 2-9-10.
0 OG POST-TREATMENT CHAN B RAD MON RTMR radiation monitor, 2-RM-90-265A on Panel 2-9-10.
0 B. VERIFY OFF-GAS SYSTEM ISOLATION VALVE, 2-FCV-66-28 has the Mechanical Restraint DISENGAGED and 2-FCV-66-28 is CLOSED.
C C. REFER TO 2-AOl-66-2.
C
References:
2-45E620-4 2-45E614-2 247E610-90-2 2-1 15D6410RE-3 GE 2-729E814-6 FSAR Sections 1.6.4.4.6, 7.12.2.2, 7.12.2.3, 7.12.3.3, 9.5.4, and 13.6.2
BFN Offgas Post-Treatment Radiation HI-HI-2-AOl-66-2 Unit 2 HI Rev. 0021 Page 5 of 8 4.0 OPERATOR ACTIONS 4.1 Immediate Actions
[1]
IF scram has not occurred, THEN PERFORM the following:
[1.1]
IF core flow is above 60%. THEN REDUCE core flow to between 50-60%.
D
[1.2]
MANUALLY SCRAM the Reactor. (Reference 2-AOl-lCD-i).
4.2 Subsequent Actions j1]
IF OFFGAS SYSTEM ISOLATION VALVE, 2-FCV-066-0028 has been mechanically restrained open due to plant conditions THEN DISENGAGE 2-FCV-066-0028 mechanical restraint by rotating the restraining handwheel fully in the counterclockwise direction, locally at the Stack. (Otherwise N!A)
D
[2]
VERIFY CLOSED OFFGAS SYSTEM ISOLATION VALVE.
2-FCV-66-28 on Panel 2-9-53 or locally.
D
[31 MONITOR area radiation levels at Panel: 2-9-il.
C
[4]
REFER TO EPIP-l for emergency classification level and response.
C
[5]
MONITOR the following parameters:
A.
MAIN STEAM LINE RADIATION, 2-RR-90-135, Panel 2-9-2.
C B.
OFFGAS RADIATION, 2-RR-90-266, Panel 2-9-2.
C C.
STACK GAS RADIATION, 0-RR-90-147, Unit 1 Panel 1-9-2.
C
BFN EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 I
Rev. 0049 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 40 OF 205 NOTES 4.1-U Peer to makrg this emergency classification based upon the WRGERMS indication, assess the release by either of the foliowing:
- 1. Actual field measurements exceed the lirnfs in table 41-U 20-Si 4,aB.l.al release fraction exceeds2.O If neither assessment can be conducted within 60 minutes then the declaration must be made on the valid WRGERMS readi 4.1-A Prior to making this emergency classification based upon the WRGERMS indication, assess the release by either of the felbiog:
1 Actual fIeld measurements exceed the itmits in table 41-A 2.0-Si 4.8.B.1.a.1 release fraction exceeds 200 ifneither assessment can be conducted within 15 minutes then the declaration must be madeon the valid WRGERMS reading.
4.1.
Prior to making this emergency classification based upon the gaseous release rate indication, assess the release by either of the fdlowing methods:
1 Actualtield measurements exceed the limits in table 41-S.
- 2. Projected or actual dose assessments exceed 100 mrern TEIJE or 500 mrem CDE.
If neither assessment can be conducted within 15 minutes then the declaration must be made based on the valid WRGERMS reading.
4,I-G Piiorto making this emergency classification based upon the gaseous release rate indication, assess the release by either of the fellowing methods:
I. Actual field measurements exceed the limits in table 4.1-G.
2 Projected or actual dose assessments exceed 1000 mrem TEDE or 5000 mrem CDE.
If neither assessmentcan be conducted within 15 minutes then the declaration must be made based on the valid WRGERMS reading.
CU RVESITABLES:
Table 4.1-U RELEASE LIMITS FOR UNUSUAL EVENT TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS) 2.88 X 10 uCi/sec 1 Hour Gaseous Release Rate 0-SI 4.8.B.1.a.1 Release Fraction 2.0 1 Hour Site Boundary Radiation Reading Field Assessment Team 0.10 MREMHR Gamma 1 Hour Table 4,1-A RELEASE LIMITS FOR LERT TYPE Gaseous Release Rate Gaseous Release Rate Site Boundary Radiation Reading riuraiiuearG METHOD Stack Noble Gas (WRGERMS) 0-SI 4.8.B.1.a.1 Field Assessment Team LIMIT 2.88 X 10 9Csec Release Fraction 200 10 MREMIHR Gamma DURATION 15 Minutes 15 Minutes 15 Minutes Table 4.1-S RELEASE LIMITS FOR SITE AREA EMERGENCY TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas tWRGERMS) 59 X 10 llCFsec 15 MinuIe Site Boundary Radiation Readinci Field Assessment Team 100 MREM HR Gaas ia 1 Hour Site Boundary Iodine-I 31 Field ssessmenI Team 3.9 X IC iCI cm 1 Hour Table 4.1-G RELEASE LIMITS FOR GENERAL EMERGENCY TYPE MONITORING METHOD LIMIT DURATION Gnsoous Roloase Rate Stack Noble Gas tWRGERMSt 5.9 X 0 10 liCi sec 15 Minutes Site Boundary Radiation Reading
Field Asse,srnent Team 000 MREM HR Gamma 1 Hour Site Boundary Iodine-l3l Field Assessment Team 3.9 X 10 pCI cm I Houi
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and seiection of procedures)
Can the question be answered solely by knowing systems knowledge. ie. how the system works Yes P0 questIon flewpath, logic, component locabon?
Can the question be answered solely by knowing immediate operator actions?
Yes P0 question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?
Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes P0 question overall mitigative strategy of a procedure?
Does the question require one or more of the following?
0 Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed 0
Knowledge of when to implement attachments and appendices including how to coordinate these items with procedure steps Yes SPO-only Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No j Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only
QUESTION 94 Unit I is at 8% power performing 1-GOI-lOO-IA, Unit Startup, and preparing to transition to MODE I.
The following alarm and pump seal pressure indications are received:
RECIRC PUMP IA NO I SEAL LEAKAGE ABN (1-9-4A, Window 25)
No. I Seal Pressure: 980 psig No. 2 Seal Pressure: 980 psig In accordance with Tech Specs, which ONE of the following completes the statements below?
The alarm/indications (1)
RCS pressure boundary leakage.
Mode I (2
be entered if the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> average TOTAL leakage stabilizes at 31 gpm.
A.
(1) represent (2) can B.
(I)represent (2) CANNOT C.
(1) do NOT represent (2) can D.
(1) do NOT represent (2) CANNOT Correct Answer: D
Level:
3 Group #
K/A#
G2.1.23 Importance Rating I
Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Explanation: D CORRECT: The alarm/indications given in the stem do NOT represent RCS pressure boundary leakage. LCO 3.0.4: When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time.
A Incorrect.- First Part: Incorrect. Plausible because the Recirc Pump Seals do contain the reactor pressure and because the interpretation of what does/does not constitute pressure boundary leakage is located in the Tech Spec Bases. (However, the 1st part of the question is RO knowledge since the RO knows which systems are routed to the equipment drain sump
- identified leakage).
Second Part: Incorrect. Plausible if the candidate does not interpret LCO 3.0.4, which is not provided as a reference.
B Incorrect First Part: Incorrect. Plausible because the Recirc Pump Seals do contain the reactor pressure and because the interpretation of what does/does not constitute pressure boundary leakage is located in the Tech Spec Bases. (However, the 1st part of the question is RO knowledge since the RO knows which systems are routed to the equipment drain sump
- identified leakage).
Second Part: Correct.
C Incorrect First Part: Correct. Second Part: Incorrect. Plausible if the candidate does not interpret LCO 3.0.4, which is not provided as a reference.
Technical Reference(s) 2-GOI-100-1, TS 3.0.4 Proposed references to be provided to applicants during examination: None Learning Objective (As available):
Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: BFN 1108 #92 Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis:
X 10 CFR Part 55 Content:
55.43 (b)(2) Facility operating limitations in the technical specifications and their bases.
BFN Panel 9-4 1-ARP-9-4A Unit I 1-XA-55-4A Rev. 0021 Page 33 of 47 RECIRC PUMP IA Sensor/Trip Point:
NO 1 SEAL LEAKAGE ABN I-Pl-068-0063A c 400 psig lowering 600 S9 rising 1 -PA-68-63 r (Page 1 of 3)
Sensor Recirculation Pump IA Location:
Drywell Elevation 549.2 Probable A.
Plugging of No. I and/or No. 2 RO (controlled pressure breakdown orifice).
Cause:
B.
Failure of no. I seal.
C. Reactor Pressure <450 psig (Alarm resets at> 650 psig).
Automatic None Action:
Operator A.
DETERMINE initiation cause by comparing No.1 and 2 seal cavity Action:
pressure indicators on 1-9-4 or ICS.
0 a
Plugging of No. 1 RO
- No. 2 seal cavity pressure indicator h.
drops toward zero.
0 a
Plugging of No. 2 RO
- No. 2 seal pressure approaches No. 1 seal pressure.
0 a
Failure of No. I seal
- No. 2 seal pressure is greater than 50%
of the pressure of No.1.
0 Failure of No. 2 seal
- No. 2 seal pressure is less than 50% of the No. I seal.
0 B.
RECORD pump seal parameters hoully on Attachment 1, Page 3 of this window response, unless other compensatory methods for recording these parameters is evaluated and approved by Engineering.
EJ C.
IF single seal failure is indicated, THEN INITIATE seal replacement as soon as possible. Continued operation is permissible if Drywell Leakrate is with T. S. limits 0
Continued on Next Page
RCS Operational LEAKAGE 3-4.4 34 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO 3.4.4 RCS operational LEAKAGE shall be limited to:
- a. No pressure boundary LEAKAGE; b.
5 gpm unidentified LEAKAGE; and c.
30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and d.
2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in MODE 1.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> not within limit, within limits.
OR Total LEAKAGE not within limit.
B. Unidentified LEAKAGE B.1 Reduce LEAKAGE 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> increase not within limit, increase to within limits.
OR (continued)
BEN-UNIT 1 3.4-9 Amendment No. 234
Clarification Guidance for SRO-only Questions RevI (03/1112010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing I
Yes hour TSITRM Action?
RO question Can question be answered solely by knowing the Yes LCOITRM information listed above-theline7 RO question Can question be answered solely by knowing the Yes TS Safety Limits?
RO question Does the question involve one or more of the following for TS, TRM, or ODCM?
Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
Application of generic LCO requirements (LCO 301 thru 3.01 and SR 401 thru 40.4)
Yes SRO-only Knowledge of TS bases that is required to analyze TS question required actions and terminology No j Question might not be linked to 10 CFR 5543(b)(2) for SRO-only
BFN 1108 NRC #92 92.202001 G12.44 NEWIH Unit I is at 8% power with the Mode switch in the Startup/Hot Standby position.
The following alarm and pump seal pressure indications are received:
REcIRC PUMP 1A NO I SEAL LEAKAGE ABN (1-9-4A. Window 25)
No. 1 Seal Pressure: 980 psig No. 2 Seal Pressure: 980 psig In accordance with Tech Specs. Which ONE of the following completes the statements below?
The alarm/indications _(1)
RCS pressure boundary leakage Mode 1 _(2_ be entered if the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> average TOTAL leakage stabilizes at 31 gpm, A. (1)represent (2) cannot B. (1) do NOT represent (2) cannot C. (1) represent (2) can D
(1) do NOT represent
- 2) can Answer: B
QUESTION 95 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A
One pump Al Restore pump 7 days inoperable, to OPERABLE
- status, 8
Required BI Be in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and MODE I associated Completion AN Time not met.
B.2 Be tn 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> MODE 4 Given the above LCO:
Pump I becomes inoperable at 0600 on May I 1 t1* Pump 2 becomes inoperable at 0100 on May 12 th Which ONE of the following completes the statements below?
(Both pumps are in the same system)
If Pump I is restored to OPERABLE at 0800 on May 1 2th1, and if Pump 2 is NOT restored to OPERABLE then Condition B is entered at time (I)
If Pump 2 is restored to OPERABLE at 0800 on May 12 th, and if Pump 1 is NOT restored to OPERABLE then Condition B is entered at time (2)
A.
(I)OlOOMayl9th (2) 0600 May 18 th B.
(I)OIOOMayI9th (2) 0600 May 19 th C.
(1)OIOOMay2Oth (2) 0600 May 18 th D.
(I)OIOOMay2Oth (2) 0600 May 19 th Answer: A
Level:
3 Group#
KIA#
G2.2.22.
Importance Rating I
Knowledge of limiting conditions for operations and safety limits.
Explanation: A CORRECT: If Pump 1 is restored to OPERABLE at 0800 on May 12th, and if Pump 2 is NOT restored to OPERABLE then Condition B is entered at 0100 May 19 th, 7 days for Pump 2 an extension from the original entry of 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br />. If Pump 2 is restored to OPERABLE at 0800 on May 12 th, and if Pump 1 is NOT restored to OPERABLE then Condition B is entered at 0600 May 18 th, 7 days for pump I.
B Incorrect First Part: Correct. Second Part: Incorrect. Plausible because the candidate may incorrectly apply the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension.
C Incorrect First Part: Incorrect. Plausible because the candidate may incorrectly apply the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension. S&ond Part: Correct.
D Incorrect First Part: Incorrect. Plausible because the candidate may incorrectly apply the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension. Second Part: Incorrect. Plausible because the candidate may incorrectly apply the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension.
Technical Reference(s)
TS 1.3 Completion Times examples Proposed references to be provided to applicants during examination: None Learning Objective (As available):
Question Source:
Bank:
Modified Bank:
New:
X Question History:
Previous NRC: None Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis:
X 10 CFR Part 55 Content:
55.43 (2) Facility operating limitations in the technical specifications and their bases.
Completion limes I.3 I.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued)
When a pump is declared inoperable, Condition A is entered.
If the pump is not restored to OPERABLE status With:ifl 7 days, Condition 8 is also entered and the Completion lime clocks for Required Actions B.i and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered.,
Condition A and B are exited, and therefore, the Required Actions of Condition B may be tern mated.
When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered for the second pump. LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump.
The Completion Time clock for Condition A does not sop after LCO 3.Q3 is entered, but continues to be tracked from the time Condition A was initially entered.
While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for condition A has not expired, 1C0 3.0.3 may be exited and operation continued in accordance with Condition A.
While in LCO 3.0.3, if one of the inoperable pumps is restored to OPER) BLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition B. The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired.
On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This Co:mpletion Time: may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension to the stated 7 days is allowed,, provided this does not result in the second pump being inoperable for> 7 days..
(continuedi
Clarification Guidance for SRO-only Questions RevI (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing 1
Yes hour TSITRM Action?
RO question Can question be answered solely by knowing the Yes LCOiTRM information listed above-the-line?
RO question Can question be answered solely by knowing the Yes TS Safety Lims?
RO question Does the question involve one or more of the following for TS.
Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
Application of generic LCO requirements (LCO 3.0 1 thru 307 and SR 4.0.1 thru 404)
Ye SRO-only Knowledge of TS bases that is required to analyze TS question required actions and terminology No j Question might not be linked to 10 CFR 55.43(b)(2) for SROonly
QUESTION 96 Unit 3 is operating at 98% power.
During a control rod surveillance a reactivity event resulted in the following annunciators:
RBM HIGH/INOP (3-9-5, Window 24)
CONTROL ROD WITHDRAWAL BLOCK (3-9-5, Window 7)
The Unit Operator (UO) observes the following values on Powerplex:
MFLCPR 0.925 MAPRAT 0.754 MFDLRX 1.20 Which ONE of the following completes both statements below?
All Tech Spec 3.2, Power Distribution Limits, Limiting Conditions for Operation (LCO) are (I)
The procedure that provides classification criteria for reactivity events is (2)
A.
(l)met (2) OPDP-I, Conduct of Operations B.
(l)met (2) NPG-SPP 10.4, Reactivity Management Program C.
(I)NOTmet (2) OPDP-1, Conduct of Operations D. (l)NOT met (2) NPG-SPP 10.4, Reactivity Management Program Correct Answer: D
Level:
3 Group #
KIA#
G2.2.38 Importance Rating I
Knowledge of conditions and limitations in the facility license.
Explanation: D CORRECT: All Tech Spec 3.2, Power Distribution Limits, Limiting Conditions for Operation (LCO) are Not met. MFDLRX is greater than 1.0 at 1.20
. The procedure that provides classification criteria for reactivity events is NPG-SPP 10.4, Reactivity Management Program.
A Incorrect First Part: Incorrect. Plausible if the applicant does not know that MFDLRX is the same as LHGR. Second Part: Incorrect. Plausible because the severe classification examples (BWR) include rods and Tech Specs.
B Incorrect-First Part: Incorrect. Plausible ifthe applicant doesnt understand that MFDLRX, MFLCPR and MAPRAT are ratios of the thermal limit value compared to its maximum, and any value >1 means the thermal limit value has exceeded its maximum. Second Part: Correct.
C Incorrect First Part: Correct. Second Part: Incorrect. Plausible because the severe classification examples (BWR) include rods) and Tech Specs.
Technical Reference(s) 3-AOl-85-7, Mispositioned Control Rod, Tech Specs 3.2.3, Linear Heat Generation Rate, NPG-SPP-10.4, Reactivity Management Program Proposed references to be provided to applicants during examination: None Learning Objective (As available):
Question Source:
Bank:
Modified Bank:
X New:
Question History:
Previous NRC: BFN 1108 #84 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis:
X 10 CFR Part 55 Content:
55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)
LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the APPLiCABILITY:
THERMAL POWER 25% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TiME A. Any LHGR not within A.1 Restore LHGR(s) to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limits.
within limits.
B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to < 25% RTP.
Time not met.
NPG-SPP 10.4 Reactivity Management F.
Reactivity management issues will be assigned an event category per the following criteria:
There are six different Significance Levels (SLs) for Reactivity Management Issues based on plant impact.
If an issue can be classified at more than one SL, then the highest SL is used.
Management discretion can be used to raise the SL of an issue, but not to lower it.
NPG Standard Reactivity Management Program NPG-SPP-1 0.4 Programs and Rev. 0003 Processes Page 21 of 69 3.2.5 Issue Identification and Trend Analysis (continued>
RMRB can raise or lower SL if, in their judgment, it was incorrectly assessed.
SL Plant Impact Short Description 1
Highest Fundamental Organizational Breakdown 2
Violation of Design or Licensing Basis 3
Violation of Process I Procedural Requirements 4
Precursor 5
Lowest Concern 6
None No impact; trending purposes only SL 1 through SL 3 events are lagging indications of performance. SL 4 and SL 5 issues, in general, may be precursors to more significant events and are considered leading indications of performance.
A Reactivity Management Issue that results in a significant plant impact or indicates a high potential for future significant events is classified as a Reactivity Management Event. There are five different classifications for Reactivity Management Events based on severity with the sixth level for issues that do not impact Reactivity Management but are monitored and trended to prevent more significant issues from occurring at a later time. Since almost all Reactivity Management-related activities are protected by at least two barriers, a Reactivity Management Event normally involves the failure of at least two barriers. Barriers include but are not limited to: redundant indications, potential or actual operator actions, procedures, and control systems.
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge. i.e. how the system works Yes flowpath, logic, component location?
RO question Can the question be answered solely by knowing 1
immediate operator actions?
Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?
Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?
Does the question require one or more of the following?
0 Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed 0
Knowledge of when to implement attachments and appendices including how to coordinate these items with procedure steps Yes SRO-only 0
Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-oniy
BEN 1108 #84
- 84. 295014G2.2.38 OO1f1/2/SROiNEWiH/3/IvIAR Unit 3 is operating at 98% power.
During a control rod surveillance, the Unit Operator (UO) was required to single notch a control rod from position 14 to 16; however, the UO continuously withdrew the control rod until the following annunciators began alarming:
CONTROL ROD WITHDRAWAL BLOCK (9-5, W7)
The control rods final position is 26 and the Unit Operator (UO) observes the following values on Powerplex:
MFLCPR 0.925 MAPRAT 0.754 MFDLRX 1.20 MFLPD 0.00 Which ONE of the following indicates:
- 1) whether all Tech Spec 3.2, Power Distribution Limits, Limiting Conditions for Operation ([00) are met and 2)the required classification for this event in accordance with NPG-SPP-10.4, Reactivity Management Program?
A. All Tech Spec 3.2 LOOs are met. (no required action statement)
Severe Reactivity Management Event (SL 1)
B. All Tech Spec 3.2 LOOs are NOT met.
Severe Reactivity Management Event (SL 1)
C. All Tech Spec 3.2 LCOs are met. (no required action statement)
Major Reactivity Management Event (SL 2)
D All Tech Spec 3.2 LCOs are NOT met.
Major Reactivity Management Event (SL 2)
QUESTION 97 After a radwaste tank was properly recirculated and sampled, o
The SRO authorized the release in accordance with O-SI-4.8.A.1-1, Liquid Effluent Permit.
Subsequently, During the batch release, the O-RM-90-l30, Radwaste Effluent Radiation Monitor is determined to be inoperable.
Which ONE of the following identifies the required actions in accordance with O-Sl-4.8.A.1-l?
Note:
Attachment I of O-SI-4.8.A. 1-1 is titled Bypassing the RM-90-1 30 Isolation Logic of O-SI-4.8.A.l-1 is titled Valve Checklist A.
The in-progress release may continue provided Attachment 2 is performed and verified.
B.
The in-progress release may continue if an independent verification of the release rate calculations and valve alignments are completed.
C.
The release would be terminated until Attachment 1 and 2 are performed and the release is re-commenced on the current O-Sl-4.8.A.1-1.
D.
The release must be terminated, a new O-S1-4.8.A.I-l Liquid Effluent Permit is required to be initiated and approved.
Answer: D
Level:
3 Group #
Examination Outline Cross-Reference KIA#
G2.3.6 Importance Rating 3.8 2.3.6 Ability to approve release permits.
Explanation: D CORRECT: If RM-90-130 is declared inoperable during the release, the release must be terminated and a new SI initiated. This is to ensure the requirement for two independent samples, independent verification ofthe release rate calculations, and independent verification of valve alignment are met.
AIncorrect. Plausible the RM-90-130 can be inoperable during the release provided certain conditions in the SI are met PRIOR to the release.
B Incorrect. Plausible the RM-90-130 can be inoperable during the release provided certain conditions in the SI are met PRIOR to the release.
C Incorrect: Plausible a release may be performed with an inoperable monitor provided attachment I and 2 are performed Technical Reference(s): O-SI-4.8.A.I-I Proposed references to be provided to applicants during examination: None Learning Objective (As available):
QuestionSource:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: River Bend 2008 #97 Question Cognitive Level:
Memory or Fundamental Knowledge: X Comprehension or Analysis 10 CFR Part 55 Content:
55.43 b (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
3.0 PRECAUTIONS AND LIMITATIONS A.
A Unit Supervisor must authorize this release and, when required, authorize the bypassing of the 0-RM-90-1 30 isolation logic.
B.
Start first sample counting within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of sample time.
C.
If the Plant Release MODE (i.e., OPEN or HELPER> changes or CCWP(s) are removed from service during the performance of this SI and the dilution flow is decreased, the release must be terminated until a new SI is initiated or until original dilution flow is restored.
If the 0-RM-90-1 30 monitor is declared inoperable during a release, terminate the release and initiate a new SI. This is to ensure the requirement for two independent samples, independent verification of the release rate calculations, and independent verification of valve alignment are met.
BFN Liquid Effluent Permit 0-SI-4.8.A.1-1 Unit 0 Rev. 0074 Page 7 of 55 3.0 PRECAUTIONS AND LIMITATIONS (continued)
E.
If 77-60 is declared inoperable during the release, the release may continue provided the time of inoperability is recorded and flow rate is estimated once every four hours during the release.
If 0-RM-90-1 30 or 77-60 is inoperable per ODCM Requirements, N/A may be recorded in the blanks requiring data from these instruments or the instrument readings may be recorded with the understanding the data is taken from ODCM inoperable equipment and is NOT intended to meet ODCM requirements for the release.
2008 River Bend Station Initial NRC License Examination Senior Reactor Operator QUESTION 97 Rev I Examination Outline Cross-
Reference:
Level ROD SR0 Tier#
3 Group #
Radiation Control K/A#
G2.3.6 Importance Rating 3.8 Ability to apprnve release peririit&
Proposed Question:
Which of the following is required to discharge an LWS tank to the Mississippi River if RMS-RE1O7 is INOPERABLE?
A. Two independent samples of the tank are analyzed. One qualified member of the Chemistry staff and one qualified member of the Radwaste staff independently verify the release rate calculations and the discharge valve lineup.
B A single sample is analyzed by two qualified members of the Chemistry staff independently. Two qualified members of the Radwaste staff independently verity the discharge valve lineup.
C. Two independent samples of the tank are analyzed. Two qualified members of the Chemistry staff independently verify the release rate calculations. Two qualified members of the Radwaste staff independently verify the discharge valve lineup
- 0. A single sample of the tank is analyzed. One qualified member of the Chemistry staff verifies the release rate calculation and one qualified member of the Radwaste staff verifies the discharge valve lineup.
Proposed Answer:
C.
IOCFR55.43 b (4)
D.
Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
(10 CFR 55.43(b)(4)1 Some examples of SRO exam items for this topic include:
Process for gaseous/liquid release approvals, i.e., release permits.
Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
QUESTION 98 Unit 2 is in a refueling outage with the following plant conditions:
A Refuel Floor overhead crane failure has led to dropping a heavy load into the Unit 2 fuel pooi during a core off-load.
2-RA-90-1A, Fuel Pool Floor Radiation Monitor, is in alarm AND reading 1000 mr/hr NO TVA Emergency Response Facilities are activated Based on the above conditions, which ONE of the following completes the statements below?
A(n)
(1) is required to be declared.
The action required to continue assessing plant conditions is to direct the performance of (2)
[REFERENCE PROVIDED]
A.
(1) Alert (2) CECC EPIP-8, Dose Assessment Staff Activities During Nuclear Plant Radiological Emergencies B.
(l)Alert (2) EPTP-13, Dose Assessment C.
(1) Site Area Emergency (2) CECC EPIP-8, Dose Assessment Staff Activities During Nuclear Plant Radiological Emergencies D.
(1) Site Area Emergency (2) EPIP-13, Dose Assessment Correct Answer: B
Level:
3 Group #
KIA#
G2.3.13 Importance Rating 3.8 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
Explanation: B CORRECT: First Part: the correct classification is an ALERT based on EAL 3.2-A.
Second Part: NO TVA EROs are activated, therefore Radcon must do the dose assessment in accordance with EPIP-13.
A Incorrect First Part: Correct. Second Part: Plausible in that if ALL required TVA EROs are activated this would be the correct answer.
C Incorrect First Part: Incorrect. Plausible in that dose level have reached SAE level in accordance with EAL 3.2-S. However, dose levels must be coupled with an unisolable Primary System leak discharging into Secondary Containment to warrant an SAE. Also, not an applicable operating mode. Second Part: Plausible in that if ALL required TVA EROs are activated this would be the correct answer.
D Incorrect First Part: Incorrect. Plausible in that dose level have reached SAE level in accordance with EAL 3.2-S. However, dose levels must be coupled with an unisolable Primary System leak discharging into Secondary Containment to warrant an SAE. Also, not an applicable operating mode. Second Part: Correct.
Technical Reference(s)
EPIP-l, EPIP-3 Proposed references to be provided to applicants during examination: EPIP-1 Section 3 Learning Objective (As available): OPL171.075 V.B.2 Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: BFN 1006 #85 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis :
X 10 CFR Part 55 Content:
55.43 (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
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BFN EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 37 OF 205 SECONDARY CONTAINMENT RADIATION uescription I
Cz Ca C>r m
.<mz
-4 3.2-Al I
I I
I Any of the following high radiation alarms on Panel 9-3:
- 1,2, or 3-RA-90-IA, Fuel Pool Floor Alarm
- 1,2, or 3-RA-90-250A, Reactor, Turbine, Refuel Exhaust
- 1,2, or 3-RA-90-1 42A, Reactor Refuel Exhaust
- 1 2 or 3-RA-90-140A Refueling Zone Exhaust 1m AND
-4 Confirmation by Refuel Floor personnel that irradiated fuel damage may have occurred.
OPERATING CONDITION:
ALL 3.2-S I I
ITABLEI US I An unisolable Primary System leak is discharging into Secondary Containment a
-I AND m
rn Any area radiation level at or above the Maximum Safe Operating Area radiation limit listed in Table 3.2.
OPERATING CONDITION:
Modelor2or3 C) 3,2-G I I
ITABLEI us I An unisolable Primary System leak is discharging into Secondary Containment m
AND Any area radiation level at or above the Maximum Safe Operating Area radiation limit listed in Table 32 AND m
Any indication of potential or significant fuel cladding failure exists. Refer to Table 3.1 -G/3.2-G with RCS Barrier intact inside Primary Containment.
z OPERATING CONDITION C)
Modelor2or3
BEN ALERT EPIP4 Unit 0 Rev 0035 Page 20 of 23 APPENDiX G Page 4 of 5 TECHNICAL SUPPORT CENTER ALERT CLASSIFICATION INSTRUCTION 4.0 Dose Assessment Evaluation (1]
IF emergency circumstances warrant dose assessment, CONTACT, TSC Radiation Protection AND DIRECT Radiation Protection to implement EPIP-13 Dose Assessment.
5.0 Notification of the Nuclear Regulatory Commission (NRC)
NOTE Notification of the NRC is required to be completed as soon as possible not to exceed 60 minutes from classification declaration.
[1]
DIRECT the TSC NRC Coordinator to implement Appendix D, Notification of the NRC.
6.0 Maintaining communications with the NRC NOTE When the TSC is staffed, the open and continuous line of communications with the NRC is managed by the TSC NRC Coordinator position.
[1)
IF REQUESTED by the NRC to maintain an open and continuous line of communications, DIRECT TSC NRC Coordinator to maintain and or manage an open and continuous line of communications as directed by NRC.
SRO Only requires analysis of radiation hazards and prescribing a procedure with which to proceed. This question is rated as C/A 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.
BEN 1006 #85 Examination Outline Cross-reference:
295033 High Secondary Containment Area Radiation Levels/9 G2.3.13 (IOCFR 55.43.4
- SRO Only)
Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
Proposed Question: #
85 Unit 2 is in a refueling outage with the following plant conditions:
A Refuel Floor overhead crane failure has led to dropping a loaded Multi Purpose Canister (MPC) into the Unit 2 fuel pool during a core off-load RA-90-IA, Fuel Pool Floor Radiation Monitor, is in alarm AND reading 1000 mr/hr NO TVA Emergency Response Facilities are activated Based on the above conditions, which ONE of the following describes the HIGHEST required Emergency Action Level AND the action required to continue assessing plant conditions?
[REFERENCE PROVIDED]
A.
Alert, Direct CECC EPIP-8, Dose Assessment Staff Activities During Nuclear Plant Radiological Emergencies, be performed.
B.
Alert, Direct EPIP-13, Dose Assessment, be performed.
C.
Site Area Emergency, Direct CECC EPIP-8, Dose Assessment Staff Activities During Nuclear Plant Radiological Emergencies, be performed.
SRO I
2 Level Tier#
Group#
KIA#
Importan ce Rating 3.8 D.
Site Area Emergency, Direct EPIP-13, Dose Assessment, be performed.
QUESTION 99 The Shift Manager / Site Emergency Director (SM/SED) has declared a General Emergency. The Central Emergency Control Center (CECC) is NOT staffed.
Besides classification, which ONE of the following duties can NOT be delegated to another emergency team member by the SM/SED?
A. Making notifications to the state B.
Directing the shutdown of the plant C.
Conducting site accountability actions D.
Determining Protective Action Recommendations Correct Answer: D
Level:
3 Group#
KIA#
G2.4.40 Importance Rating I
Knowledge of SROs responsibilities in emergency plan implementation.
Explanation: D CORRECT: Per EPIP-5, General Emergency, The Site Emergency Director must make any required recommendations until the CECC is staffed. This responsibility cannot be delegated until CECC is in operation. Recommendations are required at General Emergency.
A Incorrect The Operations Duty Specialist (ODS) should be notified by the SM/SED within five minutes of the event classification. The ODS relays the information to the Emergency Duty Officer (EDO), the State of Alabama, and the CECC Director. The EDO keeps the CECC Director informed of the situation as necessary.
B Incorrect If the event is determined to be one of the four emergency classifications, the Shift Manager assumes the responsibility of SED until relieved by the Plant Manager or designee.
C Incorrect The Shift Manager or SED shall make the decision to activate the assembly and accountability process. The actions carried out as a result of this decision can be delegated but the decision itself cannot be delegated.
Technical Reference(s)
EPIP-5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OPLI7I.075 V.B.4 Question Source:
Bank: X Modified Bank:
New:
Question History:
Previous NRC: BFN 1006 #99 Question Cognitive Level:
Memory or Fundamental Knowledge: X Comprehension or Analysis 10 CFR Part 55 Content:
55.43 (6) Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.
EPIP-5 GENERAL EMERGENCY Rev 0044 tlrnt0 Page 8 of 26 APPENDiX A Page 1 of 1 GENERAL EMERGENCY INITIAL NOTIFICATION FORM IL This is a Drlli D This (s an Actual Event-Repeat - This is an Actual Event I The SED at Browns Ferry has declared a GENERAL EMERGENCY.
I EAL Designator:
(USE ONLY ONE EPL DESIGNATOR)
LI Minor releases within federally approved limits Releases above federally approved limits 1
Release information not known Tech SpecsI000M)
LI Minor releases within federally approved limits El Releases above federally approved l[miis 1
Release Information not known
( Tech SpecsiODCM)
- 5. Event Declared: Time:
___________(
Time)
Date:
- 6. The Meteorological Conditions are:
(Use 91 mater data from the Met Tower.. if data is not available from the MET tower, contact the National Weather Service by dialing 9-1-256-890-8505 or 9-i-20562I-565O.
The National Weather Eerice will provide wind direction and wind speed A2, 52,F2, G2, F5, PlO., G5, GiG, HID From4I°-73°
A2, 52, F2, 02, F5, G5 A2,82,F2.G2,G&,G10,,H10,110 From74a92 A2,B2F2,G2,G5 A2 52 P2 G2 AS G5 HID JIG JIll Kill From 93° 137° A2 82 F2 02 AS G5 AZ 82 P2. G2,AS, AID. 140, JID, KID From 138° -203° A2, 821 F2, :G2. AS A2, 82, F2, G2, AS, AID, 85, 810 From 204°-282° A2, 52, F2, 02. AS, 85 AZ B2, P2. G2, 35, 810. ClO, 010, E&,E10 From 283°326° A2 82, F2:,:G2, B5 E5 A2 52 F2 G2 CID DID ESEIO F5 FIG From 3273l A2 82 P2 02 E5 F5 Recommendation 3 SHELTER all seotors CONSIDER 55uance of POTASSIUM IODIDE n accordance with the State Plan Completed by:
- 4. Radiological Conditions: (Check one under both Airborne and Liquid column.)
Airborne Releases Offsite Liquid Releases Offsite Wind Direction is FROM:
g (15 mm average STEP MUST BE COMPLETED BY THE SITE EMERGENCY DIRECTOR
- 7. Provide Protective Action Recommendation utilizing Appendix H: (Check either I or 2 or 3 n
n Wind Speed:
mph U Recommendation I
- EVACUATE LISTED SECTORS 2 mile Radius & 10 miles downwind
- Shelter remainder of 10 mile EPZ.
- Consider issuance of POTASSIUM IODIDE in accordance with tte State Plan, I
AZ. B2, P2. G2, E5.E10. P5. FlO, GS, G10 WIND FROM DEGREES (Mark wind direction from Step 7)
From 4-40° El Recommendation 2
- EVACUATE LISTED SECTORS (2 mile radius & 5 mile downwind I SHELTER remainder of 10 mile EPZ.
- Consider issuance of POTASSIUM IODIDE in accordance with the State
- Plan, AZ. 82, F2. 02. ES, F5. G5 Peer Checked by
EPIP-5 BFN GENERAL EMERGENCY Rev0044 UnitO Page10of26 APPENDIX B Page 2 of 3 Activation of the Emergency Response Organization (ERO)
[3]
iF... unable to establish contact with the ODS THEN... continue to perform Step 1.0 f2] for 5 minutes 4]
IF... unable to establish contact with the ODS after 5 minutes THEN., IMPLEMENT Appendix I, Activation of the Emergency Paging System concurrently with this Appendix beginning at Step 2.0,
Clarification Guidance for SRO-only Questions RevI (0311112010)
F.
Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)]
Some examples of SRO exam items for this topic include:
Q Evaluating core conditions and emergency classifications based on core conditions.
Administrative requirements associated with low power physics testing processes.
Administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities.
Administrative controls associated with the installation of neutron sources.
Knowledge of TS bases for reactivity controls.
BEN 1006 #99 SRO 3
ii G2.4.40 4.5 Besides classification, which ONE of the following duties can NOT be delegated to another emergency team member by the SM/SED?
A.
Make notifications to the state.
B.
Direct the shutdown of the plant.
C.
Conduct site accountability actions.
D.
Determine Protective Action Recommendations.
L Proposed Answer: D Explanation (Optional):
A INCORRECT: The Operations Duty Specialist (ODS) should be notified by the SM/SED within five minutes of the event classification. The ODS relays the information to the Emergency Duty Officer (EDO), the State of Alabama, and the CECC Director. The EDO keeps the CECC Director informed of the situation as necessary.
B INCORRECT: If the event is determined to be one of the four emergency classifications, the Shift Manager assumes the responsibility of SED until relieved by the Plant Manager or designee.
C INCORRECT: The Shift Manager or SED shall make the decision to activate the assembly and accountability process.
The actions carried out as a result of this decision can be delegated but the decision itself cannot be delegated.
Examination Outline Cross-reference:
G2.4.40 (1 OCFR 55.43.5 SRO Only)
Knowledge of SROs responsibilities in emergency plan implementation.
Level Tier#
Group#
KIA#
Importanc e Rating Proposed Question: #
99 The Shift Manager / Site Emergency. The Central Emergency Director (SM/SED) has declared a General Emergency Control Center (CECC) is NOT staffed.
D CORRECT: Per EPIP-5, General Emergency, The Site Emergency Director must make any required recommendations until the CECC is staffed. This responsibility cannot be delegated until CECC is in operation. Recommendations are required at General Emergency.
QUESTION 100 An ATWS has occurred on Unit I with the following conditions:
An Alert has been declared The On-Call SED has assumed the duties of the SED.
There are no radiological or other hazards in the Reactor Building Subsequently, it becomes necessary to perform I-EOI Appendix-IB, Vent and Depressurize the Scram Pilot Air Header.
Which ONE of the following completes both statements below?
The AUO will perform l-EOl Appendix-lB at the (1)
The AUO required to perform this appendix (2) requested through the TSC.
A.
(1) CRD catwalk above Hydraulic Control Units (2)is B.
(1) CRD catwalk above Hydraulic Control Units (2) is NOT C.
(1) 565 elevation north east at the CRD station (2) is D. (I) 565 elevation north east at the CRD station (2) is NOT Answer: C
Level:
3 Group #
Examination Outline Cross-Reference KIA#
G2.4.42 Importance Rating 2.4.42 Knowledge of emergency response facilities.
Explanation: C CORRECT: First Part: CORRECT-This is the correct location per appendix lB. Second Part: CORRECT-Once any REP classification has been declared, and the TSC activation is complete, the control room must request an OSC team through the TSC when it becomes necessary to dispatch an operator in to the reactor building. The candidate must know once the On-call SED has assumed responsibilities of the SED that TSC activation is complete.
A Incorrect: First Part: Incorrect-plausible because this is the location for AUO actions in 1-EOI Appendix IE for venting CRD over piston area. Second Part: Correct-see C.
B Incorrect. First Part: Incorrect-See A. Second Part: Incorrect-this is plausible as AUOs are normally directed from the MCR. In addition, the MCR may direct AUOs in to the DO rooms, Electric Board Rooms, and Control Bay as long as there are no radiological or other hazards in those areas D
Incorrect. First Part: Correct-See C
- Second Part: Incorrect-See B.
Technical Reference(s): BFN-ODM-4.2; I -EOI Appendix-i B; 1 -EOI Appendix-i E;EPIP-6 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OPL171.206 ILT Obj. l.a Question Source:
Bank:
Modified Bank:
X New:
Question History:
Previous NRC: BFN 1108 #75 Question Cognitive Level:
Memory or Fundamental Knowledge:
Comprehension or Analysis : X 10 CFR Part 55 Content:
55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
1-EOI Appendix-IS Rev I BFN VENTING AND REPRESSLJRIZING 1-EOI APPENDIX-lB UNIT I THE SCRAM PILOT AIR HEADER Rev. I Page 1 of 3 fLOCATIQN:
Unit 1 RB NE, El 565 ft, 1-LPNL-925-0018B I AHACHMENTS:
1.
Tools and Equipment
( I 1-EQI Appendix-1E Rev I BFN MANUAL INSERTION OF CONTROL RODS I-EOI APPENDIX-IE UNIT I BY VENTING THE OVER PISTON AREA Rev. I Page 2 of 7 4.a (Continued)
NOTE A ladder may be required to perform the following steps. Refer to Attachment 1.
b.
UNLOCK and CLOSE 1-SHV-085-615, WITHDRAW RISER Soy.
c.
OBTAIN necessary tools and equipment arid PERFORM the following from the catwalk area:
1)
CONNECT quick disconnect end of vent hose to coupling 1-TV-085-623, WITHDRAW RISER VENT BLOCK TEST VLV.
BFN-ODM-4.2 Rev 2 BFN petians Radiak>cai Emergency Plan (REP BFN4>tL2 Directive Manual Assinmens Rev 0002 Par5 of S ii ODM-42 DIRECThfE (continuedi C.
Control Room Dispath of ar5andArea Evcuationuihortty following REP Event CLassiiication:
May Conirol Room Order May Area Evaouatboo Control Room Order Due To Dueot Dlepalch OfOperator&..
Rathazion-Lea?
ftf :4flfl Hae Been DecLared.
YES CDT Y
AND YES TC Antadon IS NOT ConIete, Or itaa MT rcf Been Ordered Jnto YES Canfrol tj Bay?
treODM.
YES Uefa typfr O.L Board Rooma?
c+ot r
tl!n rT CSC hTu1t ftEI :33jQfl flee B YES AND NO rtçed bc ere COM.
75 UTT TC
.Jrto co iste
erwator ctc4L r rn CC em tT iCC ThTi
.Jnio 4JY Other NO R1IoIofc Confrolled Or Cac trn 41acted Areas jaB. fl. etc4?
EPP-6 Rev 34 BFN ACTIVATION AND OPERATION OF ThE TECHNICAL EPIP6 SUPPORT CENTER TSq Rev 0034 Page ief 49
1.0 INTRODUCTION
1.1 Purpose The purpose of this procedure is to describe a tialion of the Technical Support Center (15G.), define the TSC organization and provide for TSC operations by defining.taff responsitrictes.
20 REFERENCES 2.1 lndust Documents A. NUREG-0554. Criteria for Preparation and Evaluation of Radiologicat Emergency Response Plans and Preparedness in Support of Nuclear Power Plants B. 10 CFR 50.47, Code :f Federal Regulations 2.2 Plant lnstnctons A. IVA Radiological Emergency Plan B. Emergency Plan implementing Procedure EPIP)
- 1. Emergency Classification Procedure C. EP1P -2, Notification of Unusual Event
- 0. EPIP - 16, Termination and Recovery H. EPIP-i 5, Emergency Exposure5 I.
EPIP-1 1, Security and Aoess Control 3.0 INSTRUGTIDNS
- 3. 1 Activation The TSC is required to be activated at the Alert or higher emergency classification, however, activation can occur at the discretion of the Shift Manager (SM). Once an emergency classification has been declared, the SM becomes the Site Emergency Director (SED).
Depending upon the emergency classification declared, steps to activate the TSC are specified in the applicable EPIP for the emergency classification, When the TSC is activated, the on-call SED will obtain a turnover from the SMISED, ensure that minimum staffing is met for the emergency center, and assume the responsibilities of the SED from the SMSED.
Once the responsibilities of the SMISED have been assumed by the on-call SED, command and control of the emergency response transfers to the 13G. TSC activation lime is defined in the Radiological Emergency Plan.
BFN 1108 NRC
- 75. 62.4,35 NEW/H Unit 1 has experienced an A]WS and RPS cannot immediately be de-energized.
The US has dispatched an AUO to perform 1-EOl Appendix-I B. Vent and Depressunze the Scram Pilot Air Header.
Which ONE of the following completes both statements below?
The AUO will perform l-EOi Appendix-lB _tl),
The AUC will vent the scram header at _(2)_.
A.
(1) at the CR0 catwalk above Hydraulic Control Units (2) the pressure switch used for the SCRAM PILOT AIR HEADER PRESS LOW (1-9-58, Window 28) annunciator.
B.
(1) at the 565 elevation north east at the CRD station (2) one of the 3-way Alternate Rod insertion (A RI) solenoid Valves.
C.
(I) at the 565 elevation north east at the CR0 station (2) the pressure switch used for the SCRAM PILOT AIR HEADER PRESS LOW (1-9-58, Window 28) annunciator.
D.
(1) at the CR0 catwalk above Hydraulic Control Units (2) one of the 3-way Alternate Rod Insertion (ARI) solenoid valves.
CORRECT ANSWER C Tier 3: Generic.
2.435. Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR: 41 10!43.5J4513) RO IR: 3.8 Plausibility; The 1st part of A and 0 is plausible because this is the location for AUO actions in 1-EOl Appendix I E for venting CR0 over piston area. The 2nd part of B and 000 is plausible because the ARt valves are an alternate means to vent the scram air header.
References Unit 1 EOI flow chart 1-EOl Appendix 18 1-EOl Appendix 10 1-ARP-9-58 OPL171.205 Lesson Plan :DbjectPjes OPLI71,206 Objective D.1.b
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowina systems knowledge. i.e, how the system works, Yes RO uetion flowpath, logic, component location?
q Can the question be answered solely by knowing immediate operator actions?
Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RC) question that require direct entry to major EOPs?
G Can the queshon be answered solely by knowing the purpose. overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?
Does the question require one or more of the following?
Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Knowledge of when to implement attachments and appendices including how to coordinate these items with procedure steps e
SRO-only Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only