ML081010150
| ML081010150 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/31/2008 |
| From: | NRC Region 1 |
| To: | |
| Hansell S | |
| Shared Package | |
| ML073040288 | List: |
| References | |
| ES-301, NUREG-1021 | |
| Download: ML081010150 (27) | |
Text
ES-301 Administrative Topics Outline Form ES-301-1 Facility: NMP2-NRC Date of Examination: March 08 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*
Conduct of Operations M,P,R Perform administrative actions for single loop operation.
Review a completed portion of control room daily logs.
Conduct of Operations N,R Identify deficiencies and TS requirements.
Review a tagging request for the A RHR pump. Address Equipment Control N,R Tech Specs.
Radiation Control D,R Generate and approve an Emergency Exposure Authorization Emergency Plan N,R Determine EAL and complete initial notification paperwork.
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)
A1A The candidate will perform administrative actions for single loop operation. This is a modified JPM used previously on the 2002 NRC exam.
A1B The candidate will review a completed portion of control room daily logs. Identify deficiencies and TS requirements. This is a new JPM A2 The candidate will review a tagging request for the A RHR pump (or other piece of TS equipment). The request will contain several errors and TS will be addressed. This is a new JPM.
A3 The candidate will generate and approve an Emergency Exposure Authorization. This is a bank JPM A4 The candidate will perform an EAL determination given plant conditions and complete initial notification paperwork. This is a time critical modified JPM.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: NMP2 Date of Examination: MARCH 2008 Exam Level (circle one): SRO Operating Test No.: 1 Control Room Systems (8 for RO; 2 or 3 for SRO-U, including 1 ESF) SRO-U in BOLD #s S-4,6,7/P-1,2 Type Code* Safety System / JPM Title Function S-1 Reactor/Turbine Pressure Regulating / Place Main Turbine Shell N,S,L,A 3 Warming in service IAW N2-OP-21-Rev.8 K/A 241000 A4.18 2.9/2.8 S-2 Primary Containment / Initiate Division I H2/O2 monitoring to N,S,E 9 monitor sample path 4, post LOCA restart required.
K/A 223001 A4.04 3.5/3.6, A4.05 3.6/3.6 S-3 Reactor Feedwater / Transfer Feedwater Level Control to FWS- N,S,L 2 LV55A at approximately 2% power IAW N2-OP-3.
K/A 259001 A4.05 4.0/3.9 S-4 RCIC / Place RCIC in service due to a level transient, RCIC D,S,E,A 4 fails to isolate on isolation signal.
K/A 217000 A2.01 3.8/3.7 A4.01 3.7/3.7, A3.06 3.5/3.5 S-5 Standby Gas Treatment / Align SBGTS Train A to reduce Drywell D,S 5 pressure IAW N2-OP-61A K/A 295024 EA1.20 3.5/3.6 (02-OPS-SJE-261-2-02)
S-6 Emergency Diesel Generators / Manual start and load of the M,S,E,A 6 Division III EDG from Panel P-852 IAW N2-OP-100B. EDG overspeeds.
K/A 264000 A4.04 3.7/3.7 (02-OPS-SJE-264-2-67)
S-7 Resetting a Reactor Scram N,L,S 7 K/A 212000 A4.14 3.8/3.8 S-8 N/A NUREG-1021, Revision 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems (3 for RO; 3 or 2 for SRO-U)
P-1 Instrument Air / Startup of Air Dryer 2IAS -DRY1B IAW N2-OP- N,R 8 19 K/A 300000 A2.01 2.9/28 P-2 Standby Liquid Control / Boron Injection with Hydro Pump D,R,E 1 IAW EOP-16, Att.15 K/A 295037 EA1.10 3.7/3.9 (02-OPS-PJE-211-2-01)
D.C. Distribution / Place Battery Charger 2BYS-CHGR1A1 is P-3 M,A 6 placed in service.
K/A 263000 A1.01 2.5/2.8 (02 -OPS-PJE-263-2-03 Rev1)
@ All RO and SRO-I 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-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (L)ow-Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)
(R)CA 1/1/1 (S)imulator NUREG-1021, Revision 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 2007 NRC Examination Summary Description of JPMs S-1 This is a new alternate path JPM in the Reactor Pressure Control safety function area. The candidate will place Main Turbine Chest Warming in service IAW N2-OP-21-Rev.8, Section 3.0 and the Main Turbine will roll off the turning gear requiring additional operator actions.
S-2 This is a new JPM in the Radioactive Release safety function area. The candidate will initiate Division I H2/O2 monitoring to monitor sample path 4 (suppression chamber), post LOCA, IAW N2-OP-82, Section H.1.0.
S-3 This is a new JPM in the Reactor Water Inventory Control safety function area. The candidate will Transfer Feedwater Level Control to FWS-LV55A at approximately 2% power IAW N2-OP-3, Section E.3.16 S-4 This is a bank alternate path JPM in the Heat Removal From Reactor Core Safety Function area.
The candidate will be required to place RCIC in service due to a level transient. A valid isolation signal will occur and RCIC will fail to isolate. Operator action is required to isolate and trip RCIC.
S-5 This is a bank JPM in the Containment Integrity safety function area. The candidate will align SBGTS Train A to reduce Drywell pressure IAW N2-OP-61A, Section H.
S-6 This is a modified bank alternate path JPM in the Electrical safety function area. The candidate will perform a Manual start and load of the Division III EDG from Panel P-852 IAW N2-OP-100B, Section F.2.0. EDG will overspeed but will not trip requiring operator actions to trip the EDG.
S-7 This is a new JPM in the Instrumentation safety function area. The candidate will reset a reactor scram and RRCS/ARI following a scram which had occurred due to a loss of feedwater IAW N2-OP-SOP-101C and N2-OP-36B, Section H.3.0.
P-1 This is a new JPM in the Plant Systems safety function area. The candidate will startup Air Dryer 2IAS -DRY1B, IAW N2-OP-19, Section E.2.0.
P-2 This is a bank JPM in the Reactivity Control safety function area. The candidate will establish a flowpath from SLS tank to the Reactor Vessel using a hydro pump and hoses staged, complete with an air supply to the pump and then commence Boron injection.
P-3 This is an alternate path bank JPM in the Electrical safety function area. The candidate will place Battery Charger 2BYS-CHGR1A1 in service IAW N2-OP-73A Section E.4.0. Alternate actions will be required due to high charger current.
NUREG-1021, Revision 9
ES-401 Written Examination Outline Form ES-401-1 Facility: NMP2 NRC Date of Exam: March 2008 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4 *
- 1. 1 4 3 3 3 3 4 20 4 3 7 Emergency
& 2 2 1 1 1 1 1 7 1 2 3 Plant Tier Evolutions 6 4 4 4 4 5 27 5 5 10 Totals 1 3 2 2 2 3 3 2 2 2 3 2 26 2 3 5 2.
Plant 2 1 1 1 1 1 1 1 1 1 1 2 12 0 1 2 3 Systems Tier 4 3 3 3 4 4 3 3 3 4 4 38 3 5 8 Totals 1 2 3 4 1 2 3 4
- 3. Generic Knowledge & Abilities 10 7 Categories 2 3 3 2 2 2 1 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.
- 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 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 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 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 D.1.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 (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 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 10CFR55.43 ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G K/A Topic(s) Imp. Q#
AA2.06 - Ability to determine and/or 295001 Partial or Complete Loss of interpret the following as they apply to Forced Core Flow Circulation / 1 & X PARTIAL OR COMPLETE LOSS OF 3.3 76 4 FORCED CORE FLOW CIRCULATION :
Nuclear boiler instrumentation AA2.01 - Ability to determine and/or 295006 SCRAM / 1 X interpret the following as they apply to 4.6 77 SCRAM : Reactor power AA2.01 - Ability to determine and/or interpret the following as they apply to 295019 Partial or Total Loss of X PARTIAL OR COMPLETE LOSS OF 3.6 78 Inst. Air / 8 INSTRUMENT AIR : Instrument air system pressure 2.4.9 - Emergency Procedures / Plan:
Knowledge of low power/shutdown 295021 Loss of Shutdown Cooling X implications in accident (e.g., loss of 4.2 79
/4 coolant accident or loss of residual heat removal) mitigation strategies.
2.1.20 - Conduct of Operations: Ability to 295025 High Reactor Pressure / 3 X 4.6 80 interpret and execute procedure steps.
EA2.01 - Ability to determine and/or interpret the following as they apply to 295026 Suppression Pool High X SUPPRESSION POOL HIGH WATER 4.2 81 Water Temp. / 5 TEMPERATURE: Suppression pool water temperature 2.4.4 - Emergency Procedures / Plan:
Ability to recognize abnormal indications 700000 Generator Voltage and X for system operating parameters that are 4.7 82 Electrical Grid Disturbances / 6 entry-level conditions for emergency and abnormal operating procedures.
AK1.01 - Knowledge of the operational 295001 Partial or Complete Loss of implications of the following concepts as Forced Core Flow Circulation / 1 & X they apply to PARTIAL OR COMPLETE 3.5 39 4 LOSS OF FORCED CORE FLOW CIRCULATION : Natural circulation 2.4.11 - Emergency Procedures / Plan:
295003 Partial or Complete Loss of X Knowledge of abnormal condition 4.0 40 AC / 6 procedures.
AK3.01 - Knowledge of the reasons for the 295004 Partial or Total Loss of DC following responses as they apply to X 2.6 41 Pwr / 6 PARTIAL OR COMPLETE LOSS OF D.C.
POWER : Load shedding: Plant-Specific AA2.06 - Ability to determine and/or 295005 Main Turbine Generator interpret the following as they apply to X 2.6 42 Trip / 3 MAIN TURBINE GENERATOR TRIP :
Feedwater temperature AK1.01 - Knowledge of the operational implications of the following concepts as 295006 SCRAM / 1 X 3.7 43 they apply to SCRAM : Decay heat generation and removal AK2.01 - Knowledge of the interrelations 295016 Control Room between CONTROL ROOM X 4.4 44 Abandonment / 7 ABANDONMENT and the following:
Remote shutdown panel: Plant-Specific AA2.01 - Ability to determine and/or interpret the following as they apply to 295018 Partial or Total Loss of X PARTIAL OR COMPLETE LOSS OF 3.3 45 CCW / 8 COMPONENT COOLING WATER :
Component temperatures ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G K/A Topic(s) Imp. Q#
AK3.02 - Knowledge of the reasons for the following responses as they apply to 295019 Partial or Total Loss of X PARTIAL OR COMPLETE LOSS OF 3.5 46 Inst. Air / 8 INSTRUMENT AIR : Standby air compressor operation AA2.01 - Ability to determine and/or 295021 Loss of Shutdown Cooling interpret the following as they apply to X 3.5 47
/4 LOSS OF SHUTDOWN COOLING :
Reactor water heatup/cooldown rate 2.1.28 - Conduct of Operations:
295023 Refueling Acc / 8 X Knowledge of the purpose and function of 4.1 48 major system components and controls.
2.4.50 - Emergency Procedures / Plan:
Ability to verify system alarm setpoints and 295024 High Drywell Pressure / 5 X 4.2 49 operate controls identified in the alarm response manual.
EA1.03 - Ability to operate and/or monitor the following as they apply to HIGH 295025 High Reactor Pressure / 3 X 4.4 50 REACTOR PRESSURE: Safety/relief valves: Plant-Specific EA1.02 - Ability to operate and/or monitor the following as they apply to 295026 Suppression Pool High X SUPPRESSION POOL HIGH WATER 3.6 51 Water Temp. / 5 TEMPERATURE: Suppression pool spray:
Plant-Specific EK2.03 - Knowledge of the interrelations 295028 High Drywell Temperature between HIGH DRYWELL X 3.8 52
/5 TEMPERATURE and the following:
Reactor water level indication EK3.04 - Knowledge of the reasons for the 295030 Low Suppression Pool following responses as they apply to LOW X 3.5 53 Water Level / 5 SUPPRESSION POOL WATER LEVEL:
HPCS operation: Plant-Specific 2.2.44 - Equipment Control: Ability to interpret control room indications to verify 295031 Reactor Low Water Level / the status and operation of a system, and X 4.2 54 2 understand how operator actions and directives effect plant and system conditions.
EK1.07 - Knowledge of the operational implications of the following concepts as 295037 SCRAM Condition Present they apply to SCRAM CONDITION and Power Above APRM X 3.4 55 PRESENT AND REACTOR POWER Downscale or Unknown / 1 ABOVE APRM DOWNSCALE OR UNKNOWN: Shutdown margin EK2.05 - Knowledge of the interrelations 295038 High Off-site Release Rate between HIGH OFF-SITE RELEASE X 3.7 56
/9 RATE and the following: Site emergency plan AK1.02 - Knowledge of the operation applications of the following concepts as 600000 Plant Fire On-site / 8 X 2.9 57 they apply to Plant Fire On Site: Fire Fighting AA1.03 - Ability to operate and/or monitor the following as they apply to 700000 Generator Voltage and X GENERATOR VOLTAGE AND ELECTRIC 3.8 58 Electric Grid Disturbances GRID DISTURBANCES: Voltage regulator Controls K/A Category Totals: 4 3 3 3 3/4 4/3 Group Point Total: 20/7 ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G K/A Topic(s) Imp. Q#
2.4.31 - Emergency Procedures / Plan:
295002 Loss of Main Condenser X Knowledge of annunciator alarms, 4.1 83 Vac / 3 indications, or response procedures.
AA2.02 - Ability to determine and/or interpret the following as they apply to LOW 295009 Low Reactor Water Level / 2 X 3.7 84 REACTOR WATER LEVEL : Steam flow/feed flow mismatch 500000 High CTMT Hydrogen Conc. 2.4.6 Emergency Procedures / Plan:
X 4.7 85
/5 Knowledge of EOP mitigation strategies.
AA2.05 - Ability to determine and/or 295008 High Reactor Water Level /
X interpret the following as they apply to HIGH 2.9 59 2
REACTOR WATER LEVEL : Swell 2.4.2 - Emergency Procedures / Plan:
Knowledge of system set points, interlocks 295015 Incomplete SCRAM / 1 X 4.5 60 and automatic actions associated with EOP entry conditions.
AK1.01 - Knowledge of the operational implications of the following concepts as 295020 Inadvertent Cont. Isolation /
X they apply to INADVERTENT 3.7 61 5&7 CONTAINMENT ISOLATION : Loss of normal heat sink EK2.03 - Knowledge of the interrelations 295029 High Suppression Pool between HIGH SUPPRESSION POOL X 3.3 62 Water Level / 5 WATER LEVEL and the following: HPCS:
Plant-Specific EK1.02 - Knowledge of the operational 295033 High Secondary implications of the following concepts as Containment Area Radiation Levels / X they apply to HIGH SECONDARY 3.9 63 9 CONTAINMENT AREA RADIATION LEVELS : Personnel protection EA1.02 - Ability to operate and/or monitor 295035 Secondary Containment the following as they apply to SECONDARY X 3.8 64 High Differential Pressure / 5 CONTAINMENT HIGH DIFFERENTIAL PRESSURE: SBGT/FRVS EK3.04 - Knowledge of the reasons for the following responses as they apply to 295036 Secondary Containment X SECONDARY CONTAINMENT HIGH 3.1 65 High Sump/Area Water Level / 5 SUMP/AREA WATER LEVEL : Pumping secondary containment sumps 1/ 1/
K/A Category Totals: 2 1 1 1 Group Point Total: 7/3 1 2 ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A A System # / Name G Imp. Q#
1 2 3 4 5 6 1 2 3 4 2.2.40 - Equipment Control: Ability to 215003 IRM X apply Technical Specifications for a 4.7 86 system.
A2.07 - Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ;
and (b) based on those predictions, use 209001 LPCS X 3.6 87 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Core Spray Line Break 2.2.25 - Equipment Control:
Knowledge of bases in technical 211000 SLC X 4.2 88 specifications for limiting conditions for operations and safety limits.
A2.12 - Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ; and (b) 217000 RCIC X based on those predictions, use 3.0 89 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Valve openings 2.2.12 - Equipment Control:
239002 SRVs X 4.1 90 Knowledge of surveillance procedures.
A4.01 - Ability to manually operate 203000 RHR/LPCI: Injection X and/or monitor in the control room: 4.3 1 Mode Pumps A3.03 - Ability to monitor automatic operations of the SHUTDOWN 205000 Shutdown Cooling X COOLING SYSTEM (RHR 3.5 2 SHUTDOWN COOLING MODE) including: Lights and alarms K5.02 - Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN 205000 Shutdown Cooling X 2.8 3 COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) :
Valve operation K6.05 - Knowledge of the effect that a loss or malfunction of the following will 209001 LPCS X have on the LOW PRESSURE CORE 2.8 4 SPRAY SYSTEM : ECCS room cooler(s)
K5.04 - Knowledge of the operational implications of the following concepts 209002 HPCS X as they apply to HIGH PRESSURE 3.8 5 CORE SPRAY SYSTEM (HPCS):
Adequate core cooling: BWR-5,6 K2.02 - Knowledge of electrical power 211000 SLC X supplies to the following: Explosive 3.1 6 valves K4.06 - Knowledge of REACTOR PROTECTION SYSTEM design 212000 RPS X feature(s) and/or interlocks which 3.0 7 provide for the following: Select rod insertion: Plant-Specific ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A A System # / Name G Imp. Q#
1 2 3 4 5 6 1 2 3 4 K1.01- Knowledge of the physical connections and/or cause- effect 215003 IRM X relationships between INTERMEDIATE 3.9 8 RANGE MONITOR (IRM) SYSTEM and the following: RPS K5.01 - Knowledge of the operational implications of the following concepts 215003 IRM X as they apply to INTERMEDIATE 2.6 9 RANGE MONITOR (IRM) SYSTEM :
Detector operation A1.06 - Ability to predict and/or monitor changes in parameters associated with 215004 Source Range Monitor X operating the SOURCE RANGE 3.1 10 MONITOR (SRM) SYSTEM controls including: Lights and alarms K2.02 - Knowledge of electrical power 215005 APRM / LPRM X supplies to the following: APRM 2.6 11 channels K6.04 - Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE 217000 RCIC X 3.5 12 ISOLATION COOLING SYSTEM (RCIC): Condensate storage and transfer system A3.01 - Ability to monitor automatic operations of the AUTOMATIC 218000 ADS X 4.2 13 DEPRESSURIZATION SYSTEM including: ADS valve operation A2.09 - Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY 223002 PCIS/Nuclear Steam X SHUT-OFF; and (b) based on those 3.6 14 Supply Shutoff predictions, use procedures to correct, control, or mitigate the consequences of those abn cond or ops. System Initiation K1.04 - Knowledge of the physical connections and/or cause- effect 239002 SRVs X relationships between 3.6 15 RELIEF/SAFETY VALVES and the following: Main steam 2.4.18 - Emergency Procedures / Plan:
239002 SRVs X Knowledge of the specific bases for 3.3 16 EOPs.
A1.04 - Ability to predict and/or monitor changes in parameters associated with 259002 Reactor Water Level operating the REACTOR WATER X 3.6 17 Control LEVEL CONTROL SYSTEM controls including: Reactor water level control controller indications 2.2.42 - Equipment Control: Ability to recognize system parameters that are 261000 SGTS X 3.9 18 entry-level conditions for Technical Specifications.
A4.02 - Ability to manually operate 262001 AC Electrical and/or monitor in the control room:
X 3.4 19 Distribution Synchroscope, including understanding of running and incoming voltages ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A A System # / Name G Imp. Q#
1 2 3 4 5 6 1 2 3 4 K3.07 - Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER 262002 UPS (AC/DC) X 2.6 20 SUPPLY (A.C./D.C.) will have on following: Movement of control rods:
Plant-Specific K6.02 - Knowledge of the effect that a loss or malfunction of the following will 262002 UPS (AC/DC) X have on the UNINTERRUPTABLE 2.8 21 POWER SUPPLY (A.C./D.C.) : D.C.
electrical power K4.02 - Knowledge of D.C.
ELECTRICAL DISTRIBUTION design 263000 DC Electrical feature(s) and/or interlocks which X 3.1 22 Distribution provide for the following: Breaker interlocks, permissives, bypasses and cross ties: Plant-Specific K3.01 - Knowledge of the effect that a loss or malfunction of the 264000 EDGs X EMERGENCY GENERATORS 4.2 23 (DIESEL/JET) will have on following:
Emergency core cooling systems A4.04 - Ability to manually operate and/or monitor in the control room:
264000 EDGs X 3.7 24 Manual start, loading, and stopping of emergency generator: Plant-Specific A2.01 - Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those 300000 Instrument Air X predictions, use procedures to correct, 2.9 25 control, or mitigate the consequences of those abnormal operation: Air dryer and filter malfunctions K1.02 - Knowledge of the physical 400000 Component Cooling connections and / or cause-effect X 3.2 26 Water relationships between CCWS and the following: Loads cooled by CCWS 2 2 K/A Category Totals: 3 2 2 2 3 3 2 / 2 3 / Group Point Total: 26/5 2 3 ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 K K K K K K A A A A System # / Name G Imp. Q#
1 2 3 4 5 6 1 2 3 4 A2.03 - Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM ;
and (b) based on those predictions, 214000 RPIS X 3.9 91 use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Overtravel/in-out 2.4.47 - Emergency Procedures /
Plan: Ability to diagnose and recognize 245000 Main Turbine Gen. /
X trends in an accurate and timely 4.2 92 Aux.
manner utilizing the appropriate control room reference material.
2.4.41 - Conduct of operations: Ability 202001 Recirculation X to apply technical specifications for a 4.0 93 system A4.06 - Ability to manually operate 201001 CRD Hydraulic X and/or monitor in the control room: 2.8 27 SDV isolation valve test switch K6.04 - Knowledge of the effect that a loss or malfunction of the following will 215002 RBM X have on the ROD BLOCK MONITOR 2.8 28 SYSTEM : APRM reference channel:
BWR-3,4,5 K1.16 - Knowledge of the physical connections and/or cause- effect 204000 RWCU X relationships between REACTOR 2.8 29 WATER CLEANUP SYSTEM and the following: CRD system: Plant-Specific A3.01 - Ability to monitor automatic operations of the ROD POSITION 214000 RPIS X 3.4 30 INFORMATION SYSTEM including:
Full core display K2.01 - Knowledge of electrical power 256000 Reactor Condensate X supplies to the following: System 2.7 31 pumps A2.04 - Ability to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION ; and (b) based 216000 Nuclear Boiler Inst. X on those predictions, use procedures 2.9 32 to correct, control, or mitigate the consequences of those abnormal conditions or operations: Detector diaphragm failure or leakage K5.02 - Knowledge of the operational implications of the following concepts 226001 RHR/LPCI: CTMT X as they apply to RHR/LPCI: 2.6 33 Spray Mode CONTAINMENT SPRAY SYSTEM MODE : Water hammer 2.1.23 - Conduct of Operations: Ability 234000 Fuel Handling to perform specific system and X 4.3 34 Equipment integrated plant procedures during all modes of plant operation.
A1.01 - Ability to predict and/or monitor changes in parameters associated with 239001 Main and Reheat X operating the MAIN AND REHEAT 3.6 35 Steam STEAM SYSTEM controls including:
Main steam pressure ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 K K K K K K A A A A System # / Name G Imp. Q#
1 2 3 4 5 6 1 2 3 4 K4.11 - Knowledge of REACTOR FEEDWATER SYSTEM design 259001 Reactor Feedwater X feature(s) and/or interlocks which 3.5 36 provide for the following: Recirculation runbacks 2.1.32 - Conduct of Operations: Ability 268000 Radwaste X to explain and apply system limits and 3.4 37 precautions.
K3.01 - Knowledge of the effect that a loss or malfunction of the OFFGAS 271000 Off-gas X 3.5 38 SYSTEM will have on following:
Condenser vacuum 1 2 K/A Category Totals: 1 1 1 1 1 1 1 / 1 1 / Group Point Total: 12/3 1 2 ES-401 Generic Knowledge and Abilities Outline (Tier3) Form ES-401-3 Facility: 9 Mile Point Unit II Outline 1 Date: March 2008 RO SRO-Only Category K/A # Topic IR Q# IR Q#
Ability to interpret reference materials, such as 2.1.25 4.2 94 graphs, curves, tables, etc.
Knowledge of fuel handling responsibilities for 2.1.35 3.9 95 SROs
- 1. Knowledge of individual licensed operator Conduct responsibilities related to shift staffing, such as of Operations 2.1.4 medical requirements, "no-solo" operation, 3.3 66 maintenance of active license status, 10CFR55, etc.
2.1.41 Knowledge of the refueling process. 2.8 67 Subtotal 2 2 Knowledge of the process used to track 2.2.43 3.3 96 inoperable alarms.
Ability to track Technical Specification limiting 2.2.23 4.6 97 conditions for operations.
- 2. Knowledge of less than or equal to one hour Equipment 2.2.39 Technical Specification action statements for 3.9 68 Control systems.
2.2.12 Knowledge of surveillance procedures. 3.7 69 Ability to manipulate the console controls as 2.2.2 required to operate the facility between 4.6 70 shutdown and designated power levels.
Subtotal 3 2
- 3. 2.3.11 Ability to control radiation releases. 4.3 98 Radiation Control Knowledge of radiation exposure limits under 2.3.4 3.7 71 normal or emergency conditions Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, 2.3.13 3.4 72 containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable 2.3.15 2.9 73 survey instruments, personnel monitoring equipment, etc.
ES-401 Generic Knowledge and Abilities Outline (Tier3) Form ES-401-3 Subtotal 3 1 Knowledge of the parameters and logic used to assess the status of safety functions, such as 2.4.21 reactivity control, core cooling and heat removal, 4.6 99 reactor coolant system integrity, containment conditions, radioactivity release control, etc.
Knowledge of events related to system operation/status that must be reported to internal 2.4.30 organizations or external agencies, such as the 4.1 100
- 4. State, the NRC, or the transmission system Emergency operator.
Procedures /
Plan 2.4.14 Knowledge of general guidelines for EOP usage. 3.8 74 Knowledge of local auxiliary operator tasks 2.4.35 during emergency and the resultant operational 3.8 75 effects.
Subtotal 2 2 Tier 3 Point Total 10 7 ES-401 Record of Rejected K/As Form ES-401-4 Randomly Selected Tier / Group Reason for Rejection K/A
(#52) Mark III containment does not apply to NMP2.
1/1 295027 / EK2.02 Randomly selected 295028 EK2.03
(#36) MSIV Leakage system does not apply at NMP2.
2/2 239003 / K4.01 Randomly selected 259001 K4.11
(#8) Topic does not apply at NMP2. Randomly selected 2/1 215003 / K1.03 K1.07. Reselected K1.01, Could not write a discriminating RO level question for the initial topic reselection.
(#13) Overlap with other portions of the exam. Randomly 2/1 218000 / A3.05 reselected A3.01
(#14) Overlap with scenario exams. Kept same System 2/1 223002 / A2.07 223002, randomly reselected A2.09.
(#18) Original selection not related to aspects of system 2/1 261000 / 2.1.25 operation. Kept same System 261000, randomly reselected statement 2.2.42.
(#28) Turbine First Stage Pressure input to RSCS has been 2/2 201004 / K6.04 defeated at NMP2. RSCS has low operational impact.
Randomly reselected System 215002.
(#31) Double Jeopardy with Tier 2 Group 1 215005 K2.02.
2/2 215002 / K2.03 Both topics are APRM Power Supply. Randomly reselected 256000 K2.01
(#35) MSL Radiation Monitors do not initiate Group 1 2/2 239001 / A1.05 Isolations at NMP2. Randomly reselected statement A1.01
(#37) Could not write a discriminating RO Level question for 2/2 268000 / 2.1.30 local operation of Radwaste controls. Randomly replaced with statement 2.1.32
(#58) Could not write an operationally oriented RO Level 1/1 700000 / AA1.05 question. Randomly reselected statement AA1.03
(#62) Topic does not apply at NMP-2. Randomly reselected 1/2 295029 / EK2.09 EK2.03
(#69) Could not write a discriminating RO Level question.
3 2.2.14 Randomly reselected statement 2.2.12
(#79) Topic not used in conjunction with EOPs. Randomly 1/1 295021 / 2.4.8 reselected 2.4.9.
(#82) Topic not directly addressed in EOPs. Randomly 1/1 700000 / 2.4.6 reselected 2.4.4.
(#85) Could not write a discriminating SRO Level question.
1/2 500000 / 2.2.38 Randomly reselected 2.4.6.
(#86) Double Jeopardy with (#79), Shutdown Cooling Technical Specifications. Other Shutdown Cooling topics are 2/1 205000 / 2.2.40 covered on RO #2, #3, and #47. Randomly reselected System 215003.
(#90) Could not write a discriminating SRO Level question 2/1 300000 / 2.2.12 pertaining to Instrument Air / Surveillance Procedures.
Randomly reselected System 239002.
(#100) Topic covered extensively in scenarios. Randomly 3 2.4.45 reselected statement 2.4.30 ES-401 Record of Rejected K/As Form ES-401-4 Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 2 Scenario No.: NRC-01 Op-Test No.: March 2008 Examiners: _____________________ Operators: _________________________
Initial Conditions: Simulator IC-241 Reactor Power 90%
Turnover:
- 1. Swap Service Water Pumps from the 2SWP*P1B to the 2SWP*P1F for normal equipment rotation. Pre-start checks have been completed and an AO is standing by at the F pump Event Malf. No. Event Event No. Type* Description 1 N/A N (BOP) Swap operating Service Water Pumps N (SRO) N2-OP-11 Service Water 2 CW02B C (BOP) RBCLCW Pump trips. Standby Pump fails to auto-start and CW16C C (SRO) must be started manually.
N2-SOP-13 Loss or Degraded CCP System 3 PC10B C (BOP) ADS/SRV fails opens. Valve closes when fuses are pulled.
Override R (RO) Drywell Vacuum Breaker fails open.
s R/C Power decrease to 85%.
(SRO) N2-SOP-34 Stuck Open SRV TS (SRO) 4 FW03A C (ALL) Feedwater Pump Trip RR30 TS (SRO) Partial Recirc Runback (B loop)
RR31 N2-SOP-6, Feedwater Failures N2-SOP-29, Sudden Reduction in Core Flow 5 RD05- C(RO) One Control rod drifts out requiring a power decrease.
18-31 C(SRO) N2-SOP-08 Unplanned Power Changes 6 RD05- C (RO) Another control rod drifts out requiring a reactor scram 42-39 C (SRO) 7 MS04 M Steam Leak in Drywell. (EOP-RPV, EOP-PC) 8 RH01B I (BOP) DIV1 EDG, LPCS and RHR A fail to initiate and RHR B RH14A I (RO) trips when Drywell pressure exceeds 1.68 psig; Both LPCS and RHR A can be started manually.
9 RH09A C (SRO) DW Spray Valve MOV15A fails to open requiring Service C (BOP) Water in loop B for DW spray.
10 N/A PSP exceeded, RPV blowdown required (CT)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 1 March 2008
Facility: Nine Mile Point 2 Scenario No.: NRC-01 Op-Test No.: March 2008 TARGET QUANTITATIVE ATTRIBUTES ACTUAL (PER SCENARIO; SEE SECTION D.5.d) ATTRIBUTES
- 1. Total malfunctions (5-8) 7 Events 2,3,4,5,6,8,9
- 2. Malfunctions after EOP entry (1-2) 2 Events 8, 9
- 3. Abnormal events (2-4) 4 Event 2 SOP-13 Event 3 SOP-34, Event 4 SOP-6,29 Event 5 SOP-8
- 4. Major transients (1-2) 1 Event 7
- 5. EOPs entered/requiring substantive 2 actions (1-2)
- 6. EOP contingencies requiring substantive 1 actions (0-2)
Event 9 EOP-C2
- 7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:
CT-1.0 Initiate DW spray to control containment pressure CT-2.0 Initiate RPV Blowdown when PSP is exceeded and DW spray established NRC Scenario 1 March 2008
Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 2 Scenario No.: NRC-02 Op-Test No.: March 2008 Examiners: _____________________ Operators: _________________________
Initial Conditions: Simulator IC-17 or equivalent
- 1. Reactor Power 100%
Turnover:
- 2. All equipment operable.
- 2. Perform RHR Pump Operability Test IAW N2-OSP-RHS-Q@006 Event Malf. No. Event Event No. Type* Description 9 RHS*P2 N (BOP) Perform RHR Pump Operability Test IAW N2-OSP-RHS-N (SRO) Q@006 TS (SRO) RHS B/C Water Leg Pump breaker trip (TS) 10 IA02A,B C (BOP) Instrument Air Compressor A Trips, B will not start, C IA04A,B C (SRO) must be placed in service manually.
N2-SOP-19, Loss of Instrument Air 11 TC03A R (RO) Power decrease to 85% due to EHC oscillation problem R (SRO) N2-SOP-23, EHC Press Reg Failure N2-SOP-101D, Rapid Power Reduction.
12 CS01B C(BOP) HPCS spurious start. (TS)
C (SRO)
TS (SRO) 13 FW15 I (RO) Feedwater master controller fails as-is requiring manual I (SRO) control.
N2-SOP-6, Feedwater Failures 14 RR10A,B C (ALL) Recirculation FCV failure causes FCV to open.
N2-SOP-8, Unplanned Power Changes 15 TC02 M (ALL) EHC Regulator failure cause Reactor High Pressure, FW03A, ATWS, Loss of Feedwater B EOP-RPV, EOP-Failure to Scram RP02 EOP-6, Att.14 RP14A, B
16 RC07 C (BOP) RCIC controller failure. Requires manual actions to inject.
C (SRO) 9 RP08A, C (BOP) SLC pump fails to Auto-Start B C (SRO)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 2 March 2008
Facility: Nine Mile Point 2 Scenario No.: NRC-02 Op-Test No.: March 2008 TARGET QUANTITATIVE ATTRIBUTES ACTUAL (PER SCENARIO; SEE SECTION D.5.d) ATTRIBUTES
- 1. Total malfunctions (5-8) 7 Events 2,3,4,5,6,8,9
- 2. Malfunctions after EOP entry (1-2) 2 Events 8,9
- 3. Abnormal events (2-4) 4 Event 2 -SOP-19, Event 3SOP-101D, SOP-23 Event 5 - SOP-6, Event 6 SOP-8
- 4. Major transients (1-2) 1 Event 7
- 5. EOPs entered/requiring substantive 2 actions (1-2)
- 6. EOP contingencies requiring substantive 1 actions (0-2)
Event 7,8,9 EOP-C5,
- 7. Critical tasks (2-3) 4 CRITICAL TASK DESCRIPTIONS:
CT-1.0 Place ADS inhibit switches to ON to prevent injection under ATWS conditions CT-2.0 restore & maintain RPV level above the MSCWL precluding the need to perform a RPV Blowdown.
CT-3.0 Inject SLC before exceeding HCTL CT-4.0 RO inserts all control rods NRC Scenario 2 March 2008
SCENARIO
SUMMARY
The scenario begins at 100% power. The RO will perform the surveillance test for the C RHR Pump, N2-OSP-RHS-Q006. While the pump is running, the breaker will trip for the RHS B/C Water Leg Pump RHS*P2 requiring a TS entry by the SRO (TS 3.5.1.C - 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). Once TS are addressed, the in service Instrument Air Compressor will trip requiring operator action to manually start the C standby compressor. The B compressor will not start.
Oscillations of the in service EHC pressure regulator will occur and require actions to swap to the alternate regulator and lower reactor pressure. Additionally, the RO will be required to lower reactor power to 85% IAW the SOP-23.
Once conditions stabilize, a HPCS spurious start will occur requiring operator action to terminate the initiation. The SRO will address TS for HPCS inoperability and 2 other ECCS pumps inoperable (3.0.3). After addressing TS, the Feedwater Master Controller will then fail as-is. The crew will enter SOP-6 and control feedwater in manual. Additionally, a failure of the Recirculation FCV will cause the FCV to open. Operator action will be required to control the FCV and reactor level. Cram rods may be inserted or Recirc flow lowered to lower reactor power to pre-transient levels.
The backup EHC pressure regulator will fail and result in a rapid RPV pressure rise. The reactor will automatically scrams, however, all control rods will not fully insert and A and B reactor feed pumps will trip. EOPs RPV, EOP-Failure-To-Scram will be entered. The RO must inhibit ADS to prevent injection during the ATWS (CT).
The RCIC turbine can be manually controlled after a controller malfunction. SLC pumps will fail to auto-start and must be manually started prior to exceeding the HCTL (CT). RPV level must be restored with C FW pump, RCIC (or Condensate Booster Pumps with RPV pressure lowered) precluding the need to perform a RPV Blowdown (CT). The RO will implement actions to insert control rods until all rods are inserted (CT).
The scenario can be terminated when RPV level is being controlled in the required band and all control rods are inserted.
NRC Scenario 2 March 2008
Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 2 Scenario No.: NRC-03 Op-Test No.: March 2008 Examiners: _____________________ Operators: _________________________
Initial Conditions: Simulator IC-244
- 1. Reactor Power = Startup @ 900 psig
- 2. Rod 50-27 is the next rod to move
- 3. BPV #1 is approximately 15% open
- 4. Other operators will be performing SJAE startup at step 2.45 of N2-OP-101A
- 5. N2-OP signed off thru step 3.3.32
- 6. N2-OP-101A - step 2.46.3 in progress
- 7. N2-OP-101A - step 2.45 in progress (to b)e completed by other operators Turnover:
- 3. Continue Power Increase to get one bypass valve open approximately 25%
- 2. Transfer Reboiler Steam Supply to Main Steam IAW N2-OP-25, Section 5.0, then continue startup Event Malf. No. Event Event No. Type* Description 17 N/A R (RO) Continue startup R (SRO) N2-OP-101A 18 N/A N (BOP) Transfer Reboiler Steam Supply to Main Steam N (SRO) N2-OP-25 19 NM09A I (RO) IRM A Inop Trip I (SRO) N2-OP-92 Neutron Monitoring, N2-OP-97 RPS 20 ED04F C (BOP) Loss of power to Div I switchgear. (TS) Restore non-essential TS (SRO) Service Water, Drywell Cooling.
N2-SOP-3 Loss of AC Power 21 MT01 (.085) C (ALL) Small Seismic Event, Service Water Pump trip (TS); IRM G CW01F TS (SRO) Fails Upscale (TS)
NM06G N2-SOP-90 Seismic Event, N2-SOP-3 Loss of AC Power 22 MT01 (.25) M (ALL) Seismic Aftershock Event, RPV Instrument Reference Line RR35A Rupture Inside Drywell; Division II ECCS fails to auto initiate RR14B N2-SOP-90 Seismic Event, EOP-RPV, EOP-PC 23 RR34A C (BOP) Loss of Additional Level Instruments and increased RCS RR20 (0.8%) C (SRO) leakage requiring Suppression Chamber Spray 24 RR27 C (BOP) All RPV level indication lost, RPV Blowdown Required, only 5 AD08E C (SRO) ADS valves open.
AD08G EOP-C4
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 3 March 2008
Facility: Nine Mile Point 2 Scenario No.: NRC-03 Op-Test No.: March 2008 TARGET QUANTITATIVE ATTRIBUTES ACTUAL (PER SCENARIO; SEE SECTION D.5.d) ATTRIBUTES
- 1. Total malfunctions (5-8) 6 Events 3,4,5,6,8,9
- 2. Malfunctions after EOP entry (1-2) 2 Events 8,9
- 3. Abnormal events (2-4) 3 Event 3 -SOP-97, Event 4 SOP-3, Event 4-SOP-90
- 4. Major transients (1-2) 1 Event 6
- 5. EOPs entered/requiring substantive 2 actions (1-2)
- 6. EOP contingencies requiring substantive 1 actions (0-2)
Event 8 EOP-C4
- 7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:
CT-1.0 initiate an RPV blowdown when level indication is lost or if the PSP is exceeded.
CT-2.0 flood the RPV to the elevation of the main steam lines IAW RPV flooding.
NRC Scenario 3 March 2008
Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 2 Scenario No.: NRC-04 Op-Test No.: March 2008 Examiners: _____________________ Operators: _________________________
Initial Conditions: Simulator IC-17 Reactor Power 100%
Turnover:
- 4. Reduce power to 90% for a rod line adjustment which will take place on the next shift.
- 2. Perform N2-OSP-RMC-W@001 Control Rod Movement and Position Verification Surveillance Test Event Malf. No. Event Event No. Type* Description 25 N/A R (RO) Reduce power to 90% at approximately 2% per minute.
R (SRO) 26 N/A N (RO) Perform N2-OSP-RMC-W@001, Control Rod Movement &
N (SRO) Position Verification test.
27 RD18 C (RO) CRD P1A suction filter clog causes pump trip.
C (SRO) N2-SOP-30, CRD Failures 4 RD11 TS (SRO) Rod Position Indication Lost 5 override C (BOP) Control room AC unit trips (TS 3.7.2.A - 7 days, TS 3.7.3.A s TS (SRO) - 30 days.
6 EG06A C (BOP) Stator water pump trip, failure of standby to auto start, C (SRO) Generator RB. Power reduction may be required.
N2-SOP-68, Loss of Stator Cooling 7 override C (BOP) Loss of NNS-SWG015, loss of one division of RPS s C (SRO) solenoids N2-SOP-3, Loss of AC Power, N2-SOP-97 RPS Failures 8 RP03 M (ALL) Small containment leak, Mode Switch and RPS Manual PB MS03 fail, ARI successful.
EOP-RPV, EOP-C5 Failure-To-Scram 9 DG01A, C (RO) Loss of Offsite Power with EDG auto-start failures (Station C C SRO) Blackout)
N2-SOP-3, N2-SOP-1, N2-SOP-11, EOP-RPV, EOP-PC 10 MS04 Steam Leak requires DW Spray before exceeding Pressure Suppression Pressure.
NRC Scenario 4 March 2008
Facility: Nine Mile Point 2 Scenario No.: NRC-04 Op-Test No.: March 2008 TARGET QUANTITATIVE ATTRIBUTES ACTUAL (PER SCENARIO; SEE SECTION D.5.d) ATTRIBUTES
- 1. Total malfunctions (5-8) 5 Events 3,5,6,7,9
- 2. Malfunctions after EOP entry (1-2) 1 Events 9
- 3. Abnormal events (2-4) 4 Event 3- SOP-30, Event 6-SOP-68, Event 7-3,13,60,97, Event 9 - SOP Station Blackout
- 4. Major transients (1-2) 1 Event 9
- 5. EOPs entered/requiring substantive 2 actions (1-2)
- 6. EOP contingencies requiring substantive 1 actions (0-2)
EOP-Failure to Scram,
- 7. Critical tasks (2-3) 3 CRITICAL TASK DESCRIPTIONS:
CT 1.0 - Upon Mode Switch and RPS PB Failure, Scram is accomplished with RRCS CT 2.0 - a Loss of Off-Site Power with a failure of the EDGs the operators will take actions to re-power at least one vital bus (either with a local EDG start or HPCS EDG cross tie) IAW SOP-03.
CT 3.0 - drywell spray is initiated prior to exceeding the PSP.
NRC Scenario 4 March 2008