ML13297A317
| ML13297A317 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/18/2013 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| References | |
| Download: ML13297A317 (28) | |
Text
ES-401 PWR Examination Outline Form ES-401-2 Facility SEQUOYAH Date of Exam MAY2013 ROKIACajojyPoints SAC-Only_Points Tier Group KKKKKKAAAAG A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total 1 1 iL 318 3 3 6 Emergency &
Abnormal 2 1 1 2 N/A N/A j. 9 2 2 4 Plant Evolutions Tier Totals 4 4 5 5 5 4 27 5 5 10 1 3 2 3 2 2 3 3 2 3 3 2 28 3 2 5 2.
Plant 2 I+/-i_LI_L+/-+/-II2 0 2 1 3 Sys ems Tier Totals 4 3 4 2_ I 2_ I +/- 2 38 5 3 8 3* Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 3 2 3 1 2 2 2 Note: 1 Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A 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.
a 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/evolutions that are not included on the outline should be added. Refer to Section Di .b of ES-401 for guidance regarding the elimination of inappropriate K/A 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 K/As 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 K/A categories.
7* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section Di .b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) 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 ofthe K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401 -3. Limit SRO selections to K/As that are linked to 1 0 CFR 5543.
ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRO 007EK1 02 Reactor Trap Stabilization Recovery 34 38 fl LI El El LI El LI Li Shutdown margin
/1 008AK2O1 Pressurizer Vapor Space Accident I 3 2.7 2.7 [j [J j [ El LI Valves 009EK1 .02 Small Break LOCA I 3 3.5 4.2 [J fl ] Use of steam tables 01 1 EG2.1 .31 Large Break LOCA I 3 4.6 4.3 [ [J JJ ] [ Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup.
01 5AG2.1 28 RCP Malfunctions I 4 4.1 4.1 fl ] 1] ] Knowledge of the purpose and function of major system components and controls.
022AK305 Loss of Rx Coolant Makeup I 2 3.2 3.4 [ [ [j [J Need to avoid plant transients 025AK2 05 Loss of RHR System I 4 26 26 ] [ J ] [J LI El Reactor building sump 026AA1 06 Loss of Component Cooling Water I 8 2.9 2.9 Control of flow rates to components cooled by the CCWS A1O9 ATWS I 1 4 3.6 [ [ ] J j Manual rod control 040AA2.02 -
Steam Line Rupture Excessive Heat 4.6 4.7 Conditions requiring a reactor trip
[j [1 Transfer /4 054AK1.O1 Loss of Main Feedwater / 4 4.1 4.3 El El MFW line break depressurizes the S/G (similar to a steam line break)
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ES-401, REV 9 T1G1 PWR EXAMINATiON OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
AC SRQ 056AG2.2.4 Loss of Off-site Power I 6 3.6 3.6 i : , (multi-unit) Ability to explain the variations in control board layouts, systems, instrumentation and procedural actions between units at a facility.
062AA2.O1 Loss of Nuclear Svc Water 1 4 2.9 3.5 E E E Z E E E E E Location of a leak in the SWS 065AK3.08 Loss of Instrument Air I 8 3.7 3.9 fl E [ El El LI E D LZ D Actions contained in EOP for loss of instrument air O77AAlO3Generator Voltage and Electric Grid 38 37 [ [ J [ ] Voltatge regulator controls Disturbances I 6 WEO4EA2 2 LOCA Outside Containment I 3 36 42 [ [ J Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.
WEO5EK2 2 Inadequate Heat Transfer Loss of 39 42 ] fl J Facility s heat removal systems including primary Secondary Heat Sink /4 coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.
WE11EK4 Loss of Emergency Coolant Recirc. /4 3.6 3.8 E AC or SAC function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.
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ES-401, REV 9 Ti G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRO OO1AA2.05 Continuous Rod Withdrawal I 1 4.4 4.6 E i Uncontrolled rod withdrawal from available indications 036AK2 02 Fuel Handling Accident / 8 34 39 Radiation monitoring equipment (portable and installed) 037AG2.2.44 Steam Generator Tube Leak / 3 4.2 4.4 9 to interpret control room Ai indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 051 AA1 .04 Loss of Condenser Vacuum / 4 25 25 DDEiDDDLJDEE Rodposttion 059AK3.02 Accidental Liquid RadWaste Rel. /9 32 45 D El i El LI El D LI E LZ Li Implementation of E plan 074EA2.07 mad. Core Cooling /4 41 47 D 1 Z Ji LI LI E -
The difference between a LOCA and inadequate core cooling from trends and indicators WEO3EK1 .1 -
LOCA Cooldown Depress. /4 34 4.0 - Components, capacity, and function of emergency Z -
EZ systems.
WE14EA1 .1 Loss of CTMT Integrity /5 3.7 3.7 Components and functions of control and safety systems, D E fl E E H E including instrumentation, signals, interlocks, failure modes and automatic and manual features.
WE15EK3.1 Containment Flooding / 5 2.7 2.9 Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure and reactivity changes and operating limitations and reasons for these operating characteristics, Page 1 of 1 2/11/2013 1:40PM
ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRO 003K6.02 Reactor Coolant Pump 2.7 3.1 r r RCP seals and seal watersupply 004A2.06 Chemical and Volume Control 4.2 4.3 E V Inadvertent boration/dilution 004A412 Chemical and Volume Control 38 33 E Z E Li LI El El El E Boration/dilution batch control 005A1 03 Residual Heat Removal 2.5 2.6 El LI El L El El E LI El D Closed cooling water flow rate and temperature 006K5.05 Emergency Core Cooling 34 38 El D El El Li D LI D El E Effects of pressure on a solid system 007A3O1 Pressurizer Relief/Quench Tank 27 29 E E El LI LI LI D D EL Z Components which discharge to the PRT 007G2.1 .20 Pressurizer Relief/Quench Tank 4.6 4.6 E Z E Z Z Z Z Z E Ability to execute procedure steps.
008A3.06 Component Cooling Water 2.5 2.5 1 Typical CCW pump operating conditions, including vibra D E E Z D E E E tion and sound levels and motor current 010K6.O1 Pressurizer Pressure Control 2.7 3.1 E E EZ E E Pressure detection systems E E E E 0106.03 Pressurizer Pressure Control 3.2 3.6 PZR sprays and heaters 012K2.O1 Reactor Protection 3,3 3.7 RPS channels, components and interconnections Page 1 of 3 2/11/2013 1:40PM
ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRO 013K4.07 Engineered Safety Features Actuation 3.7 4.1 i p Power supply loss
[ Z 022K4.03 Containment Cooling 3.6 4.0 fl j, Z E E E E Automatic containment isolation 025A402 Ice Condenser 2.7 2.5 F , Containment vent fans fl r11 [ [] [ EZ 026A1 05 Containment Spray 31 34 [ j Chemical additive tank level and concentration 026K1 01 Containment Spray 42 4.2 El El El El El El Li El [Z El ECCS Main and Reheat Steam 28 3 1 i j ] j SDS 039K5.08 Main and Reheat Steam 3.6 3.6 i Effect of steam removal on reactivity 059A3 04 Main Feedwater 25 26 Z Z E Turbine driven feed pump E
061 K2 01 Auxiliary/Emergency Feedwater 32 33 AFW system MOVs Z J Z E Z 062A2 09 AC Electrical Distribubon 27 30 Consequences of exceeding current limitations El LI El El EL 063A1 01 DC Electrical Distribubon 25 33 El LI El El [I El [1 El El El Battery capacity as it is affected by discharge rate Page2of3 2/11/2013 1:40PM
ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETYFUNCTION: IA Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRO 063K3.02 DC Electrical Distribution 3.5 3.7 E E Components using DC control power 064K3.03 Emergency Diesel Generator 3.6 3.9 1 Z ED/G (manual loads) 073K1 .01 Process Radiation Monitoring 3.6 3.9 1 i [j E D El D E E Those systems served by PRMs 076A4.02 SeMceWater 2.6 2.6 JJ [] [J E El LI H SWS valves 078G2.435 instrument Air 3.8 4.0 El El El El El El El El El Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects 103Ki01 Containment 3.6 3.9 ]ElrEElElElElElElflfl CCS Page3of3 2/11/2013 1:40PM
ES-401, REV 9 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR KI K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRO 002K5 07 Reactor Coolant 36 39 [ L El El LI LI LI El El Reactivity effects of RCS boron pressure and temperature 015A102 Nuclear Instrumentation 3.5 3.6 ] [ j [J j SUR 016A3 01 Non nuclear Instrumentation 29 29 [ El LI El LI LI L LI El LI Automatic selection of NNIS inputs to control systems 027K2O1 Containment Iodine Removal 3.1 3.4 J [j JJ [ [ Fans 033A2O1 Spent Fuel Pool Cooling 3.0 3.5 [ [] Li El El El i El LI LI Inadequate SDM 034K6 02 Fuel Handling Equipment 26 33 ] j [ Radiation monitoring systems 071 K3 05 WasteGas Disposal 32 32 [ [ ] ARM and PRM systems 072A4.O1 Radiation Monitoring 3.0 3.3 J ] [ [J J Alarm and interlock setpoint checks and adjustments O79K4Ji Station Air 2.9 3.2 E] [ [1 [1 [1 Cross-connect with AS 086K103 AreProtection 34 35 DDDDDDE1E1DE] AFWsystem Page 1 of 1 2/11/2013 1:40 PM
ES-401, REV 9 T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETYFUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRQ G2 1 15 Conduct of operations 27 34 -
[] ] j -
[ E El El LI Knowledge of administrative requirements for temporary management directives such as standing orders, night orders, Operations memos, etc.
G2 1 43 Conduct of operations 41 43 [j [ Li LI El El LI LI Ability to use procedures to determine theeffects on reactivity of plant changes G2 2 22 Equipment Control 40 47 [ ] [ [j ] Knowledge of limiting conditions for operations and safety limits.
G2 2 42 Equipment Control 39 46 JJ fl [ Ability to recognize system parameters that are entry level conditions for Technical Specifications 2 43 Equipment Control 30 33 [ J J J j j Knowledge of the process used to track inoperable alarms G2 3 1 1 Radiation Control 38 43 El El LI LZ LI El El D D Ability to control radiation releases G2 3 5 RadiationControl 29 29 [ [ [ [ [ [ Ability to use radiation monitoring systems G24.26 Emergency Procedures/Plans 3.1 3.6 [] [] Knowledge of facility protection requirements including
]
fire brigade and portable fire fighting equipment usage.
G2 43 Emergency Procedures/Plans LI LI LI LI LI LI LI LI LI LI Ability to identify post accident instrumentation G249 Emergency Procedures/Pns 38 42 LI LI LI LI LI LI LI LI LI LI Knowledge of low power / shutdown implications in accident (eg. LOCA or loss of RHR) mitigation strategies.
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ES-401, REV 9 SRO T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IA Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
Reactor Trip Stabilization Recovery 43 4.6 Proper actions to be taken if the automatic safety tunc
/1 LiL1DDDDDE1D tions have not taken place 011 EA2.08 Large Break LOCA / 3 El El El El El El El El El El Conditions necessary for recovery when accident reaches stable phase 025AA2.05 Loss of RHR System /4 3.1 3.5 El El El El El El El El El El Limitations on LPI flow and temperature rates of change 026AG2.4.45 Loss of Component Cooling Water / 8 4.1 4.3 ElElElElElElElElElEl Ability to prioritize and interpret the significance of each annunciator or alarm.
056AG2.2.40 Loss of Off-site Power /6 3,4 4,7 ElElElElE]ElElElElEl Ability to apply technical specifications for a system.
062AG2.1 .23 Loss of Nuclear Svc Water /4 ElElElElElElElElElEl Ability to perform specific system and integrated plant procedures during all modes of plant operation.
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ES-401 , REV 9 SRO Ti G2 PWR EXAMINATION OUTLINE FORM ES-401 -2 KA NAME/SAFETYFUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRO 003AG2.2.36 Dropped Control Rod I 1 3.1 4.2 J -
T r Ability to analyze the effect of maintenance activtties, such as degraded power sources, on the status of limiting conditions of operations 028AG2.4.50 Pressurizer Level Malfunction I 2 4.2 4.0 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.
068AA2 06 Control Room Evac / 8 4 1 43 Li RCS pressure WE1OEA2.2 Natural Circ, With Seam Void/4 3.4 3.9 Li El Li Li El El LI Adherence to appropriate procedures and operation LI LI El within the limitations in the facilitys license and amendments.
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ES-401, REV 9 SRO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRO 003G2.4.34 Reactor Coolant Pump 4.2 4.1 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects 010A2.02 Pressurizer Pressure Control 3.9 3.9 bf Spray valve failures 012A2.05 Reactor Protection 3.1 3.2 E EEL L1L11EEE Faulty or erratic operation of detectors and function generators 025G2.2.25 Ice Condenser 3.2 4.2 E EEEEILiEE Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
073A2.03 Process Radiation Monitoring 2.4 2.9 E EEELEEE Calibration drift Page 1 of 1 2/11/2013 1:40PM
ES-401, REV 9 SRO T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRO 01 4A2 04 Rod Position Indication 34 39 [ [j [ [] Misaligned rod 028A2 03 Hydrogen Recombiner and Purge 34 40 j ] [ [ J] [ The hydrogen air concentration in excess of limit flame Control propagation or detonation with resulting equipment damage in containment 055G2 1 31 Condenser Air Removal 4.6 4.3 Abil[ty to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup.
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ES-401, REV 9 SRO T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:
RO SRO G2.l .4 Conduct of operations 3.3 3.8 : Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license statur, 10CFR55 etc.
G2.2.17 Equipment Control 2.6 3.8 1 Knowledge ofthe process for managing maintenance activities during power operations.
G2 2 3 Equipment Control 38 39 [1 J [ J j j E v (multi unit license) Knowledge of the design procedur&
and operational differences between units.
G2 3 6 Radiation Control 20 38 [ Li [Z El LI El EJ Li LI Ability to aprove release permits G23 7 Radiation Control 35 36 J ] Ability to comply with radiation work permit requirements
] Jj during normal or abnormal conditions G2.4.22 Emergency Procedures/Plans 3.6 4.4 fl fl Knowledge of the bases for prioritizing safety functions El LI LI during abnormal/emergency operations.
G2.4.29 Emergency Procedures/Plans 3.1 4.4 Knowledge of the emergency plan.
El Page 1 of 1 2/11/2013 1:40 PM
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination:05/13/2013 Exam Level: RO l1 SRO Operating Test No: 2013-301 Administrative Topic (see Type Describe activity to be performed Note) Code*
2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, Conduct of Operations R, M etc. (4.1/4.3)
- Given plant data, perform an RCS Deboration Calculation.
2.1.25 Ability to interpret reference materials, such as
. graphs, curves, tables, etc. (3.9/4.2)
Conduct of Operations R, D Given plant data, calculate Maximum Reactor Vessel Vent Time.
2.2.14 Knowledge of the process for controlling equipment configuration or status. (3.9/4.3)
Equipment Control R, M Perform a System Operability Checklist when the 1A RHR becomes inoperable and determine the required protected train tag placement for configuration control.
Radiation Control Not examined 2.4.39 Knowledge of RO responsibilities in emergency plan implementation. (3.9/3.8)
Emergency Procedures/Plan R N Determine the required Operator allocation and make a report to the SM for an Emergency Plan turnover to the OSC Ops Advisor SRO.
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
Type Codes & Criteria:(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1; randomly selected)
A.1.a Given that Unit 1 is in MODE 1 at End of Life Conditions and that preparations are being made to place a fresh Mixed Bed CVCS Purification Ion Exchanger in service. The examinee will perform an RCS Deboration Calculation using 1-SO-62-9, CVCS PURIFICATION SYSTEM for two cases, one with a 75 gpm Letdown orifice in service and the other with the 120 gpm Letdown orifice in service so that proper boron loading on the Mixed Bed Ion Exchanger may be accomplished with minimal impact on plant reactivity.
A.1.b Given plant data during a plant emergency, the examinee will determine by calculating and interpreting graphs the time required to vent the Reactor Vessel while maintaining containment hydrogen concentration below 3% using EA-0-7 Calculating Maximum Reactor Vessel Vent Time.
A.2 Given an emergent condition the 1A RHR pump is declared to be INOPERABLE. The examinee will identify the equipment required to be administratively protected to maintain configuration control and determine the placement of protected train tags using 0-GO-16 SYSTEM OPERABILITY CHECKLISTS.
A.3 Not examined.
A.4 Given a major plant fire is in progress in the Auxiliary Building 690 elevation penetration room with the Site Emergency Plan in progress. The examinee will direct Assistant Unit Operators (AUO) to perform local actions as required by the location of the fire. The examinee will select the correct appendix based on location for the AUO task assignment and will determine which AUO will perform the specified task based on the level of AUO qualification using AOP-N.08 APPENDIX R FIRE SAFE SHUTDOWN. The examinee will prepare a report to the Shift Manager the choices made when the Shift Manager gives a turnover for personnel accountability during the Emergency Plan when the Operations Support Center is activated.
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Seguoyah Nuclear Station 1 & 2 Date of Examination:05/13/2013 Exam Level: RO El SRO El Operating Test No: 2013-301 Administrative Topic (see Type Describe activity to be performed Note) Code*
2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.
Conduct of Operations R, M (4.3/4.6)
Determine Actions Required Following a Reactivity Management Event When at Power.
2.1.25 Ability to interpret reference materials, such as
. graphs, curves, tables, etc. (3.9/4.2)
Conduct of Operations R, D Given plant data, calculate Maximum Reactor Vessel Vent Time.
2.2.18 Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.
Equipment Control R, N (2.6/3.9)
Determine Switchyard Access Requirements During an Outage Condition.
2.3.6 Ability to approve release permits. (2.0/3.8)
Radiation Control R, D Approve a Monitor Tank Release with 0-RE-90-122 INOPERABLE.
2.4.4 1 Knowledge of the emergency action level thresholds and classifications. (4.6)
Emergency Procedures/Plan R, M Classify The Event Using The EPIP-1 and Complete a State Notification Form.
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
Type Codes & Criteria:(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1; randomly selected)
A.1.a Given a sequence of events while acting as the Unit Supervisor when a CVCS Purification mixed bed ion exchanger was placed in service with insufficient boron loading, the examinee will determine the following:
- The severity of the event as a Minor Reactivity Management Event using NPG-SPP-1 0.4 Reactivity Management Program.
- That site operations management and duty plant manager are the required internal notifications using NPG-SPP-3.5 Regulatory Reporting Requirements.
This task is based on a Sequoyah internal operating event.
A.1.b Given plant data during a plant emergency, the examinee will determine by calculating and interpreting graphs the time required to vent the Reactor Vessel while maintaining containment hydrogen concentration below 3% using EA-0-7 Calculating Maximum Reactor Vessel Vent Time.
A.2 Given a situation while acting as the Work Control Center (WCC) Supervisor, the examinee will determine the following prior to allowing a work group access to the switchyard using OPDP-2 Switchyard Access and Switching Order Execution:
- Pre-job brief requirements.
- Switchyard access gate requirements.
- Vehicle speed limit.
- Vehicle escort requirement.
- Defense in Depth concept requirements during a refueling outage condition.
The requirements listed are necessary to maintain switchyard integrity during switchyard entry and are the responsibility if a Sequoyah Senior Reactor Operator. Additionally more stringent requirements are imposed during outage conditions to ensure core cooling capability is maintained.
This task is based on a Sequoyah internal operating event.
A.3 Given a situation while acting as the Unit Supervisor when a Waste Disposal System Monitor tank liquid release is planned with Radiation Monitor O-RE-90-122 INOPERABLE, the examinee will determine the following prior to authorizing the release using O-SO-77-1 WASTE DISPOSAL SYSTEM (LIQUID):
- Valves required to be Independently Verified.
- Location of jumper placement to allow for O-RCV-77-43 operation.
Additionally the examinee will determine the following additional Chemistry Department requirements prior to authorizing the release using O-SI-CEM-077-400.1 Liquid Waste Effluent Batch Release:
- Independent sample requirement
- Independent analysis requirement
- Independent release rate verification.
The requirements listed are necessary to demonstrate the appropriate administrative controls that are in place to preclude the possibility of an inadvertent release radioactive in excess of limits to the public.
A.4 Acting as the Site Emergency Director and given data for a plant emergency, the examinee will interpret the data from a major gaseous effluent release event within the Exclusion Area Boundary (EAB) and determine the correct Emergency Classification of Site Area Emergency, and subsequently complete a state notification form.
ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination: 05/13/2013 Exam Level: RO q SRO-l El SRO-U El Operating Test No: 2013-301 Control Room Systems (8 for RO); (7 for SRO-l); (2 or 3 for SRO-U, including 1 ESF)______________
System / JPM Title Type Code* Safety Function
- a. Perform a 0-Sl-OPS-085-01 1.0 Reactivity Control Systems Moveable Control Assemblies Test. K/A 001.A2.11 (4.4/4.7) M A S 1
- f. Start up of the A Hydrogen Recombiner K/A 028 A4.01 (4.0/4.0) D, L, S 5
- g. Transfer lA-A 69 KV SD Board from Alternate to Normal Supply. 064 A4.01 (4.0/4.3) D A S 6
- h. Shutdown Containment Purge. 029 A1.03 (3.0 / 3.3) N, L, 5 8 In-Plant Systems (3 for RO); (3 for SRO-l); (3 or 2 for SRO-U)
- i. Perform a Local EDG Start with a Failure of ERCW Valve to Auto D, A, E 6 Open. K/A 064 A4.01 (4.0/4.3)
- j. Perform a Local Alignment of U-2 TDAFW Level Control Valves. K/A M, R, E,L 4S 061A2.07 (3.4/3.5)
- k. Respond to Loss of Control Air System. K/A APE 065 Ml .04 (3.5/3.4) D, E, L 8
@ All RO and SRO-l control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
Type Codes Criteria for RO / SRO-l I SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 814 (E)mergency or abnormal in-plant 1/ 1 I 1 (EN)gineered safety feature I I 1 (control room system)
(L)ow-Power I Shutdown 1 / 1 / 1 (N)ew or (M)odified from bank including 1(A) 21 2/ 1 (P)revious 2 exams 3 I 2 (randomly selected)
(R)CA 1 / 1 I 1 (S)imulator
. The examinee will perform a control rod testing in MODE 1 using 0-SI-OPS-085-01 1.0 Reactivity Control Systems Moveable Control Assemblies starting with Control Bank C. During the test an uncontrolled rod movement occurs, the examinee will use the alternate path and transition to AOP C.01, Rod Control System Malfunctions to initiate a manual Reactor trip.
- b. The examinee will assume the shift with a LOCA is in progress and ECCS pumps are running with the suction path aligned from the RWST. After assuming the shift, an RWST Low Level condition occurs and the examinee will transition to ES-i .3, Transfer to Cold Leg Recirculation to align Cold Leg Recirculation.
While performing ES-i .3, RHR suction valve 74-21 fails to auto close, the examinee will use the alternate path and manually close RHR suction valve 74-21. The examinee will subsequently complete the alignment of Cold Leg Recirculation with the ECCS pumps suction path aligned to the containment sump.
- c. The examinee will assume the shift in MODE 3 following a spurious Safety Injection. The examinee will terminate safety injection by stopping one charging pump, isolating the CCPIT and will attempt to establish charging flow using E-0, REACTOR TRIP OR SAFETY INJECTION. While aligning Charging flow, Normal Charging Isolation valve 62-85 will not operate, the examinee will use the alternate path and manually align Charging by opening Alternate Charging Isolation valve 62-85.
- d. The examinee will assume the shift in MODE 3 and will be directed to respond to plant conditions.
The #2 Reactor Coolant Pump (RCP) will develop a high temperature condition on the motor stator.
The examinee will address the ARP and transition to AOP-R.04, Reactor Coolant Pump Malfunctions. Based on plant conditions with the plant in MODE 3, the examinee will use the alternate path to stop the #2 RCP and close the Loop 2 Pressurizer Spray valve.
- e. The examinee will assume the shift in MODE 3 with a large leak in the Condensate Storage Tank and a Loss of Offsite Power. The examinee will align Essential Raw Cooling Water (Service Water) to the motor driven Auxiliary Feed Pumps (AFW) using EA-3-10 ESTABLISHING MOTOR DRIVEN AFW FLOW section 4.10 before the AFW pumps trip due to excessive cavitation.
- f. The examinee will assume the shift following an accident. The examinee will determine the Containment Pressure correction factor and will place A-A Hydrogen Recombiner in service using EA-268-1, Placing Hydrogen Recombiner in Service.
- g. The examinee will assume the shift in MODE 1 with 1A 6.9 Ky Shutdown Board to the aligned to the alternate power source. The examinee will be directed to transfer lA-A 6.9kV Shutdown Board to Normal Feeder at 1 -M-1 using 0-SO-202-4, 6900V Shutdown Boards, Section 8.1.5. The transfer to the normal power supply will fail, and the 1A Emergency Diesel Generator (EDG) will fail to auto start resulting in a loss of power to the 1A 6.9 Ky Shutdown Board. The examinee will use the alternate path to manually start the 1A EDG using 0-SO-202-4, 6900V Shutdown Boards.
- h. The examinee will assume the shift in MODE 5 with Lower Containment Purge A Train in service.
The examinee will remove Lower Containment Purge A Train from service using 0-SO-30-3 CONTAINMENT PURGE SYSTEM OPERATION.
The examinee will perform a normal, local start of the lA Diesel Generator using O-SO-82-l Diesel Generator lA-A section 8.2. While performing the start of the 1A Diesel Generator, the examinee will determine FCV-67-66 Emerg Dsl Htxs Al & A2 Sup Vlv from Hdr A failed to automatically open. The examinee will use the alternate path and manually open l-FCV-67-68D Emerg Dsl Htxs Al & A2 Sup Vlv From Hdr B to establish cooling water flow to the 1A Diesel Generator.
- j. The examinee will assume the shift in MODE 3 during a Loss of Essential Raw Cooling Water (ERCW) condition. The examinee will perform local actions to align the backup air supply isolation valves to the U-2 TDAFW Level Control Valves.
- k. The examinee will assume the shift in MODE 3 during a Loss of Control and Service Air. The examinee will perform local actions to start and manually load the Station Air Compressors using EA 32-2, Establishing Control and Service Air.
ES-301 Control Roomlln-Pbnt Svstms ()iitlin Form ES-301-2 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination: 05/13/2013 Exam Level: RO SRO-l q SRO-U Operating Test No: 2013-301 Control Room Systems (8 for RO); (7 for SRO-l); (2 or 3 for SRO-U, including 1 ESF)______________
System / JPM Title Type Code* Safety Function
- a. Perform a 0-Sl-OPS-085-01 1.0 Reactivity Control Systems Moveable Control Assemblies Test. K/A 001.A2.11 (4.4/4.7) M A S 1
- f. Not examined.
- g. Transfer lA-A 6.9 KV SD Board from Alternate to Normal Supply. 064 A4.01 (4.0/4.3) D A S 6
- h. Shutdown Containment Purge. 029 A1.03 (3.0 / 3.3) N, L, S 8 In-Plant Systems (3 for RO); (3 for SRO-l); (3 or 2 for SRO-U)
- i. Perform a Local EDG Start with a Failure of ERCW Valve to Auto D, A, E 6 Open. K/A 064 A4.01 (4.0/4.3)
- j. Perform a Local Alignment of U-2 TDAFW Level Control Valves. K/A M, R, E,L 4S 061A2.07 (3.4/3.5)
- k. Respond to Loss of Control Air System. K/A APE 065 AA1 .04 (3.5/3.4) D, E, L 8
@ All RO and SRO-l control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
Type Codes Criteria for RO / SRO-l / SRO-U (A)lternate path 4-6 / 4-6 I 2-3 (C)ontrol room (D)irect from bank 9/814 (E)mergency or abnormal in-plant 1 I 1 / 1 (EN)gineered safety feature - I - / l (control room system)
(L)ow-Power I Shutdown 1 I 1 I 1 (N)ew or (M)odified from bank including 1(A) 2121 1 (P)revious 2 exams 3 / 3 I 2 (randomly selected)
(R)CA 1 / I I 1 (S)imulator
- a. The examinee will perform a control rod testing in MODE 1 using 0-Sl-OPS-085-01 1.0 Reactivity Control Systems Moveable Control Assemblies starting with Control Bank C. During the test an uncontrolled rod movement occurs, the examinee will use the alternate path and transition to AOP C.01, Rod Control System Malfunctions to initiate a manual Reactor trip.
- b. The examinee will assume the shift with a LOCA is in progress and ECCS pumps are running with the suction path aligned from the RWST. After assuming the shift, an RWST Low Level condition occurs and the examinee will transition to ES-i .3, Transfer to Cold Leg Recirculation to align Cold Leg Recirculation.
While performing ES-i .3, RHR suction valve 74-21 fails to auto close, the examinee will use the alternate path and manually close RHR suction valve 74-21. The examinee will subsequently complete the alignment of Cold Leg Recirculation with the ECCS pumps suction path aligned to the containment sump.
- c. The examinee will assume the shift in MODE 3 following a spurious Safety Injection. The examinee will terminate safety injection by stopping one charging pump, isolating the CCPIT and will attempt to establish charging flow using E-0, REACTOR TRIP OR SAFETY INJECTION. While aligning Charging flow, Normal Charging Isolation valve 62-85 will not operate, the examinee will use the alternate path and manually align Charging by opening Alternate Charging Isolation valve 62-85.
- d. The examinee will assume the shift in MODE 3 and will be directed to respond to plant conditions.
The #2 Reactor Coolant Pump (RCP) will develop a high temperature condition on the motor stator.
The examinee will address the ARP and transition to AOP-R.04, Reactor Coolant Pump Malfunctions. Based on plant conditions with the plant in MODE 3, the examinee will use the alternate path to stop the #2 RCP and close the Loop 2 Pressurizer Spray valve.
- e. The examinee will assume the shift in MODE 3 with a large leak in the Condensate Storage Tank and a Loss of Offsite Power. The examinee will align Essential Raw Cooling Water (Service Water) to the motor driven Auxiliary Feed Pumps (AFW) using EA-3-10 ESTABLISHING MOTOR DRIVEN AFW FLOW section 4.10 before the AFW pumps trip due to excessive cavitation.
- f. Not examined.
- g. The examinee will assume the shift in MODE 1 with 1A 6.9 Ky Shutdown Board to the aligned to the alternate power source. The examinee will be directed to transfer lA-A 6.9kV Shutdown Board to Normal Feeder at 1-M-1 using 0-SO-202-4, 6900V Shutdown Boards, Section 8.1.5. The transfer to the normal power supply will fail, and the iA Emergency Diesel Generator (EDG) will fail to auto start resulting in a loss of power to the 1A 6.9 Ky Shutdown Board. The examinee will use the alternate path to manually start the 1A EDG using 0-SO-202-4, 6900V Shutdown Boards.
- h. The examinee will assume the shift in MODE 5 with Lower Containment Purge A Train in service.
The examinee will remove Lower Containment Purge A Train from service using 0-SO-30-3 CONTAINMENT PURGE SYSTEM OPERATION.
i The examinee will perform a normal, local start of the 1A Diesel Generator using O-SO-82-l Diesel Generator lA-A section 8.2. While performing the start of the 1A Diesel Generator, the examinee will determine FCV-67-66 Emerg Dsl Htxs Al & A2 Sup Vlv from Hdr A failed to automatically open. The examinee will use the alternate path and manually open l-FCV-67-68D Emerg Dsl Htxs Al & A2 Sup Vlv From Hdr B to establish cooling water flow to the 1A Diesel Generator.
- j. The examinee will assume the shift in MODE 3 during a Loss of Essential Raw Cooling Water (ERCW) condition. The examinee will perform local actions to align the backup air supply isolation valves to the U-2 TDAFW Level Control Valves.
- k. The examinee will assume the shift in MODE 3 during a Loss of Control and Service Air. The examinee will perform local actions to start and manually load the Station Air Compressors using EA 32-2, Establishing Control and Service Air.
FS-301 Control Roomiln-Plant Svstms OutIin Form ES-301 -2 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination: 0511 3/2013 Exam Level: RO LI SRO-l LI SRO-U q Operating Test No: 201 3-301 Control Room Systems (8 for RO); (7 for SRO-l); (2 or 3 for SRO-U, including 1 ESF)______________
System I JPM Title Type Code* Safety Function
- a. Perform a 0-Sl-OPS-085-01 1.0 Reactivity Control Systems Moveable Control Assemblies Test. K/A 001.A2.11 (4.4/4.7) M A S 1
- d. Not Examined
- e. Not Examined
- f. Not Examined
- g. Not Examined
- h. Not Examined In-Plant Systems (3 for RO); (3 for SRO-l); (3 or 2 for SRO-U)
- i. Not Examined
- j. Perform a Local Alignment of U-2 TDAFW Level Control Valves. K/A M, R, E,L 4S 061A2.07 (3.4/3.5)
- k. Respond to Loss of Control Air System. K/A APE 065 Ml .04 (3.5/3.4) D, E, L 8
@ All RD and SRO-l control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
Type Codes Criteria for RO I SRO-l / SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irectfrom bank 91814 (E)mergency or abnormal in-plant 1 I 1 I 1 (EN)gineered safety feature - I - I 1 (control room system)
(L)ow-Power I Shutdown 1 I 1 I 1 (N)ew or (M)odified from bank including 1(A) 2 I 2 I 1 (P)revious 2 exams 3 I 3 I 2 (randomly selected)
(R)CA 1I1I1 (S)imulator
- a. The examinee will perform a control rod testing in MODE 1 using 0-Sl-OPS-085-01 1.0 Reactivity Control Systems Moveable Control Assemblies starting with Control Bank C. During the test an uncontrolled rod movement occurs, the examinee will use the alternate path and transition to AOP C.01, Rod Control System Malfunctions to initiate a manual Reactor trip.
- b. The examinee will assume the shift with a LOCA is in progress and ECCS pumps are running with the suction path aligned from the RWST. After assuming the shift, an RWST Low Level condition occurs and the examinee will transition to ES-i .3, Transfer to Cold Leg Recirculation to align Cold Leg Recirculation.
While performing ES-i .3, RHR suction valve 74-2i fails to auto close, the exam inee will use the alternate path and manually close RHR suction valve 74-2i. The examinee will subsequently complete the alignment of Cold Leg Recirculation with the ECCS pumps suction path aligned to the containment sump.
- c. The examinee will assume the shift in MODE 3 following a spurious Safety Injection. The examinee will terminate safety injection by stopping one charging pump, isolating the CCPIT and will attempt to establish charging flow using E-0, REACTOR TRIP OR SAFETY INJECTION. While aligning Charging flow, Normal Charging Isolation valve 62-85 will not operate, the examinee will use the alternate path and manually align Charging by opening Alternate Charging Isolation valve 62-85.
- d. Not Examined.
- e. Not Examined.
- f. Not Examined.
- g. Not Examined.
- h. Not Examined.
L Not Examined.
- j. The examinee will assume the shift in MODE 3 during a Loss of Essential Raw Cooling Water (ERCW) condition. The examinee will perform local actions to align the backup air supply isolation valves to the U-2 TDAFW Level Control Valves.
- k. The examinee will assume the shift in MODE 3 during a Loss of Control and Service Air. The examinee will perform local actions to start and manually load the Station Air Compressors using EA 32-2, Establishing Control and Service Air.