ML082681838
| ML082681838 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 09/24/2008 |
| From: | Duke Energy Carolinas |
| To: | NRC/RGN-II |
| References | |
| 50-369/08-301, 50-370/08-301, ES-401-2 IR-08-301 | |
| Download: ML082681838 (324) | |
See also: IR 05000369/2008301
Text
Draft Submittal (Pink Paper)Senior Reactor Operator Written Exam..MAY 2008 EXAM-50-369 370/2008-301
.:SRO WRITTEN EXAM
a inati n ritten Exa (SRO)McGuire Nuclear Station 45-day Submittal March 17, 2008
Site-Specific
Written Examination
McGuire Units 1 and 2 Senior Reactor Operator Answer Key 1.B 26.B 51.B 76.A 2.D 27.C 52.A 77.D 3.C 28.B 53.C 78.C 4.A 29.A 54.A 79.D, 5.D 30.C 55.B 80.D 6.D 31.C 56.A 81.B 7.C 32.A 57.A 82.A 8.A 33.B 58.A 83.A 9.C 34.B 59.0 84.A 10.B 35.C 60.B 85.D 11.A 36.B 61.C 86.A 12.C 37.0 62.C 87.A 13.C 38.A 63.A 88.A 14.A 39.0 64.A 89.B 15.B 40.C 65.C 90.A 16.C 41.B 66.B 91.D 17.A 42.C 67.A 92.B 18.C 43.B 68.C 93.C 19.C 44.A 69.A 94.B 20.A 45.C 70.C 95.A 21.D 46.A 71.0 96.B 22.A 47.0 72.B 97.D 23.B 48.C 73.B 98.A 24.A 49.A 74.A 99.A 25.B 50.C 75.B 100.D-1-Draft Validation
7
Outline Form ES-401-2 c Facility: McGuire 2008 NRC Date of Exam: 5/12/2008 Exam RO KIA Category Points SRO-Only Points Tier GroupKKKKKKAA A A G G*123 4 5 6123 4*Total A2 Total 1.133333 3 1833 6 Emergency 212122 1 9 2 2 4&Plant Tier Evolutions
Totals 4 5 455 4 2755 101223 323223 3 3 2832 5 2.Plant 211111111110
10 021 3 Systems Tier Totals3344 3 43344 3 38 538 3.Generic Knowledge&Abilities12 3 4123 4 10 7 Categories
3 3221222 Note: 1.Ensure that at least two topics from every applicable
KIA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the"Tier Totals" in each KIA category shall not be less than two).2.The point total for each group and tier in the proposed outline must match that specified in the table.The final point total for each group and tier may deviate by+/-1 from that specified in the table based on NRC revisions.
The final RO exam must total 75 points and the SRO-only exam must total 25 points.3.Systems/evolutions
within each group are identified
on the associated
outline;systems or evolutions
that do not apply at the facility should be deleted and justified;
operationally
important, site-specific
systems that are not included on the outline should be added.Refer to section D.1.b of ES-401, for guidance regarding elimination
of inappropriate
KIA statements.
4.Select topics from as many systems and evolutions
as possible;sample every system or evolution in the group before selecting a second topic for any system or evolution.
5.Absent a plant specific priority, only those KAs having an importance
rating (IR)of 2.5 or higher shall be selected.Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
6.Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.
7.*The generic (G)KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable
evolution or system.Refer to Section D.1.b of ES-401 for the applicable
KIA's 8.On the following pages, enter the KIA numbers, a brief description
of each topic, the topics'importance
ratings (IR)for the applicable
license level, and the point totals (#)for each system and category.Enter the group and tier totals for each category in the table above.If fuel handling equipment is sampled in other than Category A2 or G*on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note#1 does not apply).Use duplicate pages for RO and SRO-only exams.9.For Tier 3, select topics from Section 2 of the KIA Catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#)on Form ES-401-3.Limit SRO selections
to KlAs that are linked to 10CFR55.43
ES-401 2 McGuire 2008 NRC Exam Written Examination
Outline Emergency and Abnormal Plant Evolutions
-Tier 1 Group 1 Form ES-401-2 11.,;1
,;.;K;;;.",&,21
.,;,K,;,;.,3
====,-1.::.Im;.;.lp;.;.,009/Small Break LOCA/3 X 2.1.27-Conduct of Operations:
Knowledge 4.0 76 of svstem puroose and/or function.026/Loss of Component Cooling 2.1.25-Conduct of Operations:
Ability to X interpret reference materials, such as 4.2 77 Water/8 araphs, curves, tables, etc.029/Anticipated
Transient Without EA2.01-Ability to determine or interpret 78 X the following as they apply to a ATWS: 4.7 Scram (ATWS)/1 Reactor nuclear instrumentation
EA2.03-Ability to determine or interpret 055/Station Blackout/6 X the following as they apply to a Station 4.7 79 Blackout:Actionsnecessary
to restore power AA2.18-Ability to determine and interpret the following as they apply to the Loss of 80 056/Loss of Off-site Power/6 X Offsite Power: Reactor coolant 4.0 temperature, pressure, and PZR level recorders 2.2.37 Equipment Control: Ability to 81 058/Loss of DC Power/6 X determine operability
and/or availability
of 4.6 safety related equipment 2.4.46-Emergency Procedures
/Plan: 007/Reactor Trip/1 X Ability to verify that the alarms are 4.2 39 consistent
with the plant conditions.
AA 1.03-Ability to operate and/or monitor 008/Pressurizer
Vapor Space the following as they apply to the X Pressurizer
Vapor Space Accident: Turbine 2.8 40 Accident/3
bypass in manual control to maintain header pressure EK2.03-Knowledge of the interrelations
009/Small Break LOCA/3 X between the small break LOCA and the 3.0 41 following:
S/Gs EK3.06-Knowledge of the reasons for the 011/Large Break LOCA/3 X following responses as the apply to the 4.3 42 Large Break LOCA: Actuation of Phase A and B durina LOCA initiation
AA 1.16-Ability to operate and/or monitor 015/17/Reactor Coolant Pump the following as they apply to the Reactor 3.2 43 X Coolant Pump Malfunctions (Loss of RC Malfunctions/4 Flow): Low power reactor trip block status Iiahts AA2.04-Ability to determine and interpret 022/Loss of Reactor Coolant X the following as they apply to the Loss of 2.9 44 Makeup/2 Reactor Coolant Pump Makeup: How long PZR level can be maintained
within limits AK2.03-Knowledge of the interrelations
between the Loss of Residual Heat 025/Loss of Residual Heat X Removal System and the following:
2.7 45 Removal System/4 Service water or dosed cooling water Dumps AK1.03-Knowledge of the operational
027/Pressurizer
Pressure Control implications
of the following concepts as 2.6 46 X they apply to Pressurizer
Pressure Control System Malfunction/3 Malfunctions:
Latent heat of vaporization/condensation
029/Anticipated
Transient Without EK2.06-Knowledge of the interrelations
X between the and the following an ATWS: 2.9 47 Scram (ATWS)/1 Breakers, relays, and disconnects
EK3.02-Knowledge of the reasons for the 055/Station Blackout/6 X following responses as the apply to the 4.3 48 Station Blackout: Actions contained in EOP for loss of offsite and onsite power
ES-401 2 McGuire 2008 NRC Exam Written Examination
Outline Emergency and Abnormal Plant Evolutions
-Tier 1 Group 1 Form ES-401-2 i EAPE#/Name Safety Function[ill K2 I K3 I A1 I A2 I G I KIA Topic(s)AA 1.12-Ability to operate and/or monitor 056/Loss of Off-site Power/6 X the following as they apply to the Loss of 3.2 49 Offsite Power: Reactor building cooling unit AK3.01-Knowledge of the reasons for the 057/Loss of Vital AC Electrical
following responses as they apply to the Instrument
Bus/6 X Loss of Vital AC Instrument
Bus: Actions 4.1 50 contained in EOP for loss of vital ac electrical
instrument
bus AK1.01-Knowledge of the operational
058/Loss of DC Power/6 X implications
of the following concepts as 2.8 51 they apply to Loss of DC Power: Battery charger equipment and instrumentation
AA2.06-Ability to determine and interpret 062/Loss of Nuclear Service.the following as they apply to the Loss of Water/4 X Nuclear Service Water: The length of time 2.8 52 after the loss of CCW flow to a component before that component may be damaged AA2.08-Ability to determine and interpret 065/Loss of Instrument
Air/8 X the following as they apply to the Loss of 2.9 53 Instrument
Air: Failure modes of air-operated equipment EK1.2-Knowledge of the operational
implications
of the following concepts as E04/LOCA Outside Containment
/they apply to the (LOCA Outside 3 X Containment):
Normal, abnormal and 3.5 54 emergency operating procedures
associated
with (LOCA Outside Containment).
2.1.23-Conduct of Operations:
Ability to E11/Loss of Emergency Coolant X perform specific system and integrated
4.3 55 Recirculation/4 plant procedures
during all modes of plant operation.
E12/Uncontrolled
Depressurization
X 2.1.20-Conduct of Operations:
Ability to 4.6 56 of all Steam Generators/4 interpret and execute procedure steps.KIA Category Totals:333366 Group Point Total: I 18/6
(ES-401 3 McGuire 2008 NRC Exam Written Examination
Outline Emergency and Abnormal Plant Evolutions
-Tier 1 Group 2 Form ES-401-2 16 1
...S;.;a=fe...ty...F...u_n=ct;;,;io
....
K2 I K3 IA1 I A2 I G I KIA Topic(s)Imp.@[]I AA2.02-Ability to determine and interpret 003 1 Dropped Control Rod 11 X the following as they apply to the Dropped 2.8 82 Control Rod: Signal inputs to rod control system AA2.05-Ability to determine and interpret 033 1 Loss ofIntermediateRange
the following as they apply to the Loss of Nuclear Instrumentation17 X Intermediate
Range Nuclear 3.1 83 Instrumentation:
Nature of abnormality, from rapid survey of control room data 059 1 Accidental
Liquid RadWaste 2.2.38-Equipment Control: Knowledge of Release 19 X conditions
and limitations
in the facility 4.5 84 license.E06 1 Degraded Core Cooling 14 X 2.1.20-Conduet of Operations:
Ability to 4.6 85 interpret and execute procedure steps.AA 1.02-Ability to operate and 1 or monitor 005 Ilnoperable/Stuck
Control Rod 1 X the following as they apply to the 3.7 57 1 Inoperable
1 Stuck Control Rod: Rod selection switches AA2.03-Ability to determine and interpret the following as they apply to the 0241 Emergency Boration 11 X Emergency Boration: Correlation
between 2.9 58 boric acid controller
setpoint and boric acid flow 032 1 Loss of Source Range Nuclear AK1.01-Knowledge of the operational
X implications
of the following concepts as 2.5 59 Instrumentation17 Effects of voltage changes on performance
AA 1.02-Ability to operate and 1 or monitor 0331 Loss ofIntermediateRange
X the following as they apply to the Loss of 3.0 60 Nuclear Instrumentation
17 Intermediate
Range Nuclear Instrumentation:
Level trip bypass AK2.01-Knowledge of the interrelations
0361 Fuel Handling Incidents18 X between the Fuel Handling Incidents and 2.9 61 the followina:
Fuel handlina equipment AA2.12-Ability to determine and interpret 067 1 Plant Fire On-site18 X the following as they apply to the Plant 2.9 62 Fire on Site: Location of vital equipment within fire zone AK3.01-Knowledge of the reasons for the 0691 Loss of Containment
Integrity 1 following responses as they apply to the X Loss of Containment
Integrity:
Guidance 3.8 63 5 contained in EOP for loss of containment
intearitv AK2.01-Knowledge of the interrelations
0761 High Reactor Coolant Activity 1 X between the High Reactor Coolant Activity 2.6 64 9 and the following:
Process radiation monitors 2.4.8-Emergency Procedures
1 Plan: E02 1 SI Termination
13 X Knowledge of how abnormal operating 3.8 65 procedures
are used in conjunction
with EOP's.KIA Category Totals:12124 3 Group Point Total: I 9/4
(ES-401 4 McGuire 2008 NRC Exam Written Examination
Outline Plant Systems-Tier 2 Group 1 Form ES-401-2 System#/NameKKKKKKAAAA
G Imp.Q1234561234
- A2.01-Ability to (a)predict the impacts of the following malfunctions
or operations
on the PZR PCS;and (b)010 Pressurizer
Pressure X based on those predictions, use 3.6 86 Control procedures
to correct, control, or mitigate the consequences
of those malfunctions
or operations:
Heater failures A2.03-Ability to (a)predict the impacts of the following malfunctions
or operations
on the CSS;and (b)based 026 Containment
Spray X on those predictions, use procedures
to 4.4 87 correct, control, or mitigate the consequences
of those malfunctions
or operations:
Failure of ESF 2.4.4-Emergency Procedures
/Plan: Ability to recognize abnormal 061 Auxiliary/Emergency
X indications
for system operating 4.7 88 Feedwater parameters
which are entry-level
conditions
for emergency and abnormal operating procedures.
073 Process Radiation 2.2.22-Equipment Control: KnoVv1edge
Monitoring
X of limiting conditions
for operations
and 4.7 89 safety limits A2.02-Ability to (a)predict the impacts of the following malfunctions
or operations
on the SWS;and (b)based 076 Service Water X on those predictions, use procedures
to 3.1 90 correct, control, or mitigate the consequences
of those malfunctions
or operations:
Service water header pressure K1.08-KnoVv1edge
of the physical connections
and/or cause-effect
003 Reactor Coolant Pump X relationships
between the RCPS and 2.7 1 the following systems: Containment
isolation K5.46-KnoVv1edge
of the operational
implications
of the following concepts 004 Chemical and Volume X as they apply to the CVCS: Reason for 2.5 2 Control going solid in PZR (collapsing
steam bubble): make sure no steam is in PRT when PORV is opened to drain RCS K2.03-KnoVv1edge
of bus power 005 Residual Heat Removal X supplies to the following:
RCS pressure 2.7 3 boundary motor-operated
valves A4.01-Ability to manually operate 006 Emergency Core Cooling X and/or monitor in the control room: 4.1 4 Pumps A3.01-Ability to monitor automatic 007 Pressurizer
Relief/Quench
X operation of the PRTS, including:
2.7 5 Tank Components
which discharge to the PRT K2.02-Knowledge of bus power 008 Component Cooling Water X supplies to the following:
CCW pump, 3.0 6 including emergency backup K4.01-KnoVv1edge
of CCWS design 008 Component Cooling Water X feature(s)
and/or interlock(s)
which 3.1 7 provide for the following:
Automatic start of standby pump
(ES-401 4 McGuire 2008 NRC Exam Written Examination
Outline Plant Systems-Tier 2 Group 1 Form ES-401-2 System#/NameKKKKKKAAAA G Imp.Q 12345612 3 4#K1.07-Knowledge of the physical 010 Pressurizer
Pressure connections
and/or cause-effect
Control X relationships
between the PZR PCS 2.9 8 and the following systems: Containment
A3.03-Ability to monitor automatic 012 Reactor Protection
X operation of the RPS, including:
Power 3.4 9 supply K6.03-Knowledge of the effect of a 012 Reactor Protection
X loss or malfunction
of the following will 3.3 10 have on the RPS: Trip logic circuits 013 Engineered
Safety K4.08-Knowledge of ESFAS design Features Actuation X feature(s)
and/or interlock(s)
which 3.1 11 provide for the following Redundancy
A2.05-Ability to (a)predict the impacts of the following malfunctions
or operations
on the CCS;and (b)based 022 Containment
Cooling X on those predictions, use procedures
to 3.1 12 correct, control, or mitigate the consequences
of those malfunctions
or operations:
Major leak in CCS K6.01-Knowledge of the effect of a loss or malfunction
of the following will 025 Ice Condenser X have on the ice condenser system: 3.4 13 Upper and lower doors of the ice condenser K5 01-Knowledge of operational
implications
of the following concepts 025 Ice Condenser X as they apply to the ice condenser 3.0 14 system: Relationships
between pressure and temperature
2.2.22-Equipment Control: Knowledge 026 Containment
Spray X of limiting conditions
for operations
and 4.0 15 safety limits.K3.02-Knowledge of the effect that a 026 Containment
Spray X loss or malfunction
of the CSS will have 4.2 16 on the following:
Recirculation
spray system A4.04-Ability to manually operate 039 Main and Reheat Steam X and/or monitor in the control room: 3.8 17 Emergency feedwater pump turbines K4.07-Knowledge of MRSS design 039 Main and Reheat Steam X feature(s)
and/or interlock(s)
which 3.4 18 provide for the following:
Reactor building isolation A 1.03-Ability to predict and/or monitor changes in parameters (to prevent 059 Main Feedwater X exceeding design limits)associated
2.7 19 with operating the MFW controls including:
Power level restrictions
for operation of MFW pumps and valves.061 AUXiliary/Emergency
K3.01-Knowledge of the effect that a X loss or malfunction
of the AFW will 4.4 20 Feedwater have on the following:
RCS A3.05-Ability to monitor automatic 062 AC Electrical
Distribution
X operation of the ac distribution
system, 3.5 21 including:
Safety-related
indicators
and controls A 1.01-Ability to predict and/or monitor changes in parameters
associated
with 063 DC Electrical
Distribution
X operating the dc electrical
system 2.5 22 controls inclUding:
Battery capacity as it is affected bY discharge rate
ES-401 4 McGuire 2008 NRC Exam Written Examination
Outline Plant Systems-Tier 2 Group 1 Form ES-401-2 System#/NameKKKKKKAAAA
G Imp.Q 1 23456 1 2 3 4#----f--K6.08-Knowledge of the effect of a064Emergency
Diesel X loss or malfunction
of the following will 3.2 23 Generator have on the ED/G system: Fuel oil storage tanks A2.01-Ability to (a)predict the impacts of the following malfunctions
or operations
on the PRM system;and (b)073 Process Radiation X based on those predictions, use 2.5 24 Monitoring
procedures
to correct, control, or mitigate the consequences
of those malfunctions
or operations:
Erratic or failed power supply 2.4.30-Emergency Procedures
/Plan;Knowledge of events related to system 076 Service Water X operation/status that must be reported 2.7 25 to internal organizations
or external agencies, such as the state, the NRC, or the transmission
system operator.K3.07-Knowledge of the effect that a 076 Service Water X loss or malfunction
of the SWS will 3.7 26 have on the foliowinQ:
ESF loads 2.4.35-Emergency Procedures
/Plan: 078 Instrument
Air X Knowledge of local auxiliary operator 3.8 27 tasks during emergency and the resultant operational
effects.A4.01-Ability to manually operate and/or monitor in the control room: 103 Containment
X Flow control, pressure control, and 3.2 28 temperature
control valves, including pneumatic valve controller
KIA Category Totals:22332325335
Group Point Total: I 28/5
ES-401 5 McGuire 2008 NRC Exam Written Examination
Outline Plant Systems-Tier 2 Group 2 Form ES-401-2 System#/NameKKKKKKAAAA G Imp.Q12 3 4561234#2.1.7-Conduct of Operations:
Ability to evaluate plant performance
and make 015 Nuclear Instrumentation
X operational
judgments based on 4.7 91 operating characteristics, reactor behavior, and instrument
interpretation.
A2.03-Ability to (a)predict the impacts of the following malfunctions
or operations
on the HRPS;and (b)based on those predictions, use 028 Hydrogen Recombiner
and Procedures
to correct, control, or Purge Control X mitigate the consequences
of those 4.0 92 malfunctions
or operations:
The hydrogen air concentration
in excess of limit flame propagation
or detonation
with resulting equipment dam-age in containment
A2.01-Ability to (a)predict the impacts of the following malfunctions
or operations
on the SAS;and (b)based 079 Station Air X on those predictions, use Procedures
3.2 93 to correct, control, or mitigate the consequences
of those malfunctions
or operations::
Cross-connection
with lAS K4.03-Knowledge of RPIS design 014 Rod Position Indication
X feature(s)
and/or interlock(s)
which 3.2 29 provide for the following:
Rod Bottom Iiqhts K6.01-Knowledge of the effect of a 017 In-core Temperature
X loss or malfunction
of the following ITM 2.7 30 Monitor system components:
Sensors and detectors 015 Nuclear Instrumentation
K2.01-Knowledge of bus power System X supplies to the following:
NIS channels, 3.3 31 components, and interconnections
K1.01-Knowledge of the physical connections
and/or cause-effect
028 Hydrogen Recombiner
and X relationships
between the HRPS and 2.5 32 Purge Control the following systems: Containment
annulus ventilation
system (including
pressure limits)A4.04-Ability to manually operate 029 Containment
Purge X and/or monitor in the control room: 3.5 33 Containment
Evacuation
siqnal K3.01-Knowledge of the effect that a 033 Spent Fuel Pool Cooling X loss or malfunction
of the Spent Fuel 2.6 34 Pool Cooling System will have on the following:
Area ventilation
systems A2.05-Ability to (a)predict the impacts of the following malfunctions
or operations
on the S/GS;and (b)based 035 Steam Generator X on those predictions, use procedures
3.2 35 to correct, control, or mitigate the consequences
of those malfunctions
or operations:
Unbalanced
flows to the S/Gs 041 Steam DumplTurbine
A3.03-Ability to monitor automatic Bypass Control X operation of the 5DS, including:
Steam 2.7 36 flow
(ES-401 5 McGuire 2008 NRC Exam Written Examination
Outline Plant Systems-Tier 2 Group 2 Form ES-401-2 System#/Name K K KKKKAAAA G Imp.Q123 4 5 612 3 4#A 1.06-Ability to predict and/or monitor changes in parameters(to
prevent 071 Waste Gas Disposal X exceeding design limits)associated
2.5 37 with Waste Gas Disposal System operating the controls including:
Ventilation
system K5.03-Knowledge of the operational
implication
of the following concepts as 086 Fire Protection
X they apply to the Fire Protection
3.1 38 System: Effect of water spray on electrical
components
KIA Category Totals:1111111 3 1 1 1 Group PointTotal:
I 12/3
- ES-401 Generic Knowledge and Abilities Outline (Tier3)Form ES-401-3 ((Facility: McGuire 2008 NRC Exam Date: 5/12/2008 Category KIA#Topic RO SRO-Only IR Q#IR Q#2.1.35 Knowledge of the fuel-handling
responsibilities
of 3.9 94 SRO's.2.1.18 Ability to make accurate, clear and concise logs, 3.6 66 1.records, status boards, and reports.Conduct 2.1.13 Knowledge of facility requirements
for controlling
2.5 67 of Operations
vital/controlled
access.2.1.8 Ability to coordinate
personnel activities
outside 3.4 68 the control room.Subtotal 3 1 2.2.7 Knowledge of the process for conducting
special 3.6 95 or infrequent
tests.2.2.22 Knowledge of limiting conditions
for operations
4.7 96 and safety limits.2.Ability to apply technical specifications
for a Equipment 2.2.40 system.3.4 69 Control 2.2.13 Knowledge of tagging and clearance procedures.
4.1 70 2.2.6 Knowledge of the process for making changes to 3.0 71 procedures.
Subtotal 3 2 Knowledge of radiation or containment
hazards 2.3.14 that may arise during normal, abnormal, or 3.8 97 emergency conditions
or activities.
2.3.6 Ability to approve release permits 3.8 98 3.Radiation 2.3.4 Knowledge of radiation exposure limits under 3.2 72 Control normal or emergency conditions.
2.3.11 Ability to control radiation releases.3.8 73 Subtotal 2 2 2.4.46 Ability to verify that the alarms are consistent
4.2 99 with the plant conditions.
2.4.8 Knowledge of how abnormal operating 4.5 100 4.procedures
are used in coniunction
with EOP's.Emergency Procedures
/2.4.17 Knowledge of EOP terms and definitions.
3.9 74 Plan 2.4.14 Knowledge of general guidelines
for EOP usage.3.8 75 Subtotal 2 2 Tier 3 Point Total 10 7
ES-401 Record of Rejected KIA's Form ES-401-4 ((Tier/Group Randomly Selected Reason for Rejection KJA 2/2 015 G2.1.34 Q 91 Generic topic selected provided no relationship
with system selected.Randomly reselected
G2.1.7 Q 87 Procedures
have no relationship
to, and no 2/1 026 A2.01 guidance for, phenomenon
related to topic.Randomly reselected
A2.03 1/1 015 AA1.04 Q 43 Facility does not have selected component or indication.
Randomly reselected
AA1.16 Q 33 Facility cannot operate or monitor purge flow rate 2/2 029 A4.01 from control room;can only start or stop fans.Randomly reselected
A4.04 Q 31 Facility does not have system or fans provided 2/2 027 K2.01 specifically
for iodine removal.Kept K2 category and randomly reselected
system 015 2/2 014 K4.02 Q 29 Lower Electrical
Limit is CE, not applicable
to WEC design.Randomly reselected
K4.03 2/1 012 K6.07 Q 10 Facility does not have Core Protection
Calculators, CE design.Randomly reselected
K6.03 2/1 008 K4.07 Q 7 Facility does not have swing pump breaker.Randomly reselected
K4.01 Q 79 Excessive topic overlap with RO examination
1/1 055 EA2.05 related to DC distribution.
References
did not support an SRO level test item without excessive overlap.Randomly reselected
EA2.03 from 055 topic area.Q 81 Facility and generic references
did not support any test item directly related to KA topic.Topic selected 1/1 058 G2.4.21 could not be tested at SRO level in either closed or open reference format.Randomly reselected
Generic topic 2.2.37 from required Tier 1 and Tier 2 generic topics for 058 topic area.Q 89 Facility has no difference
in system between units that could be developed into a test item, either at RO or 2/1 073 G2.2.4 SRO level.Randomly reselected
generic 2.2.22 from required Tier 1 and Tier 2 generic topics for 073 topic area
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#1----Group#LJ 1 KIA#ll fT bo9 G2.1.27 Importance
Rating 4.0----Conduct of Operations:
Knowledge of system purpose and/or function.Proposed Question: SRO 76 Given the following conditions:*A reactor trip has occurred.*Safety Injection is actuated.All equipment has actuated as designed.*The crew is performing
EP/1/A/5000/E-0, Reactor Trip or Safety Injection.
- NC System pressure is 1700 psig and lowering slowly.*Pressurizer
level is off-scale low.*Containment
pressure is 1.7 psig and rising slowly.*FWST level is 300 inches and dropping at 2 inches per minute.*SG pressures are 1050 psig and stable.*CA flow is 600 gpm.*The operators are performing
E-1, Loss of Reactor or Secondary Coolant.Which ONE (1)of the following describes the procedure that will be usedto mitigate the event in progress, and the technical specification
basis of FWST parameters
for this event?A.ES-1.2, Post LOCA Cooldown and depressurization;
FWST minimum volume ensures a sufficient
volume of water in the containment
sump after ECCS injection to initiate Cold Leg Recirculation.
B.ES-1.3, Transfer to Cold Leg Recirculation;
FWST minimum volume ensures a sufficient
volume of water in the containment
sump after ECCS injection to initiate Cold Leg Recirculation.
C.ES-1.2, Post LOCA Cooldown and depressurization;
FWST minimum volume ensures that post LOCA core cooling requirements
are met for the ECCS injection phase even with an anticipated
lossofCold Leg Recirculation.
Page 187 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 D.ES-1.3, Transfer to Cold Leg Recirculation;
FWST minimum volume ensures that post LOCA core cooling requirements
are met for the ECCS injection phase even with an anticipated
loss of Cold Leg Recirculation.
Proposed Answer: A Explanation (Optional):
A.Correct.Correct Procedure and FWST basis.ES-1.2 will be entered because the rate of change on FWST level will result in conditions
NOT being met for ES-1.3 for another 40-60 minutes.ES-1.2 transition
will come significantly
sooner B.Incorrect.
Procedure is incorrect because ES-1.3 will not be performed next, it will take too long to reach conditions
c.Incorrect.
Basis is incorrect, because a loss of cold leg recirculation
is beyonddesignbasis
for FWST operability.
D.Incorrect.
Basis and procedure are incorrect, as described in Band C above Technical Reference(s)E-1, Rev 11;ES-1.2 Rev 11;ES-1.3 Rev 23 TS 3.5.4 basis Rev 70 EP-E1 p11, 15,59 Rev 17;FH-FW p 23, 67 Rev 40 Proposed references
to be provided to applicants
during None examination:
Learning Objective:
FH-FW Obj 5;EP-E1 Obj 2 (As available)(Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Memory or Fundamental
Knowledge Page 188 of 260 Draft 7
ES-401 Level: Sample Written Examination
Question Worksheet Comprehension
or Analysis x Form ES-401-510 CFR Part 55 55.41 Content: 55.43 5 Comments: KA matched because item evaluates function of FWST (RWST)SRO level because the applicant must determine procedure entry based on plant conditiuons
and also know TS basis for operability
of FWST Page 189 of 260 Draft 7
MNS EP/1/A/5000/E-1
UNITl LOSS OF REACTOR OR SECONDARY COOLANT PAGE NO.15 of 22 Rev.11 ACTION/EXPECTED
RESPONSE 14.Check if NC System cooldown and depressurization
is required: a.NC pressure-GREATER THAN 286 PSIG.RESPONSE NOT OBTAINED a.Perform the following:
_1)IF containment
pressure has remained less than 3 PSIG, THEN GO TO EP/1/A/5000/ES-1.2 (Post LOCA Cooldown And Dep ressu rization).
_2)IF ND flow to cold legs greater than 500 GPM, THEN GO TO Step 15.b.GO TO EP/1/A/5000/ES-1.2 (Post LOCA Cooldown And Depressu rization).
15.Check transfer to Cold Leg Recirc criteria: a.FWST level-LESS THAN 180 INCHES C1FWST LEVEL LOll ALARM).b.Check S/I systems-ALIGNED FOR COLD LEG RECIRC.a.RETURN TO Step 13.b.GO TO EP/1/A/5000/ES-1.3 (Transfer To Cold Leg Recirc).
DUKE POWER 2.0 PROCEDURE SERIES BACKGROUND
2m 1.E-1, Loss of Reactor or Secondary Coolant 2m 1 m 1 Loss of Reactor Coolant MCGUIRE OPERATIONS
TRAINING In order to describe the various phenomena that can occur during a LOCA, it is convenient
to define five categories
of accidents based on the size of the break and number of SII trains.This section describes four break sizes and Safeguard equipment status as follows: 1.Breaks between 3/8 11 (::::0.1 in 2)and 1 11 (::::0.8 in 2)diameter with minimum safety injection.
NC pressure will stabilize above steam generator pressure.2.Breaks between 3/8 11 (::::0.1 in 2)and 1 11 (::::0.8 in 2)diameter with maximum safety injection.
The NC will repressurize.
3.Breaks between 1 11 (::::0.8 in 2)and 13.5 11 (::::1 ft 2)diameter.NC pressure goes below steam generator pressure.4.Breaks greater than 1 ft 2*The NC will rapidly depressurize
to close to the containment
atmospheric
pressure.Breaks smaller than 3/8 11 (::::0.1 in 2)with normal charging are considered
to be leaks rather than small LOCAs since NC pressure and pzr level do not go down.If charging flow is not available, the transient would be similar to the response described below for small LOCAs.SMALL LOCAs The flowpath through the E-1 series is dependent upon the break size, the break location, and operator/Station
Management
decisions.
For a break size of up to 1 inch diameter, the amount of S/I flow determines
the flow path in the E-1 series.If minimum S/I flow is assumed, the E-1 S/I-termination
criteria would not be met, repressurization
of the reactor coolant may not occur, and S/I flow equals the break flow.This constitutes
a safe and stable condition for the long term provided the heat sink is maintained.
As long as S/I and Auxiliary Feedwater are available, the reactor will reach equilibrium
conditions
for the steam generator pressures.
Long-term cooling may require depressurizing
to cold shutdown while stepping down S/I flow, so ES-1.2, Post LOCA Cooldown and Depressurization
would be used.If maximum SII flow is assumed such that S/I flow is greater than break flow, the reactor will rapidly repressurize, and may in fact end up with the pressurizer
filled solid.At this point, the NC system will rapidly repressurize
and the S/I termination
criteria will be met, and S/I may be terminated
using ES-1.1, S/I Termination.
However, if 8/1 is not terminated, or more realistically, if S/I termination
is delayed, the core will remain cooled and in a safe and long term stable condition.
The NC system will remain in an
although possibly not desirable, condition.
OP-MC-EP-E1
FOR TRAINING PURPOSES ONL Y Page 11 of 427 REV.17
DUKE POWER MCGUIRE OPERATIONS
TRAINING 2.2.ES-1.1, Safety Injection Termination
S/I TERMINATION
is entered based on the following criteria: 1.The NC is subcooled, 2.An adequate secondary heat sink exists, 3.NC pressure is either stable or going up, and 4.Pressurizer
level is indicating
greater than 11%(29%ACC).These conditions
combined indicate that the NC is in a safe state with adequate core cooling and that S/I flow can be reduced without jeopardizing
the safety of the plant.S/I pumps are stopped in a prescribed
sequence as long as control is maintained, until makeup is only from normal charging pump lineup.Appropriate
transitions
are provided in case all S/I pumps cannot be stopped or must be restarted.
If S/I pumps are stopped and control is maintained, then the plant configuration
is essentially
realigned to aS/I condition at no-load or some lower stable temperature.
2.3.ES-1.2, Post LOCA Cooldown and Depressurization
For a LOCA, plant design is to use makeup water from the FWST until it is drained.Recirculation
from the containment
sump to the NC is then used for long-term cooling and makeup.The time to switch over to recirculation
depends on the size of the break, the FWST water volume, and whether containment
spray is initiated.
For some smaller breaks, it is possible to cool down and depressurize
the NC to a cold shutdown condition before the FWST is drained.When doing this, it is important to maintain adequate core cooling by maintaining
inventory while also trying to minimize FWST depletion.
For any loss of NC inventory, the NC pressure will be dependent on the size of the break, the NC fluid shrink due to cooldown, and the S/I flow rate.For smaller breaks the NC pressure will remain stabilized
for a long period of time at high NC pressures (greater than 400 psig).For these breaks, transfer to cold leg recirculation
may be necessary while NC pressure remains high.Procedure ES-1.2 provides actions to reduce the NC temperature
and pressure to or below 200°F and 400 psig.This is done by establishing
a S/G cooldown and selectively
reducing S/I flow by stopping S/I pumps or establishing
normal charging flow if minimum subcooling
and pzr level can be established.
From there, the plant staff can determine how to completely
depressurize
the plant to stop NC inventory loss and effect repairs.OP-MC-EP-E1
FOR TRAINING PURPOSES ONL Y Page 15 of 427 REV.17
(DUKE POWER E-1 Loss of Reactor or Secondary Coolant 3.6.Final Plant Status MCGUIRE OPERATIONS
TRAINING (E-1 provides the actions to recover from a loss of reactor or secondary coolant.The following table summarizes
the exit guidance from E-1.The left column lists each step that provides a potential exit point from E-1.The right column lists the transition
procedure(s).
If an exit transition
is necessary, the operator should transition
to Step 1 unless otherwise directed.Other transitions
may be made as a result of the Foldout Page directives.
These are summarized
in the following table.(Cold Leg Recirc Switchover
Criteria OP-MC-EP-E1
ES-1.3, Transfer To Cold Le Recirculation
FOR TRAINING PURPOSES ONL Y Page 59 of 427 REV.17
MNS EP/1/A/5000/ES-1.2
UNITl POST LOCA COOLDOWN AND DEPRESSURIZATION
PAGE NO.1 of 55 Rev.11 A.Purpose This procedure provides actions to cool down and depressurize
the NC System to Cold Shutdown conditions
following a loss of reactor coolant inventory.
B.Symptoms or Entry Conditions
This procedure is entered from:*EP/1/A/5000/E-O (Reactor Trip Or Safety Injection), Step 26, when NC System pressure goes down after stopping all but one NV pump.*EP/1/A/5000/E-O (Reactor Trip Or Safety Injection), Step 29, when pzr level can not be maintained
using normal charging.*EP/1/A/5000/E-1 (Loss Of Reactor Or Secondary Coolant), Step 14, if symptoms of a small break LOCA exist.*EP/1/A/5000/ES-1.1 (Safety Injection Termination), Step 6, when NC System pressure goes down after stopping all but one NV pump.*EP/1/A/5000/ES-1.1 (Safety Injection Termination), Step 9, when pzr level can not be maintained
using normal charging.*EP/1/A/5000/ES-1.1 (Safety Injection Termination), Step 10, when NC System pressure is less than shutoff head pressure of the NI pumps.
DUKE POWER McGUIRE OPERATIONS
TRAINING therefore are not susceptible
to reference leg problems like losing level due to small leaks or evaporation.
Problem over the years with FWST level instrumentation
has initiated design studies and level instrumentation
redesign to make the level indication
more reliable.Site Plan MG-97-0035 addresses the FWST Level Instrumentation
Improvement
Project.It addresses potential common mode failures such as submergence, impulse line freezing and reference line blockage.As a result, modifications
are in progress under12496 and 22496.{If a transmitter
problem occurs, the operators would first notice it on cross-channel
comparison, performed periodically (SOER 97-01 Review), unless it were a gross problem which would cause a level alarm from one of the transmitters.}
FWST Pressure Atmospheric
pressure exists in the FWST since it is normally vented to atmosphere, and therefore pressure is not monitored in the FWST.FWST Level Four channels provide Control Room level indication
alarms and protection
logic utilized in Normal Operation and ECCS/NS pumps switch-over
from the FWST to the Containment
Sump following a LOCA.There are three Safety Related Level Channels required by Tech Specs under ESFAS Instrumentation
and twoSafety Related level instruments
used to monitor FWST Level during normal operations.
Each of the three Safety Related level instruments (FWP5000 Channel 4, FWP5010 Channel 1, and FWP5020 Channel 2)has a completely
separate reference and variable leg tap, and are located 120 degrees in circumference
from each other.Therefore a single failure will not affect more than one channel.Their range is from0-500" WC.NOTE: The setpoints at which the Low Level Auto-Switch-over
to Cold Leg Recirculation
is 180" H 2 0.The setpoint at which the Control Room Crew will manually swap Containment
Spray Pump Suction to the Containment
Sump is 33" H 2 0.TheSafety Related Upper Narrow Range Level Instrument
has a range of 405"-530" WC.OP-MC-FH-FW
FOR TRAINING PURPOSES ONLY Page 67 of 113 REV.40
RWST B 3.5.4 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)B 3.5.4 Refueling Water Storage Tank (RWST)BASES BACKGROUND
The RWST supplies borated water to the Chemical and Volume Control System (CVCS)during abnormal operating conditions, to the refueling pool during refueling and makeup operations, and to the ECCS and the Containment
Spray System during accident conditions.
The RWST supplies both trains of the ECCS and the Containment
Spray System through separate supply headers during the injection phase of a loss of coolant accident (LOCA)recovery.A motor operated isolation valve is provided in each header to isolate the RWST once the system has been transferred
to the recirculation
mode.The recirculation
mode is entered when pump suction is transferred
to the containment
sump following receipt of the RWST-Low Level signal.Use of a single RWSTtosupply both trains of the ECCS and Containment
Spray System is acceptable
since the RWST is a passive component, and since injection phase passive failures are not required to be assumed to occur coincidentally
with Design Basis Events.The switchover
from normal operation to the injection phase of ECCS operation requires changing centrifugal
charging pump suction from the CVCS volume control tank (VCT)to the RWST through the use of isolation valves.During normal operation in MODES1, 2, and 3, the safety injection (SI)and residual heat removal (RHR)pumps are aligned to take suction from the RWST.The ECCS pumps are provided with recirculation
lines that ensure each pump can maintain minimum flow requirements
when operating at or near shutoff head conditions.
When the suction for the ECCS and Containment
Spray System pumps is transferred
to the containment
sump, the RWST flow paths must be isolated to prevent a release of the containment
sumpcontentsto
the RWST, which could result in a release of contaminants
to the atmosphere
and the eventual loss of suction head for the ECCS pumps.This LCO ensures that: a.The RWST contains sufficient
borated water to support the ECCS during the injection phase;McGuire Units 1 and 2 B 3.5.4-1 Revision No.70
RWST B 3.5.4 BASES BACKGROUND (continued)
b.Sufficient
water volume exists in the containment
sump to support continued operation of the ECCS and Containment
Spray System pumps at the time of transfer to the recirculation
mode of cooling;and c.The reactor remains subcritical
following a LOCA.Insufficient
water in the RWST could result in insufficient
cooling capacity when the transfer to the recirculation
mode occurs.Improper boron concentrations
could result in a reduction of SDM or excessive boric acid precipitation
in the core following the LOCA, as well as excessive caustic stress corrosion of mechanical
components
and systems inside the containment.
APPLICABLE
During accident conditions, the RWST provides a source of borated SAFETY ANALYSES water to the ECCS and Containment
Spray System pumps.As such, it provides containment
cooling and depressurization, core cooling, and replacement
inventory and is a source of negative reactivity
for reactor shutdown (Ref.1).The design basis transients
and applicable
safety analyses concerning
each of these systems are discussed in the Applicable
Safety Analyses section of B 3.5.2, II ECCS-Operatingll;
B 3.5.3, IIECCS-Shutdown
ll;and B 3.6.6, IIContainment
Spray Systems.1I These analyses are used to assess changes to the RWST in order to evaluate their effects in relation to the acceptance
limits in the analyses.The RWST must also meet volume, boron concentration, and temperature
requirements
for non-LOCA events.The volume is not an explicit assumption
in non-LOCA events since the required volume is a small fractionofthe available volume.The deliverable
volume limit is set by the LOCA and containment
analyses.For the RWST, the deliverable
volume is different from the total volume contained due to the location of the piping connection.
The ECCS water boron concentration
is an explicit assumption
in the main steam line break (MSLB)analysis to ensure the required shutdown capability.
This assumption
is important in ensuring the required shutdown capability.
Although the maximum temperature
is a conservative
assumption
in the feedwater line break analysis, SI termination
occurs very quickly in this analysis and long before significant
RCS heatup occurs.The minimum temperature
is an assumption
in the MSLB actuation analyses.For a large break LOCA analysis, the RWST level setpoint equivalent
to the minimum water volume limit of 372,100 gallons and the lower boron concentration
limits listed in the COLR are used to compute the post McGuire Units 1 and 2 B 3.5.4-2 Revision No.70
RWST B 3.5.4 BASES ACTIONS (continued)
restore the RWST to OPERABLE status is based on this condition simultaneously
affecting redundant trains.C.1 and C.2 If the RWST cannot be returned to OPERABLE status within the associated
Completion
Time, the plant must be brought to a MODE in which the LCO does not apply.To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.The allowed Completion
Times are reasonable, based on operating experience, to reach the required plant conditions
from full power conditions
in an orderly manner and without challenging
plant systems.SURVEILLANCE
SR 3.5.4.1 REQUIREMENTS
The RWST borated water temperature
should be verified every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to be within the limits assumed in the accident analyses band.This Frequency is sufficient
to identify a temperature
change that would approach either limit and has been shown to be acceptable
through operating experience.
SR 3.5.4.2 The RWST water volume should be verified every 7 days to be above the required minimum level in order to ensure that a sufficient
initial supply is available for injection and to support continued ECCS and Containment
Spray System pump operation on recirculation.
Since the RWST volume is normally stable and is protected by an alarm,a7 day Frequency is appropriate
and has been shown to be acceptable
through operating experience.
SR 3.5.4.3 The boron concentrationofthe RWST should be verified every 7 days to be within the required limits.This SR ensures that the reactor will remain subcritical
following a LOCA and that the boron content assumed for the injection water in the MSLB analysis is available.
Further, it assures that the resulting sump pH will be maintained
in an acceptable
range so that boron precipitation
in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical
systems and components
will be minimized.
Since the RWST volume is normally stable,a7 day sampling Frequency to verify boron concentration
is appropriate
and has been shown to be acceptable
through operating experience.
McGuire Units 1 and 2 B 3.5.4-5 Revision No.70
DUKE POWER McGUIRE OPERATIONS
TRAINING IObjective
6 I 2.6 Refueling Water Storage Tank Design Basis The FWST has (as a minimum)a usable capacity of 372,100 gallons of borated water.The water in the FWST is electrically
heated when water temperature
decreases to<75°F.The tank capacity provides an adequate amount of borated water to insure:*A sufficient
volume of borated refueling water needed to increase the boron concentration
of the initially spilled water to a point that assures no return to criticality
with the reactor at cold shutdown and all control rods fully inserted in the core with the exception of the most reactive rod cluster control assembly.*A sufficient
volume to refill the reactor vessel above the nozzles after a LOCA.*A sufficient
volume of water in the lower compartment
of the containment
following ECCS Injection to permittheinitiation
of Cold and Hot Leg Recirculation.*A sufficient
volume of borated water to insure that the radiation dose at the surface of the refueling cavity is limited to 2.5 milli rem per hour during the period when a fuel assembly is transferred
over the reactor vessel flange.The FWST is surrounded
by a seismic wall.The basis of the seismic wall is that in the event a Tornado induced missile ruptures the FWST, the wall is high enough to retain a sufficient
volume of FWST water to provide NPSH to the Centrifugal
Charging Pumps and the Safety Injection Pumps.The Missile induced rupture assumes that there is a Main Steamline Break in conjunction
with an FWST rupture.There is no concern for the ND Pumps because it is assumed that the Steam Break Outside Containment
Event will not cause primary pressure to be reduced below the Shut-off Head of the pumps.The FWST overflows to the Spent Fuel"Pool and to the FWST trench.The following parameters
are associated
with the FWST:*Minimum Volume modes 1-4*Minimum Volume modes 5-6*Minimum Boron Concentration
- Minimum Temperature
- Maximum Temperature
372,100 gallons Cycle Dependent (See COLR)Cycle Dependent (See COLR)OP-MC-FH-FW
FOR TRAINING PURPOSES ONLY Page 23 of 113 REV.40
MNS EP/1/A/5000/ES-1.3
UNITl TRANSFER TO COLD LEG RECIRC PAGE NO.1 of 43 Rev.23 A.Purpose This procedure provides the necessary instructions
for transferring
the Safety Injection System and Containment
Spray System to the recirculation
mode.B.Symptoms or Entry Conditions
This procedure is entered from:*EP/1/A/5000/E-1 (Loss Of Reactor Or Secondary Coolant), Step 15, on low FWST level.*EP/1/A/5000/ECA-2.1 (Uncontrolled
Depressurization
Of All Steam Generators), Step 12, on low FWST level.*Other procedures
whenever FWST level reaches the switchover
setpoint.
DUKE POWER MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours}NLO NLOR LPRO LPSO LORN/AN/A 5.0 5.0 4.0 OBJECTIVES
SNN LLL E OBJECTIVELLPP 0 0 0 R S R Q R 0 0 1 Explain the purpose for each procedure in the E-1 series.XX EPE1001 2 Discuss the entry and exit guidance for each procedure in theXX E-1 series.EPE1002 3 Discuss the mitigating
strategy (major actions)of each XXX procedure in the E-1 series.EPE1003 4 Discuss the basis for any note, caution or step for each XXX procedure in the E-1 series.EPE1004 5 Given the Foldout page discuss the actions included and the XXX basis for these actions.EPE1005 6 Given the appropriate
procedure, evaluate a given scenario X X X describing
accident events and plant conditions
to determine any required action and its basis.EPE1006 7 Discuss the time critical task(s)associated
with the E-1 series XXX procedures
including the time requirements
and the basis for these requirements.
EPE1007 , OP-MC-EP-E1
FOR TRAINING PURPOSES ONL Y Page 5 of 427 REVe 17
DUKE POWER McGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours)OBJECTIVES
LOR 1.5 N NLLL OBJECTIVELLPP 000 R S R R00 1 State the purpose of the Refueling Water System.XXXX 2 Provided with the FW training drawing and OP/1 orXXXX 2/A/6200/14
and OP/1 or 2/A/6200/13, discuss the various lineups that can be utilized to transfer water, provide makeup, or purify the refueling water.3 Identify the valves/pumps/instrumentation
that can be XXX operated or monitored from the Control Room.4 Given a Limit and Precaution
associated
with the FW System,XXXXX discuss its basis and when it applies.5 Concerning
the Technical Specifications
related to the Refueling Water System;*Given the LeO title, state the LCO (including
any COLRXXX values)and applicability.
- For any LCO's that have action required within one hour, XXX state the action.*Given a set of parameter values or system conditions, XXX determine if any Tech Spec LCO's is(are)not met and any action(s)requiredwithin
one hour.*Given a set of plant parameters
or system conditions
and XXX the appropriate
Tech Specs, determine required action(s).
- Discuss the basis for a given Tech Spec LCO or Safety X*Limito*SRO Only OP-MC-FH-FW
FOR TRAINING PURPOSES ONLY Page 7 of 113 REV.40
I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#1----Group#r" fir uJ 1 KIA#L,tC)026 G2.1.25 Importance
Rating 4.2----Conduct of Operations:
Ability to interpret reference materials, such as graphs, curves, tables, etc.Proposed Question: SRO 77 Page 190 of 260 Draft 7
ES-401 Initial conditions:
Time=0 minutes Sample Written Examination
Question Worksheet Form ES-401-5*Unit 1 is at 100%power.*"A" Train KC pumps are running.*Operators have been dispatched
to initiate YM makeup to the KC Surge Tank.*"A" KC Surge Tank level is 6.5 ft.*"s" KC Surge Tank level is 6.5 ft.Current conditions:
Time=5 minutes*"A" KG Surge Tank level is 5.6 feet*"s" KC Surge Tank level is 6.4 feet.Which ONE (1)of the following describes the approximate
KC system net leak rate, and the action and procedure use required in AP/21 , Loss of KC or KC System Leakage?(Reference
Provided)A.50 GPM;Isolate KC Non-Essential
Headers in accordance
with Enclosure 2.S.50 GPM;Isolate"A" KC train from"S" KC train.C.100 GPM;Isolate KG Non-Essential
Headers in accordance
with Enclosure 2.D.100 GPM;Isolate"A" KG train from"S" KC train.Proposed Answer: 0 Explanation (Optional):
Page 191 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 D is correct per conditions.
Applicant must interpret curve and determine the leak rate indications
based on level decreases A incorrect because leak rate is wrong (1/2 of actual, as interpreted
by curve.)Also, action is incorrect, as procedure will direct splitting trains for indication
shown B incorrect because leak rate is incorrect.
Plausible because action is correct C incorrect because procedure use is incorrect.
Approximately
0.1 feet/minute, perform step 20 to split trains Technical Reference(s):
AP/21 Step 20 Rev 9 AP/21 Basis Document Rev 3 (Attach if not previously
provided)Proposed references
to be provided to applicants
during examination:
OP/1/A/61 00/22 Enclosure 4.3 Curve 7.31 None Learning Objective: (As available)
Question Source: Bank#Modified Bank X#New (Note changes or attach parent)Question History: Question Cognitive Level: Last NRC Exam Modified from 2007 NRC exam 78 Memory or Fundamental
Knowledge Comprehension
or Analysis X10 CFR Part 55 55.41 Content: 55.43 5 Comments: KA matched because use of a curve is required and interpretation
of that curve is required to determine KC (CCW)leak rate.SRO level because assessment
of Page 192 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 conditions
based on available indications, and selection of procedures (attachments)
is required Page 193 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 The following conditions
exist:*Unit 1 is in Mode 1, 100%power.*"A" and"B" Train KG pumps are running.*All available makeup has been established
to the KG surge tanks.*"A" KG Surge tank level is decreasing
0.04 ftlmin.*"B" KG Surge Tank level is decreasing
at 0.03 ftlmin.*"A" KG Surge Tank level is presently 3.2 ft.*"B" KG Surge Tank level is presently 3.4 ft.*NGP bearing temperatures
are approximately
180 degrees F and rising slowly.Which ONE (1)of the following describes the action and procedure use required?A.Enter AP/08, NG Pump Malfunctions, and trip NG Pumps.B.Trip the reactor;enter E-O, Reactor Trip or Safety Injection.
Trip NGPs and trip"A" KG Pumps.G.Enter AP/21 , Loss of KG or KG System Leakage, and Isolate KGEssential Headers in accordance
with Enclosure 2.D.Enter AP/21 , Loss of KG or KG System Leakage, and isolate"All KG train from IIB II KG train.Ans.0 Page 194 of 260 Draft 7
UN.r 1 OP/1/A.11022 ENCLOSURE 4.3 CURVE 7.31 COMPONENT COOLING SURGE TANK (VOLUME vs.COMPARTMENT
LEVEL)9 8 7 6 4 5 WATER COLUMN (FEET)3 2 I I I I8.5 Foot Level=3555.25 Gallons I./.....'tt',..",/, V./ 1/./'Total Tank Volume=7110.5 Gallons tI 0 Foot Level=72.0 Gallons(Both Compartments) 1/L/../'......o o 4000 500 1000 3000 35002500 o..J..J<<£!.2000 w::::J 5 1500>This data is also provided on the OAC.UNIT 1
MNS AP/1/A/5500/21
UNITl LOSS OF KC OR KC SYSTEM LEAKAGE PAGE NO.2 of 78 Rev.9 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED B.Symptoms*IILO KC HX A INLET FLOW II computer alarm*IILO KC HX B INLET FLOW II computer alarm*Low flow alarms on components
supplied by KC*High temperature
alarms on components
supplied by KC*Low level or level going down in KC Surge Tank*Abnormal KC pump Flow*IILO KC SURGE TANK COMPARTMENT
A LEVEL II computer alarm*IILO KC SURGE TANK COMPARTMENT
B LEVEL II computer alarm*IIKC SURGE TANK ABNORMAL LEVEL II alarm.C.Operator Actions1.Check any KC pump-ON.Perform the following:
a.Isolate:*Normal letdown*Excess letdown*ND letdown.b.Close all NM valves located on 1 MC-8 (vertical board).2.Monitor Foldout page.3.Secure any dilution in progress.4.Check ND-IN RHR MODE._GO TO Step 7.
MNS AP/1/A/5500/21
UNITl LOSS OF KC OR KC SYSTEM LEAKAGE Enclosure1-Page 1 of 1 Foldout PAGE NO.28 of 78 Rev.91.KC header isolation criteria:*IF KC surge tank level goes below 2 ft due to KC system leak, THEN immediately
isolate affected train PER Enclosure 2 (Isolation
of KC Non-essential
Headers).2.NC pump trip criteria:*IF NC pump motor bearing temperature
reaches 195°F, THEN perform the following:
a.Trip the reactor.b.WHEN reactor is tripped, THEN trip all NC pumps.c.GO TO EP/1/A/5000/E-O (Reactor Trip or Safety Injection), while continuing
in this procedure as time and conditions
allow.3.ND pump trip and flow isolation criteria (Applies if ND aligned for RHR):*IF KC cooling lost to either NO train's HX, AND NC temperature
is greater than 150°F, THEN perform the following on train of NO that lost KC flow to its NO HX: a.Stop associated
NO pump.b.IF1A NO HX lost KC flow, THEN close:*1 NO-33 (A NO Hx Bypass)*1 NO-32 (A NO Hx To Letdown Hx).c.IF1B NO HX lost KC flow, THEN close:*1 NO-18 (B NO Hx Bypass)*1 NO-17 (B NO Hx To Letdown Hx).d.IF both NO pumps off THEN REFER TO AP/1/A/5500/19 (Loss of NO or NO System Leak).4.KC pump trip criteria:*IF KC surge tank level goes below.5 ft and valid, THEN: au Trip affected pumps.b.Isolate affected train PER Enclosure 2 (Isolation
of KC Non-essential
Headers).5.VCT high temperature:
Actions).
AP/1 and 21A15500/021 (Loss of KC or KC System Leakage)STEP DESCRIPTION
FOR AP STEP 1: PURPOSE: Ensure letdown and all NM is isolated if KC pumps are off.DISCUSSION:
Since KC cools the letdown Hx, letdown is isolated if KC is lost.Note subsequent
steps that trip KC pumps or isolate cooling to the aux bldg non-ess header will also isolate letdown.Engineering
has calculated
that if KC flow through the NM Hx's is lost that it would take less than 3 minutes to flash.REFERENCES
OEDB 98-18676 STEP 2: PURPOSE: Cue the operator to monitor the foldout page.DISCUSSION:
A foldout page was chosen for this AP as a human-factors'
consideration.
Maintaining
critical items on a separate page ensures they are performed in a timely manner.The foldout page contains actions that apply throughout
the AP as described in items below: 1)"KC header isolation criteria" ensures the non-essential
headers are isolated from the KC pump and essential header prior to emptying the surge tank.If a leak occurs, theessential headers should be isolated prior to air binding the KC pumps.If the leak is on the operating train KC essential header, isolating the non-essential
headers will prevent them from draining (so they can be restored in a timely manner using other train).If the leak is on one of the non-essentialheaders,this
isolation protects the essential header.Adequate protection
of equipment cooled by the non-essential
headers is provided by isolating letdown and by other foldout page items.Note efforts to makeup to the surge tank and isolate leaks will be initiated for smaller leaks prior to having to isolate entire headers.This foldout partially addresses some concerns raised by the NRC in OEDB 98-017559, Loss of inventory from KC.A major concern was not getting a leakingessential header isolated in time to prevent the inoperability
of the safety-related
headers.2)"NC pumptripcriteria".
Isolation of the reactor bldg non-essential
header or loss of KC pumps may lead to NC pump trip criteria due to loss of cooling to motor bearings.Since Page 3 of 23 Rev 3
MNS AP/1/A/5500/21
UNITl LOSS OF KG OR KG SYSTEM LEAKAGE PAGE NO.7 of 78 Rev.9 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 15.Check both train*s KC surge tank levelGREATER THAN 3 FT._GO TO Step 20.NOTE*The following OAG points may be used to determine level drop in next step.These points are also displayed on the KG system graphic:*M1 P1317 (1A Train KG surge tank level rate)*M1 P1318 (1 B Train KG surge tank level rate).*A 0.10 ftlmin level drop in one train's surge tank equals approximately
50 GPM leak.16.Check sum of both trains*KC surge tank level drops-LESS THAN OR EQUAL TO 0.10 FT/MIN._IF level is dropping faster than 0.10 ftlmin, THEN GO TO Step 20.NOTE The next step allows maintaining
current KG system alignment for small leaks that should be within the capacity of normal makeup.Allowing level to drop to 2 ft allows more time for operators to locally align makeup, prior to taking action to isolate KG headers.17.Do not continue until at least one of the following occurs:_.KG makeup has been locally opened from RN.OR_.Either train's KG Surge Tank level is less than or equal to 2 ft.OR*Both KG surge tank levels are stable or going up.
AP/1 and 21A15500/021 (Loss of KC or KC System Leakage)If KC surge tank level is greater than 3 ft, time is given for the operators to attempt to initiate makeup and check results to see if the surge tank level can be maintained.
Per engineering, YM makeup should be sufficient
to keep up with the FSAR design basis leak of 50 GPM.Operators should be able to initiate makeup prior to reaching 2 ft in the surge tanks (assuming makeup is initiated when KC 10 level alarms at 4.5 ft).If the trains are cross-tied, allowing leaving the cross-ties
open doubles the volume (and time)to initiate makeup.Note that for larger leaks, or if level reaches 2 ft, the cross-ties
will be closed to protect the other train.If makeup is initiated and level stabilizes, operator actions are greatly simplified.
STEP 16 NOTES: PURPOSE: Give operator information
for determining
leak rate.DISCUSSION:
0.1 ftlmin level drop is equivalent
to the design basis leak (50 gpm)that YM makeup should be able to keep up with.STEP 16: PURPOSE: Procedure flow path controlling
step.DISCUSSION:
If leak is greater than design basis leak, operators need to find the leak and isolate it.Page 10 of 23 Rev 3
MNS AP/1/A/5500/21
UNITl LOSS OF KC OR KC SYSTEM LEAKAGE PAGE NO.8 of 78 Rev.9 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 18.Check KC surge tank level on both train(s)-STABLE OR GOING UP.19.GO TO Step 38.20.Isolate 1 A KC Train from 1 B KC Train as follows: a.Check any1A KC Train pumpRUNNING.b.Check the following valves-OPEN:*1 KC-3A (Trn A Rx Bldg Non Ess Ret Isol)*1 KC-230A (Trn A Rx Bldg Non Ess Sup Isol).c.Close the following valves:_1)1 KC-53B (Trn B Aux Bldg Non Ess Sup Isol)._2)1 KC-2B (Trn B Aux Bldg Non Ess Ret Isol)._3)1 KC-228B (Trn B Rx Bldg Non Ess Sup Isol)._4)1 KC-18B (Trn B Rx Bldg Non Ess Ret Isol).d.WHEN valves in Step 20.c are closed, THEN check 1A KC Surge Tank levelGOING DOWN.e.GO TO Step 21.IF KC surge tank level is still going down in an uncontrolled
manner, THEN: a.IF level goes below 2 ft, THEN ensure Foldout page item 1 is implemented.
b.GO TO Step 20.a.GO TO Step 20.1.b.GO TO Step 20.1.d.IF1A KC Surge Tank level stabilizes, AND1B KC Surge tank level continues to go down, THEN leak is on1B Essential header.
AP/1 and 21A15500/021 (Loss of KC or KC System Leakage)STEP 17 NOTE: PURPOSE: Inform operators why they're waiting.DISCUSSION:
STEP 17: PURPOSE: Establish a hold point in the AP until one of the listed items is met.DISCUSSION:
The basis for this step is to allow a chance for makeup to be established
to compensate
for the leak.If it does, or if RN has to be established
to keep up, or if the surge tank gets less than 2 ft, the hold point is released.STEPS 18&19: PURPOSE: Flow path controlling
steps.DISCUSSION:
If makeup is keeping up with the leak, actions to isolated entire headers are bypassed.If level is going down, it's time to begin isolating headers to stop the leak.STEP 20: PURPOSE: Begin the process of isolating KC headers so the leaking header can be identified.
The first step involves closing the non-operating
trains'4 cross-ties
to split the two essential headers.DISCUSSION:
If the A train pumps are running and supplying the Rx non-ess header, then the B trainties (Aux non-ess supply&return, and Rx non-ess supply&return)
are closed.Otherwise, if A train is not operating, then the A train cross-ties
are closed.After the cross-ties
are closed, the surge tank levels are checked.If the operating trains'level stabilizes, and the non-operating
Page 11 of 23 Rev 3
I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#1----Group#1 KIA#029 EA2.01----------
Importance
Rating 4.7----Abilitytodeternline
or interpret the following as they apply toaA TWS: Reactor nuclear instrurnentation
Proposed Question: SRO 78 Given the following conditions:
- An ATWS has occurred on Unit 1.*The crew is performing
FR-S.1, Response to Nuclear Power Generation/
ATWS.*NC Boration is in progress.*81 has actuated.*All SG pressures are approximately
800 psig and trending down.*NC Temperature
is approximately
490 degrees F and trending down.*Enclosure 2 (Faulted SG Isolation)
has been initiated.
- Reactor Power indicates approximately
4%and trending down slowly.Which ONE (1)of the following describes the mitigation
strategy for the event in progress?A.Remain in FR-S.1 and perform Enclosure 2.Transition
to E-O, Reactor Trip or Safety Injection ONLY after all steps of Enclosure 2 are complete and NC system temperature
is stable.B.Remain in FR-S.1 and perform Enclosure 2.Transition
to E-O, Reactor Trip or Safety Injection when Intermediate
Range amps are going down.C.Conditions
exist that allow exit from FR-S.1.When directed, exit FR-S.1 while continuing
performance
of Enclosure 2.Transition
to E-O, Reactor Trip or Safety Injection, prior to transition
D.Conditions
exist that allow exit from FR-S.1.When directed, exit FR-S.1 and terminate performance
of Enclosure 2.Transition
to E-2, Faulted Steam Generator Isolation, prior to transition
Page 195 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Proposed Answer: C Explanation (Optional):
A is incorrect.
FR-S.1 has guidance to isolate a faulted SG, but cannot transition
until power is below 5%B is incorrect.
Would go to E-O after FR-S.1 is complete and directed by FR-S.1 (Power<50/0)C is Correct.Power less than 5%, transition
may occur.Enclosure 2 is still completed if in progress.If conditions
are present, some steps of FR-S.1 may have to be performed prior to exit, because the crew may not be at step 17 D is incorrect.
Credible because a fault exists and procedure flowpath is correct, but E-O is performed first Technical Reference(s): FR-S.1, Rev10 OMP 4-3 p14, 18 (Attach if not previously
provided)Proposed references
to be provided to applicants
during None examination:
Learning Objective:
Question Source: FR-S.1 Obj 4 Bank#Modified Bank#New McGuire 2006 NRC 80 (Note changes or attach parent)Question History: Question Cognitive Level: Last NRC Exam Memory or Fundamental
Knowledge Comprehension
or Analysis x 10 CFR Part 55 55.41 Content: 55.43 5 Comments: KA is matched because transition
is made based upon PR NI indications.
Page 196 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 SRO level because the item addresses FR-S.1 strategy and compliance
with EOPs.The applicant must determine exit conditions
available and interpret use of EOP attachments
while performing
other procedures
Page 197 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Given the following conditions:
- An ATWS has occurred on Unit 1.*The crew is performing
FR-S.1, Response to Nuclear Power Generation/
ATWS.*NC Boration is in progress.*SI has actuated.*All SG pressures are approximately
800 psig and trending down.*NC Temperature
is approximately
490 degrees F and trending down.*ReactorPower
indicates approximately
7%and trending down slowly.Which ONE (1)of the following describes the mitigation
strategy for the event in progress?A.Remain in FR-S.1 and perform Enclosure 2 (Faulted SG Isolation).
Transition
to E-O, Reactor Trip or Safety Injection when Enclosure 2 is complete.B.Remain in FR-S.1 and perform Enclosure 2 (Faulted SG Isolation).
Transition
to E-O, Reactor Trip or Safety Injection when reactor power is less than 5%.C.Exit FR-S.1;Transition
to E-O, Reactor Tr,ip or Safety Injection to ensure actuated components
are in their correct alignments.
D.Exit FR-S.1;Transition
to E-O, Reactor Trip or Safety Injection and ONLY perform steps of subsequent
EOPs that do not contradict
the actions taken in FR-S.1.Ans.B Page 198 of 260 Draft 7
MNS EP/1/A/5000/FR-S.1
UNITl RESPONSE TO NUCLEAR POWER GENERATION/ATWS
PAGE NO.7 of 29 Rev.10 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 13.Check steamlines
intact:*All S/G pressures-STABLE OR GOING UP*All S/Gs-PRESSURIZED.
IF any S/G depressurized
OR pressure going down in an uncontrolled
manner, THEN: a.Ensure the following valves closed:*All MSIVs*All MSIV bypass valves.b.IF any S/G depressurized
OR pressure still going down in an uncontrolled
manner, THEN isolate any faulted S/G(s)PER Enclosure 2 (Faulted S/G Isolation).
MNS EP/1/A/5000/FR-S.1
UNIT 1 RESPONSE TO NUCLEAR POWER GENERATION/ATWS
PAGE NO.g of 29 Rev.10 ACTION/EXPECTED
RESPONSE 16.Check reactor subcritical:
- P/R channels-LESS THAN*W/R Neutron Flux-LESS THAN 5%*I/R SUR-NEGATIVE.17.Ensure adequate shutdown margin: RESPONSE NOT OBTAINED Perform the following:
a.Continue to borate.b.IF boration is not available, THEN allow NC System to heat up.c.Perform actions of any other Critical Safety Function procedures
that apply or are in effect that do not cool down NC System or add positive reactivity
to the core.d.RETURN TO Step 5.a.Obtain current NC boron concentration
from Primary Chemistry.
b.WHEN current NC boron concentration
is obtained, THEN perform shutdown margin calculation
PER OP/OI A/61 001006 (Reactivity
Balance Calculation).
c.WHEN following conditions
satisfied, THEN NC System boration may be stopped:*Adequate shutdown margin is obtained*Uncontrolled
cooldown has been stopped.18.REFER TO RP/O/Al5700/000 (Classification
of Emergency)
..19.RETURN TO procedure and step in effect.
7.15.1.5 7.15.1.6 OMP4-3 Page 18 of 35 Orange Path IF any valid orange path is encountered, the operator is expected to scan all of the remaining trees, and then, if no valid red is encountered, promptly implement the corresponding
EP.IF during the performance
of an orange path procedure, any valid red condition or higher priority valid orange condition arises, the red or higher priority orange condition is to be addressed first, and the original orange path procedure suspended.
Completion
of Red or Orange Path Procedure Once procedure is entered due to a red or orange condition, that procedure should be performed to completion, unless preempted by some higher priority condition.
It is expected that the actions in the procedure will clear the red or orange condition before all the operator actions are complete.However, these procedures
should be performed to the point of the defined transition
to a specific procedure or to the"procedure
and step in effect" to ensure the condition remains clear.At this point any lower priority red or orange paths currently indicating
or previously
started but NOT completed shall be addressed.
FR-S.1, P.1 and Z.l can be entered from either an orange or red path status.IF the color changes from orange to red while you are in one of these EPs, the crew should continue and complete the EP from where they are.Crew does NOT have to backup and restart the EP.IF the orange path is exited, and it subsequently
turns red, the EP must be re-entered
at Step 1.Upon continuation
of recovery actions in Optimal Recovery procedure, some judgment may be required by the operator to avoid inadvertent
reinstatement
of a Red or Orange condition by undoing some critical step in the Function Recovery procedure.
The Optimal Recovery procedures
are optimal assuming that safety equipment is available.
The appearance
of a Red or Orange condition in most cases implies that some equipment or function required for safety is NOT available, and by implication
some adjustment
may be required in the Optimal Recovery procedure.
7.10.5 OMP4-3 Page 14 of 35 Use of Enclosures
The decision on whether to read or hand-off an enclosure will be based on SRO judgment depending on the event.The following are some general guidelines
to help the SRO make this decision.*It is usually preferable
for SRO to read enclosure if:*The crew must waitforthe enclosure to be completed in order to continue in the EP/AP.*No more ROs are available to continue in the EP/AP, unless RO can perform enclosure concurrent
with performing
other steps.*There are no more time critical actions to be performed.
- It is usually preferable
for SRO to hand-off the enclosure if:*It is criticalforthe crew to continue in the body of the procedure in a timely manner.*It is a valve checklist.
- Actions are outside the horseshoe.
Additionally, an enclosure will be handed off if procedure specifies to hand-off the enclosure or if it is the foldout page.7.11 Place Keeping Aids EPs and APs contain a single line to the left and adjacent to the step number.The line is provided as a placekeeping
aid.Check-off the place keeping line after step is completed.
For a"check" step that requires no action, step can be checked after it is read.For steps that require action, step should be checked when action has been completed.
For example, if a valve must be closed, place keeping line should be checked when operator states that valve is closed on second three way communication.
For slow moving valves or situations
where procedure reader must move on while waiting for step completion, circle the place keeping line until step is completed;
check-off the place keeping line when performer later feeds back that step is completed.
ONE EXCEPTION to this is performance
ofES-1.3 (Transfer to Cold Leg Recirc).While performing
multiple valve manipulations
in ES-1.3, operator should proceed in EP in a timely manner and just check-off steps as they are read.This avoids excessive delays when performing
this time critical evolution.(This exception is implied by ES-1.3 note that states that double three-way communication
is NOT required.)
IF the step is a diagnostic
step that requires transition
to RNO, place right arrow (--7)next to step in lieu of, or in addition to check mark.Note that if you read an IFITHEN step that does NOT require performing
its substeps ("IF"condition
NOT met), do NOT check the substeps.The substeps will NOT be read or performed, and you only need to check steps if you READ them.Check next to IFffHEN step if it is all that is read, whether it has a place keeping line or not
I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#1----Group#1 KIA#055 EA2.03----------
Importance
Rating 4.7----Ability to determine or interpret the following as tlley apply to a Station Blackout Actions necessary to restore power Proposed Question: SRO 79 Given the following:*A LOOP has occurred on Unit 1.*Unit 2 is unaffected.
- The Unit 1 crew is performing
ECA-O.O, Loss of All AC Power.*The Standby Makeup Pump is ON.*NCS subcooling
is 8°F.*Pressurizer
level is 4%and lowering slowly.*The crew was unable to start EITHER Diesel Generator.
Which ONE of the following describes the procedure that will be required for restoring power to Bus ETA, and the subsequent
recovery procedure that will be performed upon transition
from ECA-O.O?A.AP/7, Loss of Electrical
Power;ECA-0.1, Loss of All AC Power Recovery Without SI Required B.Enclosure 9, Energizing
Unit 1 4160 V Bus from Unit2-SATA or SATB;ECA-0.1, Loss of All AC Power Recovery Without SI Required C.AP/7, Loss of Electrical
Power;ECA-0.2, Loss of All AC Power Recovery With SI Required D.Enclosure 9, Energizing
Unit 1 4160 V Bus from Unit2-SATA or SATB;ECA-0.2, Loss of All AC Power Recovery With SI Required Proposed Answer: D Explanation (Optional):
Page 199 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 A.Incorrect.
API?is plausible because it is the procedure normally used for any electrical
restoration.
In this condition, Enclosure 9 will be used.Even though Auto SI conditions
do not exist, the crew will perform ECA-O.2 based on RCS subcooling
and PZR level values requiring SI when power restored B.Incorrect.
Enclosure 9 is correct.Plausible because even though Auto SI conditions
do not exist, the crew will perform ECA-O.2 based on RCS subcooling
and PZR level values requiring SI when power restored c.Incorrect.
Incorrect restoration, but correct recovery procedure for these plant conditions
D.Correct ECA-O.O, Encl 9 Rev 24 Technical Reference(s)(Attach if not previously
provided)----------
EP-EO Rev 24 EP-ECAO Rev 12 EP-EO Rev 12 OMP 4-3 P 22 Rev 26 Proposed references
to be provided to applicants
during None examination:
Learning Objective: (As available)
(Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge Comprehension
or Analysis X 1 a CFR Part 55 55.41 Content: 55.43 2,5 Comments: KA is matched because the applicantmustidentify
where the actions are contained for restoration
of power.(title
also identifies
actions)and SRO level Page 200 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 because assessment
of conditions
and selection of procedures
is required (Requires knowledge of strategy)Page 201 of 260 Draft 7
MNS EP/1 I A/5000/ECA-0.0
UNITl LOSS OF ALL AC POWER PAGE NO.33 of 163 Rev.24 ACTION/EXPECTED
RESPONSE 40.Select recovery procedure as follows: a.Check Standby Makeup pump-ON.b.Check NC subcooling
based on core exit TICs-GREATER THAN OaF.c.Check pzr level-GREATER THAN 11°k>(29°k>ACC).RESPONSE NOT OBTAINED a.IF all NC pump seal cooling is lost, THEN notify station management
that NC pump seal cooldown will occur as the entire NC system is cooled via natural circ cooldown in subsequent
EPs.b.Perform the following:
_1)Align additional
RN valves PER Enclosure 24 (RN SII Valves)._2)GO TO EP/1/A/5000/ECA-0.2 (Loss Of All AC Power Recovery With SII Required).
c.Perform the following:
_1)Align additional
RN valves PER Enclosure 24 (RN SII Valves)._2)GO TO EP/1/A/5000/ECA-0.2 (Loss Of All AC Power Recovery With SII Required).
d.Check the following valves-CLOSED:*1 NI-9A (NC Cold Leg Inj From NV)*1NI-10B (NC Cold Leg Inj From NV).e.GO TO EP/1/A/5000/ECA-0.1 (Loss Of All AC Power Recovery Without S/I Required).
d.IF any NV pump on, THEN perform the following:
_1)Align additional
RN valves PER Enclosure 24 (RN SII Valves)._2)GO TO EP/1/A/5000/ECA-0.2 (Loss Of All AC Power Recovery With SII Required).
MNS EP/1/A/SOOO/E-O
UNITl REACTOR TRIP OR SAFETY INJECTION PAGE NO.3 of 36 Rev.24 ACTION/EXPECTED
RESPONSE c.Operator Actions 1.Monitor Foldout page.o Check Reactor Trip:*All rod bottom lights-LIT*Reactor trip and bypass breakersOPEN*I/R amps-GOING DOWN.G)Check Turbine Trip:*All throttle valves-CLOSED.-0 Check 1 ETA and 1ETB-ENERGIZED.
RESPONSE NOT OBTAINED Perform the following:
a.Trip reactor.b.IF reactor will not trip, THEN:*Implement EP/1/A/SOOO/F-O (Critical Safety Function Status Trees).*GO TO EP/1/A/SOOO/FR-S.1 (Response To Nuclear Power Generation/
A TW S).Perform the following:
a.Trip turbine.b.IF turbine will not trip, THEN:_1)Place turbine in manual._2)Close governor valves in fast action.3)IF governor valves will not close, THEN close:*All MSIVs*All MSIV bypass valves.Perform the following:
a.IF both busses de-energized, THEN GO TO EP/1/A/SOOO/ECA-O.O (Loss Of All AC Power).b.WHEN time allows, THEN try to restore power to de-energized
bus PER AP/1/A/SSOO/07 (Loss of Electrical
Power)while continuing
with this procedure.
DUKE POWER ECA-O.O Loss of All AC Power MCGUIRE OPERATIONS
TRAINING STEP 40 Select recovery procedure:
PURPOSE: To select the appropriate
loss of all AC power recovery procedure.
BASIS: This step provides the criteria by which the operator determines
which recovery procedure actions to implement.
The criteria are:1.The existence of NC subcooling
2.The existence of pressurizer
level 3.The confirmation
that SII equipment is not operating (NI-9 and NI-10 closed)Two recovery procedures
are provided based on these criteria.These are procedures
ECA-O.1 and ECA-O.2.If the operator determines
all criteria are satisfied, ECA-O.1 should be implemented
to attempt plant recovery utilizing normal operational
systems.If any criterion is not satisfied, ECA-O.2 should be implemented
to recover the plant utilizing safeguards
systems.To ensure SII has not actuated upon power restoration, the positions of the cold leg injection isolation valves(NI-9and NI-10)are checked.These valves do not II sea l in ll the SII signal and do not receive a signal through the DIG load sequencer that is deenergized.
If an SII signal was generated prior to power restoration, the procedure would reset the signal after the time delay and no equipment would reposition (NI-9 and NI-10 would remain closed).If either valve were open at this point in the procedure, it would indicate that an SII signal was generated with power restored and certain valves may have repositioned;
specifically
valves that receive direct SII signals.In this case, ECA-O.2 would direct the operator to the correct procedure to handle the accident or to terminate the spurious S/I.3.5.ECA-O.O Enclosures
Enclosure 1, Unit 1 (2)SSF Actions-ECA-O.O Actions This enclosure provides actions to be taken upon manning the SSF.These actions, if necessary, include starting the SSF DIG, loading equipment on the bus (standby makeup pump, battery chargers)and monitoring
DIG operation.
Enclosure 2, Unit 1 (2)EMXA-4 ECA-O.O Actions This one step enclosure provides the instructions
necessary to transfer EMXA-4 to the SSF.A caution provides guidance for operating Kirk-key interlocked
breakers.A note provides the fastest pathway from the Control Room to ETA room.OP-MC-EP-ECA-O
FOR TRAINING PURPOSES ONL Y Page 61 of 161 REV.12
DUKE POWER MCGUIRE OPERATIONS
TRAINING STEP 2&3 Check Reactor and Turbine Trip: (IMMEDIATE
ACTIONS)PURPOSE: To ensure the reactor and turbine are tripped.BASIS: Reactor trip must be checked to ensure the only heat being added to the NC system is from decay heat and NC pump heat.The safeguards
systems protecting
the plant during accidents are designed assuming only decay heat and pump heat are being added to the NC.If the reactor is not tripped, the RNa directs us to trip it manually.If the reactor cannot be tripped F-O, CSF Status Trees, is implemented
and a transition
is made to FR-S.1, Response to Nuclear Power Generation/ATWS, to deal with the ATWS conditions.
The turbine is tripped to prevent an uncontrolled
cooldown of the NC due to steam flow that the turbine would require.If the turbine is not tripped, the RNa directs us to trip it manually.If the turbine will not trip, steam is isolated to it by first attempting
to close the turbine governor valves.If the turbine will not runback, steam is isolated to it by closing the MSIVs and bypass valves.STEP 4 Check 1 ETA and 1ETB-ENERGIZED.(IMMEDIATE
ACTION)PURPOSE: To ensure electrical
power to at least one emergency bus.BASIS: AC power must be checked from either offsitesourcesor the diesel generators
to ensure adequate power sources to operate safeguards
equipment.
At least one train of safeguards
equipment is required to deal with emergency conditions.
If both AC emergency busses are deenergized, the RNa directs a transition
to ECA-O.O, Loss of All AC Power.OP-MC-EP-EO
FOR TRAINING PURPOSES ONL Y Page 27 of 207
OMP4-3 Page 22 of 35 7.18 Multiple Use ofEPs and APs.The Control Room SRO will determine how many procedures
can be implemented
at a time and their priority based on manpower availability
and the particular
event in progress.More than one EP shall NOT be run concurrently
unless directed by the procedure.
Generally the use of APs in conjunction
with EPs should be avoided.In some instances it would be proper to use an AP concurrently
during a major accident which is being addressed by the EPs.An example of this is upon loss of all Nuclear Service Water in the middle of an accident, the operators would need to utilize the AP for Loss of RN also.IF an AP is used during an SII event, USE CAUTION.APs are generallywrittenassuming
an SII has NOT occurred (exception
-AP/35, ECeS Actuation During Plant Shutdown).
Evaluate any AP steps in post S/I events to ensure the steps do NOT conflict with any EP in effect.NOT all AP actions would be appropriate
if an S/I occurred.(Enclosures
in EP/G-1 (Generic Enclosures)
may be used when reference by EPs or APs.)
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#1----Group#1 KIA#055 EA2.05----------
Importance
Rating 3.7----ECA-O.O, Loss of All AC Power.*A Blackout has occurred*Unit 2 is unaffected.
- The Unit 1 crew is performi Given the following:
Ability to determine or interpret the following as they apply to a Station Blackout: When battery is approaching
fully discharged
0 Proposed Question: SRO 79/().1---()--Which ONE of the following describe the technic specification
design basis for the operability
of Battery EVCA, and tH action r uired to extend the life of BatteryEVCAduring
the blackout?The battery has adequate storage capacit 10supplythe duty cycle output for...A.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />;evaluate shutting down soc ted inverter and aligning vital AC Panelboards
to KRP.B.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />;evaluate removing th C.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />;evaluate shutting own associat d inverter and aligning vital AC Panelboards
to KRP D.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />;evaluate removin the OAC from service.Proposed Answer: A Explanation (Optional):
A.Correct.B.Incorrect.
OAC is supplied from Aux Control Power (DCA/DCB)Plausible because it is an action performed if power cannot be restored C.Incorrect.
Incorrect time, though standard time for design basis battery life, Page 188 of 238
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 And also time for TS LCO action D.Incorrect.
Incorrect time, though standard time for design basis battery life, And also time for TS LCO action.OAC is supplied from Aux Control Power (DCA/DCB)Plausible because it is an action performed if power cannot be restored Technical Reference(s)ECA-O.O, API?Enclosure?TS Basis 3.8.4 Proposed references
to be provided toapplicantsduring
None examination:
Learning Objective: (As available)
(Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
KnowledgeComprehensionor
Analysis X 10 CFR Part 55 55.41 Content: 55.43 2,5 Comments: Page 189 of 238
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#1----Group#1 KIA#056 AA2.18----------
Importance
Rating 4.0----Ability to deterrnine
and interpret the following as they apply to the Loss of Offsite Power: Reactor coolant temperature, pressure, and PZR level recorders Proposed Question: Given the following:
SR080*A loss of off-site power has occurred.*Both Units have tripped.*Unit 1 SRO has been directed to initiate cooldown to Mode 5.*The following conditions
exist on Unit 1 upon transition
to ES-0.1, Reactor Trip Response.o All control
are inserted.o NC SYSTEM Tcold temperature.
- Loop1A 535°F*Loop1B 532°F*Loop1C 533°F*Loop1D 533°F Which ONE of the following choices describes (1)actions that will be required for the above conditions, and (2)the procedure required for N.C System Cooldown?A.(1)Close MSIVs ONLY;(2)OP/1/A/61 00/002, Controlling
Procedure for Unit Shutdown.B.(1)Close MSIVs ONLY;(2)ES-0.2, Natural Circulation
Cooldown.C.(1)Close MSIVs AND Initiate Emergency Boration in accordance
with AP/38, Emergency Boration;(2)OP/1/A/61 00/002, Controlling
Procedure for Unit Shutdown.Page 202 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 D.(1)Close MSIVs AND Initiate Emergency Boration in accordance
with AP/38, Emergency Boration;(2)ES-O.2, Natural Circulation
Cooldown.Proposed Answer: D Explanation (Optional):
A.Incorrect.
MSIVs are closed, but if a cooldown is required with a LOOP, then ES-O.2 would be performed instead of the Controlling
Procedure.
Also, due to Loop1D temperature, emergency boration is required B.Incorrect.
Due to Loop1D temperature, emergency boration is required C.Incorrect.
Actions are correct but procedure is incorrect as in A D.Correct.Technical Reference(s)ES-O.1, Rev 27;ES-O.2 Rev 10 EP-EO Rev 12 Proposed references
to be provided to applicants
during None examination:
Learning Objective: (As available)
(Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge Comprehension
or Analysis X10 CFR Part 55 55.41 Content: 55.43 5 Comments: KA is met because item evaluates interpretation
of RCS temperature
trends.SRO level because the assessment
requires interpretation
of indications
to take Page 203 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 action within selected EOPs/AOPs Page 204 of 260 Draft 7
MNS EP/1/N5000/ES-0.1
UNITl REACTOR TRI P RESPONSE PAGE NO.4 of 50 Rev.27 ACTION/EXPECTED
RESPONSE 5.Check NC temperatures:
- IF any NC pump on, THEN check NC T-Avg-STABLE OR TRENDING TO 557°F.OR*IF all NC pumps off, THEN check NC T-Colds-STABLE OR TRENDING TO 557°F.RESPONSE NOT OBTAINED Perform the following based on plant conditions:
a.IF temperature
less than 557°F AND going down, THEN perform the following:
_1)Ensure all steam dump valves closed.2)IF MSR"RESET" light is dark, THEN perform the following:
_a)Depress"SYSTEM MANUAL"._b)Depress"RESET"._3)Ensure all SM PORVs closed.4)IF any SM PORV can not be closed, THEN perform the following:
_a)Close SM PORV isolation valve._b)IF SM PORV isolation valve can not be closed, THEN dispatch operator to close SM PORV isolation valve._5)Ensure S/G blowdown is isolated.6)IF cooldown continues, THEN control feed flow as follows: a)IF S/G N/R level is less than 11%in all S/Gs, THEN throttle feed flow to achieve the following:
- Minimize cooldown*Maintain total feed flow greater than 450 GPM.b)WHEN N/R level is greater than 11%in at least one S/G, THEN throttle feed flow further to:*Minimize cooldown*Maintain at least one S/G N/R level greater than 11 0/0.RNO continued on nextae
MNS EP/1/A/5000/ES-0.1
UNITl REACTOR TRIP RESPONSE PAGE NO.5 of 50 Rev.27 ACTION/EXPECTED
RESPONSE 5.(Continued)
RESPONSE NOT OBTAINED 7)IF cooldown continues, THEN perform the following:
_a)Close all MSIVs._b)Close all MSIV bypass valves._c)Close 1AS-12 (U1 SM To AS Hdr Control Inlet Isol).d)IF the MSIVs will not close, THEN perform the following:
_(1)Initiate Main Steam Isolation signal.(2)IF all S/G pressures are above 775 PSIG, THEN reset the following to allow automatic SM PORV operation:
1.Main Steamline Isolation.
2.SM PORVs._8)IF cooldown continues AND faulted S/G exists, THEN stop feeding faulted S/G.9)IF cooldown continues, THEN select"CLOSEII on the following switches:*1 SM-83 (A SM Line Drain Isol)*1 SM-89 (8 SM Line Drain Isol)*1 SM-95 (C SM Line Drain Isol)*1 SM-1 01 (D SM Line Drain Isol).(RNO continued on next page)
DUKE POWER MCGUIRE OPERATIONS
TRAINING STEP 5 Check NC temperatures:
PURPOSE: To ensure that NC heat is being properly removed through the secondary side.BASIS: NC average temperature
stable or trending to the no-load value of 557°F with any NC pump running indicates that the secondary steam dump system is operating as designed.If temperature
is stable, even if not at 557°F, you can continue in the left-hand column.If no NC pump is running, then the NC average temperature
will be higher than the no-load value as natural circulation
conditions
are established.
However, if the steam dump system is working properly, the cold leg temperatures
will stabilize at the no-load value.If the cooldown is excessive, it can be controlled
by:1.Stopping all steam from being dumped, 2.Controlling
feed flow, or 3.Closing the MSIVs.Steam dump should be stopped by assuring that steam dump valves are closed, S/G PORVs are closed, and SM Line drains are closed.Excessive feed to the S/Gs can also result in cooling down the NC and it may be necessary to reduce feed flow to the minimum for decay heat removal until S/G level is in the narrow range.If the cooldown continues, the main steamlines
are isolated to stop any steam leakage downstream
of the MSIV's, such as a stuck open condenser steam dump valve.Also, AS-12 is isolated to ensure flow to the Aux Steam system is secured.If NC temperature
is greater than no-load and going up, then steam dump from the secondary must be raised for decay heat removal.OP-MC-EP-EO
FOR TRAINING PURPOSES ONL Y Page 95 of 207 REV.12
MNS EP/1/A/5000/ES-0.1
UNITl REACTOR TRIP RESPONSE PAGE NO.8 of 50 Rev.27 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 11.Check feedwater status:*Check any CA pump-ON.*Check total feed flow to S/GsGREATER THAN 450 GPM.Establish total feed flow to S/Gs greater than 450 GPM or maintain at least one S/G N/R level greater than 11%using one of the following:
_.Start CA pumps.OR*Use main feedwater PER Enclosure 4 (Reestablishing
CF Flow).12.Check if shutdown margin adequate: a.All control rods-FULL Y INSERTED.a.Perform the following:
_1)IF all rod position indication
is lost, OR greater than 5 rods not fully inserted, THEN emergency borate total of 13,200 gallons of 7000 PPM boron solution PER AP/1/A/5500/38 (Emergency
Boration).
_2)IF 2 to 5 rods not fully inserted, THEN emergency borate 2100 gallons of 7000 PPM boron solution for each rod not fully inserted PER AP/1/A/5500/38 (Emergency
Boration).
b.Stop any boron dilutions in progress.c.Check all NC T-Colds-GREATER c.Borate as follows: THAN 534°F._1)Set boric acid flow control pot at 6.5._2)Initiate emergency boration PER AP/1/A/5500/38 (Emergency
Boration).
_3)WHEN all NC T-Colds are above 534°F, THEN emergency boration may be secured._4)GO TO Step 13.d.IF AT ANY TIME any NC T-Cold goes below 534°F, THEN perform Step 12.c.
DUKE POWER MCGUIRE OPERATIONS
TRAINING STEP 10 Check NC T-Ave-GREATER THAN 553 QF STEP 11 Check feedwater status: PURPOSE: To ensure the proper feedwater alignment following a reactor trip.BASIS: T-Ave is not expected to fall to the feedwater isolation setpoint of 553°F, so feedwater isolation should not have occurred.If T-Ave is less than 553°F, then by checking the status lights lit, all feedwater isolation valves can be assured closed for a 8/G as required.Establishing
minimum feed flow to the steam generators
or minimum 8/G levels ensures a secondary heat sink for decay heat removal.The feedwater source may be from either the CA pumps or main feedwater on the bypass lines.STEP 12 Check if shutdown margin adequate: (Continuous
Action Step)PURPOSE: To ensure that the shutdown margin is adequate.BASIS: A subcritical
core is confirmed if all rods are at the bottom according to the rod bottom lights and the rod position indicators.
If these indications
reveal that one rod is not inserted, no immediate action is required since the core is designed for adequate shutdown margin with one rod stuck out.Any boron dilutions in progress should be stopped to ensure shutdown margin is not challenged.
If more than one rod fails to insert fully, the shutdown reactivity
margin must be made up through emergency boration to account for the reactivity
worth of the stuck rods.If two to five rods do not fully insert, then emergency borate 2100 gallons of 7000 ppm boron solution.If all rod position indication
is lost or more than five rods are not fully inserted, emergency borate 13,200 gallons of 7000 ppm boron solution.Also, if NC T-Colds are less than the cycle specific value at which 80M is calculated
to be challenged (typically
near 534°F), the loss in shutdown margin due to low NC temperatures
must be made up by emergency boration per AP/38 until all NC T-colds are above the specified temperature.
OP-MC-EP-EO
FOR TRAINING PURPOSES ONL Y Page 99 of 207 REV.12
MNS EP/1/N5000/ES-0.1
UNITl REACTOR TRIP RESPONSE PAGE NO.32 of 50 Rev.27 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 43.Check if Reactor Trip was performed as part of normal shutdown as follows: a.Check if OP/1/N61 001003 (Controlling
Procedure For Unit Operation), Enclosure 4.1 0 (Shutdown Via Reactor Trip)-IN EFFECT PRIOR TO TRIP.b.Checkany NC pump-ON.c.RETURN TO step in effect in OP/1/N61 001003 (Controlling
Procedure For Unit Operation), Enclosure 4.1 0 (Shutdown Via Reactor Trip).44.REFER TO OP/1/Al61 00/003 (Controlling
Procedure For Unit Operation), Enclosure 4.10 (Shutdown Via Reactor Trip)and perform applicable
steps.45.Determine if Natural Circulation
cooldown is required: a.Check if plant cooldown-REQUIRED.b.Check if all NC pumps-OFF.c.GO TO EP/1/A/5000/ES-0.2 (Natural Circulation
Cooldown).
a.GO TO Step 44.b.GO TO Step 44.a.GO TO OP/1/N61 001003 (Controlling
Procedure For Unit Operation), Enclosure 4.1 (Power Increase).
b.GO TO OP/1/N61 001002 (Controlling
Procedure For Unit Shutdown).
DUKE POWER MCGUIRE OPERATIONS
TRAINING STEPS 22-43 These steps align systems for shutdown conditions.
PURPOSE: To stop equipment not needed following a reactor trip.BASIS: Since the plant may have been operating at full power prior to the trip, certain equipment may be in operation and not needed at this time (e.g., two condensate
pumps, circulating
water pumps, etc.).STEP 44 Determine if Natural Circulation
cooldown is required PURPOSE: To determine if a cooldown must be done on natural circulation.
BASIS: If theplantstaff
determines
that a cooldown is required, then a normal cooldown should be performed if one or more NC pumps are operating.
However, if no NC pumps are operating, then a natural circulation
cooldown will be necessary.
If a naturalcirculationcooldown
is required, then a transition
to ES-O.2, Natural Circulation
Cooldown, is made.OP-MC-EP-EO
FOR TRAINING PURPOSES ONL Y Page 109 of 207 REV.12
DUKE POWER 5.6.Final Plant Status MCGUIRE OPERATIONS
TRAINING ('\(ES-O.1 provides the specific actions necessary to stabilize and control the plant following a reactor trip.ES-O.1 is also used following a reactor trip combined with either a loss of offsite power or a total loss of forced NC flow.The following table summarizes
the exit guidance from ES-O.1.The left column lists each step that provides a potential exit point from ES-O.1.The right column lists the transition
procedure(s).
If an exit transition
is necessary, the operator should transition
to Step 1 of the appropriate
procedure.
E-O, Reactor Trip or Safety Injection OP/1/Al61 00/003, Controlling
Procedure for Unit Operation, Enclosure for"Shutdown Via Reactor Trip", if Reactor Trip was performed as part of a normal shutdown.5.7.Summary/Objective
Review The objective of the recovery/restoration
technique incorporated
into procedure ES-O.1 is to stabilize and control the plant following a reactor trip without safety injection in operation.
The recovery/restoration
technique of ES-O.1 includes the following five major action categories.1.Ensure the primary system stabilizes
at no-load conditions.
2.Ensure the secondary system stabilizes
at no-load conditions.
3.Ensure necessary APs that should be run concurrently
have been addressed.
4.Maintain/establish
forced circulation
of the NC.5.Maintain stable plant conditions.
OP-MC-EP-EO
FOR TRAINING PURPOSES ONL Y Page 117 of 207 REV.12
MNS EP/1/A/5000/ES-O.2
UNITl A.Purpose NATURAL CIRCULATION
COOLDOWN PAGE NO.1 of 35 Rev.10 This procedureprovidesactions
to perform a Natural Circulation
NC System cooldown and depressurization
to Cold Shutdown, with no accident in progress, under requirements
that will preclude any upper head void formation.
B.Symptoms or Entry Conditions
This procedure is entered from:*EP/1/A/5000/ES-O.1 (Reactor Trip Response), Step 44, when it has been determined
that a Natural Circulation
cooldown is required.*EP/1/A/5000/ES-1.1 (Safety Injection Termination), Step 31, after the plant conditions
have been stabilized
and no NC pumps can be started.*EP/1/A/5000/ECA-O.1 (Loss Of All AC Power Recovery Without S/I Required), Step 29, after the plant conditions
have been stabilized
following the restoration
of AC emergency power.
DUKE POWER MCGUIRE OPERATIONS
TRAINING 6.0 ES-0.2, NATURAL CIRCULATION
COOLDOWN 6.1.Purpose ES-O.2 provides actions to performanatural circulation
NC system cooldown and depressurization
to cold shutdown, with no accident in progress, under requirements
that will preclude any upper head void formation.
6.2.Symptoms/Conditions
Upon entry to ES-O.2, natural circulation
of the NC has been established
and stable plant conditions
are being maintained.
ES-O.2 is then entered from: 1.ES-O.1, Reactor Trip Response, when it has been determined
that a natural circulation
cooldown is required.2.ES-1.1, Safety Injection Termination, after the plant conditions
have been stabilized
and no NC pumps can be started.3.ECA-O.1, Loss Of All AC Power Recovery Without S/I Required, after the plant conditions
have been stabilized
following the restoration
of AC emergency power.There are three possible transitions
out of this procedure.
1.If S/I actuation occurs, a transition
to E-O, Reactor Trip or Safety Injection, should be made.2.Since it is always desirable to have forced convection
heat transfer from the core, the first step of the procedure attempts to start a NC pump.If this attempt is successful, a transition
to the appropriate
plant procedure is in order.3.The third transition
occurs if the plant staff determines
that a natural circulation
cooldown and depressurization
must be performed at a rate that may form a steam void in the vessel.At that time a transition
should be made to ES-O.3, Natural Circulation
Cooldown with Steam Void in Vessel.OP-MC-EP-EO
FOR TRAINING PURPOSES ONL Y Page 119 of 207 REV.12
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO (LSROTier#-t ()f_1__Group#.(V 1 KIA#("u Cf'058 G2.2.37 Importance
Rating 4.6----Equiprnent
Control: Ability to determine operability
and/or availability
of safety related equiprnent
Proposed Question: SRO 81 Given the following:
- Unit 1 is at 100%power.*A loss of Charger EVDA occurred.*Battery EV-DA voltage lowered to 1 09 VDC prior to restoration
of a Charger to the battery.*Battery EVDA voltage is currently 129 VDC.*Specific gravity is 1.180 for two (2)connected cells.*Average specific gravity is 1.202 for all connected cells.*Electrolyte
temperature
is 76°F.Which ONE of the following describes the operability
status of Battery EVDA, and the TS basis for operability
of the DC electrical
power subsystem?
A.The battery is considered
operable but degraded;operability
ensures that at least ONE DC train is available assuming a loss of off-site OR on-site power coincident
with a worst case single failure.B.The battery is considered
operability
ensures that at least ONE DC train is available assuming a loss of off-site OR on-site power coincident
with a worst case single failure.C.The battery is considered
operable but degraded;operability
ensures that at least ONE DC channel is available assuming a loss of off-site ANDsite power.D.The battery is considered
operability
ensures that at least ONE DC channel is available assuming a loss of off-site AND on-site power.Page 205 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Proposed Answer: B Explanation (Optional):
A.Incorrect.
Operable but degraded would be related to Category A or B parameter out of limits.In this
the applicant must determine that specific gravity is out of limit for category C, making battery inoperable
B.Correct.C.Incorrect.
See A.Also, basis plausible because it is similar to actual
except that Loss of off-site AND on-site power is NOT design basis for battery D.Incorrect.
Operability
is correct, but basis incorrect and plausible because it is similar to actual basis, except that Loss of off-site AND on-site power is NOT design basis for battery TS 3.8.6 and basis Technical Reference(s)(Attach if not previously
provided)----------
EL-EPL Rev 22 Proposed references
to be provided to applicants
during None examination:
EL-EPL#3 Learning Objective: (As available)
(Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge Comprehension
or Analysis X 1 0 CFR Part 55 55.41 Content: 55.43 2 Comments: KA matched because the applicant must determine operability
of selected Page 206 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 equipment related to selected APE.(Loss of DC)SRO level because a determination
of operability, and basis for operability, are the required knowledge items for this test item Page 207 of 260 Draft 7
Battery Cell Parameters
3.8.6 3.8 ELECTRICAL
POWER SYSTEMS 3.8.6 Battery Cell Parameters
LCO 3.8.6 Battery cell parameters
for the channels of DC batteries shall be within the limits of Table 3.8.6-1.APPLICABILITY:
When associated
channels of DC sources are required to be OPERABLE.ACTIONS---------------------------------------------------------NOlrE------------------------------------------------------------
Separate Condition entry is allowed for each battery.CONDllrlON
REQUIRED AClrlON COMPLElrlON
TIME A.One or more batteries A.1 Verify pilot cells electrolyte
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with one or more battery level and float voltage cell parameters
not meet lrable 3.8.6-1 within Category A or B Category C limits.limits.AND A.2 Verify battery cell 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sparametersmeet
lrable 3.8.6-1 Category C AND limits.Once per 7 days thereafter
AND A.3 Restore battery cell 31 days parameters
to Category A and B limits of lrable 3.8.6-1.(continued)
McGuire Units 1 and 2 3.8.6-1 Amendment Nos.184/166
Battery Cell Parameters
3.8.6 ACTIONS (continued)
CONDITIONREQUIREDACTION
COMPLETION
TIME B.Required Action and B.1 associated
Completion
Time of Condition A not met.One or more batteries with average electrolyte
temperature
of the representative
cells<60°F.OR One or more batteries with one or more battery cell parameters
not within Category C values.Declare associated
battery Immediately
SURVEILLANCE
REQUIREMENTS
SURVEILLANCE
SR 3.8.6.1 Verify battery cell parameters
meet Table 3.8.6-1 Category A limits.FREQUENCY 7 days (continued)
McGuire Units 1 and 2 3.8.6-2 Amendment Nos.184/166
SURVEILLANCE
REQUIREMENTS (continued)
SURVEILLANCE
SR 3.8.6.2 Verify battery cell parameters
meet Table 3.8.6-1 Category B limits.Battery Cell Parameters
3.8.6 FREQUENCY 92 days Once within 7 days after a battery discharge<110 V Once within 7 days after a battery overcharge
>150 V SR3.8.6.3Verify
average electrolyte
temperature
of representative
cells is60°F.92 days McGuire Units 1 and 2 3.8.6-3 Amendment Nos.184/166
Battery Cell Parameters
3.8.6 Table 3.8.6-1 (page 1 of 1)Battery Cell Parameters
Requirements
CATEGORY A: CATEGORY C: LIMITS FOR EACH CATEGORY B: ALLOWABLE DESIGNATED
LIMITS FOR EACH LIMITS FOR EACH PARAMETER PILOT CELL CONNECTED CELL CONNECTED CELL Electrolyte
Level>Minimum level>Minimum level Above top of plates, indication
mark, and indication
mark, and and not overflowing%inch above%inch above maximum level maximum level indication
mark(a)indication
mark(a)Float Voltage2.13 V2.13 V>2.07 V Specific Gravity(b)(c)1.2001.195 Not more than 0.020 below average of all AND connected cells or1.195 Average of all connected cells AND>1.205 Average of all connected cells1.195-(a)It is acceptable
for the electrolyte
level to temporarily
increase above the specified maximum during equalizing
charges provided it is not overflowing.(b)Corrected for electrolyte
temperature
and level.Level correction
is not required, however, when battery charging is<2 amps when on float charge.(c)A battery charging current of<2 amps when on float charge is acceptable
for meetingspecificgravity
limits following a battery recharge, for a maximum of 7 days.When charging current is used to satisfy specific gravity requirements, specific gravity of each connected cell shall be measured prior to expiration
of the 7 day allowance.
McGuire Units 1 and 2 3.8.6-4 Amendment Nos.184/166
Battery Cell Parameters
B 3.8.6 B 3.8 ELECTRICAL
POWER SYSTEMS B 3.8.6 Battery Cell Parameters
BASES BACKGROUND
This LCO delineates
the limits on electrolyte
temperature, level, float voltage, and specific gravity for the DC power source batteries.
A discussion
of these batteries and their OPERABILITY
requirements
is provided in the Bases for LCO 3.8.4,"DC Sources-Operating," and LCO 3.8.5,"DC Sources-Shutdown." APPLICABLE
The initial conditions
of Design Basis Accident (DBA)and transient SAFETY ANALYSES analyses in the UFSAR, Chapter 6 (Ref.1)and Chapter 15 (Ref.2), assume Engineered
Safety Feature systems are OPERABLE.The DC electrical
power system provides normal and emergency DC electrical
power for the diesel generators, emergency auxiliaries, and control and switching during all MODES of operation.
The OPERABILITYofthe DC subsystems
is consistent
with the initial assumptions
of the accident analyses and is based upon meeting the design basis of the unit.This includes maintaining
at least one train of DC sources OPERABLE during accident conditions, in the event of: a.An assumed loss of all offsite AC power or all onsite AC power;and b...'.A worst case single failure.Battery cell parameters
satisfy the Criterion 3 of 10 CFR 50.36 (Ref.3).LCO APPLICABILITY
Battery cell parameters
must remain within acceptable
limits to ensure availability
of the required DC power to shut down the reactor and maintain it in a safe condition after an anticipated
operational
occurrence
or a postulated
DBA.Electrolyte
limits are conservatively
established, allowing continued DC electrical
system function even with Category A and B limits not met.The battery cell parameters
are required solely for the support of the associated
DC electrical
power subsystems.
Therefore, battery electrolyte
is only required when the DC power source is required to be OPERABLE.Refer to the Applicability
discussion
in Bases for LeO 3.8.4 and LCO 3.8.5.McGuire Units 1 and 2 B 3.8.6-1 Revision No.0
BASES ACTIONS Battery Cell Parameters
B 3.8.6 A.1 , A.2, and A.3 With one or more cells in one or more batteries not within limits (i.e., Category A limits not met, Category B limits not met, or Category A and B limits not met)but within the Category C limits specified in Table 3.8.6-1 in the accompanying
LCO,thebattery
is degraded but there is still sufficient
capacity to perform the intended function.Therefore, the affected battery is not required to be considered
solely as a result of Category A or B limits not met and operation is permitted for a limited period.The pilot cell electrolyte
level and float voltage are required to be verified to meet the Category C limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (Required Action A.1).This check will provide a quick indication
of the status of the remainder of the battery cells.One hour provides time to inspect the electrolyte
level and to confirm the float voltage of the pilot cells.One hour is considered
a reasonable
amount of time to perform the required verification.
Verification
that the Category C limits are met (Required Action A.2)provides assurance that during the time needed to restore the parameters
to the Category A and B limits, the battery is still capable of performing
its intended function.A period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed to complete the initial verification
because specific gravity measurements
must be obtained for each connected cell.Taking into consideration
both the time required to perform the required verification
and the assurance that the battery cell parameters
are not severely degraded, this time is considered
reasonable.
The verification
is repeated at 7 day intervals until the parameters
are restored to Category A or B limits.This periodic verification
is consistent
with the normal Frequency of pilot cell Surveillances.
Continued operation is only permitted for 31 days before battery cell parameters
must be restored to within Category A and B limits.With the consideration
that, while battery capacity is degraded, sufficient
capacity exists to perform the intended function and to allow time to fully restore the battery cell parameters
to normal limits, this time is acceptable
prior to declaring the battery inoperable.
With one or more batteries with one or more battery cell parameters
outside the Category C limit for any connected cell, sufficient
capacity to supply the maximum expected load requirement
is not assured and the corresponding
DC electrical
power subsystem must be declared inoperable.
Additionally, other potentially
extreme conditions, such as not McGuire Units 1 and 2 B 3.8.6-2 Revision No.0
Battery Cell Parameters
B 3.8.6 BASES ACTIONS (continued)
completing
the Required Actions of Condition A within the required Completion
Time or average electrolyte
temperature
of representative
cells falling below 60°F, are also cause for immediately
declaring the associated
DC electrical
power subsystem inoperable.
SURVEILLANCE
SR 3.8.6.1 REQUIREMENTS
This SR verifies that Category A battery cell parameters
are consistent
with IEEE-450 (Ref.4), which recommends
regular battery inspections (at least one per month)including voltage, specific gravity, and electrolyte
temperature
of pilot cells.SR 3.8.6.2 The quarterly inspection
of specific gravity and voltage is consistent
with IEEE-450 (Ref.4).In addition, within 7 days of a battery discharge<110 V or a battery overcharge>
150 V, the battery must be demonstrated
to meet Category B limits.Transients, such as motor starting transients, which may momentarily
cause battery voltage to drop to::;110 V, do not constitute
a battery discharge provided the battery terminal voltage andfloatcurrent
return to pre-transient
values.This inspection
is also consistent
with IEEE-450 (Ref.4), which recommends
special inspections
following a severe discharge or overcharge, to ensure that no significant
degradation
of the battery occurs as a consequence
of such discharge or overcharge.
SR 3.8.6.3 This Surveillance
verification
that the average temperature
of representative
cells is 2::: 60°F, is consistent
with a recommendation
of IEEE-450 (Ref.4), that states that the temperature
of electrolytes
in representative
cells should be determined
on a quarterly basis.Lower than normal temperatures
act to inhibit or reduce battery capacity.This SR ensures that the operating temperatures
remain within an acceptable
operating range.This limit is based on manufacturer
recommendations.
The term"representative
cells" replaces the fixed number of"six connected cells", consistent
with the recommendations
of IEEE-450 (Ref.4)to provide a general guidance to the number of cells adequate to McGuire Units 1 and 2 B 3.8.6-3 Revision No.0
Battery Cell Parameters
B 3.8.6 BASES SURVEILLANCE
REQUIREMENTS (continued)
monitor the temperature
of the battery cells as an indicator of satisfactory
performance.
For some cases, the number of cells may be less than six, in other conditions, the number may be more.Table 3.8.6-1 This table delineates
the limits on electrolyte
level, float voltage, and specific gravity for three different categories.
The meaning of each category is discussed below.Category A defines the normal parameter limit for each designated
pilot cell in each battery.The cells selected as pilot cells are those whose temperature, voltage, and electrolyte
specific gravity approximate
the state of charge of the entire battery.The Category A limits specified for electrolyte
level are based on manufacturer
recommendations
and are consistent
with the guidance in IEEE-450 (Ref.4), with the extra 1/4 inch allowance above the high water level indication
for operatingmarginto account for temperatures
and charge effects.In addition to this allowance, footnote a to Table 3.8.6-1 permits the electrolyte
level to be above the specified maximum level during equalizing
charge, provided it is not overflowing.
These limits ensure that the plates suffer no physical damage, and that adequate electron transfer capability
is maintained
in the event of transient conditions.
IEEE-450 (Ref.4)recommends
that electrolyte
level readings should be made only after the battery has been at float charge for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.The Category A limit specified for float voltage is2.13 V per cell.This value is based on the recommendations
of IEEE-450 (Ref.4), which states that prolonged operation of cells<2.13 V can reduce the life expectancy
of cells.The Category A limit specified for specific gravity for each pilot cell is1.200 (0.015 below the manufacturer
fully charged nominal specific gravity or a battery charging current that had stabilized
at a low value).This value is characteristic
of a charged cell with adequate capacity.According to IEEE-450 (Ref.4), the specific gravity readings are based on a temperature
of 77°F (25°C).McGuire Units 1 and 2 B 3.8.6-4 Revision No.0
Battery Cell Parameters
B 3.8.6 BASES SURVEILLANCE
REQUIREMENTS (continued)
The specific gravity readings are corrected for actual electrolyte
temperature
and level.For each 3°F (1.67°C)above 77°F (25°C), 1 point (0.001)is added to the reading;1 point is subtracted
for each 3°F below 77°F.The specific gravity of the electrolyte
in a cell increases with a loss of water due to electrolysis
or evaporation.
Category B defines the normal parameter limits for each connected cell.The term IIconnected
celi ll excludes any battery cell that may be jumpered out.The Category B limits specified for electrolyte
level and float voltage are the same as those specified for Category A and have been discussed above.The Category B limit specified for specific gravity for each connected cell is1.195 (0.020 below the manufacturer
fully charged, nominal specific gravity)with the average of all connected cells>1.205 (0.010belowthe manufacturer
fully charged, nominal specific gravity).These values are based on manufacturer1s
recommendations.
The minimum specific gravity value required for each cell ensures that the effects of a highly charged or newly installed cell will not mask overall degradation
of the battery.Category C defines the limits for each connected cell.These values, although reduced, provide assurance that sufficient
capacity exists to perform the intended function and maintain a margin of safety.When any battery parameter is outside the Category C limits,theassurance
of sufficient
capacity described above no longer exists, and the battery must be declared inoperable.
The Category C limits specified for electrolyte
level (above the top of the plates and not overflowing)
ensure that the plates suffernophysical
damage and maintain adequate electron transfer capability.
The Category C limits for float voltage is based on IEEE-450 (Ref.4), which states that a cell voltage of 2.07 V or below, under float conditions
and not caused by elevated temperature
of the cell, indicates internal cell problems and may require cell replacement.
The Category C limit of average specific gravity1.195 is based on manufacturer
recommendations
(0.020 below the manufacturer
recommended
fully charged, nominal specific gravity).In addition to that limit, it is required that thespecificgravity
for each connected cell must be no less than 0.020belowthe average of all connected cells.This limit ensures that the effect of a highly charged or new cell does not mask overall degradation
of the battery.McGuire Units 1 and 2 B 3.8.6-5 Revision No.0
Battery Cell Parameters
B 3.8.6 BASES SURVEILLANCE
REQUIREMENTS (continued)
The footnotes to Table 3.8.6-1 are applicable
to Category A, B, and C specific gravity.Footnote (b)to Table 3.8.6-1 requires the above mentioned correction
for electrolyte
level and temperature, with the exception that level correction
is not required when battery charging current is<2 amps on float charge.This current provides, in general, an indication
of overall battery condition.
Because of specific gravity gradients that are produced during the recharging
process, delays of several days may occur while waiting for the specific gravity to stabilize.
A stabilized
charger current is an acceptable
alternative
to specific gravity measurement
for determining
the state of charge.This phenomenon
is discussed in IEEE-450 (Ref.4).Footnote (c)to Table 3.8.6-1 allows thefloatcharge
current to be used as an alternate to specific gravity for up to 7 days following a battery recharge.Within 7 days, each connected ceilis specific gravity must be measured to confirm the state of charge.Following a minor battery recharge (such as equalizing
charge that does not follow a deep discharge)
specific gravity gradients are not significant, and confirming
measurements
may be made in less than 7 days.The value of 2 amps used in footnote (b)and (c)is the nominal value for float current established
by the battery vendor as representing
a fully charged battery with an allowance for overall battery condition.
REFERENCES
1.UFSAR, Chapter 6.2.UFSAR, Chapter 15.3.10 CFR 50.36, Technical Specifications, (c)(2)(ii).
McGuire Units 1 and 2 B 3.8.6-6 Revision No.0
DUKE POWER MCGUIRE OPERATIONS
TRAINING I Objective#12 Each battery is sized to supply the continuous
emergency loads and momentary loads fed from its distribution
center (two DC buses which includes the two inverters and their panelboards), plus supply the loads of its sister distribution
center (two DC buses which includes the two inverters and their panelboards), if required, for a period of one hour.The basis for selecting aone-hourcapacity
is a conservative
time estimate for the restoration
of power to the battery chargers under the most adverse credible conditions.
This one-hour duty cycle capacity was assumed during the plant's safety analysis (documented
in the UFSAR)and is verified every 18 months during a battery service test.The minimum design ambient temperature
in the battery room is 60 of;hence the battery is sized based on its capacity at 60°F since the battery capacity would be greater at a higher temperature.
Since each battery is, electrically, in parallel with its battery charger, and the battery charger output voltage is slightly higher than the battery voltage, during the"floating charge";the battery charger actually supplies power to the respective
DC loads during normal operation.
However, the battery will automatically
assume those DC loads, without interruption, upon loss of its respective
battery charger or AC power source.Battery bus voltage is indicated by voltmeters
located on the 125 VDC vital control distribution
centers.The battery bus voltage is also monitored by under-voltage
relays, which alarm, on Annunciator
Alarm Pane11AD-11 (Electrical), when the battery bus voltage reaches 127 volts (at this voltage the battery is still capable of performing
its intended safety function).
2.3 125 VDC Vital Instrumentation
and Control Power System Distribution
Centers Each of the four distribution
centers (EVDA, EVDB, EVDC, and EVDD)receive power from a battery and/or a battery charger, and supplies power to two of the eight 125 VDC power panelboards
(1 EVDA, 1 EVDB, 1 EVDC, 1 EVDD, 2EVDA, 2EVDB, 2EVDC, and 2EVDD), and two of the eight static inverters (1 EVIA, 1 EVIB, 1 EVIC, 1 EVID, 2EVIA, 2EVIB, 2EVIC, and 2EVID).I Objective#13 I Either of the two same train-related
buses (EVDA and EVDC/Train"A" buses or EVDB and EVDD/Train"B" buses)can be tied together through their respective
bus tie breakers.This will allow two distribution
centers to be fed from one battery/battery charger combination.
This system is shared between the two units (Unit 1 and 2)and provides four normally independent
power channels for reactor control and instrumentation.
Three of the four channels will ensure that the overall system functional
capability
is maintained, comparable
to the original design standards for safe operation.
However, a loss of any two of these channel sources will result in a reactor trip or forced reactor shutdown (Technical
Specifications)
of both units (Unit 1 and 2).OP-MC-EL-EPL
FOR TRAINING PURPOSES ONL Y Page 25 of 73 REV.22
DUKE POWER 1.0 INTRODUCTION
MCGUIRE OPERATIONS
TRAINING 1.1.Purpose I Objective#1 I The 125 VDC and 120 VAC Vital Instrumentation
and Control Power System provides a reliable source of continuous
power for the safety related controls and instrumentation
required for plant start up, normal operation, and an orderly shutdown of each unit.1.2.General Description
125 VDC Vital Instrumentation
and Control Power System I Objective#3 The 125 VDC Vital Instrumentation
and Control Power System consists of five chargers, four 125 VDC batteries, four distribution
centers (with associated
breakers), and eight separate panelboards.
The system is designed to support a manual connection
of two distribution
centers (either EVDA and EVDC or EVDB and EVDD)during periods of battery maintenance.
The DC System is divided into four independent
and physically
separated load groups.With each load group comprised of the following:
one battery, one battery charger, one DC distribution
center, and two DC power panelboards.
This system is shared between the two units (Unit 1 ,and 2)and provides four normally independent
power channels for reactor control and instrumentation.
Three of the four channels will ensure that the overall system functional
capability
is maintained, comparable
to the original design standards for safe operation.
However, a loss of any two of these channel sources will result in a shutdown of both units (Unit 1 and 2).I Objective#4 I The following is a listing of typical loads that are powered from the 125 VDC Vital Instrumentation
and Control Power System Distribution
Centers (EVDA, EVDB, EVDC, and EVDD):*Auxiliary Safeguards
Cabinets Control Power*Turbine Trip*ETA and ETB Control Power*Diesel Generator Sequencers
Control Power*Miscellaneous
NV System Solenoids*Pressurizer
PORV Solenoids*Reactor Trip Switchgear
Control Power*600 V Load Centers ELXA, ELXB, ELXC, and ELXD Control Power*Power supplies to the Reactor Vessel Head Vents*Ventilation
Units Shunt Trip Coils*NCP UF-UV Monitor Panels OP-MC-EL-EPL
FOR TRAINING PURPOSES ONL Y Page 17 of 73
DUKE POWER MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours)NLO NLOR LPRO LPSO LOR3.03.02.02.0 2.0 OBJECTIVESNN LLL OBJECTIVELL P P 000 R S R R 0 0 1 State the purpose of the 125 VDC and 120 VAC VitalXX X X Instrumentation
and Control Power Systems.2 Draw a simplified
composite of the 125 VDC and 120 VACXX X X Vital Instrumentation
and Control Power Systems as provided in Training Drawing 7.2, Simplified
125 VDC and 120 VAC Vital Instrumentation
and Control Power Drawing.3 Provide a general description
of the 125 VDC Vital X X X X Instrumentation
and Control Power System.4 List the typicalloadspowered
from the 125 VDC Vital X X X X Instrumentation
and Control Power System Distribution
Centers.5 Provide a general description
of the 120 V AC VitalXX X X Instrumentation
and Control Power System.6 List the typical loads powered from the 120 V AC VitalXXX X Instrumentation
and Control Power System Power Panelboards.
7 Describe the basis for the sizing (loading)of the batteryXX X X charger associated
with the 125 VDC Vital Instrumentation
and Control Power System.8 Discuss the normal loading demands associated
with the 125XXX X VDC Battery Chargers for the Vital Instrumentation
and Control Power System.9 Describe any of the Kirk-Key Interlocks
associated
with the X X X X X 125 VDC Vital Instrumentation
and Control Power System and state the purpose of the Kirk-Key arrangement.
10 Explain how the Standby Battery Charger is used during anXX X X X equalizing
charge of a 125 VDC Battery for the Vital Instrumentation
and Control Power System.OP-MC-EL-EPL
FOR TRAINING PURPOSES ONL Y Page 5 of 73 REV.22
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#1----Group#2 KIA#003 AA2.02----------
Importance
Rating 2.8----Ability to deternline
and interpret the follovving
as they apply to the Dropped Control Rod: Signa!inputs to rod contra!system Proposed Question: SRO 82 Given the following Unit 1 initial conditions:
- Reactor power is at 400/0*Power range NIS indicate: o 400/0 (N4*1), 41%(N42), 41%(N43), 41%(N44)*Tave for each loop indicates:
o 567°F ('AI), 567°F ('BI), 568°F ('CI), 568°F ('DI)*Turbine power is at 481 MWe*Rod control is in automatic*Group demand counters and DRPI indicate Control Bank IDI at 140 steps.Control Bank IDI Rod L-12 drops fully into the core and the following conditions
now exist:*Power range NIS indicate: o 40%(N41), 40%(N42), 42%(N43), 38%(N44)*Tave for each loop indicates:
o 564°F ('AI), 564°F ('BI), 563°F ('CI), 564°F ('DI)*Turbine power is 478 MWe Assuming NO operator action, which ONE of the following describes the effect on the rod control system, and the technical specification
action required?A.Rods withdraw due to the Tave-Tref mismatch.Verify Shutdown Margin requirements
are met or initiate boration to ensure Shutdown Margin is met, to ensure accident analysis assumptions
remain valid.B.Rods withdraw due to the Power Range NIS Mismatch Rate signal.Verify Shutdown Margin requirements
are metorinitiate
boration to ensure Shutdown Margin is met, to ensure accident analysis assumptions
remain valid.Page 208 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 C.Rods withdraw due to Power Range NIS Mismatch Rate signal.Verify AFD requirements
are met to ensure that fuel design limits and hot channel factors are maintained
within limits.D.Rods withdraw due to the Tave-Tref mismatch.Verify AFD requirements
are met to ensure that fuel design limits and hot channel factors are maintained
within limits.Proposed Answer: A Explanation (Optional):
A.Correct.Tave deviation is higher than 1.5 degrees F and rods will withdraw.TS action is correct.B.Incorrect.
Power mismatch is not high enough to overcome the Tave mismatch, and power mismatch is based on rate of change with turbine power, which is minimal c.Incorrect.
Incorrect bases and also incorrect reason for rod withdrawal.
Plausible because power mismatch is an input and AFD would be a concern above 500/0 power D.Incorrect.
Incorrect basis but AFD would be a concern at higher power, as well as action required (>50%)OP-MC-IC-IRX, Rev 23 Technical Reference(s)(Attach if not previously
provided)----------
AP/14 Rev 10 AP-14 Basis Document Rev 6 TS 3.1.4 Proposed references
to be provided to applicants
during None examination:
Learning Objective:
OP-MC-IRX-Obj
5 (Note changes or attach parent)----Bank#X----Modified Bank#New Question Source: Question History: Last NRC Exam 2002 McGuire Page 209 of 260 Draft 7
ES-401 Question Cognitive Level: Sample Written Examination
Question Worksheet Memory or Fundamental
Knowledge Comprehension
or Analysis x Form ES-401-510 CFR Part 55 55.41 Content: 55.43 2 Comments: Stem not modified but distractors
all different from original KA met because inputs to rod control are the evaluated parameters.
SRO level because the effect of the failure has implications
in TS basis that the applicant must determine Page 210 of 260 Draft 7
DUKE ENERGY 1.0 INTRODUCTION
1.1.Purpose Objective#1 MCGUIRE OPERATIONS
TRAINING The Reactor Control System (I RX)allows the reactor to follow load changes automatically
between 15 to 100%power without a reactor trip, steam dump actuation, or pressure relief with the following load changes:*Step load increase or decrease of 1 0%*Ramp increase or decrease of 5%per minute 1.2.General Description
The system matches reactor power to turbine load by controlling
reactor coolant temperature (T avg).Reference temperature (Tret)is calculated
as a function of turbine load from turbine impulse pressure.As turbine load changes, Tret changes.When coolant temperature (T avg)differs from Tret, an error signal is produced.The rate of change of the difference
between reactor power and turbine power (power mismatch)is produced to provide an anticipatory
signal.The power mismatch signal can generate rod movement prior to a TavgITret mismatch.The two error signals, temperature
mismatch and power mismatch are summed to yield a rod speed and direction demand signal (combined error)which is sent to the Rod Control System.The Reactor Control System is not safety related.2.0 COMPONENT DESCRIPTION
2.1.Loop Average Temperature (T avg)T avg for each of the four loops is derived from narrow range (NR)hot and cold leg Resistance
Temperature
Detectors (RTD's).T-Th+Tc avg-2 This derived by averaging the loops three hot leg RTD's.T avg is used in calculating
the OPilT and OTilT setpoints and in the Feedwater Isolation circuit (P-4 and Lo-T avg).Each loop T avg is indicated on the control board (530-630 OF).Isolation amplifiers
are used to isolate protection
circuits from control circuit faults.OP-MC-IC-IRX
FOR TRAINING PURPOSES ONL Y Page 11 of 65 REV.23
DUKE ENERGY MCGUIRE OPERATIONS
TRAINING The Tref signal is sent to the Steam Dump Control System to determine the output of the Load Rejection Controller, the Tavg/Tref recorder on Control Board and to the Plant computer.Tref filter provides transient suppression
prior to comparing with T avg.2.5.Temperature
Mismatch Signal I Objective#5, 12 I Auctioneered
high T avg is compared to Tref and a temperature
mismatch signal is developed.
The summer output signal is then sent to the lower scale of Control Board bargraph indicator+/-15 of.If T avg>Tref a positive temperature
mismatch exists and rod insertion may be required.If T avg<Tref a negative temperature
mismatch exists and rod withdrawal
may be required.2.6.Power Mismatch Signal I Objective#6, 12 I The auctioneered
high reactor power circuit selects the highest of all power range instruments
for the output signal.The auctioneered
High Nuclear Power is compared to the Turbine Power (impulse pressure)in order to anticipate
changes in T avg.If reactor power and turbine power are changing at different rates, there will be an output error signal.A difference
between reactor power and turbine load will not generate a mismatch signal if neither signal is changing.Reactor power could be 60%and turbine load 40%steady state and there would not be a mismatch.Any Nuclear Power Channel removed from service should be defeated using the Power Mismatch Bypass switches.2.6.1.Derivative
Circuit This provides an output which is proportional
to the rate of change of the difference
in nuclear power and turbine power.The derivative
is really a rate-lag unit (rate comparator).
If the rate of change between nuclear power and turbine power is zero, the long term output of the derivative
will be zero, even if nuclear power does not equal turbine power.When a rate of change occurs, an output results.When the rate of change returns to zero, the output will decay to zero, but it will take several minutes.NOTE: If input error signal is not changing, derivative
circuit output would be zero.OP-MC-IC-IRX
FOR TRAINING PURPOSES ONL Y Page 17 of 65 REV.23
DUKE ENERGY MCGUIRE OPERATIONS
TRAINING Power mismatch signal causes improved response (quicker)of output signal resulting in faster reaction of rod movement.It dominates initially on changing power mismatch signals.Temperature
mismatch signal dominates during any slow load increases/or
decreases.
During a rapid power mismatch transient, the temperature
mismatch signal will eventually
become the main or dominant rod movement signal after power mismatch change has subsided.MINIMUM MAXIM UM PROPORTIONAL
ROD SPEED ROD SPEED BAND BAND ROD SPEED STEPS/MIN.
DEADBAND 72 LOCKUP WITHDRAWAL
8-5-4-3-2-1.5-1 8 72 1.5 2 INSERTION 345 COMBINED ERROR SIGNAL OF (TEMP MISMATCH)+(POWER MISMATCH)Polarity of the Combined Error Signal determines
rod direction movement.If the signal is positive rods step in.*T avg>T ref*Nuclear power increasing
at a faster rate than turbine power.*Turbine power decreasing
at a faster rate than nuclear power.If the signal is negative rods ,step out.*T avg<Tref*Nuclear power decreasing
at a faster rate than turbine power.*Turbine power increasing
at a faster rate than nuclear power.OP-MC-IC-IRX
FOR TRAINING PURPOSES ONL Y Page 23 of 65 REV.23
MNS AP/11 A/5500/14 UNITl ROD CONTROL MALFUNCTION
Enclosure1-Page 2 of 12 Response To A Dropped Rod PAGE NO.7 of 44 Rev.10 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 6.Check QPTR (Tech Spec 3.2.4)-WITHIN TECH SPEC LIMITS.7.REFER TO Tech Specs:*Tech Spec 3.1.4 (Rod Group Alignment Limits)*Ensure shutdown margin calculation
performed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.Reduce reactor power as required by Tech Specs as follows: a.Do not move rods until IAE determines
rod movement is available.
b.Borate as required during power reduction to maintain T-Ave at T-Ref.c.Monitor AFD during load reduction.
d.IF AT ANY TIME AFD reaches Tech Spec limit AND reactor power is greater than THEN:_1)Trip Reactor._2)GO TO EP/1/A/5000/E-0 (Reactor Trip or Safety Injection).
e.Reduce load PER one of the following procedures:
- OP/1/A/61 001003 (Controlling
Procedure For Unit Operation), Enclosure 4.2 (Power Reduction)
OR*AP/1/A/5500104 (Rapid Downpower).
AP/1 and 21A15500/014 (Rod Control Malfunction)
Encl.1-STEP 7: PURPOSE: This step is an evaluation
of Tech Spec requirements
for Rod Group Alignment Limits Tech Spec 3.1.4 and the action requirement
for determining
Shutdown Margin with an untrippable
or immovable control rod T.S.3.1.4, action B.2.1.1.DISCUSSION:
These Tech Spec items are listed to ensure the Control Room SRO evaluates the requirements
for Rod Group Alignment Limits and Shutdown Margin when a dropped rod has occurred and complies with the appropriate
action.A SDM calculation ,must be completed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since a rod that's already inserted is not available to supply shutdown margin.Encl.1-STEP 8: PURPOSE: To perform a plant shutdown versus retrieving
a dropped rod if less than 5%power.DISCUSSION:
With the unit being in Mode 1, there is no riskofwithdrawing
a dropped control rod with the resulting power increase causing a mode change.With the unit in Mode 2, the risk of an increase in reactor power and mode change are possible when retrieving
a dropped control rod.Other factors to consider when in Mode 2 that support a unit shutdown are:*At power levels below 5%rated thermal power, the turbine is not on line and changes to T-Ave will behandledby steam dumps which does not allow for fine temperature
control,*With the unit in a shutdown condition, problems related to xenon and temperature
changes will not have to be addressed,*With the turbine not on line, the unit status allows for the rod control problem to be corrected without having the unit at risk.*Prevents recriticality
during dropped rod retrieval should the core become sub critical due to the rod drop.This step is consistent
with the guidance given in response to industry event OEDB 90-002761 (SER 90-15).In that event, Vogtle1 dropped several rods during physics testing, and withdrew rods to get back critical.This resulted in bypassing the carefully controlled
evolution of taking the reactor critical.The appropriate
response should have been to trip the reactor or drive the other Page 15 of 57 Rev 6
Rod Group Alignment Limits 3.1.4 3.1 REACTIVITY
CONTROL SYSTEMS 3.1.4 Rod Group Alignment Limits LCO 3.1.4 All shutdown and control rods shall be OPERABLE, with all individual
indicated rod positions within 12 steps of their group step counter demand position.APPLICABILITY:
MODES 1 and 2.ACTIONS CONDITION REQUIRED ACTION COMPLETION
TIME A.One or more rod(s)untrippable.
A.1.1 Verify SDM is within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> limit specified in the COLR.A.1.2 Initiate boration to restore SDM to within limit.A.2 Be in MODE 3.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 6 hours (continued)
McGuire Units 1 and 2 3.1.4-1 Amendment Nos.184/166
CONDITION B.REQUIRED ACTION B.1 Restore rod to within alignment limits.Rod Group Alignment Limits 3.1.4 COMPLETION
TIME 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OR B.2.1.2Initiate
boration to restore 80M to within limit.B.2.2 Reduce THERMAL POWER to75%RTP.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 hour 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B.2.3 Verify 80M is within the Once per limit specified in the COLR.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B.2.4 Perform 8R 3.2.1.1.B.2.5 Perform 8R 3.2.2.1.72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 72 hours B.2.6 Re-evaluate
safety 5 days analyses and confirm results remain valid for duration of operation under these conditions.(continued)
McGuire Units 1 and 2 3.1.4-2 Amendment Nos.184/166
Rod Group Alignment Limits 3.1.4 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION
TIME C.Required Action and C.1 Be in MODE 3.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated
Completion
Time of Condition B not met.D.More than one rod not 0.1.1 Verify SDM is within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within alignment limit.limit specified in the COLR.OR 0.1.2 Initiateborationto
restore required SDM to within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> limit.AND 0.2 Be in MODE 3.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> SURVEILLANCE
REQUIREMENTS
SURVEILLANCE
SR 3.1.4.1 Verify individual
rod positions within alignment limit.FREQUENCY 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter
when the rod position deviation monitor is inoperable (continued)
McGuire Units 1 and 2 3.1.4-3 Amendment Nos.184/166
Rod Group Alignment Limits 3.1.4 SURVEILLANCE
REQUIREMENTS (continued)
SURVEILLANCE
FREQUENCYSR3.1.4.2 Verify rod freedom of movement (trippability)
by moving 92 days each rod not fully inserted in the core10 steps in either direction.
OR"'*Prior to entering MODE 3 upon Unit 1 startup>following the Unit 1 end of Cycle 13 refueling outage.I SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn position, is2.2 seconds from the beginning of decay of stationary
gripper coil voltage to dashpot entry, with: a.T avg551°F;and b.All reactor coolant pumps operating.
- One time change applicable
to Unit 1 only.Prior to reactor criticality
after each removal of the reactor head McGuire Units 1 and 2 3.1.4-4 Amendment Nos.186 (Unit 1)167 (Unit 2)
MNS AP/1/A/5500/14 UNITl ROD CONTROL MALFUNCTION
Enclosure1-Page 1 of 12 Response To A Dropped Rod PAGE NO.6 of 44 Rev.10 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 1.Announce occurrence
on paging system.2.Dispatch rod control system qualified IAE to correct cause of dropped rod.3.Check uROD CONTROL URGENT FAILURE u alarm (1AD-2, A-10)-DARK.Perform the following:
a.Do not move control rods while the"ROD CONTROL URGENT FAILURE" alarm is lit, unless instructed
by IAE.b.IF AT ANY TIME IAE desires to reset"ROD CONTROL URGENT FAILURE" alarm, THEN depress the"ROD CONTROL ALARM RESET" pushbutton.
c.IF AT ANY TIME while in this procedure a runback occurs AND no rods will move, THEN perform the following:
_1)Trip Reactor._2)GO TO EP/1/A/5000/E-O (Reactor Trip or Safety Injection).
4.Use OAC point M1 P1385 (Reactor Thermal Power, Best Estimate), to determine reactor power in subsequent
steps.5.Check AFD (Tech Spec 3.2.3)-WITHIN TECH SPEC LIMITS.IF reactor power greater than 50%, THEN: a.Trip reactor.b.GO TO EP/1/A/5000/E-O (Reactor Trip or Safety Injection).
AP/1 and 21A15500/014 (Rod Control Malfunction)
If the"Rod Control Urgent Failure" (1 AD-2, A-1 0)alarm is present, the alarm is being generated by a failure in either the Logic or Power Cabinets.Control rods should not be moved until the problem has been identified
and evaluated.
If an attempt is made to move control rods in the individual"Bank" mode before the problem is identified, a dropped rod could result.This may occur from incorrect operation of the CRDM if the failure is in the Slave Cycler for the affected rod.If a problem has occurred in a Power Cabinet, dropped rods may result if the alarm is reset (using the"Rod Control Alarm Reset" pushbutton)
before the cause of the urgent alarm is identified
and repaired.The two methods to control reactivity
on a short-term (transient)
basis are by adjusting turbine load or moving control rods.If a runback occurs, adjusting turbine load is not an option for the Operator.If this occurs while rods can't be moved, there remains no quick reactivity
control method for the Operator to control reactor powerlNC temperature, and so the conservative
thing to do is trip the reactor.Encl.1-STEP 4: PURPOSE: The step provides guidance to the operator to use the OAC point for Thermal Power Best Estimate for making procedural
decisions based on power level.DISCUSSION:
Since operators typically use the OAC program that monitors the power, AFD and QPTR parameters
for each quadrant, these indications
may change significantly
from their normal indication
with a dropped control rod.It is importantthatthe operator monitor Thermal Power Best Estimate (OAC point M1 P1385)which takes in to account all parameters
of reactor power.Thermal Power Best Estimate uses heat transfer calculations
and not excore nuclear instrumentation
inputs.Thermal Power Best Estimate indication
will be used to determine if the unit should be shutdown or remain in operation based on power level.Encl.1-STEP 5: PURPOSE: This step is a check of AFD within Tech Spec limits since a dropped rod (especially
the case where the rod is misaligned
more than 50 steps below its'associated
group)can affect AFD.Page 13 of 57 Rev 6
AP/1 and 21A15500/014 (Rod Control Malfunction)
DISCUSSION:
Above 500/0 Rated Thermal Power, limits on AFD (variable from 50-100%power)are defined by Tech Specs (limits are found in Core Operating Limits Report).The limits on AFD are used to limit the amount of axial power distribution
to either the top or bottom of the core.Limiting the AFD skewing over time minimizes xenon skewing and limits excessive power distributions
that could potentially
damage the fuel.The limit ensures power distribution
remains consistent
with the design values used in the safety analysis.The limit provides a margin of protection
for both DNB and linear heat generation
rate, which contribute
to excessive power peaks.The guidance to trip the reactor if the limits are exceeded above 500/0 power is a conservative
action to take considering
that power reductions
without rod movement cause a dramatic shift toward a positive AFD and so a positive shift would cause an AFD that's out of limit in the positive direction to get even more out of spec.In either case, positive or negative, trying to restore AFD within its'limits within 30 minutes could be operationally
difficult without use of control rods.Encl.1-STEP 6: PURPOSE: Ensure compliance
with Tech Spec requirements
for aPTR.DISCUSSION:
A dropped control rod could significantly
affect aPTR and require a unit power reduction to comply with ITS.Above 50%Rated Thermal Power, limits on AFD (variable from 50-1 00%power)and aPTR ($.1.02)are defined by Tech Specs (limits for AFD are found in Core Operating Limits Report).The limit on aPTR ensures the radial powerdistributionremains
consistent
with the design values used in the safety analysis.The aPTR limit of 1.02 provides a margin of protection
for both DNB and linear heat generation
rate, which contribute
to excessive power peaks.The guidance to reduce reactor power is provided by the operating procedure for power reduction or by AP/4 (Rapid Downpower)
per the applicable
time requirements
of each Tech Spec.Since"no rod motion" is directed untillAE determines
it's available, direction is given to accomplish
the power reduction with boron to maintain T-ave at T-ref.If a power reduction is required, direction is given to monitor AFD, since it will tend to go positive on the shutdown, and to trip the reactor if it reaches its limit before getting to 50%power, since it will only get worse before it gets better.REFERENCES:
ITS 3.2.4 Page 14 of 57 Rev 6
DUKE ENERGY MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours)NLO NLOR LPRO LPSO LOR N/A 1.51.51.5 1.5 OBJECTIVES
N N LLL OBJECTIVE L L P P 0 0 0 R S R R 0 0 1 Explain the purpose of the Reactor Control System (IRX).XXX 2 Discuss the rod speedprogramfor
both rod insertion and X X X withdrawal
as per Drawing 7.4.3 Sketch the IRX block diagram, including all input and output X X X signals, per Drawing 7.6.4 Describe how the Tret program is generated, based on turbine X X X impulse pressure, including minimum and maximum values of Tret.5 Describe how the Temperature
Mismatch signal is developed X X X X and used for rod movements.
6 Describe how the Power Mismatch signal is developed and X X X X used for rod movements.
7 Explain how the Combined Error signal is used to develop rodXXXX speed and direction signals.8 State all rod speeds for both automatic and manual operation.XX X 9 Describe all interlocks
affecting rod withdrawal
to include X XXX setpoints, logic and mode of operation that is affected (Automatic
or Manual).10 Describe the system operation during transients.XXXX 11 Describe the system operation and operator response toXX X X various failed input signals.OP-MC-IC-IRX
FOR TRAINING PURPOSES ONL Y Page 5 of 65 REV.23
I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#1----Group#2 KIA#033 AA2.05----------
Importance
Rating 3.1----Ability to deterrnine
and interpret the follovving
as they apply to the Loss of!nterrnediate
Range Nuclear lnstrurnentation:
Nature of abnorrnality, frOtll rapid survey of control room data Proposed Question: SRO 83 Given the following:*A reactor startup is in progress.*SR Channel N-31 indicates 2X1 0 3 CPS.*SR Channel N-32 indicates 2X1 0 3 CPS.*IR Channel N-35 indicates 3.0X1 0-11 amps.*IR Channel N-36 indicates 9.0X1 0-11 amps.Which ONE (1)of the following describes (1)the existing plant condition, and (2)the action required in accordance
with AP/16, Malfunction
of Nuclear Instrumentation, and Technical Specifications?
A.(1)N-36 is undercompensated;
(2)maintain power stable until N-36 is repaired.B.(1)N-35 is undercompensated;
(2)maintain power stable until N-35 is repaired.C.(1)N-36 is undercompensated;
(2)Raise power to>P-1 0 or place the unit in Mode 3 until N-36 is repaired.D.(1)N-35 is undercompensated;
(2)Raise power to>P-1 0 or place the unit in Mode 3 until N-35 is repaired.Proposed Answer: A Explanation (Optional):
A.Correct.N-36 is reading approximately
0.7 decades too high for the SR counts displayed, therefore undercompensated.
AP/16 requires no positive reactivity
additions.
TS requires>P-1 0 or<P-6 Page 211 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 B.Incorrect.
Wrong NI is undercompensated.
N-35 reads correctly.
Plausible if applicant confuses overlap and indication
for IR Nls C.Incorrect.
Correct NI but incorrect action taken.Mode 3 entry is not required for the given conditions, and the AP says no positive reactivity
additions are allowed, so>P-10 is incorrect D.Incorrect.
Incorrect NI, Incorrect action taken.See A, B, C above AP/16 case 2 Technical Reference(s)(Attach if not previously
provided)----------
TS 3.3.1 IC-ENB Rev 26 Proposed references
to be provided to applicants
during None examination:
Learning Objective:
IC-ENB-Obj
7&19 (Note changes or attach parent)----Bank#Modified Bank X#New Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge Comprehension
or Analysis X 10 CFR Part 55 55.41 Content: 55.43 2,5 Comments: Modified from VC Summer 2007 NRC Exam KA is met because the applicant must determine the nature of the failure based on given indications, and SRO level because appropriate
TS action for plant conditions
is required knowledge at SRO level Page 212 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Given the following plant conditions:*A reactor startup is in progress.*SR Channel N-31 indicates 7X1 0 3 CPS.*SR Channel N-32 indicates 7X1 0 3 CPS.*IR Channel N-35 indicates 8.7X1 0-60/0 power.*IR Channel N-36 indicates 6.0X1 0-60/0 power.Which ONE (1)of the following describes (1)the existing plant condition, (2)the status of P-6, and (3)the action required in accordance
with AOP-401.8, Intermediate
Range Channel Failure?A.(1)N-36 is undercompensated;
(2)P-6 should NOT be satisfied;
(3)maintain power stable until N-36 is repaired.B.(1)N-36 is overcompensated;
(2)P-6 should be satisfied;
(3)maintain power stable until N-36 is repaired.C.(1)N-36 is undercompensated;
(2)P-6 should be satisfied;
(3)place the unit in, Mode 3 until N-36 is repaired.D.(1)N-36 is overcompensated;
(2)P-6 should NOT be satisfied;
(3)place the unit in Mode 3 until N-36 is repaired.Ans.B Page 213 of 260 Draft 7
DUKE POWER MCGUIRE OPERATIONS
TRAINING Control Power Fuses-Overcurrent
protection
for control signal circuit transformers.
Control power supplies the lights on the drawer and 118 VAC to the bistable relay drivers to the plant relays.(High flux at shutdown alarm and SR high level trip).This is true for the IR and PR drawers/circuits
also.NOTE (Reference
Figure 7.21): If either instrument
or control power fuses are removed, the bistables will trip.Level Trip Bypass will prevent bistable trip for Instrument
Power fuses only.I Objective#1 0 I Level Trip Switch-Two position switch: Normal-Switch Inactive;Bypass-Enables Operation Selector Switch for test and calibration;
Provides AC signal to prevent Rx trip signal during testing.Operation Selector Switch-Eight position switch enabled by Level Trip Switch to'Bypass'position.Channel On Test lamp lights when not in Normal.Normal-Switch Inactive;Six Test Positions with Preset cps test values;Level Adjust-Level Adjust Potentiometer
in circuit.Level Adjust Potentiometer
-Adjustable
test signal into level amp.-Enables adjustment
of the trip level of various bistables.
I Objective#10 I High Flux at Shutdown Switch-Two position switch.Normal-allows circuit to provide"High Flux at Shutdown" and"Containment
Evacuation" alarm when setpoint is exceeded;Block-used
during startup-Blocks High Flux at Shutdown Alarm and Containment
Evacuation
Alarm.2.2 Intermediate
Range 2.2.1 Intermediate
Range Detectors I Objective#6 I Reference Figure 7.6.Both intermediate
range channels use compensated
ion chambers to determine reactor power.These detectors are located just above the source range detectors in the same housing.The compensated
ion chamber (CIC)uses two concentric
Nitrogen gas filled, volumes: the"outer" is sensitive to both neutrons and gamma (boron lined);the"inner" sensitive only to gamma.As the two volumes are mounted concentrically
in one unit, both are in essentially
the same radiation field.By placing a negative potential on the inner lead, the gamma signal generated in the inner volume is made to compensate
or cancel out the gamma signal generated in the outer volume.Since the two volumes can not be manufactured
exactly the same size, the high voltage to the center electrode is variable to adjust the sensitivity
of the inner volume.Operating in the recombination
region, a change in inner volume detector voltage will vary the gamma current for a given flux level.The outer volume operates in the ion chamber region where all the ion pairs are collected.
I Objective#5 I Gamma radiation becomes a smaller percentage
of the detector interactions
as power increases and becomes insignificant
after 10-9 amps (first two decades).Above this power level gamma compensation
is no longer required for accurate indication.
OP-MC-IC-ENB
FOR TRAINING PURPOSES ONL Y Page 21 of 129 REV.26
DUKE POWER MCGUIRE OPERATIONS
TRAINING 2.2.2 Over Compensation
And Under Compensation
I Objective#7 I Reference Figure 7.7.With the inner chamber voltage set properly, inner chamber gamma current will exactly match outer chamber gamma current and the two will cancel leaving only the neutron current.With inner chamber voltage set too high, inner chamber current will exceed outer chamber gamma current canceling all gamma current plus some of the neutron current.This is"over-compensation".
The following are consequences
of over-compensation:
- The indicated power level will read lower than the actual power level.*The intermediate
range instrument
will"come on scale" at a higher source range level producing less overlap between the two ranges.*During startup, the P-6 permissive
will be received later, at a higher actual neutron flux level and the source range will be closer to the 10 5 cps, Hi Level Trip setpoint.*After a Reactor Trip, power will decay to the P-6 reset sooner than normal.*Initially, indicated SUR will be higher than actual SUR.The effects of improper compensation
are much more pronounced
at low power and become a non-factor
prior to taking critical data at 1 0-8 amps.With inner chamber voltage set too low, inner chamber current will be less than outer chamber gamma current, canceling only a portion of the gamma current.This is"under-compensation".
The following are consequences
of under-compensation:
- The indicated power level will read higher than the actual power level.*The intermediate
range instrument
will"come on scale" at a lower source range level producing more overlap between the two ranges.*During startup, the P-6 permissive
will be receivedearlier,at
a lower actual neutron flux level.*After a Reactor Trip, power will decay to the P-6 reset later than normal and may prevent automatic re-energizing
of the source range detectors.
- Initially, indicated SUR will be lower than actual SUR.2.2.3 Intermediate
Range Circuitry I Objective#4 I Reference Figure 7.8.The Intermediate
Range should normally start to indicate power at a Source Range power level of 10 3 cps and the Source Range should be blocked by the time level is 10 4 cps and Intermediate
level is at 10-10 amps.The indicating
range for the Intermediate
Range instrument
is 10-11 to 10-3 amps, which overlaps the entire power range.The current flow from the intermediate
range detectors is too low to be used directly for control purposes so the 9utput feeds a log level amplifier (log amp)for conversion
to a usable voltage.The log level amplifier also converts the detector signal to a logarithmic
output and drives the bistables, indicators
and other circuits.OP-MC-IC-ENB
FOR TRAINING PURPOSES ONL Y Page 23 of 129 REV.26
DUKE POWER 7.7 Over and Undercompensation
(03/20/97)
MCGUIRE OPERATIONS
TRAINING co ,...(((,0 ,...en w::J C\lZ ,...-:EZo 00::J:I: en a: coOo c:(w a: a: (,OwLL c:(w:E o:ri=o OP-MC-IC-ENB
FOR TRAINING PURPOSES ONL Y Page 91 of 129
DUKE POWER MCGUIRE OPERA TIONS TRAINING 7.2 Operating Ranges (01/09/02)
RANGES OF OPERATION SR IR PR (CPS)(AMPS)(0/0 PWR)WR (0/0 PWR)10 200 100 TRIP C1 P10..----------------------
10-5 10-3
_C2_10-4 109%PR 1030/0 PRlIR 25°k IR20%PR 100/0 OP-MC-IC-ENB
FOR TRAINING PURPOSES ONL Y Page 81 of 12926
RTS Instrumentation
3.3.1 3.3 INSTRUMENTATION
3.3.1 Reactor Trip System (RTS)Instrumentation
LCO 3.3.1 The RTS instrumentation
for each Function in Table 3.3.1-1 shall be OPERABLE.APPLICABILITY:
According to Table 3.3.1-1.ACTIONS----------------------------------------------------------NOTE----------------------------------------------------------
Separate Condition entry is allowed for each Function.CONDITION REQUIRED ACTION COMPLETION
TIME A.One or more Functions A.1 Enter the Condition Immediately
with one or more referenced
in Table 3.3.1-1 required channels for the channel(s).
B.One Manual Reactor B.1 Restore channel to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Trip channel inoperable.
OPERABLE status.OR B.2 Be in MODE 3.54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> C.One channel or train C.1 Restore channel or train to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable.
OPERABLE status.OR C.2 Open reactor trip breakers 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> (RTBs).(continued)
McGuire Units 1 and 2 3.3.1-1 Amendment Nos.184/166
Table 3.3.1-1 (page 1 of 7)Reactor Trip System Instrumentation
RTS Instrumentation
3.3.1 APPLICABLE
MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE
ALLOWABLE TRIP FUNCTION CONDITIONS
CHANNELS CONDITIONS
REQUIREMENTS
VALUE SETPOINT 1.Manual Reactor Trip 1,2 2 B SR 3.3.1.14 NA NA 3(a), 4(a), 5(a)2 C SR 3.3.1.14 NA NA 2.Power Range Neutron Flux a.High 1,2 4 D SR 3.3.1.1 oS.1100/0 RTP 1090/0 RTP SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 b.Low 1(b),2 4 E SR 3.3.1.1 oS.260/0 RTP 250/0 RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 3.Power Range Neutron Flux Rate High Positive Rate 1,2 4 D SR 3.3.1.7 oS.5.50/0 RTP 50/0 RTP SR 3.3.1.11 with time with time constant constant 2: 2 sec 2: 2 sec 4.Intermediate
Range 1(b),2(c)2 F,G SR 3.3.1.1 oS.30%RTP 250/0 RTP Neutron Flux SR 3.3.1.8 SR 3.3.1.11 2(d)2 H SR 3.3.1.1 oS.300/0 RTP 25%RTP SR 3.3.1.8 SR 3.3.1.11 5.Source Range 2(d)2 I,J SR 3.3.1.1 oS.1.3 E5 cps 1.0 E5 cps Neutron Flux SR 3.3.1.8 SR 3.3.1.11 3(a), 4(a), 5(a)2 J,K SR 3.3.1.1 oS.1.3 E5 1.0 E5 SR 3.3.1.7 cps cps SR 3.3.1.11 3(e), 4(e), 5(e)L SR 3.3.1.1 N/A N/A SR 3.3.1.11 (continued)(a)With Reactor Trip Breakers (RTBs)closed and Rod Control System capable of rod withdrawal.(b)Belowthe P-10 (Power Range Neutron Flux)interlocks.(c)Above the P-6 (Intermediate
Range Neutron Flux)interlocks.(d)Belowthe P-6 (Intermediate
Range Neutron Flux)interlocks.(e)With the RTBs open.In this condition, source range Function does not provide reactor trip but does provide indication.
McGuire Units 1 and 2 3.3.1-14 Amendment Nos.194/175
RTS Instrumentation
3.3.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION
TIME E.One channel inoperable.
NOTE-------------------
One channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance
testing.---------------------------------------------
E.1 Place channel in trip.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> OR E.2 Be in MODE 3.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> F.THERMAL POWER F.1 Reduce THERMAL 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s>P-6 and<P-10, one POWER to<P-6.Intermediate
Range Neutron Flux channel OR inoperable.
F.2 Increase THERMAL 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> POWER to>P-10.------------------NOTE----------------
Limited boron concentration
changes associated
with RCS inventory control or limited plant temperature
changes are allowed.--------------------------------------------
G.THERMAL POWER G.1 Suspend operations
Immediately
>P-6 and<P-1 0, two involving positive reactivity
Intermediate
Range additions.
Neutron Flux channels inoperable.
ANDG.2Reduce THERMAL 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> POWER to<P-6.H.THERMAL POWER H.1 Restore channel(s)
to Prior to increasing
<P-6, one or two OPERABLE status.THERMAL POWER Intermediate
Range to>P-6 Neutron Flux channels inoperable.
(continued)
McGuire Units 1 and 2 3.3.1-3 Amendment Nos.216/197
DUKE POWER MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours)NLO NLOR LPRO LPSO LOR 2.0 3.0 3.0 2.0 OBJECTIVES
N N L L L OBJECTIVELL P P 000 R S R R 0 0 1 State the purpose of the Nuclear Instrumentation
System.XX X 2 Explain why it is necessary to use three ranges of Excore X X X Nuclear Instrumentation.
3 Explain the operation of the detector used in each range of X X X instrumentation.
4 Sketch the outputs of each range of Nuclear Instrumentation, X X X to include all indication, control and protective
circuits.5 Explain why gamma compensation
is necessary in the Source X X X Range and Intermediate
Range but not in the Power Range.6 Describe the methods of gamma compensation
used by the X X X Source and Intermediate
Ranges.7 Describe the effects of'over'and'under'compensation
onXX X the Intermediate
Range.8 Explain the functions of thecontrolswitches
for each range ofXX X X Nuclear Instrumentation.
9 Concerning
the channel current comparator
and detector current comparator:
- Explain the function of each.X X X*List the alarm setpoints for each.X XXX 10 Explain the functions of all related bypass and block switches X XXX on the Nuclear Instrumentation
miscellaneous
panels.11 List the Reactor Trips associated
with the Nuclear X XXX Instrumentation
System.(Include setpoints, logic and interlocks)
OP-MC-IC-ENB
FOR TRAINING PURPOSES ONL Y Page 7 of 129 REV.26
DUKE POWER MCGUIRE OPERATIONS
TRAINING 12 List the Protection
and Control Interlocks (Ps and Cs)XXX X associated
with the Nuclear Instrumentation
System.(Include setpoints and logic)13 State the purpose of the Wide Range Neutron DetectionXXX System.14 Concerning
the Wide Range Neutron Detection System:*Describe the operation.XXX*Describe the indications
and controls.XXXX 15 State the purpose of the Gamma-Metrics
Shutdown Monitor X X X System.16 Concerning
the Gamma-Metrics
Shutdown Monitor System:*Describe the operation.XXX*Describe the alarms, indications
and controls.XXX X 17 Determine the validity of indicated reactor power usingXXX X alternate indications
of power level.18 Describe the Source Range instrumentation
response forXXXX voiding in the core and downcomer region.19 Concerning
the Technical Specifications
related to the Nuclear Instrumentation
System;*Given the LCO title, state the LCO (including
any COLRXX X values)and applicability.
- For any LCO's that have action required within one hour, XXX state the action.*Given a set of parameter values or system conditions,XX X determine if any Tech Spec LCO's is(are)not met and any action(s)required within one hour.*Given a set of plant parameters
or system conditions
and the appropriate
Tech Specs, determine required action(s).XX X*Discuss the basis for a given Tech Spec LCO or Safety X*Limit.*SRO Only OP-MC-IC-ENB
FOR TRAINING PURPOSES ONL Y Page 9 of 129 REV.26
I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I
ES-401 Sample Written Examination
Question Worksheet Form ES-40 1-5 4.5 SRO RO
LI059 G2.2.38 Level Tier#Group#KIA#Importance
Rating Examination
Outlinereference:
Equipment Control: Knowledge of conditions
and limitations
in the facility license, Proposed Question: SRO 84 Given the following:
Turbine Building Sump to RC Radiation Monitor, EMF-31, is discovered
to have an alarm setpoint that is set ONE decade higher than required.Which ONE of the following describes the impact of this condition?
The dose or dose commitment
to members of the public may exceed the requirements
of 1 OCFR50 of....(A.1.5 mrem whole body dose in a calendar quarter.B.5 mrem whole body dose in a calendar quarter.C.1.5 mrem wholebodydose in a calendar year.D.5 mrem whole body dose in a calendar year.Proposed Answer: A Explanation (Optional):
A.Correct.This is a memory item.Below options are plausible because the numbers supplied are all part of the SLC B.Incorrect.
5 mrem is Organ Dose allowed for a calendar quarter C.Incorrect.
Allowed WB dose for a calendar year is 3 mrem D.Incorrect.
5 mrem is Organ Dose allowed for a calendar quarter.SLC 16.11.3, Rev 0 (Technical Reference(s)(Attach if not previously
provided)-----------
Page 214 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Proposed references
to be provided to applicants
during None examination:
Learning Objective:
WE-RLR Obj 6 (Note changes or attach parent)----Bank#Modified Bank#New X----Question Source: Question History: Last NRC Exam---------------
Question Cognitive Level: Memory or Fundamental
Knowledge X Comprehension
or Analysis 10 CFR Part 55 55.41 Content: 55.43 1,2,4 Comments: KA is matched because 10CFR50 requirements
for radioactive
release are limitations
in the facility license.SRO knowledge because the item requires knowledge of SLC (TRM)conditions
that will require action by the SRO Page 215 of 260 Draft 7
(Dose-Liquid Effluents 16.11.3 16.11 RADIOLOGICAL
EFFLUENT CONTROLS 16.11.3 Dose-Liquid Effluents COMMITMENT
APPLICABILITY
The dose or dose commitment
to a MEMBER OF THE PUBLIC from radioactive
materials in liquid effluents released from each unit to UNRESTRICTED
AREAS (see Figure16.11.1-1)
shall be limited: a.During any calendar quarter, to S 1.5 mrem to the total body and to S 5 mrem to any organ, and b.During any calendar year, to S 3 mrem to the total body and to S 10 mrem to any organ.At all times.{REMEDIAL ACTIONS---------------------------------------------------------NOTES-----------------------------------------------------
Enter applicable
Conditions
and Required Actions of SLC 16.11.12,"Total Dose," when the limits of this SLC are exceeded by twice the specified limit.A.CONDITION Calculated
dose from release of radioactive
materials in liquid effluents exceeding above limits.REQUIRED ACTION---------------------NOTE--------------
The Special Report shall include the results of radiological
analyses of the drinking water source, and the radiological
impact on finished drinking water supplies with regard to the requirements
of 40 CFR 141 , Safe Drinking Water Act, as applicable.
COMPLETION
TIMEMcGuireUnits
1 and 2 A.1 Prepare and submit a 30 days Special Report to the NRC which identifies
the causes for exceeding the limits, corrective
actions taken to reduce releases, and actions taken to ensure that subsequent
releases are within limits.16.11.3-1 Revision 0
Dose-Liquid Effluents 16.11.3 TESTING REQUIREMENTS
TEST TR 16.11.3.1 Determine cumulative
dose contributions
from liquid effluents for current calendar quarter and current calendar year in accordance
with the methodology
and parameters
in the ODCM.BASES FREQUENCY 31 days ((This commitment
is provided to implement the requirements
of Sections II.A, liLA and IV.A of Appendix I, 10 CFR Part 50.The commitment
implements
the guides set forth in Section II.A of Appendix I.The REMEDIAL ACTION statements
provide the required operating flexibility
and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive
material in liquid effluents to UNRESTRICTED
AREAS will be kept"as low as is reasonably
achievable." Also, for fresh water sites with drinking water supplies that can be potentially
affected by plant operations, there is reasonable
assurance that the operation of the facility will not result in radionuclide
concentrations
in the finished drinking water that are in excess of the requirements
of 40 CFR Part 141.These requirements
are applicable
only if the drinking water supply is taken from the river 3 miles downstream
of the plant discharge.
The dose calculation
methodology
and parameters
in the ODCM implement the requirements
in Section liLA of Appendix I that conformance
with the guides of Appendix I be shown by calculational
procedures
based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate
pathways is unlikely to be substantially
underestimated.
The equations specified in the ODCM for calculating
the doses due to the actual release rates of radioactive
materials in liquid effluents are consistent
with the methodology
provided in Regulatory
Guide 1.109,"Calculation
of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating
Compliance
with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory
Guide 1.113,"Estimating
Aquatic Dispersion
ofEffluentsfrom
Accidental
and Routine Reactor Releases for the Purpose of Implementing
Appendix 1," April 1977.This commitment
applies to the release of liquid effluents from each unit at the site.For units with shared Radwaste Treatment Systems, the liquid effluents from the shared system are to be proportioned
among the units sharing that system in accordance
with the guidance given in NUREG-0133, Chapter 3.1.McGuire Units 1 and 2 16.11.3-2 Revision 0
Dose-Liquid Effluents 16.11.3 REFERENCES
1.McGuire Nuclear Station, Off site Dose Calculation
Manual 2.40 CFR Part 141, Safe Drinking Water Act 3.10 CFR Part 50, Appendix I 4.Regulatory
Guide 1.109,"Calculation
of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating
Compliance
with 10 CFR Part 50, Appendix I," Revision 1, October 1977.5.Regulatory
Guide 1.113,"Estimating
Aquatic Dispersion
of Effluents from Accidental
and Routine Reactor Releases for the Purpose of Implementing
Appendix 1," April 1977.McGuire Units 1 and 2 16.11.3-3 Revision a
DUKE ENERGY MCGUIRE OPERATIONS
TRAINING (N N L L LLLPP 0 OBJECTIVE 0 0 R S R R00 6 Concerning
the Selected Licensee Commitments (SLC)related to Liquid Waste Releases;*Given the SLC Manual, discuss any commitments
and XXX their applicability.
- For any commitments
that have action required within oneXX X hour, state the action.*Given a set of parameter values or system conditions, XXX determine if any commitment
is (are)not met and any action(s)required within one hour.*Given the SLC Manual, discuss the basis for a given X*commitment.
- SRO only WERLROO6 OP-MC-WE-RLR
FOR TRAINING PURPOSES ONL Y Page 7 of 55 REV.13
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#'./.11"\'.1 C ,d----, l!,,,!Group#GO rlt.'2 KIA#E06 G2.1.20 Importance
Rating_4_.--=.6_Conduct of Operations:
Ability to interpret and execute procedure steps.Proposed Question: Given the following:
SRO 85 (*A LOCA has occurred on"1 B" Cold Leg.*ECCS has NOT functioned
as required.*All NC Pumps are TRIPPED.*PZR PORVs are CLOSED and in AUTO.*CET's indicate 692°F and rising.*Reactor Vessel LR Level is 35%and lowering.*Containment
pressure is 3 psig and rising slowly.Which ONE of the following procedures
will the crew implement for these conditions, and the action taken if ECCS components
CANNOT be restored?A.Enter FR-C.1, Response To Inadequate
Core Cooling;NC pumps are started prior to secondary depressurization
to provide forced cooling of the NCS.B.Enter FR-C.2, Response To Degraded Core Cooling;NC pumps are started prior to secondary depressurization
to provide forced cooling of the NCS.C.Enter FR-C.1, Response To Inadequate
Core Cooling;secondary depressurization
is initiated prior to attempting
NC pump operation to depressurize
the NCS and facilitate
injection.
D.Enter FR-C.2, Response To Degraded Core Cooling;secondary depressurization
is initiated prior to attempting
NC pump operation to depressurize
the NCS and facilitate
injection.
Proposed Answer: D Page 216 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Explanation (Optional):
A.Incorrect.
Wrong procedure entry and also wrong action for NCP operation.
A LOCA is in progress but conditions
for FR-C.1 do not exist B.Incorrect.
NCP would only be operated if secondary depressurization
was ineffective
in achievingcorecooling.
C.Incorrect.
Incorrect entry but correct action with respect to secondary depressurization
and NCP operation D.Correct.F-O, FR-C.2 Rev 5 Technical Reference(s)(Attach if not previously
__________
provided)EP-FRC Rev 10 Proposed references
to be provided to applicants
during None examination:
Learning Objective:
EP-FRC Obj 2&3 (Question Source: Bank#Modified Bank#New X (WTSI)(Note changes or attach parent)-----Question History: Last NRC Exam BVPS-1 2007---=--_.....:.-=-_---------
Question Cognitive Level: Memory or Fundamental
Knowledge Comprehension
or Analysis x (10 CFR Part 55 55.41 Content: 55.43 5---Comments: KA is matched because item evaluates knowledge of procedure steps for degraded core cooling condition.
SRO level because the applicant must assess (evaluate)
plant conditions
and determine procedure entry, as well as strategy for the procedure entered Page 217 of 260 Draft 7
MNS EP/1/A/5000/F-O
UNIT 1 CRITICAL SAFETY FUNCTION STATUS TREES Core Cooling-Page 1 of 1 PAGE NO.4 of 11Rev.4......GOTO IFR-C.! REACTOR VESSEL NO LOWER RANGE LEVEL f---GREATER THAN 39%YES GO TO FR-C.!..r-.GO TO
I I NO\-------YES REACTOR VESSEL NO LOWER RANGE LEVEL f----GREATER THAN 39%YES I I...r-.GOTO
CORE EXIT TICs LESS THAN 700°F NO f---YES CORE EXIT TICs LESS THAN 1200'F AT LEAST ONE NCPUMPON NO\-------YES****FR-C.3 NC SUBCOOLlNG
BASED ON CORE EXIT TICs GREATER THANO°F NO f---YES REACTOR VESSEL DIP GREATER THAN....._-REQUIRED FOR PUMP COMBINATION (SEE TABLE NEXT PAGE)..r-.GOTO
I I NO\-------YES I
MNS EP/1/A/5000/FR-C.2
UNITl RESPONSE TO DEGRADED CORE COOLING PAGE NO.13 of 46 Rev.5 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED NOTE After the Low Pressure Steamline Isolation signal is blocked, maintaining
steam pressure negative rate less than 2 PSIG per second will prevent a Main Steam Isolation.
15.e all intact S/Gs to 110 PSIG (a.REFER TO Enclosure 6 (NC Cooldown Rate Monitoring)
to assist in monitoring
100°F in an hour cooldown rate.b.Check condenser available:
- MSIV on all intact S/Gs-OPEN*"C-9 COND AVAILABLE FOR STEAM DUMP" status light (1 SI-18)LIT.c.Check"STEAM DUMP SELECT"-IN STEAM PRESSURE MODE.d.WHEN"P-12 LO-LO TAVG" status light (1 SI-18)lit, THEN place steam dumps in bypass interlock.
b.GO TO RNO for Step 15.e.c.Perform the following to place steam dumps in steam pressure mode:_1)Place"STM PRESS CONTROLLER" in manual._2)Adjust"STM PRESS CONTROLLER" output to equal"STEAM DUMP DEMAND" signal._3)Place"STEAM DUMP SELECT" in steam pressure mode.
MNS EP/1/A/5000/FR-C.2
UNIT 1 RESPONSE TO DEGRADED CORE COOLING PAGE NO.14 of 46 Rev.5 ACTION/EXPECTED
RESPONSE 15.(Continued)
RESPONSE NOT OBTAINED (e.e.Dump steam using all intact S/G(s)SM PORV as follows:_1)Ensure Main Steam Isolation reset._2)Ensure SM PORVs reset._3)Dump steam using all intact S/G(s)SM PORVs while maintaining
cool down rate in NC T-Colds less than 100°F in an hour.4)IF any intact S/G SM PORV closed, THEN dump steam using any of the following while maintaining
cooldown rate in NC T-Colds less than 100°F in an hour:_a)Dispatch operator to operate intact S/G(s)SM PORV.b)IF any intact S/G SM PORV is unavailable, THEN evaluate using the following to dump steam:*Reopen MSIVs and dump steam to condenser PER Enclosure 8 (Condenser
Dumps).*Run TO CA pump.*Use steam drains PER EP/1/A/5000/G-1 (Generic Enclosures), Enclosure 19 (S/G Depressurization
Using Steam Drains).f.Check intact S/G pressures-LESS f.RETURN TO Step 11.THAN 110 PSIG.(-g.Check at least two NC T-Hots-LESS-g.RETURN TO Step 11.\THAN 354°F.h.Stop S/G depressurization
and maintain S/G pressures stable.
DUKE POWER FR-C.2 Response to Degraded Core Cooling MCGUIRE OPERA TIONS TRAINING\STEP 14 WHEN"P-11 PRESSURIZER
SII BLOCK PERMISSIVE" status light (1 SI-18)lit, THEN depress"BLOCK" on Low Pressure steam line Isolation block switches.PURPOSE: To prevent MSIV closure on low steamline pressure during controlled
NC system cooldown.BASIS: The Steamline Isolation signal on low steamline pressure can be blocked during cooldown once the P-11 Block Permissive
status light is lit (approximately
1955 psig).This prevents MSIV closure, thus allowing cooldown by the preferred method of steam dump to the condenser.
NOTE After Low Pressure Steamline Isolation signal is blocked, maintaining
steam pressure negative rate less than 2 PSIG per second will prevent a Main Steam Isolation..PURPOSE: To warn the operator that MSIV isolation will occur if the S/G's are depressurized
too quickly and provide guidance for controlling
the depressurization
rate.BASIS:
N/A (('.To prevent nitrogen injection, the operator is directed to stop the secondary depressurization
when the S/G pressure reaches 110 psig.STEP 16 Check NO pumps-ON.PURPOSE: To see if NO pumps are running.BASIS: In this step the operator checks if the NO pumps are running and, if not, starts them since NO injection will be used to restore long-term core cooling.The NO pumps will inject if NC system pressure is dropped below their shutoff head.OP-MC-EP-FRC
FOR TRAINING PURPOSES ONL Y Page 73 of 111 REV.10
DUKE POWER MCGUIRE OPERA TJONS TRAINING 2.1.3 NC Pump Restart and Opening pzr PORVs (Continued)
The NC Pumps cannot be expected to run indefinitely
under highly voided NC system conditions.
The operator must still take action to establish a makeup source of water to the NC system to restore adequate long term cooling.NC system pressure must, therefore, be reduced in order for the CLAs andlor NO pumps to inject.The operator should continue attempts to depressurize
the S/Gs or to establish the secondary heat sink;however, if the core exit TIC temperatures
remain above 1200°F and all available NC Pumps are running, the only other option is to effectively
enlarge the hole in the NC system to reduce pressure.This may be achieved by opening all available NC system vent paths to containment, i.e., pzr PORVs, head vents, etc.It should be noted that venting the NC system to containment
reduces NC system inventory and is not as effective in reducing NC system pressure as S/G depressurization.
Some form of low pressure flow to the NC system must be established
as soon as possible.2.2.FR-C.2, Response to Degraded Core Cooling Degraded core cooling is caused by a substantial
loss of primary coolant.If the NC Pumps are not running, the degraded core cooling symptoms indicate the core is partially uncovered.
If the NC Pumps are running, the symptoms indicate the potential for core uncovery exists if the pumps should fail or be manually tripped.Operator action is required to restore NC system inventory in either case.(Reinitiation
of high pressure SII is the most effective method to restore NC system inventory and core cooling.If some form of high pressure injection cannot be established
or is ineffective
in restoring core cooling, then the operator must take actions to reduce the NC system pressure in order for the SII accumulators
and NO pumps to inject.A controlled
secondary depressurization
is an effective method for achieving this, while at the same time avoiding a rapid NC system cooldown that could cause problems with pressurized
thermal shock.The expected system response to both of the recovery techniques
is described below.OP-MC-EP-FRC
FOR TRAINING PURPOSES ONL Y Page 15 of 111 REV.10
DUKE POWER FR-C.2 Response to Degraded Core Cooling MCGUIRE OPERA TIONS TRAINING (4.0 FR-C.2, RESPONSE TO DEGRADED CORE COOLING 4.1.Purpose This procedure provides actions to restore adequate core cooling.The major actions are to be performed sequentially.
Success, as indicated by improved core cooling and increasing
vessel inventory, is evaluated prior to performing
the next action in the sequence.4.2.Symptoms/Entry
Conditions
This procedure is entered from EP/1/A/5000/F-O (Critical Safety Function Status Trees)(Core Cooling), on any orange condition.
These conditions
are: 1.Core exit TICs greater than 700°F and vessel LR level greater than 39%, or 2.Core exit TICs less than 700°F and vessel LR level less than 39%, or 3.Subcooling
less than OaF, at least one NC pump running, and reactor vessel DIP less than required for the NC pump combination.
OP-MC-EP-FRC
FOR TRAINING PURPOSES ONL Y Page 57 of 111 REV.10
DUKE POWER FR-C.2 Response to Degraded Core Cooling MCGUIRE OPERATIONS
TRAINING (4.3.Immediate/Major
Actions The recovery/restoration
technique includes the following two major action categories:
1.Establish Safety Injection flow to the NC system.2.Initiate a controlled
S/G depressurization
to cool down and depressurize
the NC system.The following subsections
provide a more detailed discussion
of each major action category: 4.3.1 Establish Safety Injection Flow to the NC System The operator must properly align emergency S/I valves, start the S/l pumps, and then check for flow through the S/Ilines to the NC system.Core exit T/Cs and the appropriate
RVLlS indication
are checked to determine the effectiveness
of S/I in restoring core cooling and vessel inventory.
4.3.2 Initiate a Controlled
S/G Depressurization
to Cool Down and Depressurize
the NC System The operator must maintaina1 OO°F/hr cooldown of the NC system by dumping steam to the condenser or opening the S/G PORVs while maintaining
adequate feedwater to the S/Gs.The CLAs must be isolated and the NC pumps tripped once the S/Gs have been depressurized
to 110 psig and the NC system has been depressurized
until NCHots are less than 354°F.The NC system cooldown and depressurization
is continued until NO flow to the NC system has been established
and verified.Core exit T/Cs and the appropriate
RVLlS indication
are checked to determine the effectiveness
of CLA and/or NO S/I in restoring core cooling and vessel inventory.
OP-MC-EP-FRC
FOR TRAINING PURPOSES ONL Y Page 59 of 111 REV.10
DUKE POWER 2.0 PROCEDURE SERIES BACKGROUND
MCGUIRE OPERA TIONS TRAINING (2.1.FR-C.1, Response to Inadequate
Core Cooling The indication
of inadequate
core cooling requires prompt operator action.Inadequate
core cooling is caused by a substantial
loss of primary coolant resulting in a partially or fully uncovered core.Without adequate heat removal, the core decay energy will cause the fuel temperatures
to rise.Severe fuel damage will occur unless core cooling is promptly restored.Reinitiation
of high pressure SII is the most effective method to recover the core and restore adequate core cooling.If some form of highpressureinjection
cannot be established
or is ineffective, then the operator must take actions to reduce NC system pressure in order for the CLAs and NO pumps to inject.Analyses have shown that a rapid secondary depressurization
is the most effectivemeansfor achieving this.If secondary depressurization
is not possible, or primary-to-secondary
heat transfer is significantly
degraded, then the operator must start the NC Pumps.The NC Pumps will provide forced two phase flow through the core and temporarily
improve core cooling until some form of make-up flow to the NC system can be established.
The recovery techniques
applied in this procedure were developed from transient analyses.The expected system response to each of the recovery techniques
is described below.2.1.1 Reinitiation
of High Pressure Safety Injection The introduction
of subcooled SII into the highly voided NC system will cause steam in the cold legs to condense.Steam flow throughout
the NC system will go up because of this condensation
effect.Superheated
steam forced out of the core may initially cause the core exit TIC temperatures
to go up.As the vessel begins to refill, heat transfer from the fuel will cause the fluid entering the core to boil vigorously.
This will create a two phase mixture which will eventually
re-cover the entire core and cause the core exit TIC temperatures
to quickly go down to saturation
temperature.
This procedure uses the trends in core exit TIC temperatures
and indicated vessel level to determine appropriate
operator actions.The effectiveness
of SII in restoring NC system inventory is determined
by the trend in RVLlS indication.
If going up, then no further action may be necessary.
The effectiveness
of SII in restoring core cooling is determined
bythetrend in core exit TIC temperatures.
If going down, no further action isnecessary.Exit
temperatures
less than 700°F indicate success, allowing the operator to return to the procedure and step in effect.OP-MC-EP-FRC
FOR TRAINING PURPOSES ONL Y Page 11 of 111 REV.10
DUKE POWER MCGUIRE OPERA TIONS TRAINING (((2.1.2 Secondary Depressurization
If attempts to reinitiate
high pressure S/I are unsuccessful, or are ineffective
in restoring adequatecorecooling, then a rapid S/G depressurization
must be performed.
A rapid secondary depressurization
will raise primary-to-secondary
heat transfer and cause steam in the primary side of the S/G U-tubes to condense.When the condensation
rate exceedsthesteam generation
rate, the NC system will begin to depressurize.
As the NC system pressure drops, voiding of the water resident in the lower plenum and downcomer will partially recover the core with a two phase mixture.The continued depressurization
will eventually
cause S/I accumulator
injection and temporary core recovery.The operator should check the NC hot leg temperature
trend to determine the effectiveness
of the S/G depressurization
in reducing the NC system pressure.The hot leg temperatures
may initially rise as superheated
steam in the core is forced out by the advancing two phase flow, but should quickly go down to saturation
and continue to go down as the NC system depressurizes.
To prevent nitrogen injection from the S/I accumulators, the operator must isolate them.NC T-Hot less than 354°F and intact S/G pressure less than 110 psig are used to determine when the S/I accumulators
should be isolated.After the CLAs have been isolated, the secondary should be depressurized
to atmospheric
pressure.The NC system pressure should follow secondary pressure until the ND pumps begin to inject.Adequate core cooling has been restored and preparations
for long term plant recovery can be started once ND flow has been established
and the core is completely
covered.2.1.3 NC Pump Restart and Opening pzr PORVs If some form of high pressure injection cannot be established
or is ineffective
in restoring adequate core cooling, and if S/G depressurization
is not possible or ineffective, then starting the NC Pumps will provide forced two phase flow through the core and temporarily
improve core cooling.The core exit T/C temperatures
should rapidly go down and the RVLlS indication
should rapidly go up as a steam/water
mixture is forced through the core by the NC Pumps.Analysis has shown that with secondary heat sink available, the NC Pumps will maintain core cooling as long as they continue to run.However, it should be noted that a degraded core cooling condition still exists.OP-MC-EP-FRC
FOR TRAINING PURPOSES ONL Y Page 13 of 111 REV.10
DUKE POWER MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours>.NLONLORLPRO LPSO LOR 3.0 3.0 2.0 OBJECTIVES
((SNNLL L E OBJECTIVELL P P 000 R S R Q R00 1 Explain the purpose of each procedure in the FR-C series.XX EPFRCOO1 2 Discuss the entry and exit guidance for each procedure in theXX FR-C series.EPFRCOO2 3 Discuss the mitigating
strategy (major actions)of each XXX procedure in the FR-C series.EPFRCOO3 4 Discuss the basis for any note, caution or step for eachXXX procedure in the FR-C series.EPFRCOO4 5 Given the Foldout page, discuss the actions included and theXXX basis for these actions.EPFRCOO5 6 Given the appropriateprocedure,evaluate
a given scenario XXX describing
accident events and plant conditions
to determine any required action and its basis.EPFRCOO6 7 Discuss the time critical task(s)associated
with the FR-CXX X series procedures
including the time requirements
and the basis for these requirements.
EPFRCOO7 OP-MC-EP-FRC
FOR TRAINING PURPOSES ONL Y Page 5 of 111 REV.10
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#2----Group#1 KIA#010 A2.01----------
Importance
Rating 3.6----Ability to (a)predict the hnpacts of the following malfunctions
or operations
on the PZR PCS;and (b)based on those predictions, use procedures
to correct,
or nlitigate the consequences
of those rna!functions
or operations:
Heater failures Proposed Question: Giventhefollowing:
SR086*Unit 1 is at 1 00%power.*A pressurizer
pressure transient has occurred, resulting in a PZR PORV momentarily
opening.*The crew has stabilized
the unit.*Actions of AP/11, Pressurizer
Pressure anomalies, are being performed.
- NC pressure is 2120 psig and stable.*PZR heater groups 1 A, 1 B,1C are energized.
- PZR heater group1D is de-energized.
- PZR Spray Valves and PORVs indicate closed.Which ONE of the following describes the impact of the current plant conditions, and the action required in accordance
with technical specifications
and AP/11?A.NC System DNB limits are exceeding TS 3.4.1 COLR limits;place group1D PZR heater mode select switch in MANUAL and energize to raise pressure;restore NC pressure to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.B.Pressurizer
TS 3.4.9 is applicable
due to de-energized
backup heaters;Place PZR PRESS MASTER in MANUAL to control pressure manually;verify capacity of remaining Backup Heaters or initiate a plant shutdown to Mode 3 within the required action time.C.Pressurizer
TS 3.4.9 is applicable
due to de-energized
backup heaters;Place group1D PZR heater mode select switch in MANUAL and energize to raise pressure;TS 3.4.9 no longer applies when1D Backup Heaters are operating in MANUAL.Page 218 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 D.NC System DNB limits are exceeding TS 3.4.1 COLR limits;Place PZR PRESS MASTER in MANUAL to control pressure manually;restore NC pressure to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.Proposed Answer: A Explanation (Optional):
A.Correct.With NC pressure at 2120, DNB limits are not being met lAW COLR.B.Incorrect.
3.4.9 not required for loss of1D heaters.Action is plausible because it is action required if loss of1A or1B heaters occurs.PZR master in manual would be for 1 Cheaters C.Incorrect.
3.4.9 not required for loss of1D heaters.Action is plausible because it is action allowed for restoration
of1A or1B heaters D.Incorrect.
Master controller
will not operate bank 1 D, will operate 1 C.Impact is correct, however TS 3.4.1;COLR Rev 30 Technical Reference(s)(Attach if not previously
provided)----------
AP/11, Rev 10 TS 3.4.9 and Basis Proposed references
to be provided to applicants
during None examination:
Learning Objective: (As available)
(Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge Comprehension
or Analysis X 1 0 CFR Part 55 55.41 Content: 55.43 2,5 Page 219 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Comments: KA is matched because the item evaluates TS impact of failure, and also requires knowledge of action required to mitigate the consequences
of the event.SRO level because item requires knowledge of TS LCOs involved, and procedure strategy required for mitigation
Page 220 of 260 Draft 7
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)Limits LCO 3.4.1 ReS DNB parameters
for pressurizer
pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in Table 3.4.1-1.APPLICABILITY:
MODE 1.------------------------------------N()TE------------------------------------------------------
Pressurizer
pressure limit does not apply during: a.THERMAL P()WER ramp>5%RTP per minute;or b.THERMAL POWER step>10%RTP.ACTI()NS C()NDITION
REQUIRED ACTI()N C()MPLETI()N
TIME A.Pressurizer
pressure or A.1 Restore DNB parameter(s)
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> RCS average to within limit.temperature
DNB parameters
not within limits.B.RCS total flow rateB.1 Reduce THERMAL 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 99%, but<1 OO°k>of the P()WER to98%RTP.limit specified in the C()LR.AND B.2 Reduce the Power Range 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sNeutronFlux-High Trip Setpoint below the nominal setpoint by 2%RTP.(continued)
McGuire Units 1 and 2 3.4.1-1 Amendment Nos.219/201
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 Table 3.4.1-1 (page 1 of 1)RCS DNB Parameters
PARAMETER INDICATION
No.LIMITS OPERABLE CHANNELS 1.Indicated RCS meter 4The limit specified in the COLR.Average meter 3The limit specified in the COLR.Temperature
computer 4The limit specified in the COLR.computer 3The limit specified in the COLR.2.Indicated meter 4The limit specified in the COLR.Pressurizer
meter 3The limit specified in the COLR.Pressure computer 4The limit specified in the COLR.computer 3The limit specified in the COLR.3.RCS Total Flow388,000 gpm and greater than or Rate equal to the limit specified in the COLR.McGuire Units 1 and 2 3.4.1-4 Amendment Nos.219/201
MCEI-Q400-46
Page 27 of 32 Revision 30 McGuire 1 Cycle 19 Core Operating Limits Report Table 4 Reactor Coolant System DNB Parameters
PARAMETER 1.Indicated ReS Average Temperature
2.Indicated Pressurizer
Pressure 3.ReS Total Flow Rate No.Operable INDICA TION*CHANNELS LIMITS meter 4 S 587.2 OP meter 3586.9 OF computer 4:5 587.7 Of computer 3 5587.5 OP meter 4 2: 2219.8 psig meter 32222.1 psig computer 42215.8 psig
3 2: 2217.5 psig390,000 gpm**Note: The ReS minimum coolant flow rate assumed in the licensing analyses for the MIC19 core is 388 t OOO gpIn However, the flow is set at 390,000which is conservative
MNS AP/1/A/5500/11
UNITl PRESSURIZER
PRESSURE ANOMALIES PAGE NO.7 of 9 Rev.10 ACTION/EXPECTED
RESPONSE 11.Check the following pzr heaters-ON:*1A*1B*1 D.12.Check1C pzr heaters-ON.RESPONSE NOT OBTAINED IF NC pressure below desired pressure, THEN: a.Place pzr heater mode select switches in manual.b.Turn on heaters as necessary to control pressure.IF NC pressure below desired pressure, THEN: a.Place"PZR PRESS MASTER" in manual.b.Control pressure.c.WHEN pzr pressure returns to normal AND automatic pzr pressure control desired, THEN place"PZR PRESS MASTER" in auto.13.Check pzr pressure-GOING UP TO DESIRED PRESSURE.14.Check 111 NC-27 PRESSURIZER
SPRAY EMERGENCY CLOSE II switchSELECTED TO IINORMAL II*15.Check 111 NC-29 PRESSURIZER
SPRAY EMERGENCY CLOSE IISELECTED TO IINORMAL II*16.GO TO Step 24._IF pressure continues to go down, THEN REFER TO AP/1/Al5500/10 (NC System Leakage Within The Capacity Of Both NV Pumps)._Notify station management
to ensure switch restored to IINORMALII
once spray valve is repaired._Notify station management
to ensure switch restored to IINORMALII
once spray valve is repaired.
Pressurizer
3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.9 Pressurizer
LCO 3.4.9 The pressurizer
shall be OPERABLE with: a.Pressurizer
water level92%(1600 ft 3);and b.Two groups ofpressurizerheaters
OPERABLE with the capacity of each group 2:.150 kW.APPLICABILITY:
MODES 1,2, and 3.ACTIONS CONDITION REQUIRED ACTION COMPLETION
TIME A.Pressurizer
water level A.1 Be in MODE 3 with reactor 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> not within limit.trip breakers open.AND A.2 Be in MODE 4.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B.One required group of B.1 Restore required group of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> pressurizer
heaters pressurizer
heaters to inoperable.
OPERABLE status.C.Required Action and C.1 Be in MODE 3.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated
Completion
Time of Condition B not AND met.C.2 Be in MODE 4.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> McGuire Units 1 and 2 3.4.9-1 Amendment Nos.184/166
Pressurizer
B 3.4.9 BASES APPLICABLE
In MODES 1,2, and 3, the LCO requirement
for pressurizer
level to SAFETY ANALYSES remain within the required range is consistent
with the accident analyses.Safety analyses performed for lower MODES are not limiting.All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions
in the pressurizer.
In making this assumption, the analyses neglect the small fraction of noncondensible
gases normally present.Safety analyses presented in the UFSAR (Ref.1)do not take credit for pressurizer
heater operation;
however, an initial condition assumption
of the safety analyses is that the RCS is operating at normal pressure.The maximum pressurizer
water level limit satisfies Criterion 2 of10 CFR 50.36 (Ref.2).Although the heaters are not specifically
used in accident analysis, the need to maintain subcooling
in the long term during loss of offsite power, as indicated in NUREG-0737 (Ref.3), is the reason for providing an LCO.LCO APPLICABI LITY The LCO requirement
for the pressurizer
to be OPERABLE with a water volume1600 cubic feet, which is equivalent
to 92%, ensures that a steam bubble exists.Limiting the LCO maximum operating water level preserves the steam space for pressure control.The LCO has been established
to ensure the capability
to establish and maintain pressure control for steady state operation and to minimize the consequences
of potential overpressure
Requiring the presence of a steam bubble is also consistent
with safety analysis analytical
assumptions.
The LCO requires two groups of OPERABLE pressurizer
heaters, each with a capacity150 kW, capable of being powered from either the offsite power source or the emergency power supply.Only heater groups A and B are capable of being powered from the emergency power supply.The minimum heater capacity required is sufficient
to maintain the RCS near normal operating pressure when accounting
for heat losses through the pressurizer
insulation.
By maintaining
the pressure near the operating conditions, a wide margin to subcooling
can be obtained in the loops.The amount needed to maintain pressure is dependent on the heat losses.The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature, resulting in the greatest effect on pressurizer
level and RCS pressure control.Thus, applicability
has been designated
for MODES 1 and 2.The applicability
is also provided for MODE 3.The purpose is to prevent solid water RCS McGuire Units 1 and 2 B 3.4.9-2 Revision No.0
I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#2----Group#1 KIA#026 A2.03----------
Importance
Rating 4.4----Ability to (a)predict the irnpacts of the follovving
malfunctions
or operations
on the CBS;and (b)based on those predictions, use procedures
to correct, control;or nlitigate the consequences
of those rnalfunctions
or operations:
Failure of ESF Proposed Question: SRO 87 Given the following:*A Main Steam Break has occurred on Unit 1.*The Train"A" Load Sequencer is de-energized.
- "B" NS Pump did NOT automatically
start.*The crew has transitioned
to E-2, Faulted Steam Generator Isolation when the following conditions
are observed: o NC SYSTEM pressure 1400 psig and lowering.o Containment
Pressure 11 psig and rising.Enter FR-Z.1 based on a(n)...A.ORANGE CSF Status Tree;Ensure NC Pumps are off and start at least ONE NS Pump;procedure may subsequently
be completed as time allows.B.ORANGE CSF Status Tree;Perform all actions of FR-Z.1 and do NOTperformactions
of other procedures
unless a higher priority ORANGE or RED condition occurs.C.RED CSF Status Tree;Ensure NC Pumps are off and start at least ONE NS Pump;procedure may subsequently
be completed as time allows.D.RED CSF Status Tree;Perform all actions of FR-Z.1 and do NOT perform actions of other procedures
unless a higher priority RED condition occurs.Proposed Answer: A Page 221 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Explanation (Optional):
A.Correct.After step 8, procedure is treated as a yellow path for conditions
such as a steam line break.This is determined
by the SRO B.Incorrect.
SRO should know that a steam break is occurring and note will apply that procedure may be treated as a yellow path after initial actions are performed c.Incorrect.
Red path is 15 psig, but actions are correct D.Incorrect.
Red path is 15 psig and procedure is treated as a yellow path after step 8 FR-Z.1 (Rev 14)Technical Reference(s)(Attach if not previously
provided)----------
EP-FRZ Rev 15 OMP 4-3 p17, 18 Proposed references
to be provided to applicants
during None examination:
Learning Objective:
EP-FRZ Obj2&4 (Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge Comprehension
or Analysis X 10 CFR Part 55 55.41 Content: 55.43 5 Comments: KA is matched because a containment
spray failure has occurred.The impact is the result on CSF status, and the action required is also tested.SRO level because the SRO must select the appropriate
strategy for procedure use, including a judgment of when the Containment
Orange condition may be treated as a yellow condition'Page 222 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Page 223 of 260 Draft 7 Form ES-401-5
MNSEP/1/A/5000/F-0
UNITl CRITICAL SAFETY FUNCTION STATUS TREES Containment
-Page 1 of 1 PAGE NO.g of 11 Rev.4 GaIO FR-Z.1 CONTAINMENT
PRESSURE LESS THAN 15 PSla NO YES.....................
I I I I II***************
I\4/1 FR-Z.2 I I I I I CONTAINMENT
SUMP LEVEL LESS THAN 12.5 FT NO YES GOIO FR-Z.3 CONTAINMENT
RADIATION LESS THAN 35 RlHR (IEMF 51A OR B)NO YES CONTH2CONT
LESS THAN 0.5%********."..**..NO YES GOIO FR-Z.4 CSF SAT
MNS EP/1/A/5000/FR-Z.1
UNITl A.Purpose RESPONSE TO HIGH CONTAINMENT
PRESSURE PAGE NO.1 of 44 Rev.14 This procedure provides actions to respond to a high containment
pressure.B.Symptoms or Entry Conditions
This procedure is entered from EP/1/A/5000/F-O (Critical Safety Function Status Trees)(Containment), on a red or orange condition.
MNS EP/1/A/5000/FR-Z.1
UNITl RESPONSE TO HIGH CONTAINMENT
PRESSURE PAGE NO.2 of 44 Rev.14 ACTION/EXPECTED
RESPONSE C.Operator Actions RESPONSE NOT OBTAINED 1.IF loss of emergency coolant recirc has occurred, THEN this procedure may be completed as time allows.2.Monitor Foldout Page.3.Stop all NC pumps.4.Ensure all RV pumps are in manual and off.CAUTION The following breakers must be closed within 50 minutes of SII.5.Dispatch operator to remove white tags and close the following breakers:*1 EMXA-R2A (1 A ND To A&B Cold Legs Cant Outside Isol Motor (1NI-173A))(aux bldg, 750, FF-54, FF-55)*1 EMXB1-6B (1 B ND To C&D NC Cold Leg Cont Outside Isol Motor (1 NI-178B))(aux bldg, 733, GG-55, GG-56).6.Check containment
pressure-LESS THAN 15 PSIG.7.Check any NS pump-ON._GO TO Step 9._GO TO Step 9.NOTE The remainder of this EP may be completed with the priority of a yellow path EP.Completion
of this EP should be delayed if faulted S/G has occurred, or other higher priority actions are required.8.Perform the remainder of this EP as time allows.
DUKE POWER FR-Z.1 Response to High Containment
Pressure MCGUIRE OPERATIONS
TRAINING STEP 3 STEP 4 Stop all NC Pumps.Stop all RV Pumps.PURPOSE: To stop all NC and RV pumps.BASIS: The NC pumps are tripped since component cooling water to the NC pump seals and motors is isolated by the Phase B containment
isolation.
RV pumps are tripped because the suction is isolated by the Phase B containment
isolation signal.STEPS Dispatch operator to remove tags and close breakers for the following valves:*NI-173A (Train A NO to A&B Cold Leg)*NI-1788 (Train B NO to C&O Cold Leg)PURPOSE: To prepare the ND Aux containment
spray system for use if needed.BASIS: This step allows ND Aux containment
spray to be able to be aligned in subsequent
steps/procedures
at the 50 minute requirement
of the FSAR.STEP 6 STEP 7 STEP 8 Check Containment
Pressure-less 15 psig Check any NS pump-ON The remainder of this EP may be performed as time allows.PURPOSE: Allows crew to perform other procedures
in a more timely manner provided containment
pressure is less than 15 psig and any NS pump is on.BASIS: The specific scenario the WOG had in mind for this allowance is a steam break inside containment.
This will aid in terminating
81 prior to going solid in the pressurizer
for this scenario.If there are no other priority actions to complete, you may as well finish FR-Z.1 and get it out of the way.An example would be a large break LOCA.There is little to do until you need to transfer to Cold Leg Recirc.OP-MC-EP-FRZ
FOR TRAINING PURPOSES ONL Y Page 23 of 81 REV.15
7.15.1 OMP4-3 Page 17 of 35 Implementing
CSF Path Procedures
7.15.1.1 7.15.1.2 7.15.1.3 7.15.1.4 CSF procedures
are NOT to be implemented
prior to transition
from EP/1,2/A/5000/E-O (Reactor Trip or Safety Injection).
IF a CSF path is red or orange while the operating crew is in EP/1,2/A/5000/E-O, but has turned to green upon transition
from E-O, the CSF procedure which was in alarm shall NOT be implemented.
IF the CSF path is yellow, it shall be handled as any other yellow path procedure per Section 7.15.1.7.IF there are any valid red or orange path CSF's on transition
from E-O (unless transition
is to EP/1,2/A/5000/ECA-O (Loss of All AC Power), the associated
CSF procedure shall be implemented.
IF a valid red or orange path flickers into alarm on SPDS but returns to green prior to the crew validating
the condition and implementing
the procedure (implementation
of procedure being that the SRO either hands out fold-out pages or starts reading from the procedure), the CSF procedure shall NOT be implemented.
IF the CSF path is yellow, it shall be handled as any other yellow path procedure per Section 7.15.1.7.Likewise, if a valid red path or orange path goes into alarm during performance
of a higher priority CSF procedure, but returns to green prior to transition
from the higher priority CSF path procedure to the lower priority CSF procedure, the associated
CSF procedure shall NOT be implemented.
IF a CSF procedure directs the operator to return to the procedure and step in effect, AND the corresponding
status tree continues to display the offnormal conditions, the corresponding
CSF procedure does NOT have to be implemented
again, since all recovery actions have been completed.
However, if the same status tree subsequently
changes to a valid higher priority condition, OR if it changes to lower condition and returns to higher priority condition again, the corresponding
CSF procedure shall be implemented
as required by its priority.Red Path IF any valid red path is encountered
during monitoring, the operator is required to immediately
implement the corresponding
in progress shall be discontinued.
IF during the performance
of any red path procedure, a valid red condition of higher priority arises, the higher priority condition should be addressed first, and the lower priority red path procedure suspended.
7.15.1.5 7.15.1.6 OMP4-3 Page 18 of 35 Orange Path IF any valid orange path is encountered, the operator is expected to scan all of the remaining trees, and then, if no valid red is encountered, promptly implement the corresponding
EP.IF during the performance
of an orange path procedure, any valid red condition or higher priority valid orange condition arises, the red or higher priority orange condition is to be addressed first, and the original orange path procedure suspended.
Completion
of Red or Orange Path Procedure Once procedure is entered due to a red or orange condition, that procedure should be performed to completion, unless preempted by some higher priority condition.
It is expected that the actions in the procedure will clear the red or orange condition before all the operator actions are complete.However, these procedures
should be performed to the point of the defined transition
to a specific procedure or to the"procedure
and step in effect" to ensure the condition remains clear.At this point any lower priority red or orange paths currently indicating
or previously
started but NOT completed shall be addressed.
FR-S.l, P.l and Z.1 can be entered from either an orange or red path status.IF the color changes from orange to red while you are in one of these EPs, the crew should continue and complete the EP from where they are.Crew does NOT have to backup and restart the EP.IF the orange path is exited, and it subsequently
turns red, the EP must be re-entered
at Step 1.Upon continuation
of recovery actions in Optimal Recovery procedure, some judgment may be required by the operator to avoid inadvertent
reinstatement
of a Red or Orange condition by undoing some critical step in the Function Recovery procedure.
The Optimal Recovery procedures
are optimal assuming that safety equipment is available.
The appearance
of a Red or Orange condition in most cases implies that some equipment or function required for safety is NOT available, and by implication
some adjustment
may be required in the Optimal Recovery procedure.
DUKE POWER MCGUIRE OPERA TIONS TRAINING CLASSROOM TIME (Hours}NLO NLOR LPRO LPSO LOR 0.5 0.5 0.5 OBJECTIVES
SNNLL L E OBJECTIVELL P P 0 0 0 RSR Q R 0 0 1 Explain the purpose of each procedure in the FR-Z series.X X 2 Discuss the entry and exit guidance for each procedure in the X X FR-Z series.3 Discuss the mitigating
strategy (major actions)of each X X X procedure in the FR-Z series.4 Discuss the basis for any note, caution or step for eachXX X procedure in the FR-Z series.5 Given the Foldout page, discuss the actions included and the X X X basis for these actions.6 Given the appropriate
procedure, evaluate a given scenario X X X describing
accident events and plant conditions
to determine any required action and its basis.7 Discuss the time critical task(s)associated
with the FR-Z X X X series procedures
including the time requirements
and the basis for these requirements.
OP-MC-EP-FRZ
FOR TRAINING PURPOSES ONL Y Page 5 of 81 REV.15
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO 4.7 Tier#2 Group#...ll---1_KIA#AW'_0_6_1_G_2_.4_.4
_Importance
Rating Ernergency
Procedures
/Plan: Ability to recognize abnorrnal indications
for system operating pararneters
which arelevel conditions
for ernergency
and abnormal operating procedures, Proposed Question: SR088 Given the following conditions:
- A Reactor Trip with SI occurs.*The operators perform the immediate action steps, verify SI flow, and check CA flow in accordance
with EP/1/A/5000/E-0, Reactor Trip or Safety Injection.
- The RO reports all 3 CA pumps are off.*NCS pressure is 900 psig.*All SG pressures are between 825 psig and 850 psig.*All SG NR levels are off scale low.*All SG WR levels are approximately
39%.*E-O directs the crew to implement EP/1/A/5000/F-0, Critical Safety Function Status Trees.Which ONE (1)of the following actions is to be taken?A.Transition
to FR-H.1."Response to Loss of Secondary Heat Sink," and attempt to establish CA or Feedwater flow.B.Transition
to FR-H.1,Response
to Loss of Secondary Heat Sink," and initiate NCS feed and bleed.C.Transition
to FR-H.1,Response
to Loss of Secondary Heat Sink," and then return to"procedure
and step in effect" since a secondary heat sink is NOT required.D.Remain in EP-E.O, Reactor Trip or Safety Injection, until directed to EP-E.1, Loss of Reactor or Secondary Coolant since a secondary heat sink is NOT required.Page 224 of 260 Draft 7
ES-401 Answer: A Sample Written Examination
Question Worksheet Form ES-401-5 Explanation (Optional):
a.Correct.NC pressure is higher than SG pressure, therefore, use H.1 b.Plausible since these are actions that might be taken upon entry into FR-H.1.but SG levels do not meet the criteria.(24%, 36%ACC)c.Incorrect.
Since NCS pressure is higher than SG pressure, a secondary heat sink is required.d.Incorrect.
Plausible since a LOCA is in progress, and the only criteria making this incorrect is that NC pressure is higher than SG pressure Technical Reference(s):
FR-H.1 page 2 (Rev 1)E-O, Rev 24 EP-FRH Rev10 (Attach if not previously
provided)references
to be provided to applicants
during examination:
None Learning Objective:
EP-FRH Obj 2, 3, 4 Question Source: Bank#Modified Bank#New EPFRHN011 (Note changes or attach parent)Question History: Question Cognitive Level: Last NRC Exam Memory or Fundamental
Knowledge Comprehension
or Analysis x10 CFR Part 55 55.41 Content: 55.43 5 Comments: Modified 2007 SRO Retake#80 conditions
and answer.Also modified distractor
o Page 225 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 KA is matched because a failure of AFW for these conditions
results in entry to FR-H.1.SRO level because the SRO is required to evaluate procedure selection as well as strategy for the condition presented Page 226 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Given the following conditions:*A Reactor Trip with SI occurs.*The operators perform the immediate action steps, verify SI flow, and check CA flow in accordance
with EP/1/A/5000/E-O, Reactor Trip or Safety Injection.
- The RO reports all 3 CA pumps are off*NCS pressure is 400 psig.*All SG pressures are between 425 psig and 450 psig.*All SG levels are 5%NR*E-O directs the crew to implement EP/1/A/5000/F-0, Critical Safety Function Status Trees.Which ONE (1)of the following actions is to be taken?A.Transition
to FR-H.1.IIResponse
to Loss of Secondary Heat Sink,1I and attempt to establish CA or Feedwater flow.B.Transition
to FR-H.1,Response
to Loss of Secondary Heat Sink," and initiate NCS feed and bleed.C.Transition
to FR-H.1,Response
to Loss of Secondary Heat Sink," and then return to IIprocedure
and step in effect ll since a secondary heat sink is NOT required.D.Remain in EP-E.O, Reactor Trip or Safety Injection since a secondary heat sink is NOT required Ans C Page 227 of 260 Draft 7
MNS EP/1/A/5000/F-O
UNITl CRITICAL SAFETY FUNCTION STATUS TREES Heat Sink-Page 1 of 1 PAGE NO.6 of 11Rev.4 GOIO FR-H.1 TOTALFEEDWATER
NO FLOW TO lNTACT S/Gs GREATER THAN YES 450 GPM NIR LEVEL IN AT NO LEAST ONE S/G GREATER THAN 11%YES (32%ACC)**********************
- ..****GOIO FR-H.2 PRESSURE IN ALL S/Gs LESS THAN 1225 PSIG NO YES****************
- ..***GOIO FR-H.3 NIR LEVEL IN ALL S/Gs LESS THAN 830/0 NO YES**********
MNS EP/1/A/5000/E-0
UNITl REACTOR TRIP OR SAFETY INJECTION PAGE NO.12 of 36 Rev.24 ACTION/EXPECTED
RESPONSE 16.Check CA flow: a.Total CA flow-GREATER THAN 450 GPM.RESPONSE NOT OBTAINED a.Perform the following:
1)IF N/R level in all S/Gs is less than 11%(32%ACC), THEN:*Ensure correct valve alignment*Start CA pumps.2)IF N/R level in all S/Gs is less than 11%(32%ACC)AND feed flow greater than 450 GPM can not be established, THEN:*Implement EP/1/A/5000/F-0 (Critical Safety Function Status Trees).*GO TO EP/1/A/5000/FR-H.1 (Response To Loss Of Secondary Heat Sink).b.Check VI header pressure-GREATER THAN 60 PSIG.c.WHEN N/R level in any S/G greater than 11%(32%ACC), THEN control CA flow to maintain N/R levels between 11(32%ACC)and 50%.b.IF CA flow can not be throttled with CA control valves in subsequent
steps, THEN control flow PER EP/1/A/5000/G-1 (Generic Enclosures), Enclosure 16 (CA Flow Control With Loss of VI).
MNS EP/1/A/5000/FR-H.1
UNITl A.Purpose RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.1 of 81 Rev.12 This procedure provides actions to respond to a loss of secondary heat sink in all steam generators.
B.Symptoms or Entry Conditions
This procedure is entered from:*EP/1/A/5000/E-O (Reactor Trip Or Safety Injection), Step 16, when minimum CA flow is not verified AND N/R level in all S/Gs is less than 11%(32%ACC).*EP/1/A/5000/F-O (Critical Safety Function Status Trees)(Heat Sink), on a red condition.
MNS EP/1/A/5000/FR-H.1
UNITl RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.3 of 81 Rev.12 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 4.Check at least one of the following NV pumps-AVAILABLE:
_.1A NV pump OR_.1B NV pump.5.Check if NC System feed and bleed should be initiated:
a.Check W/R level in at least 3 S/GsLESS THAN(36%ACC)._b.GO TO Step 20.6.Ensure SIG BB and NM valves closed PER Enclosure 3 (S/G BB and Sampling Valve Checklist).
7.Attempt to establish CA flow to at least one SIG as follows: a.Check power to both motor driven CA pumps-AVAILABLE.
b.Ensure control room CA valves aligned PER Enclosure 4 (CA Valve Alignment).
c.Start all available CA pumps._GO TO Step 20.a.Perform the following:
_1)Monitor feed and bleed initiation
criteria._2)WHEN criteria satisfied, THEN GO TO Step 20._3)GO TO Step 6.a.Perform the following:
- IF essential power is not available, THEN restore power to the affected essential bus PER AP/1/A/5500/07 (Loss of Electrical
Power).*IF the essential bus is energized, THEN dispatch operator to determine cause of breaker failure.
MNS EP/1/A/5000/FR-H.1
UNITl RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.11 of 81 Rev.12 ACTION/EXPECTED
RESPONSE 11.Check CM System in service:*Hotwell pump(s)-ON*Condensate
Booster pump(s)-ON.RESPONSE NOT OBTAINED Perform the following:
a.IF CM System is not available to be placed in service, THEN GO TO Step 19.NOTE Hotwell and Condensate
Booster pump will be started in operating procedure.
Once these pumps are on, this EP provides steps to start available CF pump(s).12.Check CF pumps-AT LEAST ONE AVAILABLE TO START.b.Place CM System in service PER OP/1/A/6250/001 (Condensate
And Feedwater System), Enclosure 4.2 (CM System Hot Restart).c.Do not continue until condensate
booster pump is on._IF both CF pumps are known to be incapable of starting, THEN GO TO Step 15.NOTE If it appears that CF pump might not be restored prior to reaching feed and bleed criteria, it may be preferable
to hand off Enclosure 7 (Reestablishing
CFFlow)to another SRO and/or RO while continuing
with subsequent
steps.13.Establish CF flow PER Enclosure 7 (Reestablishing
CF Flow)._GO TO Step 15.
MNS EP/1/A/5000/FR-H.1
UNITl RESPONSE TO LOSS OF SECONDARY HEAT SINK Enclosure1-Page1 of 1 Foldout PAGE NO.45 of 81 Rev.12 1.Cold Leg Recirc Switchover
Criteria:*IF FWST level reaches 180 inches ("FWST LEVEL LO" alarm), THEN GO TO EP/1/A/5000/ES-1.3 (Transfer To Cold Leg Recirc).2.CA Suction Sources:*IF CA storage tank (water tower)goes below 1.5 ft, THEN perform EP/1/A/5000/G-1 (Generic Enclosures), Enclosure 20 (CA Suction Source Realignment).
3.NC System Feed and Bleed Criteria (Applies after Step 2 in the body of the procedure):
- IF W/R level in at least 3 S/Gs goes below 24%(36%ACC), THEN GO TO Step 20 in the body of the procedure.
MNS EP/1/A/5000/FR-H.1
UNITl RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.2 of 81 Rev.12 ACTION/EXPECTED
RESPONSE C.Operator Actions 1.IF total feed flow is less than 450 GPM due to operator action, THEN RETURN TO procedure and step in effect.RESPONSE NOT OBTAINED CAUTION If a non-faulted
S/G is available, then feed flow should only be established
to non-faulted
S/G(s)in subsequent
steps.2.Check if secondary heat sink is required: a.NC pressure-GREATER THAN ANY NON-FAULTED
SIG PRESSURE.b.Any NC T-Hot-GREATER THAN 350°F (347°F ACC).3.Monitor Foldout Page.a.RETURN TO procedure and step in effect.b.Perform the following while continuing
in this procedure:
1)Try to place ND in RHR mode:_a)Ensure NC pressure is less than 385 PSIG._b)IF SII has occurred, THEN place ND in RHR mode PER EP/1/A/5000/G-2 (Placing ND In RHR Mode)._c)IF SII has not occurred, THEN place ND in RHR mode PER Enclosure 2 (Placing ND in RHR mode)._2)WHEN adequate ND cooling is established, THEN RETURN TO procedure and step in effect.
DUKE POWER MCGWRE OPERA noNS TRAINING FR-H.1 Loss of Secondary Heat Sink CAUTION If a non-faulted
S/G is available, then feed flow should only be established
to non-faulted
SIGs in subsequent
steps.PURPOSE: To alert the operator to not reestablish
feed flow to a faulted S/G if an intact or ruptured S/G is available to receive the feed flow.BASIS: Reestablishment
of feed flow to a S/G may result in thermal or mechanical
shocks to the S/G tubes that could result in tube leakage or tube rupture.If feed flow is reestablished
to a faulted S/G and tube leakage resulted, control of the leakage would not be possible until the S/G secondary boundary was restored.Flow restoration
to a non-faulted
S/G will provide an effective and controllable
secondary heat sink.STEP 2 Check if a secondary heat sink is required: PURPOSE: To check if a secondary heat sink is required for heat removal.BASIS: Before implementing
actions to restore flow to the S/Gs, the operator should check if secondary heat sink is required.For larger LOCA break sizes, the NC system will depressurize
below the intact S/G pressures.
The S/Gs no longer function as a heat sink and the core decay heat is removed by the break flow.For this range of LOCA break sizes, the secondary heat sink is not required and actions to restore secondary heat sink are not necessary.
For these cases, the operator returns to the procedure and step in effect.Since Step 19 directs the operator to return to Step 1 if the loss of secondary heat sink parameters
are not exceeded, break sizes that take longer to depressurize
the NC system will be detected on subsequent
passes through Step 1.If NC system temperature
is low enough to place the NO system in service in RHR mode, then the NO system is an alternate heat sink to the secondary system.Therefore, an attempt is made to place the NO system in service (Enclosure
2, Placing NO In RHR Mode)in parallel to the attempts to reestablish
feedwater flow.NC system pressure must be below normal NO system pressure limits.When adequate NO cooling is established, then the operator is directed to return to the procedure and step in effect.Generic Enclosure G-2 (Placing NO in RHR Mode)contains guidance to align one, or both trains, of NO in RHR Mode;leaving one, or no train, available for auto swap to sump;or leaving one train on sump and one train in RHR mode.The decision for alignment will be made with concurrence/guidance
from TSC, if available.
OP-MC-EP-FRH
FOR TRAINING PURPOSES ONL Y Page 27 of 169 REV.10
DUKE POWER MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours}OBJECTIVES
SNN LLL E OBJECTIVE L LPP 000 RSR Q R 0 0 1 Explain the purpose of each procedure in the FR-H series.XX EPFRHOO1 2 Discuss the entry and exit guidance for each procedure in the X X FR-H series.EPFRHOO2 3 Discuss the mitigating
strategy (major actions)of each X X X procedure in the FR-H series.EPFRHOO3 4 Discuss the basis for any note, caution or step for each XXX procedure in the FR-H series.EPFRHOO4 5 Given the Foldout page, discuss the actions included and the X X X basis for these actions.EPFRHOO5 6 Given the appropriate
procedure, evaluate a given scenario X X X describing
accident events and plant conditions
to determine any required action and its basis.EPFRHOO6 7 Discuss the time critical task(s)associated
with the FR-H X X X series procedures
including the time requirements
and the basis for these requirements.
EPFRHOO7 OP-MC-EP-FRH
FOR TRAINING PURPOSES ONL Y Page 5 of 169 REV.10
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#2----Group#1 KIA#073 G2.2.22 Importance
Rating 4.7----Equiprnent
Control: Kno\lvledge
of limiting conditions
for operations
and safety lirnits Proposed Question: SRO 89 Which ONE of the following describes (1)the MINIMUM radiation monitor requirement
that provides the preferred means of NCS primary to secondary leak rate monitoring
in accordance
with technical specification
surveillance
requirements, and (2)the MINIMUM sensitivity
required to ensure the monitor remains OPERABLE, in accordance
with SLC and bases?A.(1)EMF-33, Condenser Evacuation
Monitor OR N-16 Monitors, EMF-71EMF-74;(2)75 GPO.B.(1)EMF-33, Condenser Evacuation
Monitor OR N-16 Monitors, EMF-71EMF-74;(2)30 GPO.C.(1)EMF-33, Condenser Evacuation
Monitor AND N-16 Monitors, EMF-71-EMF-74;(2)75 GPO.O.(1)EMF-33, Condenser Evacuation
Monitor AND N-16 Monitors, EMF-71-EMF-74;(2)30 GPO.Proposed Answer: B Explanation (Optional):
A.Incorrect.
75 GPO is leakage defined by action in AP, but minimum monitors are correct B.Correct.c.Incorrect.
Not all, just'either or'.If EMF33 is operable, then the MS line monitors are not required to be operable, and vice-versa.
D.Incorrect.
OR statement would make this correct, because minimum Page 228 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 detectable
requirement
is correct SLC 16.7.6, Rev 99 Technical Reference(s)(Attach if not previously
provided)----------
Proposed references
to be provided to applicants
during None examination:
Learning Objective:
WE-EMF Obj 10 (Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge X Comprehension
or Analysis10 CFR Part 55 55.41 Content: 55.43 2 Comments: KA is matched because the SLC LCO for process radiation monitoring
is being evaluated, and further SRO knowledge is evaluated because the SRO must know basis for operability
of the detectors Page 229 of 260 Draft 7
Radiation Monitoring
for Plant Operations
16.7.6 16.7 INSTRUMENTATION
16.7.6 Radiation Monitoring
for Plant Operations
COMMITMENT
APPLICABILITY
The radiation monitoring
instrumentation
channels shown in Table 16.7.6-1 shall be OPERABLE.As shown in Table 16.7.6-1.REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION
TIME A.One or more radiation A.1 Adjust setpoint to within the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> monitoring
channels limit.AlarmlTrip
setpoint exceeding value shown OR in Table 16.7.6-1.A.2 Declare the channel 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable.
B.One Containment
B.1 Verify containment
purge Immediately
Atmosphere
Gaseous system (VP)valves are Radioactivity
monitoring
maintained
closed.channel inoperable.
C.One Control Room Air C.1 Isolate the associated
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Intake Radioactivity
Control Room Ventilation
monitoring
channel System (VC)outside air inoperable.
intake.(continued)
McGuire Units 1 and 2 16.7.6-1 Revision 99
Radiation Monitoring
for Plant Operations
16.7.6 REMEDIAL ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION
TIME D.One or more required 0.1 Suspend all fuel movement Immediately
channels for Spent Fuel operations
in the fuel Handling Area, Reactor handling area being Building Fuel Handling monitored until Required Area or New Fuel Vault Acton 0.2 is completed.
Fuel Handling Area Radiation Monitors AND inoperable.
0.2.1 Provide a portable Immediately
continuous
monitor with same Alarm Setpoint.OR 0.2.2 Provide RP continuous
Immediately
dose rate monitoring.
AND 0.3 Restore inoperable
30 days monitors to OPERABLE status.E.One Spent Fuel Pool E.1 Verify the Fuel Handling Immediately
Radioactivity
monitoring
Ventilation
System (VF)channel inoperable.
requirements
3.7.12 are met.F.Condenser Evacuation
F.1 Ensure that all N-16 Immediately
System Noble Gas Leakage Monitor (EMF-71 , Activity Monitor (EMF-72,73,&74)channels are 33)inoperable.
OPERABLE.G.One or more N-16 G.1 Ensure that the Condenser Immediately
Leakage Monitor (EMF-Evacuation
System Noble 71,72,73,&74)Gas Activity Monitor (EMF-channels inoperable.
33)is OPERABLE.(continued)
McGuire Units 1 and 2 16.7.6-2 Revision 99
Radiation Monitoring
for Plant Operations
16.7.6 CONDITION REQUIRED ACTION COMPLETION
TIME H.Condenser Evacuation
H.1 Initiate action to restore Immediately
System Noble Gas online radiation monitor to Activity Monitor (EMF-operable.33)inoperable.
AND AND H.2 Perform TS-SR 3.4.13.2.72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> One or more N-16 Leakage Monitor (EMF-71,72,73,&74)channels inoperable.
TESTING REQUIREMENTS
NOTE-----------------------------------------------------------
Refer to Table 16.7.6-1 to determine which TRs apply for each Radiation Monitoring
channel.TEST TR 16.7.6.1 Perform CHANNEL CHECK.TR 16.7.6.2 Perform CHANNEL OPERATIONAL
TEST.TR 16.7.6.3 Perform CHANNEL OPERATIONAL
TEST.TR 16.7.6.4 Perform a CHANNEL CALIBRATION.
FREQUENCY 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 92 days 184 days 18 months McGuire Units 1 and 2 16.7.6-3 Revision 99
Radiation Monitoring
for Plant Operations
16.7.6 TABLE 16.7.6-1 RADIATION MONITORING
INSTRUMENTATION
FOR PLANT OPERATION APPLICABLE
REQUIRED ALARMfTRIP
TESTING MONITOR MODES CHANNELS SETPOINT REQU I REMENTS 1.Containment
1,2,3,4,5,6
Must meet SLC TR 16.7.6.1 Atmosphere
Gaseous 16.11-6 TR 16.7.6.2 Radioactivity-High (Low limits TR 16.7.6.4 Range EMF-39)2.Spent Fuel Pool With irradiated1.7 x 10-4 TR 16.7.6.1 Radioactivity-High (EMF-fuel in fuel JlCi/ml TR 16.7.6.2 42)storage TR 16.7.6.4 areas or fuel bUilding 3.Spent Fuel Handling With fuel in fuel
TR 16.7.6.1 Area Radiation Monitor storage See Note (b)TR 16.7.6.3 (1 EMF-17, 2EMF-4)areas or fuel TR 16.7.6.4 building 4.Reactor Building Fuel 615 mR/hr TR 16.7.6.1 Handling Area Radiation See Note (b)TR 16.7.6.3 Monitor (1 EMF-16, TR 16.7.6.4 2EMF-3)5.New Fuel Vault Fuel With fuel in New
TR 16.7.6.1 Handling Area Radiation Fuel Vault See Note (b)TR 16.7.6.3 Monitors TR 16.7.6.4 (1EMF-20,1 EMF-21 , 2EMF-7,2EMF-8)
6.Control Room Air Intake 1,2,3,4,5,6
2 per station.3.4 x 10-4 TR 16.7.6.1 Radioactivity-High (EMF-JlCi/ml TR 16.7.6.2 43aand43b)TR 16.7.6.4 7.Condenser Evacuation
See Note (a)TR 16.7.6.1 System Noble Gas TR 16.7.6.3 Activity Monitor (EMF-33)TR 16.7.6.4 8.N-16 Leakage Monitor 1 (40-1 00%4 (1/steamline)
See Note (a)TR 16.7.6.1 (EMF-71, 72, 73&74)reactor TR 15.7.6.3 power)TR 16.7.6.4 (a)The setpoint is as required by the primary to secondary leak rate monitoring
program.(b)Setpoint can be elevated above 15 mR/hr based upon direction from approved station procedures.
McGuire Units 1 and 2 16.7.6-4 Revision 99
Radiation Monitoring
for Plant Operations
16.7.6 BASES The OPERABILITY
of the radiation monitoring
instrumentation
for plant operations
ensures that: (1)the associated
action will be initiated when the radiation level monitored by each channel or combination
thereof reaches its setpoint, (2)the specified coincidence
logic is maintained, and (3)sufficient
redundancy
is maintained
to permit a channel to beservice for testing or maintenance.
The radiation monitors for plant operations
senses radiation levels in selected plant systems and locations and determines
whether or not predetermined
limits are being exceeded.If they are, the signals are combined into logic matrices sensitive to combinations
indicative
of various accidents and abnormal conditions.
Once the required logic combination
is completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Ventilation
Systems.The condenser evacuation
system noble gas activity monitor (EMF-33)and main steam line N-16 monitors (EMF-71, 72, 73,&74)are used for online monitoring
of
secondary leak rate.These radiation monitors provide the preferred means to accomplish
Technical Specification
Surveillance
SR 3.4.13.2 while in Mode 1.For the condenser evaluation
system noble gas activity monitor (EMF-33)or main steam line N-16 monitor to be considered
operable for primary to secondary leakage monitoring
the monitor must be sensitive to a least 30 gallons per day (GPD)leakage rate.Fuel assemblies
are stored and handled in areas of the plant discussed below.Radiation monitoring
is provided for these areas to detect excessive radiation levels and will provide an alarm to alert personnel if a potential radiation hazard is present.1.Unit 1 and 2 Spent Fuel Pool;includes the cask pool area, the new fuel elevator, the fuel transfer tube area and the spent fuel storage are/racks.
2.Unit 1 and 2 Reactor Building;includes the fuel transfer tube area, the reactor core and the refueling canal.3.Unit 1 and 2 Fuel Building;includes the new fuel vault area.Performance
of Required Acton D.1 shall not preclude completion
of movement of a component to a safe position.When a fuel handling area radiation monitor channel becomes inoperable, an alternate means is required for determining
dose rate and alerting individuals
to excessive radiation levels.This can be accomplished
by either a portable monitor with same alarm setpoint located within the area monitored by the inoperable
channel or using Radiation.
Protection
personnel performing
continuous
monitoring
of area dose rate using a hand-held dose rate meter.This hand-held meter will not provide an alarm, but relies upon RP personnel to alert individuals
of excessive radiation levels.Certain evolutions
may result in a higher gamma dose rate field, resulting in the need to adjust the alarm setpoint above the nominal alarm/trip
setpoint (15 mR/hr).An approved station procedure controls adjustment
of this setpont to a higher value that still ensures individuals
are alerted to the presence of excessive radiation levels.McGuire Units 1 and 2 16.7.6-5 Revision 99
Radiation Monitoring
for Plant Operations
16.7.6 REFERENCES
3.4.13-RCS Operational
Leakage.2.NSD-513-Primary to Secondary Leak Monitoring
Program, Revision 5.3.1 OCFR50.68-Criticality
Accident Requirements
4.Duke letter dates July 29, 2004-RAI Response, TS 3.7.15 and TS 4.3 Changes.5.NRC Safety Evaluation
Report dated March 17,2005-Amendments
Nos.225/207 McGuire Units 1 and 2 16.7.6-6 Revision 99
DUKE POWER MCGUIRE OPERATIONS
TRAINING 5NNLL L E OBJECTIVELL P P 0 Q00 RSR R 0 0 10 Concerning
the Technical Specifications
/SLCs related to the EMFs:*Given the LCO or SLC title, state the LCO/X X X commitment
(including any COLR values)and applicability.
X X X*For any LCOs/SLCs that have action required within one hour, state the action.X X X*Given a set of parameter values or system conditions, determine if any Tech Spec LCOs or SLCs is(are)not met and any action(s)required within one hour.XXX*Given a set of parameter values or system conditions
and the appropriate
Tech Spec/SLC, determine required action(s).
X**Discuss the bases for a given Tech.Spec.LCO or SLC.*SRO ONLY WEEMF010 OP-MC-WE-EMF
.FOR TRAINING PURPOSES ONL Y Page 11 of 129
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#2-----Group#1 KIA#076 A2.02-----------
Importance
Rating 3.1-----Ability to (a)predict the irnpacts of the following malfunctions
or operations
on the SWS;and (b)based on those predictions, use procedures
to correct, control)or r11itigate
the consequences
of those rnalfunctions
or operations:
Service water header pressure Proposed Question: SR090 Page 230 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Given the following conditions:*A plant cooldown is in progress.*Current conditions
are: o NC Pressure-1400 psig o NC Temperature
-440 degrees F o Cold Leg Accumulators
have NOT been isolated o"B" Train in service An event occurs:*NC System pressure starts to go down at approximately
2 psi per minute.*PZR level is going down at 5%per minute.*Containment
Pressure is rising at 0.1 psig per minute.*Train"B" Safety Injection actuates.*Train"A" Safety Injection did NOT actuate.Which ONE (1)of the following describes (1)the impact on the unit, and (2)the action that must be taken?A.(1)NC Pumps will overheat due to loss of RN cooling (2)Enter E-O, Reactor Trip or Safety Injection, and initiate Train A Safety Injection to restore flow to Train A Essential Header and RBEssential Header B.(1)"A" EDG will overheat due to loss of RN cooling (2)Enter E-O, Reactor Trip or Safety Injection;
Reset SI Sequencers
and open RN Cross-Connect
valves to restore Train A Essential header C.(1)NC Pumps will overheat due to loss of RN cooling (2)Enter AP-34, Shutdown LOCA, and initiate Train A Safety Injection to restore flow to Train A Essential Header and RB Non-Essential
Header D.(1)"A" EDG will overheat due to loss of RN cooling (2)Enter AP-34, Shutdown LOCA;Reset SI Sequencers
and open RN Cross-Connect
valves to restore Train A Essential header Page 231 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Proposed Answer: A Explanation (Optional):
A is correct.B is incorrect.
Correct procedure to enter;do not open RN cross connect valves on a valid SI signal, and even if the action was performed, one train would not operate since the sequencer has not actuated C is incorrect.
Credible because the procedure would be entered in Mode 4 if NC pressure was lower.Action to restore RN is correct though D is incorrect.
Wrong procedure as in C above.Also wrong action.If both sequencers
were actuated, the action could work, but not performed for valid SI Technical Reference(s):
OMP 4-3, p8 Rev 26 E-O step 5 Rev 24;AP-34 Rev 13 DG-EQB Rev 16 ECC-CLA Rev 28 EP-EO Rev 12 (Attach if not previously
provided)Proposed references
to be provided to applicants
during None examination:
Learning Objective:
DG-EQB Obj 6;PSS_RN (As available)
Obj 16 Question Source: Bank#Modified Bank#New x (Note changes or attach parent)Question History: Question Cognitive Level: Last NRC Exam 2007 McGuire Memory or Fundamental
Knowledge Comprehension
or Analysis x10 CFR Part 55 55.41 Content: Page 232 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet 55.43 5 Form ES-401-5 Comments: KA is matched because conditions
represented
by the stem indicate loss of header pressure on 1 header.SRO level because the applicant must assess plant conditions
and determine procedure use based upon selected impact Page 233 of 260 Draft 7
DUKE POWER Auxiliary Building RV loads: Auxiliary Building Ventilation
Units Reactor Building RV loads: Upper containment
ventilation
units Lower containment
ventilation
units 2.5 Discharge Paths Objective#8 MCGUIRE OPERATIONS
TRAINING The normal RN system discharge path is to the RC crossover header
returns to Lake Norman.The SNSWP is also a discharge path however it is typically only used if the suction is also from the SNSWP to prevent undesirable
changes to SNSWP level.The VC/YC chillers*RN discharge headers have been modified so that they will normally discharge into the shared RN discharge headers.This will prevent having to declare a VC/YC train inoperable
because the Unit1A or1B RN Essential header is isolated.The new flowpath will be the normally aligned path however, the old flowpath will still be available.
286 Valves I Objective#12 I 2.6.1 Blackout and Safety Injection Signals The following is a listing of the various RN valves and how they respond to Safety Injection and/or Blackout signal(s).
Valves which are shared between the units (ORN)can be powered and controlled
from either unit.(refer to Drawing 7.5)The following valves receive autoclosesignals
upon receipt of either Unit 1 or 2 blackout or safety injection:
- ORN-2B (Train1A&2A RC Supply)*ORN-3A (Train1A&2A RC Supply)*ORN-4AC (Train18&28 RC Supply)*ORN-5B (Train18&28 RC Supply)*ORN-7A (Train 1A&2A SNSWP Supply)*ORN-149A (Train 1A&2A Disch to SNSWP)*ORN-11B(Train18&28 LLI Supply)*1 RN-41B(Train 18 to Non-Ess Hdr Isol)Controlled
only from Unit 1*1 RN-43A (Train1B to Non-Ess Hdr Isol)Controlled
only from Unit 1*2RN-41B(Train 28 to Non-Ess Hdr Isol)Controlled
only from Unit 2*2RN-43A (Train 28 to Non-Ess Hdr Isol)Controlled
only from Unit 2*ORN-284B (Train18&28 Disch to RC)OP-MC-PSS-RN
FOR TRAINING PURPOSES ONL Y Page 35 of 111
0-...J t:J....-p 01III::c-0-0 0 CI)**CA,KC,KF,KD
ZC().g.g UNIT 2 en::n A ESSENTIAL HEADER TRAIN A'<:0<: RC I CA PUMP COOLERS NI PUMP RETURN tn 2B 3A KDHX NO PUMP++.....KC PUMP COOLERS NS PUMP (1)VC CHILLERS cCP 3 KCHX NSHX'" I I KFAHU RC c: SNSWP...I FOR SPECIFIC LOADS REFER TO 296A 147AC 148ACI SECTION 2.4 P::+-P P P++**-LEGEND en UNIT 2 RB NON-ESSENTIAL
HEADER Q)-h-----.....A TRAIN NCP MOTOR AIR COOLER*-CLOSES on Ss or BLACKOUT (1)12AC 13A+-OPENS on Ss or BLACKOUT.....::nS-CLOSES on Ss'<P-CLOSES on Sp-;j AB NON-ESSENTIAL
HEADERPDP.....:ti (1)n Low=(Q:E Level*01 Intake*0-0 r-Ole:: AB RVLOADS 0 o::n*SNSWP S 279B 299A Q)*RB RVLOADS C. tn.....CI)UNIT 2+m BTRAIN Q)10AC 11B RV I J 152B0 PUMP S*C.DISCHARGE<+"'<.., r--1 B ESSENTIAL HEADER Q)RC C)SAME AS TRAIN A ESS HDR<297B 283AC 284B (1)SNSWP D"<3 II;;)J UNIT 2:0 TRAIN B r-111 RETURN 0 0 (Q-0_.n RC P4ij P4iii:ti.4AC 58 0:::!0::nen:ti 111'--'" 5E(,,)C')
DUKE POWER MCGUIRE OPERATIONS
TRAINING 3.2.3 Safety Injection Alignment On receipt of a Safety Injection signal basically the same automatic actuation occurs as after a blackout.The exceptions
are that the supply to all nonessential
equipment except the NC pump motor coolers and crossovers
between essential trains are isolated.The IIAII RN pump supplies Reactor Building non-essential
header.The RV pumps will start automatically
and supply the containment
ventilation
units if a blackout does not occur concurrently
with the LOCA.Drawings 7.14 and 7.15 provides the flow path for a unit safety injection.
NOTE: An Ss signal will affect both units suction, discharge and AB non-essential
headers.Refer to Drawing 7.14 On receipt of a Phase B isolation signal (Sp)the RV pump suction is isolated to conserve water.The containment
isolation valves close to isolate the NC pump motor coolers.All nonessential
supply is isolated providing double isolation at this time between all essential and nonessential
equipment.
The NS heat exchanger inlet isolation valve is opened from the control room when required.During all modes of operation, water is available for assured makeup.Drawings 7.16 provides the flow path following a unit safety injection with a phase B signal.4.0 TECHNICAL SPECIFICATIONS
I Objective#17 I 4n1 Tech Spec 3.7.7 Nuclear Service Water System (NSWS)4.2 Tech Spec 3.7.8 Standby Nuclear Service Water Pond (SNSWP)OP-MC-PSS-RN
FOR TRAINING PURPOSES ONL Y Page 53 of 111 REV.39
DUKE POWER MCGUIRE OPERATIONS
TRAINING 3.2 Abnormal and Emergency Operation 3.2.1 Abnormal Procedure AP/1 or2/A/5500/20
AP20 purpose, Cases, Symptoms, and basis for steps is covered thoroughly
in the AP Lesson Plan.Objective#16 3.2.2 Blackout Alignment Blackout is a loss of power to the 4160 vac bus.When the low voltage condition is detected, the DIG will start and the sequencer will load the Blackout loads onto the bus.On receipt of a Blackout signal, Train A valves automatically
assume low level alignment;
Train B assumes SNSWP alignment.
Many shared valves receive signals from both units to prevent loss of water from SNSWP.Isolation valves for all heat exchangers
which are needed open automatically
and the train related RN pump will start.All nonessential
discharge is isolated except the containment
vent units and NC pump motor cooler discharge.
The containment
vent units and the NC pump motor coolers are supplied with cooling water from"A" RN pump.The IIAII RN pumps supply the containment
ventilation
units with cooling water because they have more NPSH since their suction is aligned to the LLI and because the RV pumps may not have power.Drawings 7.10 and 7.11 provides the unit blackout flow path.Drawings 7.12 and 7.13 provides the flow path for Train A and Train B Blackout respectively.
If a Blackout occurs on the opposite unit, the non-blackout
unit will have itsessential header isolated from the B RN pump as a result of RN41 Band RN43A closing (Refer to Drawing 7.5).In order to supply the non-essential
header on theblackout unit, the A Train RN pump must be started.OP-MC-PSS-RN
FOR TRAINING PURPOSES ONL Y Page 51 of 111 REV.39
DUKE POWER MCGUIRE OPERATIONS
TRAINING STEP 8 Check proper CA Pump status: PURPOSE: To ensure proper status of the CA pumps.BASIS: The MD CA pumps start automatically
on an SII signal to provide feed to the S/Gs for decay heat removal.If S/G levels drop below the appropriate
setpoint, the TD CA pump will also automatically
start to supplement
the MD pumps.STEP 9&10 Check all KC and both RN pumps-ON.PURPOSE: To ensure KC and RN pumps are running.BASIS: KC and RN pumps provide cooling to certain safeguards
components.
STEP 11 Notify Unit 2 to start 2A RN Pump.PURPOSE: To ensure required cooling.BASIS: 80th units'RN train cross ties close on a single unit 8/1.If 8 train was previously
feeding the reactor building headers on opposite unit, starting opposite unifs A RN pump ensures the reactor building headers remain cooled (only A train is aligned to reactor building headers following an S/I).Note that RV will continue to cool the opposite units reactor building headers, unless RV suction was isolated by a Phase 8 signal.Even if RV is supplying the reactor building headers, starting A train RN ensures desired flow rate to these headers.OP-MC-EP-EO
FOR TRAINING PURPOSES ONL Y Page 33 of 207 REV.12
MNS EP/1/A/5000/E-O
UNITl REACTOR TRIP OR SAFETY INJECTION PAGE NO.4 of 36 Rev.24 ACTION/EXPECTED
RESPONSE o Check if S/I is actuated: a.IISAFETY INJECTION ACTUATED II status light (1 SI-18)-LIT.RESPONSE NOT OBTAINED a.Perform the following:
1)Check if S/I is required:_.pzr pressure less than 1845 PSIG OR_.Containment
pressure greater than 1 PSIG._2)IF S/I is required, THEN initiate Silo 3)IF S/I is not required, THEN:*Implement EP/1/A/5000/F-O (Critical Safety Function Status Trees).*GO TO EP/1/A/5000/ES-O.1 (Reactor Trip Response).
b.Both LOCA Sequencer Actuated status lights (1SI-14)-LIT.6.Announce"Unit 1 Safety Injection ll*b.Initiate Silo
OMP4-3 Page 8 of 35 7.5 Manual Initiation
of Safeguards
Actions In most scenarios, ROs and SROs are expected to manually initiate safeguards
actions if an automatic action setpoint is being approached, to avoid challenging
the automatic safeguards
function.An example of this is to manually initiate safety injection if pressure is decreasing
in an uncontrolled
manner to 1845 psig.Exceptions
to this philosophy
are listed below:*Do NOT initiate Phase BfContainment
Spray earlier than required.Early initiation
of spray has the adverse affect of transferring
FWST water to the containment
sump and causing earlier transfer to Cold Leg Recirc.{NRC Bulletin 2003-01 response}*During an ATWS, it is undesirable
to initiate Sf I in"anticipation" of an Sf I signal if the reactor will NOT trip, sincethiswill cause a loss of CF flow to the SfGs.This exception is stated in the APs that manually initiate Sf I in"anticipation" of an Sf I signal.The operator is expected to manually initiate any action which should have automatically
occurred if the automatic function fails, such as the Safety Injection fails to initiate during an uncontrolled
Reactor Coolant depressurization
at 1845 psig (even during an ATWS)or an ECCS pump fails to start on a Safety Injection signal.IF directed to initiate a signal, initiate both trains unless otherwise specified.
7.6 Resetting Safety Systems IF directed to reset a signal, reset both trains unless otherwise specified.
IF a procedure directs resetting a signal that has NOT been received or that has been previously
reset, the reset pushbuttons
do NOT have to be depressed since the intent of the step has been met.Likewise, if a procedure directs the operator to stop,startor reposition
a component which is already in the desired position;the component's
control switch does NOT have to be depressed.
DUKE POWER 1.0 INTRODUCTION
MCGUIRE OPERATIONS
TRAINING 1.1 Purpose I Objective#1 The Diesel Generator Load Sequencing
System (EQB)functions to energize the necessary Blackout and/or Safety Injection loads in such a manner that the diesel generator or auxiliary transformer (ATC, ATD, SATA, SATB)is not momentarily
overloaded.
I Objective#2 I A power loss to the 4160 Volt Bus or a Safety Injection Actuation Signal from the Solid State Protection
System (SSPS)actuates the Load Sequencer.
1.2 General Description
The sequencer has basically two modes of operation:
priority and secondary.
The priority mode is actuated by a safety injection (SI)signal from the Solid State Protection
System (SSPS).The secondary mode is actuated by a loss of voltage (LOV)on the 4160 volt essential bus.The Sequencing
System is designed to be actuated automatically
without any operator action and to initiate loading of the Engineered
Safeguard bus as rapidly as loading transients
permit without overloading
the normal transformer
or diesel generator.
The controlling
parameters
of sequencer logic are the ESF signal from SSPS, the time from initial actuation, the voltage on the ESF Bus and the Diesel Generator frequency (speed).1.3 Redundancy
requirements
There are two identical systems, one associated
with each diesel.They are independent
of each other and in no way can the failure of one affect the other.The single failure is considered
to be the entire loss of one system.1 n4 Sequencer Actuation Signals Signal Setpoint Coincidence
Interlock Protection
Manual Safety 1/2 Switches Operator Injection Judgment Low Pressurizer
1845 psig 2/4 Channels P-11 LOCA Pressure High 2/3 Pressure Steam Break Containment
1.0 psig Switches LOCA Pressure 2/3 Under-voltage
on affected 4160 Volt Bus (Blackout)
OP-MC-DG-EQB
FOR TRAINING PURPOSES ONL Y Page 11 of 69 REV.16
DUKE POWER 2.0 SYSTEM DESCRIPTION
2.1 Sequencer Modes of Operation Objective#3 MCGUIRE OPERATIONS
TRAINING The Sequencer has basically two modes of operation;
The Priority Mode of operation is actuated by a Safety Injection signal from the SSPS.When Safety Injection is actuated, the signal seals in and sequencing
begins immediately.
The Secondary Mode of operation is actuated by a 2/3 phase Loss of Voltage (LOV)on the 4160 Volt Essential Bus.Upon actuation, the sequencer starts the diesel and goes through an 8 second test for verification
of a Blackout.If a Blackout does not exist, the Sequencer will automatically
reset to its initial operating state and the Diesel Generator must be manually shut down.For an actual Blackout, the signal is sealed in, the 4160V bus normal and alternate incoming breaker is tripped, the 4160 Volt Essential Bus is load shed, and the Diesel Generator Breaker is closed provided the Diesel Generating
unit has attained 95%speed.I Objective#4 I When both actuation signals (LOV and SI)are present simultaneously, the Sequencer will select the SI logic and perform those functions necessary to sequence that mode (Le., load shed, sequencer reset, removing blackout logic, and energizing
SI loads).This is also true when the Loss of Voltage condition was initiatedbythe Degraded Voltage relaying.If an SI signal were present following the completion
of the 9.7 second alarm timer cycle, the 4160 Volt Normal and Standby incoming circuit breakers would trip immediately.
This causes the SI loads to be connected to the Diesel Generator initially, therefore ensures a reliable power supply for the Essential Auxiliary loads.The Sequencer is designed to initiate loading of the 4160 Volt Essential Bus as rapidly as loadingtransientspermit
without overloading
the Normal Transformer
or Diesel Generator.
2.2 System Protection
Each unit is protected from abnormal voltage conditions
by two levels of voltage protection, Loss of Voltage and Degraded Voltage.For each train, there are three Loss of Voltage and three Degraded Voltage relays connected in 2/3 logic.Another relay is provided which is used as a permissive
for the Load Sequencer Accelerated
Sequence Mode (127 AX Special).All relays affect Sequencer operation.
The Degraded Voltage relays are set to operate at 89%of nominal bus voltage on U2 which is 3703 Volts.On U1 the Degraded Voltage relays are set to operate at 88.4%or 3678.5 Volts.The relays are a high accuracy type with a small reset dead band.These relays use time delays before initiating
any actions.With a 4160 Volt bus de-energized, the Degraded Voltage relays must be placed in TEST before the Normal and Standby circuit breakers can be operated.OP-MC-DG-EQB
FOR TRAINING PURPOSES ONL Y Page 13 of 69 REVu 16
MNS AP/1/A/5500/34
UNITl ACTION/EXPECTED
RESPONSE A.Purpose SHUTDOWN LOCA PAGE NO.1 of 119 Rev.13 RESPONSE NOT OBTAINED Provide actions for protecting
the reactor core in the event of a LOCA that occurs during either Mode 3 after the Cold Leg Accumulators
are isolated or Mode 4.
DUKE POWER 2.2.Discharge Piping MCGUIRE OPERATIONS
TRAINING A 10 inch line connects each Accumulator
to a cold leg.Each line contains two (2)check valves in series, one (1)normally open isolation valve and a flow restrictor.
The flow restrictor
is installed on the outlet of each accumulator
and ensures accumulator
discharge line resistance
is within ECCS analysis tolerance band.2.3.Isolation Valves One (1)motor operated valve per accumulator
provides isolation of CLA.The valves are normally opened with power removed prior to exceeding 1 000 psig, and with NCS temperature
between 400 and 425 degrees.Objective#4 Alarms on the Control Board alert the operator when a valve is less than fully open (ACCUM ISOL NOT FULLY OPEN).This alarm is not active<P-11.The control circuitry for each valve is equipped with a disconnect/enable
switch which allows isolation of the motor from the power source to prevent inadvertent
operation.
Removal of power to the valves is required by Tech Specs because the valves fail to meet single failure criteria.OPEN/CLOSE
pushbuttons
are located on the Control Board.Objective#3 The valves are designed to automatically
open at>P-11 setpoint (1955 psig)or on a S8 signal if closed and power is available to the valve.The valve motors are powered from EMXA and EMXB.Power supplies are as follows: Valve Number Unit 1 Unit 2 NI-54A 1 EMXA-2 Compo 3A 2EMXA-2 Compo 3A NI-65B 1 EMXB-4 Compo 2C 2EMXB-4 Compo 2C NI-76A 1 EMXA-2 Compo 3B 2EMXA-2 Compo 3B NI-88B 1 EMXB-4 Compo 3D 2EMXB-4 Compo 3D 2.4.Check Valves Swing check valves are installed in the discharge line to prevent flow from the Reactor Coolant System to the accumulator.
These valves open at of 0.5 psi (upstream to downstream)
2.5.Relief Valves Relief valves are installed on each accumulator
to prevent over-pressurization.
Sized to pass more than makeup capability, these valves are designed to pass N 2 or water.Relief valves are set at 700 psig.OP-MC-ECC-CLA
FOR TRAINING PURPOSES ONL Y Page 15 of 35 REV.28
DUKE POWER MCGUIRE OPERATIONS
TRAINING OBJECTIVESNNL L L OBJECTIVELL P P 000 R S R R 0 0 16 Explain how a Safety Injection or Blackout on either unitXXX X X affects that and the other unit during normal operations
and what action the operator must perform.17 Concerning
the Technical Specifications
related to the RN System:*Given the LCO title, state the LCO (including anyXX X COLR values)and applicability.
- For any LCO's that have action required within one X X X hour, state the action.*Given a set of parameter values or system conditions, XXX determine if any Tech Spec LCO's is(are)not met and any action(s)required within one hour.*Given a set of parameter values or system conditionsXX X and the appropriate
Tech Spec, determine required action(s).
- Discuss the bases for a given Tech.Spec.LCO or X*Safety Limit.*SRO ONLY OP-MC-PSS-RN
FOR TRAINING PURPOSES ONL Y Page 11 of 111 REV.39
DUKE POWER MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours)NLO NLOR LPRO LPSO LOR 1.01.51.51.51.5 OBJECTIVESNN LLL OBJECTIVELL P P 0 0 0 R S R R 0 0 1 State the purpose of the Diesel Generator Load Sequencing
X X X X System.2 List the Sequencer Automatic Actuation Signals.X X XXX 3 List the two Sequencer Modes of Operation and give a briefXX XXX explanation
of each mode.4 State which of the Sequencer Modes has priority.X X X X X 5 Describe the sequence of events which occur during the X X X Blackout Mode of Sequencer Operation.
6 Describe the sequence of events which occur during the XXX Safety Injection Mode of Sequencer Operation.
7 Describe the sequence of events which occur during a X X X Blackout followed by a Safety Injection.
8 Describe the sequence of events which occur during a Safety X X X Injection Actuation followed by a Blackout.(NOTE: with Ss reset and with Ss not reset).9 Describe the sequence of events required to be done in order X X X to return the 4.16 KV bus back to normal following a:*Safety Injection*Blackout*Safety Injection followed by a Blackout*Blackout followed by a Safety Injection 10 Given a Limit and/or Precaution
associated
with an operating X X X X X procedure, discuss its bases and when the it applies.OP-MC-DG-EQB
FOR TRAINING PURPOSES ONL Y Page 5 of 6916
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#2----Group#2 KIA#028 A2.03----------
Importance
Rating 4.0----Ability to (a)predict the ilnpacts of the following malfunctions
or operations
on the HRPS;and (b)based on those predictions, use Procedures
to correct, control: or mitigate the consequences
of those rnalfunctions
or operations:
The hydrogen air concentration
in excess of limit flarne propagation
or detonation
with resulting equipment dam-age in containrnent
Proposed Question: Given the following:
SR092*A LOCA has occurred on Unit 2.*Due to subsequent
failures, the crew isperformingactions
contained in FR-C.1, Response to Inadequate
Core Cooling.*Hydrogen Analyzers are in service.*Hydrogen ignitersareOFF.*NF AHUs are OFF.*Containment
Hydrogen Concentration
is currently 3%and rising slowly.Which ONE of the following describes the action required, and the reason for the action, in accordance
with FR-C.1?A.Place hydrogen igniters in service;do NOT operate Hydrogen recombiners;
recombiner
operating temperatures
may cause a challenge to containment
integrity due to hydrogen flammability.
B.Place hydrogen igniters and hydrogen recombiners
in service;containment
hydrogen concentration
is below thelimitcausing
concern for containment
integrity violations
due to hydrogen ignition.C.Do NOT place hydrogen igniters OR hydrogen recombiners
in service;consult management
for recommendation
related to hydrogen reduction.
Operation of either component may result in a challenge to containment
integrity.
D.Place hydrogen recombiners
in service;do NOT operate Hydrogen igniters;igniters must be placed in service prior to hydrogen concentration
reaching 0.5%, because ignition above that concentration
may cause a Page 236 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 challenge to containment
integrity.
Proposed Answer: B Explanation (Optional):
Per the basis document of FR-C.1, step 4, if hydrogen concentration
is between 0.5%and 6%, there is limited burn potential.
Therefore, both the igniters and the recombiners
are placed in service.If hydrogen is less than 0.5%, a flammable situation is not imminent, so the igniters are placed in service.If hydrogen is greater than 6%there is a potential explosive mixture.Hydrogen concentration
must be reduced in other ways before starting the recombiners
or igniters.A.Incorrect.
Recombiners
are allowed to be started below 6%B.Correct.C.Incorrect.
Action is correct for>6%concentration
D.Incorrect.
Igniters will be placed in service as well as recombiners
if concentration
is below 60/0 Technical Reference(s)FR-C.1 step 4 Rev 5 and basis EP-FRC Rev10 Proposed references
to be provided to applicants
during None examination:
Learning Objective:
EP-FRC Obj 6 (Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge Comprehension
or Analysis X 10 CFR Part 55 55.41 Content: 55.43 1,5 Page 237 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Comments: KA is matched because it evaluates operation of equipment to keep hydrogen concentration
below the explosive limit.SRO only because the applicant must know the design and procedural
basis for operation of hydrogen igniters and recombiners
Page 238 of 260 Draft 7
RESPONSE NOT OBTAINED ACTION/EXPECTED
RESPONSE (MNS EP/2/A/5000/FR-C.1
UNIT 2 RESPONSE TO INADEQUATE
CORE COOLING PAGE NO.5 of 50 Rev.5
DUKE POWER FR-C.1 Response to Inadequate
Core Cooling MCGUIRE OPERATIONS
TRAINING (STEP 4 PURPOSE: Check containment
H 2 concentration:
OP-MC-EP-FRC
FOR TRAINING PURPOSES ONL Y Page 29 of 111 REV.10
DUKE POWER MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours}NLO NLOR LPRO LPSO LOR 3.0 3.0 2.0 OBJECTIVES
5NN L L L E OBJECTIVELLPP 000 R 5 R Q R 0 0 1 Explain the purpose of each procedure in the FR-C series.XX EPFRCOO1 2 Discuss the entry and exit guidance for each procedure in the X X FR-C series.EPFRCOO2 3 Discuss the mitigating
strategy (major actions)of eachXX X procedure in the FR-C series.EPFRCOO3 4 Discuss the basis for any note, caution or step for eachXX X procedure in the FR-C series.EPFRCOO4 5 Given the Foldout page, discuss the actions included and the X X X basis for these actions.EPFRCOO5 6 Given the appropriate
procedure, evaluate a given scenario X X X describing
accident events and plant conditions
to determine any required action and its basis.EPFRCOO6 7 Discuss the time critical task(s)associated
with the FR-CXX X series procedures
including the time requirements
and the basis for these requirements.
EPFRCOO7 OP-MC-EP-FRC
FOR TRAINING PURPOSES ONL Y Page 5 of 111 REV.10
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#3----Group#1 KIA#2.1.35----------
Importance
Rating 3.9----Knowledge of the fuel-handling
responsibilities
of SROIS.Proposed Question: SRO 94 Unit 1 is in Mode 6, core alterations
are in progress.Which ONE of the following, by title, must approve bypass of a Fuel Handling interlock not specified in accordance
with procedures
for routine fuel handling activities?
A.Shift Manager B.Fuel Handling SRO C.Refueling Supervisor
D.Reactor Engineer Proposed Answer: B Explanation (Optional):
A.Incorrect.
Shift Manager responsible
for unit, but FH SRO is responsible
for all refueling activities
B.Correct.c.Incorrect.
Administrative
oversight required, but not approval for FH bypass D.Incorrect.
Nuclear Engineers will be involved in the core alterations, but are not part of approval for FH bypass;they are only approval authority during physics testing NSD-414 Rev 2 Technical Reference(s)(Attach if not previously
provided)----------
FH-FC Rev 18 Proposed references
to be provided to applicants
during None--------Page 242 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 examination:
Learning Objective:
FH-FC, 1 and 5 (Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge X Comprehension
or Analysis 10 CFR Part 55 55.41 Content: 55.43 6,7 Comments: KA is matched because the item evaluates a decision by refueling SROs.SRO level because knowledge of SRO responsibilities
during refueling is 10CFR55.43 (b)item 6/7 specific Page 243 of 260 Draft 7
VERIFY HARD COpy AGAINST WEB SITE IMMEDIATELY
PRIOR TO EACH USE Nuclear Policy Manual-Volume 2 NSD 414 9.Responsible
for PMs/PTs on all Fueling Handling equipment.
10.Responsible
for maintaining
and troubleshooting
all Fuel Handling equipment.
11.Responsible
for preparing, loading, and transporting
of Dry Storage Canisters.
12.Qualified/certified to Fuel Handling procedures
for assigned equipment.
13.Perform procedures
related to SNM (Special Nuclear Material)inventory control related to fuel.14.Install and maintain communications
systems required for refueling activities (installation
and checkout).
15.Maintain underwater
lights.16.Support special projects as needed.17.Perform all fuel handling activities.
18.Operate overhead cranes and hoists as necessary during fuel handling activities
19.Establish and maintain housekeeping, material condition, and FME controls of all fuel handling areas.This includes Upper Containment
Refueling Canal Area, Spent Fuel Pools and Fuel Receiving Areas.414.2.6 A.B.C.414.2.7 FUEL HANDLING ADVISORS (VENDOR)Provide expertise for fuel handling activities (cranes, hoists, tooling, including industry knowledge, etc.).Participate
as an active member of the Fuel Handling Team.Can perform the following:
- Review procedures.
- Provide"hands on" work as requested and approved by the Job Sponsor.OPERATIONS
SHIFTMANAGER(OPS)
NOTE: Operations
is responsible
for performing
the SOER 91-01 Briefing for core reload.A.During fuel movement, fuel receipt, special projects, and dry cask storage: 1.Ensure SRO's/RO's
are cognizant of all fuel handling activities
in progress or planned.2.Maintain awareness of any activities
that could impact fuel handling activities
and ensure appropriate
fuel handling personnel are aware of these activities.
3.Ensure appropriate
response and notifications
to any abnormal fuel handling event and verify any Technical Specification
implications.
4.Has ultimate responsibility
for the safety of the reactor core and fuel stored on site.5.Ensure the 91-01 Briefing is performed prior to core reload.414.2.8 CONTROL ROOM SRO AND RO (OPS)A.During fuel movement, fuel receipt, special projects, and dry cask storage: 1.Monitor the Nuclear Instrumentation
during core alterations.
2.Implement any responses required by Abnormal Procedures.
REVISION 2 5 VERIFY HARD COpy AGAINST WEB SITE IMMEDIA TEL Y PRIOR TO EACH USE
VERIFY HARD COpy AGAINST WEB SITE IMMEDIATELY
PRIOR TO EACH USE NSD 414 Nuclear Policy Manual-Volume 2 3.Log, verify, and maintain Technical Specification
for Mode 6, Core Alterations, and other Technical Specifications
for Spent Fuel Building activities.
4.Maintain awareness of fuel handling and SpentFuelBuilding
activities (i.e.-logging, turnover, etc.).5.Maintain awareness of core configuration
during core alterations.
6.Ensure reactivity
monitoring
is performed during refueling.
414.2.9 A.414.2.10 A.B.REFUELING SRO RESPONSIBLE
FOR FUEL HANDLING During core alterations:
1.Shall be present in the Reactor Building to observe and provide oversight of fuel handling activities
anytime Core Alterations
are being performed.
2.Shall have an SRO License or a SRO license limited to fuel handling.3.Maintain a working knowledge of procedures
and Technical Specifications
associated
with fuel handling and command immediate action as required.4.Approve use of fuel handling bypass interlocks
as necessary when not specified by an approved procedure.
5.Approve alternate fuel assembly moves as recommended
by Reactor Engineering.
6.The Refueling SRO should be stationed on the refueling bridge any time Fuel Assemblies
are being moved in the Reactor.7.The Refueling SRO will ensure the following:
a)Fuel Handling Procedures
are performed as written.b)All refueling personnel adhere to STAR Self-checking
techniques, procedure use and adherence, communication
standards and independent
verification.
c)Understands
the need for and approve all contingency
actions which may be required, in accordance
with Maintenance
procedures
for operating the Reactor Building Manipulator
Crane.d)Direct Reactor building Activities
during performance
of Abnormal Procedures.
e)No activities
occur that adversely affect reactivity
control.t)Foreign Material Exclusioncontrolsare
implemented
per NSD 104 in the Refueling Canal area and that all housekeeping
standards are maintained.
g)Assure approved safety practices are followed during operation of the Manipulator
Crane.h)Suspend all refueling operations
anytime he/she thinks refueling operations
are not being performed correctly or safely.TRAINING Develop and maintain initial training for designated
Maintenance, Operations, and contract personnel on fuel handling topics.Assist in the development
of Just in Time (JITT)on relevant fuel handling topics using the systematic
approach to training (SAT)process.Provide this training for the above designated
personnel to maintain a well qualified work force for safe and efficient fuel handling operations
and to maintain awareness of NSD 414 and fuel handling related issues.6 REVISION 2 VERIFY HARD COpy AGAINST WEB SITE IMMEDIA TEL Y PRIOR TO EACH USE
VERIFY HARD COpy AGAINST WEB SITE IMMEDIA TEL Y PRIOR TO EACH USE NSD 414 Nuclear Policy Manual-Volume 2*Provide periodic oversight as required.*Be present at the 91-01 briefing.414.2.4 A.414.2.5 A.REACTOR SERVICES (MNT)SUPERVISOR
During fuel movement, fuel receipt, special projects, and dry cask storage: 1.Maintain responsibility
and control of all activities
in the fuel handling areas.This includes the Spent Fuel Pools and the Fuel Receiving Areas.2.Ensure housekeeping
and material condition, and FME controls of all fuel handling areas is maintained.
This includes Upper Containment
Refueling Canal, Spent Fuel Pools, and Fuel Receiving Areas.3.Provide work direction of Fuel Handling Team in support of fuel handling activities.
4.Coordinate
interface with other groups during fuel handling activities.
5.Maintain ownership of fuel handling maintenance
procedures.
Ensure fuel handling maintenance
procedures
are developed and enhanced.6.Act as a contact for scheduling
all fuel handling PM's.Ensure all PM's are scheduled in a coordinated
manner.7.Performance
management
of the Fuel Handling Team.8.Ensures qualifications
and training requirements
are met.9.Maintain list and location of handling tools and equipment.
10.Perform all required fuel handling equipment PM's.11.Provide ownership for Fuel Handling ETQS tasks.12.Communicate
any Fuel Handling issues to Operations
in a timely manner.13.Assist in providing the following SOER 91-01 oversight for core reloading:
- Ensure Management's
expectations
are met.*Provide periodic oversight as required.*Be present at the 91-01 briefing.MAINTENANCE
FUEL HANDLING TECHNICIANS
During fuel movement, fuel receipt, special projects, and dry cask storage: 1.Operate fuel handling equipment in accordance
with approved procedures.
2.Operate the fuel transfer system.3.Operate all Fuel Handling tools.4.Operate the Spent Fuel Pool bridge during fuel handling activities.
5.Operate the Reactor Building main crane during fuel handling activities.
6.Provide Spotter for the Spent Fuel Pool bridge during fuel handling activities.
7.Provide Spotter in the Reactor Building during Fuel Handling activities.
8.Monitor the camera installed at the spent fuel pool up-enders to ensure that the correct fuel assembly is in transit to the reactor building during refueling.
4 REVISION 2 VERIFY HARD COpy AGAINST WEB SITE IMMEDIA TEL Y PRIOR TO EACH USE
VERIFY HARD COpy AGAINST WEB SITE IMMEDIA TEL Y PRIOR TO EACH USE Nuclear Policy Manual-Volume 2 414.FUEL HANDLING 414.1 INTRODUCTION
NSD 414 The purpose of this NSD is to identify the roles and responsibilities
for Fuel Handling activities
at the three nuclear sites.Mechanical
Maintenance (Reactor Services)is the owner of Fuel Handling and performs operation of all the tools and equipment used to move and manipulate
fuel and components.
Reactor Engineering
provides the core designs, configuration
control, and assists with the coordination
and oversight of fuel handling activities.
Operations
provide oversight and ensure reactivity
management
during fuel manipulations.
414.2 ROLES AND RESPONSIBILITIES
414.2.1 REACTOR ENGINEERING (RES)A.During Fuel Movement: 1.Determine fuel movement sequence.2.Determine acceptable
storage locations per Tech Specs.3.Ensure reactivity
monitoring
is performed during refueling.
4.Provide technical oversight of controlling
procedure during unload and reload.S.Assist with pre-job briefings as required.6.Provide instructions
for alternate moves as required.(must be approved by FH SRO).7.Provide technical assistance
during foreign object retrieval.
8.Perform plant engineering
roles (i.e., core verification, gap alignments, ensure SNM database is updated, etc.)in accordance
with Equipment Reliability
Program (NSDI20)NOTE: Operations
is responsible
for performing
the SOER 91-01 Briefing for core reload.9.Assist in providing the following SOER 91-01 oversight for core reload:*Ensure Management's
expectations
are met.*Provide periodic oversight as required.*Be present at the 91-01 briefing.B.During Fuel Receipt: 1.Serve as point contact for scheduling
fuel and/or component receipt (i.e.-vendor interface).
The dates are established
and provided to the FH Supervisor
for further notification
and follow up.2.Prepare documentation
for new fuel receipt.3.Ensure QA inspection
has been performed.
4.Interface with GO/vendor for evaluation
of any defects found.S.Responsible
for loading patterns in the Spent Fuel Pool.6.Ensure the SNM (Special Nuclear Materials)
database is updated C.During Special Projects: REVISION 2 1 VERIFY HARD COpy AGAINST WEB SITE IMMEDIATELY
PRIOR TO EACH USE
DUKE POWER 1.0 INTRODUCTION
1.1.Fuel Handling Overview MCGUIRE OPERATIONS
TRAINING Movement of Nuclear Fuel during core offload and core reload is a significant
plant evolution.
The fuel assemblies
and inserts are discharged
from the core into the spent fuel pool (core offload).Control rods, burnable poisons, source rods and thimble plugs are shuffled.Fuel rods are examined for leakers, leakers are reconstituted.
Fresh fuel assemblies, along with once and twice burned fuel are reloaded into the core (core reload).Several important issues need to be considered
during the performance
of fuel handling operations:
- Roles and Responsibilities
- Controlling
Core Reactivity
- Foreign Material Exclusion*Bypassing Fuel Handling Interlocks
- Abnormal Procedures
1.2.Roles and Responsibilities
Operations
Shift Manager Responsible
for the safe operation of the plant.Supervises
all of the licensed and unlicensed
Operators.
Is responsible
for responding
to any abnormal plant response including refueling problems.Objective#2 Fuel Handling SRO An SRO with no other concurrent
responsibilities
and shall direct supervision
of core alterations.
No reactivity
additions or core alterations
can be made without the direct supervision
of the Fuel Handling SRO.The fuel handling SRO should be notified of any indications
of fuel damage, unexpected
reactivity
changes or changes in refueling or spent fuel pool water levels.Core alterations
include: 1)Fuel Movement 2)Control Rod Movement (including
latching and unlatching
control rods)3)Neutron Source manipulation
4)Removal of Reactor Vessel Internals.
Who is the Fuel Handling SRO?The SRO actively in charge on the reactor building operating deck during core alteration
activities.
Although the relief SRO may be on site, all approvals shall be through the SRO actively in charge.OP-MC-FH-FC
FOR TRAINING PURPOSES ONL Y Page 9 of 47 REV.18
DUKE POWER MCGUIRE OPERATIONS
TRAINING The following is a specific list of Fuel Handling SRO responsibilities:
1.Ensure all fuel handling activities
are performed in a safe and efficient manner.2.Securing fuel handling operations
as required by Tech Specs, Plant conditions, Safety concerns, or during times of uncertainty.
3.Should monitor refueling cavity to insure FME is being maintained.
4.Maintain constant communications
with the control room during core alterations.
5.Assist the control room in monitoring
refueling canal level, audible count rate and EMF or containment
evacuation
alarms.6.Assist fuel handling crew in visually verifying fuel assemblies
are lowered and raised safely.Gives hoist operator clearance to engage or disengage on fuel assemblies.
Verifies assemblies
are aligned properly and down on core plate prior to giving concurrence
to disengage gripper.7.Gives verbal clearance prior to pulling control rods during control rod latching, unlatching,anddrag testing activities.
8.During core alterations, approve use of fuel handling bypass interlocks
as necessary when not specified by an approved procedure (NSD 414).Objective#1 Control Room Operators Direct monitoring
and manipulation
of plant and reactor controls.Including monitoring
of subcritical
multiplication
from nuclear instruments
during core alterations.
Responsible
for implementing
any necessary responses required by Abnormal Procedures.
Logging and verifying technical specifications
for MODE 6 and for core alterations.
The Reactor Operator on the headset in the back of the control room communicates
with the refueling crew.The Reactor Operator on the headset will get permission
from the"Operator At The Controls" prior to unloading each fuel assembly.The Operator at the Controls may stop fuel handling operations
if, in his/her judgement, control room indication
or communications
show warranting
conditions.
Nuclear/Reactor
Engineering
One responsibility
is coordination
of fuel movements during core loading operation by use of controlling
procedure.
Another is monitoring
nuclear instrumentation
to verify appropriate
subcritical
behavior and shutdown margin.Reactor Services Technicians
One responsibility
is operation of Fuel Handling Equipment in a safe manner moving fuel to locations recommended
by reactor engineers by procedure.
Another is the ability to recognize and properly respond to abnormal conditions.
OP-MC-FH-FC
FOR TRAINING PURPOSES ONL Y Page 11 of 47 REV.18
DUKE POWER MCGUIRE OPERATIONS
TRAINING 2.2 Bypassing Fuel Handling Interlocks
I Objective#5 I Fuel handlingproceduresdirect
bypassing an interlock when required by known specific operations.
During core alterations, the Licensed SRO for Fuel Handling is tasked with approving the use of bypasses for fuel handling interlocks
as necessary when not specified by an approved procedure (NSD 414).3.0 OPERATION 3.1.Normal Operation Refer to OP/O/A/6550
Series Procedures
Refer to Drawings 7.1 and 7.2 3.1.1 Fuel and Component Handling Fuel and Component Handling is covered by the above procedures
and includes:*Transfer of New Fuel from the Storage Vault to the Spent Fuel Pool*Transfer of New Fuel from the Spent Fuel Pool to the Storage Vault*Spent Fuel Pool Manipulator
Crane Operation*Reactor Building Manipulator
Crane Operation*Fuel Transfer System Operations
- Fuel Handling Tool Operations
3.1.2 Sequence of Refueling Operations
The first major step in refueling operations
concerns preparation.
The Reactor is shutdown and brought to COLD SHUTDOWN.RCS inventory is lowered to the vessel flange.The Fuel Handling Equipment is checked out.Next is Reactor Disassembly.
All connections
are removed from the head.The Refueling Cavity is prepared for flooding (checkout underwater
lights, tools and Fuel Transfer equipment;
close the refueling canal drain valves, and remove blind flange from the transfer tube).Then the vessel head bolts are removed.The head is raised as the canal is flooded by FWST Pumps.The head is taken to it's storage location.Next the control rod drive shafts are disconnected
and with the upper internals are removed from the vessel and stored.Now the core is free from obstructions
and the core is ready for refueling.
OP-MC-FH-FC
FOR TRAINING PURPOSES ONL Y Page 15 of 47
DUKE POWER MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours)OBJECTIVES
LOR 1.5 N N LLL OBJECTIVELL P P 0 0 0 RSR R 0 0 1 Describe the roles and responsibilities
of Control RoomXX X Operators during Fuel Handling operations.
2 Describe the roles and responsibilities
of Fuel HandlingXX SRO's during Fuel Handling operations.
3 Describe how monitoring
of core reactivity
is accomplishedXX X during Fuel Handling.4 Deleted 5 Describe the requirements
that must be met before XXX bypassing a Fuel Handling Interlock.
6 Concerning
AP-25, Spent Fuel Damage;AP-40, Loss ofXX X Refueling Canal;and AP-41 , Loss of Spent Fuel Cooling or Level:*State the purpose of the AP*Given symptoms, state the AP and Case (if applicable)
7 Concerning
the Technical Specifications
related to the FC System;*Given the LCD title, state the LCD (including
any COLRXXX values)and applicability.
- For any LCD's that have action required within one hour, state the action.XXX*Given a set of parameter values or system conditions, determine if any Tech.Spec.is (are)not met and anyXX X action(s)required within one hour.*Given a set of plant parameters
values or system conditions
and the appropriate
Tech Specs, determine X*required action(s).
- Discuss the basis for a given Tech.Spec.LCD or Safety Limit.*SRO only OP-MC-FH-FC
FOR TRAINING PURPOSES ONL Y Page 5 of 47 REV.18
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#3----Group#2 KJA#2.2.7----------
Importance
Rating 3.6----Knowledge of the process for conducting
special or infrequent
tests.Proposed Question: SRO 95 Given the following:
- Unit 1 is in Mode 1 on night shift.*The Work Window Manager and Site Risk Expert are unavailable.*A temporary test (TT)procedure is being performed on RN.*During performance
of the TT, an equipment failure occurred, resulting in a condition not evaluated during planning of the test.In accordance
with SOMP 2-2, Operations
Roles in the Risk Management
Process, who is responsible
for determining
the risk level, and what action is required if the risk level becomes ORANGE?A.WCC SRO;OSM must evaluate the restoration
plan and provide final authority on whether the plan is implemented.
B.WCC SRO;On-Shift CRS must evaluate the restoration
plan and provide final authority on whether the plan is implemented.
C.On-Shift CRS;OSM must evaluate the restoration
plan and provides final authority on whether the plan is implemented.
D.On-Shift CRS;On-Shift CRS must evaluate the restoration
plan and provide final authority on whether the plan is implemented.
Proposed Answer: A Explanation (Optional):
SOMP 02-02 summarizes
the responsibilities
of individuals
in the Operations (OPS)organization
in the processes used to assess and manage risk significant
activities
at Duke nuclear sites.Page 244 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form
A.Correct.During non-core business hours when the WWM or Site Risk Expert are not available, it is the responsibility
of the WCC SRO to evaluate the current risk when existing conditions
do not match those evaluated based on a planned schedule due to emergent work.(SOMP 02-02 Section 5.6.2)WHEN entering an orange or red condition from emergent work, the OSM will evaluate the restoration
plan and have final authority on whether the plan is implemented.(SOMP 02-02 Section 5.5.3)B.Incorrect.
While it is the responsibility
of the WCC SRO to determine the risk level as mentioned above it is the responsibility
of the OSM to evaluate the restoration
plan and who has the final authority to implement.
C.Incorrect.
On-shift CRS is not responsible
for determining
risk.On-Shift CRS responsibility
in the risk management
process is to maintain awareness of current electronic
Risk Assessment
color conditions
for each Unit.He is to immediately
notify the WCC SRO or any emergent equipment problems but is not responsible
for determining
the change is risk status for the affected Unit.(SOMP 02-02 Section 5.7)D.Incorrect.
On-shift CRS isnotresponsible
for either task SOMP 02-02 P 7 Technical Reference(s)(Attach if not previously
provided)----------
OP-MC-ADM-MRA, p49, 51 (Rev 9)Proposed references
to be provided to applicants
during None examination:
Learning Objective:
ADM-MRA, Obj#7 (Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge X Comprehension
or Analysis 10 CFR Part 55 55.41 Content: Page 245 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet 55.43 3 Form ES-401-5 Comments: KA and SRO level is matched because the item evaluates SRO responsibilities
during equipment or procedure step failure during a temporary test procedure Page 246 of 260 Draft 7
SOMP02-02 Page 7 of 19 5.5 Operations
Shift Manager (OSM): 5.5.1 The OSM maintains an awareness of current Electronic
Risk Assessment
color conditions
for each unit.5.5.2 The OSM provides guidance and direction during resolution
of any scheduling
conflicts identified
by the Electronic
Risk Assessment
tool.5.5.3 WHEN entering an orange or red condition from emergent work, the OSM will evaluate the restoration
plan and have final authority on whether the plan is implemented.
Additionally, at their discretion, the OSM may require development
of a written risk management
plan for actions to be taken in the event of further degradations.
5.5.4 The OSM is responsible
for communicating
the risk assessment
results to OPS Shift personnel at the beginning of each shift.5.6 Work Control Center SRO (WCe SRO): 5.6.1 Prior to releasing on-line work, an SRO assigned to the WCC will verify the work is part of the committed schedule, has the correct PRA code for the current plant configuration, and is being performed at the scheduled time.5.6.2 During non-core business hours when the Work Window Manager (WWM)or Risk Site Expert is unavailable, it is the responsibility
of the WCC SRO to evaluate the current risk when the existing conditions
do NOT match those evaluated based on the schedule due to emergent work or schedule carry-overs.
5.7 Control Room Supervisor (CRS): 5.7.1 The CRS maintains an awareness of current Electronic
Risk Assessment
color conditions
for each unit.5.7.2 Immediately
notifies the WCC SRO of any emergent equipment problems.5.7.3 Remains cognizant of all unavailable
equipment and any required contingency
plans.6.Reporting Requirements
None
DUKE POWER MCGUIRE OPERATIONS
TRAINING*Operations
Superintendent:
o Has the final responsibility
to ensure risk assessment
has been performed in accordance
with WPM 609 and 608.*Operations
Work Process Manager (OWPM): o Has overall responsibility
for providing operation focus into the site work scheduling
plan.*OWPM Group: o Will perform a detailed schedule review utilizing the Electronic
Risk Assessment
Tool Results to ensure compliance
with Tech Spec's, SLC, and Probabilistic
Risk Assessment (PRA)concerns.The group will provide guidance and assistance
in creating the schedule for the execution week.Also the group is responsible
for assisting the Work Control Center (WeC)Supervisor
with the final risk assessment, and assisting in conflict resolution.
Objective#7*Operation Shift Manager (OSM): o Maintains the role of command and control of the plant.Maintains an awareness of current Electronic
Risk color conditions
or overall shutdown risk level.Provide guidance and direction during resolution
of conflicts.
WHEN entering an Orange or Red condition from emergent work, the OSM will evaluate the restoration
plan and have the final authority whether the plan is implemented.
Additionally, OSM may require development
of a written risk management
plan for actions to be taken in the event of further degradations.
- wee Supervisor:
o Will utilize the Electronic
Risk results as an aid to ensure minimal risk consequences
occur from scheduled work.o Prior to releasing work, the wee Supervisor
will verify the work is part of the committed schedule, and is being performed at the scheduled time.o For emergent work, the supervisor
will review work order activities
and assign an appropriate
PRA code so that activities
are appropriately
included in the Electronic
Risk analysis.Assist the Work Window Manager (WWM)in evaluating
emergent work against the current schedule utilizing Electronic
Risk.o IF the R&R requirement
is changed from planned work or if the R&R itself is revised, evaluate any potential change in riskm o WHEN performing
any procedure, ensure the planned configuration
is evaluated for risk.IF the component is rendered IIUnavaiiable", then perform a IIWhat-lf ll scenario in the risk assessment
tool.OP-MC-ADM-MRA
FOR TRAINING PURPOSES ONL Y REV.09 Page 49 of 87
DUKE POWER MCGUIRE OPERATIONS
TRAINING o During non-core business hours when the WWM or Risk Site Expert is unavailable, it is the responsibility
of the WCC Supervisor
to evaluate the current risk when the existing SSC*s do NOT match those evaluated based on the schedule.o WHEN there is any doubt concerning
the applicability
of any PRA Code, the conservative
choice is to apply the code for that SSC.Objective#7*CONTROL ROOM SRO: o The Control Room SRO is responsible
to maintain an awareness of current Electronic
Risk Assessment
Tool color conditions
on his/her Unit.This includes an awareness of the work causing the increased level of risk along with contingency
plans for system restoration.
3.1.2 Items To Consider:*All work order tasks and/or maintenance
activities
must be risk assessed against actual plant configuration
as required by 10CFR50.65.
- Variances from the established
schedule require re-evaluation.
Changing the plant configuration
or the work sequence may invalidate
the risk assessment.
- High Safety Significant
SSCs are NOT always the same as TS/SLCs.For example, Instrument
Air or Spent Fuel Cooling is NOT a TS/SLC item, however is in Electronic
Risk Assessment
Tool.*Inoperable
items should be considered"Unavailable" unless an evaluation
to determine its'availability
has been performed (e.g.,*NO_CODE on work order task).*All risk evaluations
can NOT be predetermined.Personsperforming
the risk evaluation
should use additional
sources as necessary to perform the evaluation (Le., plant drawings, procedures, previous evaluations, knowledge,&training).
- Electronic
Risk determination
is evaluated by two distinct methods;o Deterministic
and Probabilistic.Probabilistic
Risk Assessment (PRA)is based on the adverse affect on Core Damage Frequency when a SSC is determined
to be"Unavailable".Deterministic
Risk Assessment
is based on the determined
risk (expert judgment)when a SSC is determined
to be IIUnavailable".
- As with all aspects of nuclear power operation, if in doubt, be conservative.
- WHEN coding, realize that a specific code may NOT be available for the SSC(s)which are unavailable.
In this case, if warranted, use of an alternate code that removes the same function may be necessary.
OP-MC-ADM-MRA
FOR TRAINING PURPOSES ONL Y Page 51 of 87 REV.09
DUKE POWER MCGUIRE OPERATIONS
TRAINING SNNLL L E OBJECTIVELLPP 000 R S R Q R 0 0 7 Explain the roles and responsibilities
associated
with the X XXX Electronic
Risk Assessment
Tool work release process, for the following Operations
personnel:
- OSM (Operations
Shift Manager)*Control Room SRO*wec (Work Control Center)SRO ADMMRAOO7 8 Deleted X X X X ADMMRAOO8 9 Deleted X X X X ADMMRAOO10
OP-MC-ADM-MRA
FOR TRAINING PURPOSES ONL Y Page 7 of 87 REV.09
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#3----Group#2 KIA#2.2.22----------
Importance
Rating 4.7----Knowledge of lirlliting
conditions
for operations
and safety limits.Proposed Question: SRO 96 Given the following:*Unit1 is in Mode 3.*Shutdown Banks are withdrawn.
- NC system pressure has increased to 2772 psig.In accordance
with Tech Spec Bases, which ONE of the following CORRECTLY describes all the components
that are assumed to operate at their setpoints to ensure that NC pressure remainsbelowthe Technical Specification
Safety Limit, and THE MAXIMUM TIME allowed to reduce NC pressure to below the Safety Limit?A.Pressurizer
Code Safeties, Main Steam Code Safeties, High Pressure Rx.Trip;1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.B.Pressurizer
Code Safeties, Main Steam Code Safeties, High Pressure Rx.Trip;5 minutes.C.Pressurizer
PORVs, Main Steam PORVs, High Pressurizer
Level Rx.Trip;1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.D.Pressurizer
PORVs, Main Steam PORVs, High Pressurizer
Level Rx.Trip;5 minutes.Proposed Answer: B Explanation (Optional):
A.Incorrect.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowed for Modes 1 and 2, but Mode 3 and below require pressure to be reduced below the SL within 5 minutes 8.Correct.c.Incorrect.
High PZR level trip is a backup and is not considered
for safetylimitprotection.
Plausible because it is a valid reactor trip.PORVs are not Page 247 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 credited in the accident analysis for High RCS pressure.Plausiblebecause
they will operate and do perform a safety related function D.Incorrect.
High PZR level trip is a backup and is not considered
for safety limit protection, but time is correct.PORVs are not credited in the accident analysis for High RCS pressure Technical Reference(s)TS 2.1.1 and basis;TS 3.3.1 and basis PS-NC Rev 30 IC-IPE Rev 28 Proposed references
to be provided to applicants
during None examination:
Learning Objective:
IC-IPE Obj 10, 14;PC-NC (As available)
Obj 17 (Note changes or attach parent)----Bank#X----Modified Bank#New Question Source: Question History: Last NRC Exam
_Question Cognitive Level: Memory or Fundamental
Knowledge X Comprehension
or Analysis10 CFR Part 55 55.41 Content: 55.43 1,2 Comments: NRC developed test item for Vogtle exam KA is matched because item requires knowledge of LCOs, NSSS setpoints and basis for setpoints, and action in a lower mode specific to protection
of a safety limit Page 248 of 260 Draft 7
SLs 2.0 2.0 SAFETY LIMITS (SLs)2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination
of THERMAL POWER, Reactor Coolant System (RCS)highest loop average temperature, and pressurizer
pressure shall not exceed the limits specified in the COLR for four loop operation;
and the following SLs shall not be exceeded: 2.1.1.1 The departure from nucleate boiling ratio (DNSR)shall be maintained
2:,1.14 for the WRS-2M CHF correlation.
2.1.1.2 The peak fuel centerline
temperature
shall be maintained
<5080 degrees F, decreasing
58 degrees F for every 10,000 MWd/mtU of fuel burnup.2.1.2 RCS Pressure SL In MODES 1,2,3,4, and 5, the RCS pressure shall be maintained2735 psig.2.2 SL Violations
2.2.1 If SL 2.1.1 is violated, restore compliance
and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.2.2.2 If SL 2.1.2 is violated: 2.2.2.1 2.2.2.2 In MODE 1 or 2, restore compliance
and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.In MODE 3, 4, or 5, restore compliance
within 5 minutes.McGuire Units 1 and 2 2.0-1 Amendment Nos.219 I 201
Reactor Core SLs B 2.1.1 BASES B 2.0 SAFETY LIMITS (SLs)B 2.1.1 Reactor Core SLs BASES BACKGROUND
GDC 10 (Ref.1)requires that specified acceptable
fuel design limits are not exceeded during steady state operation, normal operational
transients, and anticipated
operational
occurrences (AOOs).This is accomplished
by having a departure from nucleate boiling (DNB)design basis, which corresponds
to a 95%probability
at a 95%confidence
level (the 95/95 DNB criterion)
that DNB will not occur and by requiring that fuel centerline
temperature
stays below the melting temperature.
The restrictions
of this SL prevent overheating
of the fuel and cladding, as well as possiblecladdingperforation, that would result in the release of fission products to the reactor coolant.Overheating
of the fuel is prevented by maintaining
the transient peak linear heat rate (LHR)below the level at which fuel centerline
melting occurs.Overheating
of the fuel cladding is prevented by restricting
fuel operation to within the nucleate boiling regime, where the heat transfer coefficient
is large and the cladding surface temperature
is slightly above the coolant saturation
temperature.
Fuel centerline
melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline
temperature
to reach the melting point of the fuel.Expansion of the pellet upon centerline
melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled
release of activity to the reactor coolant.Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature
because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient.
Inside the steam film, high cladding temperatures
are reached, and a cladding water (zirconium
water)reaction may take place.This chemical reaction results in oxidation of the fuel cladding to a structurally
weaker form.This weaker form may lose its integrity, resulting in an uncontrolled
release of activity to the reactor coolant.The proper functioning
of the Reactor Protection
System (RPS)and steam generator safety valves prevents violation of the reactor core SLs.McGuire Units 1 and 2 B 2.1.1-1 Revision No.51
Reactor Core SLs B 2.1.1 BASES APPLICABLE
The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and AOOs.Thereactor
core SLs are established
to preclude violationofthe following fuel design criteria: a.There must be at least 95%probability
at a 95%confidence
level (the 95/95 DNB criterion)
that the hot fuel rod in the core does not experience
DNB;and b.The hot fuel pellet in the core must not experience
centerline
fuel melting.The Reactor Trip System setpoints (Ref.2), in combination
with all the LCOs, are designed to prevent any anticipated
combination
of transient conditions
for Reactor Coolant System (RCS)temperature, RCS Flow Rate, ill, pressure, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR)of less than the DNBR limit and preclude the existence of flow instabilities.
Automatic enforcement
of these reactor core SLs is provided by the appropriate
operation of the RPS and the steam generator safety valves.The SLs represent a design requirement
for establishing
the RPS trip setpoints identified
previously.
LCO 3.4.1 , IIRCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)Limits,1I or the assumed initial conditions
of the safety analyses (as indicated in the UFSAR, Ref.2)provide more restrictive
limits to ensure that the SLs are not exceeded.SAFETY LIMITS The Figure provided in the COLR shows the loci of points of Fraction of Rated Thermal power, RCS Pressure, and average temperature
for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline
temperature
remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, and that the exit quality is within the limits defined by the DNBR correlation.
The reactor core SLs are established
to preclude violationofthe following fuel design criteria: a.There must be at least 95%probability
at a 95%confidence
level (the 95/95 DNB criteria)that the hot fuel rod in the core does not experience
DNB;and b.There must be at least a 95%probability
at a 950/0 confidence
level that the hot fuel pellet in the core does not experience
centerline
fuel melting.The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal McGuire Units 1 and 2 B 2.1.1-2 Revision No.51
DUKE POWER Objective#15 MCGUIRE OPERATIONS
TRAINING The common discharge line from the PORVs has a temperature
element which provides indication
for PORV discharge temperature
via meter located on 1 (2)MC1 0 and an alarm on 1 (2)AD6"Pzr PORV Disch Hi Temp" (setpoint 140 0 F).This indication
is used to assist in identifying
if a PORV is leaking which has Tech Spec implications.
I Objective#16 I Each PORV has a loop seal between the PORV and its electric isolation.
These loop seals were designed to assist in preventing
the leakage of H 2 through the PORV valve seat.Industry concerns were raised over potential water slug acceleration
and subsequent
piping damage when a PORV or safety was opened.).It was determined, as documented
in PIP1-M94-1470that
in this application
a water slug would not damage the piping to the extent that the PORVs would become inoperable.
However, each loop seal between the PORV block valve and PORV has a drain line which normally drains the condensate
back to the pressurizer (Refer to Drawing 7.10).These drain valves do not have to be open for the PORVs to be operable.Each drain line has normally open isolation valve (NC269, 270, 271).Each valve is solenoid actuated and can be operated from the control room on 1 (2)MC1 O.The drain lines join to a common line which can be isolated by manual valve NC 61.Sample valve NM6A,C, and NM78 provide the flow path for this line.If a PORV is leaking, its associated
block valve and loop seal drain isolation valve will be closed to prevent bypass of the block valve function.2.8 Pressurizer
Code Safety Valves I Objective#15,17,18 I The purpose of the safety valves (NC1 ,2 and 3)is to prevent the NCS from being pressurized
above its safety limit of 2735 psig.Each unit has three totally enclosedtype, spring loaded, self-actuated
safety valves set at 2485 psig.The combined capacity of the three valves is greater than or equal to the maximum surge capacity following a complete loss of load without a reactor trip.The 6 inch pipes connecting
the pressurizer
nozzles to their respective
code safety valves are shaped in the form of a loop seal.Originally, the loops seals were designed to collect condensate, as a result of normal heat losses to the containment
atmosphere.
The condensate
was to prevent any leakage of hydrogen gas or steam through the safety valve seats.However, a concern was raised that if a water slug were to be accelerated
when the safety valve opened, the resultant water hammer could result in severe damage to the valve and/or downstream
piping which could result in an unisolable
leak from the steam space of the pressurizer.
Therefore the safety valve internals were replaced with a design that could seal on steam and drains for the loops seals were added to continuously
drain condensate
back to the pressurizer
via one of the upper pressurizer
level detector penetrations.
Each of these drain lines has a strap on RTD which provides temperature
indication
on the OAG.LO (approx.110 degrees)andLOLO (approx.100 degrees)OAC alarms are provided to notify Engineering
to assess operability
of the Safety Valves at low temperatures.
OP-MC-PS-NC
FOR TRAINING PURPOSES ONL Y Page 37 of 135 REV.30
DUKE ENERGY Objective#10 MCGUIRE OPERATIONS
TRAINING Power Range NIS Low Setpoint (214 channels=25°k)-Protects against startup accidents.
The trip can be manually blocked when 214 PR channels>10%(P-10)by using the two control board switches, one per train.The control board provides indication
of the bistable block.This trip is auto-reinstated
when 3/4 PR channels<10%(P-10).Power Range NIS High setpoint (214 channels=109°k)-protects against an overpower condition which could lead to a DNB concern.This circuit also provides a rod withdrawal
stop when 1/4 channels>103%power (C-2).Power Range Positive (+)Rate (214 channels+5°k in 2 sec)-protects against an ejected rod accident for DNB concerns.Pressurizer
High Pressure (214 channels=2385 psig)-Protects against losing NC system integrity.
Pressurizer
Low Pressure (214 channels=1945 psig)-protects against DNB due depressurization.
This"at-power" trip protection
is auto-blocked
<10%power (P-7)and is automatically
reinstated>
P-7.Pressurizer
High Level (213 channels=920/0)-protects system integrity by preventing
the passage of water through the safeties.This"at-power" trip protection
is auto-blocked
<10 0 k power (P-7)and is automatically
reinstated>
p.7.OTAT (214 channels=variable)-provides DNB protection.
DNB causes a large decrease in the heat transfer coefficient
between the fuel surface and the coolant, resulting in high fuel clad temperature.
The setpoint is a function of the 120 0 k full power AT, Tavg, Pressurizer
Pressure, and A Flux.Pressures below 2235 psig cause the setpoint to decrease while pressures above 2235 psig cause an increase in the setpoint.Tavg above 585 of causes the setpoint to decrease while Tavg below 585 of causes an increase in the setpoint.A A Flux more positive than the limit in the COLR (positive breakpoint)
causes the setpoint to decrease.This circuit also provides a rod withdrawal
stop and Turbine Runback 2°k (C-3)below the trip setpoint.OPAT (214 channels=variable)-protects against excessive fuel centerline
temperature
due to high fuel rod power density (kW/ft).The setpoint is a function of the 109°k full power AT, Tavg, Rate of Tavg increase, and A Flux.Tavg above 585 of cause the setpoint to decrease with no credit for Tavg below 585 of.A A Flux more positive than the limit in the COLR (positive breakpoint)
or more negative than the limit in the COLR (negative breakpoint)
causes the setpoint to decrease.This circuit also provides a rod withdrawal
stop and Turbine Runback 2°k (C-4)below the trip setpoint.NC Pump Bus Low Voltage (214 busses=74°k)-this anticipatory
loss of coolant flow trip protects against DNB.This"at-power" trip protection
is auto-blocked
<10 0 k power (P-7)and is automatically
reinstated>
P-7.OP-MC-IC-IPE
FOR TRAINING PURPOSES ONL Y Page 45 of 149 REVe28
DUKE ENERGY 7.5 Reactor Trips (3/27/01)MCGUIRE OPERATIONS
TRAINING MANUAL Sw.turned 45°1/2 sw.operator judgment S.R.NI HIGH 10 CPS 1/2 ch.P6, P10 uncontrolled
rod withdrawal/
startu accidents I.R.NI HIGH amps-25%power 1/2 ch.P10 uncontrolled
rod withdrawal/
startu accidents P.R.NI LOW 25%power 2/4 ch.P10 reactivity
excursion from low owers P.R.NI HIGH
power 2/4 ch.reactivityexcursionfrom
all owers DNB P.R.POS+5%/2 sec 2/4 ch.DNB (rod ejection)RATE PZR HIGH 2385 psig 2/4 ch.coolant system integrity PRESS PZR LOW 1945 psig 2/4 ch.P7 DNB PRESS PZR HIGH2/3 ch.P7 water through safeties (system LEVEL inte rit>
2/4/ch.DNB
2/4 ch.KW/FT NCP BUS 74%of normal 2/4 ch.P7 DNB (anticipatory
loss of flow)LOW VOLT NCP BUS 56 Hz 2/4 ch.P7 DNB (anticipatory
loss of flow)LOW FREQ S/G La-La 17%2/4 in loss of heat sink LVL 1/4 s/1 LOOP2/3 in P8 DNB LOSS OF 1/4 loops FLOW 2 LOOP 88%2/3 in P7 DNB LOSS OF 2/4 loops FLOW SAFETY any S/I signal 1/2 S/I trip reactor if trip not INJECTION actuated trains generated by trip instrumentation
GENERAL loose card, loss of 2/2 alarms loss of protection
WARNING voltage, train in ALARM test, by-pass bkr connected/closed, logic ground return fuse blown TURBINE low Auto-stop oil 2/3 ASO P8 trip reactor on turbine trip TRIP press<45 psig or Press all 4 stop valves switches closed 4/4 valves OP-MC-IC-IPE
FOR TRAINING PURPOSES ONL Y Page 83 of 149 REVe28
RTS Instrumentation
B 3.3.1 BASES APPLICABLE
SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
a.Pressurizer
Pressure-Low
The Pressurizer
Pressure-Low
trip Function ensures that protection
is provided against violating the DNBR limit due to low pressure.The LCO requires four channels of PressurizerLow to be OPERABLE.In MODE 1, when DNB is a major concern, the Pressurizer
Pressure-Low
trip must be OPERABLE.This trip Function is automatically
enabled on increasing
power by the P-7 interlock (NIS power range P-10 or turbine impulse pressure greater than approximately
10%of full power equivalent13)).On decreasing
power, this trip Function is automatically
blocked below P-7.Below the P-7 setpoint, power distributions
that would cause DNB concerns are unlikely.b.Pressurizer
Pressure-High
The Pressurizer
Pressure-High
trip Function ensures that protection
is provided against overpressurizing
the RCS.This trip Function operates in conjunction
with the pressurizer
relief and safety valves to prevent RCS overpressure
conditions.
The LCO requires four channels of the Pressurizer
Pressure-High
to be OPERABLE.The Pressurizer
Pressure-High
LSSS is selected to be below the pressurizer
safety valve actuation pressure and above thepoweroperated
relief valve (PORV)setting.This setting minimizes challenges
to safety valves while avoiding unnecessaryreactortrips
for those pressure increases that can be controlled
by the PORVs.In MODE 1 or 2, the Pressurizer
Pressure-High
trip must be OPERABLE to help prevent RCS overpressurization
and minimize challenges
to the safety valves.In MODE 3, 4, 5, or 6, the Pressurizer
Pressure-High
trip Function does not have to be OPERABLE because transients
that could cause an overpressure
condition will be slow to occur.Therefore, the operator will have sufficient
time to evaluate unit McGuire Units 1 and 2 B 3.3.1-16 Revision No.90
RTS Instrumentation
B 3.3.1 BASES APPLICABLE
SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
conditions
and take corrective
actions.Additionally, low temperature
overpressure
protection
systems provide overpressure
protection
when below MODE 4.9.Pressurizer
Water Level-High
The Pressurizer
Water Level-High
trip Function provides a backup signal for the Pressurizer
Pressure-High
trip and also provides protection
against water relief through the pressurizer
safety valves.These valves are designed to pass steam in order to achieve their design energy removal rate.A reactor trip is actuated prior to the pressurizer
becoming water solid.The setpoints are based on percent of instrument
span.The LCO requires three channels of Pressurizer
Water Level-High
to be OPERABLE.The pressurizer
level channels are used as input to the Pressurizer
Level Control System.A fourth channel is not required to address control/protection
interaction
concerns.The level channels do not actuate the safety valves, and the high pressure reactor trip is set below the safety valve setting.Therefore, with the slow rate of charging available, pressure overshoot due to level channel failure cannot cause the safety valve to lift before reactor high pressure trip.In MODE 1, when there is a potential for overfilling
the pressurizer, the Pressurizer
Water Level-High
trip must be OPERABLE.This trip Function is automatically
enabled on increasing
power by the7 interlock.
On decreasing
power, this trip Function is automatically
blocked below P-7.Below the P-7 setpoint, transients
that could raise the pressurizer
water level will be slow and the operator will have sufficient
time to evaluate unit conditions
and take corrective
actions.1 O.Reactor Coolant Flow-Low a.Reactor Coolant Flow-Low (Single Loop)The Reactor Coolant Flow-Low (Single Loop)trip Function ensures that protection
is provided against violating the DNBR limit due to low flow in one or more RCS loops, while avoiding reactor trips due to normal variations
in loop flowa Above the P-8 setpoint, which is approximately
48%a loss of flow in any RCS loop will actuate a reactor trip.The setpoints are based on the minimum flow specified in the McGuire Units 1 and 2 B 3.3.1-17 Revision No.90
DUKE POWER MCGUIRE OPERATIONS
TRAINING OBJECTIVESNNLl L OBJECTIVE L L P P 000 R S R R 0 0 12 Concerning
the pzr cold and hot calibrated
level indication:
- state the purpose of this indicationXXX*describe how the operator corrects the indicated levelXXXX for temperature
- state the problems which can occur if the level is notXX X X corrected for temperature
13 State the purpose of the pressurizer
power operated reliefXXX X valves.14 List the parameters
and setpoints associated
with the NCSXXX X relief valves.15 Describe the indications
which would be used to identify aXX X X leaking pzr PORV or safety.16 Concerning
the pzr PORV loop seals:*what was their original purposeXXX*why are they continuously
drained during operationXX X*describe the operational
concern of leaving the drain XXX X valve open while its associated
PORV is leaking*state from where the loop seal drain valves areXX X X operated.17 State the purpose of the pzr Code safety valve.XXX X 18 Concerning
the pzr Code safety valves loop seals:*what was their original purpose XXX*why are they continuously
drained during operationXX X X 19 State the purpose of the pressurizer
relief tank and the design XXX X features which accomplish
the purpose.OP-MC-PS-NC
FOR TRAINING PURPOSES ONL Y Page 9 of 135 REV.30
DUKE ENERGY MCGUIRE OPERATIONS
TRAINING SNN LLL E OBJECTIVELLPP 000 R S R Q R 0 0 8 Describe the function of the First-Out annunciator
panel.X X ICIPEOO8 9 Given a Limit and/or Precaution
associated
with an operatingXX X X procedure, discuss its basis and applicability.
ICIPEOO9 10 List all the Reactor Trip Signals including the setpoints, logicXXX X permissives
and bases/protection
afforded by each.ICIPE010 11 List all the protective
system permissive (UP" signal)interlocks
X X X to include input parameter(s), logic and function.For interlocks
which provide Trip block, state the Trips affected and whether Auto or Manual block.ICIPE011 12 List all the protection
system control ("e" signal)interlocksXX X including logic and functions.
ICIPE012 13 Briefly describe the incident that occurred at Salem Nuclear XXX Plant and how this event affected McGuire Reactor Trip Breaker operation.
ICIPE013 OP-MC-IC-IPE
FOR TRAINING PURPOSES ONL Y Page 11 of 149 REV.28
DUKE ENERGY MCGUIRE OPERATIONS
TRAINING 5NNLLL E OBJECTIVELL P P 0 0 0 R 5 R Q R 0 0 14 Concerning
the Technical Specifications
related to the Reactor Protection
System;*Given the LCO title, state the LCO (including
any COLR X X X values)and applicability.
- For any LCO's that have action required within one hour,XX X state the action.*Given a set of parameter values or system conditions, XXX determine if any Tech Spec LCO's is (are)not met and any action(s)required within one hour.*Given a set of plant parameters
or system conditions
and X X X the appropriate
Tech Specs, determine required action(s).
- Discuss the basis for a given Tech Spec LeO or Safety X*Limit.*SRO Only ICIPE014 OP-MC-IC-IPE
FOR TRAINING PURPOSES ONL Y Page 13 of 149 REV.28
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outline Cross-Level RO SRO reference:
Tier#3 Group#3 KIA#2.3.14 Importance
Rating 3.8 Knovvledge
of radiation or contarnination
hazards that rnay arise during normal)abnormal l or emergency conditions
or activities.
Proposed Question: Giventhefollowing:
SR097*A load reduction from 1 000/0 to 60%was performed on Unit 1 in the last 30 minutes due to a Feedwater Control problem.*The following alarms are received:*1 EMF-48 REACTOR COOLANT HIGH RAD*1EMF-18, REACTOR COOLANT FILTER 1A*Chemistry sample indicates that the high activity is due to failed fuel.*Dose-Equivalent
lodine-131
is approximately
5 microcuries
per gram.*The crew enters AP/18, High Activity in Reactor Coolant.Which ONE of the following actions will be performed in accordance
with AP/18, and which ONE of the following describes the technical specification
implications
of this condition?
A.Raise Letdown flow to 120 GPM;plant shutdown and cooldown to less than 500°F must be performed.
B.Raise Letdown flow to 120 GPM;plant operation may continue with increased NC SYSTEM sampling frequency.
C.Ensure Mixed Bed Demin is in service and evaluate use of Cation Bed Demin;plant shutdown and cooldown to less than 500°F must be performed.
D.Ensure Mixed Bed Demin is in service and evaluate use of Cation Bed Demin;plant operation may continue with increased NC SYSTEM Page 249 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 sampling frequency.
Proposed Answer: D Explanation (Optional):
A.Incorrect.
Letdown flow is raised only for crud burst.Failed Fuel is indicatedbyiodine activity.T8 shutdown required only after being above 3.4.16-1 acceptable
operation.
B.Incorrect.
Letdown flow is raised only for crud burst.Failed Fuel is indicatedbyiodine activity, as decribed by conditions
presented.
c.Incorrect.
T8 shutdown required only after being above 3.4.16-1 acceptable
operation.
This condition is above TS steady state limit but below the transient limit on the curve D.Correct.Technical Reference(s)AP/18 Rev 2 and Basis Document T83.4.16 Proposed references
to be provided to applicants
during exam i nation: TS Figure 3.4.16-1 Learning Objective: (As available)
(Note changes or attach parent)----Bank#Modified Bank#New X Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge Comprehension
or Analysis X 1 0 CFR Part 55 55.41 Content: 55.43 2,4 Comments: Page 250 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 KA is matched because the item evaluates understanding
of a fuel failure vs a crud burst.SRO level because the SRO must determine appropriate
action based upon evaluation
of this condition.
The action taken is required by technical specifications
Page 251 of 260 Draft 7
RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.16 RCS Specific Activity LCO 3.4.16 APPLICABILITY:
MODES 1 and 2, MODE 3 with RCS average temperature (T avg)500°F.ACTIONS CONDITIONREQUIREDACTION
COMPLETION
TIME A.1.1--------------------Note-------------------
LCO 3.0.4.c is applicable.
A.1 Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AND A.2 Restore DOSE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> EQUIVALENT
1-131 to within limit.B.Gross specific activity of B.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> the reactor coolant not T avg<500°F.within limit.(continued)
McGuire Units 1 and 2 3.4.16-1 Amendment Nos.221/203
RCS Specific Activity 3.4.16 ACTIONS (continued)
C.CONDITION Required Action and associated
Completion
Time of Condition A not met.OR DOSE EQUIVALENT
1-131 in the unacceptable
region of Figure 3.4.16-1.REQUIRED ACTION C.1 Be in MODE 3 with T avg<500°F.COMPLETION
TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> SURVEILLANCE
REQUIREMENTS
SURVEILLANCE
-SR 3.4.16.1 Verify reactor coolant gross specific activity.:s.1 DOlE IJCi/gm.SR 3.4.16.2------------------------------NOTE------------------------------------
Only required to be performed in MODE1.FREQUENCY 7 days Verify reactor coolant DOSE EQUIVALENT
1-131 specific 14 days activity.:s.1.0 IJCi/gm.Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of?150/0 RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period (continued)
McGuire Units 1 and 2 3.4.16-2 Amendment Nos.184/166
SURVEILLANCE
SR 3.4.16.3------------------------------N()lrE------------------------------------
Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical
for?48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.-Determine E from a sample taken in M()DE 1 after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical
for?48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.RCS Specific Activity 3.4.16 FREQUENCY 184 days McGuire Units 1 and 2 3.4.16-3 Amendment Nos.184/166
300 250 I-:J>l-s;i=(.)<C (.)200 u:::<3 ww_I-E z co:3 c, oo Co150 oI-(.)(.)0<C U w.-a:.§.,..('t)I-ffi 100..J<C>:5 a w w w o c 50\\.'\\UNACCEPTl OPERATI ON\\ACCEPT MLE OPERA'"'""'ION Res Specific Activity 3.4.16 o 20 30 40 506070 80 PERCENT OF RATED THERMAL POWER 90 100 McGuire Units 1 and 2 3.4.16-4 Amendment Nos.184/166
MNS AP/1/A/5500/18
UNITl HIGH ACTIVITY IN REACTOR COOLANT PAGE NO.2 of 3 Rev.2 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED 8.Symptoms*111 EMF-48 REACTOR COOLANT HI RADII alarm*111EMF-18 REACTOR COOLANT FILTER 1 All alarm*111EMF-19 REACTOR COOLANT FILTER 18 11 alarm*Chemistry sample results indicate an unexpected
increase in NC System activity.C.Operator Actions 1.Check 1 NV-127 A (LID Hx Outlet 3-Way Temp Cntrl)-ALIGNED TO DEMIN.2.Determine cause of high activity: a.Request Chemistry to check decontamination
factor of mixed bed demineralizer.
b.Notify Chemistry to perform an NC System isotopic analysis to determine if high activity is from a crud burst or failed fuel._Align valve to IIDEMIN II position.
MNS AP/1/A/5500/18 UNITl HIGH ACTIVITY IN REACTOR COOLANT PAGE NO.3 of 3 Rev.2 ACTION/EXPECTED
RESPONSE RESPONSE NOT OBTAINED a.b.c.IF AT ANY TIME Chemistry requests cation bed demineralizer
be placed in service, THEN place in service PER OP/1/A/6200/001
D (Chemical and Volume Control System Demineralizers), Enclosure 4.3 (Removing/Returning
the Cation Bed Demineralizer
from/to Service).d.REFER TO RP/O/A/5700/000 (Classification
of Emergency).
e.Notify Reactor Group to perform OP/O/A/6550/017 (Estimate of Failed Fuel Based on lodine-131
Concentration)
.5.Notify Radwaste to ensure VCT H2 purge flow is established.
6.REFER TO Tech Spec 3.4.16 (RCS Specific Activity).
AP/1 and 21A15500/018 (High Activity in Reactor Coolant)PURPOSE: DISCUSSION:
At the normal letdown flow rate of 75 gpm, it takes almost 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> to pass one entire volume of reactor coolant through the NV System.But a letdown flow of 120 gpm will circulate one entire volume of reactor coolant in approximately
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (at 120 gpm letdown flow, 50%of the crud is removed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />).REFERENCES:
Primary Chemistry Lesson Plan OP-MC-CH-PC
STEP 4: PURPOSE:
made.DISCUSSION:
Step 4.a ensures mixed bed demin in service to facilitate
removal of both the ion types produced by failed fuel (halogens and soluble metal ions).Step 4.b notifies Chemistry to determine if the cation bed should be placed in service so they can get with Reactor Group, RP, and themselves
to weigh the pros and cons of placing the cation bed in service.While the cation will remove the soluble metal ions like Cesium, in doing so it will also remove the Lithium ion that is used for PH control.Operating with PH out of spec must be weighed against the urgency of removing the failed fuel ions (dose control, etc.)Step 4.c gets the cation bed in serivce, if requested using the OP.Step 4.d is a reference to RP/0/A/5700/000 (Classification
of Emergency)
to ensure the proper declaration
is made.If a plant shutdown required by T.S.3.4.16 (RCS Specific Activity)is commenced, a Notification
of Unusual Event is declared based on failed fuel.For grosser failures beyond the T.S.limits, other classification
levels may be reached.Step 4.e is a quick gross guess at the extent of the failed fuel.The Reactor Group has more qualitative
tools that they'll implement as warranted, but this is a quick estimate.Basically, this procedure takes the 1-131 concentration
in uCI/ml and divides by a number depending on initial conditions (normal, clad damage, severe fuel overtemperature, or fuel melting), with correction
factors for sampling temperature
and power history: A.Normal: 1-131 uCllml+1.8 uCI/ml=Percent failed fuel B.Clad damage 1-131 uCI/ml+83.7 uCl/ml=Percent failed fuel C.Severe Fuel Overtemperature
1-131 uCl/ml+1535 uCl/ml=Percent failed fuel D.Fuel Melting 1-131 uCI/ml+2790 uCl/ml=Percent failed fuel As seen above, the more the fuel cladding is stressed by the failed fuel mechanism, the more activity is expected for a given percentage
of failed fuel.REFERENCES:
RP/OIN57001000 (Classification
of Emergency), OP/0/A/6550/17 (Estimate of Failed Fuel Based on lodine-131)
Page 4 of 6 Rev 0
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outlinereference:
Level RO SRO Tier#3----Group#3#2.3.6----------
Importance
Rating 3.8----Ability to approve release permits Proposed Question: SRO 98 Unit 1 is shutdown in mode 6 refueling.Radwaste Operator brings a liquid radiological
release permit to the SRO for approval.Given the following information
on the permit:*Release ID=WMT-B*RC Pumps running=4*RC Pumps assigned to release=3*Total RC Pumps required=1*Allowable release rate=1.61 E+05 gpm*Recommended
release rate=6.00E+01 gpm*EMF-49 (L)(LIQUID
DISCH)in service=yes*EMF background
=4.49E+03*Trip 1 setpoint=8.97E+03*Trip 2 setpoint=1.34E+04 If no other releases are in progress, which one of the following actions is correct for approval of this release permit?The release may not be approved because there is an error in the number of RC pumps required B.The release may not be approved because the EMF-49(L)trip setpoints are not correct C.The release may not be approved because the release rate is not correct D.The release may be approved as presented if a source check of49(L)is performed successfully.
ProposedPage 252 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Explanation (Optional):
A.Correct the remarks section states 4 RC pumps are required but the number of RC pumps required is listed as 3 in the RC pump data section B.Incorrect:
-nothing wrong with EMF-49L trip setpoints Plausible:
-background
<trip 1<trip 2 C.Incorrect:
-allowable release rate<recommended
release rate Plausible:
-if candidate does not understand
this requirement
D.Incorrect:
-the RC pumps required is not correct, but otherwise this is correct Technical Reference(s)OP-MC-WE-RLR, Rev 13 (Attach if not previously
provided)-----------
OP/O/B/6200/107
P 6 Proposed references
to be provided to applicants
during None examination:
Learning Objective:
OP-MC-WE-RLR
obj 3 (Note changes or attach parent)----Bank#X----Modified Bank#New Question Source: Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental
Knowledge X Comprehension
or Analysis10 CFR Part 55 55.41 Content: 55.43 4 Comments: KA is matched because the item evaluates requirements
for issuing a radioactive
liquid waste release permit.SRO level because the SRO is responsible
for authorizing
the release based on given conditions
Page 253 of 260 Draft 7
Enclosure 4.3 B WMT Release Using B WMT Pump OP/O/B/6200/107
Page 6 of 26 3.16 CR SRO performs the following steps:{PIP M-03-01124}
{PIP M-04-03470}
SRO 3.16.1 Determine operability
status of the following components
and circle"Yes" or"No" to so indicate: OWMLP5140 (B WMT Pump Disch Flow)[i.e., OWMCR5130 (Waste Mon Tank Pumps Disch Flow)or OWMFT5140 (B Waste Monitor Tank Pump Disch Flow)]1 WP-35 (WMT&VUCDT to RC Cntrl)1WP-37 (Liquid Waste to RC Cntrl)OEMF49 (Liquid Waste Disch Radiation Monitor)OWMFS5440 (OEMF49 Outlet Flow){PIP M-03-02673}(Yes/No)(YeslNo)(Yes/No)(Yes/No)(Yes/No)3.16.2 IF any component listed in Step 3.16.1 is inoperable, notify Radwaste SRO Chemistry and return L WR Document.SRO SRO 3.16.3 3.16.4 Ensure the following items on LWR Document are complete:*Number of"RC Pumps Running" is greater than or equal to"RC Pumps Assigned to this Release".*Number of"RC Pumps Running" is greater than"Total RC Pumps Required (all concurrent
Releases)".*"Recommended
Release Rate (gpm)" is less than"Allowable
Release Rate (gpm)".*OEMF49L is operable and in service.*OEMF49 source check performed.
- "Expected CPM" is less than"Trip 1 Setpoint" and"Trip 2 Setpoint".
WHEN approved for release, place signature, date, and time of SRO authorization
on L WR Document.
DUKE ENERGY MCGUIRE OPERATIONS
TRAINING*If a site assembly occurs during a release, Chemistry will secure the release.*In the event of any problem with an EMF that would require a work request, contact RP for initiation
of the work request.2.2 Releasing a WMT Refer to Drawing 7.2, WMT Subsystem.
Radwaste initiates the procedure.
They select the tank to be discharged, recirculate
it for mixing, and obtain a sample.Next, the sample is analyzed.Radwaste delivers the sample to RP for isotopic analysis.RP then generates the Release Discharge Document using the RETDAS Computer Program.RP assigns the next sequential
LWR number and calculates
the recommended
release rates.I Objective#2 I The Recommended
Release Rate is the lesser of:*Maximum System Release Rate for WMT=120 gpm, OR*Allowable Release Rate.The"Allowable
Release Rate" is determined
by the amount of activity present in the tank.RP indicates the"EMF Utilized", which is OEMF-49L for WMT releases.RP next indicates the EMF background
cpm, expected cpm, trip 1,andtrip 2 setpoints.
I Objective#3 I RP then takes the release procedure and the discharge document to the control room.The SRO authorizes
the release by signing the release document.The SRO authorizing
the release ensures the following:
- Ensuresthe
LWR document agrees with the Radwaste procedure (Le., the procedure directs releasing the same tank that is listed on the LWR.)*Operability
of EMF 49 and the discharge release valves (1 WP-35&37).The pump discharge flow meter and the EMF outlet flow meter also needs to be operable.If any of these are inoperable, then the LWR document is returned to Radwaste.OP-MC-WE-RLR
FOR TRAINING PURPOSES ONL Y Page 13 of 55
DUKE ENERGY MCGUIRE OPERATIONS
TRAINING Prior to signing the LWR document, the SRO should review the following:
- The required number of RC pumps are in operation NOTE: The RC minimum flow interlock is set to the minimum#of pumps required for the release.If the total#RC pumps running is less than the selected number, 1WP-35 and 1WP-37 will close.*The"Recommended
Release Rate" is less than or equal to the"Allowable
Release Rate".Objective#4*The proper EMF is utilized.(For a WMT release, this is EMF-49)*A source check has been performed on EMF-49.*The"Expected CPM of the EMF" and the"EMF Trip I Setpoint" are less than the"EMF Trip II Setpoint"*Any special instructions
The RO ensures the LWR number is in autolog (normally logged by Chemistry).
The purpose of the log is to maintain an account, in the control room, of all LWRIGWR releases.The information
contained in the log is:*Release#*Start Time&Date*Stop Time&Date*Volume Released*Any unusual events encountered
during the release Now the release is ready to be started.Radwaste notifies the SRO the discharge is initiated.
The Radwaste technician
aligns the WMT to be discharged
to RC and commences the release.I Objective#5 I Based on an agreement between MNS RP, GO RP, and MNS Radwaste, releases that are interrupted
by a Trip 2 on EMF49 may be reinitiated
up to a maximum of two times without resampling
before terminating
the release procedure.
Specifically, 3 release attempts are allowed.If EMF-49 Trip 2 occurs on the third release attempt, the LWR must be terminated, the WMT must be re-sampled
and new LWR paperwork must be generated.
When the release is terminated, the SRO is notified.Autolog is updated, and the Release document is closed out, with the SRO signing the Release document acknowledging
the completion.
OP-MC-WE-RLR
FOR TRAINING PURPOSES ONL Y Page 15 of 55 REV.13
DUKE ENERGY MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours)LOR 2 OBJECTIVES
N NLL L OBJECTIVELLPP 000 R S R R 0 0 1 State the systems that are used to release radioactive
liquids X X X to the environment.
WERLROO1 2 Given a completed LWR, state the recommended
releaseXX X rate.WERLROO2 3 Given the applicable
procedure and LWR paperwork, review X X X the LWR and determine if a release can be initiated.
WERLROO3 4 Given a completed LWR, state the proper EMF to be used for X X X the release.WERLROO4 5 Evaluate plant parameters
to determine any abnormal system X X X conditions
that may exist.WERLROO5 OP-MC-WE-RLR
FOR TRAINING PURPOSES ONL Y Page 5 of 55
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outline Cross-Level RO SRO reference:
Tier#3 Group#4 KIA#2.4.46 Importance
Rating 4.2 Ability to verify that the alarms are consistent
with the plant conditions.
Proposed Question: SRO 99 Given the following conditions:
A transient has occurred on Unit 2 resulting in the following alarms:*OTOT RUNBACKIROO
STOP ALERT*TREF/T-AUCT
ABNORMAL Reactor power indicates the following:
- N41-1 04.1%*N42-103.2%*N43-1 04.3%*N44-102.9%*Tavg is 590 degrees F Which ONE (1)of the following has occurred, and what is the technical specification
implication
of the event?A.Uncontrolled
Rod Withdrawal;
Linear Heat Rate and Hot Channel Factors may be challenged.
B.Uncontrolled
Rod Withdrawal;
Shutdown Margin assumptions
for anticipated
operational
may be invalid.C.Secondary Steam Leak;Linear Heat Rate and Hot Channel Factors may be challenged.
O.Secondary Steam Leak;Shutdown Margin assumptions
for anticipated
operational
may be invalid.Page 254 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Proposed Answer: A Explanation (Optional):
A is correct because core power is increasing
and LHR is a function of rods.B is incorrect because SDM is a function of several parameters, and the positive reactivity
added by rod withdrawal
is cancelled by the negative reactivity
from power defect and MTC.C is incorrect because a steam leak would result in a higher power, but Tavg would be lower, not higher.Tave is currently about 4-5 degrees above program o is incorrect for same reason as C, and basis is incorrect, but plausible because shutdown margin would be the concern if a steam leak were occurring Technical Reference(s):
AP-O 1 (Rev 14)and Basis Document (Rev 5)TS 3.2.1 Basis CTH-CP Rev 9 (Attach if not previously
provided)Proposed references
to be provided to applicants
during None examination:
Learning Objective:
Question Source: CTH-CP Obj 1 Bank#Modified Bank X#New (Note changes or attach parent)Question History: Question Cognitive Level: Last NRC Exam 2006 Exam 100 Modified Memory or Fundamental
Knowledge Comprehension
or Analysis X10 CFR Part 55 55.41 Content: 55.43 2 Comments: Page 255 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 KA matched because the item evaluates understanding
for the cause of alarms, related to current plant conditions.
SRO level because the item evaluates knowledge of accident analysis assumptions
and core operating limits as stated in TS basis Page 256 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Given the following conditions:
A transient has occurred on Unit 1 resulting in the following alarms:*OTDT RUNBACKIROD
STOP ALERT*ROD CONTROL URGENT FAILURE*OPDT REACTOR TRIP Reactor power indicates the following:
- N41-1 05.2°10*N42-1 06.2°10*N43-1 05.9°10*N44-1 06.1°10*Tavg is 581 degrees F Which ONE (1)of the following has occurred, and which procedure(s)
is/are required to be implemented?
A.Uncontrolled
Rod Withdrawal;
E-O, Reactor Trip or Safety Injection.
B.Uncontrolled
Rod Withdrawal;
AP-14, Rod Control Malfunctions.
C.SG Safety Valve opened coincident
with a rod control failure;E-O, Reactor Trip or Safety Injection.
D.SG Safety Valve opened coincident
with a rod control failure;AP-01, Steam Leak and AP-14, Rod Control Malfunctions.
Answer: C Page 257 of 260 Draft 7
DUKE POWER MCGUIRE OPERATIONS
TRAINING In some applications, heat transfer is discussed in relationship
to a heat transfer BTU/hr BTU Q rate per unit area...............
--A=q"==HEAT FLUX.ft2-hr-ft2-..Q Therefore, If..Q=UA (L\T), then A=U (L\T)=U (T c1ad-Tcoolant).
NOTE: Average HEAT FLUX, at RATED THERMAL POWER (RTP), is 189,800 BTU I hr-ft 2 while MAXIMUM HEAT FLUX, at RTP, is 440,300 BTU I hr-ft 2*Objective#5 One of the variables discussed above, local heat generation
rate, is synonymous
with another term, local power density.Local power density or power density is the term used to describe variations
in power distribution
throughout
the reactor core.Power density, quite simply, is the amount of power being produced per unit volume of the reactor.Therefore, one would expect the units of power density to be some power related term divided by some volume related term.Such as....Power Density=Watts//cm 3 The average power density for either McGuire Unit at full power (100%RTP)is approximately
340 watts/em 3.Since reactor power production
occurs solely within the fuel, power density is power production
per unit volume of fuel.Ideally, if the power produced from the reactor was evenly distributed, every fuel assembly would contribute
an equal amount of the total power, and therefore, every foot of fuel would be producing the average power density.Thus, the power distribution
term, Average Power Density.Objective#1 Since the reactor fuel rods, within tolerances, are dimensionally
identical to one another.A unit length of fuel rod, then, represents
a certain volume of fuel.Therefore, we also define IILinear Heat Generation
Rate" (a power density 1 ft term)as the power produced per linear foot of fuel rod (KW 1ft).See if you can determine average power density by performing
the example problem below.KW OP-MC-CTH-CP
FOR TRAINING PURPOSES ONL Y Page 45 of 305 REV.09
DUKE POWER MCGUIRE OPERATIONS
TRAINING 3.3 Hot Channel Factors Two hot channel factors are specified in our Technical Specifications
as core limits;the Heat Flux Hot Channel Factor and the Nuclear Enthalpy Rise Hot Channel Factor.Excessive fuel and cladding temperatures
must be avoided during reactor operation to prevent fuel rod burnout.This not only applies during normal operation but also during accident conditions, as well.Theoretically
a Hot Channel Factor represents
the specific core location with the worst possible performance
characteristics.
By controlling
this location such that the limits on core performance
are not exceeded, we are somewhat assured that the entire core is operating within limits.These Hot Channel Factors are calculated
by analyzing core data obtained during core (flux)mapping.Objective#1 HEAT FLUX HOT CHANNEL FACTOR The Heat Flux Hot Channel Factor, Fa (X, Y,Z), is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming normal fuel pellet and fuel rod dimensions:
KW peakfi F Q KW averagefi Therefore, Fo (X,Y,Z), Heat Flux Hot Channel Factor, is calculated
based on the data obtained during core or flux mapping with the incore detector system.McGuire UNIT 1 CYCLE 5 PEAKING FACTORS FROM INCORE FLUX MAPS 2.15 2.10 2.05
2.00 1.95 a LL 1.90 1.85 1.80
1.75 1.70 0*Not at full power 40 80 120 160 200 240 280 F o (X,Y,Z)varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution
and to a lesser extent, with moderator temperature.
An example of how it can change with fuel burnup is illustrated
on Training Drawing 7.17, Fa (X,Y,Z)versus Burnup Graph.OP-MC-CTH-CP
BURNUP (EFPD)FOR TRAINING PURPOSES ONL Y Page 101 of 305 REV.09
DUKE POWER MCGUIRE OPERATIONS
TRAINING Fa (X,Y,Z)is measured periodically
using the incore detector system.Approximately
620 different sets of data are taken along the length of each fuel assembly that is mapped.Each"mapped" fuel assembly will provide data from the same fuel elevation (z).This then provides a representative
slice of power distribution
throughout
the core, at various core elevations (z).These measurements
are generally taken with the core at, or near steady state conditions.
Using the measured three dimensional
power distributions, it is possible to derive a measured value for Fa (X,Y,Z).However, because this value represents
a steady state condition, it does not include the variations
in the value of Fa (X,Y,Z)that are present during non-equilibrium
situations.
To account for these possible variations, Fa (X,Y,Z)is limited by pre-calculated
factors to account for perturbations
from the steady state condition.
Objective#18 These pre-calculated
factors include:*Measurement
Uncertainty
Factor (Fa u)Accounts for uncertainties
in the flux mapping process and variations
in fuel rod dimensions.
- Engineering
Hot Channel Factor (Fa E)Provides additional
conservatism
in the hot channel estimate.Typically, a five percent conservatism
is applied to UMT (Measurement
Uncertainty
Factor);UMT=1.05 (1.04 Westinghouse
Fuel), and a three percent conservatism
is applied to MT (Engineering
Hot Channel Factor);MT=1.03 (1.033 Westinghouse
Fuel).The Measured Nuclear Heat Flux Hot Channel Factor is multiplied
by the Engineering
Hot Channel Factor and the Measurement
Uncertainty
Factor to provide the Nuclear Heat Flux Hot Channel Factor at core elevation z.F Q (z)=F Q M*UMT*MT Limits for the Nuclear Heat Flux Hot Channel Factor as specified within the COLR (Core Operating Limit Report)are related to the Rated Thermal Power Nuclear Heat Flux Hot Channel Factor, Fa RTP.OP-MC-CTH-CP
FOR TRAINING PURPOSES ONL Y Page 1 03 of 305 REV.09
Fo(X,Y,Z)B 3.2.1 B 3.2 POWER DISTRIBUTION
LIMITS B 3.2.1 Heat Flux Hot Channel Factor (Fo(X,Y,Z))
BASES BACKGROUND
The purpose of the limits on the values of Fo(X,Y,Z)is to limit the local (Le., pellet)peak power density.The value of Fo(X,Y,Z)varies axially (Z)and radially (X,Y)in the core.Fo(X,Y,Z)is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions.
Therefore, Fo(X,Y,Z)is a measure of the peak fuel pellet power within the reactor core.During power operation, the global power distribution
is limited by LCO 3.2.3, IIAXIAL FLUX DIFFERENCE (AFD),II and LCO 3.2.4,"QUADRANT TILT POWER RATIO (QPTR),II which are directly and continuously
measured process variables.
These LCOs, along with LCO 3.1.6, IIControl Bank Insertion Limits,1I maintain the core limits on power distributions
on a continuous
basis.Fo(X,Y,Z)varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution
and to a lesser extent, with boron concentration
and moderator temperature.
Fo(X,Y,Z)is measured periodically
using the incore detector system.These measurements
are generally taken with the core at, or near steady state conditions.
Using the measured three dimensional
power distributions, it is possible to derive a measured value for Fo(X,Y,Z).
However, because this value represents
a steady state condition, it does not include the variations
in the value of Fo(X,Y,Z)that are presentduringnonequilibrium
situations.
To account for these possible variations, the Fo(X,Y,Z)limit is reduced by precalculated
factors to account for perturbations
from steady state conditions
to the operating limits.Core monitoring
and control under nonsteady state conditions
are accomplished
by operating the core within the limits of the appropriate
LCOs, including the limits on AFD, QPTR, and control rod insertion.
McGuire Units 1 and 2 B 3.2.1-1 Revision No.74
Fo(X,Y,Z)B 3.2.1 BASES APPLICABLE
This LCO precludes core power distributions
that violate SAFETY ANALYSES the following fuel design criteria: a.During a loss of coolant accident (LOCA), the peak cladding temperature
must not exceed 2200°F for small breaks and there is a high level of probability
that the peak cladding temperature
does not exceed 2200°F for large breaks (Ref.1);b.The DNBR calculated
for the hottest fuel rod in the core must be above the approved DNBR limit.(TheLCO alone is not sufficient
to preclude DNB criteria violations
for certain accidents, Le., accidents in which the event itself changes the core power distribution.
For these events, additional
checks are made in the core reload design process against the permissible
statepoint
power distributions.);
c.During an ejected rod accident, the energy deposition
to the fuel must not exceed 280 cal/gm (Ref.2);and d.The control rods must be capable of shutting down the reactor with a minimum required SOM with the highest worth control rod stuck fully withdrawn (Ref.3).Limits on Fo(X,Y,Z)ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid.Other Reference 1 criteria must also be met in LOCAs (e.g., maximum cladding oxidation, maximum hydrogen generation, coolable geometry, transient strain, and long term cooling).However, the peak cladding temperature
is typically most limiting.Fo(X,Y,Z)limits assumed in the LOCA analysis are typically limiting relative to (Le., lower than)the Fo(X,Y,Z)limit assumed in safety analyses for other postulated
accidents.
Therefore, this LCO provides conservative
limits for other postulated
accidents.
Fo(X,Y,Z)satisfies Criterion 2 of 10 CFR 50.36 (Ref.4).LCO The Heat Flux Hot Channel Factor, Fo(X,Y,Z), shall be limited by the following relationships:
McGuire Units 1 and 2 F RTP Fg'(X, Y,Z)-5:_Q-K(Z)P F RTP Fg'(X, Y,Z)-5:-Q-K(Z)0.5 B 3.2.1-2 for P>0.5 for P0.5 Revision No.74
FQ(X,Y,Z)B 3.2.1 BASES LCO (continued)
where: F RTP Q is the FQ(X,Y,Z)limit at RTP provided in the COLR, and is reduced by measurement
uncertainty, K(BU), and manufacturing
tolerances
provided in the COLR, K(Z)is the normalized
FQ(X,Y,Z)as a function of core height provided in the COLR, and P=THERMAL POWER RTP The actual values of F RTP Q, K(BU), and K(Z)are given in the COLR.For relaxed AFD limit operation, FMQ(X,Y,Z)(measured
FQ(X,Y,Z))
is compared against three limits:*Steady state limit, (F RTP dP)*K(Z),*Transient operational
limit, FLQ(X,Y,Z)op, and*Transient RPS limit, FLQ(X,Y,Z)RPS.
A steady state evaluation
requires obtaining an incore flux map in MODE 1.From the incore flux map results we obtain the measured value FMQ(X,Y,Z)
of FQ(X,Y,Z).
Then, FMQ(X,Y,Z)
is adjusted by a radial local peaking factor and compared to F RTP Q which has been reduced by manufacturing
tolerances, K(BU), and flux map measurement
uncertainty.
K(BU)is the normalized
FLQ(X,Y,Z)
as a function of burnup and is provided in the COLR.FLQ(X,Y,Z)op
and FLQ(X,Y,Z)RPS
are cycle dependent design limits to ensure the FQ(X,Y,Z)is met during transients.
The expression
for FLQ(X,Y,Z)op
is:(X ,Y,Z)op=Ft (X ,Y,Z)*M Q (X ,Y,Z)/(UMT
- MT*TILT)McGuire Units 1 and 2 B 3.2.1-3 Revision No.74
Fa(X,Y,Z)B 3.2.1 BASES LCO (continued)
where: FLa(X,Y,Z)Op
is the cycle dependent maximum allowable design peaking factor which ensures that the Fa(X,Y,Z)limit will be preserved for operation within the LCO limits.FLa(X,Y,Z)op
includes allowances
for calculational
and measurement
uncertainties.
F D a(X,Y,Z)is the design power distribution
for Fa provided in the COLR.Ma(X,Y,Z)is the margin remaining in core location X,Y,Z to the LOCA limit in the transient power distribution
and is provided in the COLR for normal operating conditions
and power escalation
testing during startup operations.
UMT and MT are only included in the calculation
of FLa(X,Y,Z)op
if these factors were not included in the LOCA limit.UMT is the measurement
uncertainty.
MT is the engineering
hot channel factor.TILT is the peaking penalty that accounts for allowable quadrant power tilt ratio of 1.02 and is specified in the COLR.The expression
for FLa(X,Y,Z)RPS
is:(X,Y,Z)RPS
=Ft (X ,Y,Z)*Me (X ,Y,Z)/(UMT
- MT*TILT)where: FLa(X,Y,Z)RPS
is the cycle dependent maximum allowable design peaking factor which ensures that the centerlinefuel melt limit will be preserved for operation within the LCO limits.FLa(X,Y,Z)RPS
includes allowances
for calculational
and measurement
uncertainties.
Mc(X,Y,Z)is the margin remaining to the center line fuel melt limit in core location X,Y,Z from the transient power distribution
and is provided in the COLR for normal operating conditions
and power escalation
testing during startup operationso
UMT and MT are only included in the calculation
of FLa(X,Y,Z)RPS
if these factors were not included in the fuel melt limit.McGuire Units 1 and 2 B 3.2.1-4 Revision No.74
Fo(X,Y,Z)B 3.2.1 BASES LCO (continued)
The Fo(X,Y,Z)limits typically define limiting values for core power peaking that precludes peak cladding temperatures
above 2200°F during a small break LOCA and a high level of probability
that the peak cladding temperature
does not exceed 2200°F for a large break LOCA.This LCO requires operation within the bounds assumed in the safety analyses.Calculations
are performed in the core design process to confirm that the core can be controlled
in such a manner during operation that it can stay within the Fo(X,Y,Z)limits.If Fo(X,Y,Z)cannot be maintained
within the steady state LOCA limits, reduction of the core power is required.Violating the steady state LOCA limits for Fo(X,Y,Z)produces unacceptable
consequences
if a design basis event occurs while Fo(X,Y,Z)is outside its specified limits.APPLICABILITY
ACTIONS The Fo(X,Y,Z)limits must be maintained
in MODE 1 to prevent core power distributions
from exceeding the limits assumed in the safety analyses.Applicability
in other MODES is not required because there is either insufficient
stored energy in the fuel or insufficient
energy being transferred
to the reactor coolant to require a limit on the distribution
of core power.The exception to this is the steam line break event, which is assumed for analysis purposes to occur from very low power levels.At these low power levels, measurements
of Fo(X,Y,Z)are not sufficiently
reliable.Operation within analysis limits at these conditions
is inferred from startup physics testing verification
of design predictions
of core parameters
in general.Reducing THERMAL POWER by1%RTP for each 1%by which FMO(X,Y,Z)
exceeds its steady state limit, maintains an acceptable
absolute power density.FMO(X,Y,Z)
is the measured value of Fo(X,Y,Z)and the steady state limit includes factors accounting
for measurement
uncertainty
and manufacturing
tolerances.
The Completion
Time of 15 minutes provides an acceptable
time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable
condition for an extended period of time.McGuire Units 1 and 2 B 3.2.1-5 Revision No.74
Fa(X,Y,Z)B 3.2.1 BASES ACTIONS (continued)
A reduction of the Power Range Neutron Flux-High trip setpoints by Ok, for each 1%by which FMa(X,Y,Z)
exceeds its steady state limit, is a conservative
action for protection
against the consequences
of severe transients
with unanalyzed
power distributions.
The Completion
Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient
considering
the small likelihood
of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance
with Required Action A.1.Reduction in the Overpower 1.\T trip setpoints (valueby1%(in 1.\T span)for each 1%by which FMa(X,Y,Z)
exceeds its steady state limit, is a conservative
action for protection
against the consequences
of severe transients
with unanalyzed
power distributions
since the transient response is limited by the setpoint reduction.
The Completion
Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient
considering
the small likelihood
of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER in accordance
with Required Action A.1.Verification
that FMa(X,Y,Z)
has been restored to within its steady state and transient limits, by performing
SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.1.3 prior to increasing
THERMAL POWER above the limit imposed by Required Action A.1, ensures that core conditions
during operation at higher power levels are consistent
with safety analyses assumptions.
Since FMa(X,Y,Z)
exceeds the steady state limit, the transient operational
limit and possibly the transient RPS limit may be exceeded.By performing
SR 3.2.1.2 and SR 3.2.1.3, appropriate
actions with respect to reductions
in AFD limits and OT 1.\T trip setpoints will be performed ensuring that core conditions
during operational
and Condition 2 transients
are maintained
within the assumptions
of the safety analysis.B.1 and B.2 The operational
margin during transient operations
is based on the relationship
between FMa(X,Y,Z)
and the transient operational
limit, FLa(X,Y,Z)op, as follows: McGuire Units 1 and 2 B 3.2.1-6 Revision No.74
Fo(X,Y,Z)B 3.2.1 BASES ACTIONS (continued)
0/0 Operational
Margin=[1- (X,
- 1000/0 F Q (X, Y,Z)If the operational
margin is less than zero, then FMO(X,Y,Z)
is greater than FLO(X,Y,Z)op
and there exists a potential for exceeding the peak local power assumed in the core in a LOCA or in the loss of flow accidents.
Reducing the AFD by1 ok>from the COLR limit for each 1%by which FMO(X,Y,Z)
exceeds the operational
limit within the allowed Completion
Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> restricts the axial flux distribution
such that even if a transient occurred, core peaking factors are not exceeded.Adjusting the transient operational
limit by the equivalent
change in AFD limits establishes
the appropriate
revised surveillance
limits.C.1 and C.2 The margin contained within the reactor protection
system (RPS)OvertemperatureT setpoints during transient operations
is based on the relationship
between FMO(X,Y,Z)
and the RPS limit, FLO(X,Y,Z)RPS, as follows: 0A>RPS Margin=[1-F: (X, Y,Z))*1000/0 (X Y Z)RPSQ" If the RPS margin is less than zero, then FMO(X,Y,Z)
is greater than FLO(X,Y,Z)RPS
and there exists a potential for FMO(X,Y,Z)
to exceed peak clad temperature
limits during certain Condition 2 transients.
The OvertemperatureT K1 value is required to be reduced as follows: K1 ADJUSTED=K1-I KSLOPE*%RPS Margin I Where K1 ADJUSTED is the reduced OvertemperatureT K1 value KSLOPE is a penalty factor used to reduce K1 and is defined in the COLR%RPS Margin is the most negative margin determined
abovem McGuire Units 1 and 2 B 3.2.1-7 Revision No.74
Fa(X,Y,Z)B 3.2.1 BASES ACTIONS (continued)
Reducing the OvertemperatureT trip setpoint from the COLR limit is a conservative
action for protection
againsttheconsequences
of transients
since this adjustment
limits the peak transient power level which can be achieved during an anticipated
operational
occurrence.
Once the OTT trip setpoint is reduced, the available margin is increased.
An adjustment
is then necessary in the FLa(X,Y,Z)RPS
limit, using the increased margin, in order to restore compliance
with the LCO and exit the condition.
These adjustments
maintain a constant margin and ensure that centerline
fuel melt does not occur.The Completion
Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient
considering
the small likelihood
of a limiting transient in this time period.Adjusting the transient RPS limit by the equivalent
change in OTT trip setpoint establishes
the appropriate
revised surveillance
limit.If Required Actions A.1 through A.4, B.1, or C.1 are not met within their associated
Completion
Times, the plant must be placed in a mode or condition in which the LCO requirements
are not applicable.
This is done-by placing the plant in at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.This allowed Completion
Time is reasonable
based on operating experience
regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging
plant systems.SURVEILLANCE
REQU I REMENTS SR 3.2.1.1 , SR 3.2.1.2, and SR 3.2.1.3 are modified by a Note.The Note applies during the first power ascension after a refueling.
It states that THERMAL POWER may be increased until an equilibrium
power level has been achieved at which a power distribution
map can be obtained.This allowance is modified, however, by one of the Frequency conditions
that requires verification
that FMa(X,Y,Z)
is within the specified limits after a power rise of.2=.10%RTP over the THERMAL POWER at which it was last verified to be within specified limits.Because FMO(X,Y,Z)
could not have previously
been measured in this reload core, power may be increased to RTP prior to an equilibrium
verification
of FMO(X,Y,Z)
provided nonequilibrium
measurements
of FMa(X,Y,Z)
are performed atvariouspower
levels during startup physics testing.This ensures that some determination
of FMa(X,Y,Z)
is made at a lower power level at which adequate margin is available before going to 100%RTP.The Frequency condition is not intended to require verification
of these parameters
after every 1 o ok>increase in power level above the last McGuire Units 1 and 2 B 3.2.1-8 Revision No.74
Fa(X,Y,Z)B 3.2.1 BASES SURVEILLANCE
REQUIREMENTS (continued)
verification.
It only requires verification
after a power level is achieved for extended operation that is 10%higher than that power at which Fa was last measured.SR 3.2.1.1 Verification
that FMa(X,Y,Z)
is within its specified steady state limits involves either increasing
FMa(X,Y,Z)
to allow for manufacturing
tolerance, K(BU), and measurement
uncertainties
for the case where these factors are not included in the Fa limit.For the case where these factors are included, a direct comparison
of FMa(X,Y,Z)
to the Fa limit can be performed.
Specifically, FMa(X,Y,Z)
is the measured value of Fa(X,Y,Z)obtained from incore flux map results.Values for the manufacturing
tolerance, K(BU), and measurement
uncertainty
are specified in the COLR.The limit with which FMa(X,Y,Z)
is compared varies inversely with power above 50%RTP and directly with functions called K(Z)and K(BU)provided in the COLR.If THERMAL POWER has been increased by10%RTP since the last determination
of FMa(X,Y,Z), another evaluation
of this factor is required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium
conditions
at this higher power level (to ensure that FMa(X,Y,Z)
values have decreased sufficiently
with power increase to stay within the LCO limits).The Frequency of 31 EFPD is adequate to monitor the change of power distribution
with core burnup because such changes are slow and well controlled
when the plant is operated in accordance
with the Technical Specifications (TS).SR 3.2.1.2 and 3.2.1.3 The nuclear design process includes calculations
performed to determine that the core can be operated within the Fa(X,Y,Z)limits.Because flux maps are taken in steady state conditions, the variations
in power distribution
resulting from normal operational
maneuvers are not present in the flux map data.These variations
are, however, conservatively
calculated
by considering
a wide range of unit maneuvers in normal operation.
The maximum peaking factor increase over steady state values, is determined
by a maneuvering
analysis (Ref.5).McGuire Units 1 and 2 B 3.2.1-9 Revision No.74
Fo(X,Y,Z)B 3.2.1 BASES SURVEILLANCE
REQUIREMENTS (continued)
The limit with which FMO(X,Y,Z)
is compared varies and is provided in the COLR.No additional
uncertainties
are applied to the measured Fo(X,Y,Z)because the limits already include uncertainties.
FLO(X,Y,Z)Op
and FLO(X,Y,Z)RPS
limits are not applicable
for the following axial core regions, measured in percent of core height: a.Lower core region, from 0 toinclusive;
and b.Upper core region, from 85 to 100%inclusive.
The top and bottom 15%of the core are excluded from the evaluation
because of the low probability
that these regions would be more limiting in the safety analyses and because of the difficulty
of making a precise measurement
in these regions.This Surveillance
has been modified by a Note that may require that more frequent surveillances
be performed.
If FMO(X,Y,Z)
is evaluated and found to be within the applicable
transient limit, an evaluation
is required to account for any increase to FMO(X,Y,Z)
that may occur and cause the Fo(X,Y,Z)limit to be exceeded before the next required Fo(X,Y,Z)evaluation.
In addition to ensuring via surveillance
that the heat flux hot channel factor is within its limits when a measurement
is taken, there are also requirements
to extrapolate
trends in both the measured hot channel factor and in its operational
and RPS limits.Two extrapolations
are performed for each of these two limits:1.The first extrapolation
determines
whether the measured heat flux hot channel factor is likely to exceed its limit prior to the next performance
of the SR.2.The second extrapolation
determines
whether, prior to the next performance
of the SR, the ratio of the measured heat flux hot channel factor to the limit is likely to decrease below the value of that ratio when the measurement
was taken.Each of these extrapolations
is applied separately
to each of the operational
and RPS heat flux hot channel factor limits.If both of the extrapolations
for a given limit are unfavorable, i.e., if the extrapolated
factor is expected to exceed the extrapolated
limit and the extrapolated
factor is expected to become a larger fraction of the extrapolated
limit McGuire Units 1 and 2 B 3.2.1-10 Revision No.74
Fo(X,Y,Z)B 3.2.1 BASES SURVEILLANCE
REQUIREMENTS (continued)
than the measured factor is of the current limit, additional
actions must be taken.These actions are to meet the Fo(X,Y,Z)limit with the last FMO(X,Y,Z)
increased by the appropriate
factor specified in the COLR or to evaluate Fo(X,Y,Z)prior to the projected point in time when the extrapolated
valuesareexpected
to exceed the extrapolated
limits.These alternative
requirements
attempt to prevent Fo(X,Y,Z)from exceeding its limit for any significant
period of time without detection using the best available data.FMO(X,Y,Z)
is not required to be extrapolated
for the initial flux map taken after reaching equilibrium
conditions
since the initial flux map establishes
the baseline measurement
for future trending.Also, extrapolation
of FMO(X,Y,Z)
limits are not valid for core locations that were previously
rodded, or for core locations that were previously
withinof the core height about the demand position of the rod tip.Fo(X,Y,Z)is verified at power levels 2:: 10%RTP above the THERMAL POWER of its last verification, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium
conditions
to ensure that Fo(X,Y,Z)is within itslimitat higher power levels.The Surveillance
Frequency of 31 EFPD is adequate to monitor the change of power distribution
with core burnup.The Surveillance
may be done more frequently
if required by the results of Fo(X,Y,Z)evaluations.
The Frequency of 31 EFPD is adequate to monitor the change of power distribution
because such a change is sufficiently
slow, when the plant is operated in accordance
with the TS, to preclude adverse peaking factors between 31 day surveillances.
REFERENCES
1.10 CFR 50.46.2.UFSAR Section 15.4.8.3.10 CFR 50, Appendix A, GDC 26.4.10 CFR 50.36, Technical Specifications, (c)(2)(ii).
5.DPC-NE-2011
PA IIDuke Power Company Nuclear Design Methodology
for Core Operating Limits of Westinghouse
Reactors".
McGuire Units 1 and 2 B 3.2.1-11 Revision No.74
MNS A P/2/A/55 0%1 UNIT 2 ACTION/EXPECTED
RESPONSE B.Symptoms STEAM LEAK PAGE NO.2 of 37 Rev.14 RESPONSE NOT OBTAINED*Reactor power greater than turbine power*Reactor power greater than 1000/0*IIP/R OVER POWER ROD STOP" alarm*NC T-Ave going down in an uncontrolled
manner*High containment
pressure, temperature, humidity, or sump level without abnormal radiation*Loss of secondary inventory*Observed secondary steam leak.
AP/1 and 21A15500/001 (Steam Leak)INTRODUCTION
This procedure directs the required Operator action to be taken for a steam leak.It is written for all modes of operation, but the plant response and Operator actions are largely dependent on the mode of operation and the severity of the leak.Summary For relatively
small steam breaks, normal plant control systems are capable of maintaining
nominal or near nominal operating conditions.
For a small steamline break upstream of the turbine stop valves, the system transient response would be similar to a step load increase.The secondary system would indicate an increase in load with a resultant decrease in primary system average temperature
and pressure.The control rods would withdraw from the core in an effort to restore the primary average temperature
if the rod control system was in an automatic mode of operation.
Due to the apparent increased load, the steam flow from the steam generators
would be increasing
in at least one loop, depending upon the location of the break.If the break occurred in the steam header, all loops would experience
increased steam flow.Due to the increased steam flow, the feedwater control valves would modulate to a more open position in an attempt to maintain steam generator water level.As a result, the main feed flow in at least one loop (all loops if break is in steam header)would be increased.
Another indication
of this type of break would be a decreasing
water level in the condenser hotwell.A containment
temperature
and/or pressure increase may be observed if the break occurred inside containment.
If the break was outside containment, an audible or visual confirmation
of the break may be possible.A drop in generator MW output may also be observed.Larger size breaks may require reactor trip and/or safety injection.
A different set of symptoms might be encountered
for steam leaks that occur downstream
of the turbine (on extraction
lines, MSRl s , and feedwater heaters).For these locations, it may be possible to observe a change in plant efficiency;
however, an audible or visual indication
may be the first symptom encountered.
ENTRY CONDITIONS
This procedure can be entered any time the listed symptoms are encountered.
It should be noted that the symptom"Observed secondary steam leak" is the only symptom that definitively
identifies
a steam leak (and even then the magnitude of the leak may be considered
for entry conditions).
The other symptoms could indicate a steam leak, or some other event.In some cases the combination
of symptoms can be the best indication
the event is a steam leak and not some other event.Page 2 of 26 RevS
DUKE POWER MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours}NLO NLOR LPRO LPSO LOR 81616 16 OBJECTIVES
5NNLL LLLPP 0 E OBJECTIVE00 R 5 R Q R 0 0 1 Define the following terms associated
with core performance:XX X*Fuel rod burnout*Heat flux*Departure from nucleate boiling (DNB)*Critical heat flux (CHF)*Departure from nucleate boiling ratio (DNBR)*Linear heat generation
rate*Average power density*Local power density*Axial flux difference (AFD)*AFD Target*Quadrant power tilt ratio (QPTR)*Heat flux hot channel factor (Fa)*Enthalpy rise hot channel factor (F ilH)CTHCPOO1 2 Using a diagram of heat flux versus differential
temperatureXX X X between the cladding surface and the reactor coolant, identify and explain how the following affect fuel rod heat transfer, fuel and cladding temperature: (Refer to Training Drawing 7.1, Nucleate Boiling Curve).*Convective
heat transfer region*Nucleate boiling region*Departure from nucleate boiling*Transition (partial film)boiling region*Film boiling region*Critical heat flux CTHCPOO2 3 Describe how the Critical Heat Flux (CHF)changes with XXXX changes in reactor coolant flow, average reactor coolant temperature, and reactor coolant pressure.CTHCPOO3 OP-MC-CTH-CP
FOR TRAINING PURPOSES ONL Y Page 9 of 305 REV.09
I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 Examination
Outline Cross-Level RO SRO reference:
Tier#3 Group#4 KIA#2.4.8 Importance
Rating 4.5 Kno'vvledge
of hovv abnorrnal operating procedures
are used in conjunction
with EOpls, Proposed Question: SRO 100 Given the following:
- Unit 1 was at 100%power.*A complete loss of RN occurred.*The crew entered AP/20, Loss of RN.*The operators attempted to manually trip the reactor but the trip breakers failed to open.Which ONE of the following statements
correctly describes the proper procedural
flow path for these conditions?
A.Go directly to FR-S.1, Response to Nuclear Power Generation/ATWS, and perform concurrently
with AP/20.Go to E-O, Reactor Trip or Safety Injection, as directed by FR-S.1.B.Enter E-O and immediately
transition
to FR-S.1;continuing
in AP/20 only after exit from the EOP network.C.Enter E-O, continuing
in AP/20 until transition
to FR-S.1.AP/20 may only be performed when FR-S.1 is complete.D.Enter E-O and immediately
transition
to FR-S.1 while continuing
on in AP/20 as time and conditions
permit.Proposed Answer: 0 Explanation (Optional):
A.Incorrect.
No direct EOP entry to FR-S.1.Performance
of these 2 procedures
is opposite of what would be performed B.Incorrect.
AP/20 may be performed concurrently
because it provides Page 258 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Form ES-401-5 support for EOP use c.Incorrect.
Use of AP/20 may be restricted
when ECCS is actuated, but not by use of FR unless it clashes with steps in FR.Generally, AP use is not advisable in EPs, but may be used if required to support performance
of EPs D.Correct.OMP 4-3 p16, 17.22 Technical Reference(s)(Attach if not previously
provided)----------
AP-20, Rev 23 and Basis Document (Rev 4)EP-F)Rev 7 E-O Rev 24 Proposed references
to be provided to applicants
during None examination:
Learning Objective:
EP-FO Obj 3 (Note changes or attach parent)----Bank#X----Modified Bank#New Question Source: Question History: Last NRC Exam Wolf Creek 2007 Question Cognitive Level: Memory or Fundamental
Knowledge Comprehension
or Analysis X 10 CFR Part 55 55.41 Content: 55.43 5 Comments: Same basic question but applied to McGuire, so distractors
and procedures
are different KA is matched because the item evaluates use of an AOP with the EOPs.SRO level because the applicant must determine procedure usage requirements
for the given plant conditions.
Page 259 of 260 Draft 7
ES-401 Sample Written Examination
Question Worksheet Page 260 of 260 Draft 7 Form ES-401-5
7.15.1 OMP4-3 Page 17 of35 Implementing
CSF Path Procedures
7.15.1.1 7.15.1.2 7.15.1.3 7.15.1.4 CSF procedures
are NOT to be implemented
prior to transition
from EP/1,2/A/5000/E-O (Reactor Trip or Safety Injection).
IF a CSF path is red or orange while the operating crew is in EP/1,2/A/5000/E-O, but has turned to green upon transition
from E-O, the CSF procedure which was in alarm shall NOT be implemented.
IF the CSF path is yellow, it shall be handled as any other yellow path.procedure
per Section 7.15.1.7.IF there are any valid red or orange path CSF's on transition
from E-O (unless transition
is to EP/1,2/A/5000/ECA-O (Loss of All AC Power), the associated
CSF procedure shall be implemented.
IF a valid red or orange path flickers into alarm on SPDS but returns to green prior to the crew validating
the condition and implementing
the procedure (implementation
of procedure being that the SRO either hands out fold-out pages or starts reading from the procedure), the CSF procedure shall NOT be implemented.
IF the CSF path is yellow, it shall be handled as any other yellow path procedure per Section 7.15.1.7.Likewise, if a valid red path or orange path goes into alarm during performance
of a higher priority CSF procedure, but returns to green prior to transition
from the higher priority CSF path procedure to the lower priority CSF procedure, the associated
CSF procedure shall NOT be implemented.
IF a CSF procedure directs the operator to return to the procedure and step in effect, AND the corresponding
status tree continues to display the offnormal conditions, the corresponding
CSF procedure does NOT have to be implemented
again, since all recovery actions have been completed.
However, if the same status tree subsequently
changes to a valid higher priority condition, OR if it changes to lower condition and returns to higher priority condition again, the corresponding
CSF procedure shall be implemented
as required by its priority.Red Path IF any valid red path is encountered
during monitoring, the operator is required to immediately
implement the corresponding
in progress shall be discontinued.
IF during the performance
of any red path procedure, a valid red condition of higher priority arises, the higher priority condition should be addressed first, and the lower priority redpathprocedure
suspended.
DUKE POWER MCGUIRE OPERATIONS
TRAINING 2.0 PROCEDURE SERIES BACKGROUND (continued}
Once the Status Trees are being monitored, the following rules of usage apply:1.The Status Trees should be continuously
monitored in order of Critical Safety Function priority.2.CSF procedures
are not to be implemented
prior to transition
from E-O, Reactor Trip or Safety Injection.
If a CSF path is red or orange while the operating crew is in E-O, but has turned to green upon transition
from E-O, the CSF procedure, which was in alarm, shall not be implemented.
If the CSF path is yellow, it shall be handled as any other yellow path procedure.
If there are any valid red or orange path CSFs on transition
from E-O (unless the transition
is to ECA-O (Loss of All AC Power), the associated
CSF procedure shall be implemented.
3.If a valid red or orange path flickers into alarm on SPDS but returns to green prior to the crew validating
the condition and implementing
the procedure (implementation
of procedure being that the SRO either hands out fold-out pages or starts reading from the procedure), the CSF procedure shall not be implemented.
If the CSF path is yellow, it shall be handled as any other yellow path procedure.
Likewise, if a valid red path or orange path goes into alarm during performance
of a higher priority CSF procedure, but returns to green prior to transition
from the higher priority CSF path procedure to the lower priority CSF procedure, the associated
CSF procedure shall not be implemented.
If the CSF path is yellow, it shall be handled as any other yellow path procedure.
4.If a CSF procedure directs the operator to return to the procedure and step in effect, AND the corresponding
status tree continues to display the off-normal
conditions, THEN the corresponding
CSF procedure doesn't have to be implemented
again, since all recovery actions have been completed.
However, if the same status tree subsequently
changes to a valid higher priority condition, (ORifit changes to lower condition and returns to higher priority condition again), THEN the corresponding
CSF procedure shall be implemented
as required by its priority.5.Once status tree monitoring
is initiated, the STA should monitor status tree continuously
if an orange or red path condition exists.If no condition more serious than yellow is found, monitoring
frequency may be reduced to 10-20 minutes unless some significant
change in plant status occurs.Status tree monitoring
may be performed using the OAC SPDS display or F-O (Critical Safety Function Status Trees).If the OAC SPDS display is being used, the STA will validate the OAC SPDS status every 10-20 minutes using control board indications.
If the STA is not available, the OSM shall assume the ST A responsibilities
or delegate the ST A responsibilities
to another licensed operator.OP-MC-EP-FO
FOR TRAINING PURPOSES ONL Y Page 13 of 83 REV.07
7.14.2 OMP4-3 Page 16 of 35 The configuration
control cards filled out in Step 7.14.1 shall be handled per the following two situations:
- Without Operations
Support Center (OSC)activation
The configuration
control card will be handled by OPS shift per SOMP 02-01 (Safety Tagging and Configuration
Control).*With OSC activation
WHEN the OSC is activated, OPS will report to the OSC and shall bring with them all configuration
control cards that have been filled out.The cards taken to the OSC shall be given to the OPS SRO in the OSC.For handling cards in the OSC, refer to RP/0/A/5700/020 (Activation
of the Operations
Support Center (OSC)).7.15 Usage of Status Trees There are six different trees, each one evaluating
a separate Critical Safety Function (CSF)of the plant.Color-coding
of the status tree end points will be either red, orange, yellow, or green, with green representing
a"satisfied" safety status.Each non-green color represents
an action level that should be addressed according to the Rules of Priority as discussed below.The six Status Trees are always evaluated in the sequence:*Subcriticality
- Core Cooling*Heat Sink*Integrity*Containment
- Inventory IF identical color priorities
are found on different trees during monitoring, the required action priority is determined
by this sequence.Initial monitoring
of the status trees should begin on either of the following conditions:
- As directed by an action step in EP/1,2/A/5000/E-0 (Reactor Trip or Safety Injection).
- WHEN a transfer is made out of the Safety Injection procedure to another EP.An exception to this is that CSF procedures
are NOT required to be implemented
during the Loss of All AC Power EP since none of the electrically
powered safeguards
equipment can be used.WHEN power is subsequently
restored, EP/1,2/N5000/ECA-0.1
or 0.2 (Loss of All AC Power Recovery procedures)
will direct the operator when implementing
CSF procedures
is required.
OMP4-3 Page 22 of 35 7.18 Multiple Use ofEPs and APse The Control Room SRO will determine how many procedures
can be implemented
at a time and their priority based on manpower availability
and the particular
event in progress.More than one EP shall NOT be run concurrently
unless directed by the procedure.
Generally the use of APs in conjunction
with EPs should be avoided.In some instances it would be proper to use an AP concurrently
during a major accident which is being addressed by the EPs.An example of this is upon loss of all Nuclear Service Water in the middle of an accident, the operators would need to utilize the AP for Loss of RN also.IF an AP is used during an Sf I event, USE CAUTION.APs are generally written assuming an Sf I has NOT occurred (exception
-APf35, ECCS Actuation During Plant Shutdown).
Evaluate any AP steps in post SII events to ensure the steps do NOT conflict with any EP in effect.NOT all AP actions would be appropriate
if an Sf I occurred.(Enclosures
in EPfG-1 (Generic Enclosures)
may be used when reference by EPs or APs.)
MNS EP/1/A/SOOO/E-O
UNITl REACTOR TRIP OR SAFETY INJECTION PAGE NO.3 of 36 Rev.24 ACTION/EXPECTED
RESPONSE c.Operator Actions1.Monitor Foldout page.G)Check Reactor Trip:*All rod bottom lights-LIT*Reactor trip and bypass breakersOPEN*I/R amps-GOING DOWN.G)Check Turbine Trip:*All throttle valves-CLOSED.-0 Check 1 ETA and 1 ETB-ENERGIZED.
RESPONSE NOT OBTAINED Perform the following:
a.Trip reactor.b.IF reactor will not trip, THEN:*Implement EP/1/AJSOOO/F-O (Critical Safety Function Status Trees).*GO TO EP/1/A/SOOO/FR-S.1 (Response To Nuclear Power Generation/ATWS).
Perform the following:
a.Trip turbine.b.IF turbine will not trip, THEN:_1)Place turbine in manual._2)Close governor valves in fast action.3)IF governor valves will not close, THEN close:*All MSIVs*All MSIV bypass valves.Perform the following:
a.IF both busses de-energized, THEN GO TO EP/1/A/5000/ECA-O.O (Loss Of All AC Power)e b.WHEN time allows, THEN try to restore power to de-energized
bus PER AP/1/A/SSOO/07 (Loss of Electrical
Power)while continuing
with this procedure.
MNS EP/1/A/5000/FR-S.1
UNITl A.Purpose RESPONSE TO NUCLEAR POWER GENERATION/ATWS
PAGE NO.1 of 29 Rev.10 This procedure provides actions to add negative reactivity
to a core which is observed to be critical when expected to be shut down.B.Symptoms or Entry Conditions
This procedure is entered from:*EP/1/A/5000/E-O (Reactor Trip Or Safety Injection), Step 2, when reactor trip is not verified and manual trip is not effective.
- EP/1/A/5000/F-O (Critical Safety Function Status Trees)(Subcriticality), on either a red or orange condition.
MNS AP/1/A/5500/20
UNITl LOSS OF RN Case I Loss of Operating RN Train PAGE NO.17 of 99 Rev.23 ACTION/EXPECTED
RESPONSE 21.Check NC pumps as follows: a.Any NC pump-ON.RESPONSE NOT OBTAINED a.GO TO Step 22.b.NC pump stator winding temperatureLESS THAN 311°F.b.Perform the following:
_1)Secure any dilution in progress.2)Open the following:*1 NV-221 A (NV Pumps Suct From FWST)*1 NV-222B (NV Pumps Suct From FWST).3)Close the following:*1 NV-141A (VCT Outlet Isol)*1 NV-142B (VCT Outlet Isol)._4)Start TD CA pump._5)Maintain S/G NR levels greater than 17%to avoid auto start of MD CA pumps._6)Trip reactor._7)WHEN reactor is tripped, THEN trip all NC pumps._8)Have available operator continue to monitor bearing temperatures
on running pumps._9)WHEN time allows, THEN continue with Case I, starting at Step 220_10)GO TO EP/1/A/5000/E-0 (Reactor Trip or Safety Injection).
c.Monitor stator winding temperatures.
d.IF AT ANY TIME any NC pump stator winding temperature
reaches 311°F, THEN perform Step 21.
AP/1 and 21A15500/020 (Loss of RN)CASE I STEP 21: PURPOSE: Ensure protection
for NCPs without RN cooling.DISCUSSION:
Without RN cooling to the NCP motor coolers, the stator temperatures
will increase to the trip criteria (311°F)in about 20 minutes.According to engineering, if temps go up a couple more degrees while actions in the RND are performed, that's ok.Keep in mind that the thermocouple
location is probably not measuring the hottest spot in the NCP stator.When the OAC reaches 311, there are probably areas that are 10 degrees hotter.That's ok as long as we get the pumps off within a couple minutes.The danger zone for the hottest spot starts around 330 deg.Securing dilution prior to tripping NCPs should reduce the risks of highly diluted pockets of water from forming in the NC System (PIP M-99-0222).
Swapping charging pump suction to the FWST ensures the VCT will not heat up excessively
for NCP seal injection and NV Pump NPSH concerns.Note that as the KC System temperature
heats up, letdown and NV Pump recirc back to the VCT would cause it to heat up.The TD CA Pump is manually started prior to tripping the reactor to avoid the auto start of the MD CA Pumps.This will also avoid the subsequent
auto start on the RN pumps off the MD CA Pumps.The TO CA Pumps do not have RN cooling and will not overheat like the MD CA Pumps.Therefore, it is the preferred CA pump to run in this scenario.There is a trade-off with running the TD CA Pump in that it may contribute
to a post-trip cooldown.Waiting for the reactor to trip prior to tripping NCPs avoids loss of NC flow during an ATWS.Before going to E-O, direction is given to have another operator continue with this AP.This is as high or higher priority than many of the EP actions, since the equipment assumed available in the EPs is cooled by RN, which is not available at this point in the AP.The highest priority in this scenario is the maintenance
of NCP seal cooling and the restoration
of RN (the actions of this AP).For this reason, direction is given to continue with Case I of the AP, as a higher priority than continuing
with Case II at this point.REFERENCES:
NC Pump manual (MCM-1201.01-193)
PIP M-99-0222 Page 15 of 37 Rev 4
DUKE POWER MCGUIRE OPERATIONS
TRAINING CLASSROOM TIME (Hours}NLO I NLOR I LP;O I LP:O OBJECTIVES
LOR 2 S N N L L L E OBJECTIVELL P P 000 R S R Q R 0 0 1 State the purpose of each of the six CSF Status Trees.X X 2 Explain the priority system associated
with the CSF status X X X trees.3 Explain the IIRules of Usage ll for Critical Safety Function X X X status trees.4 Explain the bases for all blocks in the six Status Trees.X X X OP-MC-EP-FO
FOR TRAINING PURPOSES ONL Y Page 5 of 83 REV.07