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Unit Two is in MODE 5 performing fuel movements.
Unit Two is in MODE 5 performing fuel movements.
Which one of the following completes both statements below lAW Unit One Tech Spec 3.8.1, AC Sources Operating, LCO statement?
Which one of the following completes both statements below lAW Unit One Tech Spec 3.8.1, AC Sources Operating, LCO statement?
                                              -
The Unit Two SAT      (1)    required to be OPERABLE.
The Unit Two SAT      (1)    required to be OPERABLE.
(2)  Diesel Generators are required to be OPERABLE.
(2)  Diesel Generators are required to be OPERABLE.
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Page: 31
Page: 31
: 32. A Unit Two plant cooldown is being performed with the following plant conditions:
: 32. A Unit Two plant cooldown is being performed with the following plant conditions:
Reactor water level        175 inches, steady Reactor pressure band    500 700 psig
Reactor water level        175 inches, steady Reactor pressure band    500 700 psig Drywell ref leg temp      175°F (REFERENCE PROVIDED)
                                      -
Drywell ref leg temp      175°F (REFERENCE PROVIDED)
Which one of the following completes both statements below?
Which one of the following completes both statements below?
The lowering of reactor pressure causes the NOO4AIBIC (Narrow Range) reactor water level instruments indicated level error to (1)
The lowering of reactor pressure causes the NOO4AIBIC (Narrow Range) reactor water level instruments indicated level error to (1)
Line 919: Line 916:
(6) PWR auxiliary or emergency feedwater              (6) PWR auxiliary or emergency feedwater system.                                              system.
(6) PWR auxiliary or emergency feedwater              (6) PWR auxiliary or emergency feedwater system.                                              system.
(7) Containment heat removal and                      (7) Containment heat removal and depressurization systems, including                  depressurization systems, including containment spray and fan cooler systems.            containment spray and fan cooler systems.
(7) Containment heat removal and                      (7) Containment heat removal and depressurization systems, including                  depressurization systems, including containment spray and fan cooler systems.            containment spray and fan cooler systems.
__________
5 Actuation of the RPS when the reactor is critical is reportable under § 50.72(b)(2)(iv)(B).
5 Actuation of the RPS when the reactor is critical is reportable under § 50.72(b)(2)(iv)(B).
5
5
Line 996: Line 992:
----------------------------------------------------------- NOTE -----------------------------------------------------------
----------------------------------------------------------- NOTE -----------------------------------------------------------
Separate Condition entry is allowed for each channel.
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 hours channels inoperable.
CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A. One or more required                    A.1          Place channel in trip.                12 hours channels inoperable.
OR A.2          ---------------NOTE------------- 12 hours Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.
OR A.2          ---------------NOTE------------- 12 hours Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.
                                                            ------------------------------------
Place associated trip system in trip.
Place associated trip system in trip.
(continued)
(continued)
Line 1,029: Line 1,023:
: 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
: 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
: 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.
: 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.
-------------------------------------------------------------------------------------------------------------------------------
SURVEILLANCE                                                      FREQUENCY SR 3.3.1.1.1            (Not used.)                                                                12 hours SR 3.3.1.1.2            Perform CHANNEL CHECK.                                                    24 hours SR 3.3.1.1.3            --------------------------------NOTE--------------------------------
SURVEILLANCE                                                      FREQUENCY SR 3.3.1.1.1            (Not used.)                                                                12 hours SR 3.3.1.1.2            Perform CHANNEL CHECK.                                                    24 hours SR 3.3.1.1.3            --------------------------------NOTE--------------------------------
Not required to be performed until 12 hours after THERMAL POWER  23% RTP.
Not required to be performed until 12 hours after THERMAL POWER  23% RTP.
                          ------------------------------------------------------------------------
Adjust the average power range monitor (APRM)                              7 days channels to conform to the calculated power while operating at  23% RTP.
Adjust the average power range monitor (APRM)                              7 days channels to conform to the calculated power while operating at  23% RTP.
SR 3.3.1.1.4            --------------------------------NOTE--------------------------------
SR 3.3.1.1.4            --------------------------------NOTE--------------------------------
Not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
Not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
                          ------------------------------------------------------------------------
Perform CHANNEL FUNCTIONAL TEST.                                          7 days (continued)
Perform CHANNEL FUNCTIONAL TEST.                                          7 days (continued)
Brunswick Unit 2                                            3.3-4                                  Amendment No. 247
Brunswick Unit 2                                            3.3-4                                  Amendment No. 247
Line 1,044: Line 1,035:
SR 3.3.1.1.6    Verify the source range monitor (SRM) and                                Prior to withdrawing intermediate range monitor (IRM) channels overlap.                      SRMs from the fully inserted position SR 3.3.1.1.7    --------------------------------NOTE--------------------------------
SR 3.3.1.1.6    Verify the source range monitor (SRM) and                                Prior to withdrawing intermediate range monitor (IRM) channels overlap.                      SRMs from the fully inserted position SR 3.3.1.1.7    --------------------------------NOTE--------------------------------
Only required to be met during entry into MODE 2 from MODE 1.
Only required to be met during entry into MODE 2 from MODE 1.
                ------------------------------------------------------------------------
Verify the IRM and APRM channels overlap.                                7 days SR 3.3.1.1.8    Calibrate the local power range monitors.                                2000 effective full power hours SR 3.3.1.1.9    Perform CHANNEL FUNCTIONAL TEST.                                        92 days SR 3.3.1.1.10  Calibrate the trip units.                                                92 days (continued)
Verify the IRM and APRM channels overlap.                                7 days SR 3.3.1.1.8    Calibrate the local power range monitors.                                2000 effective full power hours SR 3.3.1.1.9    Perform CHANNEL FUNCTIONAL TEST.                                        92 days SR 3.3.1.1.10  Calibrate the trip units.                                                92 days (continued)
Brunswick Unit 2                                  3.3-5                                  Amendment No. 282
Brunswick Unit 2                                  3.3-5                                  Amendment No. 282
Line 1,052: Line 1,042:
: 1.      For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
: 1.      For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
: 2.      For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.
: 2.      For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.
                ------------------------------------------------------------------------
Perform CHANNEL FUNCTIONAL TEST.                                          184 days SR 3.3.1.1.12  Perform CHANNEL FUNCTIONAL TEST.                                          24 months SR 3.3.1.1.13  --------------------------------NOTES------------------------------
Perform CHANNEL FUNCTIONAL TEST.                                          184 days SR 3.3.1.1.12  Perform CHANNEL FUNCTIONAL TEST.                                          24 months SR 3.3.1.1.13  --------------------------------NOTES------------------------------
: 1.      Neutron detectors are excluded.
: 1.      Neutron detectors are excluded.
: 2.      For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
: 2.      For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
: 3.      For Functions 2.b and 2.f, the recirculation flow transmitters that feed the APRMs are included.
: 3.      For Functions 2.b and 2.f, the recirculation flow transmitters that feed the APRMs are included.
                -------------------------------------------------------------------------
Perform CHANNEL CALIBRATION.                                              24 months SR 3.3.1.1.14  (Not used.)
Perform CHANNEL CALIBRATION.                                              24 months SR 3.3.1.1.14  (Not used.)
SR 3.3.1.1.15  Perform LOGIC SYSTEM FUNCTIONAL TEST.                                    24 months (continued)
SR 3.3.1.1.15  Perform LOGIC SYSTEM FUNCTIONAL TEST.                                    24 months (continued)
Line 1,069: Line 1,057:
: 3.      For Function 5, "n" equals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.
: 3.      For Function 5, "n" equals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.
: 4.      For Function 2.e, "n" equals 8 channels for the purpose of determining the STAGGERED TEST BASIS Frequency. Testing of APRM and Oscillation Power Range Monitor (OPRM) outputs shall alternate.
: 4.      For Function 2.e, "n" equals 8 channels for the purpose of determining the STAGGERED TEST BASIS Frequency. Testing of APRM and Oscillation Power Range Monitor (OPRM) outputs shall alternate.
                ----------------------------------------------------------------------
24 months on a Verify the RPS RESPONSE TIME is within limits.                        STAGGERED TEST BASIS SR 3.3.1.1.18  Adjust the flow control trip reference card to conform                Once within 7 days to reactor flow.                                                      after reaching equilibrium conditions following refueling outage (continued)
24 months on a Verify the RPS RESPONSE TIME is within limits.                        STAGGERED TEST BASIS SR 3.3.1.1.18  Adjust the flow control trip reference card to conform                Once within 7 days to reactor flow.                                                      after reaching equilibrium conditions following refueling outage (continued)
Brunswick Unit 2                                  3.3-7                                Amendment No. 247
Brunswick Unit 2                                  3.3-7                                Amendment No. 247
Line 1,119: Line 1,106:
----------------------------------------------------------- NOTE -----------------------------------------------------------
----------------------------------------------------------- NOTE -----------------------------------------------------------
Separate Condition entry is allowed for each channel.
Separate Condition entry is allowed for each channel.
-------------------------------------------------------------------------------------------------------------------------------
CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A. One or more channels                    A.1          Enter the Condition                    Immediately inoperable.                                          referenced in Table 3.3.5.1-1 for the channel.
CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A. One or more channels                    A.1          Enter the Condition                    Immediately inoperable.                                          referenced in Table 3.3.5.1-1 for the channel.
B. As required by Required                B.1          -------------NOTES-------------
B. As required by Required                B.1          -------------NOTES-------------
Action A.1 and referenced in                          1. Only applicable in Table 3.3.5.1-1.                                            MODES 1, 2, and 3.
Action A.1 and referenced in                          1. Only applicable in Table 3.3.5.1-1.                                            MODES 1, 2, and 3.
: 2. Only applicable for Functions 1.a, 1.b, 2.a, and 2.b.
: 2. Only applicable for Functions 1.a, 1.b, 2.a, and 2.b.
                                                            ------------------------------------
Declare supported                      1 hour from feature(s) inoperable when            discovery of loss of its redundant feature ECCS            initiation capability initiation capability is              for feature(s) in both inoperable.                            divisions AND (continued)
Declare supported                      1 hour from feature(s) inoperable when            discovery of loss of its redundant feature ECCS            initiation capability initiation capability is              for feature(s) in both inoperable.                            divisions AND (continued)
Brunswick Unit 2                                            3.3-35                                  Amendment No. 233
Brunswick Unit 2                                            3.3-35                                  Amendment No. 233
Line 1,130: Line 1,115:
ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION                  REQUIRED ACTION                        COMPLETION TIME B.  (continued)                  B.2  --------------NOTE--------------
ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION                  REQUIRED ACTION                        COMPLETION TIME B.  (continued)                  B.2  --------------NOTE--------------
Only applicable for Functions 3.a and 3.b.
Only applicable for Functions 3.a and 3.b.
                                      ------------------------------------
Declare High Pressure                1 hour from Coolant Injection (HPCI)              discovery of loss of System inoperable.                    HPCI initiation capability AND B.3  Place channel in trip.                24 hours C. As required by Required      C.1  ------------NOTES--------------
Declare High Pressure                1 hour from Coolant Injection (HPCI)              discovery of loss of System inoperable.                    HPCI initiation capability AND B.3  Place channel in trip.                24 hours C. As required by Required      C.1  ------------NOTES--------------
Action A.1 and referenced in      1. Only applicable in Table 3.3.5.1-1.                        MODES 1, 2, and 3.
Action A.1 and referenced in      1. Only applicable in Table 3.3.5.1-1.                        MODES 1, 2, and 3.
: 2. Only applicable for Functions 1.c, 1.d, 2.c, 2.d, and 2.f.
: 2. Only applicable for Functions 1.c, 1.d, 2.c, 2.d, and 2.f.
                                      -------------------------------------
Declare supported                    1 hour from feature(s) inoperable when            discovery of loss of its redundant feature ECCS            initiation capability initiation capability is              for feature(s) in both inoperable.                          divisions AND C.2  Restore channel to                    24 hours OPERABLE status.
Declare supported                    1 hour from feature(s) inoperable when            discovery of loss of its redundant feature ECCS            initiation capability initiation capability is              for feature(s) in both inoperable.                          divisions AND C.2  Restore channel to                    24 hours OPERABLE status.
(continued)
(continued)
Line 1,142: Line 1,125:
CONDITION                    REQUIRED ACTION                        COMPLETION TIME D. As required by Required      D.1    --------------NOTE--------------
CONDITION                    REQUIRED ACTION                        COMPLETION TIME D. As required by Required      D.1    --------------NOTE--------------
Action A.1 and referenced in        Only applicable if HPCI Table 3.3.5.1-1                    pump suction is not aligned to the suppression pool.
Action A.1 and referenced in        Only applicable if HPCI Table 3.3.5.1-1                    pump suction is not aligned to the suppression pool.
                                        -------------------------------------
Declare HPCI System                  1 hour from inoperable.                          discovery of loss of HPCI initiation capability AND D.2.1  Place channel in trip.                24 hours OR D.2.2  Align the HPCI pump                  24 hours suction to the suppression pool.
Declare HPCI System                  1 hour from inoperable.                          discovery of loss of HPCI initiation capability AND D.2.1  Place channel in trip.                24 hours OR D.2.2  Align the HPCI pump                  24 hours suction to the suppression pool.
(continued)
(continued)
Line 1,160: Line 1,142:
: 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
: 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
: 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 3.c; and (b) for up to 6 hours for Functions other than 3.c provided the associated Function or the redundant Function maintains ECCS initiation capability.
: 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 3.c; and (b) for up to 6 hours for Functions other than 3.c provided the associated Function or the redundant Function maintains ECCS initiation capability.
-------------------------------------------------------------------------------------------------------------------------------
SURVEILLANCE                                                      FREQUENCY SR 3.3.5.1.1            Perform CHANNEL CHECK.                                                    24 hours SR 3.3.5.1.2            Perform CHANNEL FUNCTIONAL TEST.                                          92 days SR 3.3.5.1.3            Calibrate the trip unit.                                                  92 days SR 3.3.5.1.4            Perform CHANNEL CALIBRATION.                                              24 months SR 3.3.5.1.5            Perform LOGIC SYSTEM FUNCTIONAL TEST.                                      24 months SR 3.3.5.1.6            Perform CHANNEL FUNCTIONAL TEST.                                          24 months Brunswick Unit 2                                          3.3-40                                  Amendment No. 233
SURVEILLANCE                                                      FREQUENCY SR 3.3.5.1.1            Perform CHANNEL CHECK.                                                    24 hours SR 3.3.5.1.2            Perform CHANNEL FUNCTIONAL TEST.                                          92 days SR 3.3.5.1.3            Calibrate the trip unit.                                                  92 days SR 3.3.5.1.4            Perform CHANNEL CALIBRATION.                                              24 months SR 3.3.5.1.5            Perform LOGIC SYSTEM FUNCTIONAL TEST.                                      24 months SR 3.3.5.1.6            Perform CHANNEL FUNCTIONAL TEST.                                          24 months Brunswick Unit 2                                          3.3-40                                  Amendment No. 233


Line 1,232: Line 1,213:
: 4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment,"
: 4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment,"
when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.
when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.
-------------------------------------------------------------------------------------------------------------------------------
CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A.    --------------NOTE-------------        A.1          Isolate the affected                  8 hours Only applicable to                                    penetration flow path by penetration flow paths with                          use of at least one closed two PCIVs.                                            and de-activated automatic
CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A.    --------------NOTE-------------        A.1          Isolate the affected                  8 hours Only applicable to                                    penetration flow path by penetration flow paths with                          use of at least one closed two PCIVs.                                            and de-activated automatic
       ------------------------------------                  valve, closed manual valve, blind flange, or check valve One or more penetration                              with flow through the valve flow paths with one PCIV                              secured.
       ------------------------------------                  valve, closed manual valve, blind flange, or check valve One or more penetration                              with flow through the valve flow paths with one PCIV                              secured.
Line 1,240: Line 1,220:
PCIVs 3.6.1.3 ACTIONS CONDITION    REQUIRED ACTION                      COMPLETION TIME A.  (continued)    A.2  --------------NOTE--------------
PCIVs 3.6.1.3 ACTIONS CONDITION    REQUIRED ACTION                      COMPLETION TIME A.  (continued)    A.2  --------------NOTE--------------
Isolation devices in high radiation areas may be verified by use of administrative means.
Isolation devices in high radiation areas may be verified by use of administrative means.
                          -----------------------------------
Verify the affected                Once per 31 days penetration flow path is            for isolation devices isolated.                          outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment (continued)
Verify the affected                Once per 31 days penetration flow path is            for isolation devices isolated.                          outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment (continued)
Brunswick Unit 2          3.6-8                                Amendment No. 233
Brunswick Unit 2          3.6-8                                Amendment No. 233
Line 1,253: Line 1,232:
C.2  --------------NOTE--------------
C.2  --------------NOTE--------------
Isolation devices in high radiation areas may be verified by use of administrative means.
Isolation devices in high radiation areas may be verified by use of administrative means.
                                              ------------------------------------
Verify the affected                  Once per 31 days penetration flow path is isolated.
Verify the affected                  Once per 31 days penetration flow path is isolated.
(continued)
(continued)
Line 1,273: Line 1,251:
----------------------------------------------------------NOTE ----------------------------------------------------------
----------------------------------------------------------NOTE ----------------------------------------------------------
LCO 3.0.4.b is not applicable to DGs.
LCO 3.0.4.b is not applicable to DGs.
-----------------------------------------------------------------------------------------------------------------------------
CONDITION                                REQUIRED ACTION                          COMPLETION TIME A.    ----------------NOTE------------- A.1              Restore Unit 2 offsite circuit 45 days Only applicable when Unit 2                        to OPERABLE status.
CONDITION                                REQUIRED ACTION                          COMPLETION TIME A.    ----------------NOTE------------- A.1              Restore Unit 2 offsite circuit 45 days Only applicable when Unit 2                        to OPERABLE status.
is in MODE 4 or 5.
is in MODE 4 or 5.
      -------------------------------------
One Unit 2 offsite circuit inoperable.
One Unit 2 offsite circuit inoperable.
(continued)
(continued)
Line 1,285: Line 1,261:
: 1. Only applicable when                    with no power available      discovery of Unit 2 is in MODE 4                  inoperable when the          Condition B or 5.                                redundant required          concurrent with feature(s) are inoperable. inoperability of
: 1. Only applicable when                    with no power available      discovery of Unit 2 is in MODE 4                  inoperable when the          Condition B or 5.                                redundant required          concurrent with feature(s) are inoperable. inoperability of
: 2. Condition B shall not be                                          redundant required entered in conjunction                                            feature(s) with Condition A.
: 2. Condition B shall not be                                          redundant required entered in conjunction                                            feature(s) with Condition A.
    --------------------------------------
Two Unit 2 offsite circuits inoperable due to one Unit 2 balance of plant circuit path to the downstream 4.16 kV emergency bus inoperable for planned maintenance.
Two Unit 2 offsite circuits inoperable due to one Unit 2 balance of plant circuit path to the downstream 4.16 kV emergency bus inoperable for planned maintenance.
AND                                    AND DG associated with the                B.2  Perform SR 3.8.1.1 for      2 hours affected downstream                        OPERABLE offsite 4.16 kV emergency bus                      circuit(s).                  AND inoperable for planned maintenance.                                                            Once per 12 hours thereafter AND B.3  Restore both Unit 2 offsite  7 days circuits and DG to OPERABLE status.            AND 10 days from discovery of failure to meet LCO 3.8.1.a or b (continued)
AND                                    AND DG associated with the                B.2  Perform SR 3.8.1.1 for      2 hours affected downstream                        OPERABLE offsite 4.16 kV emergency bus                      circuit(s).                  AND inoperable for planned maintenance.                                                            Once per 12 hours thereafter AND B.3  Restore both Unit 2 offsite  7 days circuits and DG to OPERABLE status.            AND 10 days from discovery of failure to meet LCO 3.8.1.a or b (continued)
Line 1,328: Line 1,303:
: 2.      A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR. When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.
: 2.      A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR. When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.
: 3.      A single test at the specified Frequency will satisfy this Surveillance for both units.
: 3.      A single test at the specified Frequency will satisfy this Surveillance for both units.
                -------------------------------------------------------------------------
Verify each DG starts from standby conditions and                          31 days achieves steady state voltage  3750 V and  4300 V and frequency  58.8 Hz and  61.2 Hz.
Verify each DG starts from standby conditions and                          31 days achieves steady state voltage  3750 V and  4300 V and frequency  58.8 Hz and  61.2 Hz.
(continued)
(continued)
Line 1,340: Line 1,314:
: 4.      This SR shall be preceded by and immediately follow, without shutdown, a successful performance of SR 3.8.1.2 or SR 3.8.1.7.
: 4.      This SR shall be preceded by and immediately follow, without shutdown, a successful performance of SR 3.8.1.2 or SR 3.8.1.7.
: 5.      A single test at the specified Frequency will satisfy this Surveillance for both units.
: 5.      A single test at the specified Frequency will satisfy this Surveillance for both units.
                -------------------------------------------------------------------------
Verify each DG is synchronized and loaded and                            31 days operates for  60 minutes at a load  2800 kW and 3500 kW.
Verify each DG is synchronized and loaded and                            31 days operates for  60 minutes at a load  2800 kW and 3500 kW.
SR 3.8.1.4      Verify each engine mounted tank contains  150 gal of 31 days fuel oil.
SR 3.8.1.4      Verify each engine mounted tank contains  150 gal of 31 days fuel oil.
Line 1,352: Line 1,325:
: 1.      All DG starts may be preceded by an engine prelube period.
: 1.      All DG starts may be preceded by an engine prelube period.
: 2.      A single test at the specified Frequency will satisfy this Surveillance for both units.
: 2.      A single test at the specified Frequency will satisfy this Surveillance for both units.
                -------------------------------------------------------------------------
Verify each DG starts from standby condition and                          184 days achieves, in  10 seconds, voltage  3750 V and frequency  58.8 Hz, and after steady state conditions are reached, maintains voltage  3750 V and  4300 V and frequency  58.8 Hz and  61.2 Hz.
Verify each DG starts from standby condition and                          184 days achieves, in  10 seconds, voltage  3750 V and frequency  58.8 Hz, and after steady state conditions are reached, maintains voltage  3750 V and  4300 V and frequency  58.8 Hz and  61.2 Hz.
(continued)
(continued)
Line 1,362: Line 1,334:
: 2.      SR 3.8.1.8.a is not required to be met if the unit power supply is from the preferred offsite circuit.
: 2.      SR 3.8.1.8.a is not required to be met if the unit power supply is from the preferred offsite circuit.
: 3.      A single test at the specified Frequency will satisfy this Surveillance for both units.
: 3.      A single test at the specified Frequency will satisfy this Surveillance for both units.
                -------------------------------------------------------------------------
Verify:                                                                  24 months
Verify:                                                                  24 months
: a.      Automatic transfer capability of the unit power supply from the normal circuit to the preferred offsite circuit; and
: a.      Automatic transfer capability of the unit power supply from the normal circuit to the preferred offsite circuit; and
Line 1,374: Line 1,345:
: 2.      If performed with the DG synchronized with offsite power, it shall be performed at a power factor  0.9.
: 2.      If performed with the DG synchronized with offsite power, it shall be performed at a power factor  0.9.
: 3.      A single test at the specified Frequency will satisfy this Surveillance for both units.
: 3.      A single test at the specified Frequency will satisfy this Surveillance for both units.
                -------------------------------------------------------------------------
Verify each DG rejects a load greater than or equal to                    24 months its associated core spray pump without tripping.
Verify each DG rejects a load greater than or equal to                    24 months its associated core spray pump without tripping.
(continued)
(continued)
Line 1,382: Line 1,352:
SURVEILLANCE                                                  FREQUENCY SR 3.8.1.10    -------------------------------NOTE---------------------------------
SURVEILLANCE                                                  FREQUENCY SR 3.8.1.10    -------------------------------NOTE---------------------------------
A single test at the specified Frequency will satisfy this Surveillance for both units.
A single test at the specified Frequency will satisfy this Surveillance for both units.
                -----------------------------------------------------------------------
Verify each DG's automatic trips are bypassed on an                    24 months actual or simulated ECCS initiation signal except:
Verify each DG's automatic trips are bypassed on an                    24 months actual or simulated ECCS initiation signal except:
: a.      Engine overspeed;
: a.      Engine overspeed;
Line 1,397: Line 1,366:
: 1.      Momentary transients outside the load and power factor ranges do not invalidate this test.
: 1.      Momentary transients outside the load and power factor ranges do not invalidate this test.
: 2.      A single test at the specified Frequency will satisfy this Surveillance for both units.
: 2.      A single test at the specified Frequency will satisfy this Surveillance for both units.
                -------------------------------------------------------------------------
Verify each DG operating at a power factor  0.9                          24 months operates for  60 minutes loaded to  3500 kW and 3850 kW.
Verify each DG operating at a power factor  0.9                          24 months operates for  60 minutes loaded to  3500 kW and 3850 kW.
SR 3.8.1.12    -------------------------------NOTE---------------------------------
SR 3.8.1.12    -------------------------------NOTE---------------------------------
A single test at the specified Frequency will satisfy this Surveillance for both units.
A single test at the specified Frequency will satisfy this Surveillance for both units.
                -----------------------------------------------------------------------
Verify an actual or simulated ECCS initiation signal is                  24 months capable of overriding the test mode feature to return each DG to ready-to-load operation.
Verify an actual or simulated ECCS initiation signal is                  24 months capable of overriding the test mode feature to return each DG to ready-to-load operation.
(continued)
(continued)
Line 1,409: Line 1,376:
SURVEILLANCE                                                FREQUENCY SR 3.8.1.13    -------------------------------NOTE---------------------------------
SURVEILLANCE                                                FREQUENCY SR 3.8.1.13    -------------------------------NOTE---------------------------------
This Surveillance shall not be performed in MODE 1, 2, or 3 for the load sequence relays associated with DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.
This Surveillance shall not be performed in MODE 1, 2, or 3 for the load sequence relays associated with DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.
                -----------------------------------------------------------------------
Verify interval between each sequenced load block is 24 months within +/- 10% of design interval for each load sequence relay.
Verify interval between each sequenced load block is 24 months within +/- 10% of design interval for each load sequence relay.
(continued)
(continued)
Line 1,418: Line 1,384:
: 1.      All DG starts may be preceded by an engine prelube period.
: 1.      All DG starts may be preceded by an engine prelube period.
: 2.      This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.
: 2.      This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.
                -------------------------------------------------------------------------
Verify, on actual or simulated loss of offsite power                      24 months signal in conjunction with an actual or simulated ECCS initiation signal:
Verify, on actual or simulated loss of offsite power                      24 months signal in conjunction with an actual or simulated ECCS initiation signal:
: a.      De-energization of emergency buses;
: a.      De-energization of emergency buses;
Line 2,071: Line 2,036:
==Reference:==
==Reference:==
None Cog Level: Fundamental Explanation:  DFCS will be placed into service with the manual output set at 2550 RPM. The DFCS system will control the RFPT speed from 2450 5450 RPMs
None Cog Level: Fundamental Explanation:  DFCS will be placed into service with the manual output set at 2550 RPM. The DFCS system will control the RFPT speed from 2450 5450 RPMs
                                                            -


Distractor Analysis:
Distractor Analysis:
Line 2,110: Line 2,074:


Distractor Analysis:
Distractor Analysis:
Choice A: Correct Answer, see explanation Choice B:    Plausible because iF does auto close and SBGT Train lA/B Suction Valves (1C & 1E) on Unit One only do not auto open Choice C:    Incorrect since SBGT will auto realign from primary containment to the Reactor Building on system initiation Choice D:    Incorrect since SBGT will auto realign from primary containment to the Reactor Building on system initiation and SBGT Train lA/B Suction Valves (1 C & 1 E) on Unit One only do not auto open SRO Basis:      N/A 2.1.6        Fan A 100% capacity, heavy-duly, industrial type Fan and motor assembly is provided in each SBGT filter train. Each Fan will produce the required 2700 3300 scfm flow through its associated filter tmin
Choice A: Correct Answer, see explanation Choice B:    Plausible because iF does auto close and SBGT Train lA/B Suction Valves (1C & 1E) on Unit One only do not auto open Choice C:    Incorrect since SBGT will auto realign from primary containment to the Reactor Building on system initiation Choice D:    Incorrect since SBGT will auto realign from primary containment to the Reactor Building on system initiation and SBGT Train lA/B Suction Valves (1 C & 1 E) on Unit One only do not auto open SRO Basis:      N/A 2.1.6        Fan A 100% capacity, heavy-duly, industrial type Fan and motor assembly is provided in each SBGT filter train. Each Fan will produce the required 2700 3300 scfm flow through its associated filter tmin Each Fan is driven by a direct-drive AC motor which is energized from a redundant and separate emergency power supply. The Unit I A and 5 Fans are powered from 480 VAC MCCs I XE and 1 XF respectively and Unit 2 A and B Fans from 2XE and 2XF.
                              -
Each Fan is driven by a direct-drive AC motor which is energized from a redundant and separate emergency power supply. The Unit I A and 5 Fans are powered from 480 VAC MCCs I XE and 1 XF respectively and Unit 2 A and B Fans from 2XE and 2XF.
The filter train fans may be operated manually from controls located at RTGB XU-51.
The filter train fans may be operated manually from controls located at RTGB XU-51.
The filter train fans will automatically start if any of the following Secondary Containment isolation conditions exist: (Figure 10-2)
The filter train fans will automatically start if any of the following Secondary Containment isolation conditions exist: (Figure 10-2)
Line 2,122: Line 2,084:
Unit Two is in MODE 5 performing fuel movements.
Unit Two is in MODE 5 performing fuel movements.
Which one of the following completes both statements below lAW Unit One Tech Spec 3.8.1, AC Sources Operating, LCO statement?
Which one of the following completes both statements below lAW Unit One Tech Spec 3.8.1, AC Sources Operating, LCO statement?
                                                  -
The Unit Two SAT            (1)      required to be OPERABLE.
The Unit Two SAT            (1)      required to be OPERABLE.
(2)      Diesel Generators are required to be OPERABLE.
(2)      Diesel Generators are required to be OPERABLE.
Line 2,169: Line 2,130:


14fl.
14fl.
t)  r
t)  r r
...
r
           -]
           -]
: 29. 2150032 A reactor shutdown is in progress.
: 29. 2150032 A reactor shutdown is in progress.
Line 2,260: Line 2,219:
* Jacket Water Low pressure              12 psig The low lube oil pressure, and low jacket iwter pressure shutdowns are blocked for the first forty-five second on initiation of an engine start sequence (auto or manual). This permits the conditions to be established which will prevent these shutdowns during engine operation
* Jacket Water Low pressure              12 psig The low lube oil pressure, and low jacket iwter pressure shutdowns are blocked for the first forty-five second on initiation of an engine start sequence (auto or manual). This permits the conditions to be established which will prevent these shutdowns during engine operation
: 32. 216000 1 A Unit Two plant cooldown is being performed with the following plant conditions:
: 32. 216000 1 A Unit Two plant cooldown is being performed with the following plant conditions:
Reactor water level              175 inches, steady Reactor pressure band            500 700 psig
Reactor water level              175 inches, steady Reactor pressure band            500 700 psig Drywell ref leg temp            175°F (REFERENCE PROVIDED)
                                          -
Drywell ref leg temp            175°F (REFERENCE PROVIDED)
Which one of the following completes both statements below?
Which one of the following completes both statements below?
The lowering of reactor pressure causes the NOO4NB/C (Narrow Range) reactor water level instruments indicated level error to                (1)
The lowering of reactor pressure causes the NOO4NB/C (Narrow Range) reactor water level instruments indicated level error to                (1)
Line 2,555: Line 2,512:
                                   <<Group Isolation Checklist>>
                                   <<Group Isolation Checklist>>
Group 3 Isolation Sianals Signal                      Tech Spec Value                    Setpoint Value LowLevel2                    +101 inches                        +105 inches High Differential Flow        73 gpm                            43 gpm (after 285 minute time delay)
Group 3 Isolation Sianals Signal                      Tech Spec Value                    Setpoint Value LowLevel2                    +101 inches                        +105 inches High Differential Flow        73 gpm                            43 gpm (after 285 minute time delay)
Area High Temperature        150&deg;F                              140&deg;F Area Ventilation T High      50&deg;F                              47&deg;F Nan-Regen Hx Outlet          N/A                                135&deg;F Temp Hi SLC Initiation                N/A                                N/A RWCU Outside PumplHx          120&deg;F                              115&deg;F Rms RWCU Differential How        30 minutes                        28.5 minutes High Time Delay Group 3 Isolation Valves Control Room RTGB Panel H12-P607
Area High Temperature        150&deg;F                              140&deg;F Area Ventilation T High      50&deg;F                              47&deg;F Nan-Regen Hx Outlet          N/A                                135&deg;F Temp Hi SLC Initiation                N/A                                N/A RWCU Outside PumplHx          120&deg;F                              115&deg;F Rms RWCU Differential How        30 minutes                        28.5 minutes High Time Delay Group 3 Isolation Valves Control Room RTGB Panel H12-P607 Valve Number          Power Supply            Normal Unit 1(Unit 2)          Position Fail Position      Checked
                                                -        -
Valve Number          Power Supply            Normal Unit 1(Unit 2)          Position Fail Position      Checked
[Note 1] G31-F00i      1XC(2XC)IE1(E3)        NO        [Note 2] FAI G3J-F004              1XDB(2XDB)              NO        FAI
[Note 1] G31-F00i      1XC(2XC)IE1(E3)        NO        [Note 2] FAI G3J-F004              1XDB(2XDB)              NO        FAI
[DCI Note 1: SLC Initiation and RWCU Non-Regen Fix Outlet Temperature Hi signals do NOT isolate the RWCU Inlet Inboard Isolation Valve, G31-F001.
[DCI Note 1: SLC Initiation and RWCU Non-Regen Fix Outlet Temperature Hi signals do NOT isolate the RWCU Inlet Inboard Isolation Valve, G31-F001.
Line 2,571: Line 2,526:


Explanation;  The six switches in a row isolate DG2 engine and generator control circuitry from the control room (the fire area) since a fire induced fault in wiring in the fire area may result in loss of the DG. The seventh switch inserts redundant control power fuses to the circuitry that has been isolated in the event a fault has already resulted in blowing the normal fuses. This seventh switch must be turned last with the potentially faulted circuitry already isolated or the alternate fuses may also blow making the DG unavailable. The DG engine lockout is already tripped if the DG had been running since the operator is directed to trip the DG using emergency stop.
Explanation;  The six switches in a row isolate DG2 engine and generator control circuitry from the control room (the fire area) since a fire induced fault in wiring in the fire area may result in loss of the DG. The seventh switch inserts redundant control power fuses to the circuitry that has been isolated in the event a fault has already resulted in blowing the normal fuses. This seventh switch must be turned last with the potentially faulted circuitry already isolated or the alternate fuses may also blow making the DG unavailable. The DG engine lockout is already tripped if the DG had been running since the operator is directed to trip the DG using emergency stop.
Of the first six switches, they include; Diesel START/STOP (2 switches) Diesel Governor (2
Of the first six switches, they include; Diesel START/STOP (2 switches) Diesel Governor (2 switches) Generator Voltage Regulation (2 switches)
                                                      -                                    -
switches) Generator Voltage Regulation (2 switches)
                          -
Distractor Analysis; Choice A;    Plausible because the six switches are placed in local first and the output breaker does have redundant control power fuses.
Distractor Analysis; Choice A;    Plausible because the six switches are placed in local first and the output breaker does have redundant control power fuses.
Choice B;    Correct Answer, see explanation.
Choice B;    Correct Answer, see explanation.
Line 2,584: Line 2,536:
Voltage Adjust Switches Two three position (RAISE-NEUT-LOWER) spring return to NEUT switches are provided per engine to permit the adjustment of voltage regulators from the local panel regardless of EDG mode of operation.
Voltage Adjust Switches Two three position (RAISE-NEUT-LOWER) spring return to NEUT switches are provided per engine to permit the adjustment of voltage regulators from the local panel regardless of EDG mode of operation.
The auto adjust switch is normally used.
The auto adjust switch is normally used.
ASSD Koylock Switches Brass handled two-position NORM LOCAL ASSD keylock switches
ASSD Koylock Switches Brass handled two-position NORM LOCAL ASSD keylock switches on the local engine panels permit the operator to transfer control of the engine and generator to the local control panel. ASSD operations are performed when a fire exists in the plant and components required to be operated may be damaged by the fire.
                                                    -
on the local engine panels permit the operator to transfer control of the engine and generator to the local control panel. ASSD operations are performed when a fire exists in the plant and components required to be operated may be damaged by the fire.
These switches isolate control room controls and indications to isolate the EDG control circuitry from potential fire induced faults. There are six ASSD switches (2 for EDG mnlstop controls, 2 for governor controls, and 2 for voltage regulation controls) located on each local EDG panel. When in the IASSD!I mode, operation of the Diesel engine can only be accomplished by the LOCAL EMERGENCY STOP and LOCA1 EMERGENCY START pushbuttons.
These switches isolate control room controls and indications to isolate the EDG control circuitry from potential fire induced faults. There are six ASSD switches (2 for EDG mnlstop controls, 2 for governor controls, and 2 for voltage regulation controls) located on each local EDG panel. When in the IASSD!I mode, operation of the Diesel engine can only be accomplished by the LOCAL EMERGENCY STOP and LOCA1 EMERGENCY START pushbuttons.
In addition to the six ASSD switches, for EDG 2 and 4 only, there is a seventh ASSD switch located above the other six switches. This switch provides an alternate set of control power fuses for EDG control drwthy. This may be necessary since fire induced faults may have blown normal control fuses. When operating the ASSD switches for EDGs 2 or4, The seventh switch must be turned last after the potentially faulted circuitry has been isolated to prevent blowing the alternate fuses, making the EDG unavailable to provide paver to Safe Shutdown loads.
In addition to the six ASSD switches, for EDG 2 and 4 only, there is a seventh ASSD switch located above the other six switches. This switch provides an alternate set of control power fuses for EDG control drwthy. This may be necessary since fire induced faults may have blown normal control fuses. When operating the ASSD switches for EDGs 2 or4, The seventh switch must be turned last after the potentially faulted circuitry has been isolated to prevent blowing the alternate fuses, making the EDG unavailable to provide paver to Safe Shutdown loads.
Line 2,870: Line 2,820:
Attachment 4 of OEOP-01-SBO-01, Plant Monitoring Cog Level: Fundamental Explanation: Attachment 4 of OEOP-01 -SBO-01, Plant Monitoring, has a calculation worksheet for figuring Drywell temperature from RSDP temperature recorder readings.
Attachment 4 of OEOP-01-SBO-01, Plant Monitoring Cog Level: Fundamental Explanation: Attachment 4 of OEOP-01 -SBO-01, Plant Monitoring, has a calculation worksheet for figuring Drywell temperature from RSDP temperature recorder readings.
290 0.141
290 0.141
                    *
                             =  40.89 255
                             =  40.89 255
* 0.404 = 103.02 230
* 0.404 = 103.02 230
Line 2,902: Line 2,851:


FIGURE 19-15 Test Return Isolation Valve, E41-FOO8 (E41-FO11) Control Logic RI              fCLOSrSON dl 13W PIISS (Kb; It
FIGURE 19-15 Test Return Isolation Valve, E41-FOO8 (E41-FO11) Control Logic RI              fCLOSrSON dl 13W PIISS (Kb; It
                                                  -,
                                             <20 If          -f    CflSEC\
                                             <20 If          -f    CflSEC\
lRXLflIV 2h0 S7                                    K?
lRXLflIV 2h0 S7                                    K?
Line 2,968: Line 2,916:
This would be either TAF or the level at which downscales are received Choice D:    Plausible because since lowering level will reduce natural circulation and reduce boron mixing.
This would be either TAF or the level at which downscales are received Choice D:    Plausible because since lowering level will reduce natural circulation and reduce boron mixing.
ATWS procedure directs raising level back to the normal band (170-200 inches) once hot shutdown boron weight is injected SRO Basis:      N/A A1WS PROCEDURE BASIS DOCUMENT                                                              OOl-375 Rev. 015 Page 130162 5.4      Step RCIL-2 NOTE I
ATWS procedure directs raising level back to the normal band (170-200 inches) once hot shutdown boron weight is injected SRO Basis:      N/A A1WS PROCEDURE BASIS DOCUMENT                                                              OOl-375 Rev. 015 Page 130162 5.4      Step RCIL-2 NOTE I
                                                                  ,
Requeed immedii1xiftth 1                        recrcuiatin pumps tripped with
Requeed immedii1xiftth 1                        recrcuiatin pumps tripped with
                                           +                  -    powcr eboe 2%.
                                           +                  -    powcr eboe 2%.
IF    reaUo power i ate 2%    R AN1OT be dmiei RPV IeeI is obo.w90 inches.
IF    reaUo power i ate 2%    R AN1OT be dmiei RPV IeeI is obo.w90 inches.
Inj.clon SysThms THIN em,.rnIt awid pr.v.nt injdiun nbz the RPV unless being used *              . Ccndensoteftedvot
Inj.clon SysThms THIN em,.rnIt awid pr.v.nt injdiun nbz the RPV unless being used *              . Ccndensoteftedvot
                                                                .
* HPCI
* HPCI
* CcreSprziy
* CcreSprziy
Line 3,193: Line 3,139:
               - SW TO TBCCWHXS INBD ISOL, SW-V4 closes to a throttled position The Standby CSW pump should start and restore CSW header pressure to normal prior to the SW valves throttling closed. If the standby CSW pump fails to auto start, manually starting the pump will restore CSW header pressure. AOP-19 provides guidance to re-open the SW valves only after header pressure has been restored and the cause of low pressure is known (pump trip).
               - SW TO TBCCWHXS INBD ISOL, SW-V4 closes to a throttled position The Standby CSW pump should start and restore CSW header pressure to normal prior to the SW valves throttling closed. If the standby CSW pump fails to auto start, manually starting the pump will restore CSW header pressure. AOP-19 provides guidance to re-open the SW valves only after header pressure has been restored and the cause of low pressure is known (pump trip).
Distractor Analysis:                                                                                  -
Distractor Analysis:                                                                                  -
Choice A:    Plausible because 30 seconds is when the DG cooling water valves close and a Scram is inserted only after the SW valves have closed to the throttled position AND CSW header pressure cannot be immediately restored above 40 psig under this condition all CSW pumps
Choice A:    Plausible because 30 seconds is when the DG cooling water valves close and a Scram is inserted only after the SW valves have closed to the throttled position AND CSW header pressure cannot be immediately restored above 40 psig under this condition all CSW pumps would be shutdown.
                                                                    -
would be shutdown.
Choice B:    Plausible because 30 seconds is when the DG cooling water valves close and system pressure restored by the STBY pump start is correct.
Choice B:    Plausible because 30 seconds is when the DG cooling water valves close and system pressure restored by the STBY pump start is correct.
Choice C: Plausible because 70 seconds is correct and a Scram is inserted only after the SW valves have closed to the throttled position AND CSW header pressure cannot be immediately restored above 40 psig under this condition all CSW pumps would be shutdown.
Choice C: Plausible because 70 seconds is correct and a Scram is inserted only after the SW valves have closed to the throttled position AND CSW header pressure cannot be immediately restored above 40 psig under this condition all CSW pumps would be shutdown.
                            -
Choice D:    Correct Answer, see explanation SRO Basis: N/A CONVENTIONAL SERVICE WATER SYSTEM                                                  OAOP-19.0 FAILURE                                                          Rev. 26 Page 5 0111 3.0      AUTOMATIC ACTIONS Standby pump setected to the conventional service water header startsat4opsig                                                              U
Choice D:    Correct Answer, see explanation SRO Basis: N/A CONVENTIONAL SERVICE WATER SYSTEM                                                  OAOP-19.0 FAILURE                                                          Rev. 26 Page 5 0111 3.0      AUTOMATIC ACTIONS Standby pump setected to the conventional service water header startsat4opsig                                                              U
: 2.      IF all conventional service water pumps are tripped, THEN:
: 2.      IF all conventional service water pumps are tripped, THEN:
* SW-V36 (SW To CW Pumps Inbd Vlv), closes                              U
* SW-V36 (SW To CW Pumps Inbd Vlv), closes                              U
* SW-V37 (SW To CW Pumps Otbd Vlv), closes                              U
* SW-V37 (SW To CW Pumps Otbd Vlv), closes                              U
* CWIPs trip on low bearing lubricating water flow (5 6 gpm,
* CWIPs trip on low bearing lubricating water flow (5 6 gpm, time-delayed 15 minutes), resulting in loss of condenser vacuum                                                              U
                                                                                  -
time-delayed 15 minutes), resulting in loss of condenser vacuum                                                              U
: 3.      IF conventional service water header pressure remains less than 40 psig for 70 seconds, THEN:
: 3.      IF conventional service water header pressure remains less than 40 psig for 70 seconds, THEN:
* SW-V3 (SW To TBCCW HXs Otbd Isol), closes to a throttled position                                                            U
* SW-V3 (SW To TBCCW HXs Otbd Isol), closes to a throttled position                                                            U
Line 3,290: Line 3,231:
700000 Generator Voltage and Electric Grid Disturbances AA2      Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: (CFR: 41.5 and 43.5 I 45.5, 45.7, and 45.8) 03      Generator current outside the capability curve RO/SRO Rating: 3.5/3.6 Tier 1 / Group 1 K/A Match: This meets the K/A because the tests the ability to determine action needed to remain within capability curve.
700000 Generator Voltage and Electric Grid Disturbances AA2      Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: (CFR: 41.5 and 43.5 I 45.5, 45.7, and 45.8) 03      Generator current outside the capability curve RO/SRO Rating: 3.5/3.6 Tier 1 / Group 1 K/A Match: This meets the K/A because the tests the ability to determine action needed to remain within capability curve.
Pedigree:    Last used on 2014 NRC exam Objective:
Pedigree:    Last used on 2014 NRC exam Objective:
CLS-LP-27, Obj. 9 Given the Generator estimated capability curves, hydrogen pressure and either
CLS-LP-27, Obj. 9 Given the Generator estimated capability curves, hydrogen pressure and either MVARS, MW, or power factor, determine the limit for MW and MVARS.
                      -
MVARS, MW, or power factor, determine the limit for MW and MVARS.


==Reference:==
==Reference:==
Line 3,304: Line 3,243:
45 PSIC                                                                                      391 400 200      :
45 PSIC                                                                                      391 400 200      :
* U,                                    -
* U,                                    -
o          200
o          200 I.
                        -                                                                    -
I.
             -400 o          -600 w                      -
             -400 o          -600 w                      -
             -800 200            400              600          800              1000 1000 KILOWATTS
             -800 200            400              600          800              1000 1000 KILOWATTS
Line 3,314: Line 3,251:
A. (1) is NOT (2) one hour B. (I) is NOT (2) two hours C. (I) is (2) one hour D. (1) is (2) two hours Answer: D K/A:
A. (1) is NOT (2) one hour B. (I) is NOT (2) two hours C. (I) is (2) one hour D. (1) is (2) two hours Answer: D K/A:
G2.1.01 Knowledge of conduct of operations requirements. (CFR: 41.10/45.13)
G2.1.01 Knowledge of conduct of operations requirements. (CFR: 41.10/45.13)
RO/SRO Rating: 3.8/4.2 Tier 3 K/A Match: This meets the K/A because it is testing knowledge of the Conduct of Operations Manual Pedigree:  New Objective:  LOl-CLS-LP-201-D, Obj. lj Explain/describe the following lAW AD-OP-ALL-i 000, Conduct of
RO/SRO Rating: 3.8/4.2 Tier 3 K/A Match: This meets the K/A because it is testing knowledge of the Conduct of Operations Manual Pedigree:  New Objective:  LOl-CLS-LP-201-D, Obj. lj Explain/describe the following lAW AD-OP-ALL-i 000, Conduct of Operations, 001-01.01, BNP Conduct of Operations Supplement and OPS-NGGC-1314, Communications: Control Board walkdown and monitoring requirements
                                        -
Operations, 001-01.01, BNP Conduct of Operations Supplement and OPS-NGGC-1314, Communications: Control Board walkdown and monitoring requirements


==Reference:==
==Reference:==
Line 3,537: Line 3,472:


Explanation:  Drywell high range area monitors provide indications of gross fuel failure and are used to determine emergency plan emergency action level associated with abnormal core conditions.
Explanation:  Drywell high range area monitors provide indications of gross fuel failure and are used to determine emergency plan emergency action level associated with abnormal core conditions.
With the function switch in the ALL position the upper (red) scale is used, meter readings are taken from the upper scale between 1 1,000,000 R/h. Current indication of 200 R/h
With the function switch in the ALL position the upper (red) scale is used, meter readings are taken from the upper scale between 1 1,000,000 R/h. Current indication of 200 R/h Distractor Analysis:
                                                    -
Distractor Analysis:
Choice A:  Plausible because this is the reading on the bottom scale.
Choice A:  Plausible because this is the reading on the bottom scale.
Choice B:  Plausible because if function switch is not taken into account the answer could be 20 RIh.
Choice B:  Plausible because if function switch is not taken into account the answer could be 20 RIh.
Line 3,572: Line 3,505:
I[HHUH                            UHH1HHH
I[HHUH                            UHH1HHH
               -J w
               -J w
              >
250                                              t[fi]Itl Ui
250                                              t[fi]Itl Ui
               -J Ui U      200 z
               -J Ui U      200 z
Line 3,594: Line 3,526:
SRO Basis:      N/A GENERAL FIRE PLAN                                                        OPFP-013 Rev. 48 Page 270135 ATTACHMENT 2 Page 1 of 2
SRO Basis:      N/A GENERAL FIRE PLAN                                                        OPFP-013 Rev. 48 Page 270135 ATTACHMENT 2 Page 1 of 2
                     <<(Information Use) Control Room!Operator Fire Actions>>
                     <<(Information Use) Control Room!Operator Fire Actions>>
                                              -
Sound fire alarm, announce location of the fire 3 times, then                El
Sound fire alarm, announce location of the fire 3 times, then                El
* Announce:                                                              El Fire brigade muster at the fire house.
* Announce:                                                              El Fire brigade muster at the fire house.
Line 3,683: Line 3,614:


==REFERENCE:==
==REFERENCE:==
3.3.1.1; TRM Table 3.3.1.f-1.f I TRIP CHANNEL:                      A3-S3A B3-S3B TRIP SYSTEM:                        A3 S3A
3.3.1.1; TRM Table 3.3.1.f-1.f I TRIP CHANNEL:                      A3-S3A B3-S3B TRIP SYSTEM:                        A3 S3A B3-S3B TRIP LOGIC:                        A3 and B3      =  Reactor scram Place channel in TRIPPED condition t)y: Pull fuse
                                          -
B3-S3B TRIP LOGIC:                        A3 and B3      =  Reactor scram Place channel in TRIPPED condition t)y: Pull fuse


I.
I.
Line 3,743: Line 3,672:
Therefore, The Shear Valve in the Ball valve and shear valve assembly is inoperable. In addition, a novice applicant may assume that with the MODE switch in MANUAL, the ball valve would be inoperable as well. (However, the only position of the MODE switch that would make the TIP inoperable is OFF. ) With Two PCIVS inoperable, condition B would be entered, SRO Basis:    Facility operating limitations in the IS and their bases. [10 CFR 55.43(b)(2)]. The SRO applicant is required to select the appropriate> 1 hour TS condition based on the status of the TIP system indications.
Therefore, The Shear Valve in the Ball valve and shear valve assembly is inoperable. In addition, a novice applicant may assume that with the MODE switch in MANUAL, the ball valve would be inoperable as well. (However, the only position of the MODE switch that would make the TIP inoperable is OFF. ) With Two PCIVS inoperable, condition B would be entered, SRO Basis:    Facility operating limitations in the IS and their bases. [10 CFR 55.43(b)(2)]. The SRO applicant is required to select the appropriate> 1 hour TS condition based on the status of the TIP system indications.


Table 09.5-2 Valve Control Monitor Indications
Table 09.5-2 Valve Control Monitor Indications idication                                  Comment Squib Monitor Light          ON indicates that the iF Shear Vae squib circuit continuity has been lost Shear Valve Monitor Light    ON indicates thatthe squlb charge in the TP Shear Valve has been detonated.
                                -
idication                                  Comment Squib Monitor Light          ON indicates that the iF Shear Vae squib circuit continuity has been lost Shear Valve Monitor Light    ON indicates thatthe squlb charge in the TP Shear Valve has been detonated.
TIP Ball Valve OPEN Light    ON indicates thatTF Ball Valve is OPEN.
TIP Ball Valve OPEN Light    ON indicates thatTF Ball Valve is OPEN.
TIP Ball Valve CLOSED        ON indicates thatT Bait Valve is CLOSED.
TIP Ball Valve CLOSED        ON indicates thatT Bait Valve is CLOSED.
Line 3,753: Line 3,680:
SD-09.5                                  Rev 7                              Page 20 of 58
SD-09.5                                  Rev 7                              Page 20 of 58


Table 09.5-3 TIP Drive Control Unit Indications
Table 09.5-3 TIP Drive Control Unit Indications Indication                                        Comment DETECTOR POSITION                Dynamic digital display of detector position.
                                  -
(illuminated digits)              (0001 reference point about one foot behind the Indexer; 9750 hi Shield position)
Indication                                        Comment DETECTOR POSITION                Dynamic digital display of detector position.
(illuminated digits)              (0001 reference point about one foot behind the
                                              -
Indexer; 9750 hi Shield position)
                                                      -
CORE LIMIT                        Stafic digital display of pre-programmed core top or (illuminated dirits)              bottom limits of selected channel.
CORE LIMIT                        Stafic digital display of pre-programmed core top or (illuminated dirits)              bottom limits of selected channel.
READY Light                      Indicates that Indexer is properly aligned to selected channel.
READY Light                      Indicates that Indexer is properly aligned to selected channel.
Line 3,769: Line 3,691:
flVD (Forward) Light              Detector moving towards top of core.
flVD (Forward) Light              Detector moving towards top of core.
VALVE Light                      ON if TIP Ball Valve is CLOSED.
VALVE Light                      ON if TIP Ball Valve is CLOSED.
Table 09.54 P601 TIP Indications
Table 09.54 P601 TIP Indications Indication                                        Comment TIP Valve Status Green Light
                                            -
Indication                                        Comment TIP Valve Status Green Light
                     -                              Green Light ON indicates that each TIP Ball Valve is FUlL CLOSED.
                     -                              Green Light ON indicates that each TIP Ball Valve is FUlL CLOSED.
TIP Valve Status Red Light
TIP Valve Status Red Light
Line 3,812: Line 3,732:
CONDITiON                        REQUIRED ACTION                COMPLETION TIME A. NOTE-                  A.1      Isolate the affected          8 hours Only applicable to                        penetration flow path by penetration flow paths with              use of at least one closed two PC IVs.                              and de-activated automatic valve, closed manual valve, blind flange, or check valve One or more penetration                  with flow through the valve flow paths with one PCIV                  secured.
CONDITiON                        REQUIRED ACTION                COMPLETION TIME A. NOTE-                  A.1      Isolate the affected          8 hours Only applicable to                        penetration flow path by penetration flow paths with              use of at least one closed two PC IVs.                              and de-activated automatic valve, closed manual valve, blind flange, or check valve One or more penetration                  with flow through the valve flow paths with one PCIV                  secured.
inoperable except for MSIV leakage not within limit        AND PCIVs 3.6.1.3 ACTIONS (continued)
inoperable except for MSIV leakage not within limit        AND PCIVs 3.6.1.3 ACTIONS (continued)
CONDITION                        REQUIRED ACTION                COMPLETION TIME B.              NOTE              B.1      Isolate the affected          2 hours Only applicable to                        penetration flow path by penetration flow paths with              use of at least one closed two PCIVs.                                and de-activated automatic
CONDITION                        REQUIRED ACTION                COMPLETION TIME B.              NOTE              B.1      Isolate the affected          2 hours Only applicable to                        penetration flow path by penetration flow paths with              use of at least one closed two PCIVs.                                and de-activated automatic valve, closed manual valve, or blind flange.
 
valve, closed manual valve, or blind flange.
One or more penetration flow paths with two PCWs inoperable except for MSIV leakage not within limit.
One or more penetration flow paths with two PCWs inoperable except for MSIV leakage not within limit.
fl I      I,.,..J.,4. *1,..
fl I      I,.,..J.,4. *1,..
Line 4,044: Line 3,962:
: 1.      AOG SYSTEM BYPASS VALVE, AOG-HCV-f 02, opens CAUSES
: 1.      AOG SYSTEM BYPASS VALVE, AOG-HCV-f 02, opens CAUSES
: 1. High hydrogen Train A
: 1. High hydrogen Train A
                                -
: 2. High hydrogen Train B
: 2. High hydrogen Train B
                                -
: 3. High-high cooler condenser condensate level
: 3. High-high cooler condenser condensate level
: 4. High-high off-gas flow
: 4. High-high off-gas flow
Line 4,169: Line 4,085:
A1WS PROCEDURE BACtS DOCUMENT                                                      001-37.5 Re.. 14 Page 60 &65 5.33    Step RCIQ-6 through RCIQ-6 Ln7[J                      [LJ.                        3 If reactor power is below 2%, the operator is directed to inject boron before torus water temperature reaches I lOP. This allows sufftciert time for Hot Shutdown Boron Weight (HSBW) of boron to be injected As long as the core remans submerged (the preferred method of core cooling), fuel integrity and RPV integrttyae not directly challenged even wider failure-to-scram conditions. A scram failureccupled with an M3PJ isolation however, result in rapid heatip of the torus due to the steam dischsged from the RPV era SRVs. The challenae to containment thus becomes the tinwing factor which defines the reqi irernwit for boron injection.
A1WS PROCEDURE BACtS DOCUMENT                                                      001-37.5 Re.. 14 Page 60 &65 5.33    Step RCIQ-6 through RCIQ-6 Ln7[J                      [LJ.                        3 If reactor power is below 2%, the operator is directed to inject boron before torus water temperature reaches I lOP. This allows sufftciert time for Hot Shutdown Boron Weight (HSBW) of boron to be injected As long as the core remans submerged (the preferred method of core cooling), fuel integrity and RPV integrttyae not directly challenged even wider failure-to-scram conditions. A scram failureccupled with an M3PJ isolation however, result in rapid heatip of the torus due to the steam dischsged from the RPV era SRVs. The challenae to containment thus becomes the tinwing factor which defines the reqi irernwit for boron injection.
if tows temperatire and RPJ pressure cannot be maintained betters the Heat Capacity Temperature Limit 4HCU), rapid depressurizaton of the RPV will be required. To avoid depressudring the RPV with the reactcr at power, it is desirable to shut down the reactor prior to reaching HCTL, thL,s minirrizing the quantity of heat rejected to the tows. The Boron Injection Initiation Temperature (BItT) is defined so as to achieve ths when practicable.
if tows temperatire and RPJ pressure cannot be maintained betters the Heat Capacity Temperature Limit 4HCU), rapid depressurizaton of the RPV will be required. To avoid depressudring the RPV with the reactcr at power, it is desirable to shut down the reactor prior to reaching HCTL, thL,s minirrizing the quantity of heat rejected to the tows. The Boron Injection Initiation Temperature (BItT) is defined so as to achieve ths when practicable.
ATWO PROCEDURE BASIS DOCUMENT                                                      00l-3T.5 Rev. 14 Page 22 of 65 5.10    Step RCL4 through RCL-l0
ATWO PROCEDURE BASIS DOCUMENT                                                      00l-3T.5 Rev. 14 Page 22 of 65 5.10    Step RCL4 through RCL-l0 1
 
Ut 1  1. aeu.                      WheN Tsl4 Q.5 1a,,,i,iIt eoa P,cn,I D*Iw*lIflSt)fl I  Yn baML,U,* :-ij,..n-.,                      I Mrc(..,.,,                            1 1
1 Ut 1  1. aeu.                      WheN
                                                        -
Tsl4 Q.5 1a,,,i,iIt eoa P,cn,I D*Iw*lIflSt)fl I  Yn baML,U,* :-ij,..n-.,                      I Mrc(..,.,,                            1 1
ml Based on reactor power being above 2%. Step RCt-2 initially lowered RPV level, to the feedwater sparger. by terminating and preventing injection from identified systems if all of the condibons in Table 0-2 are met, Step RCIL-9 will lower RPV level further to suppress reactor power-When any concttion in Table 0-2 is no longer met the operator is directed to continue to subsequent steps which will establish a new RPV level band.
ml Based on reactor power being above 2%. Step RCt-2 initially lowered RPV level, to the feedwater sparger. by terminating and preventing injection from identified systems if all of the condibons in Table 0-2 are met, Step RCIL-9 will lower RPV level further to suppress reactor power-When any concttion in Table 0-2 is no longer met the operator is directed to continue to subsequent steps which will establish a new RPV level band.


Line 4,278: Line 4,191:
170                                          E                                        :r E -0.25 FT
170                                          E                                        :r E -0.25 FT
                                                                                 -1.2S FT 2,30 FT
                                                                                 -1.2S FT 2,30 FT
                                                                                     .25 FT SAFEBELOW                                      - 4.25 FT
                                                                                     .25 FT SAFEBELOW                                      - 4.25 FT U,  iao  -
 
SELECTED LINE 12D-0 1 H.                                                  5.50 FT I- 110 lin Itt-.1150 100      500    700      sco      I 1,100 0        200      400    500    tiOO      1,000 RPV PRESSURE (P510)
U,  iao  -
SELECTED LINE 12D-0 1 H.                                                  5.50 FT I- 110 lin Itt-.1150
                      -
100      500    700      sco      I 1,100 0        200      400    500    tiOO      1,000 RPV PRESSURE (P510)
(START)
(START)
AThS-?
AThS-?
Line 4,289: Line 4,198:
IF                                                            ThEN RPV Iee HfI be deem1ned                                      Exit RCI RCdP flcad 3rd          go to ECP-O1-XFP Emegeiicy da1oii leQ has De1 tequred                            Proceed to
IF                                                            ThEN RPV Iee HfI be deem1ned                                      Exit RCI RCdP flcad 3rd          go to ECP-O1-XFP Emegeiicy da1oii leQ has De1 tequred                            Proceed to
* Exit RCJP nTMPrI  and go to ECP-0i-EC Reztc te DiD WUioUt Bomn Laer    cridfliorA fte            1. TennKiate borm I cliorrijOl required sy oUn EOPs
* Exit RCJP nTMPrI  and go to ECP-0i-EC Reztc te DiD WUioUt Bomn Laer    cridfliorA fte            1. TennKiate borm I cliorrijOl required sy oUn EOPs
: 2. Exit tile flowt ar go to ECcV1RVCP
: 2. Exit tile flowt ar go to ECcV1RVCP A1.2
-
A1.2
                                                     .                a-
                                                     .                a-


                                                ;
C                          1p-
C                          1p-


Line 4,358: Line 4,264:
* EO-C1-SEP-C4 rnecessay Ia dezt LL-2 ci MJLi ayaei
* EO-C1-SEP-C4 rnecessay Ia dezt LL-2 ci MJLi ayaei
* RB vensLatliet ten&#xe7;enaite liz IIQI exceeded 135W (UA-Cd 6-2]
* RB vensLatliet ten&#xe7;enaite liz IIQI exceeded 135W (UA-Cd 6-2]
                                    -                  -
                                                                                                   -                      --a sttA-z
                                                                                                   -                      --a sttA-z
: 91. 5295038 1 A release on Unit Two is occurring with the following plant conditions:
: 91. 5295038 1 A release on Unit Two is occurring with the following plant conditions:
Line 4,379: Line 4,284:
Choice D:    Parti is plausible because it is a common stack for both units. Part 2 is plausible because it is correct, see explanation SRO Basis:    Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J The SRO candidate is required to compare the given radiation release values (including site boundary), and compare those to the EALs for rad effluent. Based on this comparison the candidate must make the correct EAL designation.
Choice D:    Parti is plausible because it is a common stack for both units. Part 2 is plausible because it is correct, see explanation SRO Basis:    Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J The SRO candidate is required to compare the given radiation release values (including site boundary), and compare those to the EALs for rad effluent. Based on this comparison the candidate must make the correct EAL designation.


,.-
While in this procedure:
While in this procedure:
IF                                        THEN Either                                          Perforni per OP-37.3
IF                                        THEN Either                                          Perforni per OP-37.3
Line 4,389: Line 4,293:
Main Steam Line Rad Hi CUA-23, 2-6)
Main Steam Line Rad Hi CUA-23, 2-6)
Process Off-Gas Rad Hi (UA-03, 5-2)
Process Off-Gas Rad Hi (UA-03, 5-2)
* Process OG Vent Pipe Rod Hi (UA-03, 6-4 Reactor building breached                        Request ERO evaluate use of mitigating sprays per EDMG-0D2
* Process OG Vent Pipe Rod Hi (UA-03, 6-4 Reactor building breached                        Request ERO evaluate use of mitigating sprays per EDMG-0D2 RRCP-2 I
--
RRCP-2 I
lF      main stack Hi-Hi isolation Monitor site boundary dose                                actuated, per AD-EP-ALL-0202.                                THEN:
lF      main stack Hi-Hi isolation Monitor site boundary dose                                actuated, per AD-EP-ALL-0202.                                THEN:
RRCP4
RRCP4
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* Process 0tt.Cas Flad Iii IUA.03 6.2)
* Process 0tt.Cas Flad Iii IUA.03 6.2)
* Pwcenn 013 Veri( P FarI Hi (IJA-03, 84)
* Pwcenn 013 Veri( P FarI Hi (IJA-03, 84)
Reactrrbuildng hr-erhd                          Request ERO eyeiu        ini of nidJIing    npmyn pi EDMC3-0U2
Reactrrbuildng hr-erhd                          Request ERO eyeiu        ini of nidJIing    npmyn pi EDMC3-0U2 nRC Step RRCP-2 is a procedure override which appLies the entire time RRCP is being executed Each of the three components specit applicable conditions and direct performance of actions as discussed below.
              -
nRC Step RRCP-2 is a procedure override which appLies the entire time RRCP is being executed Each of the three components specit applicable conditions and direct performance of actions as discussed below.
5.2.1      Step RRCP-2 First Override Continued personnel access to the turbine building may be essential for responding to emergencies or transients which may degrade into emergencies. The turbine l)uilding is not an air tight structure, and radioactivity release inside the turbine building would not only limit personnel access but would eventually lead to an unmonitored ground level release, or release via the turbine building ventilation if operating in the once-through lineup.
5.2.1      Step RRCP-2 First Override Continued personnel access to the turbine building may be essential for responding to emergencies or transients which may degrade into emergencies. The turbine l)uilding is not an air tight structure, and radioactivity release inside the turbine building would not only limit personnel access but would eventually lead to an unmonitored ground level release, or release via the turbine building ventilation if operating in the once-through lineup.
Operation of the turbine building ventilation in the recircutation lineup helps to improve turbine building accessibility. In addion, since both units share a common turbine building airspace, if the building is intact, removing turbine building ventilation from once through lineup will temiinate a large unfiltered volume discharge flow path for a leak on either unit. Due to normal operational requirements when in once through lineup, at least one Air Filter Exhaust Fan and WRGM will be in service providing a monitored and filtered discharge flowpath.
Operation of the turbine building ventilation in the recircutation lineup helps to improve turbine building accessibility. In addion, since both units share a common turbine building airspace, if the building is intact, removing turbine building ventilation from once through lineup will temiinate a large unfiltered volume discharge flow path for a leak on either unit. Due to normal operational requirements when in once through lineup, at least one Air Filter Exhaust Fan and WRGM will be in service providing a monitored and filtered discharge flowpath.
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Release Point                Monitor          GE                SAE              Alert                  UE
Release Point                Monitor          GE                SAE              Alert                  UE
  ,  Main ac                              DI2-RM-2      al3E&#xf7;opCl/ssc    2.13E+t8 pcsee    2.14t7 pcLsec        I.8O+O6 pCLsec Reactor Eicg Wet Noi1e Gas        H-124-3                                                                6.14E+04 cm C,
  ,  Main ac                              DI2-RM-2      al3E&#xf7;opCl/ssc    2.13E+t8 pcsee    2.14t7 pcLsec        I.8O+O6 pCLsec Reactor Eicg Wet Noi1e Gas        H-124-3                                                                6.14E+04 cm C,
TUUiirgVeritRad                      D12RM-23    1opCIisec        t.7E+G7 see      1a7E.iOepcUsec      1.13E+G3 Ct%ee 5er,lce V E1fltrt Rat                Dl2-RM-KC5                                                            2 X r aarm
TUUiirgVeritRad                      D12RM-23    1opCIisec        t.7E+G7 see      1a7E.iOepcUsec      1.13E+G3 Ct%ee 5er,lce V E1fltrt Rat                Dl2-RM-KC5                                                            2 X r aarm ftadwasIe EtSuent Rat                012RM-K503                                                            2 X hi-N alaIm n_
  .
ftadwasIe EtSuent Rat                012RM-K503                                                            2 X hi-N alaIm n_


GENERAL EMERGENCY SITE AREA EMERGENCY RGI Release otgasec4is radIoactl          It olTiItr dose gteater thai 1.10 mrernrEDE cr5000 ntrern 01)T0Id CDE RSI Release oi gaseous isthoacOty restttmg In 0115115 oceegrealat 100 mt5m TIDE cr500 rtrem 11frIl CDI I
GENERAL EMERGENCY SITE AREA EMERGENCY RGI Release otgasec4is radIoactl          It olTiItr dose gteater thai 1.10 mrernrEDE cr5000 ntrern 01)T0Id CDE RSI Release oi gaseous isthoacOty restttmg In 0115115 oceegrealat 100 mt5m TIDE cr500 rtrem 11frIl CDI I
Line 4,513: Line 4,411:
Nate 1:  The SEC should declare the event promptly upon determining that time limithas been exceeded, or will likely be exceeded.
Nate 1:  The SEC should declare the event promptly upon determining that time limithas been exceeded, or will likely be exceeded.
Mode Applicability:
Mode Applicability:
4 Cold Shutdown, 5 Refueling
4 Cold Shutdown, 5 Refueling Definition(s):
  -                      -
Definition(s):
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
Basis:

Latest revision as of 17:14, 24 February 2020

Final SRO Written Exam - Delay Release 2 Years
ML17017A385
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 01/27/2017
From:
NRC/RGN-II/DRS/OLB
To:
References
Download: ML17017A385 (450)
v • d • e


Text

ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U. S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:

Date: 12/13/2016 Facility! Unit: Brunswick Unit 1/2 Region: I El II El Ill El IV El Reactor Type: W ECE El SW El GEE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80 percent overall, with 70 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor receive d aid.

Applicants Signature Results RO!SRO-OnIy/Total Examination Values I I Points

/ /

Applic ants Score Points Applic ants Grade I I Percent

1. Unit Two is operating at rated power when a control rod begins to drift out from position 24.

Which one of the following identifies the first action to be taken by the operator at the controls (OATC)?

A. Initiate a single rod scram.

B. Initiate a manual reactor scram.

C. Select and attempt to arrest the control rod.

D. Select and fully insert the control rod to position 00.

Page: 1

2. Unit One is in an outage with the condensate system under clearance.

An earthquake results in damage to the CST causing level to slowly lower.

Which one of the following completes the statement below with regards to the effect on the CRD system?

The CRD system will (1) when the CST level reaches approximately (2)

A. (1) trip (2) 3 feet B. (1) trip (2) 11 feet C. (1) transfer to the backup supply (2) 3 feet D. (1) transfer to the backup supply (2) 11 feet Page: 2

3. Unit One is at rated power.

Which one of the following identifies the impact of inadvertently closing the IA Reactor Recirculation Pump I-B32-FO31A, Pump A Disch Vlv?

The 1A Reactor Recirculation pump speed will lower to approximately:

A. 20%

B. 34%

C. 45.4%

D. 48%

Page: 3

4. A line break has occurred in the Unit Two drywell with the following sequence of events:

1155 Drywell pressure rises above 1.7 psig 1202 RPV pressure drops below 410 psig 1203 RPV level drops to LL3 Which one of the following completes the statement below?

The earliest time that the operator can throttle the 2-El l-F048A, Loop 2A RHR Heat Exchanger Bypass Valve is at:

A. 1205.

B. 1206.

C. 1207.

D. 1208.

Page: 4

5. RHR Loop 2A is operating in the Shutdown Cooling mode of operation with the following parameters:

RHRSW Pump 2A Operating RHRSW Flow 4000 gpm RHR Pump 2A Operating RHR Loop A Flow 6000 gpm Which one of the following completes the statement below?

The required operator action to lower the cooldown rate lAW 20P-l 7, Residual Heat Removal System Operating Procedure, is to throttle closed:

A. 2-El l-FOO3A, HX 2A Outlet Vlv.

B. 2-Ell-FO17A, Outboard Injection Vlv.

C. 2-El l-F048A, HX 2A Bypass Vlv.

D. 2-El l-PDV-F068A, HX 2A SW Disch Vlv.

Page: 5

6. A Group 1 isolation has occurred on Unit One.

HPCI has been placed in the pressure control mode of operation lAW lOP-I 9, High Pressure Coolant Injection System Operating Procedure.

HPCI flow controller, E41-FIC-R600, is in manual with the output at midscale.

Which one of the following completes the statement below?

If the 1 -E41 -F008, Bypass To CST Valve, is throttled (1) too far, this may result in HPCI (2)

A. (1) open (2) tripping on overspeed B. (1) open (2) operation below 2100 rpm C. (1) closed (2) tripping on overspeed D. (1) closed (2) operation below 2100 rpm Page: 6

7. Unit Two is operating at rated power.

Due to a circuit malfunction an inadvertent LOCA initiation occurs in the Div II Core Spray logic causing A-03 (2-6), CORE SPRAY SYSTEM/I ACTUATED, to alarm.

Which one of the following completes both statements below?

Core Spray Pump(s) (1) will start.

(2) will start.

A. (1) 2B ONLY (2) All DGs B. (1) 2B ONLY (2) DG2 and DG4 ONLY C. (1) 2Aand2B (2) All DGs D. (1) 2A and 2B (2) DG2 and DG4 ONLY Page: 7

8. Which one of the following completes the statement below concerning Core Spray Line Break Detection differential pressure instrument?

The (1) leg of this DP instrument senses (2) core plate pressure via the SLC/Core Differential Pressure penetration.

A. (1)variable (2) below B. (1)variable (2) above C. (1) reference (2) below D. (1) reference (2) above Page: 8

9. Which one of the following completes both statements below?

The normal power supply to RPS MG Set 2B is from 480V MCC (1)

The normal alternate power supply to RPS B is from 480V Bus (2)

A. (1) 2CA (2) E7 B. (1) 2CA (2) E8 C. (1) 2CB (2) E7 D. (1) 2CB (2) E8 Page: 9

10. Which one of the following identifies the LPRM detector level that provides input to the Rod Block Monitor system for indication ONLY, and is NOT used for the purpose of generating rod blocks?

A. LevelA B. Level B C. LeveIC D. Level D Page: 10

11. Unit One is performing a startup with the reactor just declared critical.

While ranging IRM G from range 1, the IRM will not change ranges and remains on Range 1.

Which one of the following completes both statements below?

When IRM G indication first exceeds (1) on the 125 scale, annunciator A-05, 2-4, IRM UPSCALE, will alarm.

The action required lAW A-05, 2-4, IRM UPSCALE, is to (2)

A. (1) 70 (2) place the joystick on P603 for the IRM G to Bypass B. (1) 70 (2) withdraw the IRM G detector to maintain reading on scale C. (1) 117 (2) place the joystick on P603 for the IRM G to Bypass D. (1) 117 (2) withdraw the IRM G detector to maintain reading on scale Page: 11

12. Which one of the following identifies the criteria for when SRM detectors can first begin to be withdrawn from the core lAW OGP-02, Approach To Criticality And Pressurization Of The Reactor?

A. When all IRMs are above range 3.

B. When SRM counts reach 2 x io counts.

C. When RTRCT PERMIT light is illuminated.

D. When SRM/IRM overlap has been established.

Page: 12

13. Which one of the following identifies the power supply to the APRM channel NUMACs?

A. AIIAPRM channels receive 120 VAC power from UPS B. All APRM channels receive 120 VAC power from both RPS Bus A and RPS Bus B C. APRM Channels 1 & 3 receive power from ONLYI2O VAC RPS Bus A APRM Channels 2 & 4 receive power from ONLY12O VAC RPS Bus B D. APRM Channels 1 & 3 receive power from Division I 24/48 VDC APRM Channels 2 & 4 receive power from Division II 24/48 VDC Page: 13

14. Which one of the following completes the statement below?

An APRM must have at least (1) of the assigned LPRMs operable with at least (2) LPRM inputs per axial level operable.

A. (1) 18 (2) 2 B. (1) 18 (2) 3 C. (1) 17 (2) 2 D. (1) 17 (2) 3 Page: 14

15. Following a loss of feedwater, RCIC automatically initiated and subsequently tripped on low suction pressure.

Current plant status is:

Reactor water level is 150 inches RCIC flow controller in Manual set at 200 gpm Subsequently, the following actions are taken:

RCIC suction transferred to Torus E51-V8, Turbine Trip and Throttle Valve is closed E51-V8 is re-opened PF push button on the RCIC flow controller is depressed Which one of the following identifies the indicated flow on the RCIC flow controller that would be observed for these conditions?

A. Ogpm B. 200 gpm C. 400 gpm D. 500 gpm Page: 15

16. Which one of the following completes both statements below concerning the Automatic Depressurization System (ADS) reactor water level inputs from the Nuclear Boiler System?

The (1) instruments provide LL3 inputs to ADS initiation logic.

The (2) range instruments provide LL1 inputs to ADS logic.

A. (1) Fuel Zone (2) Narrow B. (1) Fuel Zone (2) Shutdown C. (1) Wide range (2) Narrow D. (1) Wide range (2) Shutdown Page: 16

17. Unit One is operating at power with Core Spray Pump I B under clearance.

A small break LOCA occurs simultaneously with a Loss of Off-site Power to both units.

DGI and DG4 fail to start and tie onto their respective E bus.

The following plant conditions exist on Unit One:

A-03 (5-1) Auto Depress Timers Initiated In alarm A-03 (6-9) Reactor Low Wtr Level Initiation In alarm RPV pressure 600 psig Drywell pressure 13 psig Which one of the following completes both statements below?

ADS (1) auto initiate.

After ADS is initiated (either automatically or manually), RPV water level (2) be restored with BOTH RHR Loops.

A. (1) will (2) will B. (1) will (2) will NOT C. (1) will NOT (2) will D. (1) will NOT (2) will NOT Page: 17

18. Which one of the following completes the statement below concerning the Fuel Zone instruments, N036 and N037, during a loss of drywell cooling?

The reference leg density will (1) causing the indicated level to read (2) than actual level.

A. (1) rise (2) higher B. (1) rise (2) lower C. (1) lower (2) higher D. (1) lower (2) lower Page: 18

19. Unit One is at 75% power.

The 1A RPS MG set trips.

No operator actions have been taken.

Which one of the following identifies the Main Steam Line Isolation Valve (MSIV) logic lamp status on P601 panel?

Inboard MSIV Logic Outboard MSIV Logic A.OC BO c *0 DO Page: 19

20. Which one of the following identifies the effect if both Refuel Bridge hoist grapple hooks are not open five seconds after placing the Engage/Release switch to Release?

A. Fuel Hoist Interlock is generated.

B. Engage amber light extinguishes.

C. Fault lockout is generated.

D. Grapple hooks will reclose.

Page: 20

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22. Which one of the following identifies the criteria for tripping the main turbine lAW the Unit Two Scram Immediate Actions of 0EOP-01-UG, Users Guide?

A. When APRMs indicate downscale trip.

B. When steam flow is less than 3 Mlbs/hr.

C. When reactor water level is 160 inches and rising.

D. When reactor mode switch is placed in SHUTDOWN.

Page: 22

23. Which one of the following completes both statements below concerning the Main Generator Voltage Regulator?

The automatic voltage regulator maintains a constant generator (1) voltage.

While in the automatic voltage regulation mode, the manual voltage regulator setting (2) automatically follow the automatic setpoint.

A. (1) field (2) does B. (1) field (2) does NOT C. (1) terminal (2) does D. (1) terminal (2) does NOT Page: 23

24. Unit One Reactor Feed Pump 1 B is operating in automatic DFCS control at 4500 RPM.

The DFCS control signal to Reactor Feed Pump 1 B woodward governor immediately fails downscale.

Which one of the following completes the statement below?

Reactor Feed Pump 1 B speed wilt:

A. lowertoorpm.

B. lower to 1000 rpm.

C. lower to 2450 rpm.

D. remain at 4500 rpm.

Page: 24

25. Which one of the following completes both statements below concerning the reactor feed pump turbine (RFPT) DFCS controls?

During a RFPT startup, transfer to DFCS control is performed when RFPT speed is approximately (1)

DFCS will automatically control the speed of the RFPT up to (2)

A. (1) 1000rpm (2) 5450 rpm B. (1) 1000 rpm (2) 6150 rpm C. (1) 2550 rpm (2) 5450 rpm D. (1) 2550 rpm (2) 6150 rpm Page: 25

26. Unit One primary containment venting is being performed lAW I OP-JO, Standby Gas Treatment System Operating System, with the following plant status:

1-VA-I F-BFV-RB, SBGT DW Suct Damper Open I-VA-JD-BFV-RB, Reactor Building SBGT Train IA Inlet Valve Closed I -VA-I H-BFV-RB, Reactor Building SBGT Train 1 B Inlet Valve Closed Which one of the following completes both statements below concerning the predicted SBGT response if drywell pressure rises to 1 .9 psig?

I-VA-1F-BFV-RB (1)

Both l-VA-JD-BFV-RB and I-VA-JH-BFV-RB (2)

A. (1) auto closes (2) auto open B. (I) auto closes (2) remain closed C. (I) remains open (2) auto open D. (1) remains open (2) remain closed Page: 26

27. Unit One is operating at rated power.

Unit Two is in MODE 5 performing fuel movements.

Which one of the following completes both statements below lAW Unit One Tech Spec 3.8.1, AC Sources Operating, LCO statement?

The Unit Two SAT (1) required to be OPERABLE.

(2) Diesel Generators are required to be OPERABLE.

A. (1) is (2) Two B. (1) is (2) Four C. (1) is NOT (2) Two D. (1) is NOT (2) Four Page: 27

28. Unit One is operating at rated power.

Subsequently, El breaker AU9, Feed to 480V Substation E5, trips.

Which one of the following completes the statement below?

120V UPS Distribution Panel IA is:

A. de-energized.

B. energized from MCC ICB.

C. energized from the Standby UPS.

D. energized from 250V DC SWBD A.

Page: 28

29. A reactor shutdown is in progress.

All lRMs on range I reading between 15 and 20.

IRM B detector is failing downscale.

Which one of the following completes both statements below?

lAW A-05 (1-4) IRM Downscale, the alarm setpoint is (1) on the 125 scale.

When the IRM downscale alarm is received, a rod block (2) be generated.

A. (1)3 (2) will B. (1) 3 (2) will NOT C. (1) 6.5 (2) will D. (1) 6.5 (2) will NOT Page: 29

30. Unit Two is operating at full power when a loss of DC Distribution Panel 4A occurs.

Which one of the following completes both statements below?

RCIC is (1) for injection from the RTGB.

RCIC (2) isolation logic has lost power.

A. (1) available (2) inboard B. (1) available (2) outboard C. (1) unavailable (2) inboard D. (1) unavailable (2) outboard Page: 30

31. Unit Two has lost off-site power.

DG3 started and tied to its respective E Bus.

Sequence of events:

1200 DG3 ties to E3 1205 DG3 lube oil temperature rises above 190°F 1206 DG3 lube oil pressure drops below 27 psig Which one of the following identifies when DG3 will trip?

A. Immediately at 1205.

B. Immediately at 1206.

C. 45 seconds after 1205.

D. 45 seconds after 1206.

Page: 31

32. A Unit Two plant cooldown is being performed with the following plant conditions:

Reactor water level 175 inches, steady Reactor pressure band 500 700 psig Drywell ref leg temp 175°F (REFERENCE PROVIDED)

Which one of the following completes both statements below?

The lowering of reactor pressure causes the NOO4AIBIC (Narrow Range) reactor water level instruments indicated level error to (1)

The reactor water level that would correspond to Low level 4 (LL4) is (2)

A. (1) increase (2) -60 inches B. (1) increase (2) -65 inches C. (1) decrease (2) -60 inches D. (1) decrease (2) -65 inches Page: 32

33. Unit Two is performing a startup PAW OGP-02, Approach to Criticality and Pressurization of the Reactor.

lAW OGP-02, which one of the following identifies the radiation monitor(s) that will require the alarm setpoints raised when HWC is placed in service?

A. D12-RM-K603A,B,C,D, Main Steam Line Rad Monitors B. ARM Channel 2-9, U-2 Turbine Bldg Breezeway C. D12-RR-4599-1,2,3, Main Stack Rad Monitors D. ARM Channel 2-4, Cond Filter-Demin Aisle Page: 33

34. Which one of the following identifies the power supply to 2D RHR Pump?

A.E1 B. E2 C.E3 D. E4 Page: 34

35. Unit One is operating at 70% power when the OATC observes indications for a failed jet pump. Subsequently, Recirc Pump IA trips.

Which one of the following completes both statements below lAW IAOP-04.0, Low Core Flow?

Performance of the jet pump operability surveillance for (1) Loop Operation is required.

If it is determined that a jet pump has failed, the required action is to (2)

A. (I) Single (2) reduce reactor power below 25% rated thermal power B. (1) Single (2) commence unit shutdown lAW OGP-05, Unit Shutdown C. (I) Two (2) reduce reactor power below 25% rated thermal power D. (1) Two (2) commence unit shutdown lAW OGP-05, Unit Shutdown Page: 35

36. Unit One is operating at rated power.

The load dispatcher reports degraded grid conditions with the following indications:

Generator frequency 59.7 hertz 230 KV Bus IA voltage 205 KV 230KV Bus lB voltage 205KV El voltage 3690 volts E2 voltage 3685 volts Which one of the following completes both statements below?

The (1) may be damaged with continued operation under these conditions.

lAW OAOP-22.0, Grid Instability, the E-Bus master/slave breakers (2) open.

A. (1) main turbine blades (2) will B. (1) main turbine blades (2) will NOT C. (1) emergency bus loads (2) will D. (1) emergency bus loads (2) will NOT Page: 36

37. Which one of the following completes both statements below?

lAW OAOP-39.0, Loss of DC Power, before 125 VDC battery voltage reaches (1) remove loads as directed by the Unit CRS.

lAW 1 EOP-01 -SBO, Station Blackout, if either division battery chargers can NOT be restored within (2) then load strip the affected battery.

A. (1) 105 volts (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. (1) 105 volts (2) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C. (1) 129 volts (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> D. (1) 129 volts (2) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Page: 37

38. Which one of the following identifies the reason an operator is directed to trip the main turbine as an immediate action lAW OAOP-32.O, Plant Shutdown From Outside Control Room?

A. To initiate a scram on TSV/TCV closure.

B. To prevent reverse power starts of the Diesel Generators.

C. The turbine cannot be tripped once the Control Room is evacuated.

D. To bring bypass valves into operation until Remote Shutdown Panel control is established.

Page: 38

39. Unit One has entered RSP with the following conditions:

Six control rods are at position 02, all others are fully inserted B Recirc Pump has tripped Which one of the following completes both statements below?

The control rods will be inserted by (1) lAW OEOP-01-LEP-02, Alternate Control Rod Insertion.

After the control rods are inserted, a CRD flow rate of approximately (2) will be established.

A. (1) placing the individual scram test switches to the Scram position (2) 30 gpm B. (1) placing the individual scram test switches to the Scram position (2) 45 gpm C. (1) driving rods using RMCS (2) 30 gpm D. (1) driving rods using RMCS (2) 45 gpm Page: 39

40. A total loss of Unit One feedwater results in reactor water level lowering to 87 inches.

Drywell pressure is 2.1 psig.

Reactor water level is being restored with RCIC and CRD.

Which one of the following completes both statements below?

RVCP (1) requiredtobeentered.

The expected response of the G31-F001, Inboard RWCU Isolation Valve, and the G31-F004, Outboard RWCU Isolation Valve, is that (2) should be closed.

A. (1) is (2) ONLY the G31-F004 B. (1) is (2) BOTH C. (1) is NOT (2) ONLY the G31-F004 D. (1) is NOT (2) BOTH Page: 40

41.

CAUTION There are seven keock NORMAULOCAL switches located on Diesel Generator 2 control panel. Six of these are located in a row. The seventh switch is located in the row above the six switches.

Which one of the following completes both statements below concerning the caution above from OASSD-02, Control Building?

The six switches in a row must be placed in LOCAL (1) placing the seventh switch in LOCAL.

The purpose of this sequence is to prevent a loss of DG2 due to a loss of the redundant power supply fuses for the (2) circuitry.

A. (1) before (2) output breaker B. (1) before (2) engine run control C. (1) after (2) output breaker D. (1) after (2) engine run control Page: 41

42. During accident conditions, the source term from the Unit One Reactor Building must be estimated. Three RB HVAC supply fans and three RB HVAC exhaust fans are running.

lAW OPEP-03.6.1, Release Estimates Based on Stack/Vent Readings, which one of the following is the calculated release rate?

ATTACHMENT 2 Page 1 of I Source Term Calculation From #1 RX Gas (1-CAC-AQH-1264-3)

METER FLOW1 EFFICIENCY12 RELEASE3 READING (cfm) FACTOR RATE TIME (cpm) (pCilsec)

I minute ago 4.0 E÷3 1275 E-5 I J IT not available use 43,200 cftn per exhaust fan times the number of fans operating.

i2) The efficiency factors can be obtained from OE&RC-2020 (contact E&RC counting room).

Release Rate (cpm) x (cfrn) x (Efficiency Factor)

A. 2.2 E+3 jiCi/sec.

B. 6.6 E+3 pCi/sec.

C. 1.3 E+4 iCi/sec.

D. 6.6 E+4 pCi/sec.

Page: 42

43. Unit Two is operating at -65% power when the following are observed:

100 1

11 p RBCCW RBCCW PIJMP HEAD TANK DISCH HEADER PUMP MOTOR PRESS LOW TEMP HI -8O LEVEL HI/to UA-3 UA-3 A-6 (In Alarm) E-60 (In Alarm) (In Alarm)

Fcc PtItP 28 2X RIXW RIP 1C 2XE E-40

-20 R9CCW O1SCRARGE PRESSURE e C-Fi-51-)

Which one of the following completes both statements below lAW OAOP-16.O, RBCCW System Failure?

A complete loss of RBCCW (1) occurred.

Areactorscram (2) requited.

A. (1) has (2) is B. (1) has (2) is NOT C. (1) has NOT (2) is D. (1) has NOT (2) is NOT Page: 43

44. Unit Two has entered OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures, due to a loss of instrument air pressure with the following annunciator status:

UA-O1 (1-1) RB lnstrAirReceiver2A Press Low Alarm sealed in UA-O1 (1-2) RB InstrAir Receiver 28 Press Low NOT in Alarm UA-O1 (3-2) Air Compr D Trip Alarm sealed in UA-O1 (4-4) Inst Air Press Low Alarm sealed in UA-O1 (5-4) Service Air Press-Low Alarm sealed in Which one of the following completes both statements below?

On a loss of instrument air, the RB HVAC Butterfly Isolation Valves will fail (1) lAW OAOP-20.O, the reactor (2) required to be scrammed.

A. (1) as-is (2) is B. (1) as-is (2) is NOT C. (1) open (2) is D. (1) open (2) is NOT Page: 44

45. l&C Techs inadvertently cause a low level 3 (LL3) signal.

Unit Two plant conditions are:

Reactor pressure 930 psig Drywell pressure 1 .7 psig, steady Drywell temp (average) 140°F, slow rise Drywell leak calculation Normal Which one of the following completes the statement below?

All Drywell Cooler Fans are:

A. tripped, but can be overridden on.

B. tripped, and cannot be overridden on.

C. running, but can be tripped at the RTGB.

D. running, and cannot be tripped at the RTGB.

Page: 45

46. Unit One in MODE 5.

The fuel pool gates are removed.

SDC Loop B is in service.

Fuel pooi cooling assist is in operation.

The RHR Loop B pumps tripped and can NOT be restarted.

Which one of the following completes both statements below?

(consider each statement separately)

Fuel pool cooling assist (I)

Fuel pool cooling assist (2) capable of being aligned to the SDC Loop A lAW I OP-I 7, Residual Heat Removal System Operating Procedure.

A. (I) remains in service (2) is B. (I) remains in service (2) is NOT C. (I) is lost (2) is D. (1) is lost (2) is NOT Page: 46

47. Unit Two is performing refueling operations when the refueling SRO reports that a spent fuel bundle has been dropped.

The following radiation monitoring alarms are received:

UA-03 (3-7) Area Rad Refuel Floor High UA-03 (4-5) Process Rx Bldg Vent Rad Hi Which one of the following identifies the Immediate Action that is required lAW OAOP-05.O, Radioactive Spills, High Radiation, and Airborne Activity?

A. Verify Group 6 isolation.

B. Evacuate all personnel from the refuel floor.

C. Place Control Room Emergency Ventilation System in operation.

D. Isolate Reactor Building Ventilation and place Standby Gas Treatment trains in operation.

Page: 47

48. Unit Two is operating at rated power when high drywell pressure switch C72-PTM-NOO2A-1 fails high resulting in the annunciation of A-05-(5-6) Pri Ctmt Press Hi Trip.

Which one of the following completes the statement below?

RPS high drywell pressure relay C72-K4A will (1)

The RSP (2) be required to be entered.

A. (1) energize (2) will B. (1) energize (2) will NOT C. (1) de-energize (2) will D. (1) de-energize (2) will NOT Page: 48

49. Unit One was operating at power when a turbine trip occurred.

85 control rods fail to insert.

Reactor pressure peaks at 1145 psig.

Which one of the following completes both statements below?

The reactor recirc pumps (1) tripped.

Tripping of the reactor recirc pumps results in a rapid decrease in reactor powerdueto (2)

A. (1) must be manually (2) voiding of the moderator B. (1) must be manually (2) a reduction in reactor water level C. (1) have automatically (2) voiding of the moderator D. (1) have automatically (2) a reduction in reactor water level Page: 49

50. Unit One failed to scram following a loss of off-site power with the following plant conditions:

Reactor Power 5%

RPV Water Level -55 inches (N036)

RPV Pressure 850 psig Which one of the following completes both statements below?

SPIMS DIV This UA-1 2 (5-4) alarm is expected to be received when suppression pooi water Dliii? UTfl TTLAD UVLI\ W JR J [NIT temperature first teaches (1)

SETPOINT TS1 lAW lOP-I 7, Residual Heat Removal System Operating Procedure, the RHR logic requirements to place torus cooling in service under the current plant conditions will requite (2)

A. (1) 95°F (2) placing the CS-S17B Think Switch to Manual first and then bypassing the 2/3rd core height interlock B. (1) 95°F (2) bypassing the 2/3rd core height interlock first and then placing the CS-SI7B Think Switch to Manual C. (1) 105°F (2) placing the CS-SI7B Think Switch to Manual first and then bypassing the 2/3rd core height interlock D. (1) 105°F (2) bypassing the 2/3rd core height interlock first and then placing the CS-SI7B Think Switch to Manual Page: 50

51. Unit Two is in MODE 3 following a Station Blackout.

lAW OEOP-01-SBO-01, Plant Monitoring, the AO has reported the following temperatures from the RSDP temperature recorder 2CAC-TR-778:

Point 1 290°F Point2 118°F Point 3 255°F Point 4 230° F Point 5 191°F Point6 117°F (REFERENCE PROVIDED)

Which one of the following represents the correct calculated Drywell temperature?

A 205°F B 249°F C 258°F D. 267°F Page: 51

52. Unit Two is performing RVCP with HPCI in pressure control.

Subsequently, A-O1 (1-5) Suppression Chamber Level Hi Hi is received.

Which one of the following completes both statements below?

The E41-F004, CST Suction Vlv, will (1)

The E41-F008, Bypass to CST Vlv, will (2)

A. (1) close (2) close B. (1) close (2) remain open C. (1) remain open (2) close D. (1) remain open (2) remain open Page: 52

53.

Unit One is operating at rated power when A-O1 (3-7)

Suppression Chamber Lvi Hi/Lo, is received.

The BOP Operator verifies the alarm using 28 CAC-Ll-4177, Supp Pool Level, indicator on Panel XU-51. (indication provided to the left)

E-33 Which one of the following identifies the action that is required lAW A-O1 (3-7) Suppression Chamber Lvi Hi/Lo?

The water level in the Unit One torus must be:

A. lowered by using Core Spray and routed to Radwaste.

B. lowered using RHR and routed to Radwaste.

C. raised by opening the HPCI suction from the CST.

D. raised by opening the Core Spray suction from the CST.

Page: 53

54. Unit One is executing the ATWS procedure with the following plant conditions:

Reactor power 12%

Reactor pressure 940 psig, controlled by EHC Reactor water level 170 inches, controlled by feedwater Which one of the following identifies the reason the ATWS procedure directs deliberately lowering RPV water level to 90 inches?

A. Reduces reactor power so that it will remain below the APRM downscale setpoint.

B. Provides heating of the feedwater to reduce potential for high core inlet subcooling.

C. Reduces challenges to primary containment if MSIVs close.

D. Promotes more efficient boron mixing in the core region.

Page: 54

55. Which one of the following identifies the reason for performing Emergency Depressurization due to exceeding Maximum Safe Operating Temperatures lAW 001-37.9, Secondary Containment Control Procedure Basis Document?

A. Prevent an unmonitored release.

B. Preserve personnel access into the reactor building.

C. Provide continued operability of equipment required for safe shutdown.

D. Ensure ODCM site boundary dose limits are not exceeded.

Page: 55

56. Which one of the following completes both statements below?

lAW OAOP-5.4, Radiological Releases, RRCP is entered when the Turbine Building Vent Rad Monitor indication exceeds an (1) EAL.

lAW RRCP, before the radioactivity release rate reaches a (2) Emergency EAL, Emergency Depressurization is required.

A. (1) Unusual Event (2) Site Area B. (1) Unusual Event (2) General C. (1) Alert (2) Site Area D. (1) Alert (2) General Page: 56

57. Following an unisolable RWCU line break in the reactor building the following conditions exist:

South Core Spray Room temperature 155°F South RHR Room temperature 300°F UA-12 (2-3) South Core Spray Room Flood Level Hi, in alarm UA-1 2 (2-4) South RHR Room Flood Level Hi, in alarm UA-12 (1-4) South RHR Room Flood Level Hi-Hi, in alarm (REFERENCE PROVIDED)

Which one of the following completes both statements below?

lAW OEOP-01-UG, Users Guide, (1) equipment required for safe shutdown will fail.

lAW SCCP, Emergency Depressurization (1) required.

A. (1) ONLY the South RHR room (2) is B. (1) ONLY the South RHR room (2) is NOT C. (1) the South RHR room AND Core Spray room (2) is D. (1) the South RHR room AND Core Spray room (2) is NOT Page: 57

58. The RO has attempted to manually scram Unit One with the following actions taken:

All rods are noted to be greater than position 02 Reactor mode switch is placed in shutdown ARI was initiated.

Both recirculation pumps were tripped.

Reactor power reported at 12%

SLC is injecting RPV level is 80 inches and stable Rod insertion attempts are unsuccessful Which one of the following completes both statements below?

Reactor power (1) expected to be lowering.

Assuming no rod insertion, SLC injection (2)

A. (1) is (2) can be secured when all APRMs are downscale B. (1) is (2) must be continued until the reactor is shutdown under all conditions C. (1) is NOT (2) can be secured when all APR Ms are downscale D. (1) is NOT (2) must be continued until the reactor is shutdown under all conditions Page: 58

59. A radioactive release has occurred in the Turbine Building.

Which one of the following completes both statements below?

lAW OAOP-05.4, Radiological Releases, the Unit Two turbine building ventilation must be in the (1) operating mode.

This discharge will be monitored by the (2)

A. (1) recirc (2) Main Stack Radiation Monitor B. (1) recirc (2) Wide Range Gaseous Monitor (WRGM)

C. (1) once through (2) Main Stack Radiation Monitor D. (1) once through (2) Wide Range Gaseous Monitor (WRGM)

Page: 59

60. Unit One is operating at rated power when the following alarms are received:

UA-01 (4-4) lnstr Air Press-Low UA-01 (5-1) Air Dryer 1A Trouble The AC reports that the cause of the alarms is due to filter blockage.

Which one of the following completes both statements below?

The Service Air Dryer malfunction will cause SA-PV-5067, Service Air Dryer Bypass Valve, to open when pressure first lowers to (1) lAW OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures, the required action isto (2)

A. (1) 105 psig (2) place the 1 B Service Air Dryer in service B. (1) 105 psig (2) set the service air dryer maximum sweep value to zero C. (1) 98psig (2) place the I B Service Air Dryer in service D. (I) 98psig (2) set the service air dryer maximum sweep value to zero Page: 60

61. Unit One is in MODE 3 following a seismic event and reactor scram with the following plant conditions:

Reactor level 55 inches Reactor pressure 500 psig Drywell pressure 9 psig Division I PNS header pressure 93 psig Division II PNS header pressure 98 psig Which one of the following completes both statements below?

Div I Backup N2 Rack Isol Vlv, RNA-SV-5482 is (1)

Div II Backup N2 Rack Isol Vlv, RNA-SV-5481 is (2)

A. (1) open (2) open B. (1) open (2) closed C. (1) closed (2) open D. (1) closed (2) closed Page: 61

62. Unit One is operating at rated power with the following conditions:

CSW Pump JA trips Conventional header pressure lowers to 35 psig Which one of the following completes both statements below?

If CSW header pressure remains at this pressure for (1) seconds, the SW-V3, SW To TBCCW HXs Otbd Isol Vlv, and SW-V4, SW To TBCCW HXs lnbd Isol Vlv, will close to a throttled position.

lAW OAOP-1 9, Conventional Service Water System Failure, the SW-V3 and SW-V4 are reopened (2)

A. (1) 30 (2) ONLY after a reactor Scram is inserted B. (1) 30 (2) if system pressure is restored by starting the standby CSW pump C. (1) 70 (2) ONLY after a reactor Scram is inserted D. (1) 70 (2) if system pressure is restored by starting the standby CSW pump Page: 62

63. Unit Two Nuclear Service Water (NSW) pumps are aligned as follows in preparation for equipment realignment:

DISCHARGE NUCLEAR SERVICE DISCHARGE NUCLEAR SERVICE VLV SWV20 VLV SWV 19 YATER PUMP 2A \NATER PUMP 2B 2PA 2P B E4 1

I

© Subsequently, Off-site power is lost.

Which one of the following completes the statement below?

(1) NSW pump(s) will auto start (2) associated E Bus is re-energized.

A. (1) 2A and 2B (2) immediately when their B. (1) 2A and 2B (2) five seconds after their C. (1) 2B ONLY (2) immediately when its D. (1) 2B ONLY (2) five seconds after its Page: 63

64. Which one of the following identifies the potential consequence of failing to place backup nitrogen in service by placing RNA keylock switches in LOCAL lAW OASSD-02, Control Building?

RNA keylock switch noun names:

2-RNA-CS-OO1, Override Switch For Valve RNA-SV-5482 2-RNA-CS-002, Override Switch For Valve RNA-SV-5253 A. Misoperation of RCIC.

B. Loss of drywell cooling.

C. Inability to operate SRVs.

D. Spurious operation of MSIVs.

Page: 64

65. A grid disturbance occurs with the following Unit One plant parameters:

Generator Load 980 MWe Generator Reactive Load 160 MVARs, out Generator Gas Pressure 50 psig (REFERENCE PROVIDED)

Which one of the following identifies both available options that will place the Unit within the Estimated Capability Curve?

A. Raise gas pressure to 58 psig or lower power to 940 MWe.

B. Raise gas pressure to 58 psig or raise reactive load to 240 MVARs.

C. Raise gas pressure to 58 psig or lower reactive load to 70 MVARs.

D. Lower power to 940 MWe or raise reactive load to 240 MVARs.

Page: 65

66. Which one of the following completes both statements below lAW AD-OP-ALL-WOO, Conduct of Operations?

With the Unit operating at rated, steady state power, steam flow/feed flow (1) a key parameter that the OATC must monitor to assure a constant awareness of its value and trend.

An end to end control panel walk down shall be performed every (2) and documented in the Narrative Logbook.

A. (1) is NOT (2) one hour B. (1) is NOT (2) two hours C. (1) is (2) one hour D. (1) is (2) two hours Page: 66

67. Which one of the following completes the statement below?

1 OP-I 0, Standby Gas Treatment System Operating Procedure, prohibits venting the drywell and the suppression pool chamber simultaneously with the reactor at power because this would cause the:

A. unnecessary cycling of reactor building to torus vacuum breakers.

B. unnecessary cycling of torus to drywell vacuum breaker.

C. SBGT Train water seal to blow out of the trough.

D. pressure suppression function to be bypassed.

Page: 67

68. A core reload is in progress during a refueling outage. The initial loading of fuel bundles around each SRM centered 4-bundle cell was completed with all four SRMs fully inserted and reading 50 cps.

It is now approximately half way through the core loading sequence and SRMs read 80 cps.

Which one of the following completes the statement below lAW OFH-1 1, Refueling?

Fuel movement must be suspended when any SRM reading first rises to upon insertion of the next fuel bundle.

A. lOOcps B. 160 cps C. 250 cps D. 400 cps Page: 68

69. Unit Two is conducting a routine power reduction for rod pattern improvement.

The Reactivity Management Plan contains actions for the RO to insert a group of four rods from position 24 to position 12.

Which one of the following completes the statement below lAW AD-OP-ALL-0203, Reactivity Management?

The movement of these rods should be:

A. single notched for the entire movement.

B. continuously inserted to the final intended position.

C. continuously inserted to settle four notches prior to reaching the intended position and then single notched into the final intended position.

D. continuously inserted to settle one notch prior to reaching the intended position and then single notched into the final intended position.

Page: 69

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ouiqin U!BW OLfiJO 6U1dd1J+/- V touO 1!Ufl uo ueq oouEwioped J04 E!JOpJD UeJo44ip E SeZ!I!lfl TBL.II EIUOI1DV J0TJOdQ OWIpOWWI W8J391 OM+/- !Ufl eqT SO!4!1UOP! 6U!M01104 OIfl 40 OUO 1pIq 0L

71. The OATC obser thei liowing indications after initiating SLC during an ATWS.

76 SLC A/B z15 SOUB VALVE ONflNUItY 15 P

PE S

72-SE SC PUMP ZA pup 28 C41-cCOlA I C4CCO18 2X [ 2XJ GE x

Xz E-6 lz 0

O3 0 SLC PUMP A&B ct1s z-n PREURE £%SCHARCE PRESSURE Which one of the following completes both statements below?

Squib valve (1) has failed to fire.

lAW 20P-05, Standby Liquid System Operating Procedure, the OATC is required to (2)

A. (1)A (2) place the CS-SI, SLC Pump A & B, in the PUMP B RUN position B. (1) A (2) leave the CS-Si, SLC Pump A & B, in the PUMP NB RUN position C. (I) B (2) place the CS-SI, SLC Pump A & B, in the PUMP A RUN position D. (I) B (2) leave the CS-SI, SLC Pump A & B, in the PUMP NB RUN position Page: 71

72. Two operators are required to enter a room that is posted as a Locked High Radiation Area (LHRA) to hang a clearance for scheduled work.

Which one of the following completes both statements below?

The radiation level at which a LHRA posting is required is (1) in one hour at 30 centimeters from the radiation source.

The LHRA key is obtained from (2 A. (1) >l00mrem (2) the Shift Manager B. (1) >J00mrem (2) a RP Technician C. (1) >l000mrem (2) the Shift Manager D. (1) >l000mrem (2) a RP Technician Page: 72

73.

Which one of the following identifies the DW radiation value indicated above?

A. 1OR/hr B. 20 R/hr C. 100 R/hr D. 200 R/hr Page: 73

74. A transient has occurred on Unit Two with the following plant conditions:

RPV pressure 1000 psig Drywell ref leg area temp 197°F Rx Bldg 50 temp 135°F Wide Range Level 170 inches (NO26AIB)

Shutdown Range Level 160 inches (N027NB)

(REFERENCE PROVIDED)

Which one of the following completes both statements below concerning the level instruments that can be used to determine reactor water level lAW EOP Caution 1?

Wide Range Level instruments NO26NB (1) be used.

Shutdown Range Level instruments N027A/B (2) be used.

A. (1) can (2) can B. (1) can (2) can NOT C. (1) can NOT (2) can D. (1) can NOT (2) can NOT Page: 74

75. A fire has been reported and confirmed in the turbine building breezeway.

A fire hose is being used to control/suppress the fire.

Which one of the following completes both statements below lAW OPFP-O1 3, General Fire Plan?

The RD is required to sound the fire alarm and announce the location of the fire (1)

A call for offsite assistance to the Brunswick County 91 1 Center (2) required.

A. (1) ONLY once (2) is B. (1) ONLY once (2) is NOT C. (1) three times (2) is D. (1) three times (2) is NOT Page: 75

76. During a LOCA and LOOP on Unit One, the following plant conditions exist:

An Emergency Depressurization has been performed due to RPV water level The Reactor Building -17 foot and 20 foot elevations are NOT accessible due to radiation levels.

ALL ECCS pumps are unavailable.

Which one of the following completes the statement below?

The CRS will direct demin water injection to the RPV, lAW OEOP-01-LEP-01, Alternate Coolant Injection, Section:

A. 2.4.3.3a, Demineralized Water Actions, Inject demineralized water through Core Spray Loop A B. 2.4.3.3c, Demineralized Water Actions, Inject demineralized water through RHR Loop A C. 2.4.3.3d, Demineralized Water Actions, Inject demineralized water through HPCI D. 2.4.3.3e, Demineralized Water Actions, Inject demineralized water through RCIC Page: 76

77. Unit One is at rated power performing OPT-O1.1.6, Reactor Protection System Manual Scram Test.

The Reactor Scram System A pushbutton has been depressed.

RPS Trip System A Scram Groups light for groups one, two, three, and four are illuminated (REFERENCE PROVIDED)

Which one of the following completes both statements below?

The scram pilot valve solenoids associated with these lights are (1)

Tech Spec 3.3.1 .1, Reactor Protection System Instrumentation, Condition B (2) requited to be entered.

A. (1) energized (2) is B. (1) energized (2) is NOT C. (1) de-energized (2) is D. (1) de-energized (2) is NOT Page: 77

78. Unit Two is at rated power. A TIP trace is in progress.

TIP D Valve Control Unit and Monitor indications are as follows:

(REFERENCE PROVIDED)

Which one of the following completes both statements below?

Tip Valves are Group (1) PCIVs.

Tech Spec 3.6.1 .3, Primary Containment Isolation Valves (PC/Vs), Condition(s) (2) is/are required to be entered.

A. (1)2 (2) A ONLY B. (1) 2 (2) A and B C.(1)6 (2) A ONLY D. (1) 6 (2) A and B Page: 78

79. Unit One is operating at rated power.

Unit Two is in MODE 5 with UAT backfeed established.

A main generator backup lockout occurs on Unit One.

(REFERENCE PROVIDED)

Which one of the following completes both statements below?

All four diesels (1) automatically start.

lAW Unit One Tech Spec 3.8.1, AC sources Operating, Condition E (2) required to be entered.

A. (1) will (2) is B. (1) will (2) is NOT C. (1) will NOT (2) is D. (1) will NOT (2) is NOT Page: 79

80. Unit One was operating at power when a Group 1 isolation and reactor scram occurred.

Reactor pressure is 950 psig and being manually controlled by SRVs.

An SRV is stuck open with a stuck open SRV tailpipe vacuum breaker.

Torus and Drywell sprays have been initiated lAW PCCP (REFERENCE PROVIDED)

Which one of the following completes both statements below?

The SRV is discharging through the open vacuum breaker directly into the (1)

The highest EAL classification for this event is a(n) (2)

A. (1) drywell (2) Alert B. (1) drywell (2) Site Area Emergency C. (1) suppression chamber air space (2) Alert D. (1) suppression chamber air space (2) Site Area Emergency Page: 80

81. Unit Two is operating at rated power with RHR Loop A operating in suppression pool cooling mode.

A-O1 (2-8) RHR Relay Logic Pwr Failure, is in alarm due to a blown fuse affecting RHR Logic A ONLY.

(REFERENCE PROVIDED)

Which one of the following completes both statements below?

CwW lAW Tech Spec 3.3.5.1, ECCS Instrumentation, required channels (1) required to be placed in trip within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If a LOCA signal were to occur, 2-E11-FO15A, Inboard Injection Vlv, (2) open automatically on low reactor pressure.

A. (1) are (2) will B. (1) are NOT (2) will C. (1) are (2) will NOT D. (1) are NOT (2) will NOT Page: 81

82. Unit Two is operating at rated power.

Subsequently, a Div I pneumatic leak occurs causing drywell pressure to rise to 1 .9 psig.

Which one of the following completes both statements below?

The SBGT trains (1) running.

The Div I pneumatics are required be isolated lAW (2)

A. (1) are NOT (2) OEO P-O 1-SEP-I 6, Drywell Systems Isolation B. (1) are NOT (2) OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures C. (1) are (2) OEOP-O1 -SEP-I 6, Drywell Systems Isolation D. (1) are (2) OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures Page: 82

83. Unit Two is operating at rated power.

UA-48 (5-4) AOG System Bypass, has been alarming for 1 minute due to High-High off gas flow (REFERENCE PROVIDED)

Which one of the following completes both statements below?

AOG-XCV-142, Guard Bed Isolation Valve, (1) automatically close.

ODCM 7.3.10, Gaseous Radwaste Treatment System, Condition A entry (2) required.

A. (1) will (2) is B. (1) will (2) is NOT C. (1) will NOT (2) is D. (1) will NOT (2) is NOT Page: 83

84. Unit One is operating at 72% power with the following conditions:

Jet Pump Flow Loop A (B21-R61 IA) 25 Mlbs/hr Jet Pump Flow Loop B (B21-R611B) 29 Mlbs/hr Total Core Flow (U1CPWTCF) 54 Mlbs/hr Which one of the following completes both statements below lAW Tech Spec 3.4.1, Recirculation Loops Operating, and Bases? (consider each statement separately)

The current Jet Pump Flow mismatch (I)

If Jet Pump Flows are not matched within limits, then the loop with the (2) must be considered not in operation.

A. (1) is within limits (2) lower flow B. (1) is within limits (2) higher flow C. (1) is not within limits (2) lower flow D. (1) is not within limits (2) higher flow Page: 84

85. Unit One was at full power when all offsite power was lost.

The following is the Emergency Diesel Generator status:

DGI Locked out on fault DG2 Running and loaded DG3 Running and loaded DG4 Locked out on fault Which one of the following completes the statements below?

The (I) CRD pump must be started to re-establish the CRD system.

OAOP-36.1 Loss Of Any 4760 V Buses or 480V E-Buses, (2) contain the step for placing the CRD Flow Control, C11-FC-R600, in manual with manual potentiometer at minimum setting following the loss of the CRD pump?

A. (1) IA (2) does B. (1) IA (2) does NOT C. (1) lB (2) does D. (1) lB (2) does NOT Page: 85

86. Unit One is performing the ATWS Procedure with the following conditions:

A-05 (2-6) Reactor Vess Lo Level Trip, is illuminated A-06 (1-6) Reactor Vess Lo Lo Water Level Sys A, is NOT illuminated A-06 (2-6) Reactor Vess Lo Lo Water Level Sys B, is NOT illuminated MSIVs are closed Reactor pressure peaked at 1141 psig and is now being controlled 800-1000 psig.

Torus water temperature is 105°F and rising Reactor power is 25%

lAW 001-37.5, ATWS Procedure Basis Document, which one of the following identifies the action that will have the highest priority?

A. SLC initiation.

B. Inhibiting ADS.

C. Trip both Reactor Recirc Pumps.

D. Termination and prevention of RPV injection.

Page: 86

87. Which one of the following completes both statements below?

lAW Tech Spec 3.9.6, Reactor Pressure Vessel(RPV) Water Level, the minimum water level over the top of irradiated fuel assemblies seated within the RPV during movement of irradiated fuel assemblies in the RPV is (1)

The Tech Spec bases for the minimum water level is to provide for (2) during a fuel handling accident.

A. (1) l9feetll inches (2) iodine retention B. (1) 19 feet 11 inches (2) shielding of radioactive decay particles C. (1) 23 feet (2) iodine retention D. (1) 23 feet (2) shielding of radioactive decay particles Page: 87

88. Unit Two is in an ATWS executing RXFP, with the following plant conditions:

Injection to the RPV has been terminated and prevented The Minimum Number of SRVs Required for Emergency Depressurization are open.

Table P4 Minimum Steam Cooling Pressure Open SRVs Pressure (psig) 7ormore 120 6 145 5 175 4 220 3 300 2 455 1 915 lAW RXFP, which one of the following completes the statement below?

The CRS should direct injection to the RPV when EITHER:

(1) SRV remains open OR when reactor pressure lowers below the Minimum Steam Cooling Pressure of (2)

A. (1) NO (2) 175 psig B. (1) NO (2) 455 psig C. (1) ONLY one (2) 175 psig D. (1) ONLY one (2) 455 psig Page: 88

89. An event on Unit One has resulted in the following plant conditions:

Reactor pressure: 1000 psig Reactor Water Level 120 inches Control Rod position Unknown APRMs Downscale Drywell pressure: 3 psig Torus pressure: 2 psig Torus water temp: 152°F Torus water level: -36 inches (REFERENCE PROVIDED)

Which one of the following identifies the required actions for reactor pressure control?

A. Exit the RC/P flowpath of ATWS, and go to OEOP-01 -EDP, Emergency Depressurization.

B. Exit the RC/P flowpath of RVCP, and go to OEOP-01-EDP, Emergency Depressurization.

C. Remain in the RC/P flowpath of ATWS, and exceed 100°F/hr cooldown rate if necessary.

D. Remain in the RC/P flowpath of RVCP, and exceed 100°F/hr cooldown rate if necessary.

Page: 89

90.

Unit Two is operating at rated power.

PCCP has been entered due to high torus water temperature with the c E.. 04 following plant conditions:

UA-12 (3-3) Rx Bldg Duff Press High/Low, is in alarm.

7 E UA-05 (6-10) Rx Bldg Isolated, is in alarm.

C o 2 Reactor Building Pressure (indication on the left)

Which one of the following completes both statements below?

Reactor Building pressure is (1) 0

The CR5 will direct Reactor Building HVAC restarted lAW (2)

A. (1) positive (2) 20P-37.1, Reactor Building Heating and Ventilation System Operating Procedure B. (1) positive (2) OEOP-01 -SEP-04, Reactor Building HVAC Restart Procedure C. (1) negative (2) 20P-37.1, Reactor Building Heating and Ventilation System Operating Procedure D. (1) negative (2) OEOP-01 -SEP-04, Reactor Building HVAC Restart Procedure Page: 90

91. A release on Unit Two is occurring with the following plant conditions:

Main Stack Rad Monitor, D12-RM-23S, is reading 2.3E+08 pCI/sec Turbine Building Vent Rad Monitor, D12-RM-23, is reading 2.5E+07 pCI/sec Real-time dose assessment using actual meteorology indicates 0.92 Rem TEDE and 5.1 Rem thyroid CDE at the site boundary (REFERENCE PROVIDED)

Which one of the following completes both statements below?

lAW RRCP, Unit One (1) override and reset the main stack hi-hi isolation signal.

The highest EAL classification for this event is (2)

A. (1) can (2) Site Area Emergency B. (1) can (2) General Emergency C. (1) can NOT (2) Site Area Emergency D. (1) can NOT (2) General Emergency Page: 91

92. Unit One and Unit Two are executing OASSD-O1, Alternative Sale Shutdown Procedure Index, due to a fire in Main Control Room back panels requiring Main Control Room evacuation. Current plant conditions are:

Unit One and Two have scrammed All MSIVs are shut Which one of the following completes both statements below?

The CRS will enter OASSD-02, Control Building, and (1) OASSD-O1.

The CRS will direct actions to achieve a safe shutdown using (2)

A. (1) exit (2) HPCI B. (1) exit (2) RCIC C. (1) concurrently perform (2) HPCI D. (1) concurrently perform (2) RCIC Page: 92

93. Unit One is in MODE 4, when a loss of SDC occurs due to RCS leakage.

cuf UNPLANNED loss of RPV inventory for 15 minutes or longer I 14151 I cutl UNPLANNED loss of reactor coolant results in RP\ water level less than a required lower limit rot 15 mm. (Note 1)

Which one of the following completes both of the statements below?

The minimum required RPV water level to support natural circulation is (1) lAW OPEP-02.2.1, EmergencyAction Level Technical Bases, the Unusual Event required lower limit is defined as RPV water level less than (2)

A. (1) 200 inches (2) 105 inches B. (1) 200 inches (2) 166 inches C. (1) 254 inches (2) 105 inches D. (1) 254 inches (2) 166 inches Page: 93

94. Which one of the following completes both statements below?

(Consider each statement separately.)

lAW Tech Spec 5.2.2, Facility Staff, the shift crew composition may be less than the minimum requirement for a period of time not to exceed (1) for an unexpected absence of on-duty shift crew members.

lAW 001-01.01, BNP Conduct of Operations Supplement, the minimum required number of Auxiliary Operators for manning a shift at BNP is (2)

A. (1) one hour (2) three B. (1) one hour (2) nine C. (1) two hours (2) three D. (1) two hours (2) nine Page: 94

95. Following the bypass of Unit Two feedwater heaters 4A and 5A, the following plant conditions exist:

Reactor Power is 60%

Feedwater Temperature is 330°F Final Feedwater Temperature vs Power Nominal FW Temp 11 0.3F Reduced RX PWR Nominal FW Temp Reduced lOT FRVT 65% 3944 384.4 296A 64% 393.1 383.1 295.5 63% 391.7 381.7 294.6 62% 390.4 380.4 293.7 61% 389.0 379.0 2928 60% 387.6 3776 291.9 mrs n ne_s n (REFERENCE PROVIDED) lAW 001-01.01, BNP Conduct of Operations Supplement, which one of the following completes both statements below? (consider each statement separately)

The CRS (1) required to implement the thermal limit penalties for FHOOS (feedwater heater out of service).

Entry into Tech Spec 3.O3 (2) required if final feedwater temperature is less than the 110.3°F reduced final feedwater temperature value.

A. (1) is (2) is B. (1) is (2) is NOT C. (1) is NOT (2) is D. (1) is NOT (2) is NOT Page: 95

96. Unit One is operating at rated power.

A-03 (2-2) Auto Depress Control Pwr Failure, is in alarm due to Fuse F5 being blown.

(REFERENCE PROVIDED)

Which one of the following completes both statements below?

Fuse F5 (Dl on l-FP-05687) is located on ADS Logic (1)

ADS (2) operable.

A. (1)A (2) is B. (1) A (2) is NOT C. (1) B (2) is D. (1) B (2) is NOT Page: 96

97. Unit Two is operating at rated power.

While performing OPT-07.2.4A, Core Spray Loop A Operability, Core Spray Room Cooler A fails to start when Core Spray Pump A is started.

The reactor building AO reports that the room cooler tripped on thermal overload.

lAW AD-OP-ALL-I 000, Conduct of Operations, which one of the following completes both statements below? (consider each statement separately)

Core Spray Loop A is (I)

A one time reset of the thermal overload (2) allowed before a Maintenance and Engineering evaluation.

A. (I) OPERABLE (2) is B. (1) OPERABLE (2) is NOT C. (1) INOPERABLE (2) is D. (I) INOPERABLE (2) is NOT Page: 97

98. Following a small steam line break in the drywell plant conditions are as follows:

Drywefl pressure: 25 psig and rising Drywell hydrogen: 1.3%

Suppression Chamber hydrogen: 1.2%

Torus level: 42 inches Which one of the following completes both statements below?

The CRS is required to direct venting containment lAW OEOP-O1-SEP-O1, Primary Containment Venting, using (1)

Venting of the (2) will be directed first.

A. (1) Section 2.1, Containment Pressure Control (2) drywell B. (1) Section 2.1, Containment Pressure Control (2) torus C. (1) Section 2.2, Containment Hydrogen Control (2) drywell D. (1) Section 2.2, Containment Hydrogen Control (2) torus Page: 98

99. Unit Two is operating at rated power with LPCI A inoperable and the following sequence of events occurs:

0000 7 day completion time for LCO 3.5.1, ECCS Operating, Condition A expires and Condition C is entered requiring that the Unit be placed in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

0030 Plant shutdown is commenced per LCO 3.5.1, Condition C.

0050 LPCI A is repaired and declared operable; LCO 3.5.1 Conditions A and C are exited.

0100 Management decides to continue the plant shutdown as planned to complete other maintenance items.

0230 Unit Two in MODE 3 (REFERENCE PROVIDED)

Which one of the following completes both statements below?

lAW 01-01 .07, Notifications, an Emergency Notification System (ENS) report to the NRC must be submitted no later than (1) lAW 10 CFR 50.73, Licensee Event Reporting System, an LER (2) required.

A. (1) 0400 (2) is B. (1) 0400 (2) is NOT C. (1) 0430 (2) is D. (1) 0430 (2) is NOT Page: 99

100. Unit One and Unit Two have entered SBO procedures at time 1300 due to a loss of all onsite and ofisite power.

Which one of the following completes both statements below?

lAW IEOP-01-SBO, Station Blackout, opening the reactor building roof hatch is required to be performed no later than (1) lAW 00 1-37.14, Station Blackout Procedure Basis Document, the reactor building doors and roof hatch are opened to ensure (2)

A. (1) 1330 (2) equipment availability B. (1) 1330 (2) habitability C. (1) 1500 (2) equipment availability D. (1) 1500 (2) habitability Page: 100

SRO Written Exam Reference Index

1. 0EOP-01-NL, EOP/SAMG Numerical Limits and Values, Attachment 3, Containment Parameters, Secondary Containment Area Temperature Limits, Table 3-B
2. 0EOP-01-SBO-01, Attachment 4, Drywell Temperature Calculation Using RSDP Recorder Inputs
3. 0EOP-01-UG, Users Guide, Attachment 7, Heat Capacity Temperature Limit
4. 0EOP-01-UG, Users Guide, Attachments 19 (RPV Saturation Limit), 22 Shutdown Range Level Instrument (N027A, B) Caution), and 31 (RPV Level Caution, pages 1 & 2)
5. 0EOP-01-UG, Users Guide, Attachment 26, Unit 2 RPV Level at LL4
6. 0OI-01-07, Notifications, Attachment 1, Reportability Evaluation Checklist
7. NUREG 1022, Event Report Guidelines, Table 1, Reportable Events
8. 1OP-27, Attachment 2, Estimated Capability Curves
9. ODCM 7.3.10, Gaseous Radwaste Treatment System
10. Tech Spec 3.2.1, Average Planar Linear Heat Generation Rate
11. Tech Spec 3.2.2, Minimum Critical Power Ratio
12. Tech Spec 3.2.3, Linear Heat Generation Rate
13. Tech Spec 3.3.1.1, Reactor Protection System (RPS) Instrumentation
14. Tech Spec 3.3.5.1, Emergency Core Cooling System (ECCS)

Instrumentation

15. Tech Spec 3.6.1.3, Primary Containment Isolation Valves (PCIVs)
16. Tech Spec 3.8.1, AC Sources - Operating
17. 1-FP-05887, Auto Depressurization System Elementary Diagram Unit
18. 10PEP-02.1, Brunswick Nuclear Plant Initial Emergency Actions

ATTACHMENT 3 Page 73 of 87 Containment Parameters Secondary Containment Area Temperature Limits Table 3-B PLANT PLANT LOCATION MAX NORM MAX SAFE AUTO GROUP AREA DESCRIPTION OPERATING OPERATING ISOLATION VALUE (°F) VALUE (°F)

N CORE N CORE SPRAY 120 175 N/A SPRAY ROOM S CORE S CORE SPRAY 120 175 N/A SPRAY ROOM RWCU PMP ROOM A PMP ROOM B 140 225 3 HX ROOM N RHR N RHR EQUIP ROOM 175 295 N/A S RHR S RHR EQUIP ROOM 175 295 N/A RCIC EQUIP ROOM 165 295 5 HPCI HPCI EQUIP ROOM 165 165 4 STEAM RCIC STM TUNNEL 190 295 5 TUNNEL HPCI STM TUNNEL 190 295 4 20 FT 2O FT NORTH 140 200 N/A 20 FT SOUTH 140 200 N/A 50 FT 50 FT NW 140 200 N/A 50 FT SE 140 200 N/A REACTOR MULTIPLE AREAS ALARM N/A 3, 4, AND/OR 5 BLDG ANNUN. SETPOINT A-02 5-7 REACTOR MSIV PIT ANNUN. ALARM N/A 1 BLDG A-06 6-7 SETPOINT 0EOP-01-NL Rev. 27 Page 158 of 258

PLANT MONITORING 0EOP-01-SBO-01 (PAS)

Rev. 0 Page 16 of 18 ATTACHMENT 4 Page 1 of 1 Drywell Temperature Calculation Using RSDP Recorder Inputs Values obtained from Recorder CAC-TR-778 Above 70' Elevation PT 1 x 0.141 = °F Between 28' and 45' Elevation PT 3 x 0.404 = °F Between 10' and 23' Elevation PT 4 x 0.455 = °F Average Drywell Temperature °F (Sum of 3 Regional Weighted Areas)

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 74 of 156 ATTACHMENT 7 Page 1 of 1

<< Heat Capacity Temperature Limit >>

Torus water temperature is determined by:

  • CAC-TR-4426-1A, Point Wtr Avg OR
  • CAC-TR-4426-2A, Point Wtr Avg OR
  • Computer point G050 OR
  • Computer point G051 OR
  • CAC-TY-4426-1 OR
  • CAC-TY-4426-2 Select graph line immediately below torus water level as the limit.

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 87 of 156 ATTACHMENT 19 Page 1 of 1

<< RPV Saturation Limit >>

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 90 of 156 ATTACHMENT 22 Page 1 of 1

<< Shutdown Range Level Instrument (N027A, B) Caution >>

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 99 of 156 ATTACHMENT 31 Page 1 of 4

<< RPV Level Caution >>

Caution 1 A RPV level instrument may be used to determine RPV level only when the conditions for use specified below are satisfied for that instrument.

NOTE

  • Reference leg area drywell temperature is determined using Attachment 18, Level Instrument Reference Leg Area Drywell Temperature Calculations, ERFIS or Instructional Aid based on Attachment 18. ...............................
  • If the temperature near any instrument run is in the UNSAFE region of the Attachment 19, RPV Saturation Limit, the instrument may be unreliable due to boiling in the run. ...........................................................................................................
  • Immediate reference leg boiling is not expected to occur for short duration excursions into the unsafe region due to heating of the drywell. The thermal time constant associated with the mass of metal and water in the reference leg will prohibit immediate boiling of the reference leg. Reference leg boiling is an obvious phenomenon. Large scale oscillations of all water level instruments associated with the reference leg that is boiling will occur. This occurrence will be obvious and readily observable by the operator. Additionally, if the operator is not certain whether boiling has occurred, he can refer to plant history as provided on water level recorders or ERFIS. Reference leg boiling is indicated by level oscillations without corresponding pressure oscillations. .................................

Instrument Conditions for Use Narrow Range Level Instruments Unit 1 Only: The indicated level is in the C32-LI-R606A, B, C (N004A, B, C) SAFE region of Attachment 20.

C32-LPR-R608 (N004A, B) Unit 2 Only: The indicated level is in the Indicating Range 150-210 Inches SAFE region of Attachment 21.

Cold Reference Leg Shutdown Range Level Instruments The indicated level is in the SAFE region of B21-LI-R605A, B (N027A, B) Attachment 22.

Indicating Range 150-550 Inches Cold Reference Leg To determine RPV level at the Main Steam Line Flood Level (MSL), see Attachment 30.

Attachment 30 has two curves: The upper curve is for reference leg area drywell temperature equal to or greater than 200°F.

The lower curve is for reference leg area drywell temperature less than 200°F.

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 100 of 156 ATTACHMENT 31 Page 2 of 4

<< RPV Level Caution >>

Caution 1 (Continued)

Instrument Conditions for Use Wide Range Level Instruments

  • Temperature on the Reactor Building B21-LI-R604A, B (N026A, B) 50' below 140°F (B21-XY-5948A C32-PR-R609 (N026B) A2-4, B21-XY-5948B A2-4, Indicating Range 0-210 Inches ERFIS Computer Point B21TA102, Cold Reference Leg OR B21TA103)

AND

  • IF the reference leg area drywell temperature is in the UNSAFE region of Attachment 19, RPV Saturation Limit, THEN the indicated level is greater than 20 inches OR IF the reference leg area drywell temperature is in the SAFE region of Attachment 19, RPV Saturation Limit, THEN the indicated level is greater than 10 inches.

USER'S GUIDE 0EOP-01-UG Rev. 067 Page 94 of 156 ATTACHMENT 26 Page 1 of 1

<< Unit 2 RPV Level at LL 4 (Minimum Steam Cooling RPV Level) >>

When RPV pressure is less than 60 psig, use indicated level. LL-4 is -27.5 inches.

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 26 of 46 ATTACHMENT 1 Page 1 of 8

<< Reportability Evaluation Checklist >>

NOTE If the answer to the following question is YES, then Accelerated Verbal Notification to the NRC is required within 15 minutes. Reference 0AOP-40.0, Security Events, for notification content.

15 MINUTE REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION NA Has a Hostile Action occurred? [NRC Bulletin 2005-02]

NOTE

  • NUREG-1022, Rev. 3 is a reference to provide additional guidance on reportability.
  • If the answer to any of the following questions is YES, the event is reportable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
  • If all answers to the following questions are NO, the event is not reportable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

1 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION Is the event a deviation from technical specifications as per 10 CFR 50.54(X)?

1.1

[10 CFR 50.72(b)(1)]

1.2 Does the event involve by-product, source or special nuclear material possessed by the licensee that might have or threatens to cause:

Any individual's exposure to reach or exceed 25 Rems total effective dose 1.2.1 equivalent (TEDE); 75 Rems eye dose equivalent; or 250 Rads shallow-dose equivalent to the skin or extremities? [10 CFR 20.2202(a)(1)]

The release of radioactive material inside or outside of a restricted area, such that, 1.2.2 had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the individual could have received an intake 5 times the occupational annual limit on intake? [10 CFR 20.2202(a)(2)]

ISFSI - Does the event involve accidental criticality or loss of any special nuclear 1.3 material? [10 CFR 72.74(a)]

Does the event involve the discovery of a cyber attack that adversely impacted safety related or important-to-safety functions, security functions, or emergency preparedness functions (including offsite communications); or that compromised 1.4 support systems and equipment resulting in adverse impacts to safety, security, or emergency preparedness functions within the scope of 10 CFR 73.54? (Note 1)

[10 CFR 73.77(a)(1)]

Notes:

1. Assistance with 10 CFR 73.77 reporting can be provided by the CSIRT.

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 27 of 46 ATTACHMENT 1 Page 2 of 8

<< Reportability Evaluation Checklist >>

NOTE

  • If the answer to any of the following questions is YES, the event is reportable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
  • If all answers to the following questions are NO, the event is not reportable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION Is plant shutdown required by technical specifications being initiated? (Note 1) 2.1

[10 CFR 50.72(b)(2)(i)]

Has the event resulted in or should have resulted in an Emergency Core Cooling 2.2 System (ECCS) discharge into the Reactor Coolant System as a result of a valid signal, except when the actuation resulted from and was part of a pre-planned sequence during testing or reactor operation? [10 CFR 50.72(b)(2)(iv)(A)]

Did the event or condition result in actuation of the reactor protection system (RPS) 2.3 when the reactor was critical, except when the actuation resulted from and was part of a pre-planned sequence during testing or reactor operation?[10 CFR 50.72(b)(2)(iv)(B)]

Is the event a situation, as related to the health and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or 2.4 notification to other government agencies has been or will be made? (Note 2) (Note 3)

[10 CFR 50.72(b)(2)(xi)]

[10 CFR 72.75(b)(2)]

Notes:

1. Includes any Safety Limit violation (Tech Spec 2.2)
2. Such an event may include an on-site fatality or an inadvertent release of radioactively contaminated materials.
3. The North Carolina Wildlife Commission's Sea Turtle Coordinator (NCSTC) is notified of each sea turtle recovery. A report per 10 CFR 50.72(b)(2)(xi) is required 1) when a dead turtle is recovered OR 2) when, after consultation with the NCSTC, it is determined that an injured turtle requires rehabilitation versus release. The NRC notification is required no later than 4 hrs after consultation with the NCSTC when either of these conditions is met.

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 28 of 46 ATTACHMENT 1 Page 3 of 8

<< Reportability Evaluation Checklist >>

4 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION Has any licensed material been lost, stolen, or missing in an aggregate quantity equal to or greater than 1,000 times the quantity specified in 10 CFR 20 Appendix C under 2.5 such circumstances that it appears that an exposure could result to persons in unrestricted areas? (Note 1)

[10 CFR 20.2201(a)(i)]

ISFSI - Departure from License Condition.

Has an action been taken in an emergency that departs from a condition or a technical specification contained in a license or certificate of compliance issued under 10 CFR 72 2.6 when the action was immediately needed to protect the public health and safety and no action consistent with license conditions or technical specifications that could provide adequate or equivalent protection was immediately apparent as per 72.32(d)?

[10 CFR 72.75(b)(1)]

Does the event involve discovery of a cyber attack that could have caused an adverse impact to safety related or important-to-safety functions, security functions, or emergency preparedness functions (including offsite communications); or that could 2.7 have compromised support systems and equipment, which if compromised could have adversely impacted safety, security, or emergency preparedness functions within the scope of 10 CFR-73.54? (Note 2)

[10 CFR 73.77(a)(2)(i)]

Does the event involve discovery of a suspected or actual cyber attack initiated by personnel with physical or electronic access to digital computer and communication 2.8 systems and networks within the scope of 10 CFR 73.54? (Note 2)

[10 CFR 73.77(a)(2)(ii)]

Does the event involve notification of a local, State, or other Federal agency (e.g., law enforcement, FBI, etc.) of an event related to implementation of the cyber security program for digital computer and communication systems and networks within the 2.9 scope of 10 CFR 73.54 that does not otherwise require a notification under paragraph (a) of this section? (Note 2)

[10 CFR 73.77(a)(2)(iii)]

Notes:

1. Further information is located in AD-SY-ALL-0150, Reporting Safeguards, Security, and Fitness for Duty Events.
2. Assistance with 10 CFR 73.77 reporting can be provided by the CSIRT.

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 29 of 46 ATTACHMENT 1 Page 4 of 8

<< Reportability Evaluation Checklist >>

NOTE

  • If the answer to any of the following questions is YES, the event is reportable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
  • If all the answers to the following questions are NO, the event is not reportable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

8 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION Has the event or condition resulted in the condition of the nuclear power plant, 3.1 including its principal safety barriers, being seriously degraded?

[10 CFR 50.72(b)(3)(ii)(A)]

Has the event or condition resulted in the nuclear power plant being in an unanalyzed 3.2 condition that significantly degrades plant safety?

[10 CFR 50.72(b)(3)(ii)(B)]

Did the event or condition result in valid actuation of any of the systems listed below 3.3 except when the actuation resulted from and is part of a pre-planned sequence during testing or reactor operation? (Note 1)

[10 CFR 50.72(b)(3)(iv)(A)]

These systems are:

3.3.1 Reactor protection system (RPS) including: reactor scram and reactor trip.

[10 CFR 50.72(b)(3)(iv)(B)(1)]

Notes:

1. Automatic OR Manual initiation of the system listed is reportable. NUREG-1022, Section 3.2.6 discussion, should be referenced for additional information.

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 30 of 46 ATTACHMENT 1 Page 5 of 8

<< Reportability Evaluation Checklist >>

8 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 3.3.2 General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).

  • HPCI Steam Line Isolation.
  • RCIC Steam Line Isolation.
  • RWCU Suction Isolation.
  • Combustible Gas Control (CAD).

[10 CFR 50.72(b)(3)(iv)(B)(2)]

3.3.3 Emergency core cooling systems (ECCS), including:

  • Automatic Depressurization (ADS) System

[10 CFR 50.72(b)(3)(iv)(B)(4)]

3.3.4 Reactor Core Isolation Cooling (RCIC)

[10 CFR 50.72(b)(3)(iv)(B)(5)]

3.3.5 Containment heat removal and depressurization systems including containment spray and fan cooler systems.

  • RHR Suppression Pool Cooling.
  • Drywell Spray System Actuation.

[10 CFR 50.72(b)(3)(iv)(B)(7)]

3.3.6 Emergency Diesel Generators (DGs)

[10 CFR 50.72(b)(3)(iv)(B)(8)]

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 31 of 46 ATTACHMENT 1 Page 6 of 8

<< Reportability Evaluation Checklist >>

8 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 3.4 Could the event or condition at the time of discovery have prevented the fulfillment of the safety function of structures or systems that are needed to:

[10 CFR 50.72(b)(3)(v)]

These criteria cover an event or condition in which scoped in SSCs could have failed to perform their intended function because of one or more personnel errors, including procedure violations; equipment failures; inadequate maintenance; or design, analysis, fabrication, equipment qualification, construction, or procedural deficiencies and no redundant equipment in the same system was OPERABLE.

However, individual component failures need not be reported if redundant equipment in the same system was OPERABLE and available to perform the required safety function. (Note 1) [10 CFR 50.72(b)(3)(vi)]

Shut down the reactor and maintain it in a safe shutdown 3.4.1 condition? (Note 1) [10 CFR 50.72(b)(3)(v)(A)]

Remove residual heat?

3.4.2

[10 CFR 50.72(b)(3)(v)(B)]

Control the release of radioactive material?

3.4.3

[10 CFR 50.72(b)(3)(v)(C)]

3.4.4 Mitigate the consequences of an accident? (Note 2)

[10 CFR 50.72(b)(3)(v)(D)]

Does the event require the transport of a radioactively contaminated person to an off-site medical facility for treatment?

3.5 [10 CFR 50.72(b)(3)(xii)]

[10 CFR 72.75(c)(3)]

Notes:

1. No Event Notification [i.e., per 10 CFR 50.72(b)(3)(v)] is required for conditions which could have prevented fulfillment of the safety function that are discovered when the affected system is INOPERABLE or when the affected system is INOPERABLE but considered available. If the condition is discovered when the system is OPERABLE, an EN will be made per 10 CFR 50.72(b)(3)(v).
2. RCIC INOPERABILITY is not reportable as a single train system per 10 CFR 50.72(b)(3)(v)(d). TS Basis 3.5.3 states that the RCIC System is not an ESF system and no credit is taken in the safety analysis for RCIC System operation. As such, consistent with Example 2 on NUREG 1022, Revision 3, RCIC Failure is not reportable under 10 CFR 50.72(b)(3)(v)(d).

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 32 of 46 ATTACHMENT 1 Page 7 of 8

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NOTE

  • Additional reportability guidance concerning loss of emergency preparedness capabilities is contained in Section 3.2.13 of NUREG-1022, Rev. 3. Consultation with an Emergency Preparedness representative is advised when assessing the significance of the loss of capability.

8 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 3.6 Has the event resulted in a major loss of emergency assessment capability, off-site response capability, or communications capability (i.e., significant portion of the Main Control Room indication, emergency notification system, or off-site notification system)?

[10 CFR 50.72(b)(3)(xiii)]

Major loss of emergency or off-site notification system is considered to be/but not limited to:

a. Loss of:
1) DEMNET; OR NRC Emergency Notification System (ENS);

AND

2) Commercial telephone network.
b. INOPERABILITY for greater than or equal to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of:
1) Seven or more off-site sirens; OR
2) All off-site sirens in one county.

ISFSI -Important to Safety Defect Has a defect been discovered in any Independent Spent Fuel Storage structure, 3.7 system, or component that is important to safety?

[10 CFR 72.75(c)(1)]

ISFSI - Reduction in Effectiveness Has a condition been discovered which results in a significant reduction in the 3.8 effectiveness of any Independent Spent Fuel Storage cask confinement system during use?

[10 CFR 72.75(c)(2)]

NOTIFICATIONS 0OI-01.07 Rev. 38 Page 33 of 46 ATTACHMENT 1 Page 8 of 8

<< Reportability Evaluation Checklist >>

NOTE Additional cyber security event reportability guidance is contained in NEI 15-09, Revision 0.

Consultation with Nuclear Information Technology and Licensing is advised when assessing the significance of cyber security events.

8 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 3.9 Does the event involve receipt or collection of information regarding observed behavior, activities, or statements that may indicate intelligence gathering or pre-operational planning related to a cyber attack against digital computer and communication systems and networks within the scope of 10 CFR 73.54? (Note 1)

[10 CFR 73.77(a)(3)]

NOTE If the answer to any of the following questions is YES, the event is reportable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

24 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION 4.1 Does the incident involve the loss of control of licensed material possessed by BNP which might have caused or threatens to cause:

4.1.1 Any individual's exposure in a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to exceed: 5 Rems total effective dose equivalent (TEDE); or 15 Rems eye dose equivalent; or 50 Rems shallow-dose equivalent to the skin or extremities? [10 CFR 20.2202(b)(1)]

The release of radioactive material inside or outside of a restricted area, so that, had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the individual could have received 4.1.2 an intake in excess of one occupational annual limit on intake?

[10 CFR 20.2202(b)(2)]

4.2 ISFSI - Equipment Important to Safety Disabled or Failed to Function Does the event involve equipment important to safety which is disabled or fails to function as designed when:

The equipment is required by certificate of compliance to be available and OPERABLE to prevent releases that could exceed regulatory limits, to prevent exposures to radiation or radioactive materials that could exceed regulatory limits, or to mitigate the consequences of an accident; and, No redundant equipment was available and OPERABLE to perform the required safety function. [10 CFR 72.75(d)(1)]

Notes:

1. Assistance with 10 CFR 73.77 reporting can be provided by the CSIRT.

NUREG-1022, Rev. 3 Event Report Guidelines 10 CFR 50.72 and 50.73 Final Report Manuscript Completed: January 2013 Date Published: January 2013 Prepared by:

Aron Lewin Office of Nuclear Reactor Regulation

EXECUTIVE

SUMMARY

Two of the many elements contributing to the safety of nuclear power are emergency response and the feedback from operating experience into plant operations. These are achieved partly by the licensee event reporting requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73, Licensee Event Report System. In 10 CFR 50.72, the U.S. Nuclear Regulatory Commission (NRC) provides for immediate notification requirements through the emergency notification system, and in 10 CFR 50.73 provides for 60-day written licensee event reports.

The NRC staff uses the information reported under 10 CFR 50.72 and 10 CFR 50.73 in responding to emergencies, monitoring ongoing events, confirming licensing bases, studying potentially generic safety problems, assessing trends and patterns of operational experience, monitoring performance, identifying precursors of more significant events, and providing operational experience to the industry.

NUREG-1022 contains guidelines that the staff of the NRC considers acceptable for use in meeting the requirements of 10 CFR 50.72 and 10 CFR 50.73. Several identified reporting issues could not be quickly resolved given certain ambiguities in the guidance in Revision 2 of NUREG-1022. In developing Revision 3 to NUREG-1022, the NRC held numerous public and internal meetings to solicit stakeholder input and feedback. In resolving the ambiguities, the NRC considered the provisions of the rule itself, the associated statements of consideration, and other available guidance in that hierarchal order. Revision 3 to NUREG-1022 revises the event reporting guidelines to provide clearer guidance.

vii

Table 1 Reportable Events Declaration of an Emergency Class (See Section 3.1.1 of this report)

§ 50.72(a)(1)(i) The declaration of any of the Emergency Classes specified in the licensees approved Emergency Plan.

Plant Shutdown Required by Technical Specifications (See Section 3.2.1 of this report)

§ 50.72(b)(2)(i) The initiation of any nuclear § 50.73(a)(2)(i)(A) The completion of any plant shutdown required by the plants Technical nuclear plant shutdown required by the plants Specifications. Technical Specifications.

Operation or Condition Prohibited by Technical Specifications (See Section 3.2.2 of this report)

§ 50.73(a)(2)(i)(B) Any operation or condition which was prohibited by the plants Technical Specifications except when:

(1) The Technical Specification is administrative in nature; (2) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions; or (3) The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event.

Deviation from Technical Specifications Authorized under § 50.54(x)

(See Section 3.2.3 of this report)

§ 50.72(b)(1) ... any deviation from the plants § 50.73(a)(2)(i)(C) Any deviation from the plants Technical Specifications authorized pursuant to Technical Specifications authorized pursuant to

§ 50.54(x) of this part. § 50.54(x) of this part.

Degraded or Unanalyzed Condition (See Section 3.2.4 of this report)

§ 50.72(b)(3)(ii) Any event or condition that 50.73(a)(2)(ii) Any event or condition that 3

Table 1 Reportable Events (continued) results in: resulted in:

(A) The condition of the nuclear power plant, (A) The condition of the nuclear power plant, including its principal safety barriers, being including its principal safety barriers, being seriously degraded; or seriously degraded; or (B) The nuclear power plant being in an (B) The nuclear power plant being in an unanalyzed condition that significantly unanalyzed condition that significantly degrades plant safety. degraded plant safety.

External Threat or Hampering (See Section 3.2.5 of this report)

§ 50.73(a)(2)(iii) Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.

System Actuation (See Section 3.2.6 of this report)

§ 50.72(b)(2)(iv)(A) Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

§ 50.72(b)(2)(iv)(B) Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

§ 50.72(b)(3)(iv)(A) Any event or condition that § 50.73(a)(2)(iv)(A) Any event or condition that results in valid actuation of any of the systems resulted in manual or automatic actuation of any listed in paragraph (b)(3)(iv)(B) of this section, of the systems listed in paragraph (a)(2)(iv)(B) of except when the actuation results from and is this section, except when:

part of a pre-planned sequence during testing or reactor operation. (1) The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or 4

Table 1 Reportable Events (continued)

(2) The actuation was invalid and; (i) Occurred while the system was properly removed from service; or (ii) Occurred after the safety function had been already completed.

§ 50.72(b)(3)(iv)(B) The systems to which the § 50.73(a)(2)(iv)(B) The systems to which the requirements of paragraph (b)(3)(iv)(A) of this requirements of paragraph (a)(2)(iv)(A) of this section apply are: section apply are:

(1) Reactor protection system (RPS) including: (1) Reactor protection system (RPS) including:

reactor scram and reactor trip.5 reactor scram or reactor trip.

(2) General containment isolation signals (2) General containment isolation signals affecting containment isolation valves in more affecting containment isolation valves in more than one system or multiple main steam than one system or multiple main steam isolation valves (MSIVs). isolation valves (MSIVs).

(3) Emergency core cooling systems (ECCS) for (3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including: pressurized water reactors (PWRs) including:

high-head, intermediate-head, and low-head high-head, intermediate-head, and low-head injection systems and the low pressure injection systems and the low pressure injection function of residual (decay) heat injection function of residual (decay) heat removal systems. removal systems.

(4) ECCS for boiling water reactors (BWRs) (4) ECCS for boiling water reactors (BWRs) including: high-pressure and low-pressure including: high-pressure and low-pressure core spray systems; high-pressure coolant core spray systems; high-pressure coolant injection system; low pressure injection injection system; low pressure injection function of the residual heat removal system. function of the residual heat removal system.

(5) BWR reactor core isolation cooling system; (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater isolation condenser system; and feedwater coolant injection system. coolant injection system.

(6) PWR auxiliary or emergency feedwater (6) PWR auxiliary or emergency feedwater system. system.

(7) Containment heat removal and (7) Containment heat removal and depressurization systems, including depressurization systems, including containment spray and fan cooler systems. containment spray and fan cooler systems.

5 Actuation of the RPS when the reactor is critical is reportable under § 50.72(b)(2)(iv)(B).

5

Table 1 Reportable Events (continued)

(8) Emergency ac electrical power systems, (8) Emergency ac electrical power systems, including: emergency diesel generators including: emergency diesel generators (EDGs); hydroelectric facilities used in lieu of (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs. dedicated Division 3 EDGs.

(9) Emergency service water systems that do not normally run and that serve as ultimate heat sinks.

Event or Condition that Could Have Prevented Fulfillment of a Safety Function (See Section 3.2.7 of this report)

§ 50.72(b)(3)(v) Any event or condition that at § 50.73(a)(2)(v) Any event or condition that the time of discovery could have prevented the could have prevented the fulfillment of the safety fulfillment of the safety function of structures or function of structures or systems that are needed systems that are needed to: to:

(A) Shut down the reactor and maintain it in a (A) Shut down the reactor and maintain it in a safe shutdown condition; safe shutdown condition; (B) Remove residual heat; (B) Remove residual heat; (C) Control the release of radioactive material; or (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. (D) Mitigate the consequences of an accident.

§ 50.72(b)(3)(vi) Events covered in § 50.73(a)(2)(vi) Events covered in paragraph (b)(3)(v) of this section may include paragraph (a)(2)(v) of this section may include one or more procedural errors, equipment one or more procedural errors, equipment failures, and/or discovery of design, analysis, failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural fabrication, construction, and/or procedural inadequacies. However, individual component inadequacies. However, individual component failures need not be reported pursuant to failures need not be reported pursuant to paragraph (b)(3)(v) of this section if redundant paragraph (a)(2)(v) of this section if redundant equipment in the same system was operable and equipment in the same system was operable and available to perform the required safety function. available to perform the required safety function.

Common Cause Inoperability of Independent Trains or Channels (See Section 3.2.8 of this report)

§ 50.73(a)(2)(vii) Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; 6

Table 1 Reportable Events (continued)

(B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

Radioactive Release (See Section 3.2.9 of this report)

§ 50.73(a)(2)(viii)(A) Any airborne radioactive release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in appendix B to part 20, table 2, column 1.

§ 50.73(a)(2)(viii)(B) Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, exceeds 20 times the applicable concentrations specified in appendix B to part 20, table 2, column 2, at the point of entry into the receiving waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases.

Internal Threat or Hampering (See Section 3.2.10 of this report)

§ 50.73(a)(2)(x) Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.

Transport of a Contaminated Person Offsite (See Section 3.2.11 of this report)

§ 50.72(b)(3)(xii) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.

News Release or Notification of Other Government Agency (See Section 3.2.12 of this report)

§ 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 7

Table 1 Reportable Events (continued) 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 radioactively contaminated materials.

Loss of Emergency Preparedness Capabilities (See Section 3.2.13 of this report)

§ 50.72(b)(3)(xiii) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, emergency notification system, or offsite notification system).

Single Cause that Could Have Prevented Fulfillment of the Safety Functions of Trains or Channels in Different Systems (See Section 3.2.14 of this report)

§ 50.73(a)(2)(ix)(A) Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

(1) Shut down the reactor and maintain it in a safe shutdown condition; (2) Remove residual heat; (3) Control the release of radioactive material; or (4) Mitigate the consequences of an accident.

§ 50.73(a)(2)(ix)(B) Events covered in paragraph (ix)(A) of this section may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to paragraph (ix)(A) of this section if the event results from:

(1) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (2) Normal and expected wear or degradation.

8

GENERATOR AND EXCITER SYSTEM OPERATING 1OP-27 PROCEDURE Rev. 63 Page 60 of 70 ATTACHMENT 2 Page 1 of 1

<< Estimated Capability Curves >>

GASEOUS RADWASTE TREATMENT SYSTEM 7.3.10 7.3.10 GASEOUS RADWASTE TREATMENT SYSTEM ODCMS 7.3.10 The GASEOUS RADWASTE TREATMENT SYSTEM shall be in operation.

APPLICABILITY: Whenever the Main Condenser Air Ejector (evacuation) System is in operation.

COMPENSATORY MEASURES CONDITION REQUIRED COMPENSATORY COMPLETION MEASURE TIME A. GASEOUS RADWASTE A.1 Place GASEOUS 7 days TREATMENT SYSTEM not RADWASTE in operation. TREATMENT SYSTEM in operation.

B. B.1 Submit a Special Report to 30 days the NRC that identifies the NOTE required inoperable Required Compensatory equipment and the Measure B.1 shall be reasons for the completed if this Condition is inoperability, corrective entered. actions taken to restore Required Compensatory the required inoperable measure and associated equipment to OPERABLE Completion Time not met. status, and a summary description of the corrective actions taken to prevent recurrence.

TEST REQUIREMENTS TEST FREQUENCY TR 7.3.10.1 Verify GASEOUS RADWASTE TREATMENT 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SYSTEM in operation by checking the readings of the relevant instruments.

Brunswick Units 1 and 2 7.3.10-1 Rev. 25

APLHGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

LCO 3.2.1 All APLHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER 23% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any APLHGR not within A.1 Restore APLHGR(s) to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 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 Time POWER to < 23% RTP.

not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 23% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Brunswick Unit 2 3.2-1 Amendment No. 247

MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.

APPLICABILITY: THERMAL POWER 23% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A.1 Restore MCPR(s) to within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limits. limits.

B. Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion to < 23% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the limits Once within specified in the COLR. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 23% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter (continued)

Brunswick Unit 2 3.2-2 Amendment No. 247

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 COLR.

APPLICABILITY: THERMAL POWER 23% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 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 < 23% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 23% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Brunswick Unit 2 3.2-4 Amendment No. 274

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------------- 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

Place associated trip system in trip.

(continued)

Brunswick Unit 2 3.3-1 Amendment No. 243

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. ------------NOTE---------------- B.1 Place channel in one trip 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Not applicable for Functions system in trip.

2.a, 2.b, 2.c, 2.d, or 2.f.

OR One or more Functions with B.2 Place one trip system in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> one or more required trip.

channels inoperable in both trip systems.

C. One or more Functions with C.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 /> RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion Time referenced in of Condition A, B, or C not Table 3.3.1.1-1 for the 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 D.1 and referenced in POWER to < 26% RTP.

Table 3.3.1.1-1.

(continued)

Brunswick Unit 2 3.3-2 Amendment No. 247

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 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 D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.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 /> Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully insert Immediately Action D.1 and referenced in all insertable control rods in Table 3.3.1.1-1. core cells containing one or more fuel assemblies.

I. As required by Required I.1 Initiate alternate method to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action D.1 and referenced in detect and suppress Table 3.3.1.1-1. thermal hydraulic instability oscillations.

J. Required Action and J.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 Time POWER to < 20% RTP.

of Condition I not met.

Brunswick Unit 2 3.3-3 Amendment No. 243

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

NOTES ----------------------------------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 (Not used.) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.1.1.2 Perform CHANNEL CHECK. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.3.1.1.3 --------------------------------NOTE--------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER 23% RTP.

Adjust the average power range monitor (APRM) 7 days channels to conform to the calculated power while operating at 23% RTP.

SR 3.3.1.1.4 --------------------------------NOTE--------------------------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 7 days (continued)

Brunswick Unit 2 3.3-4 Amendment No. 247

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.5 Perform a functional test of each automatic scram 7 days contactor.

SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to withdrawing intermediate range monitor (IRM) channels overlap. SRMs from the fully inserted position SR 3.3.1.1.7 --------------------------------NOTE--------------------------------

Only required to be met during entry into MODE 2 from MODE 1.

Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.8 Calibrate the local power range monitors. 2000 effective full power hours SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.10 Calibrate the trip units. 92 days (continued)

Brunswick Unit 2 3.3-5 Amendment No. 282

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.11 --------------------------------NOTES------------------------------

1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.
2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.

Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.1.1.13 --------------------------------NOTES------------------------------

1. Neutron detectors are excluded.
2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.
3. For Functions 2.b and 2.f, the recirculation flow transmitters that feed the APRMs are included.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.14 (Not used.)

SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months (continued)

Brunswick Unit 2 3.3-6 Amendment No. 243

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.16 Verify Turbine Stop ValveClosure and Turbine 24 months Control Valve Fast Closure, Trip Oil PressureLow Functions are not bypassed when THERMAL POWER is 26% RTP.

SR 3.3.1.1.17 --------------------------------NOTES----------------------------

1. Neutron detectors are excluded.
2. For Functions 3 and 4, the sensor response time may be assumed to be the design sensor response time.
3. For Function 5, "n" equals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.
4. For Function 2.e, "n" equals 8 channels for the purpose of determining the STAGGERED TEST BASIS Frequency. Testing of APRM and Oscillation Power Range Monitor (OPRM) outputs shall alternate.

24 months on a Verify the RPS RESPONSE TIME is within limits. STAGGERED TEST BASIS SR 3.3.1.1.18 Adjust the flow control trip reference card to conform Once within 7 days to reactor flow. after reaching equilibrium conditions following refueling outage (continued)

Brunswick Unit 2 3.3-7 Amendment No. 247

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.19 Verify OPRM is not bypassed when APRM Simulated 24 months Thermal Power is 25% and recirculation drive flow is 60%.

Brunswick Unit 2 3.3-8 Amendment No. 243

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors
a. Neutron FluxHigh 2 3 G SR 3.3.1.1.2 SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.15 5(a) 3 H SR 3.3.1.1.2 SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.13 SR 3.3.1.1.15
b. Inop 2 3 G SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.15 5(a) 3 H SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.15
2. Average Power Range Monitors (c)
a. Neutron FluxHigh (Setdown) 2 3 G SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 3(c) (b)
b. Simulated Thermal PowerHigh 1 F SR 3.3.1.1.2 SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) [0.55 (W - W) + 62.6% RTP] when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating." The value of W is defined in plant procedures.

(c) Each APRM channel provides inputs to both trip systems.

Brunswick Unit 2 3.3-9 Amendment No. 247

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors (continued)
c. Neutron FluxHigh 1 3(c) F SR 3.3.1.1.2 SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13
d. Inop 1,2 3(c) G SR 3.3.1.1.5 SR 3.3.1.1.11
e. 2-Out-Of-4 Voter 1,2 2 G SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.17
f. OPRM Upscale 20% RTP 3(c) I SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18 SR 3.3.1.1.19
3. Reactor Vessel Steam Dome Pressure 1,2 2 G SR 3.3.1.1.2 High SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17
4. Reactor Vessel Water LevelLow Level 1 1,2 2 G SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17
5. Main Steam Isolation ValveClosure 1 8 F SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17
6. Drywell PressureHigh 1,2 2 G SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 (continued)

(c) Each APRM channel provides inputs to both trip systems.

(d) See COLR for OPRM period based detection algorithm (PBDA) setpoint limits.

Brunswick Unit 2 3.3-10 Amendment No. 243

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

7. Scram Discharge Volume 1,2 2 G SR 3.3.1.1.5 Water LevelHigh SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 5(a) 2 H SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15
8. Turbine Stop ValveClosure 26% RTP 4 E SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17
9. Turbine Control Valve Fast Closure, 26% RTP 2 E SR 3.3.1.1.5 Control Oil PressureLow SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17
10. Reactor Mode Switch Shutdown Position 1,2 1 G SR 3.3.1.1.12 SR 3.3.1.1.15 5(a) 1 H SR 3.3.1.1.12 SR 3.3.1.1.15
11. Manual Scram 1,2 1 G SR 3.3.1.1.9 SR 3.3.1.1.15 5(a) 1 H SR 3.3.1.1.9 SR 3.3.1.1.15 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

Brunswick Unit 2 3.3-11 Amendment No. 247

ECCS Instrumentation 3.3.5.1 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.5.1-1.

ACTIONS

NOTE -----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.1-1 for the channel.

B. As required by Required B.1 -------------NOTES-------------

Action A.1 and referenced in 1. Only applicable in Table 3.3.5.1-1. MODES 1, 2, and 3.

2. Only applicable for Functions 1.a, 1.b, 2.a, and 2.b.

Declare supported 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from feature(s) inoperable when discovery of loss of its redundant feature ECCS initiation capability initiation capability is for feature(s) in both inoperable. divisions AND (continued)

Brunswick Unit 2 3.3-35 Amendment No. 233

ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 --------------NOTE--------------

Only applicable for Functions 3.a and 3.b.

Declare High Pressure 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from Coolant Injection (HPCI) discovery of loss of System inoperable. HPCI initiation capability AND B.3 Place channel in trip. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. As required by Required C.1 ------------NOTES--------------

Action A.1 and referenced in 1. Only applicable in Table 3.3.5.1-1. MODES 1, 2, and 3.

2. Only applicable for Functions 1.c, 1.d, 2.c, 2.d, and 2.f.

Declare supported 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from feature(s) inoperable when discovery of loss of its redundant feature ECCS initiation capability initiation capability is for feature(s) in both inoperable. divisions AND C.2 Restore channel to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

(continued)

Brunswick Unit 2 3.3-36 Amendment No. 233

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.1 --------------NOTE--------------

Action A.1 and referenced in Only applicable if HPCI Table 3.3.5.1-1 pump suction is not aligned to the suppression pool.

Declare HPCI System 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from inoperable. discovery of loss of HPCI initiation capability AND D.2.1 Place channel in trip. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OR D.2.2 Align the HPCI pump 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> suction to the suppression pool.

(continued)

Brunswick Unit 2 3.3-37 Amendment No. 233

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. As required by Required E.1 Declare Automatic 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from Action A.1 and referenced in Depressurization System discovery of loss of Table 3.3.5.1-1. (ADS) valves inoperable. ADS initiation capability in both trip systems AND E.2 Place channel in trip. 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> from discovery of inoperable channel concurrent with HPCI or reactor core isolation cooling (RCIC) inoperable AND 8 days (continued)

Brunswick Unit 2 3.3-38 Amendment No. 233

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. As required by Required F.1 Declare ADS valves 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from Action A.1 and referenced in inoperable. discovery of loss of Table 3.3.5.1-1. ADS initiation capability in both trip systems AND F.2 Restore channel to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> from OPERABLE status. discovery of inoperable channel concurrent with HPCI or RCIC inoperable AND 8 days G. Required Action and G.1 Declare associated Immediately associated Completion Time supported feature(s) of Condition B, C, D, E, or F inoperable.

not met.

Brunswick Unit 2 3.3-39 Amendment No. 233

ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS

NOTES ----------------------------------------------------------

1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for Function 3.c; and (b) for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for Functions other than 3.c provided the associated Function or the redundant Function maintains ECCS initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.1.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.5.1.3 Calibrate the trip unit. 92 days SR 3.3.5.1.4 Perform CHANNEL CALIBRATION. 24 months SR 3.3.5.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.5.1.6 Perform CHANNEL FUNCTIONAL TEST. 24 months Brunswick Unit 2 3.3-40 Amendment No. 233

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 4)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Core Spray System
a. Reactor Vessel Water LevelLow 1,2,3, 4 B SR 3.3.5.1.1 13 inches Level 3 4(a), 5(a) SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
b. Drywell PressureHigh 1,2,3 4 B SR 3.3.5.1.1 1.8 psig SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
c. Reactor Steam Dome PressureLow 1,2,3 4 C SR 3.3.5.1.1 402 psig SR 3.3.5.1.2 and SR 3.3.5.1.3 425 psig SR 3.3.5.1.4 SR 3.3.5.1.5 4(a), 5(a) 4 B SR 3.3.5.1.1 402 psig SR 3.3.5.1.2 and SR 3.3.5.1.3 425 psig SR 3.3.5.1.4 SR 3.3.5.1.5
d. Core Spray Pump StartTime Delay 1,2,3, 2 C SR 3.3.5.1.4 14 seconds Relay 4(a), 5(a) 1 per pump SR 3.3.5.1.5 and SR 3.3.5.1.6 16 seconds
2. Low Pressure Coolant Injection (LPCI)

System

a. Reactor Vessel Water LevelLow 1,2,3, 4 B SR 3.3.5.1.1 13 inches Level 3 4(a), 5(a) SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
b. Drywell PressureHigh 1,2,3 4 B SR 3.3.5.1.1 1.8 psig SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5 (continued)

(a) When associated subsystem(s) are required to be OPERABLE.

Brunswick Unit 2 3.3-41 Amendment No. 233

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 4)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

2. LPCI System (continued)
c. Reactor Steam Dome PressureLow 1,2,3 4 C SR 3.3.5.1.1 402 psig SR 3.3.5.1.2 and SR 3.3.5.1.3 425 psig SR 3.3.5.1.4 SR 3.3.5.1.5 4(a), 5(a) 4 B SR 3.3.5.1.1 402 psig SR 3.3.5.1.2 and SR 3.3.5.1.3 425 psig SR 3.3.5.1.4 SR 3.3.5.1.5
d. Reactor Steam Dome PressureLow 1(b),2(b), 4 C SR 3.3.5.1.1 302 psig (Recirculation Pump Discharge Valve 3(b) SR 3.3.5.1.2 Permissive) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
e. Reactor Vessel Shroud Level 1,2,3 2 B SR 3.3.5.1.1 -50 inches SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
f. RHR Pump StartTime Delay Relay 1,2,3, 4 C SR 3.3.5.1.4 9 seconds 4(a), 5(a) 1 per pump SR 3.3.5.1.5 and SR 3.3.5.1.6 11 seconds
3. High Pressure Coolant Injection (HPCI)

System

a. Reactor Vessel Water LevelLow 1, 4 B SR 3.3.5.1.1 101 inches Level 2 2(c), 3(c) SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
b. Drywell PressureHigh 1, 4 B SR 3.3.5.1.1 1.8 psig (c) (c) 2 ,3 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5 (continued)

(a) When associated subsystem(s) are required to be OPERABLE.

(b) With associated recirculation pump discharge valve or recirculation pump discharge bypass valve open.

(c) With reactor steam dome pressure > 150 psig.

Brunswick Unit 2 3.3-42 Amendment No. 233

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 4)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

3. HPCI System (continued)
c. Reactor Vessel Water LevelHigh 1, 2 C SR 3.3.5.1.1 207 inches 2(c), 3(c) SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
d. Condensate Storage Tank LevelLow 1, 2 D SR 3.3.5.1.2 23 feet 4 inches (c) (c) 2 ,3 SR 3.3.5.1.4 SR 3.3.5.1.5
e. Suppression Chamber Water Level 1, 2 D SR 3.3.5.1.2 -2 feet High 2(c), 3(c) SR 3.3.5.1.4 SR 3.3.5.1.5
4. Automatic Depressurization System (ADS)

Trip System A

a. Reactor Vessel Water LevelLow 1, 2 E SR 3.3.5.1.1 13 inches (c) (c)

Level 3 2 ,3 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

b. ADS Timer 1, 1 F SR 3.3.5.1.4 108 seconds (c) (c) 2 ,3 SR 3.3.5.1.5 SR 3.3.5.1.6
c. Reactor Vessel Water LevelLow 1, 1 E SR 3.3.5.1.1 153 inches (c) (c)

Level 1 2 ,3 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5

d. Core Spray Pump Discharge 1, 2 F SR 3.3.5.1.2 102 psig PressureHigh 2(c), 3(c) SR 3.3.5.1.4 and SR 3.3.5.1.5 130 psig
e. RHR (LPCI Mode) Pump Discharge 1, 4 F SR 3.3.5.1.2 102 psig PressureHigh 2(c), 3(c) 2 per pump SR 3.3.5.1.4 and SR 3.3.5.1.5 130 psig (continued)

(c) With reactor steam dome pressure > 150 psig.

Brunswick Unit 2 3.3-43 Amendment No. 233

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 4 of 4)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

5. ADS Trip System B
a. Reactor Vessel Water LevelLow 1, 2 E SR 3.3.5.1.1 13 inches Level 3 2(c), 3(c) SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
b. ADS Timer 1, 1 F SR 3.3.5.1.4 108 seconds (c) (c) 2 ,3 SR 3.3.5.1.5 SR 3.3.5.1.6
c. Reactor Vessel Water LevelLow 1, 1 E SR 3.3.5.1.1 153 inches Level 1 2(c), 3(c) SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
d. Core Spray Pump Discharge 1, 2 F SR 3.3.5.1.2 102 psig PressureHigh 2(c), 3(c) SR 3.3.5.1.4 and SR 3.3.5.1.5 130 psig
e. RHR (LPCI Mode) Pump Discharge 1, 4 F SR 3.3.5.1.2 102 psig (c) (c)

PressureHigh 2 ,3 2 per pump SR 3.3.5.1.4 and SR 3.3.5.1.5 130 psig (c) With reactor steam dome pressure > 150 psig.

Brunswick Unit 2 3.3-44 Amendment No. 233

PCIVs 3.6.1.3 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment Isolation Instrumentation."

ACTIONS

NOTES ----------------------------------------------------------

1. Penetration flow paths may be unisolated intermittently under administrative controls.
2. Separate Condition entry is allowed for each penetration flow path.
3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.
4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment,"

when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.

CONDITION REQUIRED ACTION COMPLETION TIME A. --------------NOTE------------- A.1 Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Only applicable to penetration flow path by penetration flow paths with use of at least one closed two PCIVs. and de-activated automatic

valve, closed manual valve, blind flange, or check valve One or more penetration with flow through the valve flow paths with one PCIV secured.

inoperable except for MSIV leakage not within limit. AND (continued)

Brunswick Unit 2 3.6-7 Amendment No. 233

PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 --------------NOTE--------------

Isolation devices in high radiation areas may be verified by use of administrative means.

Verify the affected Once per 31 days penetration flow path is for isolation devices isolated. outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment (continued)

Brunswick Unit 2 3.6-8 Amendment No. 233

PCIVs 3.6.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. --------------NOTE-------------- B.1 Isolate the affected 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Only applicable to penetration flow path by penetration flow paths with use of at least one closed two PCIVs. and de-activated automatic

valve, closed manual valve, or blind flange.

One or more penetration flow paths with two PCIVs inoperable except for MSIV leakage not within limit.

C. --------------NOTE-------------- C.1 Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> except for Only applicable to penetration flow path by excess flow check penetration flow paths with use of at least one closed valves (EFCVs) only one PCIV. and de-activated automatic

valve, closed manual valve, AND or blind flange.

One or more penetration 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for EFCVs flow paths with one PCIV AND inoperable.

C.2 --------------NOTE--------------

Isolation devices in high radiation areas may be verified by use of administrative means.

Verify the affected Once per 31 days penetration flow path is isolated.

(continued)

Brunswick Unit 2 3.6-9 Amendment No. 233

PCIVs 3.6.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One or more penetration D.1 Restore leakage rate to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> flow paths with one or more within limit.

MSIVs not within MSIV leakage rate limits.

E. Required Action and E.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, B, C, or D AND not met in MODE 1, 2, or 3.

E.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 /> F. Required Action and F.1 Initiate action to suspend Immediately associated Completion Time operations with a potential of Condition A, B, C, or D for draining the reactor not met for PCIV(s) required vessel (OPDRVs).

to be OPERABLE during MODE 4 or 5. OR F.2 Initiate action to restore Immediately valve(s) to OPERABLE status.

Brunswick Unit 2 3.6-10 Amendment No. 233

AC SourcesOperating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC SourcesOperating LCO 3.8.1 

 



APPLICABILITY:

ACTIONS

NOTE ----------------------------------------------------------

LCO 3.0.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. ----------------NOTE------------- A.1 Restore Unit 2 offsite circuit 45 days Only applicable when Unit 2 to OPERABLE status.

is in MODE 4 or 5.

One Unit 2 offsite circuit inoperable.

(continued)

Brunswick Unit 1 3.8-1 Amendment No. 233

AC SourcesOperating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. --------------NOTES-------------- B.1 Declare required feature(s) Immediately from

1. Only applicable when with no power available discovery of Unit 2 is in MODE 4 inoperable when the Condition B or 5. redundant required concurrent with feature(s) are inoperable. inoperability of
2. Condition B shall not be redundant required entered in conjunction feature(s) with Condition A.

Two Unit 2 offsite circuits inoperable due to one Unit 2 balance of plant circuit path to the downstream 4.16 kV emergency bus inoperable for planned maintenance.

AND AND DG associated with the B.2 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> affected downstream OPERABLE offsite 4.16 kV emergency bus circuit(s). AND inoperable for planned maintenance. Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND B.3 Restore both Unit 2 offsite 7 days circuits and DG to OPERABLE status. AND 10 days from discovery of failure to meet LCO 3.8.1.a or b (continued)

Brunswick Unit 1 3.8-2 Amendment No. 205

AC SourcesOperating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One offsite circuit inoperable C.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for reasons other than OPERABLE offsite Condition A or B. circuit(s). AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND C.2 Declare required feature(s) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from with no offsite power discovery of no available inoperable when offsite power to one the redundant required 4.16 kV emergency feature(s) are inoperable. bus concurrent with inoperability of redundant required feature(s)

AND C.3 Restore offsite circuit to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

AND 17 days from discovery of failure to meet LCO 3.8.1.a or b (continued)

Brunswick Unit 1 3.8-3 Amendment No. 264

AC SourcesOperating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One DG inoperable for D.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> reasons other than OPERABLE offsite Condition B. circuit(s). AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND D.2 Evaluate availability of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> supplemental diesel generator (SUPP-DG) AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND D.3 Declare required feature 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from (s), supported by the discovery of inoperable DG, inoperable Condition D when the redundant concurrent with required feature (s) are inoperability of inoperable. redundant required feature (s)

AND D.4.1 Determine OPERABLE 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DG(s) are not inoperable due to common cause failure.

OR D.4.2 Perform SR 3.8.1.2 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE DG(s).

AND (continued)

Brunswick Unit 1 3.8-4 Amendment No. 264

AC SourcesOperating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.5 Restore DG to OPERABLE 7 days from status. discovery of unavailability of SUPP-DG AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of Condition D entry 6 days concurrent with unavailability of SUPP-DG AND 14 days AND 17 days from discovery of failure to meet LCO 3.8.1.a or b E. Two or more offsite circuits E.1 Declare required feature(s) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from inoperable for reasons other inoperable when the discovery of than Condition B. redundant required Condition E feature(s) are inoperable. concurrent with inoperability of redundant required feature(s)

AND E.2 Restore all but one offsite 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> circuit to OPERABLE status.

(continued)

Brunswick Unit 1 3.8-5 Amendment No. 264

AC SourcesOperating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. One offsite circuit inoperable ----------------------NOTE-------------------

for reasons other than Enter applicable Conditions and Condition B. Required Actions of LCO 3.8.7, "Distribution SystemsOperating,"

AND when Condition F is entered with no AC power source to any 4.16 kV One DG inoperable for emergency bus.

reasons other than -------------------------------------------------

Condition B.

F.1 Restore offsite circuit to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OPERABLE status.

OR F.2 Restore DG to OPERABLE 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> status.

G. Two or more DGs G.1 Restore all but one DG to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> inoperable. OPERABLE status.

H. Required Action and H.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, B, C, D, E, F AND or G not met.

H.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 /> I. One or more offsite circuits I.1 Enter LCO 3.0.3. Immediately and two or more DGs inoperable.

OR Two or more offsite circuits and one DG inoperable for reasons other than Condition B.

Brunswick Unit 1 3.8-6 Amendment No. 264

AC SourcesOperating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power 7 days availability for each offsite circuit.

SR 3.8.1.2 -------------------------------NOTES-------------------------------

1. All DG starts may be preceded by an engine prelube period.
2. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR. When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.
3. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG starts from standby conditions and 31 days achieves steady state voltage 3750 V and 4300 V and frequency 58.8 Hz and 61.2 Hz.

(continued)

Brunswick Unit 1 3.8-7 Amendment No. 264

AC SourcesOperating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.3 -------------------------------NOTES-------------------------------

1. DG loadings may include gradual loading.
2. Momentary transients outside the load range do not invalidate this test.
3. This Surveillance shall be conducted on only one DG at a time.
4. This SR shall be preceded by and immediately follow, without shutdown, a successful performance of SR 3.8.1.2 or SR 3.8.1.7.
5. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG is synchronized and loaded and 31 days operates for 60 minutes at a load 2800 kW and 3500 kW.

SR 3.8.1.4 Verify each engine mounted tank contains 150 gal of 31 days fuel oil.

SR 3.8.1.5 Check for and remove accumulated water from each 31 days engine mounted tank.

SR 3.8.1.6 Verify the fuel oil transfer system operates to transfer 31 days fuel oil from the day fuel oil storage tank to the engine mounted tank.

(continued)

Brunswick Unit 1 3.8-8 Amendment No. 205

AC SourcesOperating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.7 -------------------------------NOTES-------------------------------

1. All DG starts may be preceded by an engine prelube period.
2. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG starts from standby condition and 184 days achieves, in 10 seconds, voltage 3750 V and frequency 58.8 Hz, and after steady state conditions are reached, maintains voltage 3750 V and 4300 V and frequency 58.8 Hz and 61.2 Hz.

(continued)

Brunswick Unit 1 3.8-9 Amendment No. 205

AC SourcesOperating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.8 -------------------------------NOTES-------------------------------

1. SR 3.8.1.8.a shall not be performed in MODE 1 or 2 for the Unit 1 offsite circuits. However, credit may be taken for unplanned events that satisfy this SR.
2. SR 3.8.1.8.a is not required to be met if the unit power supply is from the preferred offsite circuit.
3. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify: 24 months

a. Automatic transfer capability of the unit power supply from the normal circuit to the preferred offsite circuit; and
b. Manual transfer of the unit power supply from the preferred offsite circuit to the alternate offsite circuit.

(continued)

Brunswick Unit 1 3.8-10 Amendment No. 205

AC SourcesOperating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.9 -------------------------------NOTES-------------------------------

1. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.
2. If performed with the DG synchronized with offsite power, it shall be performed at a power factor 0.9.
3. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG rejects a load greater than or equal to 24 months its associated core spray pump without tripping.

(continued)

Brunswick Unit 1 3.8-11 Amendment No. 205

AC SourcesOperating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.10 -------------------------------NOTE---------------------------------

A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG's automatic trips are bypassed on an 24 months actual or simulated ECCS initiation signal except:

a. Engine overspeed;
b. Generator differential overcurrent;
c. Low lube oil pressure;
d. Reverse power;
e. Loss of field; and
f. Phase overcurrent (voltage restrained).

(continued)

Brunswick Unit 1 3.8-12 Amendment No. 205

AC SourcesOperating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.11 -------------------------------NOTES-------------------------------

1. Momentary transients outside the load and power factor ranges do not invalidate this test.
2. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG operating at a power factor 0.9 24 months operates for 60 minutes loaded to 3500 kW and 3850 kW.

SR 3.8.1.12 -------------------------------NOTE---------------------------------

A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify an actual or simulated ECCS initiation signal is 24 months capable of overriding the test mode feature to return each DG to ready-to-load operation.

(continued)

Brunswick Unit 1 3.8-13 Amendment No. 205

AC SourcesOperating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.13 -------------------------------NOTE---------------------------------

This Surveillance shall not be performed in MODE 1, 2, or 3 for the load sequence relays associated with DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.

Verify interval between each sequenced load block is 24 months within +/- 10% of design interval for each load sequence relay.

(continued)

Brunswick Unit 1 3.8-14 Amendment No. 205

AC SourcesOperating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.14 -------------------------------NOTES-------------------------------

1. All DG starts may be preceded by an engine prelube period.
2. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.

Verify, on actual or simulated loss of offsite power 24 months signal in conjunction with an actual or simulated ECCS initiation signal:

a. De-energization of emergency buses;
b. Load shedding from emergency buses; and
c. DG auto-starts from standby condition and:
1. energizes permanently connected loads in 10.5 seconds,
2. energizes auto-connected emergency loads through load sequence relays,
3. maintains steady state voltage 3750 V and 4300 V,
4. maintains steady state frequency 58.8 Hz and 61.2 Hz, and
5. supplies permanently connected and auto-connected emergency loads for 5 minutes.

Brunswick Unit 1 3.8-15 Amendment No. 205

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ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U. S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:

Date: 12/13/2016 aciIity I Unit: Brunswick Unit 1/2 Region: I El II [] Ill El IV El Reactor Type: W LICE El BW [] GEE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80 percent overall, with 70 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results RO/SRO-Onlyrrotal Examination Values / I Points Applicants Score I I Points Applicants Grade / /

Percent

1. 2010031 Unit Two is operating at rated power when a control rod begins to drift out from position 24.

Which one of the following identifies the first action to be taken by the operator at the controls (OATC)?

A. Initiate a single rod scram.

B. Initiate a manual reactor scram.

C. Select and attempt to arrest the control rod.

D. Select and fully insert the control rod to position 00.

Answer: C K/A:

201003 Control Rod and Drive Mechanism G2.4.49Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.10 / 43.2 /45.6)

RO/SRO Rating: 4.6/4.4 Tier2/Group2 K/A Match: This meets the K/A because the question is testing the operator action required to control a drifting control rod (Chief Examiner agreed that operation of the RMCS for a rod drift would meet this K/A)

Pedigree: New Objective: LOl-CLS-LP-07. Obj. lib Describe the possible cause(s) and required operator actions for the following alarms:

A-5 3-2. Control Rod Drift

Reference:

None Cog Level: Fundamental Explanation: This abnormal positive reactivity addition requires response from the APP before entering the AOP. The APP requires that the operator attempt to arrest the drift at the intended position first, if it cannot be arrested but responds to RMCS to insert to 00, if it does not respond to RMCS to perform a single rod scram. If more than 1 rod drifts then a manual scram is required.

Distractor Analysis:

Choice A: Plausible because if the rod does not move then this is the appropriate action.

Choice B: Plausible because if more than 1 rod is drifting then this would be correct Choice C: Correct Answer, see explanation.

Choice D: Plausible because this is the correct action if the rod is drifting in or the rod continues to drift after attempting to arrest.

SRO Basis: N/A

APP ATh 32 Pa;e 1 of at: oaisi A710 ACTICHS I. RW! withdraw or insert errors possibly causing rod block if reactor power is below the low power setpoint.

2. The reactor power will respond to the drifting rod depending upon the direction of the rod draft and rod worth, and could result in a reactor Scram if the plant is at low power operation -

CArJSE

1. Rod in uneven position due to:
a. Leaking Scram valve.
b. High cooling water pressure.
c. Failure cf directional control valves.
d. Slow to settle due to fuel bundle channel bow.
2. Nalfunction an alarm circuit.

03 SERVATI OHS I. Rod Drift indication on the full core display.

2. RWM error indications and Rod Block if reactor power is below the low power setpoant.
3. A change in neutron monitorang system neter readings as a result of the drafting rod with possible high flux alarms.
4. if drifting rod is selected, the fourrod group display will andicate an odd control rod posation, a blank window, or a changing control rod position in the direction of the drift.

S. High control rod cooling water pressure and/or flow.

C If control rod drifts to the full in position, a green backlight on the full core display wall illuminate wath no position readout on RTGB.

?. Rot OUT BLOCK alarm A-OS (22) and no wathdraw permissive light.

8. Greater than normal settle times causing an odd or noposition to be present when the RHOS timer tanes out.

ACTIONS

1. Determine if the affected control rod (s) as drifting or af the rod(s) has scrammed using full core display, RPIS, and RWN.
2. Select the drifting rod and determine direction of drift.
a. Attempt to arrest the drift and latch rod by performang the following:
1) Apply appropriate insert or withdrawal signals to the rod using PNCS.
2) If RN!! is causing rod blocks, then bypass RN!! if directed by Unit CR5.

2 AP A3S 32 rage 2 of E

3. If he rod c trnues to 2AZFT CUT, then perforn the following:

CAUTIU A control rod c:.llett praton stuck in the withdraw unlatched) pcsiti:.n will allow the rod to drift full out due to its own weiqht when insert pressure iS removed either by the PNCS or by closing Valve CIClOl.

a. Notify Acactor Engineer.
b. !onit.or core rarameters, marc 5tCam rifle ramration fl:nitorS, and offgas actrvify.
c. If r:d responds no an P.MCS insert signal, then fully insert the rod to tosinion CC If r:od fails to laith at position CC, then reapply insert signal to drve the rod fulZ. ln.
e. If rod fails to tespond to P.lC.3, then inrtiate a single control rod cotem.

P.efer to AOPJ3.O.

g. Pefer to Technical Specificatrons 3.3.
4. If od contInues to ZP.I3T N, then perform the following:

a Amply an PY2E insert signal and fully insert rod to p:sinion

CONTROL ROD MALFUNCTION1MISPOSON OAOP-020 Rev. 28 Page 6 0125 3.0 AUTOMATIC ACTIONS

1. Possible rod block or select block from a faded reed switch or a loss of power
2. CRD pumps trip alter a 3 second delay on low suction pressure 4.0 OPERATOR ACTIONS NOTE The following should be considered for establishment as critical parameters during performance of this procedure 0
  • Reactor power
  • Thermal limits 4.1 Immediate Actions
1. Stop any power changes in progress 0 NOTE Detected control rod motion without a withdraw or insert command will cause annunciator A-05 3-2, Rod Dntt. to alanm IF the annunciator alarms AND NO blue scram light(s) are lit on the full core display, the conservative assumption is that rod(s) are drifting 0
2. IF more than one control rod is drifting.

THEN insert a manual scram AND enter 1 EOP-01 -RSP(2EOP-Ot-RSP), Reactor Scram Procedure 0

2.201001 1 Unit One is in an outage with the condensate system under clearance.

An earthquake results in damage to the CST causing level to slowly lower.

Which one of the following completes the statement below with regards to the effect on the CRD system?

The CRD system will (1) when the CST level reaches approximately (2)

A. (1) trip (2) 3 feet B. (1) trip (2) 11 feet C. (1) transfer to the backup supply (2) 3 feet D. (1) transfer to the backup supply (2) 11 feet Answer: B KJA:

201001 Control Rod Drive Hydraulic System K6 Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROD DRIVE HYDRAULIC System: (CFR: 41.7 / 45.7) 02 Condensate storage tanks RO/SRO Rating: 3.0/3.1 Tier 2 / Group 2 K/A Match: This meets the KJA because the student has determine the effect of the loss of the CST on the CRD system.

Pedigree: New Objective: LOl-CLS-LP-008, Obj. 8 Given plant conditions, predict the effect that a loss or malfunction of the following will have on the CRDH System: b. Condensate Storage Tank

Reference:

None Cog Level: High Explanation: Under normal system operations the CRD system suction is from the condensate system.

The alternate supply is from the CST, which will transfer automatically. With the condensate system under clearance these valves would be isolated. The standpipe for the CRD suction is at 1 1 feet. The auto transfer for the suction for ECCS is at 3 feet.

Distractor Analysis:

Choice A: Plausible because the pumps will lose NPSH and trip but the suction is at 11 feet not 3 feet.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because an auto transfer to the CST would occur but in this case an auto transfer to the condensate system is not possible. 3 feet is the suction height for the ECCS system.

Choice D: Plausible because an auto transfer to the CST would occur but in this case an auto transfer to the condensate system is not possible. The second part is correct.

SRO Basis: N/A CONTROL ROD DRIVE HYDRAULIC SYSTEM 1 OP-08 OPERA11NG PROCEDURE Rev 96 Page 6 of 377 3.0 PRECAUTIONS AND LIMITATIONS

1. This procedure is Reactivity Management related per AD-OP-ALL-0203, Reactivity Management. Those portions of this procedure that move control rods in MODES 1 or 2 are considered a Direct Reactivity manipulation and Reactivity Evolution Category R2 (Reactivity Manipulation, R2) I]
2. CST level is maintained greater than 11 feet to prevent CRD pumps from losing suction
3. 202002 1 Unit One is at rated power.

Which one of the following identifies the impact of inadvertently closing the IA Reactor Recirculation Pump 1-B32-FO3IA, Pump A Disch Vlv?

The 1A Reactor Recirculation pump speed will lower to approximately:

A. 20%

B. 34%

C. 45.4%

D. 48%

Answer: B K/A:

202002 Recirculation Flow Control System K6 Knowledge of the effect that a loss or malfunction of the following will have on the RECIRCULATION FLOW CONTROL SYSTEM: (CFR: 41.7/45.7) 03 Recirculation system RO/SRO Rating: 2.8/2.8 Tier 2 I Group 2 K/A Match: This meets the K/A because the student has to determine the effect of closing the discharge valve (which causes a loss of recirc) will have on the recirc flow control system.

Pedigree: new Objective: LOI-CLS-LP-002. 1, Obj. 17 Explain the operation of the following VFD limiters and controls: a. Limiter #1 b. Limiter #2

Reference:

None Cog Level: Fundamental Explanation: Closing of the discharge valve will cause the pump to runback to limiter #1(34%).

Distractor Analysis:

Choice A: Plausible because this is the minimum speed setting.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because this is limiter #2 setting for Unit One Choice D: Plausible because this is limiter #2 setting for Unit Two SRO Basis: N/A

4. RFCS VFD Runback #1 Logic (Figure 02-1 8D and 02-1 BE)

The logic for VFD A Runback #1 is shovrn on Figure 02-180; the logic for VFD B Runback #1 shown on Figure Q2-18E is functionally identical to that for VFD A. The initiating conditions for Runback #1 are:

  • Recirculation Pump A Discharge Valve B32-FO31A Limit Switch LS-2 opens (equivalent to Discharge Valve Not Full Open);
  • Total Feedwater flow as sensed by DECS is less than 16.4% for 15 seconds or mote.

Unit 1 Specific VFD Parameters Parameter Value Function 1170 98.9%(i661.5 rpm) VFD OverSpeed Trip (Over Speed Alarm at 93.95% or 1578.4 rpm) 2080 92.5% (1554.0 rpm) Maximum running motor speed (based upon achieving 104.5% Core flow) 2120 45.4% (762.7 rpm) Runback #2 Active Maximum Motor Speed 4250 50.8% (853.0 rpm) Manual Runback Motor Speed Low Limit Unit 2 Specific VFD Parameters Parameter Value Function 1170 103.7% VFD Over Speed Trip (1742.2 rpm) (Over Speed Alarm at 98.5% or 1655.1 rpm) 2080 97.9% Maximum running motor speed (1644.7 rpm) (based upon achieving 104.5% Core Flow) 2120 48% (806.4 rpm) Runback #2 Active Maximum Motor Speed 4250 53.6% (900.5 rpm) Manual Runbadc Motor Speed Low Limit VFD Parameters Common to Both Units Parameter Value Function 2090 20% (336 rpm) Minimum Running Motor Speed 2100 34% (571.2 rpm) Runback #1 Active Maximum Motor Speed

4. 203000 1 A line break has occurred in the Unit Two drywell with the following sequence of events:

1155 Drywell pressure rises above 1.7 psig 1202 RPV pressure drops below 410 psig 1203 RPV level drops to LL3 Which one of the following completes the statement below?

The earliest time that the operator can throttle the 2-El l-F048A, Loop 2A RHR Heat Exchanger Bypass Valve is at:

A. 1205.

B. 1206.

C. 1207.

D. 1208.

Answer: A K/A:

203000 RHR I LPCI: Injection Mode A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /45.5 to 45.8) 04 Heat exchanger cooling flow RO/SRO Rating: 3.6/3.6 Tier2lGroup 1 K/A Match: This meets the K/A because the student has to determine when the HX cooling flow can be operated.

Pedigree: Bank Objective: LOl-CLS-LP-017, Obj. 09 Given an RHR pump or valve, list the interlocks, permissives and/or automatic actions associated with the RHR pump or valve, including setpoints.

Reference:

None Cog Level: High Explanation: The heat exchanger bypass valve has a 3 minute timer that starts on a LOCA signal.

Drywell pressure greater than 1 .7# and reactor pressure is less than 41 0# is the first LOCA signal. The injection valve has a 5 minute interlock initiated by the same conditions.

Another LOCA signal is introduced when reactor water level less than LL3 which provides the plausibility of the distractors.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because this is 3 minutes from the LL3 LOCA signal.

Choice C: Plausible because this is 5 minutes from the low pressure setpoint which is time limit interlock for the injection valve.

Choice D: Plausible because this is 5 minutes from the LL3 LOCA signal.

SRO Basis: N/A Alter an hiltiation signal is received the foLlowing actions will occur:

  • all tour RHR pumps wilt start 10 seconds alter power is available to the E-buses.
  • Recirculation pumps are tripped via LL#2
  • All valves not needed for LPCI injection automatically isolate and are interlocked shut as previously described.
  • Heat exchanger bypass valve FO48NB opens and cannot be throttled for 3 minutes alter an initiation signal is received. This ensures a discharge path for the RHR pumps.

a Permissives sent to ADS as RHR pump pressure is sensed

> 100 psig. Both pumps in either loop are required to satisfy the ADS permissive, or one core spray loop.

  • Minimum flow valve opens if injection flow in loop is < 1000 gpm decreasing after a 10 sec time delay. It automatically shuts as injection valves open and injection flow raises to> 3000 gpm increasing.
  • Reactor pressure decreases through the break and/or with actuation of ADS.
  • As reactor pressure decreases to 410 psig, the LPCI injection valves EDt 5AfB) auto open. The outboard injection valve FO1ZA(B) can be throttled 5 minutes after the RPV pressure is below 410 psig.
  • As pressure reaches 310 psig, recirculation pump discharge and discharge bypass valves shut and are interlocked shut in the attempt to re-flood the core.
  • As pressure reaches 200 psig, the RHR system injects into both recirculation system loops by lifting the check valves and overcoming reactor pressure.
5. 205000 1 RHR Loop 2A is operating in the Shutdown Cooling mode of operation with the following parameters:

RHRSW Pump 2A Operating RHRSW Flow 4000 gpm RHR Pump 2A Operating RHR Loop A Flow 6000 gpm Which one of the following completes the statement below?

The required operator action to lower the cooldown rate lAW 20P-l 7, Residual Heat Removal System Operating Procedure, is to throttle closed:

A. 2-El 1-FOO3A, HX 2A Outlet Vlv.

B. 2-Ell-FOI7A, Outboard Injection Vlv.

C. 2-E11-F048A, HX 2A Bypass Vlv.

D. 2-El l-PDV-F068A, HX 2A SW Disch VIv.

Answer: D K/A:

205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

K5 Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): (CFR: 41.5/45.3) 03 Heat removal mechanisms RO/SRO Rating: 2.8/3.1 Tier 2 I Group 1 K/A Match: This meets the K/A because the student has to know which valve would need to be operated to control the heat removal for SDC.

Pedigree: New Objective: LOI-CLS-LP-01 7, Obj. 15 Describe how the reactor cool down rate is controlled when the RHR system is in the Shutdown Cooling mode

Reference:

None Cog Level: High Explanation: The procedure allows throttling closed the F003, FOl 7 or F068 or throttling open the F048.

Throttling open the F048 will bypass some of the RHR flow around the heat exchanger thereby lowering cooldown rate. RHR flow is limited to greater than 6000 gpm, so closing the F017 or F003 is not an option.

Distractor Analysis:

Choice A: Plausible because if the valve was throttled closed this would lower cooldown, but flow must be greater than 6000 gpm.,

Choice B: Plausible because if the valve was throttled closed this would lower cooldown, but flow must be greater than 6000 gpm..

Choice C: Plausible because if the valve was throttled open this would be correct.

Choice D: Correct Answer, see explanation SRO Basis: N/A

5. jf less cooling is desired.

THEN perform the following, in order of preference CAUTION When Eli -FOO3A(B) [Hx 2A(2B) Outlet VIvj is CLOSED, RHR Heat Exchanger 2A(2B) inlet temperature, located on E41-TR-R605 (HPCI Turb Brg Oil Temp Recorder), Point 1(2), is NOT a valid indication of reactor coolant temperature U

a. Throttle close El 1-FOQ3AfB) [Hx 2A(28) Outlet VlvJ not lower than 6000 gpm. as necessary
b. Reduce RHRSW loop flow by performing the following as necessary:
  • Throttle closed Eii-PDV-FO68AfB) [Hx 2A(2B) SW Disch VlvJ to reduce RHRSW flow rate
c. Bypass a portion of RHR toop flow around the HX as follows:

(1) Station an operator at El 1-EO48AfB) [Hx 2A(2B)

Bypass Vlv] to monitor for severe vibration/cavitation during throttling evolutions (2) Adjust E11-FDD3A(B) [Hx 2A(2B) Outlet Vlv] as required to achieve 6500 gpm RHR loop flow (3) Throttle close El l-FOJ7AfB) (Outboard Injection Vlv) as required to achieve 6000 gpm RHR loop flow (4) Throttle open El 1-F048A(B) [Hx 2A(2B) Bypass VlvJ as necessary

6. 206000 1 A Group 1 isolation has occurred on Unit One.

HPCI has been placed in the pressure control mode of operation lAW IOP-19, High Pressure Coolant Injection System Operating Procedure.

HPCI flow controller, E41-FIC-R600, is in manual with the output at midscale.

Which one of the following completes the statement below?

lithe 1-E41-F008, Bypass To CST Valve, is throttled (1) too far, this may result in HPCI (2)

A. (1) open (2) tripping on overspeed B. (1) open (2) operation below 2100 rpm C. (1) closed (2) tripping on overspeed D. (1) closed (2) operation below 2100 rpm Answer: B K/A:

206000 High Pressure Coolant Injection System Ki Knowledge of the physical connections and/or cause-effect relationships between HIGH PRESSURE COOLANT INJECTION SYSTEM and the following:

(CFR: 41.2 to 41.9 /45.7 to 45.8) 10 Condensate storage and transfer system RO/SRO Rating: 2.8/3.1 Tier 2 I Group 1 K/A Match: This meets the K/A because this is testing the cause-effect relationship between HPCI and the flowpath to the CST.

Pedigree: Bank Objective: LOl-CLS-LP-019. Obj. 8 Describe the methods available for controlling RPV pressure and/or RPV cooldown when operating the HPCI System in the Pressure Control mode. (LOCT)

Reference:

None Cog Level: high Explanation: Opening F008 will increase flow, causing turbine speed control to lower turbine speed to maintain desired flow. Opening valve too far can result in RPM below 2100 (OP-i 9, Section 8.2). Closing F008 will cause turbine speed to increase, but the governor limits turbine speed to a maximum value (4100 RPM) below the overspeed trip.

Distractor Analysis:

Choice A: Plausible because opened is correct and an overspeed condition may be thought correct if the flowpath is considered incorrectly.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because throttling closed will increase the speed of the turbine.

Choice D: Plausible because may be thought correct if the flowpath is considered incorrectly.

SRO Basis: N/A From OP-19:

CAUTION Throttling E41-FOO& open may cause turbine speed reduction to less than 2100 rpm, if opened too far.

From the SD:

Operation of the HPCI Turbine betow the minimum rated speed of 2100 rpm may result in a failure of the auxiliary oil pump from repeated startup cycles. A loss of the auxiliary oil pump Mlt prevent starting of the HPCI Turbine.

7. 209001 1 Unit Two is operating at rated power.

Due to a circuit malfunction an inadvertent LOCA initiation occurs in the Div II Core Spray logic causing A-03 (2-6), CORE SPRAY SYSTEM II ACTUATED, to alarm.

Which one of the following completes both statements below?

Core Spray Pump(s) (1) will start.

(2) will start.

A. (1) 2BONLY (2) All DGs B. (1) 2B ONLY (2) DG2 and DG4 ONLY C. (1) 2A and 2B (2) All DGs

0. (1) 2A and 2B (2) DG2 and DG4 ONLY Answer: A K/A:

20900 1 Low Pressure Core Spray System K3 Knowledge of the effect that a loss or malfunction of the LOW PRESSURE CORE SPRAY SYSTEM will have on following: (CFR: 41.7 /45.4) 03 Emergency generators RO/SRO Rating: 2.9/3.0 Tier 2 I Group 1 K/A Match: This meets the K/A because it is testing the knowledge of a malfunction of the CS logic has on the EDG.

Pedigree: New Objective: LOl-CLS-LP-01 8, Obj. 14 List three systems, other than the Core Spray System, which are initiated or isolated by the Core Spray System logic.

Reference:

None Cog Level: High Explanation: For CS the logic will only start that divisions pump (RHR would start the other divisions pump) for the CS logic to the DGs either divisions signal will start all DGs.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because the first part is correct and since it is divisional for the pump starts the student may think that it would start only the Div II DGs. There are signals that would start divisional DGs.

Choice C: Plausible because the student may think the CS logic is similar to the RHR logic for pump starts and the second part is correct.

Choice D: Plausible because the student may think the CS logic is similar to the RHR logic for pump starts and since it is a Div Illogic the student may think that it would start only the Div II DGs. There are signals that would start divisional DGs.

SRO Basis: N/A Unit 2 APP ACS D Page 1 of S cca sy :1 ACTUATED AUTO ACTIONS

1. if H bus was not deenergired, Core Spray Pump 23 starts 25 seconds after receipt of initiation signal
2. If H bus was deenergired, Core Spray pump 23 starts 25 seconds after diesel generator ties onto H bus
3. If open, Full Flow Test Byp Vlv, E2lF3ISB, closes
4. If closed, Outboard injection Vlv, E2iF2043, opens
5. When reactor pressure drops to 410 psig, Inboard Injection Tic, E22FDDsa, opens E. When loop flow is greater than 15CC gpm, Mtn Flow Bypass Vlv, E22FJ3lB, closes
7. Dlv ii Nonintrpt P2A, PJTASV5261, and Div 2 NcnIntrpt PITA, P.LASV52E2, close
8. Div ii Backup N2 P.ack Isol Vlv, fliASV5431, and Div i Backup ND Rac:c Isol Vlv, RiTASV542, open S. Fans for Drywell coolers B and C trip ID. All diesel generators start
11. Nuclear Service Water To Vital Header Valve, SWVll?, opens
12. RBCCW HZ Service Water Inlet Valve, SWVlOS, closes CAUSE
1. Reactor low level three 445 inches)
2. High drywell pressure (1.7 psig) in conjunction with lcw reactor pressure (413 psig)
3. circuit malfunction
8. 2110001 Which one of the following completes the statement below concerning Core Spray Line Break Detection differential pressure instrument?

The (1) leg of this DP instrument senses (2) core plate pressure via the SLC/Core Differential Pressure penetration.

A. (1) variable (2) below B. (1) variable (2) above C. (1) reference (2) below D. (1) reference (2) above Answer: D K/A:

211000 Standby Liquid Control System Ki Knowledge of the physical connections and/or cause-effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) 01 Core spray line break detection RO/SRO Rating: 3.0/3.3 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the physical connection of SBLC and CS line break detection.

Pedigree: Last used on 10-1 NRC Exam Objective: CLS-LP-18, Obj. 10 Explain the principle of operation of the CS Line Break Detection Instrumentation

Reference:

None Cog Level: fundamental Explanation: This system is comprised of a differential pressure detector which provides Control Room annunciation on detected high DP. The high pressure reference leg of this instrument is exposed to above core plate pressure via the SLC/Core Differential Pressure penetration.

The low pressure of this instrument is normally exposed to above core pressure via the Core Spray injection line. This results in the instrument normally measuring core DP (not including core plate DP).

Distractor Analysis:

Choice A: Plausible because the examinee may confuse the reference and variable legs and SLC does discharge below the core plate Choice B: Plausible because the examinee may confuse the reference and variable legs Choice C: Plausible because it is the reference leg and SLC does discharge below the core plate.

Choice D: Correct Answer, see explanation SRO Basis: N/A

This system is comprised of a differential pressure detector which provides Control Room annunciation on detected high P. The hgh pressure reference leg of this instrument is exposed to above core plate pressure via the SLCIC0re Differential Pressure penetration. The low pressure of this instrument is normally exposed to above core pressure via the Core Spray injection line. This results in the instrument normaity measuring core P (not including core plate aP).

A break in the Core Spray injection line between the reactor vessel penetration and the core shroud would expose the low pressure side of the instrument to the lower pressure of the region outside the shroud.

This would be sensed as an increased differential pressure and Control Room annunciator would alert the Operator. Although other indtcations would be available, this atami would also indicate a break in the line between the E21-FOQ6GfA) check valve and the reactor vessel penetration.

The Core Spray pipe break detection instruments are located on the Reactor Building 20 elevation.

SD-18 Rev6 I Page29of53

9. 2120001 Which one of the following completes both statements below?

The normal power supply to RPS MG Set 2B is from 480V MCC (1)

The normal alternate power supply to RPS B is from 480V Bus (2)

A. (1) 2CA (2) E7 B. (1) 2CA (2) E8 C. (1) 2CB (2) E7 D. (1) 2CB (2) E8 Answer: C K/A:

212000 Reactor Protection System K2 Knowledge of electrical power supplies to the following: (CFR: 41 .7) 01 RPS motor-generator sets RO/SRO Rating: 3.2/3.3 Tier2/Group 1 K/A Match: This meets the K/A because it is testing the power supply to the RPS MG Set Pedigree: New Objective: CLS-LP-03, Obj 18b State the power supplies for the following: RPS MG Set B

Reference:

None Cog Level: Fundamental Explanation: Power for the Unit 2B Motor Generator Sets is tapped off two phases of the normal 480 VAC MCC 2CB power supply for the motor through a stepdown transformer (480V to 1 20V) from E8 (the 2A MG Set is powered from 2CA). Normal alternate power to the RPS Bus is provided from E7 with Alternate alternate power to the RPS Bus provided from E8. In the event that either RPS M-G Set fails to operate, the alternate power sources must be manually selected.

Distractor Analysis:

Choice A: Plausible because 2CA supplies RPS MG Set A and E7 is the normal alternate power supply.

Choice B: Plausible because 2CA supplies RPS MG Set A and E8 is the alternate alternate power supply.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because 2CB is the RPS MG Set B power supply and E8 is the alternate alternate power supply.

SRO Basis: N/A I .?I CCkT5 -TI T E

10. 2150021 Which one of the following identifies the LPRM detector level that provides input to the Rod Block Monitor system for indication ONLY, and is NOT used for the purpose of generating rod blocks?

A. LevelA B. Level B C. LeveIC D. Level D Answer: A KJA:

215002 Rod Block Monitor System Ki Knowledge of the physical connections and/or cause-effect relationships between ROD BLOCK MONITOR SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) 02 LPRM RO/SRO Rating: 3.2/3.1 Tier 2 / Group 2 K/A Match: This meets the K/A because it is testing the connection between RBM and LPRM5 Pedigree: Bank Objective: LOl-CLS-LP-09.6, Obj 5a List the PRNMS system signals/conditions that will cause the following actions: APRM / RBM Rod Blocks

Reference:

None Cog Level: Fundamental Explanation: The level A inputs are sent to RBM-A for processing/output to the LPRM Display Meters on the 4-Rod Display. Level A is for indication only RBM-A Receives all four level C inputs lower left and upper right level B inputs upper left and lower right level D inputs RBM-B Receives all four level C inputs upper left and lower right level B inputs lower left and upper right level D inputs

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because LPRMs have a B level that input to the RBMs Rod Blocks.

Choice C: Plausible because LPRMs have a C level that input to the RBMs Rod Blocks.

Choice D: Plausible because LPRM5 have a D level that input to the RBMs Rod Blocks.

SRO Basis: N/A The A level LPRM detectors are not used for RBM input processing, while both RBM channels use art C level detectors.

This gives an accurate representation of actual power around the control rod. The B and D detectors are distributed evenly between the two REM channels. An example of LPRf input to a both REM channels vith a four-string rod selected is two B level LPRMs, four C level LPRMs, and two D level LPRMS for each channel.

The REM circuitry undergoes a nulling and filtering sequence when a rod is selected and therefore a delay of at least 2.5 seconds must be allowed between selection and rod movement. A Rod Inhibit signal is SD-09.6 Rev. 12 Page 25 of 95

II. 2150031 Unit One is performing a startup with the reactor just declared critical.

While ranging IRM G from range 1, the IRM will not change ranges and remains on Range 1.

Which one of the following completes both statements below?

When IRM G indication first exceeds (1) on the 125 scale, annunciator A-05, 2-4, IRM UPSCALE, will alarm.

The action required lAW A-05, 2-4, IRM UPSCALE, is to (2)

A. (1) 70 (2) place the joystick on P603 for the IRM G to Bypass B. (1) 70 (2) withdraw the IRM G detector to maintain reading on scale C. (1) 117 (2) place the joystick on P603 for the IRM G to Bypass D. (1) 117 (2) withdraw the IRM G detector to maintain reading on scale Answer: A K/A:

215003 Intermediate Range Monitor System A2 Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR fIRM)

SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) 06 Faulty range switch RO/SRO Rating: 3.0/3.2 Tier2/Group 1 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL.

This meets the K/A because it is testing what will happen with a faulty range switch and the action required.

Pedigree: New Objective: LOI-CLS-LP-009.1, Obj. 3a List the SRM/IRM system signals/conditions that will cause the following actions and the conditions under which each is bypassed: Rod Blocks (LOCT)

LOI-CLS-LP-009.1, Obj. 14a Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event: SRM/IRM Upscale alarm (LOCT)

Reference:

None Cog Level: High

Explanation: With the reactor critical the indication will continue to rise. The Upscale alarm will come in at 70 on the 0-125 scale. The Upscale Hi/Inop alarm comes in at 117 on the 0-125 scale. lAW with the APP the action to take is to bypass the IRM. In the case of the SRMs an action to take could be to withdraw the SRM to maintain on scale readings.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because the first part is correct and the second part could be correct if this was an SRM.

Choice C: Plausible because 117 is an alarm setpoint for the IRMs and the second part is correct.

Choice D: Plausible because 117 is an alarm setpoint for the IRMs and the second part could be correct if this was an SRM.

SRO Basis: N/A

tnjt 1 APP ACE 24 Page 1 of 2 1PM UEALE A7t0 AICTS

1. Rod withdrawal bl:rs byoassed when Pector Mode Switch n RUN)

CAUSE

1. 1PM chonel indicates g ater tSar, or equal to 73 on 315 soal
2. Improper ranging of 1PM channels dliring rea:cor startup or shutdown
3. 1PM detector failure
4. turing refuel outages. 1PM spiking due to noise generation from worc tivities in drywell. such as welding
5. Cirruit malfunctior.

03 SE RYATIONS

1. 1PM channel indicating greater than or equal to 7C on 0125 scale
2. 1PM channel upscale (UPSC ALAPM) anher indtoating light on 3 -

Q!J BIOCE A3S 2-2 a1ars 4 Rod withdrawal permissive indicating light off ACTIONS

1. If in progress, stop withdrawal of control rods.
2. Monitor 1PM indicattons to determine afferted channels).

CA.UTION switches should be remositioned carefully in order to prevent e

3. Reposition affected 1PM range switch to ner higher range.
4. If a sudden rise in indicated reartor power occurred on more than one 1PM channel, verify correct rod withdrawal sequence is Seing used and insert inseauenoe control rods as necessary to turn power rise.

S - If 3PM detector failure or circuit malfunrticn is suspected, perform the following:

a. Refer to Technical Specification 3.3.1.1 and 2PM 3.3 for 1PM channel operability requirements.
5. Notify Unit CPS
c. 3ypass affected channel using 1PM bypass switch.
d. Ensure a WR is prepared.

JAPP-A-05 Rev. 75 Page 24 of 94

12. 2150041 Which one of the following identifies the criteria for when SRM detectors can first begin to be withdrawn from the core lAW OGP-02, Approach To Criticality And Pressurization Of The Reactor?

A. When all IRMs are above range 3.

B. When SRM counts reach 2 x io counts.

C. When RTRCT PERMIT light is illuminated.

D. When SRM/IRM overlap has been established.

Answer: D K/A:

215004 Source Range Monitor System K5 Knowledge of the operational implications of the following concepts as they apply to SOURCE RANGE MONITOR (SRM) SYSTEM: (CFR: 41.5 / 45.3) 03 Changing detector position RO/SRO Rating: 2.8/2.8 Tier 2 / Group I K/A Match: This meets the K/A because it is testing when SRMs can change detector positions.

Pedigree: New Objective: LOI-CLS-LP-009-A, Obj 3b List the SRM/IRM system signals/conditions that will cause the following actions and the conditions under which each is bypassed: Retract Permissive (SRM) only (LOCT)

Reference:

None Cog Level: Fundamental Explanation: When SRM/IRM overlap has been established then SRM can be withdrawn to maintain an indicated SRM count rate between 100 cps and 200,000 cps.

Distractor Analysis:

Choice A: Plausible because this is the logic setpoint at which the SRM can be fully withdrawn Choice B: Plausible because this is the point at which the SRM must be fully withdrawn.

Choice C: Plausible because this is an indication that is used during the withdrawal of the SRMs Choice D: Correct Answer, see explanation.

SRO Basis: N/A

NOTE

  • SRM/IRM overlap is required to be demonstrated for all operable IRM channels prior to withdrawing SRMs from the fully inserted position. SRMIIRM overlap exists when IRM channels show an increase to at least twice their pre-staftup levels and indicate at least 10% of scale (Le, 12.5 on the digital readout 0-125 scale) before the first SRM channel reaches 5 x JO cps (Technical Specifications, SR 3.3.1 :1.6) U
  • If desired, the level of the highest reading IRM (pre-startup) may be doubled and that value used as overlap criteria for all IRMs. This method will allow the operator to compare IRM channel response to a single value which is at least twice the pre-startup levels of the individual IRMs U APPROACH TO CRITICALITY AND PRESSURIZATION OGP-02 OF THE REACTOR Rev. 109 Page 16 of 54 6.2 Pulling Rods To Achieve Criticality (continued)

NOTE

  • With IRM channels betow Range 3, the SRM channels will initiate a rod withdrawal block when either of The following conditions exists:

SRM channel indicates greater than 2 x iO cps U O SRM channel indicates less than 102 cps with its detector NOT fult in U

  • SRM detectors are withdrawn two at a time so that the reactor flux level conditions are being monitored by channels that are NOT being affected by detector movement U
32. WHEN SRMIIRM overlap has been confirmed, THEN withdraw SRM detectors as required to maintain an indicated SRM count rate between .102 cps and 2 x cps CAUTION Repositioning IRM range switches is performed by one operator, using one hand, on one trip system at a time {8. 1 .6} U
33. As reactor power rises, reposition the IRM range switches to maintain IRM indication on recorders between 15 and 50 on the 0-125 scale
34. WHEN all OPERABLE IRM channels are above Range 3 AN prior to reaching Range 7, THEN fully withdraw all SRM detectors
13. 2150051 Which one of the following identifies the power supply to the APRM channel NUMACs?

A. All APRM channels receive 120 VAC power from UPS B. All APRM channels receive 120 VAC power from both RPS Bus A and RPS Bus B C. APRM Channels 1 & 3 receive power from ONLY12O VAC RPS Bus A APRM Channels 2 & 4 receive power from ONLYI2O VAC RPS Bus B D. APRM Channels 1 & 3 receive power from Division I 24/48 VDC APRM Channels 2 & 4 receive power from Division II 24/48 VDC Answer: B KJA:

215005 Average Power Range Monitor/Local Power Range Monitor K2 Knowledge of electrical power supplies to the following: (CFR: 41.7) 02 APRM channels RO/SRO Rating: 2.6/2.8 Tier2/Group 1 K/A Match: This meets the K/A because it is testing the power supply to the NUMACs.

Pedigree: Modified from 2015 NRC Exam Objective: LOI-CLS-LP-09.6, Objective 7a Describe the operational relationships between the PRNMS and the following:

Reactor Protection System

Reference:

None Cog Level: Fundamental Explanation: Each APRM channel NUMAC is equipped with a dual power supply arrangement with one supply from RPS Bus A and the other supply from RPS Bus B. All four APRM channels maintain power on loss of either supply as long as the other supply is available Distractor Analysis:

Choice A: Plausible because UPS supplies power to the APRM ODA and recorder Choice B: Correct Answer, see explanation.

Choice C: Plausible because this is the power supply arrangement for the voters.

Choice D: Plausible because other ranges of nuclear instrumentation (SRM/IRM) receive their power from here.

SRO Basis: N/A

2015 Exam Question:

Which one of the following is the power supply to APR.1 Channel 4 NUMAC on P608?

A 120 VAC RPS B. 12OVAC UPS C. 24148 VDC Dlvi ft 24/48 VDC Div Ii 2.8.8 PRNMS Power Supphes The Power Range Neutron monitoring System uses one Quadwple Voltage Power Supply (QLVPS) chassis and four Dual Low Voltage Power Supplies (DLVPS), one for each bay of the PRNMS panel, to provide redundant power to the NUMAC instruments. These LVPS convert 120 VAC to low voltage DC. See Figure 09.6-it.

Each APRM instrument receives power from two power supplies, LVPS 1 and LVPS 4. LVPS 1 is fed from RPS Bus A while LVPS 4 is fed from RPS Bus B. Therefore, a toss of an RPS Bus will not affect operation of the APRM NUMACS. Each RBM instrument also SD-09.6 Rev. 11 Page 32 of 94 4.3.1 Reactor Protection System APRM channels provide signals to open contacts in the scram trip logic ot the RPS System under various conditions discussed previou sly.

The RPS System provides power to each of the four APRM instruments, which in turn provide power to all subsystems driven from the APRM instruments or NUMAC. Both RPS busses, A and B, provide power to each APRM instrument, as well as, each RBM.

Therefore, a loss of one RPS bus will not affect operation of the PRNMS.

The reactor mode switch provides input to each APRM instrument to determine when to enforce the fixed or flow biased scram trip and rod block settings. OPRM circuitry is enabled only when power/flow conditions are met and the mode switch in RUN.

SD-09.6 Rev. ii Page48of94

14. 2150052 Which one of the following completes the statement below?

An APRM must have at least (1) of the assigned LPRMs operable with at least (2) LPRM inputs per axial level operable.

A. (1) 18 (2) 2 B. (1) 18 (2) 3 C. (1) 17 (2) 2 D. (1) 17 (2) 3 Answer: D K/A:

215005 Average Power Range Monitor/Local Power Range Monitor K5 Knowledge of the operational implications of the following concepts as they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: (CFR: 41.5/45.3) 04 LPRM detector location and core symmetry RO/SRO Rating: 2.9/3.2 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the knowledge of the LPRM inputs per axial level that are required and the minimum number of inputs for core symmetry that are required.

Pedigree: New Objective: LOl-CLS-LP-09.6, Obj. 13b Given plant conditions, predict the effect of a single or multiple LPRM failure on the following:

APRM

Reference:

None Cog Level: Fundamental Explanation: An APRM channel must have a minimum of 3 LPRM inputs per level and a total of 17 LPRM inputs to be operable Distractor Analysis:

Choice A: Plausible because an OPRM requires 18 LPRMs with at least 2 LPRM inputs to each cell.

Choice B: Plausible because an OPRM requires 18 LPRMs and 3 per level is correct for APRMS.

Choice C: Plausible because 17 is correct for APRMs and OPRMs require at least 2 LPRM inputs to each cell.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A

4.2.1 LPRM LPRM System failure, depending on the extent or failure type, can cause the loss of LPRM functions including the loss of indication, incorrect operation of rod block or scram protection. Generally, the following symptoms are exhibited for LPRM failure for the affected LPRM:

  • Indicates upscale, accompanied by an upscale alarm.
  • Indicates downscale, accompanied by a downscale alarm.
  • Indicator reads erratically.

The results of an LPRM failure may lead to an APRM or OPRM becoming inoperable- An APRM channel must have a minimum of 3 LPRM inputs per level and a total of 17 LPRM inputs to be operable.

SD-09.6 Rev. 12 Page 45 of 95 An OPRM cell must have a minimum of 2 LPRM inputs to each cell and a total of 18 cells to be operable.

15. 2170001 Following a loss of feedwater, RCIC automatically initiated and subsequently tripped on low suction pressure.

Current plant status is:

Reactor water level is 150 inches RCIC flow controller in Manual set at 200 gpm Subsequently, the following actions are taken:

RCIC suction transferred to Torus E51-V8, Turbine Trip and Throttle Valve is closed E51-V8 is re-opened PF push button on the RCIC flow controller is depressed Which one of the following identifies the indicated flow on the RCIC flow controller that would be observed for these conditions?

A. Ogpm B. 200 gpm C. 400 gpm D. 500 gpm a-1 :3---

Answer:

217000 Reactor Core Isolation Cooling System Al Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) controls including: (CFR: 41.5 /45.5) 01 RCIC flow RO/SRO Rating: 3.7/3.7 Tier2/Group 1 K/A Match: This meets the K/A because it is testing the prediction of what RCIC flow will be when operating the RCIC system.

Pedigree: New Objective: CLS-LP-016-A, Obj.16c Describe how the following evolutions are performed during operation of the RCIC System:

Adjusting RCIC flow in the Reactor Level Control mode.

Reference:

None Cog Level: high

Explanation: The RCIC Turbine is provided with a solenoid operated remote electrical tripping device, which when actuated (in this case by low suction pressure), will close the Turbine Trip and Throttle Valve, E51-V8. Resetting of the remote electrical tripping device may be accomplished from the RTGB. The RCIC system is restarted after auto initiation and turbine trip by fully closing the V-8, and re-opening the V-8. Located on the controller face is a PF (programmable function) pushbutton which when depressed an automatic transfer from manual to automatic at a predetermined setpoint of 400 GPM will result. This button (PF) has no function if the controller is already in automatic.

Distractor Analysis:

Choice A: This is plausible because this answer would be correct for these actions following a high RPV water level trip of RCIC Choice B: Plausible because this would be correct if the operator did not depress the PF pushbutton.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because the PF push button would raise RCIC flow to rated (400 gpm) and not maximum per procedure (500 gpm). Achieving 500 gpm would require the flow control setpoint to be manually raised.

SRO Basis: N/A From SD-16:

Also located on the controller lace is a PF (programmable fUnction) pushbutton. When depressed an automatic transfer from MANUAL to AUTOMATIC at a predetermined setpoint of 400 GPM will result.

NOTE: This button (PF) has no function if the controller is already in AUTOMATIC.

For various internal processing failures, the controller is designed to hold the last output and automatically switch to MANUAL giving the operator manual control capability. Barring operator intervention, this failure could result in rising or lowering RCIC 110w and would be indicated by the red FAIL lamp on the controller face. Failure display code can then be checked using the side panel keypad. A down scale failure of the controller is possible and would result in turbine operation at well below the normal minimum speed of 2000 rpm. An upscale failure is highly unlikely but would result in turbine speed at or above the maximum running speed of 4600 rpm. Failures associated with the dynamic response are also highly unlikely but would produce either excessively sluggish responses or dynamic instability (full scale oscillations) when in the Automatic mode. Programmable settings internal to the controller are maintained during a loss 0124 Vdc power stipply by a lithium battery. If this battery voltage drops to a pre-determined low value, the yellow ALARM light will flash. If the input signals are not within the limits 01-6.3% to 106.3% or if the input or output signals are not intact, the Yellow ALARM light will come on solid.

j SD-16 Rev. 12 Page29of120

REACTOR CORE ISOLATION COOLING SYSTEM 20P-1 6 OPERATING PROCEDURE Rev. 120 Page 99 of 99 ATTACHMENT 9 Page 1 0 1

<<RCIC Instructional Aid for EOPs>>

RESTARTING RCIC AFTER AUTO INITIATION AND TURBINE TRIP (20P-16 Section 8.?)

1. ENSURE THE E51-V8 (VALVE POSITION) AND E51-V8 (MOTOR OPERATOR) ARE CLOSED El
2. PLACE RCIC FLOW CONTROL IN MANUAL (M) AND ADJUST OUTPUTTOO% El
3. JOG OPEN E51-V8 UNTIL THE TURBINE SPEED IS CONTROLLED BY THE GOVERNOR El
4. FULLY OPEN E51-V8 El
5. SLOWLY RAISE TURBINE SPEED UNTIL FLOW RATE OF AT LEAST 120 GPM El
6. ENSURE E51-F019 IS CLOSED WITH FLOW GREATER ThAN 8OGPM El
7. WHEN SYSTEM CONDITIONS ARE STABLE, THEN ADJUST SETPOINT, AND TRANSFER RCIC FLOW CONTROL TO AUTO (A) El
8. SLOWLY ADJUST FLOW RATE USING RCIC FLOW CONTROL IN AUTO (A) El
9. ENSURE THE FOLLOWING:

BAROMETRIC CNDSR VACUUM PUMP HAS STARTED El SBGT STARTED (2OP-10) El SGT-V8 AND SGT-V9 ARE OPEN El

16. 2180001 Which one of the following completes both statements below concerning the Automatic Depressurization System (ADS) reactor water level inputs from the Nuclear Boiler System?

The (1) instruments provide LL3 inputs to ADS initiation logic.

The (2) range instruments provide LU inputs to ADS logic.

A. (1) FuelZone (2) Narrow B. (1) FuelZone (2) Shutdown C. (1) Wide range (2) Narrow D. (1) Wide range (2) Shutdown Answer: C K/A:

217000 Automatic Depressurization System KI Knowledge of the physical connections and/or cause-effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) 03 Nuclear boiler instrument system RO/SRO Rating: 3.7/3.8 Tier2/Group 1 K/A Match: This meets the K/A because it is testing the connection between ADS and level indicators.

Pedigree: New Objective: LOI-CLS-LP-0001.2, Obj 4a List the systems which receive input from the Vessel Instrumentation system for the following:

Level signal

Reference:

None Cog Level: Fundamental Explanation: B21-LT-N031(Wide Range) provide LL3 initiation from NO31A and C for Logic B and from N031 B and D for Logic A.

B21-LT-N042 (Narrow Range) provide LL1 confirmatory from N042A for Logic B and from N042B for Logic A.

Distractor Analysis:

Choice A: Plausible because the fuel zone instruments covers LL3 (45 inches) and the second part is correct.

Choice B: Plausible because fuel zone instruments covers LL3 (45 inches) and the shutdown range covers the LL1 setpoint (166 inches).

Choice C: Correct Answer, see explanation.

Choice D: Plausible because the wide range is correct and the shutdown range covers the LL1 setpoint (166 inches).

SRO Basis: N/A 4.1.2 Automatic Operation The ADS logic automatically opens the ADS valves in the event the HPCI System fails to maintain reactor level during a LOCA. The seven ADS valves open automatically when all the following conditions are met on either of tvo logic channels (A or B) associated with ADS:

Reactor low water level (113 from B21-LTS-ND31A and C or B and D).

Reactor confirmatory low water level (LL1 from B21 -LTS-N042A or B).

Operation of both pumps of an RHR loop or one Core Spray pump as indicated by a pump discharge pressure of 115 psig (either El l-PS-NQ16A AND C or B AND D or El 1-PS-NQ2OA AND C or B AND D for RHR or either E21-PS-N008A AND El i-PS-NOO9A or E21-PS-NQO8B AND E21-PS-NOO9B for CS).

A time delay of 83 seconds has elapsed (timer B21-TDPU-K5A or B).

AUTO/INHIBIT switches in AUTO for either or both logic channels A and B.

SD-20 Rev.3 PAGE26 of 62

a

  • 2 ci FIGURE 20-10
17. 2180002 Unit One is operating at power with Core Spray Pump I B under clearance.

A small break LOCA occurs simultaneously with a Loss of Off-site Power to both units.

DGJ and DG4 fail to start and tie onto their respective E bus.

The following plant conditions exist on Unit One:

A-03 (5-1) Auto Depress Timers Initiated In alarm A-03 (6-9) Reactor Low Wtr Level Initiation In alarm RPV pressure 600 psig Drywell pressure 13 psig Which one of the following completes both statements below?

ADS (1) auto initiate.

After ADS is initiated (either automatically or manually), RPV water level (2) be restored with BOTH RHR Loops.

A. (1) will (2) will B. (1) will (2) will NOT C. (1) wiIINOT (2) will D. (1) will NOT (2) will NOT Answer: D K/A:

218000 Automatic Depressurization System K3 Knowledge of the effect that a toss or malfunction of the AUTOMATIC DEPRESSURIZATION SYSTEM will have on following: (CFR: 41.7 /45.4) 01 Restoration of reactor water level after a break that does not depressurize the reactor when required RO/SRO Rating: 4.4/4.4 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the knowledge of the effect of the malfunction on auto initiation of ADS and how level will be restored.

Pedigree: Last used on 10-1 NRC exam Objective: CLS-LP-20 Obj. 16b Given plant conditions, predict how the following will be affected by a loss or malfunction of ADS/SRVs: Reactor water level

Reference:

None

Cog Level: high Explanation: With the loss of offsite power and 1 B CS pump under clearance this would leave only one pump available in each RHR loop. Therefore ADS logic is lost. Level will continue to lower until the ADS valves are manually opened (emergency depressurization) at which time the running low pressure pumps will be able to add water. Injection would be from the A Loop of RHR as the B Loop injection valves do not have power.

Distractor Analysis:

Choice A: Plausible because ADS does have initiation conditions except that the logic will not have the appropriate pumps lined up for injection. B Loop of RHR does not have power to the injection valves Choice B: Plausible because ADS does have initiation conditions except that the logic will not have the appropriate pumps lined up for injection. B Loop of RHR does not have power to the injection valves Choice C: Plausible because ADS will not auto initiate but the B Loop of RHR does not have power to the injection valves. B Loop of RHR does not have power to the injection valves Choice D: Correct Answer, see explanation.

SRO Basis: N/A SD-20 4.1.2 Automatic Operation The ADS logic automatically opens the ADS valves in the event the HPCI System fails to maintain reactor level during a LOCA. The seven ADS valves open automatically when all the following conditions are met on either of two logic channels (A or B) associated with ADS:

  • Reactor confirmatory low water level (LL1 from B21-LTS-N042A or B).
  • Operation of both pumps of an RHR loop or one Core Spray pump as indicated by a pump discharge pressure of 115 psig (either Eli -PS-NO1 6A AND C or B AND D or Eli -PS-NO2OA AND C or B AND D for RHR or either E21 -PS-NOO8A AND Eli -PS-NOO9A or E2 1 -PS-NOO8B AND E21 -PS-NOO9B for CS).
  • A time delay of 83 seconds has elapsed (timer B21-TDPU-K5A or B).
  • AUTO/INHIBIT switches in AUTO for either or both logic channels A and B. Reactor low water level (LL3 from B21-LTS-NO31A and C or B and D).
18. 223001 1 Which one of the following completes the statement below concerning the Fuel Zone instruments, N036 and N037, during a loss of drywell cooling?

The reference leg density will (1) causing the indicated level to read (2) than actual level.

A. (1) rise (2) higher B. (1) rise (2) lower C. (1) lower (2) higher D. (1) lower (2) lower Answer: C K/A:

223001 Primary Containment System and Auxiliaries K3 Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following: (CFR: 41.7 / 45.4) 09 Nuclear boiler instrumentation RO/SRO Rating: 2.8/3.1 Tier 2 / Group 2 K/A Match: This meets the K/A because it is testing the knowledge of a loss of DW cooling has on instrumentation.

Pedigree: Bank Objective: LOl-CLS-LP-001.2, Obj. 05c Explain the effect that the following will have on reactor vessel level and/or pressure indications:

High containment (primary and secondary) temperatures.

Reference:

None Cog Level: High Explanation: The reference leg length is longer than the variable leg length, therefore secondary temp increasing makes the instrument read higher than actual level.

Distractor Analysis:

Choice A: Plausible because density is a function of temperature and the temperature is rising. The second part is correct.

Choice B: Plausible because density is a function of temperature and the temperature is rising. The second part is plausible because if the first part is seen as right then this would be correct.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because the first part is correct and the second part is the opposite of the answer.

SRO Basis: N/A

19. 223002 1 Unit One is at 75% power.

The 1A RPS MG set trips.

No operator actions have been taken.

Which one of the following identifies the Main Steam Line Isolation Valve (MSIV) logic lamp status on P601 panel?

Inboard MSIV Logic Outboard MSIV Logic AC A.Q B.Q Q

C.Ø 0

D.,Q Q

Answer: C K/A:

223002 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off Al Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including: (CFR: 41.5 / 45.5) 01 System indicating lights and alarms RO/SRO Rating: 3.5/3.5 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the ability to predict the light status on a loss of a power supply Pedigree: New Objective: LOI-CLS-LP-0l 2, Objective 12 Given plant conditions, determine how the following will affect PCIS:

c. Loss of RPS

Reference:

None Cog Level: High Explanation: See Notes Section. RPS A provides power to PCIS Logic A. PCIS Logic A is Inboard AC and Outboard DC indicating lights on P601.

Distractor Analysis:

Choice A: Plausible because first part is correct. Outboard light is DC.

Choice B: Plausible because second part is correct. Inboard light is AC.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because the lights are just reversed. This would be true for Loss of RPS B.

SRO Basis: N/A 4.3.10 AC Distribution RPS MG sets supply power to the following PCIS related components:

RPS Bus A PCIS Trip System A logic PCIS Trip Channels Al and A2 logic Inboard isolation logic for valves:

Inboard reactor water sample valve Main Steam Line drains Shutdown cooling suction RWCU Inboard RHR Sample valves Drywell floor and equipment drains CAC/CAMS/PASS for LL1 and High Drywell pressure Valve operating power:

Inboard reactor water sample valve Inboard RHR Sample valves Drywell floor and equipment drains Inboard AC MSIV solenoids Reactor Building Vent Exh Rad Monitor NO1OA Main Steam Line Rad Monitors A and C (alarm function only)

SD-12 Rev. ii Page65of208 P601 panel. These lights are arranged above the MSIV control switches as follows:

TABLE 25-3, MSIV ISOLATION SIGNAL STATUS Light INBD DC INBD AC OUTUD DC OUTBD AC Solenoid 125 VDC RPS A 125 VDC RPS B Power A B PCIS Logic B A A B SD-25 Rev. 14 Pagei6of79

20. 234000 1 Which one of the following identifies the effect if both Refuel Bridge hoist grapple hooks are not open five seconds after placing the Engage/Release switch to Release?

A. Fuel Hoist Interlock is generated.

B. Engage amber light extinguishes.

C. Fault lockout is generated.

D. Grapple hooks will reclose.

Answer: D K/A:

234000 Fuel Handling A3 Ability to monitor automatic operations of the FUEL HANDLING EQUIPMENT including:

(CFR: 41.7 /45.7) 01 Crane/refuel bridge movement RO/SRO Rating: 2.6/3.1 Tier 2 I Group 2 K/A Match: This meets the K/A because it is testing the ability to monitor the crane grapple hooks auto re-close feature.

Pedigree: New Objective: LOl-CLS-LP-58.1, Obj 13 Describe the operation of the grapple if the ENGAGE/RELEASE Switch is positioned to RELEASE and both grapple hooks are not open within 5 seconds when the main hoist is loaded.

Reference:

None Cog Level: Fundamental Explanation: If the grapple does not indicate released (open) within 5 seconds, the solenoid is de-energized and the grapple hooks re-close. The switch must then be taken to the ENGAGE position to reset the logic prior to making another attempt to release the grapple.

Distractor Analysis:

Choice A: Plausible because a Fuel Hoist Interlock is generated for a number of reasons.

Choice B: Plausible because this is an indication of operation of the grapple hooks.

Choice C: Plausible because a fault lockout is generated for a number of reasons.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A

21. 2390021 Which one of the following identifies the SRV component that will prevent siphoning of water into the SRV discharge piping?

A. Vacuum breaker B. Check Valve C. I-Quencher D. Sparger Answer: A K/A:

239002 Safety Relief Valves K4 Knowledge of RELIEF/SAFETY VALVES design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) 03 Prevents siphoning of water into SRV discharge piping and limits loads on subsequent actuation of SRVs RO/SRO Rating: 3.1/3.3 Tier 2 I Group 1 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL.

This meets the K/A because it is testing the knowledge of the design feature that prevents siphoning of water.

Pedigree: New Objective: LOI-CLS-LP-020, Obj. 7d State the purpose of the following: SRV tailpipe vacuum breakers

Reference:

None Cog Level: Fundamental Explanation: Following operation of the valve, a vacuum is created in the SRV tailpipe as the steam condenses. Water in the line above the suppression pool water level would cause excessive pressure at the SRVs discharge when and if the valve reopened. For this reason, a vacuum relief valve is provided on each SRV tailpipe to prevent drawing water up into the line due to this steam condensation following SRV operation.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because this is a component that is typically used to provide an anti-siphon break.

Choice C: Plausible because this is a component on the SRV that the steam discharges through and has holes in the pipe which could be thought of an anti-siphon type break.

Choice D: Plausible because this is a component on the SRV that the steam discharges through and has holes throughout the pipe which could be thought of an anti-siphon type break. (The supplemental fuel pool cooling sparger has this design to prevent siphoning of water)

SRO Basis: N/A

22. 241000 1 Which one of the following identifies the criteria for tripping the main turbine lAW the Unit Two Scram Immediate Actions of OEOP-01-UG, Users Guide?

A. When APRMs indicate downscale trip.

B. When steam flow is less than 3 Mlbs/hr.

C. When reactor water level is 160 inches and rising.

D. When reactor mode switch is placed in SHUTDOWN.

Answer: A K/A:

241 000 Reactor/Turbine Pressure Regulating System A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 I 45.5 to 45.8) 14 Turbine trip ROISRO Rating: 3.8/3.7 Tier 2 / Group 2 K/A Match: This meets the K/A because it is testing the ability of tripping the turbine from the control room.

Pedigree: New Objective: LOl-CLS-LP-300-C, Obj. 2 List the immediate operator actions for a reactor scram.

Reference:

None Cog Level: Fundamental Explanation: The main turbine is tripped after reactor power is below 2% which is indicated by APRM downscale trip lights illuminated.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because this is a criteria for placing the mode switch to shutdown which is an immediate operator action.

Choice C: Plausible because this is a criteria for a reactor feed pump which is an immediate operator action.

Choice D: Plausible because this is an immediate operator action that is performed on the scram.

SRO Basis: N/A

ATTACHMENT 38 Page 7 oIl Unit 2 Scram Immediate Actions (OEOP-Of -UG)

SCRAM IMMEDIATE ACTIONS

1. Ensure SCRAM valves OPEN by manual SCRAM or ARI initiation.
2. WHEN steam 110w less than 3 x 1O lb/hr.

THEN place reactor mode switch in SHUTDOWN.

3, if reactor power below 2% (APRM downscale trip),

THEN trip main turbine.

4. Ensure master RPV level controller setpoint at +170 inches.
5. IF:
  • Two reactor feed pumps running AND
  • RPV le,el above +160 inches AND
  • RPV level rising, THEN trip one.
23. 245000 1 Which one of the following completes both statements below concerning the Main Generator Voltage Regulator?

The automatic voltage regulator maintains a constant generator (1) voltage.

While in the automatic voltage regulation mode, the manual voltage regulator setting (2) automatically follow the automatic setpoint.

A. (1) field (2) does B. (1) field (2) does NOT C. (1) terminal (2) does D. (1) terminal (2) does NOT Answer: D K/A:

245000 Main Turbine Generator and Auxiliary Systems K4 Knowledge of MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS design feature(s) and/or interlocks which provide for the following: (CFR: 41 .7) 07 Generator voltage regulation RO/SRO Rating: 2.5/2.6 Tier 2 / Group 2 K/A Match: This meets the K/A because this is testing the design of the auto regulator as to what it controls and whether the manual regulator automatically follows the auto regulator.

Pedigree: Bank Objective: LOl-CLS-LP-027.0, Obj 7c Given a simplified diagram of the Main Generator Voltage Regulator, explain how:

a. the MANUAL regulator controls Generator output voltage
b. the AUTOMATIC regulator controls Generator output voltage
c. to transfer from one Voltage Regulator to the other

Reference:

None Cog Level: Fundamental Explanation: The AVR controls terminal voltage while the manual regulator controls field voltage. The manual voltage regulator does not track the setpoint of the AVR, this must be manually adjusted in the control room.

Distractor Analysis:

Choice A: Plausible because the MVR controls field voltage and the DG manual voltage regulator does track the auto regulator setpoint.

Choice B: Plausible because the MVR controls field voltage and the second part is correct.

Choice C: Plausible because the first part is correct and the DC manual voltage regulator does track the auto regulator setpoint.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A 2.15 Excitation Control (Refer to Figure 27-12)

The Silicon Controlled Rectifier (SCR) bridge circuit is used as a variable DC voltage source to control the exciter field current as required by the AC or DC regulator. The source of the control signal for the SCRs is determined by the Regulator Mode Selector Switch (43CS) located on Panel XlJ-1.

When Manual is selected, the DC regulator maintains a constant generator field voltage that is determined by the Manual Volts Adjust Rheostat. When the Automatic regulator is selected, the AC regulator maintains a constant generator terminal voltage.

2.17.8 Generator Voltage Regulator Differential Voltmeter This is a standard voltmeter that measures the magnitude and polarity of the difference between the DC regulator output signal and the AC regulator output signal. When shining control from the DC voltage regulator to the AC regulator or back, it is important to ensure that the signals are the same. As an example, if the meter reads to the clockwise of zero, then the manual regulator output is less than the automatic regulator If the meter reads counter clockwise of zero, then the manual signal is larger than the automatic signal. The meter indicates 0-10 volts in both directions.

Failure to have the regulator control signals matched when shifting regulator modes may result in transients on the generator output.

The severity of the transient would be determined by the direction and magnitude of the mismatch.

SD-27 Rev. 19 Page22of129

24. 259001 1 Unit One Reactor Feed Pump I B is operating in automatic DFCS control at 4500 RPM.

The DFCS control signal to Reactor Feed Pump 1 B woodward governor immediately fails downscale.

Which one of the following completes the statement below?

Reactor Feed Pump I B speed will:

A. lowerto0 rpm.

B. lower to 1000 rpm.

C. lower to 2450 rpm.

D. remain at 4500 rpm.

Answer: D K/A:

259001 Reactor Feedwater System Al Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including: (CFR: 41.5 / 45.5) 04 RFP turbine speed: Turbine-Driven-Only RO/SRO Rating: 2.8/2.7 Tier 2 / Group 2 K/A Match: This meets the K/A because it is testing the ability to predict the response in parameters.

Pedigree: New Objective: LOl-CLS-LP-032.2, Obj. 1 3d Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event:

Loss of signal interface between controllers and processor.

Reference:

None Cog Level: High Explanation: If RFPT A(B) MAN/DFCS selector switch is in DFCS, and DFCS control signal subsequently drops below 2450 rpm, or increases to greater than 5450 rpm, then Woodward 5009 digital controls will automatically assume RFPT speed control and maintain current pump speed.

Distractor Analysis:

Choice A: Plausible if the student believes that a loss of input signal will cause the controller to use 0 as the input for the speed of the pump. (i.e. HPCI/RCIC controllers will fail to zero)

Choice B: Plausible because an idled RFP is maintained at 1000 rpm.

Choice C: Plausible because 2450 is the low end of the controller function.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A

NOTE: If RFPT A(B) MAN/DFCS selector switch is in DFCS, and DECS control signal subsequently drops below 2450 rpm, or increases to greater Than 5450 rpm, Then Woodward 5009 digital controls witl automatically assume REPT speed control and maintain current pump speed. In this conditton, The REPT wilt only respond to LQWEPJRAISE speed control switch commands until MAN/DFCS selector switch is placed in MAN, DFCS CTRL RESET pushbutton is depressed, and MAN/DFCS selector switch returned to DFCS 3.13 Plant management has recommended one RFPT be idled at 1000 rpm with The discharge valve closed, during conditions with one RFPT in service.

25. 259002 1 Which one of the following completes both statements below concerning the reactor feed pump turbine (RFPT) DFCS controls?

During a RFPT startup, transfer to DFCS control is performed when RFPT speed is approximately (1)

DFCS will automatically control the speed of the RFPT up to (2)

A. (1) 1000 rpm (2) 5450 rpm B. (1) 1000 rpm (2) 6150 rpm C. (1) 2550 rpm (2) 5450 rpm D. (1) 2550 rpm (2) 6150 rpm Answer: C K/A:

259002 Reactor Water Level Control System A3 Ability to monitor automatic operations of the REACTOR WATER LEVEL CONTROL SYSTEM including: (C FR: 41.7 / 45.7) 01 Runout flow control RO/SRO Rating: 3.0/3.0 Tier2/Group 1 K/A Match: This meets the K/A because this is testing the upper limit of the auto (DFCS) controls which in essence prevent pump runout of the reactor feed pumps.

Pedigree: new Objective: LOl-CLS-LP-032.2, Obj. 5d Describe the operation of the DFCS in the following operating modes:

Master Level Control Mode (auto and manual)

Reference:

None Cog Level: Fundamental Explanation: DFCS will be placed into service with the manual output set at 2550 RPM. The DFCS system will control the RFPT speed from 2450 5450 RPMs

Distractor Analysis:

Choice A: Plausible because 1000 RPM is the idle speed of the RFPT and the second part is correct.

Choice B: Plausible because 1000 RPM is the idle speed of the RFPT and 6150 is the overspeed setpoint of the woodward controls.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because the first part is correct and 6150 is the overspeed setpoint of the woodward controls.

SRO Basis: N/A

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev 206 Page 50 of 408 6.1.5 Reactor Feed Pump Startup from Idle Speed to Injection at Low Pressure Conditions (continued)

NOTE When using RFPTA(B) Lor/Raise speed control switch. reactor feed pump turbine speed wilt change at a rate of 50 rpm per second. If switch is held in LOWER or RAISE for greater than 3 seconds, the rate of change will rise to 375 rpm per second El

6. Maintain RFPT A(B) discharge pressure at least 100 psig greater than reactor pressure by adjusting RFPT A(B) LoweriRaise speed control stch until REPT speed is approximately 2550 rpm END RM. LEVEL R3 REACTIVITY EVOLUTION
7. Direct Radwaste Operator to monitor effluent conductivity for each in service CDD
8. WHEN REPT A(B) speed is approximately 2550 rpm, THEN raise C32-SIC-R6O1A(B) [RFPT A(B) Sp CII) output to match DECS Stpt and Speed Stpt on Panel P603 to within 100 rpm NOTE
  • When kEPT A(B) Man/DECS control switch is placed in DFCS, C32-SIC-R60 1A(B) [RE PT At B) Sp CtlJ will control REPT speed El
  • When REPT A(B) Mari/DECS control switch is placed in DECS, and DFCS is in control, the REPT AfB) DFCS Ctrl light will be ON El
  • If REPT At B) Man/DFCS selector switch is in DFCS and DFCS control signal subsequently drops to less than 2450 rpm or rises to greater than 5450 rpm.

Woodward 5009 digital controls will automatically assume REPT speed control and maintain current pump speed. In this condition, the RFPT will only respond to LowriRaise speed control switch commands unW the Man/DECS selector switch is placed in MAN, DFCS Ctrt Reset pushbutton is depressed, and the Man/DECS selector switch returned to DFCS El

9. Confirm the following RFPT A(B) speed signals on Panel P603 agree within 100 rpm:
  • DECS Stpt (speed demand from DECS)
  • Speed Stpt (speed demand from 5009 control)
  • Act Spd (actual REPT speed)
10. Place ManJDECS control switch in DECS

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev. -4)6 Page Zof 406 3.0 PRECAUTIONS AND LIMITATIONS (continued) 5 Any of the following conditions will automatically trip a mactot feed pump turbine:

RFPT Woadward 5009 overspeed greater than or equal to 6150rpm

26. 261000 1 Unit One primary containment venting is being performed lAW lOP-b, Standby Gas Treatment System Operating System, with the following plant status:

i-VA-i F-BFV-RB, SBGT DWSuct Damper Open 1-VA-i D-BFV-RB, Reactor Building SBGT Train IA Inlet Valve Closed 1-VA-i H-BFV-RB, Reactor Building SBGT Train 1 B Inlet Valve Closed Which one of the following completes both statements below concerning the predicted SBGT response if drywell pressure rises to 1 .9 psig?

i-VA-iF-BFV-RB (1)

Both i-VA-i D-BFV-RB and i-VA-i H-BFV-RB (2)

A. (1) auto closes (2) auto open B. (1) auto closes (2) remain closed C. (I) remains open (2) auto open D. (1) remains open (2) remain closed Answer: A K/A:

261 000 Standby Gas Treatment System A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) 02 Suction valves RO/SRO Rating: 3.1/3.1 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the ability to monitor SBGT suction valves.

Pedigree: Last used on 2014 NRC Exam Objective: LOl-CLS-LP-004.1, Obj 5 List the signals and setpoints that will cause a Secondary Containment isolation

Reference:

None Cog Level: High Explanation: The filter train fans will automatically start on High Drywell Pressure. The following actions occur: 1) SBGT Reactor Building suction dampers (1D-BFV-RB and 1H-BFV-RB) open, 2)

SBGT DW Suct Damper (F-BFV-RB) closes. The SBGT Train A/B Suction & Discharge Valves on Ui do not auto open. These valves on U2 do auto open, so there could be a misconception on these valves (inlet vs. suction dampers).

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because iF does auto close and SBGT Train lA/B Suction Valves (1C & 1E) on Unit One only do not auto open Choice C: Incorrect since SBGT will auto realign from primary containment to the Reactor Building on system initiation Choice D: Incorrect since SBGT will auto realign from primary containment to the Reactor Building on system initiation and SBGT Train lA/B Suction Valves (1 C & 1 E) on Unit One only do not auto open SRO Basis: N/A 2.1.6 Fan A 100% capacity, heavy-duly, industrial type Fan and motor assembly is provided in each SBGT filter train. Each Fan will produce the required 2700 3300 scfm flow through its associated filter tmin Each Fan is driven by a direct-drive AC motor which is energized from a redundant and separate emergency power supply. The Unit I A and 5 Fans are powered from 480 VAC MCCs I XE and 1 XF respectively and Unit 2 A and B Fans from 2XE and 2XF.

The filter train fans may be operated manually from controls located at RTGB XU-51.

The filter train fans will automatically start if any of the following Secondary Containment isolation conditions exist: (Figure 10-2)

1. Low Reactor Water Level, LL #2 2 High Drywell Pressure
3. Reactor Building Ventilation Radiation (Figure 10-3) 3.2.6 Automatic
1. Upon receipt of an automatic initiation signal both trains of SECT will starL Unit 1 ONLY The dampers associated with Unit 1 SECT System will receive automatic open signals when an initiation signal is received EXCEPT for the train inlet and outlet dampers, (BFIs-f 5,1 C, 1 E,and 1 G)

Should these normally open dampers be manually closed locally via Their CLOSE/OPEN pushbuttons, they will NOT automatically reopen and the associated SBGT will not automatically start SD-b Rev7 Page i8ot38

27. 262001 1 Unit One is operating at rated power.

Unit Two is in MODE 5 performing fuel movements.

Which one of the following completes both statements below lAW Unit One Tech Spec 3.8.1, AC Sources Operating, LCO statement?

The Unit Two SAT (1) required to be OPERABLE.

(2) Diesel Generators are required to be OPERABLE.

A. (1) is (2) Two B. (I) is (2) Four C. (1) is NOT (2) Two D. (1) is NOT (2) Four Answer: B K/A:

262001 A.C. Electrical Distribution G2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 /43.2 /43.5 / 45.3)

RO/SRO Rating: 3.4/4.7 Tier 2 / Group 1 K/A Match: This meets the K/A because this is testing the items above the line for TS 3.8.1.

Pedigree: New Objective: LOI-CLS-LP-050, Obj. 16 Given plant conditions, determine whether given plant conditions meet minimum Technical Specifications requirements associated with the 230 KV Electrical Distribution system.

Reference:

None Cog Level: Fundamental Explanation: Unit One Tech Specs require with Unit One in Mode 1, both SATs and both UATs and all four DGs are required to be operable. This would change if Unit One was not in Mode 1, 2, or

3. Unit Two Tech Specs do not require the SAT, it only requires one offsite circuit.

Distractor Analysis:

Choice A: Plausible because the first part is correct and there are only two Unit One DGs but all four are required for the LCO to be met.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because this is the Unit 2 SAT and asking if it is required for Unit 1 TS (it is not required forthe Unit Two TS) and whether only the 2 Unit One DGs are required or all four of the DGs.

Choice D: Plausible because this is the Unit 2 SAT and asking if it is required for Unit 1 TS (it is not required for the Unit Two TS) and the second part is correct.

SRO Basis: N/A AC SourcesOperating 3.8,1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC SourcesOperating LCD 3.8.1 The following AC electrical poer sources shall be OPERABLE:

a. Two Unit 1 qualified circuits between the ofisite transmission network and the onsite Class 1E AC Electrical Power Distribution System,
b. Four diesel generators (DGs), and
c. Two Unit 2 qualified circuits between the offsite transmission netork and the onsite Class 1E AC Electrical Power Distribution System APPLICABILITY: MODES 1,2, and 3.

ACTIONS NOTE LCD 3.O.4.b is not applicable to DGs.

28. 262002 1 Unit One is operating at rated power.

Subsequently, El breaker AU9, Feed to 480V Substation E5, trips.

Which one of the following completes the statement below?

120V UPS Distribution Panel 1A is:

A. de-energized.

B. energized from MCC ICB.

C. energized from the Standby UPS.

D. energized from 250V DC SWBD A.

Answer: D K/A:

262002 Uninterruptable Power Supply (A.C. /D.C.)

A3 Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (AC/D.C.)

including: (CFR: 41.7 / 45.7) 01 Transfer from preferred to alternate source RO/SRO Rating: 2.8/3.1 Tier 2 I Group 1 K/A Match: This meets the K/A because it is testing the ability to monitor the transfer to the alternate power source.

Pedigree: New Objective: LOI-CLS-LP-052, Obj. 5 Given plant conditions, determine the lineup of the primary UPS, the Standby UPS, and their reserve sources.

Reference:

None Cog Level: High Explanation: The UPS system is normally aligned such the primary inverter is powering UPS loads. The standby inverter is energized but bypassed with the Manual Bypass switch in Bypass Test.

The static transfer switch of the Primary inverter (and also the Standby inverter) is receiving an input from the alternate (hard) source. If the primary power source is lost f in this case the loss of E5 which powers MCC CA) the alternate power source from the 250V batteries will keep the loads energized with no need for the inverter to swap to the hard source.

Distractor Analysis:

Choice A: Plausible because the normal power source is lost.

Choice B: Plausible because this is the hard source for the Distribution Panel.

Choice C: Plausible because this is an available power source for the Distribution Panel.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A

14fl.

t) r r

-]

29. 2150032 A reactor shutdown is in progress.

All IRMs on range I reading between 15 and 20.

IRM B detector is failing downscale.

Which one of the following completes both statements below?

lAW A-05 (1-4) IRM Downscale, the alarm setpoint is (1) on the 125 scale.

When the IRM downscale alarm is received, a rod block (2) be generated.

A. (1) 3 (2) will B. (1) 3 (2) will NOT C. (1) 6.5 (2) will D. (1) 6.5 (2) will NOT Answer: D K/A:

215003 Intermediate Range Monitor fIRM) System K6 Knowledge of the effect that a loss or malfunction of the following will have on the INTERMEDIATE RANGE MONITOR fIRM) SYSTEM: (CFR: 41.7/45.7) 04 Detectors RO/SRO Rating: 3.0/3.0 Tier 2 I Group 1 K/A Match: This meets the K/A because this is testing a failure/malfunction of a detector effect on the IRM system (whether it generates a rod block)

Pedigree: New Objective: LOI-CLS-LP-009-A, Obj. 3a List the SRM/IRM system signals/conditions that will cause the following actions and the conditions under which each is bypassed: Rod Blocks

Reference:

None Cog Level: High Explanation: The downscale setpoint for the IRMs is 6.5 on the 125 scale. The rod block is bypassed under these conditions because the IRMs are all on Range 1.

Distractor Analysis:

Choice A: Plausible because 3 is the downscale tech spec setpoint for SRM5 and if the IRMs were not all on range 1 a rod block would be generated.

Choice B: Plausible because 3 is the downscale tech spec setpoint for SRMs and the second part is correct.

Choice C: Plausible because the first part is correct and if the IRMs were not all on range 1 a rod block would be generated.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A Unot 2 APP ATh 14 Page 1 of 2 1PM DOWNS CALE AUTO ACTIONS

1. Rod withdrawal blcck (bypassed when 1PM range switch f:.r the affected channel is on Range 1 cr when the reactor node swatch is an P.UN).
2. Computer printout.

CAUSE I. 1PM channel(s) indicatang less than or equal to E.S on the 0-125 scale when its range switch is not on Range 1.

2. Improper ranging cf 1PM channels during reactor startup or shutdcwn.
3. 1PM detectc-r not fully inserted.
4. 1PM detecccr failure.

S. Circuit malfunction.

03 SERVATI ONS

1. 1PM channel indacating less than cr equal to CS cn the 0125 scale.
2. 1PM downscale (ONSC1 white indicating laght is on.
3. ROD OUT BLOCK (A-CS 2-21 alarm, if affected 1PM channel is not on Range 1.
4. If the affected 1PM channel(s) is not on Range 1, the rod withdrawal permissive indicating light will he off.
30. 263000 1 Unit Two is operating at full power when a loss of DC Distribution Panel 4A occurs.

Which one of the following completes both statements below?

RCIC is (1) for injection from the RTGB.

RCIC (2) isolation logic has lost power.

A. (1) available (2) inboard B. (1) available (2) outboard C. (1) unavailable (2) inboard D. (1) unavailable (2) outboard Answer: A K/A:

263000 D.C. Electrical Distribution G2.2.37Ability to determine operability and/or availability of safety related equipment.

(CFR: 41.7 /43.5 /45.12)

RO/SRO Rating: 3614.6 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the operability of RCICIADS.

Pedigree: New Objective: LOl-CLS-LP-051, Obj. 7 Given plant conditions, determine the effect that a loss of DC power will have on the following:

d. Reactor Core Isolation Cooling.
e. Automatic Depressurization System.

Reference:

None Cog Level: High Explanation: RCIC will not shutdown on reactor water level and the inboard isolation logic is powered from Division 11125 VDC panels 4B for Unit 2.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because the first part is correct and a loss of 4B would cause the outboard logic to be lost.

Choice C: Plausible because a loss of 4B would cause RCIC to be unavailable for injection and the second part is correct.

Choice D: Plausible because this would be correct for a loss of 4B.

SRO Basis: N/A LOSS OF DC POWER OAOP-39.0 Rev. 042 Page 320134 ATFACHF1ENT 6 Page 1 of 2

<<Plant Effects from Loss of DC Distribution Panel 35(4B)>>

RCIC: Will .NI auto initiate, outboard isolation logic INOPERABLE (E51-F008, -F029, and -F066 will NOT auto close), RCIC turbine will NOT trip except on overspeed, RCIC flow controller and EGM INOPERABLE (no flow control or indication), E5i-F045 ll NOT auto close on high water level, E51-F004, -F054, and -F026 fail closed RCIC isolation is required in accordance with APP 1(2)-A-03 1-4.

<<Plant Effects from Loss of DC Distribution Panel 3A(4A)>>

RCIC: Will IAI shutdown on reactor high water level, inboard isolation logic INOPERABLE fE5l-E007, -F031, and -F062 will NOT auto close) Valves E5l -F005 and -F025 fail closed.

31. 2640001 Unit Two has lost off-site power.

DG3 started and tied to its respective E Bus.

Sequence of events:

1200 DG3 ties to E3 1205 DG3 lube oil temperature rises above 190°F 1206 DG3 lube oil pressure drops below 27 psig Which one of the following identifies when DG3 will trip?

A. Immediately at 1205.

B. Immediately at 1206.

C. 45 seconds after 1205.

D. 45 seconds after 1206.

Answer: B KJA:

264000 Emergency Generators (Diesel/Jet)

K6 Knowledge of the effect that a loss or malfunction of the following will have on the EMERGENCY GENERATORS (DIESEL/JET): (CFR: 41 .7/45.7) 03 Lube oil pumps RO/SRO Rating: 3.5/3.7 Tier 2 / Group I K/A Match: This meets the K/A because it is testing the effect of a loss of lube oil on the EDG.

Pedigree: Bank Objective: LOl-CLS-LP-039, Obj. 4a Given plant conditions, determine if EDGs will trip: After an auto start (LOCT)

Reference:

None Cog Level: High Explanation: Hi lube oil temperature bypassed by auto start signal (LOOP and LOCA). Low lube oil pressure trip never bypassed. On a start of the DG the low lube oil trip is bypassed for 45 seconds.

Distractor Analysis:

Choice A: Plausible because hi lube oil temperature is a trip, but it is bypassed on the LOOP.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because there is a 45 second time delay associated with the lube oil trip on an initial start of the EDG.

Choice D: Plausible because there is a 45 second time delay associated with the lube oil trip on an initial start of the EDG.

SRO Basis: N/A

SD-39 I EMERGENCY DIESEL GENERATORS Rev. 20 SYSTEMDESCRIPT1ON Page46of166 3.3.3 Automatic Stop Control (Figure 39-14)

Under conditions vtiere continued Diesel Generator operation may cause damage to the Diesel itself automatic shutdowns are provided.

The shutdown signals will vary dependent upon whether the engine has been started manually or automatically.

When operating due to receipt of an automatic start signal the following trips and lockout are provided:

  • Overspeed 575 (561 to 589) rpm When operating as a result of an initiation from a normal non-emergency start the folloMng rnps and lockouts are enforced in addition to those listed above:
  • High jacket water temperature 200°F
  • Jacket Water Low pressure 12 psig The low lube oil pressure, and low jacket iwter pressure shutdowns are blocked for the first forty-five second on initiation of an engine start sequence (auto or manual). This permits the conditions to be established which will prevent these shutdowns during engine operation
32. 216000 1 A Unit Two plant cooldown is being performed with the following plant conditions:

Reactor water level 175 inches, steady Reactor pressure band 500 700 psig Drywell ref leg temp 175°F (REFERENCE PROVIDED)

Which one of the following completes both statements below?

The lowering of reactor pressure causes the NOO4NB/C (Narrow Range) reactor water level instruments indicated level error to (1)

The reactor water level that would correspond to Low level 4 (LL4) is (2)

A. (1) increase (2) -60 inches B. (1) increase (2) -65 inches C. (1) decrease (2) -60 inches D. (1) decrease (2) -65 inches Answer: A K/A:

216000 Nuclear Boiler Instrumentation A2 Ability to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 I 45.6) 11 Heatup or cooldown of the reactor vessel ROISRO Rating: 3.2/3.3 Tier 2 / Group 2 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL.

The first part of the question deals with predicting the effect of a cooldown on indicated level error while the second part has the student has to determine based on the lowering pressure what LL4 value would be which is the value that emergency depressurization would be required. They must utilize the lower end of the pressure band to determine LL4. If LL4 cannot be maintained then ED is required.

Pedigree: New Objective: LOl-CLS-LP-001.2, Objective 5a Explain the effect that the following will have on reactor vessel level and/or pressure indications:

Plant heatup/cooldown

Reference:

OEOP-01-UG, Attachment 26

Cog Level: High Explanation: The indicated level error is sensitive to changes in the saturation density of the bulk water as a function of system pressure. The amount of the indicated level error is also a function of the difference in the actual water level and the variable leg instrument tap elevation. As the saturation density increases (pressure decreases) the indicated level error will increase for the narrow and wide range instruments and decrease for the fuel zone and shutdown range instruments due to calibration criteria.

From 01-37.11 TAF, LL4, and LL5 values should be determined based on the reference leg area temperature and RPV pressure compensation curves, using RPV pressure at the low end of the established RPV pressure control band. Based on the low end of the band of 500 psig and <200°F in the drywell the LL4 value would be -60 inches.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because the first part is correct and the second part would be correct for 700 psig.

Choice C: Plausible because this would be correct for the fuel zone or shutdown range instruments and the second part is correct.

Choice D: Plausible because this would be correct for the fuel zone or shutdown range instruments and the second part would be correct for 700 psig.

SRO Basis: N/A

USERS GUIDE QEOP-Of-UG Rev. 067 Page 94 of 156 ATTACHMENT 26 Page 1 of 1

<<Unit 2 RPV Level at LL 4 (Minimum Steam Cooling RPV Level)>>

0

-10

-20 (0

U]

-30 0

+I{++4 44-f 4-I.3-444 ,-l 4 4 J 4_f 4.1.4

-40

-J Ui

> -50 Ui

-J REFLEG

UJ -60 41.3_f1..I43443_f;1-fI41.1-1.43_f1.3.I1.iJ 4.4 41.4 tEMP A5OE 0 -70 zitI[I1UllIll[lltllt111ttt1.ifl 111111 I r JP TEMP LEG z -80 z______ Ill[ll[% 1 ELQQR EQUAL TO

-90 100 H IITI]IW I I I I I L150 jioo 300 500 700 900 1,100 60 200 400 600 800 1,000 RPV PRESSURE (PSIG)

When RPV pressure is less than 60 psig, use indicated level. LL-4 is -27.5 inches.

4.1.2 System Pressure (Heat-up and Cool-down)

The indicated level error is sensitive to changes in the saturation density of the bulk water as a function of system pressure. The amount of the indicated level error is also a function of the difference in the actual water level and the variable leg instrument tap elevation.

As the saturation density increases (pressure decreases) the indicated level error will increase for the narrow and wide range instruments and decrease for the fuel zone and shutdown range instruments due to calibration criteria. As actual water level decreases, the amount of error will decrease because less vessel water level is acting on the instrument.

SD-Ui .2 Rev. 10 Page 36 of 85

TRANSIENT MGATI0N GUIDELINES 001-37.11 Rev 4 Page 17 cr25 5.3.3 1(2)EOP-01-RVCP, Reactor Vessel Control Procedure Level Leg

a. The CRS directs an initial RPV level band of +166 to +206 inches. The reactor operator actually maintains a RPV level band of +170 to

+200 inches to provide additional margin to the reactor scram and turbine trip set points. The CRS may direct a widened band bed on plant conditions and other controlling procedures associated with the transient.

b. If RPV level is above TAF, injection flow should be controlled so as to control the cooldowTl rate below i00°FIhr.
c. It RPV level is below TAF, RPV level should be rapidly restored to above TAF. and then injection flow reduced so as to control The cooldown rate below 100°F/hr.
d. TAF. LL4, and LL5 values should be detemlined based on The reference leg area temperature and RPV pressure compensation curves, using RPV pressure at the low end of the established RPV pressure control band.
33. 272000 1 Unit Two is performing a startup lAW OGP-02, Approach to Criticality and Pressurization of the Reactor.

lAW OGP-02, which one of the following identifies the radiation monitor(s) that will require the alarm setpoints raised when HWC is placed in service?

A. D12-RM-K603A,B,C,D, Main Steam Line Rad Monitors B. ARM Channel 2-9, U-2 Turbine Bldg Breezeway C. D12-RR-4599-1,2,3, Main Stack Rad Monitors D. ARM Channel 2-4, Cond Filter-Demin Aisle Answer: A K/A:

272000 Radiation Monitoring System K5 Knowledge of the operational implications of the following concepts as they apply to RADIATION MONITORING SYSTEM: (CFR: 41.7/45.4) 01 Hydrogen injection operations effect on process radiation indications RO/SRO Rating: 3.2/3.5 Tier 2 / Group 2 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL.

This meets the K/A because this is testing the operational implication as to which rad monitor, if asked as the operational effect on the individual rad monitor this would provide no discrimatory value.

Pedigree: New Objective: LOI-CLS-LP-059, Obj. 8 Explain why Chemistry must be notified when starting and securing the HWC System.

Reference:

None Cog Level: Fundamental Explanation: The excess Hydrogen injected into the reactor coolant creates the driving force to shift the Nitrogen-16 distribution ratio, resulting in a larger fraction of the Nitrogen-16 forming volatile Ammonia and a smaller fraction forming Nitrites and Nitrates. This additional volatile Ammonia is then carried over in the reactor steam resulting in higher background radiation levels. Any increase in Hydrogen injection rates will result in a proportional increase in background radiation levels and vise-versa.

OGP-02 has a step for ensuring that the rad monitors are adjusted based on this background rad level increase.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because when HWC is placed in service the rad levels will increase minimally and HWC H2 is injected in the reactor feed pumps.

Choice C: Plausible because sufficient decay time is available for N-16 such that radiation levels wouldnt raise that much in this area.

Choice D: Plausible because when HWC is placed in service the rad levels will increase minimally and this is downstream of the HWC 02 injection point.

SRO Basis: N/A APPROACH TO CRFRCALFFY AND PRESSURIZATION OGP-02 OF THE REACTOR Rev. 110 Page 6 of 54 3.0 PRECAUTIONS AND LIMITATIONS (continued)

15. B21-F032A and B21-F032B (Feedwater Supply Line Isolation Valves), are stop-check valves These valves are designed to prevent leakage from the reactor into the feedwater system. These valves are not designed to positivety close against condensate system pressure. As such, with the reactor depressurized and the condensate system in service, these valves may teak by. causing reactor water level to rise 0
16. The Main Steam Line Radiation Monitor (MSLR) Htgh-High Radiaon setpoint is adjusted assuming HWC is in service. If HWC is removed from service for an extended period of time (greater than one week), 1(2)OP-59, Hydrogen Water Chemistry System Operating Procedure requires BESS determine if a MSLRM High-High Radiation setpoint adjustment is required 0
17. The HWC System wilt normally be placed in service immediately after establishing the following conditions:
  • At least one Condensate Booster Pump feeding the reactor with minimum flow valve closed 0
  • At least one SJAE operating at greater than or equal to half-load 0
34. 226001 1 Which one of the following identifies the power supply to 2D RHR Pump?

A.E1 B. E2 C. E3 D. E4 Answer: B K/A:

226001 RHR/LPCI: Containment Spray System Mode K2 Knowledge of electrical power supplies to the following: (CFR: 41 .7) 03 Pumps RO/SRO Rating: 2.9/2.9 Tier 2 / Group 2 K/A Match: This meets the K/A because this is testing the power supply to RHR pumps which are the pumps for the containment sprays.

Pedigree: Modified from the 2012 NRC Exam (changed to the D RHR Pump)

Objective: LOI-CLS-LP-017-A Obj. 173 List the normal and emergency power sources for the following: RHR Pumps.

Reference:

None Cog Level: Fundamental Explanation: 2D RHR pump is a Div II pump with a power supply from E2.

Distractor Analysis:

Choice A: Plausible because El is a Unit One bus that supplies power to Unit One and Unit Two loads.

RHR Pumps 1C and 2C are supplied from this bus.

Choice B: Correct Answer, see explanation Choice C: Plausible because E3 is a Unit Two bus that supplies power to Unit One and Unit Two loads.

RHR Pumps 1A and 2A are supplied from this bus.

Choice D: Plausible because E4 is a Unit Two bus that supplies power to Unit One and Unit Two loads.

RHR Pumps lB and 28 are supplied from this bus SRO Basis: N/A

9GURE 17-2B Low Pressure Injection Systems and Power UNIT I LOW PRESSURE ECOS UNIT 2 LOW PRESSURE EGGS Lz:r NOTE INJEC11ON FLOW PATH AND POWER SUPPLIES SHOWN.

LOGIC & OTF FLOW PAThS NOT SHOWN.

SD-17 Rev. 19 Page1OOofi28

35. 295001 1 Unit One is operating at 70% power when the OATC observes indications for a failed jet pump. Subsequently, Recirc Pump 1A trips.

Which one of the following completes both statements below lAW IAOP-04.0, Low Core Flow?

Performance of the jet pump operability surveillance for (1) Loop Operation is required.

If it is determined that a jet pump has failed, the required action is to (2)

A. (1) Single (2) reduce reactor power below 25% rated thermal power B. (1) Single (2) commence unit shutdown lAW OGP-05, Unit Shutdown C. (1) Two (2) reduce reactor power below 25% rated thermal power D. (1) Two (2) commence unit shutdown lAW OGP-05, Unit Shutdown Answer: B K/A:

295001 Partial or Complete Loss of Forced Core Flow Circulation G2.2.l2Knowledge of surveillance procedures. (CFR: 41.10 /45.13)

RO/SRO Rating: 3.7/4.1 Tier 1 I Group 1 K/A Match: This meets the K/A because this is testing knowledge of which surv. is required and the action if it has failed the surv.

Pedigree: New Objective: LOl-CLS-LP-302-C, Obj 4 Given plant conditions and AOP-04.0, determine the required supplementary actions.

Reference:

None Cog Level: High Explanation: The indications given are for a failed jet pump which lAW the AOP require the surveillance performed for determination of a failed jet pump. Unlike the selection of the power to flow map the PT only looks at the recirc pumps for determination of single loop or two loop operation. The power to flow maps for single loop are not used until the APRM setpoint adjustments are made.

Distractor Analysis:

Choice A: Plausible because single loop is correct and 25% is the requirement for when the PT is required to be performed.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because APRM setpoint adjustments have not been made which is a determination of how to use the power to flow maps and 25% is the requirement for when the PT is required to be performed.

Choice D: Plausible because APRM setpoint adjustments have not been made which is a determination of how to use the power to flow maps and the second part is correct.

SRO Basis: N/A

LOW CORE FLOW 2AOP-04.O Rev. 37 Page 18 of 25 4.2 Supplementary Actions (continued)

NOTE Jet pump failure is indicated by the following-

  • Reduction in generator megawatt output on GEN-WMR-760 (Net Unit Megawatts)
  • Reduction in core thermal power
  • Rise in indicated total core flow on B21-R613 (Core A Pressure/Core flow) recorder
  • Reduction in core plate differential pressure on 621-R613 (Core A Pressure/Core Flow) recorder
  • Rise in recirculation loop flow in the loop with a failed jet pump on 332-R614 (Recirculation flow) recorder CAUTION Under conditions of jet pump failure, indicated core flow on Process Computer Point U2CPWTCF and 621-R613 (Core A Pressure/Core Flow) recorder, will NOT be accurate. Accurate core flow is available from Process Computer Point U2NSSWDP (Core Plate Differential Pressure) or Attachment I, Esilmated Total Core Flow vs. Core Support Plate Delta P for B2C22. Until Step 23.b(1), the operating point on the Power-to-Flow Map will NQI be accurate. Indicated total core flow on B21-R613 (Core A Pressure/Core Flow) recorder will continue to be inaccurate until the failed jet pump is repaired 0
23. ffijet pump failure is suspected, THEN perform the following:
a. IF reactor power is greater than or equal to 25%,

THEN ensure the following:

  • OPT-i 3.1. Reactor Recirculation Jet Pump Operability, is performed for two loop operation 0 OR
  • OPT-i 3.4. Reactor Recirculation Jet Pump Operability for Single Loop Operation, is performed for single loop operation 0
b. IF any jet pump is determined to be INOPERABLE, THEN perform the following:

(1) Ensure the input to the Power-to-How Map has been changed from WTCF to core plate differential pressure LI (2) Notify the Duty Reactor Engineer the input to the Power-to-Flow Map has been changed from WTCF to core plate differential pressure LI (3) Commence unit shutdown in accordance with OGP-05, Unit Shutdown, in compliance with Technical Specification 3.42 LI

36. 295003 1 Unit One is operating at rated power.

The load dispatcher reports degraded grid conditions with the following indications:

Generator frequency 59.7 hertz 230 KV Bus 1A voltage 205 KV 230 Ky Bus 1 B voltage 205 KV El voltage 3690 volts E2 voltage 3685 volts Which one of the following completes both statements below?

The (1) may be damaged with continued operation under these conditions.

lAW OAOP-22.0, Grid Instability, the E-Bus master/slave breakers (2) open.

A. (1) main turbine blades (2) will B. (1) main turbine blades (2) will NOT C. (1) emergency bus loads (2) will D. (1) emergency bus loads (2) will NOT Answer: C K/A:

295003 Partial or Complete Loss of A.C. Power AK1 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF AC. POWER: (CFR: 41 .8 to 41.10) 03 Under voltage/degraded voltage effects on electrical loads RO/SRO Rating: 2.9/3.2 Tier 1 / Group 1 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL.

This meets the K/A because this is testing the degraded voltage conditions.

Pedigree: Last used on the 10-2 NRC Exam Objective: LOl-CLS-LP-302-G, Obj. 4b Given plant conditions and any of the following AOPs, determine the required supplemental actions: AOP-22.0, Grid Instability

Reference:

None Cog Level: High

Explanation: There are frequency based criteria in AOP-22.0 (Caution directly preceding step 3.2.1) for tripping the turbine to prevent resonance vibration of low pressure blades due to off frequency operation. Time limits include, 5 minute ranges and 1 minute ranges. At this current frequency, the Main Turbine can be operated indefinitely, which will not cause turbine damage. Sustained low voltage provides for higher running currents which will damage running ESF motors. Per the automatic actions section of AOP-22.0, the degraded voltage relays will actuate when emergency bus voltage has dropped below 3700 VAC for 10 seconds. This trips the Master/Slave breakers (BOP bus supply to E Buses) and the DGs start and load.

Distractor Analysis:

Choice A: Plausible because turbine blade damage can occur due to off frequency operation and the second part is correct.

Choice B: Plausible because turbine blade damage can occur due to off frequency operation and the second part is plausible because the frequency is within range.

Choice C: Correct Answer, see explanation Choice D: Plausible because damage to E Bus loads is correct and the second part is plausible because the frequency is within range.

SRO Basis: N/A dSyIIIIJILIIt.uI SI lUlL L..JIL.UIt t.UIIIL UI I.UUftM tWIs).

2.4 Protective Relaying Protective relaying is designed to isolate any faulted component or portion of the electrical system, while maintaining continuity of power to the unfaulted portion of the system. The most commonly used protective devices include:

1. Undervoltage (27 Device) Relays. Undervoltage relays actuate on a low voltage condition, and usually are time delayed to account for momentary transient conditions, such as fault cleanng and bus transfers. The degraded grid voltage relays are provided with a substantially longer time delay to prevent actuation due to motor starting transients. Undervoltage relays provide a variety of protective functions including supply breaker trips and closure permissives, large motor breaker trips and closure permissives, and automatic starting of the Emergency Diesel Generators.

SD-50.1 Rev. 19 Page 27 of 131

4.2 Abnormal Operation 4.2.1 Abnormal Frequency Conditions When system frequency reaches 59.8 hertz. Annunciator, UA-06, window 1-2. GEN Bus UNDER FREO RELAY is activated.

Operators are directed to respond per AOP-22.0, Generator Abnormal Frequency Conditions. This is done to stabilize loads on the system. One of the most probable causes of an under frequency condition would be the loss of another large generating unit, or units, when the on-line reserve capacity is inadequate for current system loads. Rapid response and close coordination with the load dispatcher are required to ensure system stability.

Abnormal frequency operation can develop resonant frequencies that may induce vibrations in the low pressure turbine blades. The vibration can cause turbine blades to fatigue and possibly fail during operation. The effect increases proportionally in relation to the magnitude of the frequency difference, and the length of time at the abnormal frequency.

SD-27 Rev.15 Page51of127 GRID INSTABILITY OAOP-22.O Rev. 27 Page 5 of 14 3.0 AUTOMATIC ACTIONS

1. iF emergency bus voltage has lowered to less than 3700 volts (approximately equal to BOP bus voltage) for greater than 10 seconds, THEN the master/slave breakers to the E bus open and associated diesel generator starts and loads C

GRID INSTABILITY OAOP-22.0 Rev. 27 Page 6 of 14 4.2 Supplementary Actions NOTE A sudden rise in system frequency may be observed due to additional generation or load shedding. Automatic load shedding (10% of system load) occurs at each of the following frequencies: 59.3, 59.0, and 585 Hz. El CAUTION The maximum allowable time at a given frequency is as follows El

  • Below 58.1 Hz, operation is prohibited
  • Between 58.1-58.5 Hz, operation for 1 minute is allowed
  • Between 5&6-59.3 Hz, operation for 5 minutes is allowed
  • Between 59.4-60.6 Hz, operation is allowed indetinitely
  • Between 60.7-61.4 Hz, operation for 5 minutes is allowed
  • Between 61.5-61.9 Hz, operation fort minute is allowed
  • Above 61.9 Hz, operation is prohibited CAUTION
  • Off-frequency operation can stimulate resonance vibration in low pressure blades El
  • A total loss of off-site power (LOOP) should be anticipated if the turbine is tripped El
  • With grid voltage or frequency unstable or grid vulnerability identified, diesel generators should NOT be paralleled with any E bus connected to the grid since severe load swings may occur and possibly overload the diesel generators El
1. IF the maximum allowable time at a given frequency is exceeded, THEN perform the following:
a. jf reactor power is greater than or equal to 26%,

THEN insert a manual scram El

b. Trip the main turbine El
c. IF the unit was scrammed, THEN enter 1EOP-01-RSP(2EOP-01-RSP), Reactor Scram Procedure El

GRID INSTABILITY OAOP-220 Rev. 27 Page 9 of 14 4.2 Supplementary Actions (continued)

10. IF system frequency is high, THEN:
a. Establish communication with the Load Dispatcher C
b. Continue unit generation as directed by the Unit CR5 coordinate with the Load Dispatcher C c IF tripping the turbine becomes imminent THEN rapidly reduce power in an attempt to lower frequency to less than 601 Hz prior to tripping the main turbine C it lFnotifled by the Load Dispatcher system voltage is unable OR will be unable to support a LOCA, PR abnormal frequency conditions persist, THEN follow the guidelines in OOI-OLQ1, BNP Conduct of Operations Supplement C
12. IF any diesel generator is loaded to an E bus connected to the grid, THEN restore the diesel generator to standby in accordance with applicable procedures C
13. IF system voltage is less than 3700 volts for greater than 10 seconds, THEN ensure:
  • The affected E bus master/slave breakers OPEN C
  • The affected diesel generator starts and loads C
37. 295004 1 Which one of the following completes both statements below?

lAW OAOP-39.0, Loss of DC Power, before 125 VDC battery voltage reaches (1) remove loads as directed by the Unit CRS.

lAW 1 EOP-01 -SBO, Station Blackout, if either division battery chargers can NOT be restored within (2) then load strip the affected battery.

A. (1) 105 volts (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. (1) 105 volts (2) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C. (1) 129 volts (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> D. (1) 129 volts (2) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Answer: A K/A:

295004 Partial or Complete Loss of D.C. Power AK2 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following: (CFR: 41.7 I 45.8) 01 Battery charger ROISRO Rating: 3.1/3.1 Tier 1 / Group 1 K/A Match: This meets the K/A because this is testing knowledge of the relationship between the loss of DC power and time requirement to re-energize the battery charger.

Pedigree: New Objective: LOI-CLS-LP-051, Obj. 14 Describe the consequences/problems associated with the following: a. Battery chargers remaining out of service during a loss of off-site power I station blackout.

Reference:

None Cog Level: Fundamental Explanation: AOP-39.0 directs to load strip before teaching 105 VDC to prevent cell reversal. The alarm for undervoltage comes in at 129 VDC. The station Blackout procedure states that if the battery charger is not energized in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to load strip the batteries. There is a time critical 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> action in the SBO procedure for opening the Reactor Building doors.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because the first part is correct and the second part is a time critical action time limit in the SBO procedure.

Choice C: Plausible because 129 volts is the annunciator setpoint for the batteries and the second part is correct.

Choice D: Plausible because 129 volts is the annunciator setpoint for the batteries and the second part is a time critical action time limit in the SBO procedure.

SRO Basis: N/A LOSS OF DC POWER OAOP-39.0 Rev 41 Page 70136 6.2 Supplementary Actions Loss of Battery Chargers:

a. Monitor 125V and 24V DC battery voltages U
b. IF power has been removed from the battery chargers for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, THEN remove selected loads from the battery based on 001-50, 1251250 and 24/48 VDC Electrical Load List and Untt CRS direction U
c. Before 125V DC battery voltage reaches the low voltage limit of 105 volts, remove loads as directed by the Unit CRS as necessary to maintain battery voltage greater than 105 volts U
d. Before 24V battery voltage reaches the low voltage limit of 21 volts, remove loads as directed by the Unit CRS as necessary to maintain battery voltage greater than 21 volts U e IF battery charger AC power has been lost due to Station Blackout, THEN enter IEOP-fl 1-SB0(2EOP-01-SBO), Station Blackout U

(IF ral,1 tffyrgera Time Critic3T , CUOTbet 6r.acy ioaa OnthLt9a.

99 reJrml WtJlLl I n -

EOP1Otl IY P Time

/ WHEN ELAP II DElm1nTe1 ElOR ML fI be eimI a ania

\ THEN /

Defeat ICIC ai.qac Ixl çer (IF RCIClsjIavaza ThEN efeah-tPClsbNrtIogc J

reactor bLalcIng locra pf Eop-aI-o-C4. Time Serisilive stage aterne iEl poc waiieutsfay lime______________

e.fçiieat pacE P-()iCEP-12.

(If tIreodbyERQ THEN ccrEacrtpeT EOPl-O-l5.

38. 295005 1 Which one of the following identifies the reason an operator is directed to trip the main turbine as an immediate action lAW OAOP-32.0, Plant Shutdown From Outside Control Room?

A. To initiate a scram on TSV/TCV closure.

B. To prevent reverse power starts of the Diesel Generators.

C. The turbine cannot be tripped once the Control Room is evacuated.

D. To bring bypass valves into operation until Remote Shutdown Panel control is established.

Answer: B K/A:

295005 Main Turbine Generator Trip AK3 Knowledge of the reasons for the following responses as they apply to MAIN TURBINE GENERATOR TRIP: (CFR: 41.5/45.6) 04 Main generator trip RO/SRO Rating: 3.2/3.2 Tier I I Group 1 K/A Match: This question requires the operator to have knowledge of the reason for turbine/generator trip.

AOP-32 was used to include plausibility of distractors.

Pedigree: Bank Objective: LOl-CLS-LP-302E, Obj. 6 Given plant conditions and entry into OAOP-32.0, Plant Shutdown From Outside Control Room, explain the basis for a specific caution, note, or series of procedure steps.

Reference:

None Cog Level: Fundamental Explanation: Following a reactor scram, the turbine control valves throttle shut in an effort to control RPV pressure at the setpoint of 928 psig. Without operator action, the turbine control valves will fully close, causing the generator to motor. Reverse power on the generator will cause a generator primary lockout and auto start of the diesel generators. The main turbine is therefore manually tripped to prevent it from automatically tripping on generator reverse power. This also reduces the number of cold start demands on the diesel generators.

Distractor Analysis:

Choice A: Plausible because a reactor scram is inserted as a step in the AOP, but it is performed earlier.

Choice B: Correct Answer, see explanation Choice C: Plausible because the procedure states to perform the step prior to exiting the control room but it could still be done at the turbine front standard.

Choice D: Plausible because this would allow use of the bypass valves, but MSIVs are manually closed prior to leaving the control room. This brings SRVs into operation. If MSIVs are not closed prior to leaving the control room, RPS EPA breakers are opened prior to establishing control at Remote Shutdown panel, which would close MSIVs.

SRO Basis: N/A

REACTOR SCRAM PROCEDURE I 001-37.3 I BASIS DOCUMENT I Rev. 016 I Page 7of23 5.2 Step RSP-2 I

Perform crcim mmIiaLo actions.

I RSP2 Step RSP-2 includes the potential for multiple sensor and sensor relay failures in the automatic RPS logic where an automatic reactor scram should have initiated but did not. If needed a manual scram is inserted to accomplish an automatic action v.hich should have taken place. A manual reactor scram is also required when directed trom other EOP5 and no condition exists which would have automatically initiated a reactor scram (e.g., entry from PCCP because of high torus temperature).

Step RSP-2 also addresses other Reactor Operator scram immediate actions and includes:

  • ARI initiation is an additional means of inserting controt rods if needed.
  • Placing the reactor mode switch to shutdown. When the reactor mode switch is placed in SHUTDOWN position, a diverse and redundant reactor scram signal is generated by the RPS logic. If the mode switch is taken out of RUN prior to RPV pressure decreasing to 835 psig, The MSIV closure due to low main steam line pressure is prevented.

For Unit 2 only, if the mode switch is taken out of RUN when steam flow is above 33%, the MSIVs will close. Therefore, for Unit 2 the mode switch is placed in SHUTDOWN after steam flow is below 3xi0 lb/hr.

  • Following a reactor scram, the turbine control valves throttle shut in an effort to control RPV pressure at the setpoint of 928 psig. Without operator action, the turbine control valves will fully close, causing the generator to motor. Reverse power on the generator will cause a generator primary lockout and auto start of the diesel generators. The main turbine is therefore manually rnpped to prevent it from automatically tripping on generator reverse power. This also reduces the number of cold start demands on the diesel generators.
39. 295006 1 Unit One has entered RSP with the following conditions:

Six control rods are at position 02, all others are fully inserted B Recirc Pump has tripped Which one of the following completes both statements below?

The control rods will be inserted by (1) lAW OEOP-01-LEP-02, Alternate Control Rod Insertion.

After the control rods are inserted, a CRD flow rate of approximately (2) will be established.

A. (1) placing the individual scram test switches to the Scram position (2) 30 gpm B. (1) placing the individual scram test switches to the Scram position (2) 45 gpm C. (1) driving rods using RMCS (2) 30 gpm D. (1) driving rods using RMCS (2) 45 gpm Answer: C K/A:

295006 Scram AA1 Ability to operate and/or monitor the following as they apply to SCRAM: (CFR: 41.7 / 45.6) 06 CRD hydraulic system RO/SRO Rating: 3.5/3.6 Tier 1 / Group I K/A Match: This meets the K/A because this is testing operation of CRD controls after a scram.

Pedigree: new Objective: LOl-CLS-LP-300-C, Obj. 10 Given plant conditions and the Reactor Scram Procedure, determine the required operator actions

Reference:

None Cog Level: High Explanation: Even if the reactor will remain shutdown under all conditions without boron the LEP is used to insert the control rods using RMCS. If more control rods were out then the scram test switches would be an option. If a recirc pump is tripped then CRD flow is set to 30 gpm to minimize the stratification in the bottom head region.

Distractor Analysis:

Choice A: Plausible because this is an option used to insert the control rods in the LEP. The second part is correct.

Choice B: Plausible because this is an option used to insert the control rods in the LEP. The second part is the nominal setting for the CRD flowrate.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because the first part is correct and the second part is the nominal setting for the CRD flowrate.

SRO Basis: N/A

7. WHEN either:
  • Reactor engineering has determined the reactor wilt remain shutdown under all conditions without boron Li RD THEN perform Section 22, Control Rod Insertion Verification on Page 7 Li RD
10. lFy control rod NOT fully inserted, THEN insert control rods:
a. Record in Control Room log the control rod number and position ofy rods NOT fully inserted RD ALTERNATE CONTROL ROD INSERTION . OEOP-01-LEP-02 Rev. 029 Page 120137 2.2a Control Rod Verification Actions (continued)
b. Bypass RWM RO
c. Insert control rods with Emergency Rod In Notch Override switch RO

ALTERNATE CONTROL ROD INSERTION DEOP-.01-LEP-02 Rev. 029 Page 24 0137 2.6.3 Scram Individual Control Rods Actions (continued)

JO. Unit 1 Only: Insert control rods with individual scram test switches:

a. Identify control rod NOT inserted to or beyond Position DO C RO NOTE
  • A sound powered phone jack is located on the column beside Panel XU-76 and in Panels XU-12, 58, 49 and 61 U
  • The preferred sound-powered phone switchboard bus for use is Bus 1 U
b. Establish communication beten Panel P610 and Control Room U RO NOTE The individual scram test switch SCRAM position Is dowii U
c. Place individual scram test switch to SCRAM position for py control rod NOT inserted to or beyond Position 00 U RO

ALTERNATE CONTROL ROD INSERTION OEOP-Q1-LEP-02 Rev. 029 Pageilof37 22.3 Control Rod Verification Actions (continued)

(3) IF CRD pumps running, THEN stop one CRD pump RO (4) Set the setpoint tape on Ci ifCl2)-FC-R600 (CR0 flow Control) to 30 gpm RO NOTE The actions in Section 2.2.3 Step 9.h(5) may be repeated as necessary D (5) Adjust cooling water differential pressure, CRD flow rate and drhe pressure:

  • Cl 1(C12)-FC-R600 (CRD Flow Control) to maintain cooling water diflerential pressure between 10 and 26 psid RO
  • jf a reactor recirwiation pump is tripped, THEN establish a CRD flow rate ot approximately 30 gpm RO
40. 295009 1 A total loss of Unit One feedwater results in reactor water level lowering to 87 inches.

Drywell pressure is 2.1 psig.

Reactor water level is being restored with RCIC and CRD.

Which one of the following completes both statements below?

RVCP (1) requiredtobeentered.

The expected response of the G31-F001, Inboard RWCU Isolation Valve, and the G31-F004, Outboard RWCU Isolation Valve, is that (2) should be closed.

A. (1) is (2) ONLY the G31-F004 B. (1) is (2) BOTH C. (1) is NOT (2) ONLY the G31-F004 D. (1) is NOT (2) BOTH Answer: B K/A:

295009 Low Reactor Water Level AK2 Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following: (CFR:

41.7 /45.8) 04 Reactor water cleanup RO/SRO Rating: 2.6/2.6 Tier I / Group 2 K/A Match: This meets the K/A because this is testing the LL2 relationship to Group 3 (RWCU) isolation.

Pedigree: New Objective: LOl-CLS-LP-014, Obj 8 Given plant conditions, determine if the RWCU system should have isolated, including expected changes in RWCU System components

Reference:

None Cog Level: High Explanation: Based on conditions RVCP should be entered. By knowing the entry conditions for RVCP (2# DW pressure) this eliminates the RSP. The low level condition will isolate the FOOl and F004. There are some signals that will isolate only the F004 only.

Distractor Analysis:

Choice A: Plausible because the first part is correct and some of the Group 3 signals do only close the F004.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because the RSP would be entered, but there is an entry condition for RVCP (2# in the DW). Some of the Group 3 signals do only close the F004.

Choice D: Plausible because the RSP would be entered, but there is an entry condition for RVCP (2# in the DW). The second part is correct.

SRO Basis: N/A USERS GUIDE OEOP-Oi-UG Rev. 067 Page 5201156 ATTACHMENT 1 Page 5 of 15

<<Group Isolation Checklist>>

Group 3 Isolation Sianals Signal Tech Spec Value Setpoint Value LowLevel2 +101 inches +105 inches High Differential Flow 73 gpm 43 gpm (after 285 minute time delay)

Area High Temperature 150°F 140°F Area Ventilation T High 50°F 47°F Nan-Regen Hx Outlet N/A 135°F Temp Hi SLC Initiation N/A N/A RWCU Outside PumplHx 120°F 115°F Rms RWCU Differential How 30 minutes 28.5 minutes High Time Delay Group 3 Isolation Valves Control Room RTGB Panel H12-P607 Valve Number Power Supply Normal Unit 1(Unit 2) Position Fail Position Checked

[Note 1] G31-F00i 1XC(2XC)IE1(E3) NO [Note 2] FAI G3J-F004 1XDB(2XDB) NO FAI

[DCI Note 1: SLC Initiation and RWCU Non-Regen Fix Outlet Temperature Hi signals do NOT isolate the RWCU Inlet Inboard Isolation Valve, G31-F001.

41. 2950161 CAUTION There are seven ke1ock NORMAULOCAL switches located on Diesel Generator 2 control panel. Sbc of these are located in a row. The seventh switch s located in the row above the six switches Which one of the following completes both statements below concerning the caution above from OASSD-02, Control Building?

The six switches in a row must be placed in LOCAL (1) placing the seventh switch in LOCAL.

The purpose of this sequence is to prevent a loss of DG2 due to a loss of the redundant power supply fuses for the (2) circuitry.

A. (1) before (2) output breaker B. (1) before (2) engine run control C. (1) after (2) output breaker D. (1) after (2) engine run control Answer: B K/A:

295016 Control Room Abandonment AK3 Knowledge of the reasons for the following responses as they apply to CONTROL ROOM ABANDONMENT: (CFR: 41.5/45.6) 03 Disabling control room controls RO/SRO Rating: 3.5/3.7 Tier 1 I Group 1 K/A Match: This meets the K/A because the six local switches remove control room controls and the seventh switch supplies an alternate power supply to the equipment.

Pedigree: Bank Objective: LOl-CLS-LP-304, Obj. 21 Explain why the Diesel Generator NORMAL/LOCAL switches must be placed in LOCAL in a particular sequence.

Reference:

None Cog Level: Fundamental

Explanation; The six switches in a row isolate DG2 engine and generator control circuitry from the control room (the fire area) since a fire induced fault in wiring in the fire area may result in loss of the DG. The seventh switch inserts redundant control power fuses to the circuitry that has been isolated in the event a fault has already resulted in blowing the normal fuses. This seventh switch must be turned last with the potentially faulted circuitry already isolated or the alternate fuses may also blow making the DG unavailable. The DG engine lockout is already tripped if the DG had been running since the operator is directed to trip the DG using emergency stop.

Of the first six switches, they include; Diesel START/STOP (2 switches) Diesel Governor (2 switches) Generator Voltage Regulation (2 switches)

Distractor Analysis; Choice A; Plausible because the six switches are placed in local first and the output breaker does have redundant control power fuses.

Choice B; Correct Answer, see explanation.

Choice C; Plausible because this is the opposite of the correct sequence and the output breaker does have redundant control power fuses.

Choice D: Plausible because this is the opposite of the correct sequence and the second part is correct.

SRO Basis: N/A

SD-39 I EMERGENCY DIESEL GENERATORS Rev. 20 I SYSTEM DESCRIPTiON Page 39 of 166 The Governor Control At Setpoint indicator light provides a status or the DRU speed reference for the 23CM governor The light is an indicator that the Governor Control System is ready to operate at the Setpoint speed. During actual operation of the DG, the Governor Control At Setpoint indicator light may or may not be illuminated depending on the speed of the DG.

Voltage Adjust Switches Two three position (RAISE-NEUT-LOWER) spring return to NEUT switches are provided per engine to permit the adjustment of voltage regulators from the local panel regardless of EDG mode of operation.

The auto adjust switch is normally used.

ASSD Koylock Switches Brass handled two-position NORM LOCAL ASSD keylock switches on the local engine panels permit the operator to transfer control of the engine and generator to the local control panel. ASSD operations are performed when a fire exists in the plant and components required to be operated may be damaged by the fire.

These switches isolate control room controls and indications to isolate the EDG control circuitry from potential fire induced faults. There are six ASSD switches (2 for EDG mnlstop controls, 2 for governor controls, and 2 for voltage regulation controls) located on each local EDG panel. When in the IASSD!I mode, operation of the Diesel engine can only be accomplished by the LOCAL EMERGENCY STOP and LOCA1 EMERGENCY START pushbuttons.

In addition to the six ASSD switches, for EDG 2 and 4 only, there is a seventh ASSD switch located above the other six switches. This switch provides an alternate set of control power fuses for EDG control drwthy. This may be necessary since fire induced faults may have blown normal control fuses. When operating the ASSD switches for EDGs 2 or4, The seventh switch must be turned last after the potentially faulted circuitry has been isolated to prevent blowing the alternate fuses, making the EDG unavailable to provide paver to Safe Shutdown loads.

42. 295017 1 During accident conditions, the source term from the Unit One Reactor Building must be estimated. Three RB HVAC supply fans and three RB HVAC exhaust fans are running.

lAW OPEP-03.6.1, Release Estimates Based on Stack/Vent Readings, which one of the following is the calculated release rate?

ATTACHMENT 2 Page 1 of I Source Term Calculation From #1 RX Gas tl-CAC-AQH-12644)

METER FLOW1 EFFICIENCY12 RELEASE3 READING fcfrn) FACTOR RATE TIME (cpm) (iCi1sec) 43,200 CFM per 1 minute ago 4.0 E+3 1.275 E-5 e,jaijst tan LU If not available use 43,200 cftn per exhaust tan times the number of fans operating.

The efficiency factors can be obtained from OE&RC-2020 (contact E&RC counting room).

(3j (cpm) x (cfrn) x (Efficiency Factor)

Release Rate A. 2.2 E+3 pCi/sec.

B. 6.6 E+3 pCi/sec.

C. 1.3 E+4 p.Ci/sec.

D. 6.6 E+4 liCi/sec.

Answer: B K/A:

295017 High Off-Site Release Rate AA2 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.10/43.5/45.13) 03 Radiation levels RO/SRO Rating: 3.1/3.9 Tier 1 / Group 2 K/A Match: This meets the K/A because it is testing the source term for a release off-site.

Pedigree: Bank Objective: LOl-CLS-LP-301A, Obj. 6 Determine data required for offsite dose projection in accordance with AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment, and PEP-03.6.1, Release Estimates Based Upon StackNent Readings.

Reference:

None

Cog Level: High Explanation: Per Attachment 2 the calculated release rate is:

Meter reading (CPM) X Flow (43,200 per fan X no of discharge fans) X efficiency factor or (4 E+3) (43,200 X 3) (1.275 E-5) = 6.6 E+3 mCi/sec Distractor Analysis:

Choice A: Plausible because it is the calculation without multiplying times the number of running exhaust fans.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because it uses the total number of fans running vs. the number of exhaust fans.

Choice D: Plausible because it is the correct numerical value but is off by a factor of 10.

SRO Basis: N/A

43. 2950181 Unit Two is operating at 65% power when the following are I 0 I 0 V 100 ccw RBCCW PUMP HEAD TANK DISCH HEADER PUMP MOTOR PRESS LOW TEl1P HI -80 LEVEL Hi/LO UA-3 UA-3 A-6 (In Alarm) (In Alarm) (In Alarm) E-G0 E-40

-20 Z()

R8CCW DISCWR GE PRESSURE X-P-1-)

Which one of the following completes both statements below lAW OAOP-1 6.0, RBCCW System Failure?

A complete loss of RBCCW (1) occurred.

Areactorscram (2) required.

A. (1) has (2) is B. (1) has (2) is NOT C. (1) has NOT (2) is D. (1) has NOT (2) is NOT Answer: A K/A:

295018 Partial or Complete Loss of Component Cooling Water AK2 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER and the following: (CFR: 41.7 /45.8) 02 Plant operations RO/SRO Rating: 3.4/3.6

Tier 1 / Group I K/A Match: This meets the K/A because it is testing the relationship of the loss of RBCCW and the actions required for plant operations Pedigree: Last used on the 04 NRC Exam Objective: LOl-CLS-LP-302-H, Obj. 4a Given plant conditions, determine the required supplementary actions in accordance with the following AOPs: OAOP-16.0, RBCCW System Failure

Reference:

None Cog Level: High Explanation: A complete loss of RBCCW is defined as discharge header pressure below 60 psig and all available RBCCW pumps running (AOP-16.0). A complete loss requires a manual scram.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because the first part is correct and for a loss of TBCCW only a performing power reduction is allowed.

Choice C: Plausible because the pumps are running and system pressure is available and a reactor scram could be thought of to be needed due to alarms.

Choice D: Plausible because the pumps are running and system pressure is available and for a loss of TBCCW only a performing power reduction is allowed.

SRO Basis: N/A

RBCCW SYSTEM FAILURE OAOP-16.O Rev. 31 Page 9 of 18 4.2 Supplementary Actions (continued)

b. IF ADHR Mode piping is NOT the source of the leakage.

THEN re-align RBCW Pumps A and D from ADHR Mode to RBCCW Mode, as necessary LI NOTE A complete loss of RBCCW is defined as discharge header pressure less than 60 psig, high temperature alarms on components supplied by RBCCW, and all available (no more than three) RBCCW pumps operating on the RBCW header LI

4. IF there is a complete toss of RBCCW, THEN LI
a. Trip all RBCCW pumps (including RBCCW Drywell HVAC Cooling Pump if operating on the affected unit and pumps operating in ADHR Mode) LI
b. Close the following valves:
d. Isolate RWCU System by closing the following valves:
  • G31-FO01 (RWCU Inboard 1501 VIv) LI
  • G31 -E004 (RWCU Outboard isol VIv) LI e Reduce reactor power with recirc flow in accordance with OENP-24.5, Form 2, Immediate Reactor Power Reduction Instructions LI
f. Insert a manual scram LI
9. Enter 1 EOP-01 -RSP(OP-O1 -RSP), Reactor Scram Procedure, AND perform concurrently with this procedure LI
h. Trip both reactor recirculation pumps by performing the following:

(1) Depress VED A Emerg Stop LI (2) Depress VFD B Emerg Stop LI

44. 295019 1 Unit Two has entered OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures, due to a loss of instrument air pressure with the following annunciator status:

UA-01 (1-1) RB Instr Air Receiver 2A Press Low Alarm sealed in UA-01 (1-2) RB InstrAir Receiver 28 Press Low NOT in Alarm UA-01 (3-2) Air Compr 0 Trip Alarm sealed in UA-01 (4-4) Inst Air Press Low Alarm sealed in UA-01 (5-4) Service Air Press-Low Alarm sealed in Which one of the following completes both statements below?

On a loss of instrument air, the RB HVAC Butterfly Isolation Valves will fail (1) lAW OAOP-20.0, the reactor (2) required to be scrammed.

A. (1) as-is (2) is B. (1) as-is (2) is NOT C. (1) open (2) is D. (1) open (2) is NOT Answer: B K/A:

295019 Partial or Complete Loss of Instrument Air AA2 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: (CFR: 41.10 / 43.5 / 45.13) 02 Status of safety-related instrument air system loads RO/SRO Rating: 3.6/3.7 Tier 1 /Group 1 K/A Match: This meets the K/A because it is testing status of equipment on a loss of air and the action that is required from the AOP Pedigree: New Objective: LOI-CLS-LP-302K, Objective 6 Summarize the consequences associated with improper equipment operation specified in OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures

Reference:

None Cog Level: Fundamental Explanation: A loss of instrument air the BFIVs fail as-is. Other equipment will fail open or closed. A reactor scram is required if unable to maintain at least one division non-interruptible instrument air pressure greater than 95 psig.

Distractor Analysis:

Choice A: Plausible because the first part is correct and a scram is required if both divisions are pressure cannot be maintained. The BFIVs are closed if either divisions air pressure cannot be maintained.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because air operated valves can be designed to fail open, closed or as-is. Part 2 because a reactor scram is required if unable to maintain at least one division non-interruptible instrument air pressure greater than 95 psig.

Choice D: Plausible because air operated valves can be designed to fail open, closed or as-is. Part 2 is correct.

SRO Basis: N/A PNEUMATIC (AIR/NITROGEN) SYSTEM FAILURES OAOP-200 Rev. 46 Page 70128 4.0 OPERATOR ACTiONS NOTE The following should be considered for establishment as critical parameters during pertormance of this pro cedure LI

  • Instrument alt pressure
  • Condensate and Feedwater System minimum flow valve status
1. IF any of the following conditions exist LI
  • Unable to maintain at feast one division non-interruptible instrument air pressure greater than 95 psig LI
  • Unable to maintain at least one division dryweti pneumatic pressure greater than 95 psig LI
  • Instrumentation indicates unsafe reactor operation LI THEN:
a. Insert a manual scram LI
b. Enter 1 EOP-O1-RSP(QP-O1-RSP), Reactor Scram Procedure and perform concurrently with this procedure LI PNEUMATIC (AIR/NITROGEN) SYSTEM FAILURES OAOP-20.O Rev. 46 Page 10 0128

a IF UA-O1 1-1, RB InsW Air Receiver 1A(2A) Press Low OR UA-Q1 1-2, RB lnstr Air Receiver 1B(2B) Press Low, alarm is received, THEN perform the following:

NOTE Isolation of the reactor building supply and exhaust valves renders the building ventilation system INOPERABLE Standby Gas Treatment System operation may be required to maintain reactor building negative pressure 0 (1) if required to maintain reactor building negative pressure, THEN start the Standby Gas Treatment System in accordance with lOP-i D(20P-1 0), Standby Gas Treatment System 0 (2) Close the following valves:

Unit I Only:

  • 1B-BFIV-RB and ID-BFIV-RB (RB Vent Outbd Valves) 0 Unit 2 Only:
45. 295020 1 l&C Techs inadvertently cause a low level 3 (LL3) signal.

Unit Two plant conditions are:

Reactor pressure 930 psig Drywell pressure 1.7 psig, steady Drywell temp (average) 140°F, slow rise Drywell leak calculation Normal Which one of the following completes the statement below?

All Drywell Cooler Fans are:

A. tripped, but can be overridden on.

B. tripped, and cannot be overridden on.

C. running, but can be tripped at the RTGB.

D. running, and cannot be tripped at the RTGB.

Answer: A K/A:

295020 Inadvertent Containment Isolation AAJ Ability to operate and/or monitor the following as they apply to INADVERTENT CONTAINMENT ISOLATION: (CFR: 41.7 / 45.6) 02 Drywell ventilation/cooling system ROISRO Rating: 3.2/3.2 Tier 1 I Group 2 K/A Match: This meets the K/A because it is testing what the DW coolers do on an isolation signal.

Pedigree: Bank Objective: LOl-CLS-LP-04, Obj. 20 Given plant conditions determine if the drywell coolers should auto start or trip

Reference:

None Cog Level: High Explanation: LOCA signal on LL3 closes Group 10 which fails dampers open, but also trips fan motors.

Override for LOCA trip can be performed as long as a LOCA does not really exist which is overridden in back panels (XU-271XU-28). The low level condition also is a scram signal which provides an auto start signal for the DW Coolers which is prioritized by the trip signal.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because the fans do trip and if the conditions were different they would not be able to be overridden.

Choice C: Plausible because the fans do auto start on a scram signal or usually when the dampers are opened and under different conditions they would be able to be tripped from the RTGB.

Choice D: Plausible because the fans do auto start on a scram signal or usually when the dampers are opened and under different conditions they would not be able to be tripped from the RTGB.

SRO Basis: N/A Placing a Unit 2 DrW/elI Cooling Fan control switch in START causes the tans discharge damper to open. WI-lEN the discharge damper is full open, the fan Will start. The control switch should be held in the START position until the discharge damper is full open. The RBCW cooling water valve to the coils will open concurrently with a fan start.

Placing the Drywell Cooling lB Fan control switch in START causes the fans discharge damper to open. WHEN the discharge damper is full open, the fan will start. The control switch should be held in the START position until the fan starts. The common air inlet damper and The RBCCW cooling water valve to the coils will open concurrently with a fan start.

WHEN the control switch for Drywell Cooling Fan 1A, 1C, or 10 is placed in START, the associated fan starts and the discharge backdraff damper opens from the fan air flow. The discharge damper position indication does not input to the start logic for Drywell Cooling Fans IA, 1 C, and 1 D. The RBCCW cooling water valve to the coils will open concurrently with a fan start.

The Drywell Lower Vent dampers can be positioned to either MIN or MAX position by a two-position control switch on Panel XU-3. Normal plant operating position for these dampers is the MIN position. Placing these dampers to MAX position during plant operation may produce extreme temperature excursions in the upper dryvell regions. Low scram air header pressure will reposition these dampers to the MAX position and automatically start any idle drywell cooling fan selected for AUTO.

Drywell Cooler Override Switches, VA-CS-5993!5994, are provided in Panels XU-27128 to facilitate various modes of Diyweil cooler operation as required by the EOP5.

Tile Pneumatic Nitrogen System or Reactor Building Non-lntermptible Instrument Air pneumatically operates the drywell cooling fans discharge dampers. These dampers will fail open on loss of pneumatics. Unit 2 and lB drywell cooling fans discharge dampers fail closed on loss of the associated 120 VAC distribution panel.

A contactor in the associated fans 480 VAC breaker provides drywell cooler FAN ON indication on RTGB Panel XU-3.

SD-04 Rev. 9 Page 17 of 103 The drywell coolers receive a LOCA trip signal from the Core Spray initiation relays.

46. 295021 1 Unit One in MODE 5.

The fuel pool gates are removed.

SDC Loop B is in service.

Fuel pool cooling assist is in operation.

The RHR Loop B pumps tripped and can NOT be restarted.

Which one of the following completes both statements below?

(consider each statement separately)

Fuel pool cooling assist (I)

Fuel pool cooling assist (2) capable of being aligned to the SDC Loop A lAW 1 Op-I 7, Residual Heat Removal System Operating Procedure.

A. (1) remains in service (2) is B. (1) remains in service (2) is NOT C. (1) is lost (2) is D. (1) is lost (2) is NOT Answer: D K/A:

295021 Loss of Shutdown Cooling AK2 Knowledge of the interrelations between LOSS OF SHUTDOWN COOLING and the following:

(CFR: 41.7 / 45.8) 05 Fuel pool cooling and cleanup system RO/SRO Rating: 2.7/2.8 Tier 1 I Group 1 K/A Match: This meets the K/A because it is testing the relationship of using SDC and the Fuel Pool.

Pedigree: New Objective: LOI-CLS-LP-017, Obj 5 Given a drawing of the RHR system, trace the flow path for all of the six (6) modes of operation.

Reference:

None Cog Level: High Explanation: Fuel pool cooling assist mode utilizes the B Loop of RHR so that when it is lost so too will the fuel pool cooling assist operations. If the gates were installed then the A Loop of SDC could be used with the B loop discharge flowpath, but with the gates removed this is NOT an option.

Distractor Analysis:

Choice A: Plausible because the students may think that the FPC pumps provide the motive force for this mode of operation and if the gates were installed then this would be correct.

Choice B: Plausible because the students may think that the FPC pumps provide the motive force for this mode of operation and the second part is correct.

Choice C: Plausible because the first part is correct and if the gates were installed then this would be correct.

Choice D: Correct Answer, see explanation SRO Basis: N/A 8.11 Fuel Pool Cooling Assist Mode With Fuel Pool Gates Removed CAUTION The following section has the potential to significantly raise area dose rate&

8.11.1 Initial Conditions Datefrime Started Initials

1. Reactor in Mode 5 with fuel pool gates removed.
2. Fuel pool temperature can NOT be maintained less than 125°F.
3. OPT-08.OC has been completed satisfactorily within previous 92 days.
4. Fuel Pool Cooling system in operation in accordance with 1OP-13 with available fuel pool cooling heat exchangers in operation.
5. RHR Loop B is operating iti shutdown cooling in accordance with Section 5.7 or 5.8.
47. 295023 1 Unit Two is performing refueling operations when the refueling SRO reports that a spent fuel bundle has been dropped.

The following radiation monitoring alarms are received:

UA-03 (3-7) Area Rad Refuel Floor High UA-03 (4-5) Process Rx Bldg Vent Rad Hi Which one of the following identifies the Immediate Action that is requited lAW OAOP-05.0, Radioactive Spills, High Radiation, and Airborne Activity?

A. Verify Group 6 isolation.

B. Evacuate all personnel from the refuel floor.

C. Place Control Room Emergency Ventilation System in operation.

D. Isolate Reactor Building Ventilation and place Standby Gas Treatment trains in operation.

Answer: C K/A:

295023 Refueling Accidents AA1 Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS:

(CFR: 41.7 / 45.6) 04 Radiation monitoring equipment RO/SRO Rating: 3.4/3.7 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing the immediate operator actions for a radiation event.

Pedigree: Bank Objective: LOl-CLS-LP-302J, Obj. 5 List the immediate operator actions required to be performed in accordance with OAOP-05, Radioactive Spills, High Radiation, and Airborne Activity

Reference:

None Cog Level: Fundamental Explanation: This is an Immediate Action identified in AOP-05.0.

Distractor Analysis:

Choice A: Plausible because this is an auto action not an immediate operator action of the AOP Choice B: Plausible because lAW OAOP-05.0 this is the first supplemental action.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because RBHVAC isolation and SBGT start requires PROCESS RXBLDG VENT RAD HI-HI fUA-03 3-5) in alarm and these are supplementary actions in the procedure.

SRO Basis: N/A

RADIOACTIVE SPILLS, HIGH RADIATION, AND OAOP05.O AIRBORNE ACTIV1W Rev 32 Page 5 of 15 4.0 OPERATOR ACTIONS NOTE The following should be considered for estabftshment as critical parameters during perrormance of this procedure I]

  • Area radiation levels
  • Personnel habitability in the affected area 4.1 Immediate Actions
1. IF a fuel assembly was dropped or damaged, THEN ensure the Control Room Emergency VenUlatton System (CREVS) is in Operation. {71 1} C]
48. 295024 1 Unit Two is operating at rated power when high drywell pressure switch C72-PTM-NOO2A-1 fails high resulting in the annunciation of A-05-(5-6) Pri Ctmt Press Hi Trip.

Which one of the following completes the statement below?

RPS high drywell pressure relay C72-K4A will (1)

The RSP (2) be required to be entered.

A. (1) energize (2) will B. (1) energize (2) will NOT C. (1) de-energize (2) will D. (1) de-energize (2) will NOT Answer: D K/A:

295024 High Drywell Pressure EA1 Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE:

(CFR: 41.7 / 45.6) 05 RPS RO/SRO Rating: 3.9/4.0 Tier 1 I Group 1 K/A Match: This meets the K/A because it is testing the ability to monitor RPS (half scram condition) for a high DW pressure condition Pedigree: New Objective: LOl-CLS-LP-003, Objectives:

7.g Given plant conditions state the Normal, Initiation, and Fail position/condition of the following components: (Open/Closed Energized/De Energized) RPS Logic

9. Given any scram signal, describe the logic arrangement for the signal including what combination of signals will cause a Full Scram.

Reference:

None Cog Level: High Explanation: The RPS relays are de-energize to actuate and a single relay actuates the alarm and will cause a half scram.

Distractor Analysis:

Choice A: Plausible because there are logics that are energize to actuate and there are also logics that only require one instrument to actuate (Nuclear instrumentation).

Choice B: Plausible because there are logics that are energize to actuate and the half scram is the result.

Choice C: Plausible because the first part is correct and some logics do cause a full scram (Nuclear instrumentation).

Choice D: Correct Answer, see explanation.

SRO Basis: N/A 1

APP AD E Page 1 cf 2 PRI CT1T PRESS HI TRIP AJTCMATIC ACTIONS

1. If the primary ccntair zit pressure high trip signal is recivd in only one P.PS Trip Systezi, a half Scran will occur.
2. If the primary c:ntainclent pressure high trip sgnal is received in both RPS Trip Systems, a react:r Scram will occur.

DEVICE! SETPOINTS Relay C72K4A eenezgized Pressure Switch C71FTMNclA1, Bi, Cl, or 21 17 p5lg

FIGURE 03-15 High Drywell Pressure Trip TRIP CHANNEL A1 TRIP CHANNEL A2

- LC71f2)-PTM C71(2)-PTM -

NOO2C-1 NOO2A1 PANEL PANEL XU65 XU6 K4A >K4C NOTE PRESSURE SWITCH CONTACTS OPEN ON HIGH DRYWELL PRESSURE CONDITION TRIP CHANNEL Hi TRIP CHANNEL B2

C72(2)-PTM  : C71(2)-PTM N002S-1 NOG2D-1

>K4D K45 SD-03 Rev. 12 Page73ot9Q

49. 295025 1 Unit One was operating at power when a turbine trip occurred.

85 control rods fail to insert.

Reactor pressure peaks at 1145 psig.

Which one of the following completes both statements below?

The reactor recirc pumps (1) tripped.

Tripping of the reactor recirc pumps results in a rapid decrease in reactor power due to (2)

A. (1) must be manually (2) voiding of the moderator B. (1) must be manually (2) a reduction in reactor water level C. (1) have automatically (2) voiding of the moderator D. (1) have automatically (2) a reduction in reactor water level Answer: C K/A:

295025 High Reactor Pressure EK3 Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE: (CFR: 41.5 /45.6) 02 Recirculation pump trip RO/SRO Rating: 3.9/4.1 Tier 1 /Group 1 K/A Match: This meets the K/A because it is testing the reason the recirc pump is tripped.

Pedigree: Bank Objective: LOl-CLS-LP-002, Obj. 30 Given Plant conditions determine if the ATWS-RPT protection logic should have actuated

Reference:

None Cog Level: Fundamental Explanation: The Anticipated Transient Without Scram circuit provides an alternate means of reducing reactor power in the unlikely event that the control rods fail to insert into the core following a Reactor Protection System actuation signal. Tripping of the VFD Input Circuit Breakers (ICE) will rapidly reduce recirculation flow. This results in a rapid decrease in reactor power because of the voiding of the moderator. Setpoints for ATWS trip are high reactor pressure 1137.8 psig and low reactor level LL2 105

Distractor Analysis:

Choice A: Plausible because the ATWS procedure directs the pumps to be tripped and the second part is correct.

Choice B: Plausible because the AIWS procedure directs the pumps to be tripped and level is reduced in the ATWS procedure which does lower power.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because the first part is correct and level is reduced in the ATWS procedure which does lower power.

SRO Basis: N/A 3.2.6 Anticipated Transient Without Scram Recirculation Pump Trip (ATiNS-RPT)

The Anticipated Transient Without Scram circuit provides an alternate means of reducing reactor power in the unlikely event that the control rods fait to insert into the core following a Reactor Protection System actuation signaL Tripping of the VFD Input Circuit Breakers fICB)witt rapidly reduce recirculaion flow. This results in a rapid decrease in reactor power because ot the voiding of the moderator.

Two signals ate used for the initiatk)n of ATWS-RPT. These signals are LL2 reactor vessel weter level and high reactor vessel pressure Each of these parameters is monitored by four sensors. Two level or pressure instruments in one of tv.o logic trains are required to energize relays which trip both Recircutation Pumps.

SD-OZ1 Rev 0 Page 72 of 182

50. 295026 1 Unit One failed to scram following a loss of off-site power with the following plant conditions:

Reactor Power 5%

RPV Water Level -55 inches (N036)

RPV Pressure 850 psig Which one of the following completes both statements below?

This UA-12 (5-4) alarm is expected to be received when suppression pool water temperature first reaches (1) lAW I OP-I 7, Residual Heat Removal System Operating Procedure, the RHR logic requirements to place torus cooling in service under the current plant conditions will require (2)

A. (1) 95°F (2) placing the CS-Si 7B Think Switch to Manual first and then bypassing the 2/3rd core height interlock B. (1) 95°F (2) bypassing the 2/3rd core height interlock first and then placing the CS-S17B Think Switch to Manual C. (1) 105°F (2) placing the CS-Si 7B Think Switch to Manual first and then bypassing the 2/3rd core height interlock D. (1) 105°F (2) bypassing the 2/3rd core height interlock first and then placing the CS-S17B Think Switch to Manual Answer: B K/A:

295026 Suppression Pool High Water Temperature G2.4.5oAbility to verify system alarm setpoints and operate controls identified in the alarm response manual. fCFR: 41.10/43.5 /45.3)

ROISRO Rating: 4.2/4.0 Tier 1 I Group 1 K/A Match: This meets the K/A because it is testing when the torus temperature alarm setpoint and what controls need to be operated to establish cooling.

Pedigree: New

Objective: LOI-CLS-LP-017, Obj 09 Given an RHR pump or valve, list the interlocks, permissives and/or automatic actions associated with the RHR pump or valve, including setpoints.

Reference:

None Cog Level: High Explanation: LOCA signal is sealed in due to being less than LL3 (45 inches) RPV water level is less than 2/3rd core height (-47 inches) therefore the keylock switch and then the Think switch is required (sequencing is essential). When the torus reaches 95°F this alarm will come in, 105°F is the TMax alarm.

Distractor Analysis:

Choice A: Plausible because the first part is correct and the second part is opposite of the required actions.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because this is the alarm setpoint for the SPTMS DIV I BULK WTR TEMP SETPT TMAX, and the second part is opposite of the required actions.

Choice D: Plausible because this is the alarm setpoint for the SPTMS DIV I BULK WTR TEMP SETPT TMAX, and the second part is correct SRO Basis: N/A

Unit 1 PSP UA22 54 Page 1 of 2 SPINS DIV I BULK WIR TEXP SEUPOINI U9.

NOTE: Inoperability of this anncnriatcr nay result in a Th1 Required Ccmpensatory Meascre AUtO TICNS NONE t:AJSE

1. High sucpression pool bul:< average water temperature OBSERVATIONS l Rerorder Channel 1 on CACTR442IA indicates inrreasing scppression pool tenerature.
2. 151 indicator illuminated (CACTY4426U -

ACTIONS

1. If suppressn:n pot1 temperature is approarhing 95°F and no testong is in progress that cccld add heat to the su;pressicn pool, then refer to AOPl4.O, Abnormal Primary Ccntairnent Conditions and AOP3L0, Safety/Relief Valve Failures If suttressoon pool temperature is greater than 99°F dcc to adtng heat to the suppression pool from approved testing roce*iures, then refer to the approprtate test procedure to naontain scppression pro1 temperature below l35
3. If suttression pot1 temperature is greater than 95°F and no testong is in progress that rould add heat to the suppressirn pool, then enter EOP02PCCP, Primary Containment Control, and ACPlCO, Abnormal Primary Containment Conditions
4. If a circuit or equipment malfcnction is susperted, enscre that a WR!WO is prepared.

DEVICE! SEIPCThTE SPINS Microprocessor CACfl442l 99°F

tnit 1 APP UA12 52 Page 1 of I SPTMS DIV I BULE WIR TEMP SETPT THAT AUTO ACTIONS NONE CAUSE I. High supPression pool bulk average water temperature OBSERVATIONS

1. Recorder Channel 1 on CACTR442IA indicates increasing suppression pooi ternerature.
2. THAT indicator illuminated tCACfl442El).

ACTI OHS

1. If suppression pool temperature is greater than SE°F and no testing is in progress that could add heat to the suppressIon pool, then enter EOPOPCCP. Primary Containment Control, and AOPl4.O, Abnormal Primary Containnent Conditions, if not already entered.
2. If suppression pool temperature is aocroarhing lorE due to adding heat to the suppression pool from approved testing procedures, then refer to the appropriate test procedure to naintain suppression pool tenperature helow lO5E.
3. If suppression pool temperature is greater than acrE, then stop all testing and enter EOPC2PCCP, Primary Containment Control, and AOPl4.C, Abnormal Prrnary Containment Conditions.
4. If a circuit or ecuigment malfunction is suspected, then ensure that a WR!WO is prepared.

DEVICE! SETPO lUTE SPTMS Microprocessor CACfl442l lOrE

LI I4I>4 ll3ftAAdO NO casoio 1VNOIS NOIIVIIINI CHAD

.LHOILIH (NI ins) 969>4 (991.5>

(NI ins)

NDJJNiW 1)69>4 9VflNVW

.i4NIHI,. I NI C3SOO ..L iVflNVLJ NISIV na-s I - j 30)01 GMSS!uued Aeids:Ou4IooD fl-IL B?JAOH

51. 2950281 Unit Two is in MODE 3 following a Station Blackout.

lAW OEOP-01 -SBO-01, Plant Monitoring, the AO has reported the following temperatures from the RSDP temperature recorder 2CAC-TR-778:

Point 1 290°F Point2 118°F Point 3 255°F Point 4 230°F Point 5 191°F Point6 117°F (REFERENCE PROVIDED)

Which one of the following represents the correct calculated Drywell temperature?

A -205°F B -P249°F C 258°F D. 267°F Answer: B K/A:

295028 High Drywell Temperature EA2 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE: (CFR: 41.10/43.5/45.13) 01 Drywell temperature RO/SRO Rating: 4.0/4.1 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing the students ability to determine drywell temperature.

Pedigree: Bank Objective: LOI-CLS-LP-303-B, Obj. 3 Given plant conditions, control room or remote shutdown panel indications, and SBO-04, calculate the following parameters: a. Drywell Temperature

Reference:

Attachment 4 of OEOP-01-SBO-01, Plant Monitoring Cog Level: Fundamental Explanation: Attachment 4 of OEOP-01 -SBO-01, Plant Monitoring, has a calculation worksheet for figuring Drywell temperature from RSDP temperature recorder readings.

290 0.141

= 40.89 255

  • 0.404 = 103.02 230
  • 0.455 = 104.65 248.56

Distractor Analysis:

Choice A: Plausible because this is the average of points 1 - 3 used in calculation.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because this is the average of points 1, 3, & 4.

Choice D: Plausible because this is performing the calculation backwards (points 4, 3, 1)

SRO Basis: N/A PLANT MONITORING OEOP-0i-SBO-.01 Rev. 0 Page 16 of 18 ATTACHMENT 4 Page 1 off Drywell Temperature Calculation Using RSDP Recorder Inputs Values obtained from Recorder CAC-TR-778 Above 70 Elevation PT1 290 X0.14t 40,89 F Between 28 and 45 Elevation PT3 255 xOAO4= 103,02 °E Between 10 and 23 Elevation PT4 230 x0455= 104.65 F Average Drywell Temperature 248.56 (Sum of 3 Regional Weighted Areas)

52. 295029 1 Unit Two is performing RVCP with HPCI in pressure control.

Subsequently, A-01 (1-5) Suppression Chamber Level Hi Hi is received.

Which one of the following completes both statements below?

The E41-F004, CST Suction Vlv, will (1)

The E41-F008, Bypass to CST Vlv, will (2)

A. (1) close (2) close B. (1) close (2) remain open C. (1) remain open (2) close D. (1) remain open (2) remain open Answer: A K/A:

295029 High Suppression Pool Water Level EA1 Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: (CFR: 41.7 / 45.6) 01 HPCI RO/SRO Rating: 3.4/3.5 Tier 1 / Group 2 K/A Match: This meets the K/A because it is testing operation of HPCI on high torus level Pedigree: New Objective: LOI-CLS-LP-019. Obj. 3p Given plant conditions, predict how the HPCI System will respond to the following events:

High/low Suppression Pool water level

Reference:

None Cog Level: High Explanation: The torus water high level condition (>-25 inches) will cause the torus suction valves to open. When either valve is full open the F008 and F004 will close.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because the first part is correct. The F008 will get a close signal when the F041 or F042 is full open.

Choice C: Plausible because the high/low level alarm does not affect the valves (annunciation only). The second part is correct.

Choice D: Plausible because the high/low level alarm does not affect the valves (annunciation only).

SRO Basis: N/A FIGURE 19-7 CST Suction Valve, E41-F004, Control Logic I CiOSF-SWHi-N f Ct).Hf Lr4 TaLL OPENJ (N-SHIN IT42 FULL OPE%

NOT rIJLL opr,__] 1rt42 T PULL OLN KI ..____(LCSLON NY. LO VL TO 1 Kt 1 I Ct ost S K2C - II D1/4 41 (OPEN: (c_OSLI OPEN CLOSE

FIGURE 19-15 Test Return Isolation Valve, E41-FOO8 (E41-FO11) Control Logic RI fCLOSrSON dl 13W PIISS (Kb; It

<20 If -f CflSEC\

lRXLflIV 2h0 S7 K?

  • f CLOSS?dHbN K16 Lro4i ruLLoPcs If (S9 (CLOSE; 4 OPEN 4 CLOSE
53. 295030 1 18 Unit One is operating at rated power when A-01 (3-7)

Suppression Chamber Lvi Hi/Lo, is received.

E---23 The BOP Operator verifies the alarm using CAC-Ll-41 77, 26 Supp Pool Level, indicator on Panel XU-51. (indication provided to the left)

E-33 Which one of the following identifies the action that is required lAW A-01 (3-7) Suppression Chamber Lvi f38 Hi/Lo?

The water level in the Unit One torus must be:

A. lowered by using Core Spray and routed to Radwaste.

B. lowered using RHR and routed to Radwaste.

C. raised by opening the HPCI suction from the CST.

D. raised by opening the Core Spray suction from the CST.

Answer: D K/A:

295030 Low Suppression Pool Water Level G2.4.5oAbility to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10/43.5 /45.3)

ROISRO Rating: 4.2/4.0 Tier 1 I Group 1 K/A Match: This meets the K/A because it is testing ability to know whether the alarm is due to high or low level and knowledge of how to correct.

Pedigree: New Objective: LOl-CLS-LP-302-D, Obj 2 Given plant conditions and AOP-14.0, determine the required supplementary actions.

Reference:

None Cog Level: High

Explanation: The student will verify that level is low using the provided indication and then will determine that the level must be raised lAW the APP. The low level alarm comes in at -30.5 inches and the high level alarm comes in at -27.5 inches. Level can be raised using RHR or the Core Spray systems.

Distractor Analysis:

Choice A: Plausible because it is a combined alarm and if it is assumed that a high water level condition exists the CS system can take a suction from the torus to correct the level condition, but is not allowed in the procedure.

Choice B: Plausible because it is a combined alarm and if it is assumed that a high water level condition exists the RHR system is utilized in the procedure to lower level.

Choice C: Plausible because level is low requiring it to be raised and the HPCI system could gravity drain to the torus, but is not allowed by the procedure.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A

Unit 1 app a:: 37 Page 1 of 2 37PPPEEEION CHA>2ER LVI HI/IC agyc ACflCTTE ITOTTE CAUSE

1. Suvpression pool water level high (2Th inches)
2. Suppressioc. pc:1 water level low (ZC inches)
3. Circuit malfunction CEEERThTICITS I. Suppression Dccl water level TCACMEQ1, CACL:4l77, CACLPEO2)

NOTE: 2apid changes in suptression pccl pressure due to conditioc.s such as inerting or air inleakage can cause level fluctuations in suppression poo1 up to 1 inch or more.

ACTI CITE TE: ECCS keepfill stations makeup flow to the suppression Dccl IS I approximately 27 ppm.

1. If the cause cf the annunciatcr is a planned evoluticn, then refer to the appropriate operatng procedure to maintaIn suppcessicn pool water level.
2. If the cause cf the annunctator is nc-t a planned evclutioc., thec.

determine the cause of additicn or lcss of water to suppression pool and ninimire evclutions which add or remove water to cr from the suppression pool.

3. If suppression pool water level is high or low, then enter IAOP14.O to drain or fill the suppression pool as necessary.
4. If suppression pool water level is greater than 27 inches or less than 31 inches, then enter 3ECPQ2PCP.

S. If a circuit malfunction is suspected, ensure a WO is prepared.

ABNORMAL PRIMARY CONTAINMENT CONDITIONS OAOP-14.O Rev 30 Page 15of36 4.2A Suppression Pool Level HighILow IF suppression pool level is approaching -27 inches, THEN lower suppression pool level to Radwaste in accordance with 1OP-17(20P-17), Residual Heat Removal System Operating Procedure D

2. if. suppression pool level is approaching -31 inches THEN raise suppression pool level in accordance with the following applicable procedure: 0 Unit 1 Only:
  • lOP-i 8. Core Spray System Operaung Procedure 0 Unit 2 Only:
54. 295031 1 Unit One is executing the ATWS procedure with the following plant conditions:

Reactor power 12%

Reactor pressure 940 psig, controlled by EHC Reactor water level 170 inches, controlled by feedwater Which one of the following identifies the reason the ATWS procedure directs deliberately lowering RPV water level to 90 inches?

A. Reduces reactor power so that it will remain below the APRM downscale setpoint.

B. Provides heating of the feedwater to reduce potential for high core inlet subcooling.

C. Reduces challenges to primary containment if MSIVs close.

D. Promotes more efficient boron mixing in the core region.

Answer: B K/A:

295031 Reactor Low Water Level EK1 Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL: (CFR: 41.8 to 41.10) 03 Water level effects on reactor power RO/SRO Rating: 3.7/4.1 Tier 1/ Group 1 K/A Match: This meets the K/A because it is testing the knowledge of why level is lowered in a ATWS Pedigree: Bank Objective: LOl-CLS-LP-300-E, Obj 7 Explain the reason for lowering reactor water level while performing the Anticipated Transient Without Scram Procedure.

Reference:

None Cog Level: fundamental Explanation: To prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities, reactor water level is lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude. Twenty-four inches below the lowest nozzle in the feedwater sparger (i.e. 90 inches) has been selected as the upper bound of the reactor water level control band. This water level is sufficiently low that steam heating of the injected water will be at least 65% to 75% effective (i.e., the temperature of the injected water will be increased to 65% to 75% of its equilibrium value in the steam environment). This water level is sufficiently high that even without bypassing the low reactor water level MSIV isolation, reactor water level can be controlled with the feedwater pumps to preclude the isolation.

Distractor Analysis:

Choice A: Plausible because since the operator can re-establish injection at 90 inches irrespective of power level. Power will lower as level is lowered but 90 inches will not guarantee APRMs are downscale Choice B: Correct Answer, see explanation.

Choice C: Plausible because since there is no current challenge to containment from heat input. If level is lowered due to containment heat input, 90 inches is not specified as the top of the level band.

This would be either TAF or the level at which downscales are received Choice D: Plausible because since lowering level will reduce natural circulation and reduce boron mixing.

ATWS procedure directs raising level back to the normal band (170-200 inches) once hot shutdown boron weight is injected SRO Basis: N/A A1WS PROCEDURE BASIS DOCUMENT OOl-375 Rev. 015 Page 130162 5.4 Step RCIL-2 NOTE I

Requeed immedii1xiftth 1 recrcuiatin pumps tripped with

+ - powcr eboe 2%.

IF reaUo power i ate 2% R AN1OT be dmiei RPV IeeI is obo.w90 inches.

Inj.clon SysThms THIN em,.rnIt awid pr.v.nt injdiun nbz the RPV unless being used * . Ccndensoteftedvot

  • CcreSprziy

- LpcI

  • Tle-2pRCIC If reactor power is greater than 23% with both reactor recircutation pumps tripped and RPV level above 90 inches, RPV level needs to be promptly reduced below The teedwater nozzles, to avoid thermal hydraulic instabilities. This is accomplished by termination and prevention of injection systems, from identified systems, particularly feedwater, within 120 seconds.

To prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutroniclthermal-hydraulic instabilities, RPV level is initially lowered sufficiently below the etevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing et[ective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, initiation and growth of oscillations is principally dependent upon subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude.

55. 295032 1 Which one of the following identifies the teason for performing Emergency Depressurization due to exceeding Maximum Safe Operating Temperatures lAW 001-37.9, Secondary Containment Control Procedure Basis Document?

A. Prevent an unmonitored release.

B. Preserve personnel access into the reactor building.

C. Provide continued operability of equipment required for safe shutdown.

D. Ensure ODCM site boundary dose limits are not exceeded.

Answer: C K/A:

295032 High Secondary Containment Area Temperature EK3 Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: (CFR: 41.5/45.6) 01 Emergency/normal depressurization RO/SRO Rating: 3.5/3.8 Tier 1 / Group 2 K/A Match: This meets the K/A because it is testing the reason ED is performed for high secondary containment temperatures.

Pedigree: Bank Objective: LOl-CLS-LP-300-M, Obj 13a Given plant conditions and the SCCP, determine the required actions if the following limits are exceeded: Maximum Safe operating values with a primary system discharging into secondary containment.

Reference:

None Cog Level: Fundamental Explanation: The MSOT values are the area temperatures above which equipment necessary for the safe shutdown of the plant will fail. These area temperatures are utilized in establishing the conditions which reactor depressurization is required. The criteria of more than one area specified in this step identifies the rise in reactor building parameters as a wide spread problem which may pose a direct and immediate threat to secondary containment integrity, equipment located in the RB, and continued safe operation of the plant.

Distractor Analysis:

Choice A: Plausible because this is a purpose of SCCP not the reason for ED on Temperature.

Choice B: Plausible because this is the reason for max safe operating rad levels.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because this is a purpose of SCCP not the reason for ED on Temperature.

SRO Basis: N/A

SECONDARY CONTAINMENT CONTROL 001-37.9 PROCEDURE BASIS DOCUMENT Rev. 004 Page 8 of 33 5.1 Step SCCP-1 cbn., :o.o&.I I

  • .,,Wne,u, S,fl,jtOW,p,n flNfl(c,.,yrg 1I*fl fr..W$,,Lr.I C. *.

The conditions Iuich require entry to SCCP are symptomatic of conditions wtüch, if not colTected, could degrade into an emergency. Adverse effects on the operability of equipment located in the reactor building and conditions directly challenging secondary containment integnty or spent fuel pool cooling were specifically considered in the selection of these entry conditions. In addition, personnel accessibility to some of the areas may be required to perform certain actions specified in the procedure. This was also considered in making these determinations.

An area temperature or area differential temperature above its maximum normal operating level is an indication that steam from a primary system may be discharging into the reactor building. As temperatures continue to increase, the continued operability of equipment needed to carry out E0P actions may be compromised.

56. 295034 1 Which one of the following completes both statements below?

lAW OAOP-5.4, Radiological Releases, RRCP is entered when the Turbine Building Vent Rad Monitor indication exceeds an (1) EAL.

lAW RRCP, before the radioactivity release rate reaches a (2) Emergency EAL, Emergency Depressurization is required.

A. (1) Unusual Event (2) Site Area B. (1) Unusual Event (2) General C. (1) Alert (2) Site Area D. (1) Alert (2) General Answer: D K/A:

295034 Secondary Containment Ventilation High Radiation G2.4.O8Knowledge of how abnormal operating procedures are used in conjunction with EOPs. (CFR: 41.10

/43.5 /45.13)

RO/SRO Rating: 3.8/4.5 Tier 1/ Group 2 K/A Match: This question matches the KA because it tests the knowledge if the AOP and EOP are performed in conjunction with each other.

Pedigree: new Objective: LOl-CLS-LP-302-J, Obj. 3c Given plant conditions, determine the required Supplementary Actions in accordance with:

OAOP-05.4, Radiological Release

Reference:

None Cog Level: Fundamental Explanation: The AOP states that when an Alert EAL is entered then ENTER RRCP. Before a GE is declared ED is required to be performed. (A scram is required before a SAE is declared)

Distractor Analysis:

Choice A: Plausible because an Unusual Event is the first declaration in the EAL network and a SAE is the criteria for a scram in RRCP.

Choice B: Plausible because an Unusual Event is the first declaration in the EAL network and the second part is correct.

Choice C: Plausible because the first part is correct and the SAE is the criteria for a scram in RRCP.

Choice D: Correct Answer, see explanation SRO Basis: N/A RADIOLOGICAL RELEASE OAOP-05.4 Rev. 0 Page 6 of 13 3.0 AUTOMATIC ACTIONS (continued)

  • Group 6 isolation valves close 0
4. IF UA-03 2-8, Radwaste Effluent Rad Hi Hi, in ALARM, THEN D12-V27A(B) (RW Liq Effluent Disch Vlvs) dose 0
5. IF UA-23 3-6, Main Steam Line Red Hi-Hi/mop, in ALARM, THEN:
  • Mechanical vacuum pumps trip 0
  • OG-V7 (Cndsr Hogging Valve) ctoses 0 4.0 OPERATOR ACTIONS 4.1 Immediate Actions None 4.2 Supplementary Actions
1. IF AT ANY TIME elevated radiation levels are deterimned to be from resin injection pjjy, THEN go to OAOP-26.0, High Reactor Coolant or Condensate Conductivity 0
2. IF AT ANY TIME gaseous release rate exceeds an Alert level, THEN enter OEOP-04-RRCP, Radioactivity Release Control Procedure 0

2 0 d ci 1 0 q

N w

0uJ U

Ui

57. 295036 1 Following an unisolable RWCU line break in the reactor building the following conditions exist:

South Core Spray Room temperature 155°F South RHR Room temperature 300°F UA-1 2 (2-3) South Core Spray Room Flood Level Hi, in alarm UA-12 (2-4) South RHR Room Flood Level Hi, in alarm UA-1 2 (1-4) South RHR Room Flood Level Hi-Hi, in alarm (REFERENCE PROVIDED)

Which one of the following completes both statements below?

lAW OEOP-01-UG, Users Guide, (1) equipment required for safe shutdown will fail.

lAW SCCP, Emergency Depressurization (1) required.

A. (1) ONLY the South RHR room (2) is B. (1) ONLY the South RHR room (2) is NOT C. (1) the South RHR room AND Core Spray room (2) is D. (1) the South RHR room AND Core Spray room (2) is NOT Answer: B KJA:

295036 Secondary Containment High Sump / Area Water Level EK1 Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: (CFR: 41.8 to 41.10) 02 Electrical ground! circuit malfunction RO/SRO Rating: 2.6/2.8 Tier 1 I Group 2 K/A Match: This meets the K/A because this is testing the implication of high water level on equipment and whether ED is required.

Pedigree: New Objective: LOl-CLS-LP-300-M, Obj, 13a Given plant conditions and the Secondary Containment Control Procedure, determine the required action if the following limits are exceeded: Maximum Safe operating values WITH a primary system discharging into Secondary Containment

Reference:

OEOP-01-NL, EOP/SAMG Numerical Limits And Values, Attachment 3, Containment Parameters, Table 3-B, Secondanj Containment Area Temperature Limits

Cog Level: High Explanation:

Distractor Analysis:

Choice A: Plausible because the first part is correct and for ED two areas in the same parameter must be at max safe conditions, while this question has two parameters in the same area.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because both areas have a max normal condition and for ED two areas in the same parameter must be at max safe conditions, while this question has two parameters in the same area.

Choice D: Plausible because both areas have a max normal condition and the second part is correct.

SRO Basis: N/A

USERS GUIDE OEOP-01-UG Rev. 067 Page lOof 156 3.0 DEFINITIONS (continued)

  • RHR Loop A (one or two pumps running)
  • RHR Loop 5 (one or iwo pumps running)
32. Maximum Normal Operating (Parameter): The highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning propeily.

33 Maximum Pressure Suppression Primary Containment Water Level: The highest primary containment water level at tflch the pressure suppression capability of the containment can be maintained. This corresponds to the bottom of the ring header.

34. Maximum Safe Operating Radiation Level: The radiation level above which personnel access necessary for the safe shutdown of the plant ill be precluded.

If the maximum safe operating radiation level is exceeded in an area (but is within the EQ envelope as contained in DR-227, Document Reference for Environmental Qualification Service Conditions) and then later clears and is subsequently followed by another area exceeding maximum safe operating radiation level, action for one area exceeding maximum safe operating radiation level should be taken.

35. Maximum Safe Operating Temperature: The temperature above which equipment necessary for the safe shutdown of the plant may fail. This temperature is utilized in establishing the conditions under which RPV depressunzation is required. Separate temperatures are provided for each Secondary Containment area. If the maximum safe operating temperature is exceeded in an area and then later clears and is subsequently followed by another area exceeding maximum safe operating temperature, action for two areas exceeding maximum safe operating temperature should be taken.
36. Maximum Safe Operating Water Level: The water level above which equipment necessary for the safe shutdown of the plant may fail. This water level is utilized in establishing the conditions under which RPV depressunzation is required. Separate water levels are provided for each Secondary Containment area. If the maximum safe operating water level is exceeded in an area and then later clears and is subsequently followed by another area exceeding maximum safe operating water level, action for twa areas exceeding maximum safe operating water level should be taken.

WHEN II paarni.flr Max Safe OR EQ I envelope an nwre than one &eo

\ THEN

&CCP-9 F,.

(E.MERGNCY DEPRSSURIZATION REQUIRED.

ATTACHMENT 3 Page 73 of 87 Containment Parameters Secondary Containment Area Temperature Limits Table 3-B PLANT PLANT LOCATION MAX NORM MAX SAFE AUTO GROUP AREA DESCRIPTION OPERATING OPERATING ISOLATION VALUE (F) VALUE (F)

N CORE N CORE SPRAY 120 175 N/A SPRAY ROOM S CORE S CORE SPRAY 120 175 N/A SPRAY ROOM RWCU PMPROOMA PMPROOMB 140 225 3 IIX ROOM N RHR N RHR EQUIP ROOM 175 295 N/A S RI-fR S RHR EQUIP ROOM 175 295 N/A RCIC EQUIP ROOM 165 295 5 HPCI HPCIEOUIPROOM 165 165 4 STEAM RCICSTMTUNNEL 190 295 5 TUNNEL HPCI STM TUNNEL 190 295 4 20 FT 20 Fr NORTh 140 200 NJA 20 rr SOUTH 140 200 N/A SOFT 5OFTNW 140 200 NIA 50Ff SE 140 200 N/A REACTOR MULTIPLE AREAS ALARM N/A 3, 4, AND/OR 5

&DG ANNUN. SETPOINT A-02 5-7 REACTOR MSIV PIT ANNUN. ALARM N/A 1 SLOG A-O6 6-7 SETPOINT

58. 295037 1 The RD has attempted to manually scram Unit One with the following actions taken:

All rods are noted to be greater than position 02 Reactor mode switch is placed in shutdown ARI was initiated.

Both recirculation pumps were tripped.

Reactor power reported at 12%

SLC is injecting RPV level is 80 inches and stable Rod insertion attempts are unsuccessful Which one of the following completes both statements below?

Reactor power (1) expected to be lowering.

Assuming no rod insertion, SLC injection (2)

A. (1) is (2) can be secured when all APRMs are downscale B. (1) is (2) must be continued until the reactor is shutdown under all conditions C. (1) is NOT (2) can be secured when all APRMs are downscale D. (1) is NOT (2) must be continued until the reactor is shutdown under all conditions Answer: B K/A:

295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown EK1 Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

(CFR: 41.8 to 41.10) 03 Boron effects on reactor power (SBLC)

RO/SRO Rating: 4.2/4.4 Tier 1 / Group 1 K/A Match: This meets the K/A because the student will have to know the effects that boron has shutting down the reactor during an ATWS.

Pedigree: Bank Objective: LOI-CLS-LP-005, Obj 3 List the positive reactivity effects that must be overcome by SLC injection

Reference:

None Cog Level: Fundamental

Explanation: Injection of the CSBW into the RPV will provide adequate assurance that the reactor is and will remain shutdown. It is the least weight of soluble boron which, if injected into the RPV and mixed uniformly, will maintain the reactor shutdown under all conditions. This weight is utilized to assure the reactor will remain shutdown irrespective of control rod position or RPV water temperature. Boron injection is continued until the entire tank is injected or all rods are inserted.

Distractor Analysis:

Choice A: Plausible because the first part is correct and the second part is plausible because in some procedures the reactor is called shutdown if power is downscale on the APRMs.

Choice B: Correct Answer, see explanation Choice C: Plausible because rods are not being inserted and the second part is plausible because in some procedures the reactor is called shutdown if power is downscale on the APRMs.

Choice D: Plausible because rods are not being inserted and the second part is correct.

SRO Basis: N/A 1.2 System Design Basis The design basis for the SLC System is as llows:

1.2.1 Backup capability for reactivity control is provided, independent of the normal reactivity control provisions in the nuclear reactor, to permit shutdown of the reactor if the normal control ever becomes inoperative.

1.2.2 To assure complete shutdown from the most reactive condition at any time in core life, this backup system has the capacity to control the reactivity difference between the steady state rated operating condition of the reactor with voids and the cold shutdown condition, including shutdown margin.

1.2.3 The time required to actuate and effect the backup control is consistent with the nuclear reactivity rate of change predicted between rated operating and cold shutdown conditions. A fast scram of the reactor or operational control of fast reactivity transients is not specified for this system.

1.2.4 Means are provided by which the functional performance capability of the backup control system components can be verified under conditions approaching actual use requirements.

1.2.5 The neutron absorber is dispersed within the reactor core in sufficient quantity to provide a reasonable margin for leakage, dilution, or SD-Q5 Rev. 11 Page5of43

59. 295038 1 A radioactive release has occurred in the Turbine Building.

Which one of the following completes both statements below?

lAW OAOP-05.4, Radiological Releases, the Unit Two turbine building ventilation must be in the (1) operating mode.

This discharge will be monitored by the (2)

A. (1) recirc (2) Main Stack Radiation Monitor B. (1) recirc (2) Wide Range Gaseous Monitor (WRGM)

C. (1) once through (2) Main Stack Radiation Monitor D. (1) once through (2) Wide Range Gaseous Monitor (WRGM)

Answer: B K/A:

295038 High Off-Site Release Rate EAJ Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE:

(CFR: 41.7 I 45.6) 01 Stack-gas monitoring system RO/SRO Rating: 3.9/4.2 Tier 1 / Group 1 K/A Match: This meets the K/A because the student will have to determine the procedural requirement for the turbine building ventilation operational mode and the rad monitor that monitors it. (ability to monitor)

Pedigree: New Objective: LOl-CLS-LP-302-J, Obj 3c Given plant conditions, determine the required Supplementary Actions in accordance with:: c.

OAOP-05.4, Radiological Release

Reference:

None Cog Level: Fundamental Explanation: The turbine building ventilation can be lined up for once through or recirc mode, the AOP has the operator ensure that it is lined up in the recirc mode. The discharge is monitored by the turbine building WRGM.

Distractor Analysis:

Choice A: Plausible because the first part is correct and the second part is a common radiation monitor for other ventilation systms.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because once through is a mode of operation for the TB Ventilation system and the second part is a common radiation monitor for other ventilation systms.

Choice D: Plausible because once through is a mode of operation for the TB Ventilation system and the second part is correct.

SRO Basis: N/A RADIOLOGICAL RELEASE OAOP-05.4 Rev. 001 Page BoriS 4.2 Supplementary Actions (continued)

NOTE

  • Turbine Building Fiabitability shouLd be consklered for establishment as critical parameters during performance of this procedure U
  • Emergency Plan requirements mandate securing once through ventilation for any site radiological release U
10. if any site radiological release occumng, THEN ensure Unit 2 turbine building ventilation in recirculation mode per 20P-37.3, Turbine Building Ventilation System Operating Procedure U
60. 300000 1 Unit One is operating at rated power when the following alarms are received:

UA-01 (4-4) lnstr Air Press-Low UA-01 (5-1) Air Dryer IA Trouble The AO reports that the cause of the alarms is due to filter blockage.

Which one of the following completes both statements below?

The Service Air Dryer malfunction will cause SA-PV-5067, Service Air Dryer Bypass Valve, to open when pressure first lowers to (1) lAW OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures, the required action isto (2)

A. (1) 105 psig (2) place the 1 B Service Air Dryer in service B. (1) 105 psig (2) set the service air dryer maximum sweep value to zero C. (1) 98psig (2) place the 1 B Service Air Dryer in service D. (1) 98psig (2) set the service air dryer maximum sweep value to zero Answer: C K/A:

300000 Instrument Air System A2 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: (CFR: 41.5 /45.6) 01 Air dryer and filter malfunctions RO/SRO Rating: 2.9/2.8 Tier 2 / Group 1 K/A Match: This meets the KA because it is predicting the response on the system and then using procedure (AOP-20) determine the action required.

Pedigree: New Objective: LOI-CLS-LP-046, Obj. 6 Given plant conditions, determine if the following automatic actions should occur:

a. Service Air Isolation g. Air Dryer bypass.

Reference:

None Cog Level: High

Explanation: 98 psig is when the bypass valve auto opens, the 105 psig is the isolation setpoint for Service Air. The AOP will direct placing the standby Air Dryer in service.

Distractor Analysis:

Choice A: Plausible because 105 is the isolation setpoint for the service air system and the second part is correct.

Choice B: Plausible because 105 is the isolation setpoint for the service air system and this is an action in the AQP but would not be performed for this failure. If there is a high demand then this is performed to limit the amount of air that is used for the blowdown of the air dryer filter when cycling filters.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because the first part is correct and this is an action in the AOP but would not be performed for this failure. If there is a high demand then this is performed to limit the amount of air that is used for the blowdown of the air dryer filter when cycling filters.

SRO Basis: N/A

PNEUMATIC (AIRJNITROGEN) SYSTEM FAILURES OAOP-20.0 Rev. 46 Page 13of28 4.2 Supplementary Actions (continued)

NOTE

  • Service Air System pre-filter or affer-fUter dflerential pressure should NOT exceed 15 psid D
  • In service air compressor high discharge pressure (Unit 1: greater than or equal to 125 psig, Unit 2: greater than or equal to 130 psig) or relief valves lifting could be an indication of air dryer high differential pressure potentially caused by power failures resulting in valves in the flow path failing closed 0
  • 1 (2)SA-PV-5067 [Serv Air Dryer I (2)A G,pass Pressure Control Valve), is located in the Turbine Building air compressor area 0
i. IF UA-01 5-3, Air Dryer if2)A Trouble, is in alarm, THEN perform the following:

(1) Unit I Only: Confirm i-SA-PV-5067 fSeIv Air Dryer JA Bypass Pressure Control Valve), is OPEN 0 CAUTION The service air dryer provides a low dew point pneumatic source to downstream components. A low dew point is necessary to insure long term reliability of these components. The time the dryer is bypassed should be minimized {7.1 .1} 0 (2) Unit I Only: IF 1-SA-PV-5067 is NOT open, THEN open 1-SA-V5089 fServ Air Dryer Manual Bypass Valve) 0 (3) Unit 2 Only: Confirm 2-SA-PV-5067 (Serv Air Dryer 2A Bypass Pressure Control Valve), is OPEN 0 (4) Unit 2 Only: IF 2-SA-PV-5067 is NOT open, THEN open 2-SA-V5089 fServ Air Dryer Manual Bypass Valve) 0 (5) IF availabte, THEN place 1 B Service Air Dryer in service AND shutdown 1 (2)A Service Air Dryer in accordance with QOP-46, Instrument and Service Air System Operating Procedure 0

PNEUMATIC (AIR/NITROGEN) SYSTEM FAILURES OAOP-20.O Rev. 46 Page 9 0128 4.2 Supplementary Actions (continued)

c. IF air is NOT cross-tied, AND cross-tie operation will .NI cause a loss of instrument air on the unafFected unit, THEN perform the following:

(1) Obtain permission from the non-affected unit C (2) Ensure 1-SA-PV-5071 (Cross-Tie Valve), located on Unit 1, Panel XU-2, is OPEN C (3) Ensure 2-SA-PV-5071 (Cross-Tie Valve), located on Unit 2, Panel XU-2, is OPEN C (4) IF opening the cross-tie valve degrades the non-affected unit, THEN return to Step 1 .b(4) C ci. IF the in service air dryer is in sweep mode, THEN consider securing sweep mode in accordance with Attachment 1, Setting Service Air Dryer(s) Maximum Sweep Value ToZero C

61. 3000002 Unit One is in MODE 3 following a seismic event and reactor scram with the following plant conditions:

Reactor level 55 inches Reactor pressure 500 psig Drywell pressure 9 psig Division I PNS header pressure 93 psig Division II PNS header pressure 98 psig Which one of the following completes both statements below?

Div I Backup N2 Rack Isol Vlv, RNA-SV-5482 is (1)

Div II Backup N2 Rack Isol Vlv, RNA-SV-5481 is (2)

A. (1) open (2) open B. (1) open (2) closed C. (1) closed (2) open D. (1) closed (2) closed Answer: B K/A:

300000 Instrument Air System K3 Knowledge of the effect that a loss or malfunction of the (INSTRUMENT AIR SYSTEM) will have on the following: (CFR: 41.7 I 45.6) 01 Containment air system RO/SRO Rating: 2.7/2.9 Tier 2 I Group 1 K/A Match: This meets the KA because it is testing the effect of the low pressure (loss or malfunction) of the air system on containment air (N2 backup).

Pedigree: Last used on 2007 NRC Exam Objective: LOl-CLS-LP-046-A, Obj. 8 Given plant conditions, determine the effects that the following conditions will have on the Pneumatic System: (LOCT) b. Low Instrument Air/Pneumatic Nitrogen (IANIRNNPNS)

Header Pressure

Reference:

None Cog Level: High

Explanation: No LOCA signal is present so the Backup N2 valves will not be open on a Core Spray initiation signal. The Backup N2 valves open at 95 psig or lower in the PNS header. This would result in Division I Backup N2 valve (5482) being open and Division 11(5481) being closed.

Distractor Analysis:

Choice A: Plausible if the student believes that either division will open both valves.

Choice B: Correct Answer, see explanation.

Choice C: Plausible if the student uses the valves for the division separation.

Choice D: Plausible if the student only checks the LOCA signal and not the low pressure signal.

SRO Basis: N/A Unit I APP UA-DJ 1-1 Page 1 of 2 RB INSTR AIR RECEIVER 1A PRESS LOW AUTO ACTIONS

1. RNA-SV-5482, High Pressure Bottle Rack Isolation VaIe, opens, suppIing SRVs and CAC-16 with a pneumatic source.

DEVICEISETPOINTS RNA-PSL-3596 95 psg decreasing Unit I APP UA-O1 1-2 Page 1 of 2 RB INSTR AIR RECEIVER lB PRESS LOW AUTO ACTIONS I. RNA-SV-548f, High Pressure Bottle Rack Isolation Valve opens supplying SRVs and CAC-f 7 with a pneumatic source.

DEVICEISETPOINTS RNA-PSL-3597 95 psig decreasing

62. 400000 1 Unit One is operating at rated power with the following conditions:

CSW Pump IA trips Conventional header pressure lowers to 35 psig Which one of the following completes both statements below?

If CSW header pressure remains at this pressure for (1) seconds, the SW-V3, SW To TBCCW HXs Otbd Isol Vlv, and SW-V4, SW To TBCCW HXs InbU Isol VIv, will close to a throttled position.

lAW OAOP-I 9, Conventional Service Water System Failure, the SW-V3 and SW-V4 are reopened (2)

A. (1) 30 (2) ONLY after a reactor Scram is inserted B. (1) 30 (2) if system pressure is restored by starting the standby CSW pump C. (1) 70 (2) ONLY after a reactor Scram is inserted D. (1) 70 (2) if system pressure is restored by starting the standby CSW pump Answer: D K/A:

400000 Component Cooling Water System (CCWS)

A2 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 operation:

(CFR: 41.5 /45.6) 01 Loss of CCW pump RO/SRO Rating: 3.3/3.4 Tier 2 / Group 1 K/A Match: This meets the KA because it is testing the auto start signal/logic for a cooling water system.

Pedigree: Last used on the 2010 NRC exam Objective: CLS-LP-302-H, Obj. 4 Given plant conditions and any of the following AOPs, determine the required supplementary actions: U. OAOP-19.0, Conventional Service Water System Failure

Reference:

None Cog Level: High

Explanation: IF conventional service water header pressure remains below 40 psig for 70 seconds, THEN:

- SW TO TBCCWHXS OTBD ISOL, SW-V3 closes to a throttled position

- SW TO TBCCWHXS INBD ISOL, SW-V4 closes to a throttled position The Standby CSW pump should start and restore CSW header pressure to normal prior to the SW valves throttling closed. If the standby CSW pump fails to auto start, manually starting the pump will restore CSW header pressure. AOP-19 provides guidance to re-open the SW valves only after header pressure has been restored and the cause of low pressure is known (pump trip).

Distractor Analysis: -

Choice A: Plausible because 30 seconds is when the DG cooling water valves close and a Scram is inserted only after the SW valves have closed to the throttled position AND CSW header pressure cannot be immediately restored above 40 psig under this condition all CSW pumps would be shutdown.

Choice B: Plausible because 30 seconds is when the DG cooling water valves close and system pressure restored by the STBY pump start is correct.

Choice C: Plausible because 70 seconds is correct and a Scram is inserted only after the SW valves have closed to the throttled position AND CSW header pressure cannot be immediately restored above 40 psig under this condition all CSW pumps would be shutdown.

Choice D: Correct Answer, see explanation SRO Basis: N/A CONVENTIONAL SERVICE WATER SYSTEM OAOP-19.0 FAILURE Rev. 26 Page 5 0111 3.0 AUTOMATIC ACTIONS Standby pump setected to the conventional service water header startsat4opsig U

2. IF all conventional service water pumps are tripped, THEN:
  • SW-V36 (SW To CW Pumps Inbd Vlv), closes U
  • SW-V37 (SW To CW Pumps Otbd Vlv), closes U
3. IF conventional service water header pressure remains less than 40 psig for 70 seconds, THEN:
  • SW-V3 (SW To TBCCW HXs Otbd Isol), closes to a throttled position U
  • SW-V4 (SW To TBCCW HX5 Inbd Isol), closes to a throttled position U

CONVENTIONAL SERVICE WATER SYSTEM OAOP-f 9.0 FAILURE Rev. 26

. Page9ofll 4.2 Supplementary Actions (continued)

d. Attempt to isolate any source of leakage I]
e. Ensure discharge valves are CLOSED on shutdown pump(s) El
f. Check service water traveling screens for excessive build-up AN. wash if excessive buildup is occurring El
9. Check service water trash racks for excessive build-up AND notify Maintenance to clean it excessive buildup is occurring El
h. Locally monitor each pump discharge strainer differential pressure U
i. Check Annunciator Panel UA-01 for lit annunciators El
10. Refer to Technical Specirication 3.7.2, Service Water (SW) System and Ultimate Heat Sink (UHS) for operability requirements El
11. WHEN conventional service water header pressure is restored to normal AND the cause of low header pressure has been corrected, THEN:
a. Open SW-V3 (SW To TECCW HX5 Otbd Isol) El
b. Open SW-V4 (SW To ThCCW HXs lnbd lsol) U
63. 400000 2 Unit Two Nuclear Service Water (NSW) pumps are aligned as follows in preparation for equipment realignment:

DISCHARGE NUCLEAR SERVICE DISCHARGE NUCLEAR SERVICE VLV SW-V20 VLV SWV19 VATER PUMP 2A WATER PUMP 2B 2PB E4

© Subsequently, Off-site power is lost.

Which one of the following completes the statement below?

(1) NSW pump(s) will auto start (2) associated E Bus is re-energized.

A. (1) 2A and 25 (2) immediately when their B. (1) 2A and 2B (2) five seconds after their C. (1) 2BONLY (2) immediately when its D. (1) 2B ONLY (2) five seconds after its Answer: A K/A:

400000 Component Cooling Water System (CCWS)

K4 Knowledge of CCWS design feature(s) and or interlocks which provide for the following: (CFR:

41.7) 01 Automatic start of standby pump ROISRO Rating: 3.4/3.9 Tier 2 I Group 1

K/A Match: This meets the KA because it is testing the auto start signal/logic for a cooling water system.

Pedigree: Bank Objective: LOl-CLS-043, Objective 8a State the power supply (bus and voltage) for the following Service Water System components:

Nuclear Service Water Pumps.

Reference:

N/A Cog Level: High Explanation: NSW pumps auto start immediately after LOOP signal regardless of mode selector switch or discharge valve position. 5 second timer applies only on a LOCA.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because this would be the case with a LOCA signal present.

Choice C: Plausible because examinee must know the power supply scheme.

Choice D: Plausible because examinee must know the power supply scheme.

SRO Basis: N/A Each pump is powered by a 4160 VAC motor supplied from the emergency bus power supplies:

Component Power Supply 1ACSW pump E4 lBCSWpump El lCCSWpump E2 lANSWpump El lBNSWpump E2 2ACSWpump E3 28 CSW pump E4 2C CSW pump Ef 2A NSW pump E3 28 NSW pump E4 SD-43 Rev. 25 Page 10 of 87

In addition to the low header pressure auto start, the NSW pumps will start live seconds after receipt of a LOCA signal, regardless of mode selector switch or discharge valve position. For example, a Division I LOCA signal from either Unit 1 or Unit 2 will auto start the 1A and 2A NSW pumps; the Division II LOCA logic wilt auto start 16 and 26 NWS pumps.

The NSW pumps, powered through the 4160 VAC emergency buses, will also automatically start immediately after the start of the diesel generators and reenergization of the emergency buses on loss of off-site power (LOOP), regardless of mode selector switch or discharge valve position. For example, a Division I LOOP signal from either Unit 1 or Unit 2 will auto start the IA and 2A NSW pumps; the Division II LOOP logic will auto start 16 and 26 NSW pumps. If a LOCA signal exists on the division sensing the LOOP, auto start will occur after live seconds, provided that a LOOP signal is not present on the opposite unit. On a dual unit LOOP the NSW pump(s) of the LOCA (and non LOCA) unit start immediately after the emergency buses are reenergized by their respective diesel generators without the live second delay.

64. 600000 1 Which one of the following identifies the potential consequence of failing to place backup nitrogen in service by placing RNA keylock switches in LOCAL lAW OASSD-02, Control Building?

RNA keylock switch noun names:

2-RNA-CS-001, Override Switch For Valve RNA-SV-5482 2-RNA-CS-002, Override Switch For Valve RNA-SV-5253 A. Misoperation of RCIC.

B. Loss of drywell cooling.

C. Inability to operate SRVs.

D. Spurious operation of MSIVs.

Answer: C K/A:

600000 Plant Fire On Site AK3 Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE:

(CFR: 41.5/45.6) 04 Actions contained in the abnormal procedure for plant fire on site RO/SRO Rating: 2.8/3.4 Tier 1 / Group 1 K/A Match: This matches the KA because it tests the reason a step in the ASSD procedure is performed.

The ASSD procedures are the plant fire procedures.

Pedigree: Bank Objective: LOI-CLS-LP-304, Obj. 25k Given ASSD procedures and plant conditions, predict the consequences of FAILURE to perform the following actions: Deenergize RNA-SV-5482 and RNA-SV-5253 via keylock switches RNA-CS-001 and RNA-CS-002.

Reference:

None Cog Level: fundamental Explanation: The Reactor Building MCC Operator places the key lock switches to the LOCAL position to ensure Nitrogen System is lined up to provide reliable operation of the SRVs.

Distractor Analysis:

Choice A: Plausible because actions for the operation of RCIC are contained in the ASSD procedures.

Choice B: Plausible because a loss of pneumatics would cause the DW cooler dampers to close.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because actions to prevent the spurious operation of the MSIV is contained in the ASSD procedure.

SRO Basis: N/A

SECTION Bi UNIT 2 RX BLDG MCC OPERATOR ACTIONS Initial Actions and RCIC Operations 1.6.5 WHEN directed to start RCIC, THEN PERFORM the following at MCC 2XDB:

1. OPEN RCIC TURB TR & fliP VLV, E51-V8, at Compt B37 (Row Cl).
2. OPEN RCIC TURB STh1 SPLY VLV, E51-F045, at Compt B44 (Row F2).
3. INFORM Unit 2 CR5 RCIC should be running. E 1.7 WHEN directed, THEN PERFORM the following at Unit 2 Reactor Building 50 foot elevation:

1.71 PLACE keylock switch 2-PNA-CS-OO1 in LOCAL for valve 2-RNA-SV-5482.

1.7.2 PLACE keylock switch 2-PNA-CS-002 in LOCAL for valve fl 2-PNA-SV-5253.

1.7.3 INFORM the Unit 2 CR5 that backup nitrogen has been made available for SRV operation.

1.8 IF directed, THEN TRANSFER RCIC suction from CST to suppression pool at MCC 2XDB as follows:

1.8.1 OPEN PCIC SUPP POOL SUCT VLV, E51-F031, at Compt B45 (Row Gi).

1.8.2 OPEN PC1C SUPP POOL SUCT VLV, E51-FO2& at El Compt B46 (Row G2).

1.8.3 CLOSE RC!C CST SUCT VLV, E51-FOic at Compt B38 (Row C2).

OASSD-02 Rev. 57 Page 31 of 156

65. 700000 1 A grid disturbance occurs with the following Unit One plant parameters:

Generator Load 980 MWe Generator Reactive Load 160 MVARs, out Generator Gas Pressure 50 psig (REFERENCE PROVIDED)

Which one of the following identifies both available options that will place the Unit within the Estimated Capability Curve?

A. Raise gas pressure to 58 psig or lower power to 940 MWe.

B. Raise gas pressure to 58 psig or raise reactive load to 240 MVARs.

C. Raise gas pressure to 58 psig or lower reactive load to 70 MVARs.

D. Lower power to 940 MWe or raise reactive load to 240 MVARs.

Answer: A K/A:

700000 Generator Voltage and Electric Grid Disturbances AA2 Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: (CFR: 41.5 and 43.5 I 45.5, 45.7, and 45.8) 03 Generator current outside the capability curve RO/SRO Rating: 3.5/3.6 Tier 1 / Group 1 K/A Match: This meets the K/A because the tests the ability to determine action needed to remain within capability curve.

Pedigree: Last used on 2014 NRC exam Objective:

CLS-LP-27, Obj. 9 Given the Generator estimated capability curves, hydrogen pressure and either MVARS, MW, or power factor, determine the limit for MW and MVARS.

Reference:

JOP-27 Attachment 2, Estimated Capability Curves Cog Level: High Explanation: Based on the conditions the student should plot the current location on the graph. Plot MWe along the bottom and MVARs up the side. Where these two points intersect, based on 50 psig gas pressure line is outside of the safe area. (Must be inside the curve to be safe) Lowering MWe or raising gas pressure are the only options. For this case lowering or raising MVARs would still be outside the curve.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because raising pressure will move the plant within the limits of the curve. Raising MVARS will not move the plant within the limits of the curve.

Choice C: Plausible because raising pressure will move the plant within the limits of the curve. Lowering MVARS will not move the plant within the limits of the curve.

Choice D: Plausible because raising MWe will move the plant within the limits of the curve. Raising MVARS will not move the plant within the limits of the curve.

SRO Basis: N/A o roio - - it 5BPSIG 48PSIG  :

45 PSIC 391 400 200  :

  • U, -

o 200 I.

-400 o -600 w -

-800 200 400 600 800 1000 1000 KILOWATTS

66. G2.1.O1 I Which one of the following completes both statements below lAW AD-OP-ALL-i 000, Conduct of Operations?

With the Unit operating at rated, steady state power, steam flow I feed flow (1) a key parameter that the OATC must monitor to assure a constant awareness of its value and trend.

An end to end control panel walk down shall be performed every (2) and documented in the Narrative Logbook.

A. (1) is NOT (2) one hour B. (I) is NOT (2) two hours C. (I) is (2) one hour D. (1) is (2) two hours Answer: D K/A:

G2.1.01 Knowledge of conduct of operations requirements. (CFR: 41.10/45.13)

RO/SRO Rating: 3.8/4.2 Tier 3 K/A Match: This meets the K/A because it is testing knowledge of the Conduct of Operations Manual Pedigree: New Objective: LOl-CLS-LP-201-D, Obj. lj Explain/describe the following lAW AD-OP-ALL-i 000, Conduct of Operations, 001-01.01, BNP Conduct of Operations Supplement and OPS-NGGC-1314, Communications: Control Board walkdown and monitoring requirements

Reference:

None Cog Level: Fund Explanation: lAW the conduct of operations document board walk downs must be completed every two hours and section 5.5.6 lists the key parameters to watch.

Distractor Analysis:

Choice A: Plausible because jet pump flow has a daily surveillance requirement and if a watchstander is relieved for greater than one hour it must be entered in narrative logbook.

Choice B: Plausible because jet pump flow has a daily surveillance requirement and part two is correct.

Choice C: Plausible because part one is correct and if a watchstander is relieved for greater than one hour it must be entered in narrative logbook.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A CONDUCT OF OPERATIONS AD-OP-ALL-i 000 Rev. 5 Page 27 of 85 5.5.6 Control Board Monitoring (continued) k Unless involved in activities where Reactor Operator involvement is required by the Conduct of Operations (for example reactivity manipulations, peer checks or deiled panel reviews), the operator shalt monitor the following key parameters at a frequency to assure a constant awareness of their value and end:

  • RxPowor
  • RCS temperature
  • Steam flow I feed flow
  • Pressunzer level (PWR)

CONDUCT OF OPERATIONS AD-OP-ALL-i 000 Rev. 5 Page 28 of 85 5.5.6 Control Board Monitoring (continued)

3. The CR3 ensures that a hcensed operator perfomis an end to end control panel walk down every two hours. The watk down shall be documented in the Narrative Logbook CONDUCT OF OPERATIONS AD-OP-ALL-i 000 Rev. 5 Page 40 of 85
4. Whenever a watch stalon is retieved for greater than one hour, this information shall be entered in a Narrative Log Program, a format turnover and shift turnover sheet will be completed, including the togs signed over.
67. G2.l.321 Which one of the following completes the statement below?

lOP-JO, Standby Gas Treatment System Operating Procedure, prohibits venting the drywell and the suppression pool chamber simultaneously with the reactor at power because this would cause the:

A. unnecessary cycling of reactor building to torus vacuum breakers.

B. unnecessary cycling of torus to drywell vacuum breaker.

C. SBGT Train water seal to blow out of the trough.

D. pressure suppression function to be bypassed.

Answer: D K/A:

G2.1 .32 Ability to explain and apply system limits and precautions. (CFR: 41.10 /43.2 / 45.12)

RO/SRO Rating: 3.8/4.0 Tier 3 K/A Match: This meets the K/A because it is testing the ability to explain the system precaution.

Pedigree: Last used on 2012 NRC exam Objective:

Reference:

None Cog Level: Fundamental Explanation: Per OP-b, torus and drywell cannot be vented at the same time in Modes 1, 2 or 3. per the LER reference, this could result in bypassing pressure suppression function.

Distractor Analysis:

Choice A: Plausible because these vacuum breakers prevent drawing a negative pressure in the suppression pool. Cross connecting the drywell and the suppression pool free air space will not cause a negative pressure in the suppression pool.

Choice B: Plausible because this lineup equalizes pressure between the drywell and the suppression pool free air space since the vacuum breakers operate on a d/p between the spaces this would bypass them, not open them.

Choice C: Plausible because venting containment through large valves with an elevated pressure may blow out the SBGT water seal.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A

STANDBY GAS TREATMENT SYSTEM OPERATING lop-iD PROCEDURE Rev 66 Page4of49 1.0 PURPOSE

1. This procedure provides instructional guidance for operation of the Standby Gas Treatment System and its associated deluge system.

2.0 SCOPE

1. This procedure provides the prerequisites, precautions, limitations, and instructional guidance for startup, normal operation, shutdown, and infrequent operation of the Standby Gas Treatment System and its associated deluge system.

3.0 PRECAUTIONS AND LIMITATIONS

1. The Standby Gas Treatment System will N.QI automatically start if the control switch is in STBY 0
2. Venting the drywell and suppression pool simultaneously is NOT performed wflen the plant is in MODE 1, 2, or 3. {8.1.i} 0 STANDBY GAS TREATMENT SYSTEM OPERATING 1 OP-i 0 PROCEDURE Rev 66 Page 31 of 49

8.0 REFERENCES

8.1 Commitments

1. LER 1-97-01 1, Drywell and Torus Inerting/Deinerting Lineup Results in Unanalyzed Suppression Pool Bypass Path
68. G2.1.361 A core reload is in progress during a refueling outage. The initial loading of fuel bundles around each SRM centered 4-bundle cell was completed with all four SRMs fully inserted and reading 50 cps.

It is now approximately halfway through the core loading sequence and SRMs read 80 cps.

Which one of the following completes the statement below lAW OFH-1 1, Refueling?

Fuel movement must be suspended when any SRM reading first rises to upon insertion of the next fuel bundle.

A. 100 cps B. 160 cps C. 250 cps D. 400 cps Answer: B K/A:

G2.1 .36 Knowledge of procedures and limitations involved in core alterations. (CFR: 41.10 / 43.6 / 45.7)

RO/SRO Rating: 3.0/4.1 Tier 3 K/A Match: This meets the K/A because it is testing the fuel movement requirements that an RO would monitor.

Pedigree: New Objective: LOI-CLS-LP-305, Objectives 18 Given the conditions during a refueling outage state the operator actions required for rising SRM count rates and/or inadvertent criticality.

Reference:

None Cog Level: High Explanation: An increase in counts by a factor of two during a single bundle insertion is reason to suspend fuel movements. An increase by a factor of five from the baseline is also a reason.

Distractor Analysis:

Choice A: Plausible because this is a doubling of the baseline counts which is used for a different criteria for suspension of fuel movements.

Choice B: Correct Answer, see explanation Choice C: Plausible because this is an increase of the baseline counts by a factor of five which is a reason to suspend fuel movements.

Choice D: Plausible because this is an increase of the counts by a factor of five which is a reason to suspend fuel movements.

SRO Basis: N/A FH-11:

24. Suspension of fuel movement and notification of the Reactor Engineer is required if either of the following occur:

An SRM reading rise by a factor of two upon insertion of any single bundle. During a spiral reload, this restriction applies only after the initial loading of fuel bundles around each SRM is complete. During a Core Shuffle, this restriction does I apply to the SRM that is having an adjacent fuel bundle inserted or removed 0

  • An SRM rise by a factor of five relative to the SRM baseline count rate recorded on Attachment 6, Documentation for SRM Baseline 0
25. SRM count rate may drop to less than 3 cps during either of the following conditions:
  • With less than or equal to four fuel assemblies adjacent to the SRM and NO other fuel assemblies in the associated core quadrant 0
  • During a core spiral offload 0
69. G2.2.02 1 Unit Two is conducting a routine power reduction for rod pattern improvement.

The Reactivity Management Plan contains actions for the RO to insert a group of four rods from position 24 to position 12.

Which one of the following completes the statement below lAW AD-OP-ALL-0203, Reactivity Management?

The movement of these rods should be:

A. single notched for the entire movement.

B. continuously inserted to the final intended position.

C. continuously inserted to settle four notches prior to reaching the intended position and then single notched into the final intended position.

D. continuously inserted to settle one notch prior to reaching the intended position and then single notched into the final intended position.

Answer: D K/A:

G2.2.02 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. (CFR: 41.6 / 41.7 /45.2)

ROISRO Rating: 4.6/4.1 Tier 3 K/A Match: This question matches the KA because it tests the genetic requirements of control rod movement during any power level.

Pedigree: Bank Objective: LOl-CLS-LP-201-D, Obj. 22f Explain the following regarding AD-OP-ALL-0203, Reactivity Management: The procedural requirements for positioning intermediate control rods

Reference:

None Cog Level: High Explanation: If a rod is to be moved between 46 and 02 three notches or less, it must be single notched the entire move. When moving a control rod four notches or more, the control rod should be stopped one notch prior to reaching the intended position and then single notched into the final intended position.

Distractor Analysis:

Choice A: Plausible because this would apply if the movement was < four notches.

Choice B: Plausible because this would apply under emergency conditions.

Choice C: Plausible because the rod does have to be single notched into its final position but the rod can be continuously move if greater than four notches not for four notches.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A REACTIVIW MANAGEMENT AD-OP-ALL-0203 Rev. 2 Page 47 0190 5.2.8 [BWR] Single Recirculation Loop Operation

{7.1.5}

Standards

a. Single-Loop operation for extended periods of time is discouraged.
2. Expectations
a. Plant procedures that address Single Recircutation Loop Operation will identify applicable limitations and trip criteria.
b. For operations not covered by an approved procedure the Operational Decision Making process will be used to evaluate continued operation in Single-Loop.
c. The risk associated with single recirculation loop operations shall be carefully considered and appropriate contingencies will be developed.
d. Operator JIlT shall be conducted for planned Single-Loop Operations.

5.2.9 Control Rod Manipulations Standards

a. Ensure all control rod movements are made in a deliberate, carefully controlled manner while constantly monitoring nuclear instrumentation and redundant indications of reactor power and neutron flux.
2. Expectations
a. [BWRI To minimize the possibility of mispositioning a control rod when inserting or withdrawing to an intermediate position (notch positions 02 through 46), the following practices shall be followed:

(1) When moving a control rod four notches or more, the control rod should be stopped one notch prior to reaching the intended position and then single notched into the final intended position.

This guidance does not supersede any other requirement to single notch control rods.

(2) When moving a control rod three notches or less, the control rod should be single notched for the entire move.

70. G2.2.04 1 Which one of the following identifies the Unit Two Scram Immediate Operator Action that utilizes a different criteria for performance than on Unit One?

A. Tripping of the main turbine.

B. Tripping one of the running feed pumps.

C. Master level controller setpoint setdown.

D. Placing the reactor mode switch to Shutdown.

Answer: D K/A:

G2.2.04 Ability to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility. (CFR: 41.6 / 41.7 / 41.10 / 45.1 / 45.13)

RO/SRO Rating: 3.6/3.6 Tier 3 K/A Match: This meets the K/A because it is testing the differences between the Units Pedigree: Bank Objective: LOI-CLS-LP-300-C, Obj. 2 List the immediate operator actions for a Reactor Scram. (LOCT)

Reference:

None Cog Level: fund Explanation: On Unit Two the mode switch cannot be placed to shutdown until MSL flow is less than 3 Mlbms. This restriction does not exist on Unit One.

Distractor Analysis:

Choice A: Plausible because this is an immediate operator action.

Choice B: Plausible because this is an immediate operator action.

Choice C: Plausible because this is an immediate operator action.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A

Unit 2 Scram Immediate Actions (OEOP-O1-UG)

SCRAM IMMEDIATE ACTIONS Ensure SCRAM valves OPEN by manual SCRAM orARI initiation.

2. WHEN steam flow less than 3 x iO lb/br, THEN place reactor mode switch in SHUTDOWN.
3. iF reactor power below 2% (APRM downscale trip),

THEN trip main turbine.

4. Ensure master RPV level controller setpoint at +170 inches.
5. IF:
  • Two reactor feed pumps running AND
  • RPV level above +160 inches AND
  • RPV level rising.

THEN trip one.

Unit I Scram Immediate Actions (OEOP-O1-UG)

SCRAM IMMEDtATE ACTIONS

1. Ensure SCRAM valves OPEN by manual SCRAM or ARI initiation.
2. Place reactor mode switch in SHUTDOWN.
3. IF reactor power below 2% (APRM downscale trip),

THEN trip main turbine.

4. Ensure master RPV level controller setpoint at +170 inches.
5. IF:
  • Two reactor feed pumps running AND
  • RPV level above +160 inches AND
  • RPV level rising, THEN trip one.
71. G2.2.44 1 The OATC observes the loll j indications after initiating SLC during an ATWS.

18 SLC A/B z15 I SUB VALVE CONTINUITY 15-*

P P2 S j 12H SE I =-9 I SIC RIMP ?

C41CCOIA R)I SICC4I1B PIJIP 28 L 2X x

E-6 0

1 ri o 0 SLO PIU4P A&B C41S I

DrrTtO I SI_c

-I-A-RCA I U Which one of the following completes both statements below?

Squib valve (1) has failed to fire.

lAW 20P-05, Standby Liquid System Operating Procedure, the OATC is required to (2)

A.(l) A (2) place the CS-SI, SLC Pump A & B, in the PUMP B RUN position B. (1) A (2) leave the CS-SI, SLC Pump A & B, in the PUMP NB RUN position C. (1) B (2) place the CS-Si, SLC Pump A & B, in the PUMP A RUN position D. (1) B (2) leave the CS-S 1, SLC Pump A & B, in the PUMP NB RUN position Answer: C

K/A:

G2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 /43.5 /45.12)

ROISRO Rating: 4.2/4.4 Tier 3 K/A Match: This meets the K/A because it is testing knowledge of the indications and what action is required based on the system lineup.

Pedigree: Previously used on the 2014 NRC exam Objective: LOl-CLS-LP-005, Obj 13 -

Predict the effect of the following on the Standby Liquid Control System, and based on those predictions use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: a. Failure of one or both squib valves to fire.

Reference:

None Cog Level: Hi Explanation: The SLC squib valve continuity lights are normally lit and go out when fired on SLC initiation.

Per OP-05, if one squib valve fails to fire, two pump SLC operation may still continue provided reactor pressure is below 1184 psig, which it is not.

Distractor Analysis:

Choice A: Plausible because the student may think that the light is illuminated when the squib valve fires and securing 1 pump is correct.

Choice B: Plausible because the student may think that the light is illuminated when the squib valve fires and if reactor pressure was lower this would be correct.

Choice C: Correct Answer, see explanation Choice D: Plausible because the B squib did not fire and if reactor pressure was lower this would be correct.

SRO Basis: N/A NOTE: The SLC pump discharge relief valve should NOT actuate with two pumps operating and only one squib valve open unless reactor pressure exceeds 1184 psig, which is possible during an ATWS even with 10 SRVs open.

2. IF SLC A SQUIB VALVE CONTINUITY OR SLC B SQUIB VALVE CONTINUITY indicating light on Panel P603 remains on AND reactor pressure is greater than or equal to 1184 psig, THEN PERFORM the following:
a. PLACE SLC PUMP A & B Control Switch, C41-CS-$1, to the SLC PUMP A OR SLC PUMP B position.
b. ENSURE the selected SLC pump red indicating light on.
72. G2.3.12 I Two operators are required to enter a room that is posted as a Locked High Radiation Area (LHRA) to hang a clearance for scheduled work.

Which one of the following completes both statements below?

The radiation level at which a LHRA posting is required is (1) in one hour at 30 centimeters from the radiation source.

The LHRA key is obtained from (2)

A. (1) >l00mrem (2) the Shift Manager B. (1) >100 mrem (2) a RP Technician C. (1) >1000 mrem (2) the Shift Manager D. (1) >1000 mrem (2) a RP Technician Answer: D K/A:

G2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 45.9 I 45.10)

ROISRO Rating: 3.2/3.7 Tier 3 K/A Match: This question matches the KA because it is testing the rad requirements for entering a LHRA.

Pedigree: Bank (from Farley)

Objective: LOl-CLS-LP-201 -F, Obj. 10 Explain the requirement regarding control of High Radiation Areas per E&RC-0040.

Reference:

None Cog Level: Fundamental Explanation: Locked High Radiation Area (LHRA) criteria is an area, accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess of 1.0 rem (1000 mrem) (10 mSv) in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates or an area accessible to individuals with dose rates in excess of 1.0 rem per hour at 30cm from the radiation source or 30cm from any surface that the radiation penetrates but less than 500 rads in one hour at one meter from the radiation source or from any surface penetrated by the radiation. The Shift Manager has a LHRA key for emergency use.

Distractor Analysis:

Choice A: Plausible because this is the limit for a high radiation area not a LHRA. The Shift manager has a key for LHRA but it is for emergency use, not scheduled work.

Choice B: Plausible because this is the limit for a high radiation area not a LHRA. The second part is correct.

Choice C: Plausible because the first part is correct and the Shift manager has a key for LHRA but it is for emergency use, not scheduled work.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A

14. High Radiation Area (HRA): An area, accessible to individuals in which radiation levels from radiation sources external to the body coutd result in an individual receiving a dose equivalent in excess of 0:1 rem (100 mrem) (lmSv) in one hour at 30 cm from the radiation source or 30 cm from any surface the radiation penetrates.
15. Hot Spot (HS): An accessible, tocalized source of radiation with a contact dose rate of greater than 100 mrem per hour and greater than five times the general area dose rate at 30 cm.
16. Licensed Material: Source material, special nuclear material, or byproduct material received, possessed, used, transferred or disposed of under a general or specific license.
17. Locked High Radiation Area (LHRA): An area, accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess 011.0 rem (1000 mrem) (10 mSv) in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates or an area accessible to individuals with dose rates in excess of 1.0 rem per hour at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates but less than 500 rads in one hour at one meter from the radiation source or from any surface penetrated by the radiation.

ACCESS CONTROLS FOR HIGH, LOCKED HIGH, AND AD-RP-ALL2O17 VERY HIGH RADIATION AREAS Rev 2 Page 9of29 5.1 General Instructions (continued)

12. Entry into HRAs, LHRAs, or VHRAs require a briefing per AD-RP-ALL-2011, Radiation Protection Briefings. {7.1 .2}
13. HRA. LHRA less than 10 R/hr, and LHRA greater than or equal to 10 R/hr master keys may be under the control of the Operations Shift Manager for emergency usc.
73. G2.3.15 1 Which one of the following identifies the DW radiation value indicated above?

A. IOR/hr B. 20 R/hr C. -100 R/hr D. 200 R/hr Answer: D K/A:

G2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.12 /43.4 I 45.9)

RO/SRO Rating: 2.9/3.1 Tier 3 K/A Match: This question matches the KA as it requires knowledge of the DW rad monitoring system to answer question.

Pedigree: Bank Objective: CLS-LP-11.1, Obj. 03a Describe the function/operation of the following: Drywell High Range Radiation Monitors

Reference:

None Cog Level: Fundamental

Explanation: Drywell high range area monitors provide indications of gross fuel failure and are used to determine emergency plan emergency action level associated with abnormal core conditions.

With the function switch in the ALL position the upper (red) scale is used, meter readings are taken from the upper scale between 1 1,000,000 R/h. Current indication of 200 R/h Distractor Analysis:

Choice A: Plausible because this is the reading on the bottom scale.

Choice B: Plausible because if function switch is not taken into account the answer could be 20 RIh.

Choice C: Plausible because if the reading on the bottom scale is adjusted by afactorof 10.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A

74. G2.4.20 1 A transient has occurred on Unit Two with the following plant conditions:

RPV pressure 1000 psig Drywell ref leg area temp 197°F Rx Bldg 50 temp 135°F Wide Range Level 170 inches (NO26NB)

Shutdown Range Level 160 inches (NO27NB)

(REFERENCE PROVIDED)

Which one of the following completes both statements below concerning the level instruments that can be used to determine reactor water level lAW EOP Caution 1?

Wide Range Level instruments NO26NB (1) be used.

Shutdown Range Level instruments N027A/B (2) be used.

A. (1) can (2) can B. (1) can (2) can NOT C. (1) can NOT (2) can D. (1) can NOT (2) can NOT Answer: B K/A:

G2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

(CFR: 41.10/43.5/45.13)

RO/SRO Rating: 3.8/4.3 Tier 3 K/A Match: The question meets the KA because it is testing the knowledge of EOP Caution 1 which deals with the water level instruments availability to determine level.

Pedigree: New Objective: LOI-CLS-LP-300-B, Objective 16 Given Plant conditions, determine if the RPV water level instrument is providing valid trending information lAW Caution 1.

Reference:

Caution 1 (EOP-01-UG, Aft. 19, Aft. 22 & Att. 31 pages 1 and 2)

Cog Level: High Explanation: N026s can be used since reading >20 and RB 50 temp is <140 degrees and N027s cannot be used since in unsafe region for minimum indicated level

Distractor Analysis:

Choice A: Plausible because the first part is correct and if Attachment 19 is only looked at then this is plausible.

Choice B: Correct Answer, see explanation.

Choice C: Plausible because if the temperature was a little higher on the RB 50 foot this would be correct and if Attachment 19 only is looked at for the second part this would be correct.

Choice D: Plausible because if the temperature was a little higher on the RB 50 foot this would be correct and the second part is correct.

SRO Basis: N/A USERS GUIDE CEOP-Ol-UG Rev. 067 Page 90 of 156 AUACHMENT 22 Page 1 of 1

<<Shutdown Range Level Instrument (NO2YA, B) Caution>>

300 z

I[HHUH UHH1HHH

-J w

250 t[fi]Itl Ui

-J Ui U 200 z

UNSAFE 150 150 250 350 450 100 200 300 400 REFERENCE LEG AREA DRYWELL TEMP f°F

75. G2.4.27 1 A fire has been reported and confirmed in the turbine building breezeway.

A fire hose is being used to control/suppress the fire.

Which one of the following completes both statements below lAW OPFP-01 3, General Fire Plan?

The RO is required to sound the fire alarm and announce the location of the fire (1)

A call for offsite assistance to the Brunswick County 911 Center (2) required.

A. (1) ONLY once (2) is B. (1) ONLY once (2) is NOT C. (1) three times (2) is D. (1) three times (2) is NOT Answer: C K/A:

G2.4.27 Knowledge of fire in the plant procedures. (CFR: 41.10 /43.5 /45.13)

ROISRO Rating: 3.4/3.9 Tier 3 K/A Match: This meets the K/A because it is testing knowledge of the actions contained in the plant fire procedure Pedigree: New Objective: FPT-CLS-LP-205 Lesson plan discusses the actions for the control room but no objective is listed.

Reference:

None Cog Level: Fundamental Explanation: The operator aid (from the General Fire Plan, PFP-013) for the control room operators states to announce the fire location 3 times. The procedure also states to request off site assistance if a fire hose is used for extinguishing the fire.

Distractor Analysis:

Choice A: Plausible because EP announcements are performed once and the second part is correct.

Choice B: Plausible because EP announcements are performed once and the second part because the stem says that the fire is under control.

Choice C: Correct Answer, see explanation Choice D: Plausible because EP announcements are performed once and the second part because the stem says that the fire is under control.

SRO Basis: N/A GENERAL FIRE PLAN OPFP-013 Rev. 48 Page 270135 ATTACHMENT 2 Page 1 of 2

<<(Information Use) Control Room!Operator Fire Actions>>

Sound fire alarm, announce location of the fire 3 times, then El

  • Announce: El Fire brigade muster at the fire house.
  • IF fire is outside the Protected Area, THEN announce:

All personnel NOT involved in fire righting or direct support activities are to evacuate the involved area Immediately. El

  • IF tire is inside the Protected Area, THEN announce:

All personnel in the affected area are to evacuate the involved area immediately and report to your normal work location, If your normal work location is inaccessible, report to the O&M lunch room or TAC auditorium as conditions dictate. El

  • Announce:

Use of the PA and radio is restricted to emergency tire communications, except as directed by the Unit CRS for operational safety concerns. El

2. Announce the fire over Unit 1 and Unit 2 radio channels El
c. IF the investigating operator confirms a fire AND any of the following conditions exist, THEN immediately request off site assistance by calling 911: El
  • Extreme force is necessary to gain entry into fire area El
  • A fire hose is requited for fire suppression El
  • Fire is located outside the Protected Area, but within the Owner-Controlled Area El
76. S209001 I During a LOCA and LOOP on Unit One, the following pflant conditions exist:

An Emergency Depressurization has been performed due to RPV water level The Reactor Building -17 foot and 20 foot elevations are NOT accessible due to radiation levels.

ALL ECCS pumps are unavailable.

Which one of the following completes the statement below?

The CR5 will direct demin water injection to the RPV, lAW OEOP-01-LEP-01, Alternate Coolant Injection, Section:

A. 2.4.3.3a, Demineralized Water Actions, Inject demineralized water through Core Spray Loop A B. 2.4.3.3c, Demineralized Water Actions, Inject demineralized water through RHR Loop A C. 2.4.3.3d, Demineralized Water Actions, Inject demineralized water through HPCI D. 2.4.3.3e, Demineralized Water Actions, Inject demineralized water through RCIC Answer: A K/A:

209001 Low Pressure Core Spray System G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR: 41.10/43.5 /45.13)

RC/SRC Rating: 3.8/4.0 Tier 2 / Group 1 K/A match: This question meets the K/A because an emergency condition exists in the stem (ED followed by no HP injection/building inaccesible). In addition, the SRC is required to have knowledge of the local emergency procedure and that it contains actions the AC must take in the field in order to inject (AC must locally open demin keepfill bypass valves for various injection sources). Contrasting the given conditions, the procedural knowledge of AC field actions, and system knowledge of valve locations will have the operational effect of determining which injection source is viable.

Pedigree: New Cbjective: LCI-CLS-LP-300-J Cbj 4a Given plant conditions, determine which system should be utilized to restore RPV water level and/or pressure when executing the following:

a. Alternate Cooling Injection Procedure with ECP-01-LEP-01.

Reference:

None Cog Level: High

Explanation: Demineralized water injection requires knowledge from the LEP that the system Keepfill Bypass Valves are to be opened. The only keepfill Ipypass valve that is not inaccessible is the Core Spray Loop A Keepfill Bypass valve. Therefore, Core Spray Loop A is the section to use in order to inject demin water to the RPV.

Distractor Analysis:

Choice A: Correct answer, see explanation.

Choice B: Plausible because it is a section in 2.4.3, it is incorrect because the keepfill bypass valve for RHR loop A is inaccessible.

Choice C: Plausible because it is a section in 2.4.3, it is incorrect because the keepfill bypass valve for HPCI is inaccessible.

Choice D: Plausible because it is a section in 2.4.3, it is incorrect because the keepfill bypass valve for RCIC is inaccessible.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]. Requires the SRO to evaluate the emergency conditions in the stem, and contrast those conditions with the given local emergency procedure sections and then select the appropriate procedure section.

2.4.3 Deminerat;zed Watet Actions NOTE OE&RC-0040 states that the Shitt Manager isltj grant immediate access to a locket high raitiatiog area to mantain the health arid safety of citant personnel or the generat pubSc per IOCFRSd 54(s) using the keys m3rntarr5d in the Rathatori Prcitection key locker near the Unit 2 AOG panel.. ............... 0 1 (ELAYJiIIE a loclied hgh radiation area is entered, 1ttrout Riaton Protectcn support JJf promptly nobly RP ............................... 0 RO 2 MonitOr arid control MUtt tank lesisi greater than 14 teet ,..,,........ 0 A0 NOTE Oemmera5:ed water transfer pump capacity is 4C0 gpm 0 3 Perform for systems (f(J operating ,jf(2 available to provide Injection to RPV . 0 RO 3 triject demineralCed water thscsgh Core Spray Loop A (1) Ensure E21-FEOSA (inboard Injecticis \)v) OPEN 0 RO (2) Ensure E21FOO4A (Outboard Injection Vtv) OPEN 0 RO ALTERNATE COOLANT INJECTiON CEOP.dl-LEP-O1 Rev 33 Page 20 of 47 243 D5rnineratized Water ActIons (continued)

NOTE E2 l-FQ28A is located on Reactor Building 50. . .. .0 t3 Open B2t.FO2BA (Core Spray Loop A (esEtit Station B1pass Vatwe) 0 AO

Inject dertrinerakzed waler tlrED5h RI-IR Loop A (1) Enture El 1-FOISA (Inboard loecticel VIe) OPEN ci RD (2) Throttle El l-FO17A (Outboyd Injection tv) ci RD NOTE 1 l.FO)2A 61 1-FOBS ar 6l1-FCS6A are located on tIre KPCI nv2Zanoe .,.., ci (3) Open S I l-FCB2A RHR Lo A t(eeplD StaSoo Bpaso VawO) ci AD (3) Open El 13QBS 1RHR Sdem Demnora:e,1 Wa)

Fril Valve) ci AO 0 Inject Oemrflcralrzed waler tOroel3Ir HPl:

(1) Ensure E41-F012 ((-(PCI Minnow 8,assTcTorvs VlsI CLOSED ,,

0 RD (2) Al MCC XDA. Row HI. Cornpl 824 (HPCI Mitt Plo OPV To SuppChambcc Valve 641 -F0121 place tlleakef OFF AD MOTE E4l-VtOOctedeiNRld) ..., ,,. ci t3t Open 53 1-ViED CHPCI <ceptilt Slalon Bps Valeci .0 AD (3i EnsLceE4l-F47lPunrtoDischargevTv(OPEN RO e Inject demzreralred wJier thrcvli RCIC (II Enewe 651-FOb IRCIC Mill Flow Ejpass To Tceu5

\le)CLOSSD 0 80 3LTER1I3JE COOLAIJT INJECTION I OEOP-Ol-LEP-Ol Re-v 33 I I Po3e22cA47j

.4.3 Deminerairred Water Actions (continued)

(2) At MCC XDB, Row Hi, Conpt 637 IRCIC Mn FIll EpaSC To Supp Pod VIe ESi-FOtO). plice Efealer OFF ,.,..

. ci AD NOTE ESl-7OiSICcaledrflSRHR ...,,.,...,.

0

13) Open ES I -V70 (RCIC Keeplil Slatcet 61p355 Valvel ci AD (4) EnsureE5l-F012(Pumpcilschargevlv)OPEN ... .0
77. 5212000 1 Unit One is at rated power performing OPT-Ol .1.6, Reactor Protection System Manual Scram Test The Reactor Scram System A pushbutton has been depressed.

RPS Trip System A Scram Groups light for groups one, two, three, and four are illuminated (REFERENCE PROVIDED)

Which one of the following completes both statements below?

The scram pilot valve solenoids associated with these lights are (1)

Tech Spec 3.3.1 .1, Reactor Protection System Instrumentation, Condition B (2) required to be entered.

A. (1) energized (2) is B. (1) energized (2) is NOT C. (1) de-energized (2) is D. (1) de-energized (2) is NOT Answer: B K/A:

212000 Reactor Protection System G2.2.44Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41 .5 I 43.5/45.12)

ROISRO Rating: 4.2/4.4 Tier2lGroup 1 K/A match: The applicant must interpret the indications (scram group lights) in comparison with the expected conditions for the given action (Sys A pushbutton depressed). The applicant must then use that knowledge to determine whether the system meets the given limited condition for operability as stated in the TS.

Pedigree: New Objective: LOl-CLS-LP-003 Rev 3 Obj 27 Given plant conditions, determine whether given plant conditions meet minimum Technical Specification requirements associated with the Reactor Protection System.

Reference:

TS 3.3.1.1 Cog Level: High

Explanation: Part I When group lights are OFF that is indication that the solenoid valves are de-energized. All groups remain lit therefore, they are energized. Part 2 ONLY Condition A and C are required to be entered, due to a failure of the A3 scram channel. Only one required channel is inoperable, but a loss of manual scram function has occurred resulting in RPS trip capability not maintained.

Distractor Analysis:

Choice A: Part I is the correct Answer, see explanation. Part 2 is plausible because Groups I through 4 are also in trip System B (although for the SV-1 18s), a novice applicant may assume that failure of these solenoids in Trip System A would mean they would not function in Trip System B. In addition, one channel per trip system is required, and since there are only lights for groups 1 through 4 in both trip systems, a novice applicant may assume that all required channels are mop.

Choice B: Correct Answer, see explanation.

Choice C: Part I is plausible because ARI system uses energize to function valves. Part 2 is plausible because Groups 1 through 4 are also in trip System B (although for the SV-1 18s), a novice applicant may assume that failure of these solenoids in Trip System A would mean they would not function in Trip System B. In addition, one channel per trip system is required, and since there are only lights for groups 1 through 4 in both trip systems, a novice applicant may assume that all required channels are mop.

Choice D: Part us plausible because ARI system uses energize to function valves. Part 2 is the correct Answer, see explanation.

SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. Requires the SRO to evaluate the failure of PT for RPS, and select the appropriate> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> IS condition.

RPS nstrumentation 3.31.1 33 1-I e.IUIII H..n, v4ctiu 5ç.wy, fl4a,.rftu, APILCA3tE C39C11t445 UI221303 550.25113 IEFSSEI4CEO 01553 CHllsal ss 395215E0 .9 lee irEO e.511J)Z4Lt ftnCllce. c\rULUrK:ztlS 515mW .t2FCHD1 510.19551311 7 SariOSiVtrs 12 2 C 39 11115 WSLs,WflJ. 39 3111 U 39 311112 39 3.31 1 l U 2 H 3931115 39 31114 33 111113 39 311.111 S 119.tlezeVE.eCceas i%RrP 4 1 39 13115 ,ilC%r.Iaal 39 33154 39 3.11.1 12 39 331111 39 321111 39 1.1.111?

5 risbr.C39t3SdnFntCS,w& i25%RtP 2 3 39 1111 lSS1tuiu C4,UnU CS ft.,.w.4&w 33 531 1 5 39 111111 39 111113 39 331115 39 111 1 tI IC Sn+/-s 33391Sk3a.woPla39flI I2 I C 39 131112 145 39111111 5$ 1 H 3912i111 145 39 331 1 11 W*.U1,cn 12 1 2 5333115 145 53 131115 5 1 H 59 31115 145 39131115 Is 5119 ry caflU 1 455451 3o( 4 fin 39 t194S1,5 nfl flu. L RPS irsbtwentalon 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reacior Pmtecflon System (RPS) Msdnimentatlon LCO 3.3.1.1 The RPS instrurnenIatort for each runcton In Tame 3.3.1.1-I snail te OPERABLE.

APPLICABILITY: According to TaBe 3.3.1.1-1.

ACTIONS NOTE separate ConEton Intl) Is alisleed for eacfl channel, CONOITION REOUIRED ACTION COMPLETION TIME A. One or more required A.l Race charnel in h1. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> cflanne!s Iroperab:e.

fig A.2 NOTE 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Nol applcab15 for Functans 2.a, 213. 2,c, Zd. cr21 Race associated tltp system in tip.

fconUrued)

RPC Irne.ntation

.1 ACTIONtI CONDITION REOUIRED ACTION COMPLEtION TIME B. NOTE 6.1 Place charnel Irl one trip it hours Not appicabte for Functons system In trip.

2.3, 2.0, 2.c. 2.d, or If.

E One or mare Functions with 6,2 Place one trip system Ii 6 hourS one or more required trip.

channels Inoperable In both trip systems.

C. One or more Functions with 0.1 Restore RPO trip capabilIty. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> RPS top capability not maintained.

0. RequIred Action and 0.1 Enter the COnd.ton lunniedlately associated Completion Time referenced In or Condition A, 5, or C not Tabe 3.3.1.1-I for the met channel.

E. As required by Requred El Reduce THERMAL 4 hOurs Aciton 0.1 and referenced In POWER to 26% RTP.

Table 3.3.1.1-1.

(conUnued DEFINITION OF INSTRUMENT CHANNELS AND TRIP 001 SYSTEMS FOR SELECTED INSTRUMENTS Rev.

Page 58 of AUACHMENI Page 1 C7IA-S3A, B; C72A-S3A, B INSTRUMENT NUMBER: C7JA-S3A, B; C72A-S3A, B INSTRUMENT NAME: Manual Scram TS

REFERENCE:

3.3.1.1; TRM Table 3.3.1.f-1.f I TRIP CHANNEL: A3-S3A B3-S3B TRIP SYSTEM: A3 S3A B3-S3B TRIP LOGIC: A3 and B3 = Reactor scram Place channel in TRIPPED condition t)y: Pull fuse

I.

The Reactor Manual Scram relays deenergize the Scram pilot valve solenoids for the RPS Trip System.

  • The SV-f 17 valves are in RPS Trip System A.
  • The SV-f 18 valves are in RPS Trip System B.

Shorting links are nomially installed around the auxiliary trip relay contacts in the Manual Scram circuits allowing tliis Scram signal to t]e bypassed. These shorting links are located in the back panels and are color coded red for identification.

rTE: The shorting links are removed prior to and during the time any control rod is withdrawn (except for control rods removed per Technical Specifications) during operation in refueling mode or during shutdown margin demonstration.

There are two shorting links per RPS Trip System, a total of four for the entire Reactor Protection System.

SD-03 Rev. 12 Page 27 of 90

REACTOR PROTECTION SYSTEM MANUAL OFI-Ol .1 .fi SCRAM TEST Rev. 18 Page 4 of 11 1.0 PURPOSE This test is performed to determine the OPERAB1LITY of the Reactor Protection System Manual Scram function.

2.0 SCOPE This procedure performs the following:

  • A quarterly Channel Functional Test perTS SR 3.3.1.1.9 for Table 3.3.1.1-1 Function 11, Manual Scram.
  • Satisfies a portion of the 24 month TS SR 3.3.1.1.15 Logic System Functional Test fcr Table 3.3.1.1-1 Function 11. Manual Scram.

3.0 PRECAUTIONS ANO LIMITATIONS

1. A half-scram signal elI exist until RESET C 4.0 GENERAL INFORMATION The follcwfng annunciators will alarm during the performance of this test:
  • A-OS. 2-8, Reactor Manual Scram Sys B 5.0 ACCEPTANCE CRITERIA
1. This test may be consicered satsfauow when a] of the fcllcwiing criteria are met:
a. A tip is indicated on RPS A and alarmed on RTGB Panel H12-P803 when C71(C72)-S3A (Manual Reactor Scram System A) push butcn is depressed.
b. A tip is indicated on RPS S and alarmed on RTGB Panel H12-P603 when C71(C72)-S3B (Manual Reactor Scram System B) push button is depressed.
c. The Scram valve solenoids are DE-ENERG1ZED when the associated RPS is tripped.

REACTOR PROTECTION SYSTEM MANUAL OPT-01.1.6 SCRAM TEST Rev. 10 Page 6 of 1 1 7.0 INSTRUCTIONS 7.1 Test Preparation

1. Obtain Unit CR5 perwission to perform this test
2. Ensure all prerequisites listed in Section 6.0. Prerequisites are met NOTE The length of time a hall-scram is sealed-in to be minimized C
3. J[j during the performance of this procedure, the expected test results from a half-scram initiaton are NOT observed, THEN immediately reset the half-scram and notify the Unit CR3 72 Manual Scram A Test Depress C71(C72)-S3A (Manual Reactor Scram System A), push button and observe the following actons occur
a. Plant Process Computer Event Log displays Manual Scram Channel A Trip (Computer Point 0533)
b. IF the proper Plant Process Computer Event Log (Computer Point 0533) was jQ received in Step l.a.

THEN generate aWO

c. Manual Reactor Scram System A push button light comes ON
d. A-OS, 1-B, Reactor Manual Scram Sys A. ALARMS
e. RPS Trip System A Scram Group lights 1.2.3, and 4 located on Panel H12-P603 are OFF, indicating Scram valve solenoids are DE-ENERGIZED
f. RPS Trip System A Scram Group lights 1. 2,3. and 4 located on RPS A Panel Ht2-P8D9 are OFF, indicating Scram valve solenoids are DE-ENERGIZED -
78. 5215001 1 Unit Two is at rated power. A TIP trace is in progress.

IN-CORE (REFERENCE PROVIDED)

Which one of the following completes both statements below?

Tip Valves are Group (1) PCIVs.

Tech Spec 3.6.1.3, Primary Containment Isolation Valves (PCIVs), Condition(s) (2) is/are required to be entered.

A. (1) 2 (2) A ONLY B. (1) 2 (2) AandB C.(1) 6 (2) A ONLY D.(1) 6 (2) AandB Answer: A KJA:

215001 Traversing In-Core Probe G2.2.44Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 /43.5 /45.12)

ROISRO Rating: 4.2/4.4 Tier 2 I Group 2 K/A match: The applicant must interpret the given TIP system indications and controls to verify the status

of the system. The applicant must then compare those indications and controls with the required conditions for the given mode as stated in the TS to determine the appropriate condition.

Pedigree: New Objective: LOI-CLS-LP-009.5 Obj 9 Determine whether given plant conditions meet minimum Technical Specification requirements, including the Bases, associated with the Traversing Incore Probe System.

Reference:

TS 3.6.1.3 Cog Level: High Explanation: Part 1: Tip Valves are Group 2 PCIVs. Part 2: Squib Valve Monitor Light On indicates that the squib valve continuity is lost. Therefore, The Shear Valve in the Ball valve and shear valve assembly is inoperable. The ball valve remains operable. Therefore, Condition A is entered for one PCIV inoperable.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because squib Valve Monitor Light On indicates that the squib valve continuity is lost. Therefore, The Shear Valve in the Ball valve and shear valve assembly is inoperable. In addition, a novice applicant may assume that with the MODE switch in MANUAL, the ball valve would be inoperable as well.

(However, the only position of the MODE switch that would make the TIP inoperable is OFF.)

With Two PCIVS inoperable, condition B would be entered, Choice C: Part 1 is plausible because group six PCIVs also isolate on LL1 and High DW pressure. Part 2 is plausible because it is correct, see explanation.

Choice D: Part 1 is plausible because group six PCIVs also isolate on LU and High DW pressure. Part 2 is plausible because squib Valve Monitor Light On indicates that the squib valve continuity is lost.

Therefore, The Shear Valve in the Ball valve and shear valve assembly is inoperable. In addition, a novice applicant may assume that with the MODE switch in MANUAL, the ball valve would be inoperable as well. (However, the only position of the MODE switch that would make the TIP inoperable is OFF. ) With Two PCIVS inoperable, condition B would be entered, SRO Basis: Facility operating limitations in the IS and their bases. [10 CFR 55.43(b)(2)]. The SRO applicant is required to select the appropriate> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS condition based on the status of the TIP system indications.

Table 09.5-2 Valve Control Monitor Indications idication Comment Squib Monitor Light ON indicates that the iF Shear Vae squib circuit continuity has been lost Shear Valve Monitor Light ON indicates thatthe squlb charge in the TP Shear Valve has been detonated.

TIP Ball Valve OPEN Light ON indicates thatTF Ball Valve is OPEN.

TIP Ball Valve CLOSED ON indicates thatT Bait Valve is CLOSED.

Light lime DelayLight ON ndicatesthatthe TF Ball Valve wasOPEN within 6 seconds from when the detector left the in-shield position and The W Drive Motor should have stopped.

Purge Light dicates t1t the solenoid forthe hdexer Purge System should be energized OPEN.

Fuse F5 Continuity Light htdicates powerto PCIS Groi 2 Bus in drawer is available (Fuse F5 is not blon.

SD-09.5 Rev 7 Page 20 of 58

Table 09.5-3 TIP Drive Control Unit Indications Indication Comment DETECTOR POSITION Dynamic digital display of detector position.

(illuminated digits) (0001 reference point about one foot behind the Indexer; 9750 hi Shield position)

CORE LIMIT Stafic digital display of pre-programmed core top or (illuminated dirits) bottom limits of selected channel.

READY Light Indicates that Indexer is properly aligned to selected channel.

CORE TOP Light Detector is at top of core.

N CORE Light Detector is above core bottom hmiL N SHIELD Light Detector is in Shield Chamber.

SCAN Light Axial Flux Profile is being recorded.

LOW Speed Light Detector is being driven at 3 inches per second, (15 feet per minute).

REV (Reverse) Light Detector moving away from top of core.

flVD (Forward) Light Detector moving towards top of core.

VALVE Light ON if TIP Ball Valve is CLOSED.

Table 09.54 P601 TIP Indications Indication Comment TIP Valve Status Green Light

- Green Light ON indicates that each TIP Ball Valve is FUlL CLOSED.

TIP Valve Status Red Light

- Red Light ON indicates that a TIP Ball Valve is NOT FULL CLOSED.

SD-09.5 Rev. 7 Page 21 of 58

Low Speed OFF Makes tow-speed drive a function of detector position and independent of operator control.

ON Initiates continuous low-speed detector drive.

Core Limit TOP Permits digital display of selected channel pre-prog rammed top-core limit which corresponds to top of active fuel.

BOTTOM As above, except pre-programmed core bottom limit is displayed. (The core top and bottom numbers are different for each TIP channel because of different lengths of guide tube run).

Mode (Switch 5-7)

OFF Deenergizes power supplies in Drive Control Unit.

MANUAL Positions detector in conjunction with the FWD and REV position ot manual switch S-3.

AUTO Permits automatic mode ot operation when Auto start 5-2 is pressed.

Manual Valve Control (Spring Return to Closed Position)

CLOSED Pemilts TIP Ball Valve to open automaticaity when mechanism is operated.

OPEN Opens TIP Ball Valve without energizing the Drive in the TIP Drive Mechanism.

X-Y Recorder (Figure 09.5-14)

Alternate plotting capability via AFORA software.

Controls on the X-Y recorder drawer are as follows:

SD-09.5 Rev. 7 Page 24 of 58

4.0 SYSTEM OPERATION 4.1 Normal Operational Relationships TIP traces are required to he performed periodically. Technical Specifications require calitration of LPRM detectors at least once every effective full power month (EFPM). Traces are done to obtain new Gain Adjustment Factors (GAF) as calculated by the process computer for each LPRM. These GAFs can be used to adjust the current applied to the LPRM detectors. Current adjustment is required due to loss of detector sensitivity which results from exposure to the fission process in the core. The TIP Detector is run through the dry tube within the LPRM assembly containing the LPRM to tJe calibrated. A comparison is made between the TIP output and the existing LPRM reading and a GAF is calculated. The current applied to the LPRM detector is adjusted, it necessary.

TIP Detector calibration is also periodically required. Capability for calibration of TIP probes is provided by a common channel which can align each TIP to the center LPRM assembly (28-29) (Figure 09.5-15) and TIP Dry Tube. Each TIP is run through the center LPRM assembly. Readings from the TIPs are compared and the gains are adjusted to meet an average value of the four TIP readings and so that the gains for the TIP channels fall within a specified band.

The TIP System can be operated in an Automatic mode or a Manual mode.

OP-09.1 in conjunction with OENP-24,f 5 covers precautions, initial conditions, and specific instructions related to the operation of the TIP System.

Regardless of the mode of operation there are some precautions that must be observed. The TIP Machine should never be fumed off with the detector inserted past the TIP Ball Valve. This conditions prevents the isolation logic from retracting the detector and closing the TIP Ball Valve should an isolation signal be received. Also, the TIP Machines should not be left unattended if the detectors are in motion.

41.1 Automatic Operation Sequence The mode switch on the selected Drive Control Unit is placed in AUTO and the manual switch and low speed switch remain in OFF. This gives a permissive to the TIP logic to be wn automatically.

The auto start pushbutton (Drive Control Unit) is then pressed. The detector will automatically move from Shield Chamber to the entrance of the Indexer at low speed f37sec or 15 Wmin) and will then stop.

SD-09.5 Rev. 7 Page 27 of 58

3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PC IV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1 Primary Containment Isolation Instrumentation.

ACTIONS NOTES _________________

1. Penetration flow paths may be unisolated intemithently under administrative controls.
2. Separate Condition entry is allowed foreach penetration flow path.
3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.
4. Enter applicable Conditions and Required Actions oILCO 3.6.1.1, Primary Containment, when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.

CONDITiON REQUIRED ACTION COMPLETION TIME A. NOTE- A.1 Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Only applicable to penetration flow path by penetration flow paths with use of at least one closed two PC IVs. and de-activated automatic valve, closed manual valve, blind flange, or check valve One or more penetration with flow through the valve flow paths with one PCIV secured.

inoperable except for MSIV leakage not within limit AND PCIVs 3.6.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. NOTE B.1 Isolate the affected 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Only applicable to penetration flow path by penetration flow paths with use of at least one closed two PCIVs. and de-activated automatic valve, closed manual valve, or blind flange.

One or more penetration flow paths with two PCWs inoperable except for MSIV leakage not within limit.

fl I I,.,..J.,4. *1,..

ThELE 12-2 Primary Containment Isolatioa System GroW Isctation Insmimentation Setpoints ISOLATION ISOIA71ON TRiP SEIFOINT NOTES GROUP SIGNAL iect sc. Ac Aikwable ote1)

GroupS I-lirSteamflow s2Th% 0% NoteS Low Steam Pressure S3peg 70 ps I-içr Tuib Exh Pressure 6 psi9 5 psig -

Steam Lie kea K Twnp 175°F 165°F 4 SteamLireTcernelHgh s2lC°F 165°Fi1OF Note4 Mt Tenç Steam Line Tunnel dT Hi . SDF 47°F Note 4 Equip Area HTemnp 175F 165°F EquipAreadTKgh s&WF 47°F Group 6 Low Level #1 15Y 168 Ii,CryweIIPtessure l.Bpsrg 17psig Fbr Bklg Eahaust Hi Ract s 1 mR,Thr 4 mRhr I8ldgEshaustHiTerrp N/A 135P No1e6 K ktiin Stick Rad C*)CM COCM Note2 Grtup7 LowSteamPressureAND a lQ4psig 115 p&g KyUPressure sl.8pe 1.7i:rsi GroupS LowLevel#1 153 166 Ktj Steam Done Pnture ° 137 psici 130.8 peg Group 9 Low Steam Pressure AND aS3 psg 70 psig K [ywe Pressure 1.8 peg 1.7 psig Group 10 Low Level #3 N/A 45 KgIi Drywefl Pressure AND 17 peg Low Reactor Ptessure 41D peg Note 1: AJI Actuaf vues from ThM Note 2: Smack radiaton high level is c craced r accordacice with the OSsite Dose CajIticn Manu.

Note 3: Me, a 28.5 mrinute tine delay Note 4: Mer 27 minute time delay Note 5: Mer a 5 second time delay Note 6: SpecilIc °Actu vues from EOP Users Guide. Allathrrent I JSD-12 Rev.11 Page84c205J TABLE 12-2 Primary Containment Isolation System Gmq Isolation Insumenbtion Selpoints ISOLATION ISOLATION TRiP SETPOINT NOTES GROUP SIGNAL Tech Spec Actual Aicwable (Note 1)

Value Grc1jpl LowLevel#3 2+13 45 Main Steam Le Hi s 197°F lçC°F Ten-c ThrbineBAreaHiTerrp N/A lt(IP MainSteamLwreHighRow 13S% 137%

NotinRUN 33% 30% Uift2 Low Condenser Vacuum 7.5 Hg 1JHg Low Steam Pressure a 825 psg 835 peg Group 2 Low Level #1 a +153 166 Kgh Drywe Pressure 7.8 peg 1.7 psig Group 3 Low Level #2 a 101 105 Kçi Duff Row s 73 9pm 43 m Note 3 keaHiiTemp sff°F 140°F Area Vent dT f s 50°F 47°F 1-ELBlsolation s12)°F 115F HXOutietTerrpHgh N/A 135°P Note6 SLClrutiation N/A N/A Group 4 ugh Steam Flow 275% Z% NoteS LowSteamPressure 104psig llSpsg KghTsebExhPreseure 9psig 7psig Steamt.ineAreaKTernp s27°F 165°F SteamLineTunnelHigh £2C0°F 165°F/190°F Ant Teirc SteamLineTunneldTHigh sSO°F 47°F EciipAseaHi&rTemp 175°F 165°F

79. 5262001 1 Unit One is operating at rated power.

Unit Two is in MODE 5 with UAT backfeed established.

A main generator backup lockout occurs on Unit One.

(REFERENCE PROVIDED)

Which one of the following completes both statements below?

All four diesels (1) automatically start.

lAW Unit One Tech Spec 3.8.1, AC sources Operating, Condition E (2) required to be entered.

A. (1) will (2) is B. (1) will (2) is NOT C. (1) will NOT (2) is D. (1) wilINOT (2) is NOT Answer: D KJA:

262001 AC. Electrical Distribution A2 Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) 09 Turbine/generator trip RO/SRO Rating: 2.9/3.0 Tier 2 I Group 1 K/A Match: This meets the K/A because the candidate is required to predict the status of the EDGs and determine whether the appropriate IS condition is entered.

Pedigree: New Objective: LOl-CLS-LP-050 Obj 17 Given plant conditions, determine the required action(s) to be taken in accordance with Technical Specifications associated with the 230 Ky Electrical Distribution System. (LOCT) (SRO Only)

Reference:

T.S. 3.8.1 (blank out the LCO statement and Applicability)

Cog Level: High Explanation: Part 1: DGs do not auto start on a generator backup lockout. Part 2: Condition E is not entered because a loss of two offsite circuits has not occurred. Only one offsite circuit is lost, the Unit One UAT.

Distractor Analysis:

Choice A: Part 1 is plausible because all four DGs start on a generator primary lockout. Part 2 is plausible because if UAT was not in backfeed on Unit Two this would be the case.

Choice B: Part 1 is plausible because all four DGs start on a generator primary lockout. Part 2 is plausible because it is correct, see explanation.

Choice C: Part I is plausible because it is correct, see explanation. Part 2 is plausible because if UAT was not in backfeed on Unit Two this would be the case.

Choice D: Correct Answer, see explanation.

SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)f2)]. Part two requires knowledge of the TS 3.8.1 bases to determine if Unit Two on UAT backfeed can qualify as an offsite source. In addition it requires the appropriate determination of the TS condition application.

3.2.6 Main Transformer, UAT, and Main Generator Protection

1. The Main Generator, MPT, and UAT are alt three protected by the Generatorfrransformer Primary (86GP) and Backup (86GB) Lockout Relays.

The Main Generator is provided additional protection by the Main Generator Differential Lockout Relays (8GG).

a. Generatorifransformer Primary, Backup, and Main Generator Differential Relay trip actions are as follows:

(1 )Main Generator Output breakers trip and lock out.

f2)Main turbine trips.

(3)Main generator exciter fteid breaker trips and locks out (4)UAT 4160 supply breakers to B, C and D buses trip and lock out (5)SAT feeders to C and D buses auto close.

(6)Four diesel generators auto start for the Main Generator Differential Lockout or the Generator/Transformer Primary Lockout. They do not auto start for a Generator/Transformer Backup Lockout SD-50 Rev.23 Page4Gofl4O

Offsite power is supplied to the 230 kV switchyards from the transnjission network by eight transmission lines. From the 230 kV switchyards, two qualified electrically and physically separated circuits provide AC power, through either a startup auxiliary transformer (SAT) or backteedthg via a unit auxiliary transformer (UAT), to 4.16 kV BOP buses. A single circuit path fmastertslave breakers and interconnecting cables) from each BOP bus provides offsite power to its associated downstream 4.16kV emergency bus. A detailed description of the offsite power network and circuits to the onsite Class IE emergency buses is found in the UFSAR.

Sections 8.2 and 8.3 (Ref. 2).

A qualified offsite circuit consists of all breakers, transformers, switches, interrupting devices, cabling, and controls required to transmit power from either 230 kV bus (bus A or B) to the onsite Class 1 E emergency buses.

The Unit I main generator provides the nomial source of power to 4.16 kV emergency buses El and E2 via its respective UAT. The Unit 2 main generator provides the normal source of power to 4.16 kV emergency buses E3 and E4 via its respective UAT. In the event of a Brunswick Unit I B 3.8.1-1 Revision No. 31 I AC SourcesOperating B 3.8.1 BASES BACKGROUND unit trip, an automatic transfer from the normal circuit (main generator (continued) output via the UAT) to the respective unit SAT occurs resulting in the SAT supplying power to two 4.16 kV emergency buses. As such, the Unit 1 SAT provides the preferred source of power to emergency buses El and E2 and the Unit 1 UAT (backfeed mode) is the alternate source of power to emergency buses El and E2. The Unit 2 SAT provides the preferred source of power to emergency buses E3 and E4 and the Unit 2 UAT (backfeed mode) is the alternate source of power to emergency buses E3 and E4. Each UAT can only be considered a qualified offsite source if it is capable of being powered from the 230 kV switchyard fRet 3).

AC SourcesOperating 3.81 18 ELECTRICAL POWER SYSTEMS 181 AC SourcesOperating LCO 3.8.1 The folleing AC eledrical power sources shall be OPERABLE:

a. Two Unt 1 qualified circuits between the offste transmission network and the onste Class 1E AC Eledrical Pawer Distribution System:
b. Four diesel generators (DGs): and
c. Two Unit 2 qualified circuits between the offsite transmission network and the onste Class 1E AC Eledrical Power Distribution System.

APPLICABILITY: MODES 1,2, and 3.

ACTIONS LCO 3.O.4.b is not applicable to DG5.

E. Two or more offsfte drcuits E.1 Declare required feature(s) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from inoperable for reasons other inoperable when the discoverj of than Condition B. redundant required Condition E feature(s) are inoperable. cencurrent with inoperability of redundant required feature(s)

AND E.2 Restore all but one offste 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> drcuit to OPERABLE status.

(centinued)

Brunswick Unit 1 3.8-5 Amendment No. 284

80. 5239002 1 Unit One was operating at power when a Group I isolation and reactor scram occurred.

Reactor pressure is 950 psig and being manually controlled by SRVs.

An SRV is stuck open with a stuck open SRV tailpipe vacuum breaker.

Torus and Drywell sprays have been initiated lAW PCCP (REFERENCE PROVIDED)

Which one of the following completes both statements below?

The SRV is discharging through the open vacuum breaker directly into the (1)

The highest EAL classification for this event is a(n) (2)

A. (I) drywell (2) Alert B. (1) drywell (2) Site Area Emergency C. (1) suppression chamber air space (2) Alert D. (1) suppression chamber air space (2) Site Area Emergency Answer: A KJA:

239002 Safety Relief Valves A2 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 I 45.6) 01 Stuck open vacuum breakers ROISRO Rating: 3.0/3.3 Tier 2 I Group 1 K/A match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL.

The applicant must determine that the effect of the stuck open vacuum breaker on the SRV is that it is now bypassing the pressure suppression function and discharging directly into the DW.

The applicant must also determine that the ultimate effect of the given conditions is that the RCS barrier has been lost, and that the consequences of this conditions and its potential effects on the health and safety of the public are mitigated by declaring an ALERT lAW with OPEP-2. 1.

Pedigree: New Objective: LOl-CLS-LP-020 Obj 1 5e Given plant conditions, predict how ADS/SRVs will be affected by the following: Failure of the SRV tailpipe vacuum breakers.

Reference:

OPEP-02.1

Cog Level: High Explanation: Part 1: SRV tailpipe vacuum breakers relieve to the drywell, therefore a stuck open tailpipe breaker with a stuck open SRV would discharge steam directly into the drywell. Part 2: With a stuck open relief valve and stuck open vacuum breaker, a LOCA is occurring. With Torus pressure sufficient to warrant drywell sprays (1 1.5 psig) drywell pressure has more than exceeded 1 .7 psig. Therefore the conditions for a loss of the RCS barrier ( Primary Containment Pressure >1 .7psig due to RCS leakage) are met and an ALERT should be declared based on FA1.1 loss of RCS battier.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Part I is plausible because it is correct, see explanation. Part 2 is plausible because a group 1 isolation has occurred which is a primary containment isolation signal, an unisolable LOCA is occurring, and a novice candidate might assume that this meets the conditions for a loss of the containment barrier (Unisolable direct downstream pathway to the environment exists after primary containment isolation signal), this loss coupled with the loss in the explanation would meet the criteria for a loss of two barriers, and a declaration of an SAE lAW FS1 .1 They also might think that this condition would not be consistent with a LOCA.

Choice C: Part 1 is Plausible because with a failed open Suppression Chamber to Drywell Vacuum Breaker would allow DW steam to go directly to the suppression Chamber Air Space. in addition, a failed open SRV with a broken tailpipe and broken downcomer would directly pressurize the torus air space. Part 2 is plausible because it is correct, see explanation.

Choice D: Part 1 is Plausible because with a failed open Suppression Chamber to Drywell Vacuum Breaker would allow DW steam to go directly to the suppression Chamber Air Space. in addition, a failed open SRV with a broken tailpipe and broken downcomer would directly pressurize the torus air space. Part 2 is plausible because a group 1 isolation has occurred which is a primary containment isolation signal, an unisolable LOCA is occurring, and a novice candidate might assume that this meets the conditions for a loss of the containment barrier (Unisolable direct downstream pathway to the environment exists after primary containment isolation signal), this loss coupled with the loss in the explanation would meet the criteria for a loss of two barriers, and a declaration of an SAE lAW FS1 .1. They also might think that this condition would not be consistent with a LOCA.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)f 5)] The SRO applicant is required to select the appropriate ALERT emergency classification based on the given conditions which equates to a loss of the RCS barrier.

U)

C O

C U)

-no

-c CD C.,

3 0

-U a,

CD WPPRZS 0l 0

0) 1J permissives.

4.22 Failure of the SRV Tailpipe Vacuum Breakers In the event an SRV tailpipe vacuum breaker fails in the open position.

the result isa direct path of steam from the reactor to the drjwell (Le.,

LOCA) In the event an SRV vacuum breaker fails in the closed position, the result is the possible creation of a vacuum in the tailpipe upon closure of an SRV, resulting in the drawing of water into the SRV tailpipe.

SD-20 Rev. 3 PAGE 27 of 62

F5111112131 I FAIiIII2I3I Loss or potential loss of any two bamers (Table F-i) Any loss or any potntiaI toss of either Fuel Clad or RCS (Table F-i) rable F-I Fission Product Barrier Threshold Matrix Reactor Coolant System Barrier Containment ss Loss Potential Loss Loss

1. RPV level cannot be restored and red maintained> TAF or cannot be ie determined
1. UNlSOLA8LEbreakinanyofthe 1. UNlSOLABLEprimarysystem 1. UttlSOLABLEprimarysystemleakagethat following: leakage that results in exceeding results in exceeding one or mcre Secondary

- Main steam line EITHER of the following: Containment area temperature Maximum

- HPCI steam line One or more Secondary Safe Operating Limits

- RClCsteamline Ccntainmentarearadalicn OEOP-Q3-SCCPTableSC-1)

- RWCU I Maximum Normal Operating Limits

- Feedwater (OEOP-03-SCCP Table SC-3)

  • One or more Secondary
2. Emergency Depressurization is Containment area temperature required Maximum Normal Operating Limits I (OEOP-O3-SCCP Table SC-i Primary Containment pressure I. UNPLANNED rapid drop in Pnmary 1

Contarriment pressure following Pnmary

> 13 psi9 due to RCS leakage Containment pressure rise Nxre

2. Primary Containment pressure response not consistent with LOCA conditions I - Drywell radiation >27 Rihr with reactor shutdown Ncr-c NJce
1. UNISOLABLE direct downstream pathway to the environment exists after Primacy Containment isolation signal

JIU dSSUITIptIUtIS COflS[Sttlt Wi[n uje L)b#-LuLp aIi.uysIs.

4.2.7 Containment Response with Vacuum Breaker Failure

1. Suppression Chamber to Drywell Vacuum Breakers Failed Open Steam flows from drywell to suppression chamber through the open vacuum breakers equalizing pressure between the two immediately.

The steam is not forced through the water of the suppression pool:

therefore it will operate only as a surface condenser. As a result, the drywell pressure will probably exceed the design pressure. To prevent this occurrence, light indication is provided for each vacuum breaker. It indicates if the va]ve is off its seat and is displayed in the Control Room.

2. Suppression Chamber to Drywell Vacuum Breakers Failed Closed The steam in the drywell will condense and drjweLl pressure will decrease. With vacuum breakers failed shut, pressure cannot equalize between the suppression pool and the dryNell. The pressure in the drywell may decrease such that the suppression chamber to drywell differential pressure limit (10 psid) is exceeded.

Vent pipe buckling can occur if suppression chamber pressure is 10 psia greater than drywell pressure. To prevent this situation, the vacuum breakers are operationally checked periodically and 133%

capacity is provided with ten vacuum breakers.

SD-04 Rev. 9 Page 48 of 103

81. 52190001 Unit Two is operating at rated power with RHR Loop A operating in suppression pool cooling mode.

A-01 (2-8) RHR Relay Logic Pwr Failure, is in alarm due to a blown fuse affecting RHR Logic A ONLY.

(REFERENCE PROVIDED)

Which one of the following completes both statements below?

lAW Tech Spec 3.3.5.1, ECCS lnstrumentation, required channels (1) required to be placed in trip within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If a LOCA signal were to occur, 2-El I -FOl 5A, Inboard Injection VIv, (2) open automatically on low reactor pressure.

A. (1) are (2) will B. (1) are NOT (2) will C. (1) are (2) will NOT D. (1) are NOT (2) will NOT Answer: A K/A:

219000 RHR/LPCI: Torus/Suppression Pool Cooling Mode A2 Ability to (a) predict the impacts of the following on the RHR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) 12 Valve logic failure RO/SRO Rating: 3.0/3.1 Tier 2 / Group 2 K/A Match: This meets the K/A because it is testing whether RHR in suppression pool cooling mode will transfer to LPCI injection mode with a failure of divisional valve logic, and whether the LCD action statement for condition B should be applied.

Pedigree: New Objective: LOl-CLS-LP-017 Obj 7. Given plant conditions, determine if the RHR system should automatically initiate in the LPCI mode. (LOCT)

Reference:

1.8. 3.3.5.1 Cog Level: High Explanation: Part 1: With the plant at rated power and torus cooling in service on loop A, RHR loop A is not

in its normal standby lineup. The loss of Div I RHR relay logic power, will result in the failure of the A loops suppression cooling valves (f024128) to automatically close on a LPCI initiation.

However, the FOl 5A will auto open from the div 2 logic. Part 2: The loss of power to the Div I RHR logic will result in the loss of the functions applicable to condition C.

Distractor Analysis:

Choice A: This is the Correct answer, see explanation.

Choice B: Part 1 is plausible because a novice candidate might believe the note that required action B.2.

is only applicable to functions 3a and 3b also applies to condition 6.3. Part 2 is plausible because it is correct, see explanation.

Choice C: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because a candidate might believe a loss of Div I logic would prevent the auto function of the FOl 5a, since the F024 and F028 will not automatically close under these conditions.

Choice D: Part 1 is plausible because a novice candidate might believe the note that required action B.2.

is only applicable to functions 3a and 3b also applies to condition 8.3. Part 2 is plausible because a candidate might believe a loss of Div I logic would prevent the auto function of the FO1 5a, since the F024 and F028 will not automatically close under these conditions.

SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. The SRO applicant is required to select whether an action statement is applicable given a loss of Div I RHR logic.

ECCS Instrumentation 3.3.5_I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

8. (continued) 8.2 ------NOIE-----

Only applicable for Functions 3.a and lb.

Declare High Pressure 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from Coolant Injedion (HPCI) discoveryof loss of System inoperable. HPCI initiation capability AND 8.3 Place channel in trip- 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. As required by Required Ci --NOTES--

Actionki and referenced in I. Only apphcable in Table 3.3.5.1-1. MODES 1, 2, and 3.

2. Only applicable for Functions 1.c. I.d. 2.c.

2.d. and 2.f.

Declare supported 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from featu to(s) inoperable when discovery of less of its redundant feature ECCS initiation capability initiation capability is for feature(s) in both inoperable. dMsions AND C2 Restore channel to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

(canfin ued)

ECCS Instrumentation 13.51 Tati*335.1-1 )D1OIM 5 3fl C.-o Cc,s r S rza Lcai.j*t.z.n

. I APJCAaL5 COJDITfl1S MCES REaLR43 RSRERS3CD OR OThER C&NNE.S SIECIFIED 1R R5-ZJ SyEtLftia, AUZ4/.2LE HJICTIOJ CaDmO JCTl N AC71Y *1 REJR54Ewr3 t Co-ofl-aen a xVozeWa+/-oLa,cLcrw 123, 4 3 SR 3.311,1 E13n100 4S!5* SR33512 f**g3 SR 33.51 .3 SR 33514 SR 3351.5 S 113 4 6 SR 3351.1 SR 3.3.512 SR 35.513 SR 35514 SR 33515 c RaSx Eoai&aaest.Law 123 4 C SR 3.351 I SR 3.3512 aid SR 33.51.3 SR 3351.4 SR 33.51.5 45 4 6 SR 33.51 I SR 33112 aid SR 33.513 SR 3 3 51 4 SR 33.513

4. Szra,FapS5a1_Tci.Dta, 123, 2 C SR 3.3114 wl4oxn3 a1 4&54 lpeepac SR 13513 aid SR 3)516 %16w1n3s 2 Low Coon1 lseaarillpClI Sylom a 1,23, 4 5 SR 3.3.51.1 L.a 3 3S SR 33.512 SR 33.51.3 SR 33513 SR 3351.5
5. volRw.noWØ 123 4 6 SR 33111 tl43; SR 33112 SR 3351.3 SR 33512 SR 3351.5 flae3

I Page29of34 ATtACHMENT 4 Page 1 ol 2

<<Plant Effects from Loss of DC Distribution Panel 3A(4A)>>

RCIC: Will NOT shutdown on reactor high water level, inboard isolation logic INOPERABLE (E5 1-F007, -F03 1, and -F062 will J4QI auto close). Valves E51-F005 and -F025 fail closed.

ADS: ADS Logic B is INOPERABLE. ADS will initiate from ADS Logic A if Core Spray Pump B or both RHR Loop B pumps are operating.

HPCI: Will NOT auto initiate, outboard isolation logic INOPERABLE (E41-F003, -F041, and -F075 wiILNQI auto close), HPCI flow controller and EGM INOPERABLE (no tlow control or indication),

HPCI trip logic INOPERABLE, valves E4f-F053, -F054, and -F026 fail closed. HPCI isolation is required in accordance with APP f(2)-A-Q1 5-5.

CS Loop A: Will NOT auto initiate (manual operation possible but minimum flow valve will NOT auto open, and injection valves can NOT be opened simultaneously).

RPS Logic A: Will NOT have 10 second time delay prior to reset of full scram, power lost to backup scram valves.

RI-JR Loop A: Will auto initiate from RHR Logic B, however the following effects exist: Pumps can NQI be restarted if stopped by control switch, pumps will NI trip on No Suction Path Interlock, LOCA interlocks NOT functional, mm flow valve will .tIQI auto open, Loop A Containment Spray can NOT be initiated. If a toss of DC Distribution Panel 38(48) has also occurred, RHR Loop A will NOT auto initiate (manual operation possible).

82. 52610001 Unit Two is operating at rated power.

Subsequently, a Div I pneumatic leak occurs causing drywell pressure to rise to 1 .9 psig.

Which one of the following completes both statements below?

The SBGT trains (1) running.

The Div I pneumatics are requited be isolated lAW (2)

A. (1) are NOT (2) OEOP-01-SEP-16, Drywell Systems Isolation B. (1) are NOT (2) OAO P-20 .0, Pneumatic (Air/Nitrogen) System Failures C. (1) are (2) OEOP-01-SEP-16, Drywell Systems Isolation D. (1) are (2) OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures Answer: C KJA:

261000 Standby Gas Treatment System A2 Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 /45.6) 09 Plant air system failure RO/SRO Rating: 2.4/2.6 Tier 2 I Group 1 K/A match: A failure of plant air in the drywell would require that the drywell be vented due to increasing DW pressure. In this case, since the impact of the failure is DW pressure increasing> 1 .7 psig, the specific impact on SBGT is that it can not be used to vent containment. The procedure to mitigate the consequences of the failure is the leak is required to be isolated using SEP-16.

(NOTE: the conditions are NOT entry conditions for the procedures)

Pedigree: New Objective: LOl-CLS-LP-046A Obj 13. Predict the effect that a loss or malfunction of the Pneumatic System would have on plant operation.

Reference:

None Cog Level: High Explanation: With drywell pressure >1.7 psig, 2OP-10 cannot be used to vent the drywell. SBGT has automatically started at 1.7 psig. With drywell pressure> 1 .7 psig, PCCP directs the use of SEP-16 to isolate containment leaks, and SEP 16 contains the steps to isolate DIV I pneumatics.

Distractor Analysis:

Choice A: Part 1 is plausible because PCCP directs 20P-10 to control DW pressure <1.7 psig. Part 2 is plausible because it is correct, see explanation.

Choice B: Part 1 is plausible because PCCP directs 20P-1 0 to control DW pressure <1.7 psig. Part 2 is plausible because AOP-20 directs the isolation of various pneumatic sources, but while executing the EOPs the EOPs take precedence.

Choice C: Correct Answer, see explanation.

Choice D: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because AOP-20 directs the isolation of various pneumatic sources, but while executing the EOP5 the EOPs take precedence.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]. The applicant is required to select the appropriate procedure for venting the drywell and isolating pneumatics based on the given conditions.

DRWVELL SYSTEMS ISOLATION OEOP-0f -SEP-f 6 Rev. 0 Page 6 at 7 2.2 Drywell Pneumatics Isolation 2.2.1 Manpower Required

  • 1 Reactor Operator 2.2.2 Special Equipment None 2.2.3 Operator Actions NOTE If both divisions are isolated MSIV and SRV operation will be limited to accumulators C IF Division I pneumatic leakage suspected, THEN:
a. Notify CRS C RD
b. Close RNA-SV-5262 (Div I Non-lntrpt RNA) C RD
c. Close RNA-SV-5253 (Dlvi Bu N2 Supp To DW lsd VIv) C RD

? IF fliuicinn II niimti Ik+/-n cI,np4fcd

STANDBY GAS TREATMENT SYSTEM OPERATING 20P-f 0 PROCEDURE Rev. 81 Page 19of49 6.3.2 Venting Containment Via SBGT Date/Time Started_______________

Confirm the following Initial Conditions are met:

  • Drywell pressure has risen to greater than 0.15 psfg
  • SBGT System is in STANDBY in accordance with Section 6.1.1
  • One of the following:

Plant stack radiation monitor is in service and CAC-CS-5519 tCAC Purge Vent Isol Ovrd) is in OFF 0 E&C has sampled the drywetl atmosphere and has determined that it is suitable for release

  • Unit CRS approval is obtained prior to venting ci 1

Control corlrmertpcessxe bec I.? pig uin 53GTp GP-iO.

/WHEN\

I==I tlowl.7pelg N THEN /

IF THEN 1e65L15 tedudlcn Is rered Vent roninnt per EOP-1-EP-O1

  • Exceed ol?Ste raoacMty rel3ae r If oct maLn ateq.9e coce neosssa,y
  • IE p etrmtc upy dmri a fe tatIabon dose EOP-Oj4O-CO ienng CAPLOT oe psetcaTied Pertomi EDUGOO vert ccni9rrerIt

çLig pceiecits roo1r.mern Agn trec5ta po.eet per EOP-O1-SEC-14

  • 75KWfeqtred e dmç to 2.5 pelQ Teimmate redr5 tr2.SpeIg Terminate cy.e1 sproy Isolate ten conlarinent Ies Sources per EOt1-5EP-16, -...

. Rero seas I *

  • .reed-e1ec I RFFORF
83. S271000 1 Unit Two is operating at rated power.

UA-48 (5-4) AOG System Bypass, has been alarming for 1 minute due to High-High off gas flow (REFERENCE PROVIDED)

Which one of the following completes both statements below?

AOG-XCV-142, Guard Bed Isolation Valve, (1) automatically close.

ODCM 7.3.10, Gaseous Radwaste Treatment System, Condition A entry (2) required.

A. (1) will (2) is B. (1) will (2) is NOT C. (1) will NOT (2) is D. (1) will NOT (2) is NOT Answer: C K/A:

271000 Offgas System A2 Ability to (a) predict the impacts of the following on the OFFGAS SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41 .5 I 45.6) 10 Offgas system high flow RO/SRO Rating: 3.1/3.3 Tier 2 /Group2 K/A match: The applicant is required to predict the status of the AOG system (XCV-142) based on high-high offgas flow, and the required procedural actions (i.e ODCM).

Pedigree: New Objective: LOl-CLS-LP-030 Obj 9 Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM and COLR, determine the required action(s) to be taken in accordance with Technical Specifications, the TRM or ODCM associated with the Condenser Air Removal/Augmented Offgas System.

Reference:

ODCM 7.3.10 Cog Level: High

Explanation: Part 1: With XCV-142 remaining open the probable cause for UA-48 (5-4) is off gas flow high, all other conditions (besides circuit failure) for this alarm would cause a closure of the XCV-142. Part 2: with UA-48 (5-4) in alarm, HCV-102 is open. This would bypass the AOG portion of the Gaseous Radwaste Treatment System, leading to reduced hold up times and increased main stack rad levels. ODCM 7.3.10 requires AOG in operation so the comp measure is required.

Distractor Analysis:

Choice A: Part 1 is plausible because high-high cooler condenser condensate level would also cause UA-48 (5-4) to alarm, but it would also close the XCV-142. Part 2 is plausible because it is correct, see explanation.

Choice B: Part 1 is plausible because high-high cooler condenser condensate level would also cause UA-48 (5-4) to alarm, but it would also close the XCV-142. Part 2 is plausible because with the opening of the HCV-1 02, an additional flowpath is provided around the charcoal adsorbers and the normal flowpath remains in service.

Choice C: This is the Correct answer, see explanation.

Choice D: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because with the opening of the HCV-1 02, an additional flowpath is provided around the charcoal adsorbers and the normal flowpath remains in service.

SRO Basis: Conditions and limitations in the facility license. [10 CFR 55.43(b)(1). Requires the SRO applicant to have basis knowledge to determine whether ODCM compensatory actions are required based on plant status of the AOG system.

Unit 2 2APP-UA-48 5-4 Page 1 of 2 AOG SYSTEM BYPASS AUTO ACTIONS

1. AOG SYSTEM BYPASS VALVE, AOG-HCV-f 02, opens CAUSES
1. High hydrogen Train A
2. High hydrogen Train B
3. High-high cooler condenser condensate level
4. High-high off-gas flow
5. Circuit failure Unit 2 2APP-UA-48 1-4 Page 1 of 2 COOLER CNDSR DRN LEVEL HI AOG SYS BYP AUTO ACTIONS
1. GUARD BED iSOLATION VALVE, A0c3 XC V-142, closes
2. AOG SYSTEM BYPASS VALVE, AOG-HCV-102, opens

Unit 2 APP UA-44 2-2 Page 1 of 2 DISC11AKGE 112 CONC 11:011 AUTO ACTIONS

1. isolation to AGO System. Icloses XCV-1$a, 14?, 142, 143, and 141 after a 30 second time delay)
2. Cpen AOC-WDV-102.

GASEOUS RADWASTE TREATMENT SYSTEM 87.3.10 B 7.3.10 GASEOUS RADWASTE TREATMENT SYSTEM BASES This requirement provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as reasonably achievable. This specification implements the requirements of 10 CFR Part 5D.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section ll.D of Appendix Ito 10 CFR Part

50. The GASEOUS RADWASTE TREATMENT SYSTEM refers to the 3D-minute offgas holdup line, stack filter house flifration, and the Augmented Off-Gas-Treatment System.

GASEOUS RAE WASTE TREATMENT SYSTEM 7.3.10 7.3.10 GASEOUS RAG WASTE TREATMENT SYSTEM OSCMS 7.3W The GASEOUS RASWASTE TREATMENT SYSTEM sEal tie in OFue9 APPUCAEI LIT?: Whenever me Main Cor.denaer Air Ejector eacuatSn System Is n operation..

COMPENSATORY MEASURES CONOON REQUiRED COMPENSATORY COMPLErION MEASURE TIME A. GASEOUS RAE WASTE &I Place GASEOUS 7 Ca)t TREATMENT SYSTEM not RAC4VASTE in operaton. TREATM ENT SYSTEM tn operation

5. 8.5 SuomitaSpeniai Reportio &]days NRC that ieenfffles the NOTE required inonerabie

! Measure Required Compensatory 8.1 shal tie equipment and inc reasons ror line

competed irIRs Condition ta Iroperabt.ty. corrective
entered, actors been to restore me required inoperabie Required Compensatory measure and associated equipment to OPERASLE Compieson TUne not met status, ard a summary descripton ot tine someoNe actions taten to prevent recurrence.

WATER CHEMISTRY GUIDELINES DA.l-8f I Rev. 78 I Page 63 o 87 ATTACHMENT 23 Page I of I Condenser Air Inleakage CONDENSER AIR INLEAKAGE: MODE I SAMPLE DIAGNOSTIC REMARKS REQUIRED ACTIONS IF LMt IS DIAGNOSTIC PARAMErER LIMIT EXCEEDED FREQUENCY Unit pplies abose 50% power At ifO odin the AOG System Bypass Deue4op sod implement so in-leakage test Air In-leakage. sdm eatue tAOG4-tCV-1GI siN open p4so Compensatory measures are required byODCM7. 10.

84. S295001 I Unit One is operating at 72% power with the following conditions:

Jet Pump Flow Loop A (B21-R61 JA) 25 Mlbs/hr Jet Pump Flow Loop B (B21-R61IB) 29 Mlbs/hr Total Core Flow (UICPWTCF) 54 Mlbs/hr Which one of the following completes both statements below lAW Tech Spec 3.4.1, Recirculation Loops Operating, and Bases? (consider each statement separately)

The current Jet Pump Flow mismatch (1)

If Jet Pump Flows are not matched within limits, then the loop with the (2) must be considered not in operation.

A. (1) is within limits (2) lower flow B. (1) is within limits (2) higher flow C. (1) is not within limits (2) lower flow D. (1) is not within limits (2) higher flow Answer: A K/A:

295001 Partial or Complete Loss of Forced Core Flow Circulation AA2 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : (CFR: 41.10 / 43.5 / 45.13) 05 Jet pump operability RO/SRO Rating: 3.1/3.4 Tier 1 / Group 1 K/A match: This question has the SRO candidate determine jet pump operability based on given core flow conditions.

Pedigree: Bank NRC 10-1 Objective: CLSLP002*34 Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR determine the required action(s) to be taken in accordance with Technical Specifications associated with the Reactor Recirculation System. (SRO/STA only)

Reference:

None Cog Level: High

Explanation: Two recirculation loops ate normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. Jet pump loop flow mismatch should be maintained within the following limits:

-jet pump loop flows within 10% (maximum indicated difference 6.0 x106 lbs/hr) with total core flow less than 57.5 x106 lbs/hr

-jet pump loop flows within 5% (maximum indicated difference 3.0 x106 lbs/hr) with total core flow greater than or equal to 57.5 x106 lbs/hr Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because flow mismatch is within limits for lower reactor power level and because the belief that the higher flow loop will experience excessive vibration could cause them to select the

°hig her flow response.

Choice C: Plausible because flow mismatch is not within limits for a higher reactor power level. Part 2 is correct.

Choice D: Plausible because flow mismatch is not within limits for a higher reactor power level. Part 2 is plausible because the belief that the higher flow loop will experience excessive vibration could cause them to select the higher flow response.

SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. The candidate is required to determine whether jet pump flow is within the limits and then use TS bases information to determine which loop is considered not in operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 NOTE Not required to be perfomied until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after both recirculation loops are in operation.

Verifj recirculation loop jet pump flow mismatch th 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> both recirculation loops in operation:

a 10% of rated core flow when operating at

<75% of rated core flow; and

b. 5% of rated core flow when operating at 75% of rated core flow.

Re circulation Loops Operating B 3.41 BASES (continued)

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., 75% of rated core flow), the MCPR requirements provide laiger margins to the fuel cladding integrity Safety Limit such that the potential adverse effe of early boiling transition during a LOCA Is reduced. A larger flow mismatch can, therefore, be allowed when core flow is <75% of rated core flow. The redrculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps assodated with a single rodrculation loop.

The mismatch is measured in terms of the percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered not in operation. The SR is not required when both loops are not in operation since the mismatch limits are meeningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after both loops are in operation. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating expenenceto be adequate to deted off normal jet pump loop flows in a timely manner.

REFERENCES 1. UFSAR, Section 5.4.1.3.

2. UFSAR. Chapter 15.
3. 10 CFR 50.36fc)(2)i).

LOW CORE FLOW 1AOP-04.0 L Rev. 040 I Page8of25 4.2 Supplementary Actions

1. Jf the yellow Speed Hold light is lit, THEN confirm the applicable VFD is maintaining stable recirculation pcimp speed 0 NOTE Jet pump loop flows should be maintained within the following limits: 0 Jet pump loop flows within 10% (maximum indicated difference 6.0 x i0 lbslhr) with total core flow less than 57.5 x 106 lbslhr Jet pump loop flows within 5% (maximum indicated difference 3.0 x 106 lbs/br) with total core flow greater than or equal to 57.5 x 1 0 lbs/hr
85. S295022 I Unit One was at full power when all offsite power was lost.

The following is the Emergency Diesel Generator status:

DGI Locked out on fault DG2 Running and loaded DG3 Running and loaded DG4 Locked out on fault Which one of the following completes the statements below?

The (1) CRD pump must be started to re-establish the CRD system.

OAOP-36.1, Loss Of Any 4760 V Buses or 480V E-Buses, (2) contain the step for placing the CRD Flow Control, CII -FC-R600, in manual with manual potentiometer at minimum setting following the loss of the CRD pump?

A. (I) JA (2) does B. (1) 1A (2) does NOT C. (1) lB (2) does D. (1) lB (2) does NOT Answer: D K/A:

295022 LOSS OF CRD PUMPS AA2 Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS:

(CFR: 41.10 /43.5 /45.13) 02 CRD system status RO/SRO Rating: 3.3/3.4 K/A match: The first part can be answered by knowing the power supply to the pump (systems knowledge.

The second part of the question (SRO Knowledge) is not systems knowledge, is not an immediate operator action, is not an entry condition for AOP/EOP, and is not purpose or mitigative strategy of the procedure. It assesses plant abnormal conditions and then selects a procedure to recover or with which to proceed.

Pedigree: Bank, last used on the 2014 NRC Exam Objective: CLS-LP-302G, Obj. 4c.

Given plant conditions and any of the following AOPs, determine the requited supplementary actions: AOP-36. 1.

Reference:

None Cog Level: High

Explanation: With a loss of all offsite power the E-Buses will strip the loads (CRD Pumps), there are no auto starts for these pumps, so both CRD pumps will be off. DG1 is lost which means El is lost and A CRD pump will not be able to be started. The 15 does however have power, and can be restarted. The steps for restart are located in the OP, and AOP2.0, however they are not located in A0P36.l. DG-2 and DG3 provide power to the I B and 2A CRD Pumps. The DG4 loss is a loss of the 2B CRD pump.

Distractor Analysis:

Choice A: Part 1 is plausible because the JA CRD pump is tripped but it cannot be re-started due to El remaining de-energized (DG1 Locked ouot on fault). In addition, Unit 2 has power for the 2A CRD pump. Part 2 is plausible because an AOP (AOP-2.0) does contain the steps for re-energizing the pump, and A0P36.l is an AOP we would be in due to the given electrical power loss, however AOP 36.1 does nto contain the steps for restarting the pump.

Choice B: Part 1 is plausible because the JA CRD pump is tripped but it cannot be re-started due to El remaining de-energized (DGI Locked ouot on fault). In addition, Unit 2 has power for the 2A CRD pump. Part 2 is correct, see explanation.

Choice C: Part 1 is correct, see explanation. Part 2 is plausible because an AOP (AOP-2.0) does contain the steps for re-energizing the pump, and A0P36.l is an AOP we would be in due to the given electrical power loss, however AOP 36.1 does nto contain the steps for restarting the pump.

Choice D: The answer is correct, see explanation.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)J LOSS OF ANY 4160! BUSES OR 480V E-BUSES OAOP-36.1 Rev. 64 Page 95 of 101 ATTACHMENT 4 Page 1 of 6 4160V and 4$OV Emergency Bus Loads Load Load Bus El (kw) Bus E2 (kw)

RHR Pump 1C 750 RHR Pump 1D 750 RHRSW Pump 1C 600 RHRSW Pump 1D 600 CS Pump JA 940 CS Pump lB 940 CRD Pump IA 190 CRD Pump lB 190 NSW Pump JA 225 NSW Pump lB 225 CSWPump 18 225 CSWPump1C 225 RHR Pump 2C 750 RHR Pump 2D 750 RHRSW Pump 2C 600 RHRSW Pump 2D 600 CSW Pump 2C 225 Fire Pump (normal) 190

CONTROL ROD DRIVE HYDRAULIC SYSTEM 1 OP-08 OPERATING PROCEDURE Rev. 91 Page 104 of 384 63.2Q Restarting CRD Hydraulic System Foflowing Loss Of CRD Pump 1 Ensure the following Initial Conditions are met:

  • CRD System was in operation per Section 6.1.1
  • The operating CRD pump has STOPPED
2. Close B32-V22 (Seal Injection Vlv) tot Recirc Pump A
3. Close B32-V30 (Seal Injection Vlv) tot Recitc Pump B
4. Place Ci 1-FC-R600 (CRD Flow Control) in MAN
5. Reduce CI i-FC-R600 (CRD Flow Control) potentiometer to minimum setting
6. Ensure Cl 1-PCV-F003 (Drive Press Vlv) is OPEN
7. Ensure RBCCW is in operation to supply cooling v.iater to CRD pumps
8. Start non-operating (desired) CRD Pump A or B LOSS OF ANY 4160V BUSES OR 480V E-BUSES OAOP-36.i Rev. 69 Page 12 of 108
4. Start the CRD System in accordance 1 OP-08(20P-08) Control Rod Drive 1-tydraulic System Operating Procedixe, using the section for restarting the system following loss of the operating CR0 pump D
86. S295015 I Unit One is performing the ATWS Procedute with the following conditions:

A-05 (2-6) Reactor Vess Lo Level Trip, is illuminated A-06 (1-6) Reactor Vess Lo Lo Water Level Sys A, is NOT illuminated A-06 (2-6) Reactor Vess Lo Lo Water Level Sys B, is NOT illuminated MSIVs are closed Reactor pressure peaked at 1141 psig and is now being controlled 800-1000 psig.

Torus water temperature is 105°F and rising Reactor power is 25%

lAW 001-37.5, ATW$ Procedure Basis Document, which one of the following identifies the action that will have the highest priority?

A. SLC initiation.

B. Inhibiting ADS.

C. Trip both Reactor Recirc Pumps.

D. Termination and prevention of RPV injection.

Answer: D K/A:

295015 Incomplete SCRAM G2.4.31 Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 /45.3)

RO/SRO Rating: 4.2/4.1 Tier 1 I Group 2 K/A match: THIS QUESTION WAS PRE-SUBMIHED FOR APPROVAL.

This question requires knowledge of the annunciator response procedure status (where level is at) to determine the appropriate course of action while executing the EOP. Also for prioritization of the action to direct.

Pedigree: New Objective: 300E-17e Given plant conditions and the Anticipated Transient Without Scram Procedure, determine the following: Priority of execution given to each leg of the procedure.

Reference:

None Cog Level: Hi Explanation: With Reactor vessel lo lo water level not illuminated, RPV water level is >90. Reactor power is also >23% with the RR pumps tripped, lAW RC/Q-8 and RCIL-2, terminate and prevent is immediately required in order to prevent THI.

Distractor Analysis:

Choice A: Plausible because SLC initiation is required with >2% power and rising torus temperatures in order to not exceed the HCTL. However, it is not the first step required because power is >23%.

Choice B: Plausible because Inhibiting ADS is required, However, it is not the first step required because power is >23%.

Choice C: Plausible because tripping the recirc pumps would be done prior to terminate and prevent but with RPV pressure peaking at> 1138 ARI would have tripped the pumps, Level is not at LL2 yet according to the alarms.

Choice D: Correct Answer, see explanation.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J. This SRO applicant is required to direct the appropriate actions out of the ATWS EOP based on plant conditions. The action is contained in a note in the EOP and not a general strategy of the EOP.

EVCE! ETOTT!

I35.5 Auiiiary Rty 171K21,. tatecJ orn 2LLTIIO24i1) u2iiry .T1K1C ELLTltC251)

Lere1 Tan.ttr Tri U,+/-t and QS3IEt. L1T EFFECTS

1. incrab u.p1nnt ma-: a Tzh pe LC.

i_ LL:4 Lc o CC Pcwer 1 APP-A-06 Rev. 69 Page 11 ot 85

ATWS PftOCEDURE BASIS D-OCUMEPJr 001-37.5 Rev. 14 Page 13 d65 5.4 Step RCL-2 If reactor power is greater than 23% with both reactor reel rculaton pun-cs tripped and RPV level above 90 inches, RPV level needs to be prompdy reduced below the feedwater notee. to avoid thermal hydraulo instabilities. Ths is accomplished by termiraticn and prevention & injection systems. from identified systems, paticuIaiyfeedwatw within 120 seconds.

To prevent or mitigate the consequences of any lage irregular neuton flux oscillations induced by neuoonicthenial-hydrauld instabilities, RPV level is initially lowered sufficiently below the elevat;on of the kedwaterspsger nozzles. This place the feedwater spagers in the steam space providing effective heating of the retatively cold feedwater and eltininating the potential for high core inlet subcoding. For ocndticns that are susceptible to oscillations, initiation and groMh of oscillations is principally dependent upon subcocling at the core inlet the greater the subcooling. the more likely oscillations will commence and increase in magnitude.

If reactor power is at or below the APRM downsc.ale mp setpisnt (2%), ft is highly unirkely that the core bulk boiling bounday would be below that which provides sritable stability margin for operation at high powers and low flows. (A minimum b&rng boundary cr14 ft above the bottom of active fuel has been shown to be effective as a stabllrty control because a relatively long two-phase c&umn is required to develop a coupled neutonic! thermal-hydrautc instability.) Furthermore, flow!density variations would be limited with reactor power this low since the core has a relatively low average void content.

A1WS PROCEDURE BACtS DOCUMENT 001-37.5 Re.. 14 Page 60 &65 5.33 Step RCIQ-6 through RCIQ-6 Ln7[J [LJ. 3 If reactor power is below 2%, the operator is directed to inject boron before torus water temperature reaches I lOP. This allows sufftciert time for Hot Shutdown Boron Weight (HSBW) of boron to be injected As long as the core remans submerged (the preferred method of core cooling), fuel integrity and RPV integrttyae not directly challenged even wider failure-to-scram conditions. A scram failureccupled with an M3PJ isolation however, result in rapid heatip of the torus due to the steam dischsged from the RPV era SRVs. The challenae to containment thus becomes the tinwing factor which defines the reqi irernwit for boron injection.

if tows temperatire and RPJ pressure cannot be maintained betters the Heat Capacity Temperature Limit 4HCU), rapid depressurizaton of the RPV will be required. To avoid depressudring the RPV with the reactcr at power, it is desirable to shut down the reactor prior to reaching HCTL, thL,s minirrizing the quantity of heat rejected to the tows. The Boron Injection Initiation Temperature (BItT) is defined so as to achieve ths when practicable.

ATWO PROCEDURE BASIS DOCUMENT 00l-3T.5 Rev. 14 Page 22 of 65 5.10 Step RCL4 through RCL-l0 1

Ut 1 1. aeu. WheN Tsl4 Q.5 1a,,,i,iIt eoa P,cn,I D*Iw*lIflSt)fl I Yn baML,U,* :-ij,..n-., I Mrc(..,.,, 1 1

ml Based on reactor power being above 2%. Step RCt-2 initially lowered RPV level, to the feedwater sparger. by terminating and preventing injection from identified systems if all of the condibons in Table 0-2 are met, Step RCIL-9 will lower RPV level further to suppress reactor power-When any concttion in Table 0-2 is no longer met the operator is directed to continue to subsequent steps which will establish a new RPV level band.

AV.VG PROCEDURE BAtiO DOCUMENT Ol-37.5 Rev. 14 Page23 o5 5.10 Step RCL-8 through RC&-l0 (continued)

Terminating and preventing injection from:

  • Condensate and Feedwater is addressed in IOP-32 (20P.321, Condensate And Feedwater System Operating Procedure, and covers terminating and preventing ejection by either hipping budt Reactor Feed Pumps (RFPs) or by idling one RFP.
  • Core Spray is accomplished bytnpping the associated Ccre Spray ioops operating pump.
  • HPCI is addressed ti, iop-ig (QP19j, High Pressure Ccolar,t Injection System Operating Procedure, and covers tentinating and preventing injection when HPCI is either operating or not operating.
  • RHR is accomplished by thpping the associated RHR loops operating pump(s).

The Boron Injection Initiation Temperature (BItT) a function reactor power and is the torus temperature before which boron injection must be initiated if a reactor depressurizabon, due to exceeding the Heat Capacity Temperature Limit (HCTL). is to be precluded. This temperature is 11 OF.

The combination of high reactor power (above the APRM dcwiscale hip),

high torus temperature (above BItT), and an open SRV or high drpisll pressure (above the scram setpoint). are symptomatic of heat being rejected to the torus at a rate in excess of that which can be removed by the torus coding system. Unless mitigated, these conditions ultimately result in loss of NPSH for ECCS pimps taking suction on the torus, containment overpressztzation, and (ultimately) loss of Pnmaty Containment integrity, which in turn could lead to a loss dl adequate core cooling and uncontrofted release of radioactivity to the environment.

The conditions listed in Table 0-2. combried with the inabiitty to shut down the reactor through control rod insertion, dictate a requirement to pcorrØy further reduce reactor pcwer in order to preserve Primaty Containment integrity since, as long as these conthiorw exist. torus heatup wIt contfnue.

Since RPV level is only atowed to drop to TAF before injection is restarted.

if RPV level is already below TAF. then the jective of the step has been accomplished. Further lowering of RPV level is not necessaty, and the steps which deliberately lower RPV level e bypassed.

87. S295023 I Which one of the following completes both statements below?

lAW Tech Spec 3.9.6, Reactor Pressure Vessel (RPV) Water Level, the minimum water level over the top of irradiated fuel assemblies seated within the RPV during movement of irradiated fuel assemblies in the RPV is (1)

The Tech Spec bases for the minimum water level is to provide for (2) during a fuel handling accident.

A. (I) I9feetIl inches (2) iodine retention B. (1) I9feetII inches (2) shielding of radioactive decay particles C. (1) 23 feet (2) iodine retention D. (I) 23 feet (2) shielding of radioactive decay particles Answer: C K/A:

295023 Refueling Accidents G2.2.25Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 /41.7 / 43.2)

RO/SRO Rating: 3.2/4.2 Tier 1 / Group 1 K/A Match: THIS QUESTION WAS PRE-SUBM1HED FOR APPROVAL.

This meets the K/A because it is testing the TS bases for the RPV water level during movement of irradiated fuel assemblies in the RPV.

Pedigree: New Objective: LOI-CLS-LP-200-B Obj 12. Identify conditions and limitations in the facility license. (SRO/STA only)

Reference:

None Cog Level: Fundamental Explanation: Part 1: LCO 3.9.6 states, RPV water level shall be = 23 ft above the top of irradiated fuel assemblies seated within the RPV. Part 2: The minimum water level of 23 ft allows a decontamination factor of 200 to be used in the accident analysis for iodine. This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the damaged fuel assembly rods is retained by the water.

Distractor Analysis:

Choice A: Part 1 is plausible because lAW TS 3.7.7 the spent fuel storage pool water level shall be = 19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks. Part 2 is plausible because it is correct, see explanation.

Choice B: Part I is plausible because lAW IS 3.7.7 the spent fuel storage pool water level shall be = 19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks. Part 2 is plausible because the water does provide shielding from radioactive decay particles, but that is not the TS bases for the minimum water level.

Choice C: Correct Answer, see explanation Choice D: Part 2 is plausible because it is correct, see explanation. Part 2 is plausible because the water does provide shielding from radioactive decay particles, but that is not the TS bases for the minimum water level.

SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)J. Requires the SRO applicant to know the TS bases for RPV water level.

RPV Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Reactor Pressure Vessel (RPV) Water Level LCO 3.9.6 RPV water level shall be 23 ft above the top of irradiated fuel assemblies seated within the RPV APPLICABILITY: During movement of irradiated fuel assemblies vithin the RPV.

During movement of new fuel assemnbhes or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPV.

Spent Fuel Storage Pool Water Level 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Spent Fuel Storage Pool Water Level LCO 3.7.7 The spent fuel storage pool water level shall be 19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool.

B 3.9 REFUELING OPERATIONS B 3.9.6 Reactor Pressure Vessel (RPV) Water Level BASES BACKGROUND The movement of fuel assemblies or handling of control rods within the RPV requires a minimum water level of 23 ft above the top of irradiated fuel assemblies seated within the RPV. During refueling, this maintains a sufficient water level in the reactor vessel. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to well below the 10 CFR 50.67 exposure guidelines (Ref. 3).

APPLICABLE During movement of fuel assemblies or handling of control rods, the SAFETY ANALYSES water level in the RPV is an initial condition design parameter in the analysis of a fuel handhng accident in containment postulated by Regulatory Guide 1.183 (Ref. 1). A minimum water level of 23 ft allows a decontamination factor of 200 to be used in the accident analysis for iodine (Ref. 1). This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the damaged fuel assembly rods is retained by the water.

Analysis of the fuel handling accident loside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and that offsite doses are maintained well below the allowable limits of Reference 3.

RPV water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(H) (Ref. 4).

(continued)

88. 5295037 1 Unit Two is in an ATWS executing RXFP, with the following plant conditions:

Injection to the RPV has been terminated and prevented The Minimum Number of SRVs Required for Emergency Depressurization are open.

Table P3 Minimum Steam Cooling Pressure Open SRVs Pressure (psig) 7or more 120 6 145 5 175 4 220 3 300 2 455 I 516 lAW RXFP, which one of the following completes the statement below?

The CRS should direct injection to the RPV when EITHER:

(1) SRV remains open OR when reactor pressure lowers below the Minimum Steam Cooling Pressure of (2)

A. (1) NO (2) 175 psig B. (1) NO (2) 455 psig C. (1) ONLY one (2) 175 psig D. (1) ONLY one (2) 455 psig Answer: A K/A:

295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown G2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7/43.5 /45.12)

RO/SRO Rating: 4.0/4.6 Tier 1 / Group 1

K/A Match: This meets the K/A because it is testing the required steam cooling pressure for adequate core cooling.

Pedigree: New Objective: LOl-CLS-LP-300-F Obj.3 Given the Reactor Flooding Procedure, which steps have been completed and plant parameter values, determine the required operator actions.

Reference:

None Cog Level: High Explanation: Part 1: lAW RXFP-9/10, injection is reestablished when either no SRVs are open or Part 2:

Reactor is below the MSCP, which in this case for 5 SRVs (MNSRED) is 175 psig.

Distractor Analysis:

Choice A: Correct answer, see explanation.

Choice B: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because this is the minimum steam cooling pressure for having the Minimum Number of SRVs Required for Decay Heat Removal open.

Choice C: Part 1 is plausible because once injection is re-established, RXFP-10 directs injection to continue until at least one SRV is open. In addition, with no SRVs open or below the MSCP adequate core cooling could possibly not exist, so a candidate might assume we would never wait for that condition to reestablish injection. Part 2 is plausible, because it is correct, see explanation.

Choice D: Part 1 is plausible because once injection is re-established, RXFP-10 directs injection to continue until at least one SRV is open. In addition, with no SRVs open or below the MSCP adequate core cooling could possibly not exist, so a candidate might assume we would never wait for that condition to reestablish injection. Part 2 is plausible because this is the minimum steam cooling pressure for having the Minimum Number of SRVs Required for Decay Heat Removal open.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]. This requires the SRO to have knowledge of more than the overall sequence of events in the EOPs or mitigating strategy.

Since it requires the SRO to know the MNSRED and how to implement Table-P3.

USERS GUIDE QEOP-Ol-UG Rev 067 Page ii of 156 3.0 DEFINITIONS (continued)

37. Maximtim Subcritical Banked Withdrawal Position: The lowest control rod position to which all controls rods may be withdrawn in bank and the reactor wit nonetheless remain shutdown under all conditions. This position is utilized to assure the reactor will remain shutdown irrespective of reactor water temperature.
38. Minimum Core Steam Flow: The lowest core steam flow which is sufficient to preclude any clad temperature from exceeding 15Q0F even if the reactor core is not completely cove red
39. Minimum Debris Retention Injection Rate: The lowest RPV injection rate at which it is expected that core debris wilt be retained in the RPV when RPV level cannot be determined to be above the bottom of active fuel. (Attachment f 7)
40. Minimum Indicated Level: The highest RPV level instrument indication which results from off-calibration instrument run temperature conditions when RPV level is actually at the elevation of the instrument variable leg tap.
41. Minimum Ncimber of SRVS Required for Decay Heat Removal:

The least number of SRVs (2) which will remove all decay heat from the core at a pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flw.

4Z Minimum Number of SRVS Required for Emergency Depressurizatlon: The number of SRVs (5) which correspond to a minimum steam cooling pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow.

,/ WHEN (Either NO SRVs are OPEN

  • RPV pressure is below MSCP (Table P-3)

\ THEN continue

/RXFP-9 Commence and slowly raise RPV injection using Table 1-3 systems until: 5 At least one SRV is OPEN AND RPV pressure above MSCP (Table P-3) but as tow as practicable

89. S295026 I An event on Unit One has resulted in the following plant conditions:

Reactor pressure: 1000 psig Reactor Water Level 120 inches Control Rod position Unknown APRMs Downscale Drywell pressure: 3 psig Torus pressure: 2 psig Torus water temp: 152°F Torus water level: -36 inches (REFERENCE PROVIDED)

Which one of the following identifies the required actions for reactor pressure control?

A. Exit the RC/P flowpath of ATWS, and go to OEOP-01-EDP, Emergency Depressurization.

B. Exit the RC/P flowpath of RVCP, and go to OEOP-01-EDP, Emergency Depressurization.

C. Remain in the RC/P flowpath of ATWS, and exceed 100°F/hr cooldown rate if necessary.

D. Remain in the RC/P flowpath of RVCP, and exceed 100°F/hr cooldown rate if necessary.

Answer: C K/A:

295026 Suppression Pool High Water Temperature G2.1 .23Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 /43.5 /45.2 /45.6)

RO/SRO Rating: 4.3/4.4 Tier 1 I Group 1 K/A match: The applicant is required to determine which procedure to implement based on HCTL Pedigree: New Objective: CLS-LP-300L, Obj. 5a Given the Primary Containment Control Procedure, determine the appropriate operator actions if any of the following limits are approached or exceeded: Heat Capacity Temperature Limit

Reference:

OEOP-01 -UG, Attachment 7, Heat Capacity Temperature Limit Cog Level: Hi Explanation: With rods at an unknown position, an ATWS has occurred. Since HCTL is close to the unsafe region (but not violating it), exceeding the cooldown rate in the RC/P flowpath of ATWS is warranted.

Distractor Analysis:

Choice A: Plausible because a novice applicant may misinterpret the graph and believe HCTL is in the unsafe region, and an ED is warranted.

Choice B: Plausible because a novice applicant may believe with APRMs downscale an ATWS has not occurred, and may misinterpret the graph believing an ED is warranted.

Choice C: Correct answer, see explanation.

Choice D: Plausible because a novice applicant may believe with APRMs downscale an ATWS has not occurred. Exceeding the cooldown rate is correct because of the operating point on the HCTL graph.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5) SRO is required to select the appropriate EOP action based on HCTL.

UER GUIDE CEOP-QI-UG Rev. 85 Page 77 of 152 ATTACHMENT 7 Page 1 of I Heat Capacity Temperature limit T

21O -!H+/- UNSAFE ABOVE -

UU]

i+/- +/- SELECTED LINE -,

4n-FZJ I

-t-r. i I-,

170 E :r E -0.25 FT

-1.2S FT 2,30 FT

.25 FT SAFEBELOW - 4.25 FT U, iao -

SELECTED LINE 12D-0 1 H. 5.50 FT I- 110 lin Itt-.1150 100 500 700 sco I 1,100 0 200 400 500 tiOO 1,000 RPV PRESSURE (P510)

(START)

AThS-?

r While in this procedure:

IF ThEN RPV Iee HfI be deem1ned Exit RCI RCdP flcad 3rd go to ECP-O1-XFP Emegeiicy da1oii leQ has De1 tequred Proceed to

  • Exit RCJP nTMPrI and go to ECP-0i-EC Reztc te DiD WUioUt Bomn Laer cridfliorA fte 1. TennKiate borm I cliorrijOl required sy oUn EOPs
2. Exit tile flowt ar go to ECcV1RVCP A1.2

. a-

C 1p-

F THEN Pe plesetre seçiades 440 pzig Ccntrc4 Irecc41 1r:

1 Ccnrse a ce pcay

. LPc KCIL C.SM4OT t.e srt3e0 1r Ue rebr. biIt Maintsai RPV prseswe tose 1CJL n IflJX Ecled 1G2Ffl coen re IT recseeq MIVs are CLOD Equanse pceesew. ard open MCP-]:

DeateRPJse.se QicuD 4 IscIcnDer EoP.oi-Ep-ia Sceon 5em reqaired t

Mr cend&ereiS asaIIe asa heatell AND tiQ Irco1 of a main &te3n INC treas erets A orUnuor pneumatic &lquly INQI avatLe to RJ Eec1Qirg ceeseure.

. Ii place ccn ewt, JJtCCLCE EdepreesrIrg RP, JIjOj nlrlrrlze cycs

90. S295035 1 0.5 Unit Two is operating at rated power.

PCCP has been entered due to high torus water temperature with the following plant conditions:

c=-04 UA-12 (3-3) Rx Bldg Duff Press High/Low, is in alarm.

UA-05 (6-10) Rx Bldg Isolated, is in alarm.

Gz

-0 2 Reactor Building Pressure (indication on the left)

H Which one of the following completes both statements below?

H Reactor Building pressure is (1)

O The CRS will direct Reactor Building HVAC restarted lAW (2)

A. (1) positive (2) 20P-37.1, Reactor Building Heating and Ventilation System Operating Procedure B. (1) positive (2) OEOP-01 -SEP-04, Reactor Building HVAC Restart Procedure C. (1) negative (2) 20P-37. 1, Reactor Building Heating and Ventilation System Operating Procedure D. (1) negative (2) OEOP-01 -SEP-04, Reactor Building HVAC Restart Procedure Answer: C K/A:

295035 Secondary Containment High Differential Pressure EA2 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: fCFR: 41.8 to 41.10) 01 Secondary containment pressure RO/SRO Rating: 3.8/3.9 Tier 1 / Group 2 K/A match: The applicant is required to interpret secondary containment pressure and select the appropriate procedure based on this interpretation.

Pedigree: New Objective: LOl-CLS-LP-300-M Obj 11 Given plant conditions involving Reactor Building HVAC system isolation and the Secondary Containment Control Procedure, determine if the Reactor Building HVAC system should be restarted.

Reference:

None Cog Level: Hi

Explanation: Part 1: With indications at upscale> +0.5 inches of h20, reactor building pressure is negative.

Part 2: lAW with UA-5 (6-JO) and UA-12 (3-3) RBHVAC is restarted 20P37.1.

Distractor Analysis:

Choice A: Part 1 is plausible because the reading is +05 inches of h2O and upscale high, a novice candidate could mistake this for a positive pressure indication. Part 2: is plausible because it is correct, see explanation.

Choice B: Part 1 is plausible because the reading is +0.5 inches of h20 and upscale high, a novice candidate could mistake this for a positive pressure indication. Part 2: is plausible because RBHVAC is restarted using SEP-04, when in SCCP when LL2 and high drywell pressure needs to be defeated. In addition the title is rbhvac restart procedure, and it is a SEP.

Choice C: Correct Answer, see explanation.

Choice D: Part 1 is plausible because it is correct, see explanation. Part 2: is plausible because RBHVAC is restarted using SEP-04, when in SCCP when LL2 and high drywell pressure needs to be defeated. In addition the title is rbhvac restart procedure, and it is a SEP.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [JO CFR 55.43(b)(5)J The SRO applicant is required to interpret secondary containment pressure and select the appropriate procedure based on this interpretation.

AP UA12 33 Page 1 f 1 PX BLDG DIFF PRESS H1GH,/LCW (Reactor Building Differential Pressure High/l:w)

AVID ACTIONS

1. Reactor Building supply and ezhaus. fans trip
1. High or low differential pressure between the Ret-cr Building and atmosctheric ores sure.
2. Circuit inalfuncti:n.

OBSERVATIONS

1. Reactor Building Static Pressure Indicator, 2VAPI1257, on P.1GB Panel XU3 -

ACTIONS

1. If secondary containment integrity is required and differential pressure is lc-w, enter OEOP33SCCP, Secondary Containment Control, and e.ecute concurrently with thts procedure.
2. Inform E&P.C Chemistry Reactor Building Ventilation is not in service.
3. Verify that the valve uncut is correct per 037.1, P.eart:r Building Heating and Ventilation System.
4. Start up the system per Section 3.1 cf 0P37.l.

S. If a circuit malfunction is suspected, ensure that a WP.IWD is submitted.

Lint 2 it?? tit-C5 e:;

?ars 2 cf 2 itCTICNE

& if area radiation iees or. Tabe I of GED?025CD? enreed naniru norna operating tal:es, enter G!C?QIfCC?, Secondarr Containr.ent oztro_ -

If the .eartor Stilding has izolated and it is desired to restart ventiaticn, enter 20?272 Peartor Erildin; Eeating and Ventnlatron lysten.

B. flot:v ESRC Counting Porn that reactor building trentilaticn hs been scoured.

I -

While in this procedure:

IF THEN

  • ESSLNS RB HVJC soi1ed
  • RB verfllalicr eflatLNt rxl3Bcn eroee 4 nThr
  • Ejisies SBGT oçenung
  • RB verUblicc terwfle etee 1WF 4UA03, 6-2]

95 FWAC sc4ates.Ij ciondlicre est Restart RB tJAC per

  • RB verUoneflalSrLNlizcnDaw4mMw OP-37.I
  • RB venttstco atiaS rniem nxn ties maloed or soae
  • EO-C1-SEP-C4 rnecessay Ia dezt LL-2 ci MJLi ayaei
  • RB vensLatliet tençenaite liz IIQI exceeded 135W (UA-Cd 6-2]

- --a sttA-z

91. 5295038 1 A release on Unit Two is occurring with the following plant conditions:

Main Stack Rad Monitor, D12-RM-23S, is reading 2.3E+08 pCI/sec Turbine Building Vent Rad Monitor, D12-RM-23, is reading 2.5E+07 pCI/sec Real-time dose assessment using actual meteorology indicates 0.92 Rem TEDE and 5.1 Rem thyroid CDE at the site boundary (REFERENCE PROVIDED)

Which one of the following completes both statements below?

lAW RRCP, Unit One (1) override and reset the main stack hi-hi isolation signal.

The highest EAL classification for this event is (2)

A. (1) can (2) Site Area Emergency B. (1) can (2) General Emergency C. (1) can NOT (2) Site Area Emergency D. (1) can NOT (2) General Emergency Answer: B K/A:

295038 High Off-Site Release Rate EA2 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.10/43.5/45.13) 01 Off-site RO/SRO Rating: 3.3/4.3 Tier 1 / Group 1 K/A match: The candidate is required to compare the given radiation release values (including site boundary), and compare those to the EALs for rad effluent. Based on this comparison the candidate must make the correct EAL designation.

Pedigree: NEW Objective: CLS-LP-301-B Obj 9: Given a hypothetical abnormal event and plant operating mode, use OPEP-02.1 to properly classify or re-classify the event

Reference:

OPEP-02.1 Cog Level: Hi

Explanation: Part 1: lAW RRCP if the release is not from Unit One they are allowed to override and then reset the isolation signal. Part 2: Due to Site Area Boundary dose >5000 mrem thyroid CDE, a GE is the highest classification.

Distractor Analysis:

Choice A: Parti is plausible because it is correct, see explanation. Part 2 is plausible because the main stack and turbine building rad monitors are reading > the SAE setpoint. However, the dose assessments results are above a GE classification.

Choice B: Correct Answer, see explanation.

Choice C: Parti is plausible because it is a common stack for both units. Part 2 is plausible because the main stack and turbine building rad monitors are reading > the SAE setpoint. However, the dose assessments results are above a GE classification.

Choice D: Parti is plausible because it is a common stack for both units. Part 2 is plausible because it is correct, see explanation SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J The SRO candidate is required to compare the given radiation release values (including site boundary), and compare those to the EALs for rad effluent. Based on this comparison the candidate must make the correct EAL designation.

While in this procedure:

IF THEN Either Perforni per OP-37.3

  • TB ventilation is SHUTDOWN
  • PLace TB HVAC in Recirculation Mode

. TB ventilation is in Once Through Mode

  • EnsureQfl TB Air Filter Exhaust Fans operating Fuel failure indicated by: Ensure Control Building Emergency

. . Recirculalton operating (OP-37)

Main Steam Line Rad Hi CUA-23, 2-6)

Process Off-Gas Rad Hi (UA-03, 5-2)

  • Process OG Vent Pipe Rod Hi (UA-03, 6-4 Reactor building breached Request ERO evaluate use of mitigating sprays per EDMG-0D2 RRCP-2 I

lF main stack Hi-Hi isolation Monitor site boundary dose actuated, per AD-EP-ALL-0202. THEN:

RRCP4

  • Override isolation
  • Reset Group 6 isolation RRCP-11 WHEN iI ni primary system is discharging outside Exit procedure.

reactor building RRCP-12 N THEN /

continue fRRCP.5

RADIOACTIVITY RELEASE CONTROL 001-37.10 PROCEDURE BASIS DOCUMENT Rev. Oil Page 7 of 16 5.2 Step RRCP-2 r

I -

White in this procedure:

IF THEII Ethc Perfonn r OP31.3

  • 18 eiititin e SHUTOO\VN
  • Place T6 I1i.AC in 1e eiiM:ion I.1)tli3
  • TB venflblion is in Once Through Mode Eneure tour TB Air Filer Etuist Fans epefalilig Fuel FaLire ndcated by. !nsure Contici Bud Emergency nec;lcui.tci UIJJ.iniiI] uJP-37
  • t1ain 5e1tn ljne Raii HI (UA.23.
  • Process 0tt.Cas Flad Iii IUA.03 6.2)
  • Pwcenn 013 Veri( P FarI Hi (IJA-03, 84)

Reactrrbuildng hr-erhd Request ERO eyeiu ini of nidJIing npmyn pi EDMC3-0U2 nRC Step RRCP-2 is a procedure override which appLies the entire time RRCP is being executed Each of the three components specit applicable conditions and direct performance of actions as discussed below.

5.2.1 Step RRCP-2 First Override Continued personnel access to the turbine building may be essential for responding to emergencies or transients which may degrade into emergencies. The turbine l)uilding is not an air tight structure, and radioactivity release inside the turbine building would not only limit personnel access but would eventually lead to an unmonitored ground level release, or release via the turbine building ventilation if operating in the once-through lineup.

Operation of the turbine building ventilation in the recircutation lineup helps to improve turbine building accessibility. In addion, since both units share a common turbine building airspace, if the building is intact, removing turbine building ventilation from once through lineup will temiinate a large unfiltered volume discharge flow path for a leak on either unit. Due to normal operational requirements when in once through lineup, at least one Air Filter Exhaust Fan and WRGM will be in service providing a monitored and filtered discharge flowpath.

Table R I Effluent Monitor Classification Thresholds i_,i_=

Release Point Monitor GE SAE Alert UE

, Main ac DI2-RM-2 al3E÷opCl/ssc 2.13E+t8 pcsee 2.14t7 pcLsec I.8O+O6 pCLsec Reactor Eicg Wet Noi1e Gas H-124-3 6.14E+04 cm C,

TUUiirgVeritRad D12RM-23 1opCIisec t.7E+G7 see 1a7E.iOepcUsec 1.13E+G3 Ct%ee 5er,lce V E1fltrt Rat Dl2-RM-KC5 2 X r aarm ftadwasIe EtSuent Rat 012RM-K503 2 X hi-N alaIm n_

GENERAL EMERGENCY SITE AREA EMERGENCY RGI Release otgasec4is radIoactl It olTiItr dose gteater thai 1.10 mrernrEDE cr5000 ntrern 01)T0Id CDE RSI Release oi gaseous isthoacOty restttmg In 0115115 oceegrealat 100 mt5m TIDE cr500 rtrem 11frIl CDI I

I I I 2 I 3 I 4 I 5 IDEFI I I I 2 I 3 I 4 I 5 IDEFI RGI .1 RS 1.1 In the absence & real-time dose assessment, rsadir on any In the absence of real-time dose assessment, reading on any Table R-l effluent radiation monitor> column GE for 15 Table R-1 effluent radiation monitor : coii]rnn SAE for 15 mm. (Notes 1.2.3,4) mm. (Notes 1, 2,3,4)

RGI.2 RS 1.2 Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorotogy indicates doses

> 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond > 100 mrem TEOE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) the SITE BOUNDARY (Note 4)

RGI.3 RSI.3 Field survey results indicate EITHER of the following at or Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: beyond the SITE BOUNDARY:

- Closed window dose rates> 1000 mR/hr expected to - Closed window dose rates> 100 mRhr expected to continue for 60 miii. continue for SOmin.

- Analyses of field sur.ey samples uiidicate thyroid CDE - Analyses of field survey samptes indicate thyroid CDE

> 5000 rnrem for 60 mm. of toha!ation. > 500 mrem for 60 mm. of inhalation.

(Notes 1,2) (Notes 1.2)

92. S600000 I Unit One and Unit Two are executing OASSD-OI, Alternative Safe Shutdown Procedure Index, due to a fire in Main Control Room back panels requiring Main Control Room evacuation. Current plant conditions are:

Unit One and Two have scrammed All MSIVs are shut Which one of the following completes both statements below?

The CRS will enter OASSD-02, Control Building, and (I) OASSD-O1.

The CRS will direct actions to achieve a safe shutdown using (2)

A. (1) exit (2) HPCI B. (1) exit (2) RCIC C. (1) concurrently perform (2) HPCI D. (I) concurrently perform (2) RCIC Answer: B K/A:

600000 Plant Fire On Site AA2 Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE:

(CFR: 41.10/43.5/45.13) 07 Whether malfunction is due to common-mode electrical failures RO/SRO Rating: 2.6/3.0 Tier 1 / Group 1 K/A match: THIS QUESTION WAS PRE-SUBMIHED FOR APPROVAL.

The applicant is required to determine that based on a fire in the MCR requiring evacuation, common mode failures of electrical equipment could occur and interpret the procedure steps that require exiting ASSD-01 and entering the standalone ASSD-02. In addition, the applicant will determine which train of ASSD equipment (HPCI/RCIC) will remain unaffected by the potential common mode electrical failure.

Pedigree: Modified 14 NRC Objective: LOI-CLS-LP-301 Obj 20 Given a fire in an ASSD area, describe the potential impact that the fire may have on Safe Shutdown Equipment

Reference:

None Cog Level: Fundamental Explanation: Part 1: OASSD-02, Control Building, is an outside Control Room shutdown procedure for both units. This procedure is a stand-alone post fire shutdown procedure for a Control Room evacuation and requires the reactors to be in hot shutdown/manually scrammed prior to

leaving the Control Room. There are parts of the control building that this procedure is not used for, i.e. battery rooms. Part 2: The safe shutdown strategy for control room evacuation requires the B train of ASSD equipment (RCIC).

Distractor Analysis:

Choice A: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because the A train of Safe shutdown equipment (HPCI) is used in other ASSDs.

Choice B: Correct Answer, see explanation Choice C: Part I is plausible because for every other ASSD fire OASSD-01 is performed concurrently with specific ASSD sub procedures. Part 2 is plausible because the A train (HPCI) of Safe shutdown equipment is used in other ASSDs.

Choice D: Part 1 is plausible because for every other ASSD fire OASSD-01 is performed concurrently with specific ASSD sub procedures. Part 2 is plausible because it is correct, see explanation.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]. requires the SRO applicant to decide the appropriate transition to an event specific subprocedure based on fire in the control building.

2014 Question:

A fire in the control building fire area requires entry into OASSD-01, Alternative Safe Shutdown Procedure Index. The CRS has determined that alternate safe shutdown actions are required. Both Unit One and Unit Two have been manually scrammed.

Which one of the following completes the statements below lAW OASSD-0i?

The next action that is required is to (1) -

Following this action both units will (2) -

A. (1) place MSIV control svitches in close (2) perform OASSD-0I, Aitemative Safe Shutdown Procedure Index concurrently with OASSD-02, Control Building.

B. (1) trip both Reactor Recirc pumps (2) perlomi OASSD-01, Alternative Safe Shutdown Procedure Index concurrently with QASSD-02. Control Building.

O (1) place MSIV control switches in close (2) exit OASSD-0 1, Alternative Safe Shutdown Procedure Index and enter OASSD-02, Control Building D. (1) trip both Reactor Recirc pumps (2) exit OASSD-01, Alternative Safe Shutdown Procedure Index and enter OASSD-02, Control Building

15.2 IF the fire is in the Control Building fire area, AND control room evacuation is required, ThEN PERFORM the following:

a. MANUALLY SCRAM Unit 1 reactor. El
b. PLACE Unit I MSIV control switches in CLOSE. El
c. MANUALLY SCRAM Unit 2 reactor.
d. PLACE Unit 2 MSIV control switches in CLOSE.
e. Both units EXIT this procedure AND ENTER OASSD-02, Control Building.

OASSD-O1 Rev. 41 Page 50115

10 OPERATOR ACTIONS 3.5.3 IF the fire is NOT in the Control Building, THEN ENTER the applicable ASSD procedure AND EXECUTE concurrently with this procedure.

NOTE: A loss of drywell cooling can be determined using CAC-TR-4426-1A, CAC-TR-4426-2A, in the Control Room or CAC-TR-770, points 1 3, and 4, at the Remote Shutdown Panel. The time the loss of drywell cooling occurred can be determined from the recorder display information.

3.5.4 IF RCIC or HPCI is injecting AND drywell cooling has been lost, THEN START reactor vessel cootdown at 100F/hr or greater within 60 minutes of the loss of drywell cooling.

3.5.5 IF drywell temperature control is lost, THEN PERFORM the following to preserve containment overpressure for RHR pump net positive suction head:

1. STOP A, B, C, and D RBCCW pumps for the affected unit.
2. CLOSE the following valves for the affected unit

- DW EQUiP DRAIN INBD ISOL VLV, G76-F019 Li

- DW EQUiP DRAIN OTBD (SQL VLV, G16-F020

- DW FLOOR DRAIN INBD (SQL VLV, G16-F003 C

- DWFLOOR DRAIN OTBD (SQL VLV. 616-f004 OASSD-01 Rev. 41 Page 6 of 16

2.1 This procedure is entered from A[tematve Safe Shutdown Procedure Index, OASSD-O1, AND 2.2 Unit CR5 has determined 1)0th reactors are to be brought to safe and stable conditions from outside the Control Room using ASSD Train B.

3.0 OPERATOR ACTIONS 3.1 CONTINUE implementation of this procedure by perfom,ing the steps w Section A.

3.2 IF the fire is extinguished while executing this procedure AND the Unit CRS determines no action within this procedure is required.

THEN EXIT this procedure.

4.0 RESTORATION 4.1 RETURN plant to general operating condition as directed by plant management OASSD-02 Rev. 57 Page 3 of 154

93. S295021 I Unit One is in MODE 4, when a loss of SDC occurs due to RCS leakage.

cul UNPLANNED loss of RPV invenkny for 15 minutes or longer I I I 14151 I cu1.1 UNPLANNED loss of reactor coolant results in RPV water level less than a required lower limit for 15 mm. (Note 1)

Which one of the following completes both of the statements below?

The minimum required RPV water level to support natural circulation is (1) lAW OPEP-02.2.1, Emergency Action Level Technical Bases, the Unusual Event required lower limit is defined as RPV water level less than (2)

A. (1) 200 inches (2) 105 inches B. (1) 200 inches (2) 166 inches C. (1) 254 inches (2) 105 inches D. (1) 254 inches (2) 166 inches Answer: B K/A:

295021 Loss of Shutdown Cooling AA2 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING: fCFR: 41.10/43.5/45.13) 03 Reactor water level RO/SRO Rating: 3.5/3.5 Tier 1 I Group 1 K/A Match: This meets the K/A because it is testing the ability of the applicant to determine the water level designated for a UE with a loss of SDC.

Pedigree: Modified from 10-1 Objective: LO-CLS-LP-301-B Obj. 6 Define the relationship between fission product barrier loss/potential loss and each of the four emergency classifications.

Reference:

None Cog Level: Fundamental

Explanation: Part 1: lAW OAOP-15.0, the minimum required level for natural circulation is 200 inches. Part 2: lAW OPEP-02.2.1, the RPV water level lower limit is the low end of any established band.

Distractor Analysis:

Choice A: Parti is plausible because this is correct, see explanation. Part 2 is plausible because this would be the lower limit if no band was established for SDC in MODE 4.

Choice B: Correct Answer, see explanation Choice C: Part 1 is plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC.

Choice D: Correct Answer, see explanation.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J. The applicant is required to have bases knowledge of the EAL procedure to select the appropriate method of implementing the UE criteria for a lowering RPV level in MODE 4. EAL determination is an SRO only task.

Question from NRC 10.1 exam:

While in Mode 4 a loss of Shutdown Cooling (SDC) occurs.

Which one of the following completes both statements?

The minimum required Reactor Water Level to support Natural Circulation is (1) inches.

An Alert declaration is first required after an unplanned RPV pressure increase greater than (2) psig due to a toss of RCS cooling.

A. (1) 200 (2) 135 B (1) 200 (2) 10 C. (1) 254 (2) 135 D. (1) 254 (2) 10 LOSS OF SHUTDOWN COOLING OAOP-15.0 Rev. 31 Page 7 ot25 4.2 Supplementary Actions (continued)

2. IF forced circulation has been lost, AND natural circulation has NOT been established, THEN ensure reactor vessel water level is being maintained between 200 inches and 220 inches as read on B2f-LI-R605A(B)

(RPV Water Level),

OR as directed by the Unit CR5 based on plant conditions until forced circulation is restored El

ATTACHMENT I Page 47 of 200 EAL Eases Category: C Cold Shutdown / Refueling System Maltunction Subcategory: 1 RPV Level Initiating Condition: UNPLANNED toss of RPV inventory for 15 minutes or longer EAL:

CULl Unusual Event UNPLANNED loss of reactor coolant results in RPV water level less than a required lower limit for 15 mm. (Note 1)

Nate 1: The SEC should declare the event promptly upon determining that time limithas been exceeded, or will likely be exceeded.

Mode Applicability:

4 Cold Shutdown, 5 Refueling Definition(s):

UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

Figure C-i illustrates the elevations of the RPV level instrument ranges (ret. 2).

With the plant in Cold Shutdown, RPV water level is normally maintained above the RPV tow level scram setpoint of 166 in. above TAF (ref. 1, 3). However, if RPV level is being controlled below the RPV low level scram setpoint, or if level is being maintained in a designated band in tile reactor vessel it is tile inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.

Figure c-i RPV Levels (ref. 2)

Ricto Water Levd Iiistmwent Rançs 17 FRDf VESSEL ZERO PPRQX ai I

w ST&AMUNE N SPILLOVERI I824ON R N

NARPOW E

I N004 NO?

ThF 1 z CORE E NOS3NQ7

94. SG2.1 .05 1 Which one of the following completes both statements below?

(Consider each statement separately.)

lAW Tech Spec 5.2.2, Facility Staff, the shift crew composition may be less than the minimum requirement for a period of time not to exceed (1) for an unexpected absence of on-duty shift crew members.

lAW 001-01.01, BNP Conduct of Operations Supplement, the minimum required number of Auxiliary Operators for manning a shift at BNP is (2)

A. (1) one hour (2) three B. (1) one hour (2) nine C. (1) two hours (2) three D. (1) two hours (2) nine Answer: D K/A:

G2.1 .05 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (CFR: 41.10 /43.5 /45.12)

RO/SRO Rating: 2.9/3.9 Tier 3 K/A match: knowledge of required tech spec and conduct of ops shift manning requirements.

Pedigree: 2012 BNP NRC Objective: LOI-CLS-LP-200-B Obj.12.-ldentify conditions and limitations in the facility license.

Reference:

None Cog Level: Low Explanation: lAW the procedure, 9 AO makeup the minimum shift staffing and two hours is the time to find a replacement. One hour is the time on stepping out limitation of the control room personnel. The tech Specs 5.2 only address the number of AOs for the Units which is 3, this does not take into account ASSD and Fire Brigade.

Distractor Analysis:

Choice A: Plausible because IS 5.2.2 requires 3 AOs for both Units which does not take into account ASSD and Fire Brigade requirements. One hour is the stepping out time limit for control room personnel.

Choice B: Plausible because nine is correct but one hour is the stepping out time limit for control room personnel.

Choice C: Plausible because IS 5.2.2 requires 3 AOs for both Units which does not take into account ASSD and Fire Brigade requirements.

Choice D: Correct Answer, see explanation SRO Basis: Conditions and limitations in the facility license. (10 CFR 55.43(b)(1)). Requires the SRO applicant to know the limitations for shift staffing in the license.

BNP CONDUCT OF OPERATIONS SUPPLEMENT 001-01.01 Rev. 73 Page 15 of 191 5.5 Operations Shift Staffing 5.5.1 General

1. In addition to the requirements of AD-OP-ALL-i 000, the following requirements apply:
a. The following table outlines the administrative guideline fot the normal Operations shift complement. Any deviation from the normal shift complement must remain in accordance with Section 5.2.2 of Technical Specifications, applicable sections of OASSD-00, Users Guide OFPP-031 Fire Bngade Staffing Roster and Equipment Requirements, and OERP, Radiotogical Emergency Response Plan (ERP).

(Attachment 13, Operations Staffing Roster contains a listing of required ERO Watch Stattons and qualifications for each and ASSD positions.)

BNP Watchstations BNP Shift Complement License Shift Manager (SM) I Shift Manager SRO Control Room Supervisor (CR5) 2 CRSs tf for each unit) SRO Reactor Operator (RO) 4 Reactor Operators (typically, 2 ROISRO for each unit)

Auxiliary Operator fAQ) 9 (includes 2 in Radwaste) NA Operations Center SRO 1 Operations Center SRO SRO STA [Note 1] 1 STA STA Qualified Notes:

Organization 5.2 5.2 Organization 5.2.2 Facility Staff (continued)

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, when either unit is in MODE 1 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room. With one unit in MODE 1,2, or 3 and the other unit defueled, the minimum shift crew shall include a total of two SROs and two ROs.
c. Shift crew composition may be less than the minimum requirement of 10 CFR 5054(rn){2)(i) and Specifications 5.IZa and 522.g for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
d. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
e. Deleted.
95. SG2.1.43 I Following the bypass of Unit Two feedwater heaters 4A and 5A, the following plant conditions exist:

Reactor Power is 60%

Feedwater Temperature is 330°F Final Feedwater Temperature vs Power

. Nominal FW Temp 11 0.3F Reduced RX PWR Nominal FW Temp Reduced 1 ODF FFWT 65% 394A 384A 296A 64% 393.1 383.1 295.5 63% 391.7 381.7 294.6 62% 390.4 380.4 293.7 61% 389.0 379.0 292.8 60% 387.6 377.6 291.9 (REFERENCE PROVIDED) lAW 001-01.01, BNP Conduct of Operations Supplement, which one of the following completes both statements below? (consider each statement separately)

The CRS (1) required to implement the thermal limit penalties for FHOOS (feedwater heater out of service).

Entry into Tech Spec 3.0.3 (2) required if final feedwater temperature is less than the 110.3°F reduced final feedwater temperature value.

A. (1) is (2) is B. (1) is (2) is NOT C. (1) is NOT (2) is D. (1) is NOT (2) is NOT Answer: B K/A:

G2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc. (CFR: 41.10/43.6/45.6)

ROISRO Rating: 4.1/4.3 Tier 3

K/A match: The applicant is requited to use the final feedwater temperature reduction attachment to determine if the effect of the feedwater teduction is severe enough on reactivity to require implementation of thermal limit penalties.

Pedigree: New Objective: CLS-LP-032 obj 27 Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR, determine whether given plant conditions meet minimum Technical Specifications requirements associated with the Condensate and Feedwater System.

Reference:

T.S. 3.2.1, T.S. 3.2.2, and IS. 3.2.3 Cog Level: High Explanation:  : Parti: A final fw temp of 385°F is less than the nominal FW temp for 60% power, but >10°F reduced from nominal, therefore the thermal limit penalties for FHOOS do not need to be implemented. Part 2:There are NO core operating limits specified in the COLR for operation beyond 110.3°F Final Feedwater Temperature. Thermal limits CANNOT be verified to be within the limits specified in the COLR, which requires entry into the Actions of LCO 3.2.1, 3.2.2, and 3.2.3. These LCOs require thermal limits to be restored within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, LCO 3.0.3 is not entered, Distractor Analysis:

Choice A: Part 1 is correct, see explanation. Part2 is plausible because a candidate may believe since there are no thermal limits specified in the COLR for this condition, LCO 3.0.3 would be applicable.

Choice B: Correct Answer, see explanation Choice C: Part 1 is plausible because a final fw temp of 330°F is less than the nominal FW temp reduced by 10°F for 60% power, but greater than the 110.3°F reduced FFWT, a novice applicant may believe thermal limit penalties are only applied at the 110.3°F value. Part2 is plausible because a candidate may believe since there are no thermal limits specified in the COLR for this condition, LCO 3.0.3 would be applicable.

Choice D: Part 1 is plausible because a final fw temp of 330°F is less than the nominal FW temp reduced by 10°F for 60% power, but greater than the 110.3°F reduced FFWT, a novice applicant may believe thermal limit penalties are only applied at the 110.3°F value. Part 2 is correct, see explanation.

SRO Basis: Facility operating limitations in the IS and their bases. [10 CFR 55.43(b)(2)] This question requires that the applicant determines whether the TS thermal limits should incur a penalty. In addition, it also requires that the candidate determines whether LCO 3.0.3 applies for a given condition.

LCO 3.0.3 When an LCD is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed bythe associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCD is not applicable. Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:

a. MODE 2 v.ithin 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />s:
b. MODE 3 within f3 hours: and
c. MODE 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

Exceptions to this Specification are stated in the individual Specifications.

Where corredivo measures are completed that permit operation in accordance with the LCD or ACTIONS, completion of the actions required by LCD 3.0.3 is not required.

LCD 3.0.3 is only applicable in MODES 1,2. and 3.

8NP CONDUCT OF OPERATIONS SUPPLEMENT 001-01.01 Rev. 76 Page 131 of 191 ATTACHMENT 19 Page 3 of 4

<< Equipment Out Of Service Contingencies>>

EOOS Power Condition Required Action (Note 1.2)

. Reduce reactor pcwerta .

. Imptement OGP-14.

SLO Any

  • Implement applicable SIC power to flaw map.

. IF 23% RTP.THEN implement thermal limit penalty.

  • ReferenceTS3.4.1.
  • Implement thermal limit penatty.

TBVQOS 23% RTP

. IF < the value in the 1 t0.3F Reduced FiN Temp column of 112)OP-32, 23%RTP Att6.

FHOOS THEN enter LCO 3.2.1, 3.2.2 and 3.2.3.

fFWTR)

  • IF> 1OF below nominal FW temperature, (FFTR) 23% RTP 0 Implement applicable FWTR power to flow map.

0 Implement thermal btt penafty.

. I.,.nI,nn+ nnI,..kI V.flO r.,,r I,, fln, n,..

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev. 206 Page 400 of 408 ATTACHMENT 6 Page 2 of 2 Final Feedwater Temperature vs Power Nominal FWIemp 110.3DF Reduced RXPWR Nominal FWTemp Reduced 10SF FFWT 65% 394.4 384.4 296.4 64% 393.1 383.1 295.5 63% 391.7 381.7 294.6 62% 390.4 380.4 293.7 61% 389.0 379.0 292,8 60% 387.6 377.6 291.9 59% 386.2 3762 290.9

96. SG2.2.15 1 Unit One is operating at rated power.

A-03 (2-2) Auto Depress Control Pwr Failure, is in alarm due to Fuse F5 being blown.

(REFERENCE PROVIDED)

Which one of the following completes both statements below?

Fuse F5 (Dl on 1-FP-05887) is located on ADS Logic (1)

ADS (2) operable.

A.(1) A (2) is B. (1) A (2) is NOT C. (1) B (2) is D. (1) B (2) is NOT Answer: C K/A:

G2.2.15 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc. (CFR: 41.10 /43.3 /45.1 3)

RO/SRO Rating: 3.9/4.3 Tier 3 K/A match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL.

This question requires the candidate to use a drawing to determine operability of ADS.

Pedigree: New Objective: LOI-CLS-LP-020 OBJ 15d. Given plant conditions, predict how ADS/SRVs will be affected by the following: Loss of DC power

Reference:

1-FP-05887 (Block out references to which logic string is logic A and B)

Cog Level: High Explanation: Part 1: Fuse F5 is located on the alternate power source, only logic B has an alternate power source. Therefore, the fuse is on logic B, Part2: Since the drawing is shown in the de-energized state, fuse 5 being blown will have no impact on ADS instrumentation. ADS remains on its normal power source.

Distractor Analysis:

Choice A: Part 1 is plausible the fuse is located on 125V DC 3A power or the operator might forget which train of logic has two power supplies. Part 2 is plausible because it is correct, see explanation.

Choice B: Part 1 is plausible the fuse is located on 125V DC 3A power or the operator might forget which train of logic has two power supplies. Part 2 is plausible because if the drawing was shown in the energized state, this would be correct.

Choice C: Correct Answer, see explanation.

Choice D: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because if the drawing was shown in the energized state, this would be correct.

SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Requires the SRO candidate to use a drawing to determine the status of ADS power and to know whether that loss effects ADS operability.

ctci?S w,+ ANb C {Ar 1 FC shown energized here:

-U-B LOGIC BUS F 12% IUC PNLVA B K1B

+

pr..[ 3[/. S KlL Ii B

125VDC PNL (4)A

97. SG2.2.22 I Unit Two is operating at rated power.

While performing OPT-07.2.4A, Core Spray Loop A Operability, Core Spray Room Cooler A fails to start when Core Spray Pump A is started.

The reactor building AC reports that the room cooler tripped on thermal overload.

lAW AD-OP-ALL-bOO, Conduct of Operations, which one of the following completes both statements below? (consider each statement separately)

Core Spray Loop A is (1)

A one time reset of the thermal overload (2) allowed before a Maintenance and Engineering evaluation.

A. (I) OPERABLE (2) is B. (I) OPERABLE (2) is NOT C. (1) INOPERABLE (2) is D. (1) INOPERABLE (2) is NOT Answer: D K/A:

G2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 /43.2 /45.2)

ROISRO Rating: 4.0/4.7 Tier 3 K/A match: Requires knowledge of conduct of ops procedure to determine whether Core Spray A meets the conditions for operability in the tech specs based on cooler operation.

Pedigree: Bank NRC 08 Objective: CLS-LP-1 8, Obj. 18. Given plant conditions and TS, including bases, TRM, ODCM, and COLR, determine the required actions to be taken in accordance the IS associated with the Core Spray System.

Reference:

None Cog Level: Hi

Explanation: Part 1: Per 001-01.01, When any ECCS Room Cooler is determined to be INOPERABLE, then the ECCS equipment associated with that room cooler is to be declared INOPERABLE per the applicable Technical Specifications. Part 2: Per AD-OP-ALL-i 000, the breaker should only be reset once the condition is identified and corrected, and plant conditions dictate the reset before maint and eng personnel are available.

Distractor Analysis:

Choice A: Part 1 is plausible, because the room cooler is not part of the Core Spray system listed in the tech spec bases. Part 2 is plausible because during transient conditions the breaker could be reset, however, the plant is in a stable condition.

Choice B: Part 1 is plausible, because the room cooler is not part of the Core Spray system listed in the tech spec bases. Part 2 is correct, see explanation.

Choice C: Part 1 is correct, see explanation. Part 2 is plausible because during transient conditions the breaker could be reset, however, the plant is in a stable condition.

Choice D: Correct Answer, see explanation.

SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Requires the SRO candidate to have knowledge of conduct of ops procedure to determine whether Core Spray A meets the conditions for operability in the tech specs based on cooler operation.

CONDUCT OF OPERATIONS AD-OP-ALL-i 000 Rev. 4 Page 80 of 133 5.19 Resetting Protective Devices

{7f.4}

5.19.1 Standards

1. Protective devices should not be reset without a clear understanding of the reason for the protective device trip.
2. The overriding priority for the operating crew upon the trip of any protective device is to stabilize the plant and restore the systems to the safest possible condition.

5.19.2 Expectations

1. Protection devices which have actuated (breakers, fuses, bistables, MOV themial overloads, lockouts, etc.) should only be restored with shift supervision approval, under the following conditions. The following conditions do not apply to 120 volt breakers that only supply lighting or receptacles.
a. The cause of The actuation has been identified and corrected.
b. Restoring the protective device is not recommended unless plant conditions dictate that the component repositioning must be completed before Maintenance and Engineering personnel are available. Remote operation of the component with no personnel in the immediate area after resetting the protective device is recommended if repositioning is required prior to completion of The evaluation by Maintenance and Engineering.
2. The SM may approve additional protective device resetting after consultation with Engineering.

BNP CONDUCT OF OPERATIONS SUPPLEMENT COt-al .01 Rev. 73 Page 36 ot 191 5.16.2 Degraded Equipment Controls System)Component Related Guidance (continued)

(1) Reference TRM Appenthx F, Safety Function Determination Program (SFDP), Attachments 1 and 2 to assist with determination of Technical Specification 3.81 and 3.8.7 requirements and to assess the possible impact on supported systems.

(2) If an evaluation of the SFDP is performed, then document the evaluation and the results in the the narrative log or on Attachment 26, if the narrative log is not available.

4. ECCS Room Coolers U.13}

NOTE

  • The following step is not required to be performed if the ECCS Room Cooler is INOPERABLE due to the toss of a 4160V or 480V E-Bus. E-Bus INOPERABILITY impacts the OPERABILITY of ECCS subsystems. Technical Specifications and the SFDP will provide Requited Actions to be taken for the loss of the E-Bus.
a. When any ECCS Room Cooter is determined to be INOPERABLE, then the ECCS equipment associated with that room cooler is to be declared INOPERABLE per the applicable Technical Specifications.

EXAMPLE The RHR Room Coolers are to be considered redundant components required to support the operation of RHR. Therefore, should a room cooler be found or made INOPERABLE, a 7 day Active LCO is required to be established on the RHR system. Likewise, should both room coolers be found INOPERABLE, the action required is the same as if both RHR loops and HPCI were INOPERABLE. Should it be identified that one RHR Room Cooler is INOPERABLE and one RHR Loop is also INOPERABLE (specific combinations do not matter), the action is as if only one RHR Loop is INOPERABLE (7 days).

flnntrr Ri,ilrtinn WVA(. Air (nmnmcnr

98. SG2.3.11 I Following a small steam line break in the drywell plant conditions are as follows:

Drywell pressure: 25 psig and rising Drywell hydrogen: 1.3%

Suppression Chamber hydrogen: 1 .2%

Torus level: 42 inches Which one of the following completes both statements below?

The CRS is required to direct venting containment lAW OEOP-O1-SEP-O1, Primary Containment Venting, using (1)

Venting of the (2) will be directed first.

A. (1) Section 2.1, Containment Pressure Control (2) drywell B. (1) Section 2.1, Containment Pressure Control (2) torus C. (1) Section 2.2, Containment Hydrogen Control (2) drywell D. (1) Section 2.2, Containment Hydrogen Control (2) torus Answer: D K/A:

G2.3.1 1 Ability to control radiation releases. (CFR: 41.11/43.4 / 45.10)

RO/SRO Rating: 3.8/4.3 Tier 3 K/A match: Requires the ability to determine the procedure section for venting, and the correct sequence of termination of venting.

Pedigree: New Objective: CLSLP300L*08d Given the Primary Containment Control Procedure and plant conditions, determine if the following actions are required: Venting the primary containment IRRESPECTIVE of radioactivity release rate limits

Reference:

None Cog Level: High Explanation: Part 1: Following the H2 leg of the PCCP with the given conditions will drive you to step PC/G-9 which directs you to Vent Containment per EOP-01 -SEP-01, since H2 is the driving condition for venting, then section 2.2 is the appropriate section to implement. Part 2: lAW SEP-01, the torus is vented first as long as the torus water level is less than 6 feet.

Distractor Analysis:

Choice A: Part 1 is plausible because H2 concentration is less than the entry limit into PCCP (entry at 1.5%), and Containment pressure is >11 .5 psig (pressure for DW sprays), therefore a novice applicant might believe that the appropriate procedure section required is for containment pressure control. Part 2 is plausible since venting of the drywell is performed first if torus water level is >6 feet.

Choice B: Part 1 is plausible because H2 concentration is less than the entry limit into PCCP, and Containment pressure is >11 .5 psig (pressure for DW sprays), therefore a novice applicant might believe that the appropriate procedure section required is for containment pressure control. Part 2 is correct.

Choice C: Part 1 is plausible since it is correct, see explanation. Part 2 is plausible since venting of the drywell is performed first if torus water level is >6 feet.

Choice D: Correct Answer, see explanation..

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J Requires knowledge of diagnostic step in EOP, and selection of appropriate emergency contingency procedure.

/WH EN\

Orywel.QR tows H2reachesl%

THEN /

\continue/ PCG-5 Notify E&C to sample coritanment for venttog.

i1r Trip Hydrogen Water Chemistry.

PCiG-7

()

Vent containment per EOP-O1-SEP-al.

PCIG-9

PRIMARY CONTAINMENT VENTING OEOP-0 1-SEP-01 Rev. 026 Page 120121 2.2.3 Containment Hydrogen Control Actions (continued)

e. Close VA-D-BFV-RB (SBGT A Isol Damper) El RO
f. Close VA-H-B FV-RB (SBGT B Isol Damper) El RO
g. IF venting the tows, THEN open:

(1) CAC-V7 fToms Purge Exh Vie) El RO (2) CAC-V$ (Tows Purge Exh VIv) El RO

h. IF venting the dr1weli.

THEN open:

ff) CAC-V9 tDrywell Purge Exh VIv) El RO (2) CAC-V1O fDr,weII Purge Exh Vlv) El RO

i. Open VA-F-BFV-RB (SBGT OW Suct Damper) El RO
13. IF directed to terminate tows venting, THEN:
a. Ensure primary containment purging terminated per EOP-Oi-SEP-05 El RO
99. SG2.4.30 I Unit Two is operating at rated power with LPCI A inoperable and the following sequence of events occurs:

0000 7 day completion time for LCO 3.5.1, ECCS Operating, Condition A expires and Condition C is entered requiring that the Unit be placed in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

0030 Plant shutdown is commenced per LCO 3.5.1, Condition C.

0050 LPCI A is repaired and declared operable; LCO 3.5.1 Conditions A and C are exited.

0100 Management decides to continue the plant shutdown as planned to complete other maintenance items.

0230 UnitTwoin MODE3 (REFERENCE PROVIDED)

Which one of the following completes both statements below?

lAW 01-01 .07, Notifications, an Emergency Notification System (ENS) report to the NRC must be submitted no later than (1) lAW 10 CFR 50.73, Licensee Event Reporting System, an LER (2) required.

A. (1) 0400 (2) is B. (1) 0400 (2) is NOT C. (1) 0430 (2) is D. (1) 0430 (2) is NOT Answer: D K/A:

G2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11)

RO/SRO Rating: 2.7/4.1 Tier 3 K/A match: Applicant required to determine report status for the given condition.

Pedigree: Bank Objective: LOl-CLS-LP-201-D, Obj 11 Explain the following regarding NRC Reporting requirements per AD-LS-ALL-0006, Notification/Reportability Evaluation: d. Determination of clock start time for reportable events (LOCT)

Reference:

001-01.07 Attachment 1, NUREG 1022 Table 1 Cog Level: High Explanation: A IS required shutdown requires a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> NRC report. The time starts when the shutdown is actually started, Completion of the shutdown required by a IS is an LER.

Distractor Analysis:

Choice A: Plausible because this would be 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from when the TS shutdown condition was entered.

Part 2 plausible because an LER is required after completing a TS required shutdown, but in this case the shutdown was not TS required.

Choice B: Plausible because this would be 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from when the TS shutdown condition was entered.

Choice C: Plausible because an LER is required after completing a TS required shutdown, but in this case the shutdown was not IS required.

Choice D: Correct Answer, see explanation.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]. Requires SRO administration procedure knowledge of reportability requirements based on plant conditions.

NOTIFICATIONS 001-01.07 Rev. 35 Page 250143 ATTACHMENT 1 Page 2 of 7 Reportabi I ity Evaluation Checklist NOTE

  • If the answer to any of the following questions is YES, the event is reportable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
  • If all answers to the following questions are NO, the event is not reportable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4 HOUR REPORTABILITY ITEM # YES NO DESCRIPTIVE QUESTION NOTE Includes any Safety Limit vioIaon (Tech Spec 2.2).

2 Is plant shutdown requited by technical specifications being initiated?

[10 CFR 50.72fb)2)(i)I Plant Shutdown Required by Technical Specifications (See Section 3.2.1 of this report)

§ 50.72(b)(2)(i) The initiation of any nuclear § 50.73fa)(2)(i)(A) The completion of any plant shutdown required by the plants Technical nuclear plant shutdown required by the plants Specifications. Technical Specifications.

Discussion The 10 CER 50.72 reporting requirement is intended to capture those events for which TS require the initiation of reactor shutdown to provide the NRC with early warning of safety-significant conditions serious enough to warrant that the plant be shut down. For 10 CFR 50.72 reporting purposes, the phrase initiation of any nuclear plant shutdown includes action to start reducing reactor power; i.e., adding negative reactivity to achieve a nuclear plant shutdown requited by TS. This includes initiation of any shutdown due to expected inability to restore equipment prior to exceeding the LCO action time. As a practical matter, in order to meet the time limits for reporting under 10 CFR 50.72, the reporting decision should sometimes be based on such expectations. (See Example 4.)

The initiation of any nuclear plant shutdown does not include mode changes required by TS if they are initiated after the plant is already in a shutdown condition.

A reduction in power for some other purpose, not constituting initiation of a shutdown required by TS, is not reportable under this criterion.

For 10 CFR 50.73 reporting purposes, the phrase completion of any nuclear plant shutdown is defined as the point in time during a TS-required shutdown when the plant enters the first shutdown condition required by an LCO (e.g., hot standby (Mode 3) for PWRs with the Standard Technical Specifications (STS)). For example, if at 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> a plant enters an LCO action statement that states, restore the inoperable channel to operable status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the plant must be shut down (i.e., at least in hot standby) by 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />. An LER is required if the inoperable channel is not returned to operable status by 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> and the plant enters hot standby.

An LER is not required if a failure was or could have been corrected before a plant has completed shutdown (as discussed above) and no other criteria in 10 CFR 50.73 apply.

NOTIFICATIONIREPORTABILIW EVALUATION AD-LS-ALL-0006 Rev. 0 Page 14 of 17 5.4 Making Emergency Notification System and LER Reports (continued)

Table 1. Emergency Notification System Reporting Overview Event or ENS notification ENS notification ENS notification 60-day LER Job Aid condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> within B hours Section Plant shutdown Initiation of SD compleon of a A.3. A.4 (310) required by required by Tech S1D required by Tech Specs Specs [50.72 Tech Specs [50.73 (b)(2)fi)] (a12)fi)(A)]

100. SG2.4.35 I Unit One and Unit Two have entered SBO procedures at time 1300 due to a loss of all onsite and offsite power.

Which one of the following completes both statements below?

lAW I EOP-01 -SBO, Station Blackout, opening the reactor building roof hatch is required to be performed no later than (I) lAW 001-37.14, Station Blackout Procedure Basis Document, the reactor building doors and roof hatch are opened to ensure (2)

A. (1) 1330 (2) equipment availability B. (1) 1330 (2) habitability C. (1) 1500 (2) equipment availability D. (1) 1500 (2) habitability Answer: D K/A:

G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR: 41.10/43.5 /45.13)

RO/SRO Rating: 3.8/4.0 Tier 3 K/A match: Requires the applicant to have knowledge of when to implement the AO task in a EOP sub procedure (opening RB roof hatch during SBO), and the operational effects if not completed (jeopardized habitability).

Pedigree: New Objective: LOI-CLS-LP-303-B Obj 2 Given plant conditions, EOP-01-SBO Flowchart, and SBO Support Procedures, determine the required operator actions. Temperature analysis states that access would be prohibited in the RB building due to 117 elevation ceiling temperature if the hatch was not opened.

Reference:

None Cog Level: Fundamental Explanation: This is a time sensitive action from the SBO procedure that directs the RB roof hatch to be opened within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the start of the SBO.

Distractor Analysis:

Choice A: Part 1 is plausible because this is the time critical action time limit in SEQ procedure for opening control panel doors. Part 2 is plausible because high temperatures could be thought to jeopardize equipment availability, however the hatch and doors are opened to ensure habitability.

Choice B: Part 1 is plausible because this is the time critical action time limit in SEQ procedure for opening control panel doors. Part 2 is correct, see explanation.

Choice C: Part 1 is correct, see explanation. Part 2 is plausible because high temperatures could be thought to jeopardize equipment availability, however the hatch and doors are opened to ensure habitability.

Choice D: Correct, see explanation.

SRQ Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)] Knowledge of when to implement attachments and the basis for the step.

STATION BLACKOUT 001-37.14 PROCEDURE BASIS DOCUMENT Rev. 001 Page 40 of 42 5.25 Steps SBO-27 and SBO-28 4

Open reactor budding doors per EOF-O 1 -SBO-04. Time SensItive 5[>-2T RB roof hatch required open whn 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

11r Stag. alternate fuel pool rnakeup(spray Time equipment per EOP-O1-SEP-12 I SBO28 It ELAP conditions exist, reactor building temperatures lI rise rapidly due to the loss of building ventilation. The refuel floor roof hatch and 20 elevation personnel access doors are blocked opened to provide alternate ventilation. The refuel floor roof hatch should be opened as soon as resources are available and is required to be open within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the SBO start time recorded at Step SBO-I. The 20 elevation personnel access doors should be opened as soon as resources are available and are required to be open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the SEQ start time recorded at Step SBO-1 by the text procedure. The reactor building temperature analysis (SNP-MECH-FLEX-0001) shows 117 elevation ceiling temperature will reach 114°F at time 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, which is approaching the temperature that access would be prohibited. Alternate ventilation should be established as eally as possible based on priorities and available resources.

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