ML053330588
| ML053330588 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/03/2005 |
| From: | Lantz R E Operations Branch IV |
| To: | Williams G A Entergy Operations |
| References | |
| 50-416/05-301 50-416/05-301, ES-301, es-301-1 | |
| Download: ML053330588 (34) | |
Text
ES-301Administrative Topics OutlineForm ES-301-1 Facility: Grand Gulf Nuclear StationDate of Examination:
16 August 2005Examination Level (circle one)
RO / SROOperating Test Number:Administrative Topic (see Note)
Type Code*Describe activity to be performed Conduct of Operations D; SGiven plant conditions, perform the Idle Loop Startup Surveillance for Recirculation System.
GJPM-RO-ADM-1 K/A 2.1.20: 4.3 Safety Function 1 Conduct of Operations M; SGiven plant conditions, complete documentation for Shift Turnover GJPM-RO-ADM-2 K/A 2.1.3: 3.0Equipment Control MGiven a component to be isolated for a work order, prepare a tagout using the eSOMS program.
GJPM-RO-ADM-3 K/A 2.2.13: 3.6 Radiation Control DComplete entry and egress from the CAA with access requirements for a High Radiation Area.
GJPM-OPS-ADM-26 K/A 2.3.1: 2.6; 2.3.4: 2.5; 2.3.5: 2.3Emergency PlanN/AN/ANOTE: 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 (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) (S)imulator ES-301Control Room/In-Plant Systems OutlineForm ES-301-2 Facility: Grand Gulf Nuclear StationDate of Examination:
15 August 2005Exam Level (circle one)
RO / SRO-I / SRO-UOperating Test Number:
Control Room Systems
@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) System / JPM TitleType Code*Safety Functiona.202001 Recirculation System - Startup idle Recirculation Pump <30% power with incomplete start actuation.S; M; A4b.201001 Control Rod Drive Hydraulic System - Rotate operating CRD pumps, trip of newly operating pump.S; M; A1c.259001 Reactor Feed Water System - Startup Second Reactor Feed Pump and place on Master level control, with failure of Automatic controller.S; N; A2d.226001 RHR Containment Spray - Secure Containment Sprays and align for injection to RPV with failure of one RHR injection
valve.S; N; A5 ESFe.264000 Emergency Diesel Generators
- Start, parallel, and load the Diesel Generator with trip of SSW.S; D; A6 ESFf.290003 Control Room HVAC System - Secure Control Room Standby Fresh Air System.C; D9 ESFg.239001 Main & Reheat Steam System - Open Main Steam Isolation Valves.S; D; L3 ESFh.201005 Rod Control & Information System
- Operate RCIS to bring the reactor critical.S; D; L7 In-Plant Systems
@ (3 for RO; 3 for SRO-I; 3or2 for SRO-U)i.295003 Partial Loss of AC
- Reset undervoltage lockouts on buses when power is restored, with one lockout failing to reset.
R; M; E; A 6j.286000 Fire Protection System
- Manually initiate fire suppression for the Standby Gas Filter Train with the train operating.R; N; E8k.295016 Control Room Abandonment
- Startup RHR in Suppression Pool Cooling.N; E; L7 ESF@ 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 CodesCriteria for RO / SRO-I / SRO-U (A)lternate path (C)ontrol Room (D)irect from bank (E)mergency or abnormal in-plant (L)ow-Power (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator4-6 / 4-6 / 2-39 / 8 / 41 / 1 / 11 / 1 / 12 / 2 / 13 / 3 / 2 (randomly selected)1 / 1 / 1 ES-301Administrative Topics OutlineForm ES-301-1 Facility: Grand Gulf Nuclear StationDate of Examination:
16 August 2005Examination Level (circle one) RO /
SROOperating Test Number:Administrative Topic (see Note)
Type Code*Describe activity to be performed Conduct of Operations MGiven a completed AC/DC Lineup following a failure of a Diesel Generator, determine the LCO action requirements and generate an eSOMS LCO.
GJPM-SRO-ADM-1 K/A 2.1.12: 4.0 Safety Function 6 Conduct of Operations MGiven a failed relay, determine the impact on plant operations using facility drawings.
GJPM-SRO-ADM-2 K/A 2.1.24: 3.1Equipment Control MGiven a work order and prepared tagout, determine the adequacy of the tagout and the impact on plant operations.
GJPM-SRO-ADM-3 K/A 2.2.13: 3.8; 2.2.17: 3.5 Radiation Control NGiven plant conditions, determine Protective Action Recommendations and Radiological Considerations for On-Site Personnel.
GJPM-SRO-A&E-41 K/A 2.3.8: 3.2Emergency Plan MGiven plant conditions, determine entry into the Site Emergency Plan and complete the initial notification forms.
GJPM-SRO-A&E-42 K/A 2.4.41: 4.1; 2.4.38: 4.0; 2.4.40: 4.0NOTE: 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 (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) (S) imulator ES-301Control Room/In-Plant Systems OutlineForm ES-301-2 Facility: Grand Gulf Nuclear StationDate of Examination:
15 August 2005Exam Level (circle one)
RO / SRO-I / SRO-UOperating Test Number:
Control Room Systems
@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)System / JPM TitleType Code*Safety Functiona.202001 Recirculation System - Startup idle Recirculation Pump <30% power with incomplete start actuation.S; M; A4b.201001 Control Rod Drive Hydraulic System - Rotate operating CRD pumps, trip of newly operating pump.S; M; A1c.259001 Reactor Feed Water System - Startup Second Reactor Feed Pump and place on Master level control, with failure of Automatic controller.S; N; A2d.226001 RHR Containment Spray - Secure Containment Sprays and align for injection to RPV with failure of one RHR injection
valve.S; N; A5 ESFe.264000 Emergency Diesel Generators
- Start, parallel, and load the Diesel Generator with trip of SSW.S; D; A6 ESFf.290003 Control Room HVAC System - Secure Control Room Standby Fresh Air System.C; D9 ESFg.239001 Main & Reheat Steam System - Open Main Steam Isolation Valves.S; D; L3 ESFh.N/A In-Plant Systems
@ (3 for RO; 3 for SRO-I; 3or2 for SRO-U)i.295003 Partial Loss of AC
- Reset undervoltage lockouts on buses when power is restored, with one lockout failing to reset.
R; M; E; A 6j.286000 Fire Protection System
- Manually initiate fire suppression for the Standby Gas Filter Train with the train operating.R; N; E8k.295016 Control Room Abandonment
- Startup RHR in Suppression Pool Cooling.N; E; L7 ESF@ 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 CodesCriteria for RO / SRO-I / SRO-U (A)lternate path (C)ontrol Room (D)irect from bank (E)mergency or abnormal in-plant (L)ow-Power (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator4-6 / 4-6 / 2-39 / 8 / 41 / 1 / 11 / 1 / 12 / 2 / 13 / 3 / 2 (randomly selected)1 / 1 / 1 ES-301Control Room/In-Plant Systems OutlineForm ES-301-2 Facility: Grand Gulf Nuclear StationDate of Examination:
15 August 2005Exam Level (circle one)
RO / SRO-I /
SRO-UOperating Test Number:
Control Room Systems
@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)System / JPM TitleType Code*Safety Functiona.202001 Recirculation System - Startup idle Recirculation Pump <30% power with incomplete start actuation.S; M; A4b.201001 Control Rod Drive Hydraulic System - Rotate operating CRD pumps, trip of newly operating pump.S; M; A1c.N/Ad.N/Ae.N/Af.290003 Control Room HVAC System - Secure Control Room Standby Fresh Air System.C; D9 ESFg.N/Ah.N/A In-Plant Systems
@ (3 for RO; 3 for SRO-I; 3or2 for SRO-U)i.N/Aj.286000 Fire Protection System
- Manually initiate fire suppression for the Standby Gas Filter Train with the train operating.R; N; E8k.295016 Control Room Abandonment
- Startup RHR in Suppression Pool Cooling.N; E; L7 ESF@ 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 CodesCriteria for RO / SRO-I / SRO-U (A)lternate path (C)ontrol Room (D)irect from bank (E)mergency or abnormal in-plant (L)ow-Power (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator4-6 / 4-6 / 2-39 / 8 / 41 / 1 / 11 / 1 / 12 / 2 / 13 / 3 / 2 (randomly selected)1 / 1 / 1 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEEMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 (RO/SRO)Form ES-4 0 E/APE #/NAME/SAFETY FUNCTIONK1K2K3A1A2G TOPIC(S)IR295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4
CFR 2.4.
4Given plant conditions, parameters, and a loss of the recirculation system, determine appropriate actions.4.0295003 Partial or Complete Loss of AC Power/ 6
CFR 01Given plant conditions and a loss of AC power, determine the necessary actions to restore vital
busses.3.3295004 Partial or Complete Loss of DC Power / 6
CFR 02Given plant conditions and a loss of DC power, determine the effect to the SSW system.3.8295005 Main Turbine Generator Trip / 3 CFR 03Following a reactor scram and subsequent main turbine generator trip, determine the effects of manual bypass valve operation on reactor water level.3.5295006 SCRAM / 1CFR02 Given plant conditions following a reactor scram, determine if adequate shutdown margin exists.3.4295016 Control Room Abandonment / 7
CFR 01Describe the method used to manually scram the reactor after the control room has been abandoned.3.8295018 Partial or Complete Loss of CCW / 8
CFR 01Given plant conditions and a partial loss of Component Cooling Water, determine the
necessary actions to ensure the plant remains/returns to a safe condition.
3.5295019 Partial or Complete Loss of Inst. Air / 8
CFR 01Given indications of a partial loss of Instrument Air determine a method to restore Instrument Air system pressure.
3.5295021 Loss of Shutdown Cooling / 4
CFR 01Given specific plant conditions following a loss of Shutdown Cooling, determine the reason for
raising reactor water level.
3.3295023 Refueling Accidents / 8CFR03Determine the correct operator response to inadvertent criticality following a refueling
accident.3.7295024 High Drywell Pressure /
5 CFR 2.1.
23Given plant conditions and high drywell pressure, determine the method to lower drywell pressure.3.9 295025 High Reactor Pressure /
3 CFR 05 Describe RCIC operation following a reactor scram where the SRVs are used to control reactor
pressure.3.6295026 Suppression Pool High Water Temp. / 5
CFR 01Given an ATWS condition, describe the EP bases for lowering reactor pressu re as Suppression Pool temperature rises.
3.8295027 High Containment Temperature / 5
CFR 03Given rising Containment temperature, describe the necessary actions to maintain the plant/containment in a safe condition.
3.5 PAGE 1 TOTAL TIER 1 GROUP 1404312 PAGE TOTAL # QUESTIONS 14 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEEMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 (RO/SRO)Form ES-4 0 E/APE #/NAME/SAFETY FUNCTIONK1K2K3A1A2G TOPIC(S)IMP295028 High Drywell Temperature / 5
CFR 02Given plant conditions and elevated drywell temperature, determine the effects to control room reactor water level indication.
2.9295030 Low Suppression Pool Water Level / 5
CFR 2.2.
12 Given a low suppression pool level condition, determine the effects to other plant systems.
3.0295031 Reactor Low Water Level / 2
CFR 04Given plant conditions, describe the operation of the High Pressure Core Spray system following
a LOCA.4.3 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1
CFR 06Given plant conditions and an ATWS condition, determine the availability of the main condenser as a heat sink.3.8295038 High Offsite Release
Rate / 9 CFR 01Given a radioactive release from the plant, determine when it is considered to be offsite.3.3600000 Plant Fire On Site / 804Determine the required procedural actions for a fire on the plant site.2.8 PAGE 2 TOTAL TIER 1 GROUP 1102111 PAGE TOTAL # QUESTIONS 6 PAGE 1 TOTAL TIER 1 GROUP 1404312 PAGE TOTAL # QUESTIONS 14TIER 1 GROUP 1 TOTALS50642320 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEEMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 (RO/SRO)Form ES-4 E/APE #/NAME/SAFETY FUNCTIONK1K2K3A1A2G TOPIC(S)IMP295002 Loss of Main Condenser Vacuum / 3
CFR 01Given plant conditions and degrading main condenser vacuum, determine the automatic plant response (RPS actuation).
3.5 295007
High Reactor Pressure /
3 CFR 2.4.
35Determine the conditions necessary to require connection of an alternate air source to the
SRVs.3.3295008 High Reactor Water Level / 2295009 Low Reactor Water Level / 2295010 High Drywell Pressure /
5295011 High Containment Temperature / 5295012 High Drywell Temperature / 5295013 High Suppression Pool Water Temp. / 5
CFR 02Describe the preferred method to minimize localized suppression pool heating when using
the SRVs to control reactor pressure without suppression cooling in service.
3.2 295014
Inadvertent Reactivity Addition / 1295015 Incomplete SCRAM / 1 295017 High Offsite Release
Rate / 9295020 Inadvertent Cont.
Isolation / 5 & 7295022 Loss of CRD Pumps / 1295029 High Suppression Pool Water Level / 5 PAGE 1 TOTAL TIER 1 GROUP 2010011 PAGE TOTAL # QUESTIONS 3
GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEEMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 (RO/SRO)Form ES-4 E/APE #/NAME/SAFETY FUNCTIONK1K2K3A1A2G TOPIC(S)IMP295032 High Secondary Containment Area Temperature /
5 CFR 02Given plant conditions including elevated Auxiliary Building temperatures, describe the
conditions that would require a reactor scram.3.5295033 High Secondary Containment Area Radiation Levels / 9295034 Secondary Containment Ventilation High Radiation / 9
CFR 03Given plant conditions including elevated Auxiliary Building radiation levels, describe the conditions that would automatically start the Standby Gas Treatment system.
4.3295035 Secondary Containment High Differential Pressure / 5
CFR 02 Given accident conditions and a Standby Gas Treatment system failure, determine the type of
release.295036 Secondary Containment High Sump/Area Water Level / 5
CFR 01Describe the system logic used by the Auxiliary Building Floor Drain system to contain a significant CCW system rupture.
3.2500000 High CTMT Hydrogen Conc. / 5 PAGE 2 TOTAL TIER 1 GROUP 2111100 PAGE TOTAL # QUESTIONS 4 PAGE 1 TOTAL TIER 1 GROUP 2010011 PAGE TOTAL # QUESTIONS 3TIER 1 GROUP 2 TOTALS1211117 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)Form ES-4 0SYSTEM #/NAMEK1K2K3K4K 5 K 6 A 1 A 2 A 3 A 4GTOPIC(S)IMP203000 RHR/LPCI: Injection Mode
CFR 02Given plant conditions, describe the design features and limits of the RHR pump manual override feature.
3.5 8205000
Shutdown Cooling CFR 04Describe the RHR Shutdown Cooling system NPSH interlocks.
2.6 8206000
HPCI N/A GGNS207000 Isolation (Emergency) Condenser N/A GGNS209001 LPCS CFR01Given degraded plant conditions during a LOCA, describe LPCS manual operation.
3.8 8209002
HPCSCFR09 Describe available methods to raise/lower suppression pool level using HPCS.
3.4 8209002
HPCS CFR 2.1.
2 8 Describe the bases for the HPCS injection
valve high reactor water level interlock.
3.2 8 211000 SLCCFR02Predict the SLC system indication and response with indication the squib valve failed to actuate and follow up actions.
3.6 8212000
RPSCFR12Given plant conditions including a partial main turbine stop/control valve closure, determine
the effect to RPS.
4.0 8215003
IRMCFR03 Describe the reason for the precaution concerning driving IRMs during surveillance activities.
3.0 8215004
Source Range Monitor CFR 2.2.
3 3Describe the SRM precaution warning of a potential control rod
block even if the channel is bypassed.
2.5 8215005
APRM /
LPRM CFR 02Given a partial loss of plant electrical power, determine the effect to the APRMs.
2.6 8217000
RCICCFR02 Predict how a reactor pressure change will affect RCIC system
flow.3.3 8 PAGE 1 TOTAL TIER 2 GROUP 101012012022 PAGE TOTAL # QUESTIONS11 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)Form ES-4 0SYSTEM #/NAMEK 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4GTOPIC(S)IMP 218000 ADSCFR01Describe the relationship between ADS Logic power and the operation of the ADS logic.
3.1 8223002
PCIS / Nuclear Steam Supply Shutoff
CFR 03Determine the operator actions required to mitigate a NSSSS logic
failure.3.0 4 8239002 SRVsCFR09Describe the design features available to determine if a SRV is
open.3.7 4 8259002 Reactor Water Level Control
CFR 04Describe potential problems associated with operating a RFP in Emergency Manual 3.0 4 8259002 Reactor Water Level Control
CFR 06Describe prerequisites for transferring the Feedwater system to 3-element control.
3.1 4 2 261000 SGTSCFR03 Describe the SGTS damper logic following system initiation.
3.0 4 8262001 AC Electrical Distribution
CFR 01Given plant conditions and a partial loss of DC power, determine the
affect to the AC distribution system.
3.1 4 8262002 UPS (AC/DC)CFR01Given plant conditions and degraded AC power, determine the status of plant inverters.
3.1 4 8263000 DC Electrical Distribution
CFR 01Given a loss of AC power to battery chargers, determine the
affects to the DC distribution system.
4 8 264000 EDGs CFR10Describe EDG response to a LOCA.3.9 4 8 264000 EDGs
CFR 2.4.
4 8Determine EDG status from control room alarms and indications and any required operator actions to improve plant conditions.
3.5 4 8300000 Instrument AirCFR01Determine the effect on the plant given a loss of Instrument Air to the containment.
2.7 8 PAGE 2 TOTAL TIER 2 GROUP 101130103111 PAGE TOTAL # QUESTIONS12 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)Form ES-SYSTEM #/NAMEK1K2K3K4K5K6A1A2A3A4GTOPIC(S)
IMP300000 Instrument AirCFR13Determine the affect of a clogged filter on the Instrument Air system.
2.8400000 Component Cooling Water
CFR 04Determine the method used to confirm a reactor coolant leak into the CCW system.2.9400000 Component Cooling Water
CFR 02Determine the affect to the plant if the CCW temperature control fails.
2.8 PAGE 3 TOTAL TIER 2 GROUP 110000110000 PAGE TOTAL # QUESTIONS3 PAGE 1 TOTAL TIER 2 GROUP 101012012022 PAGE TOTAL # QUESTIONS11 PAGE 2 TOTAL TIER 2 GROUP 101130103111 PAGE TOTAL # QUESTIONS12 TIER 2GROUP 1 TOTALS1214222513326 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)Form ES-4SYSTEM #/NAMEK1K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4GTOPIC(S)IMP201001 CRD Hydraulic CFR201002 RMCS N/A GGNS201003 Control Rod and
Drive Mechanism
CFR201004 RSCS N/A GGNS201005 RCIS CFR10Describe the purpose for the rod withdrawal limiter.3.2201006 RWM N/A GGNS202001 Recirculation CFR 2.2.
2 5Given degraded plant conditions determine any
applicable Recirculation Loop LCOs.
2.5202002 Recirculation Flow Control CFR41.6 01Given plant conditions, determine any automatic
actions associated with the Recirculation System
HPUs. 3.5204000 RWCUCFR06Determine the correct flow path to use RWCU
as an alternate shutdown cooling. 2.6214000 RPIS N/A GGNS215001 Traversing In-Core Probe
CFR215002 RBM N/A GGNS PAGE 1 TOTAL TIER 2 GROUP 210011000001 PAGE TOTAL # QUESTIONS4 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)Form ES-SYSTEM #/NAMEK1K2K3K4K5K6A1A2A3A4GTOPIC(S)
IMP216000 Nuclear Boiler Instrumentation
CFR219000 RHR /LPCI Suppression Pool Cooling Mode
CFR223001 Primary CTMT and Auxiliaries
CFR 08Determine the limitations to SRV usage given
a reduced suppression pool level.3.6226001 RHR/LPCI: CTMT Spray Mode
CFR230000 RHR/LPCI:
Torus/Pool Spray Mode N/A GGNS233000 Fuel Pool Cooling and Cleanup
CFR234000 Fuel Handling Equipment CFR239001 Main and Reheat
Steam CFR 04Given plant conditions including a MSIV closure, determine the
affect to the Offgas system.
2.8239003 MSIV Leakage Control CFR 02Explain the relationship
between the
MSIV Leakage Control system
and SGTS.
2.9 241000
Reactor/Turbine Pressure Regulator
CFR 2.4.
6 Describe the bases for each of the Scram ONEP immediate actions.3.1 PAGE 2 TOTAL TIER 2 GROUP 220100000001 PAGE TOTAL # QUESTIONS4 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)Form ES-SYSTEM #/NAMEK 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4GTOPIC(S)IMP245000 Main Turbine Gen./Aux.
CFR 02Determine main turbine critical speeds as it is rolled to rated speed.
2.8 256000
Reactor Condensate
CFR 259001 Reactor Feedwater CFR03Determine necessary actions and priorities immediately after a single condensate pump trips with the plant at rated conditions.
3.6268000 RadwasteCFR04Determine the Drywell Floor Drains indications
available to detect drywell
general area leakage.
2.7 271000
Offgas CFR272000 Radiation Monitoring
CFR286000 Fire Protection
CFR 288000 Plant Ventilation
CFR290001 Secondary CTMT CFR03Determine inputs to the Fuel Pool leak detection standpipe.
2.8290003 Control Room HVAC CFR 290002 Reactor Vessel
Internals CFR PAGE 3 TOTAL TIER 2 GROUP 210010001100 PAGE TOTAL # QUESTIONS4 PAGE 1 TOTAL TIER 2 GROUP 210011000001 PAGE TOTAL # QUESTIONS4 PAGE 2 TOTAL TIER 2 GROUP 220100000001 PAGE TOTAL # QUESTIONS4 TIER 2 GROUP 2 TOTALS4012100110212 ES-401Generic Knowledge and Abilities Outline (Tier 3)Form-4 Facility:
Grand Gulf Nuclear StationDate of Exam:
12 August 2005CategoryK/ A#TopicROSRO-OnlyIR#IR#2.1.19Given plant conditions and the PDS computer, determine necessary actions based on PBDS counts.3.0 66 8982.1.25Given plant conditions and EOP-3 graphs, determine the correct mitigation strategy.2.8 67 8991.2.1.29Determine the correct locking device color coding for locked components. 3.4 68 237aConduct2.1Of Operations2.1 2.1Subtotal 32.2.1Given plant conditions, determine proper operation of the IRMs.3.7 69 9002.2.30Discuss the duties of the operator assigned to communicate with the refueling floor SRO during core alterations.
3.5 70 9012.2.2Equipment 2.2Control2.2 2.2Subtotal 22.3.1Given the need to enter a high radiation area, determine the allowed time in the area to prevent exceeding the administrative exposure limits.
2.6 71 9022.3.4Given plant conditions and applicable Emergency Planning Procedures, determine the radiation exposure limits that are in effect.
2.5 72 9033.2.3Radiation 2.3Control2.3 2.3Subtotal 22.4.20Given plant conditions, determine the bases for any applicable EOP cautions.3.3 73 9042.4.25Given plant conditions including a fire, determine the proper response.2.9 74 9054.2.4.43Given plant conditions and Emergency Plan Procedures, determine the available emergency communications systems.
2.8 75 906Emergency2.4Procedures /2.4Plan2.4Subtotal 3Tier 3 Point Total 10 7 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEEMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 (RO/SRO)Form ES-4 0 E/APE #/NAME/SAFETY FUNCTIONK1K2K3A1A2G TOPIC(S)IR295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4
CFR 2.4.
4Given plant conditions, parameters, and a loss of the recirculation system, determine appropriate actions.4.3295003 Partial or Complete
Loss of AC Power/ 6
CFR 01Given plant conditions and a loss of AC power, determine the necessary actions to restore vital
busses.3.5295004 Partial or Complete Loss of DC Power / 6
CFR 02Given plant conditions and a loss of DC power, determine the effect to the SSW system.4.1295005 Main Turbine Generator Trip / 3 CFR 03Following a reactor scram and subsequent main turbine generator trip, determine the effects of manual bypass valve operation on reactor water level.3.7295006 SCRAM / 1CFR02 Given plant conditions following a reactor scram, determine if adequate shutdown margin exists.3.7295016 Control Room Abandonment / 7
CFR 01Describe the method used to manually scram the reactor after the control room has been abandoned.3.9295018 Partial or Complete Loss of CCW / 8
CFR 01Given plant conditions and a partial loss of Component Cooling Water, determine the
necessary actions to ensure the plant remains/returns to a safe condition.
3.6295019 Partial or Complete Loss of Inst. Air / 8
CFR 01Given indications of a partial loss of Instrument Air determine a method to restore Instrument Air system pressure.
3.6295021 Loss of Shutdown Cooling / 4
CFR 01Given specific plant conditions following a loss of Shutdown Cooling, determine the reason for
raising reactor water level.
3.4295023 Refueling Accidents / 8CFR03Determine the correct operator response to inadvertent criticality following a refueling
accident.4.0295024 High Drywell Pressure /
5 CFR 2.1.
23Given plant conditions and high drywell pressure, determine the method to lower drywell pressure.4.0 295025 High Reactor Pressure /
3 CFR 05 Describe RCIC operation following a reactor scram where the SRVs are used to control reactor
pressure.3.7295026 Suppression Pool High Water Temp. / 5
CFR 01Given an ATWS condition, describe the EP bases for lowering reactor pressu re as Suppression Pool temperature rises.
4.1295027 High Containment Temperature / 5
CFR 03Given rising Containment temperature, describe the necessary actions to maintain the plant/containment in a safe condition.
3.8 PAGE 1 TOTAL TIER 1 GROUP 1404312 PAGE TOTAL # QUESTIONS 14 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEEMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 (RO/SRO)Form ES-4 0 E/APE #/NAME/SAFETY FUNCTIONK1K2K3A1A2G TOPIC(S)IMP295028 High Drywell Temperature / 5
CFR 02Given plant conditions and elevated drywell temperature, determine the effects to control room reactor water level indication.
3.1295030 Low Suppression Pool Water Level / 5
CFR 2.2.
12 Given a low suppression pool level condition, determine the effects to other plant systems.
3.4295031 Reactor Low Water Level / 2
CFR 04Given plant conditions, describe the operation of the High Pressure Core Spray system following
a LOCA.4.2 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1
CFR 06Given plant conditions and an ATWS condition, determine the availability of the main condenser as a heat sink.4.1295038 High Offsite Release
Rate / 9 CFR 01Given a radioactive release from the plant, determine when it is considered to be offsite.4.3600000 Plant Fire On Site / 804Determine the required procedural actions for a fire on the plant site.3.4295004 Partial or Complete
Loss of DC Power / 6
CFR 02Given a loss of Division 1 DC logic power, determine the affect to the Division 1 ECCS.3.9*295005 Main Turbine Generator Trip / 3
CFR 2.3.
5Given plant data including current area dose rates, determine the required personnel monitoring equipment needed to enter the main turbine/generator area to investigate the cause for a trip.
2.5*295026 Suppression Pool High Water Temp. / 5
CFR 03Given plant conditions including rising Suppression Pool temperature, interpret HCTL and determine appropriate actions.
4.0*295027 High Containment Temperature / 5
CFR 2.2.
22Explain the bases for th e Technical Specification Containment average air temperature limit. 4.1*295030 Low Suppression Pool Water Level / 5
CFR 02Given low suppression pool water level, determine if suppression pool temperature can/cannot be measured and why.
3.9*295038 High Offsite Release Rate / 9 CFR 2.2.
28 Given a severe case fu el handling accident, explain the processes designed to prevent high
offsite release rates.
3.5*600000 Plant Fire On Site / 816Explain the automatic response of the plant Fire Protection system to a main transformer fire.3.5** SRO Only Questions PAGE 2 TOTAL TIER 1 GROUP 1102154 PAGE TOTAL # QUESTIONS 13 PAGE 1 TOTAL TIER 1 GROUP 1404312 PAGE TOTAL # QUESTIONS 14TIER 1 GROUP 1 TOTALS50646627 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEEMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 (RO/SRO)Form ES-4 E/APE #/NAME/SAFETY FUNCTIONK1K2K3A1A2G TOPIC(S)IMP295002 Loss of Main Condenser Vacuum / 3
CFR 01Given plant conditions and degrading main condenser vacuum, determine the automatic plant response (RPS actuation).
3.5 295007
High Reactor Pressure /
3 CFR 2.4.
35Determine the conditions necessary to require connection of an alternate air source to the
SRVs.3.5295008 High Reactor Water Level / 2295009 Low Reactor Water Level / 2295010 High Drywell Pressure /
5295011 High Containment Temperature / 5295012 High Drywell Temperature / 5295013 High Suppression Pool Water Temp. / 5
CFR 02Describe the preferred method to minimize localized suppression pool heating when using
the SRVs to control reactor pressure without suppression cooling in service.
3.5 295014
Inadvertent Reactivity Addition / 1295015 Incomplete SCRAM / 1 295017 High Offsite Release
Rate / 9295020 Inadvertent Cont.
Isolation / 5 & 7295022 Loss of CRD Pumps / 1295029 High Suppression Pool Water Level / 5 PAGE 1 TOTAL TIER 1 GROUP 2010011 PAGE TOTAL # QUESTIONS 3
GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEEMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 (RO/SRO)Form ES-4 E/APE #/NAME/SAFETY FUNCTIONK1K2K3A1A2G TOPIC(S)IMP295032 High Secondary Containment Area Temperature /
5 CFR 02Given plant conditions including elevated Auxiliary Building temperatures, describe the
conditions that would require a reactor scram.3.8295033 High Secondary Containment Area Radiation Levels / 9295034 Secondary Containment Ventilation High Radiation / 9
CFR 03Given plant conditions including elevated Auxiliary Building radiation levels, describe the conditions that would automatically start the Standby Gas Treatment system.
4.5295035 Secondary Containment High Differential Pressure / 5
CFR 02 Given accident conditions and a Standby Gas Treatment system failure, determine the type of
release.4.2295036 Secondary Containment High Sump/Area Water Level / 5
CFR 01Describe the system logic used by the Auxiliary Building Floor Drain system to contain a significant CCW system rupture.
3.3500000 High CTMT Hydrogen Conc. / 5295011 High Containment Temperature / 5
CFR 01Given LOCA conditions, determine when containment spray should be initiated.3.9*
295014 Inadvertent Reactivity Addition / 1
CFR 2.1.
14Given a control rod drifting out with the plant at power, determine any necessary notifications.3.3*295020 Inadvertent Cont.
Isolation / 5 & 7
CFR 03Given a partial MSIV closure, determine the affect on reactor power.3.7** SRO Only Questions PAGE 2 TOTAL TIER 1 GROUP 2111121 PAGE TOTAL # QUESTIONS 7 PAGE 1 TOTAL TIER 1 GROUP 2010011 PAGE TOTAL # QUESTIONS 3TIER 1 GROUP 2 TOTALS12113210 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)Form ES-401
-SYSTEM #/NAMEK1K2K3K4K 5 K 6 A 1 A 2 A 3 A 4GTOPIC(S)IMP#203000 RHR/LPCI: Injection Mode
CFR 02Given plant conditions, describe the design features and limits of the RHR pump manual override feature.
3.7205000 Shutdown Cooling CFR 04Describe the RHR Shutdown Cooling system NPSH interlocks.
2.6206000 HPCI N/A GGNS207000 Isolation (Emergency) Condenser N/A GGNS209001 LPCS CFR01Given degraded plant conditions during a LOCA, describe LPCS manual operation.
3.6209002 HPCSCFR09 Describe available methods to raise/lower suppression pool level using HPCS.
3.5209002 HPCS CFR 2.1.
2 8 Describe the bases for the HPCS injection
valve high reactor water level interlock.
3.3 211000
SLCCFR02Predict the SLC system indication and response with indication the squib valve failed to actuate and follow up actions.
3.9212000 RPSCFR12Given plant conditions including a partial main turbine stop/control valve closure, determine
the effect to RPS.
4.1215003 IRMCFR03 Describe the reason for the precaution concerning driving IRMs during surveillance activities.
3.1215004 Source Range Monitor CFR 2.2.
3 3Describe the SRM precaution warning of a potential control rod
block even if the channel is bypassed.
2.9215005 APRM /
LPRM CFR 02Given a partial loss of plant electrical power, determine the effect to the APRMs.
2.8217000 RCICCFR02 Predict how a reactor pressure change will affect RCIC system
flow.3.3 PAGE 1 TOTAL TIER 2 GROUP 101012012022 PAGE TOTAL # QUESTIONS11 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)Form ES-40 1SYSTEM #/NAMEK1K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4GTOPIC(S)IMP#218000 ADSCFR01Describe the relationship between ADS Logic power and the operation of the ADS logic.
3.3223002 PCIS / Nuclear Steam Supply Shutoff
CFR 03Determine the operator actions required to mitigate a NSSSS logic
failure.3.3239002 SRVsCFR09Describe the design features available to determine if a SRV is
open.3.6259002 Reactor Water Level Control
CFR 04Describe potential problems associated with operating a RFP in Emergency Manual 3.1259002 Reactor Water Level Control
CFR 06Describe prerequisites for transferring the Feedwater system to 3-element control.
3.2 261000
SGTSCFR03 Describe the SGTS damper logic following system initiation.
2.9262001 AC Electrical Distribution
CFR 01Given plant conditions and a partial loss of DC power, determine the
affect to the AC distribution system.
3.4262002 UPS (AC/DC)CFR01Given plant conditions and degraded AC power, determine the status of plant inverters.263000 DC Electrical Distribution
CFR 01Given a loss of AC power to battery chargers, determine the
affects to the DC distribution system.
3.4 264000
EDGsCFR10Describe EDG response to a LOCA.4.2 264000 EDGs
CFR 2.4.
4 8Determine EDG status from control room alarms and indications and any required operator actions to improve plant conditions.
3.8300000 Instrument AirCFR01Determine the effect on the plant given a loss of Instrument Air to the containment.
2.9 PAGE 2 TOTAL TIER 2 GROUP 101130103111 PAGE TOTAL # QUESTIONS12 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)Form ES-SYSTEM #/NAMEK1K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4GTOPIC(S)IMP300000 Instrument AirCFR13Determine the affect of a clogged filter on the Instrument Air system.
2.3400000 Component Cooling Water
CFR 04Determine the method used to confirm a reactor coolant leak into the CCW system.3.1400000 Component Cooling Water
CFR 02Determine the affect to the plant if the CCW temperature control fails.
2.8203000 RHR/LPCI: Injection Mode
CFR 2.3.
1 1Given LOCA conditions, determine how LPCI works in conjunction with the other ECCS to control
radiation releases.
3.2*209001 LPCS
CFR 2.1.
1 5Given a short-term problem associated with LPCS that does not affect operability, determine the most effective method to provide the information to operations personnel.
3.0*215003 IRM
CFR 2.4.
1 6Given plant conditions requiring entry into the
EOPs and the need to insert the IRMs, determine
the correct procedure hierarchy to accomplish the task.4.0*215004 Source Range Monitor CFR 2.2.
2 1 Given the applicable Tech Specs and a repaired SRM detector, determine the surveillance requirements to ensure operability.
3.5*217000 RCIC
CFR 2.1.
2 5Given plant conditions and procedures, determine Suppression Pool Level using the RCIC System.
3.1** SRO Only Questions PAGE 3 TOTAL TIER 2 GROUP 110000110005 PAGE TOTAL # QUESTIONS8 PAGE 1 TOTAL TIER 2 GROUP 101012012022 PAGE TOTAL # QUESTIONS11 PAGE 2 TOTAL TIER 2 GROUP 101130103111 PAGE TOTAL # QUESTIONS12 TIER 2GROUP 1 TOTALS1214222513831 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)Form ES-401
-SYSTEM #/NAMEK1K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4GTOPIC(S)IMP#201001 CRD Hydraulic CFR201002 RMCS N/A GGNS201003 Control Rod and
Drive Mechanism
CFR201004 RSCS N/A GGNS201005 RCIS CFR10Describe the purpose for the rod withdrawal limiter.3.3201006 RWM N/A GGNS202001 Recirculation CFR 2.2.
2 5Given degraded plant conditions determine any
applicable Recirculation Loop LCOs or safety limits.3.7202002 Recirculation Flow Control CFR41.6 01Given plant conditions, determine any automatic
actions associated with the Recirculation System
HPUs. 3.6204000 RWCUCFR06Determine the correct flow path to use RWCU
as an alternate shutdown cooling. 2.8214000 RPIS N/A GGNS215001 Traversing In-Core Probe
CFR215002 RBM N/A GGNS PAGE 1 TOTAL TIER 2 GROUP 210011000001 PAGE TOTAL # QUESTIONS4 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)Form ES-4 0SYSTEM #/NAMEK1K2K3K4K5K6A1A2A3A4GTOPIC(S)
IMP216000 Nuclear Boiler Instrumentation
CFR219000 RHR /LPCI Suppression Pool Cooling Mode
CFR223001 Primary CTMT and Auxiliaries
CFR 08Determine the limitations to SRV usage given
a reduced suppression pool level.3.8226001 RHR/LPCI: CTMT Spray Mode
CFR230000 RHR/LPCI:
Torus/Pool Spray Mode N/A GGNS233000 Fuel Pool Cooling and Cleanup
CFR234000 Fuel Handling Equipment CFR239001 Main and Reheat
Steam CFR 04Given plant conditions including a MSIV closure, determine the
affect to the Offgas system.
2.8239003 MSIV Leakage Control CFR 02Explain the relationship
between the
MSIV Leakage Control system
and SGTS.
3.0 241000
Reactor/Turbine Pressure Regulator
CFR 2.4.
6 Describe the bases for each of the Scram ONEP immediate actions.4.0 PAGE 2 TOTAL TIER 2 GROUP 220100000001 PAGE TOTAL # QUESTIONS4 GRAND GULF NUCLEAR STATIONBWR EXAMINATION OUTLINEPLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)Form ES-SYSTEM #/NAMEK 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4GTOPIC(S)IM P245000 Main Turbine Gen./Aux.
CFR 02Determine main turbine critical speeds as it is rolled to rated speed.
2.8 256000
Reactor Condensate
CFR 259001 Reactor
Feedwater CFR 03Determine necessary actions and priorities immediately after a single condensate pump trips with the plant at rated conditions.
3.6268000 RadwasteCFR04Determine the Drywell Floor Drains indications available
to detect drywell general area
leakage.2.9 271000 Offgas CFR272000 Radiation Monitoring
CFR286000 Fire Protection
CFR 288000 Plant Ventilation
CFR290001 Secondary CTMT CFR03Determine inputs to the Fuel Pool leak detection standpipe.
2.8290003 Control Room HVAC CFR 290002 Reactor Vessel
Internals CFR 2.4.
1 4 Given a severe accident condition, describe the bases for why the transition is made from the EOPs to the
SAPs.3.9*226001 RHR/LPCI: CTMT Spray Mode
CFR 13Determine the affects to the Containment Spray mode of RHR given a valve interlock
failure.2.9*234000 Fuel Handling Equipment CFR 01Determine the affects to fuel handling operations given a Refueling Bridge interlock
failure.3.7** SRO Only Questions PAGE 3 TOTAL TIER 2 GROUP 210010003101 PAGE TOTAL # QUESTIONS7 PAGE 1 TOTAL TIER 2 GROUP 210011000001 PAGE TOTAL # QUESTIONS4 PAGE 2 TOTAL TIER 2 GROUP 220100000001 PAGE TOTAL # QUESTIONS4 TIER 2 GROUP 2 TOTALS401210031021 5
ES-401Generic Knowledge and Abilities Outline (Tier 3)Form-4 Facility:
Grand Gulf Nuclear StationDate of Exam:
12 August 2005CategoryK/ A#TopicSROSRO-OnlyIR#IR#2.1.19Given plant conditions and the PDS computer, determine necessary actions based on PBDS counts.3.02.1.25Given plant conditions and EP3 graphs, determine the correct mitigation strategy.3.11.2.1.29Determine the correct locking device color coding for locked components. 3.3Conduct2.1.2Given conditions, determine when an act of sabotage or tampering should be suspected.4.0Of Operations2.1.24Given the need to generate a protective tagging clearance, discuss any procedural guidance for
use of electrical and mechanical drawings.
3.1Subtotal 3 22.2.1Given plant conditions, determine proper operation of the IRMs.3.62.2.30Discuss the duties of the operator assigned to communicate with the refueling floor SRO during core alterations.
3.32.2.2.19Describe the process for generating a maintenance work request.3.1Equipment Control2.2.16Determine who is responsible for reviewing the installation and removal of temporary alterations.
2.6Subtotal 2 22.3.1Given the need to enter a high radiation area, determine the allowed time in the area to prevent exceeding the administrative exposure limits.
3.02.3.4Given plant conditions and applicable Emergency Planning Procedures, determine the radiation exposure limits that are in effect.
3.1 3.Radiation Control2.3.6Given liquid radwaste batch release data, determine which does not require Operations approval or a discharge permit.
3.1Subtotal 2 12.4.20Given plant conditions, determine the bases for any applicable EOP cautions.4.02.4.25Given plant conditions including a fire, determine the proper response.3.44.2.4.43Given plant conditions and Emergency Planning Procedures, determine the available emergency communications systems.
3.5Emergency Procedures /2.4.47Given plant conditions and indications from the recirculation pump shaft seals, analyze the condition and determine the probable failure mechanism.
3.7Plan2.4.44Given plant conditions that warrant a General Emergency, determine the correct protective action recommendations.
4.0Subtotal 3 2Tier 3 Point Total 10 7 Appendix DScenario OutlineForm ES-D-1 Facility:
GRAND GULF NUCLEAR STATION Scenario No.:
1 Op-Test No.: Day 1Examiners: _________________________Operators:__________________________
_________________________ __________________________
_________________________ __________________________
Objectives:
To evaluate the candidates' ability to operate the facility in response to the following evolutions:1.Start RCIC for testing per EPI CST to CST.2.Respond to a failure of 1C34-LI-R606C RPV Narrow Range Level 'C' downscale.3.Take actions in response to a Low Pressure Feedwater Heater 3C Tube leak and Failure of the Heater String to Isolate. Complete actions of
the Loss of Feedwater Heating ONEP and Reduction in Recirculation
System Flowrate ONEP.4.Respond to a trip of RCIC.5.Respond to a loss of RPS normal power supply.6.Take actions for a double Recirculation Pump downshift to manually scram the reactor.7.Take actions per the EOPs in response to an ATWS and mitigate the consequences of the ATWS with Main Steam Bypass Valves.8.Respond to a failure of Division II ECCS to manually initiate via the Manual Initiation pushbutton.
Initial Conditions:
Reactor Power is at 100 %.
INOPERABLE EquipmentSRMs 'E' & 'F' are INOP APRM 'H' is INOP due to a failed power supply card.
LPCS Pump is tagged out of service for motor oil replacement.
ESF Transformer 12 is tagged out of service Entergy - Mississippi maintenance.
Appropriate clearances and LCOs are written.
Turnover: The plant is operating at 100% power. Operate RCIC CST to CST at rated flow per a controlled startup in the EPI to allow taking of engineering data with RCIC operating 800 gpm at 1000 psig Standby Service Water 'A' is operating. Containment Ventilation is operating in High Volume Purge.
There are scattered thundershowers reported in the Tensas Parish area.
Event No.Malf. No.Event Type*Event Description1N (BO P)Start RCIC and operate CST to CST per EPI.(EPI 04-1-03-E51-2)
Appendix DScenario OutlineForm ES-D-1 Scenario 1 Day 1 (Continued)
Event No.Malf. No.Event Type*Event Description 2 1 fw126c@ 0 TS (SS)Respond to RPV Narrow Range Level 'C' instrument failure downscale. Complete Technical Specification determination.
3 2 fw232i @
50% ramp to 80%R (R O)Respond to a tube failure in LP FW Heater 3C.
Perform actions per ONEP 05-1-02-V-5 and ONEP 05-
1-02-III-3. Lower Reactor power with Recirc flow.
C (BO P)With a failure to isolate the Condensate System.
Perform actions per ARI 04-1-02-1H13-P870 6A-B3 to
isolate LP Feedwater Heater String 'C'.
4 3 e51047 C (BO P)
TS (SS)RCIC Turbine Trip. Complete Technical Specification determination.
5 4 c71077b C (R O/
BOP)Respond to a RPS 'B' Motor Generator EPA Breaker Trip per the ONEP 05-1-02-III-2.
6 5 fw201; c71076 C (R O)Respond to a double Reactor Recirculation Pump down shift, Automatic RPS actuation fails requiring insertion
of a manual Reactor Scram.
7 6 c11164 @
0.2%M (AL L)Upon Reactor Scram recognize the failure of all control rods to fully insert and take actions per EOPs for ATWS
with Main Steam Bypass Valves.
7 di_1e12 m617@ NORM I (BO P)Upon orders to initiate and override Low Pressure ECCS, recognize the failure of Division II to initiate via
Manual Initiation pushbutton. Take actions upon
automatic initiation to override Division II Low Pressure
ECCS. * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Critical TasksTerminate and prevent injection from Feedwater and ECCS when conditions require entry into Level/Power Control.Commence injection into the reactor using Feedwater or RHR 'A' or 'B' through Shutdown Cooling to restore and maintain level > -192 inches.Insert Control Rods in response to ATWS conditions.
Appendix DScenario OutlineForm ES-D-1 Facility:
GRAND GULF NUCLEAR STATION Scenario No.:
2 Op-Test No.: Day 1Examiners: _________________________Operators:__________________________
_________________________ __________________________
_________________________ __________________________Objectives:
To evaluate the candidates' ability to operate the facility in response to the following evolutions:1. Start SSW 'A' in support of chemical addition.2. Raise Reactor Power by withdrawing control rods. Respond to single control rod drift per ONEP 05-1-02-IV-1.3. Respond to ESF Transformer 21 trouble and subsequent trip with a failure of DG 12 to start.4. Take actions to mitigate a large break failure of Feedwater piping in the Drywell per EOPs. (LOCA is NOT severe enough to result in depressurization of RPV.)5. Respond to a failure of Division 1 ECCS to automatically initiate on High Drywell Pressure.6. Respond to a failure of High Pressure Core Spray to inject. (LOCA with degraded high pressure sources.)
Initial Conditions:
Reactor Power is at 45 %. Plant startup is in progress following an outage.
Reactor Recirculation pumps in Fast Speed; a single Reactor Feed Pump in Three element Master
Level Control; both Heater Drain Pumps are pumping forward.
INOPERABLE EquipmentSRMs 'E' & 'F' are INOP and bypassed.
APRM 'H' is INOP due to a failed power supply card.
LPCS Pump is tagged out of service for pump seal replacement.
ESF 12 Transformer is tagged out of service for maintenance.
Appropriate clearances and LCOs are written.
Turnover: Chemistry requires SSW 'A' in operation to support a chemical addition. Continue plant startup per IOI-2. There are scattered thunder showers reported in the Tensas Parish area.
Event No.Malf. No.Event Type*Event Description1N (BO P)Place Standby Service Water 'A' in service for chemical addition. (EPI 04-1-03-P41-1)2R (RO)Raise Reactor power using control rods to between 40 and 45%. (Control Rod Pull Sheet) 3 1 z161161_24_17 C (RO)TS (SS)Respond to single control rod drift taking actions to insert the control rod. (ONEP 05-1-02-IV-1)
Disarm Control Rod. Complete Technical Specification determination.
Appendix DScenario OutlineForm ES-D-1 Scenario 2 Day 1 (Continued)
Event No.Malf. No.Event Type*Event Description 4 2 p807_4a_f_2 ON r21180 n41141b C (BO P)
TS (SS)Respond to trouble and trip of ESF Transformer 21 with a failure of DG 12 to Start. Complete Technical Specification determination.(ONEP 05-1-01-I-4) 5 3 fw0171b @ 70%rr063b @ 1%
ramp to 4%
M (ALL)Respond to indications of large break LOCA on Feedwater Line 'B' per EOPs. (B21-F065B will close if attempted.)
4 rr040e@ 0 rr041e @ 83%I (BOP)Respond to a failure of Division 1 ECCS to automatically initiate on High Drywell Pressure.
5 e22159a@0 C (BO P)Respond to a failure of High Pressure Core Spray to inject.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Critical Tasks-Recognize failure of Division 1 to initiate and manually initiate Division 1.-Isolate the failed Feedwater line and re-establish Condensate/Feedwater or when RPV level reaches -160 inches wide range, Emergency Depressurizes the
RPV to allow injection from Low Pressure systems (if level cannot be restored
and maintained above -192 inches).
Appendix DScenario OutlineForm ES-D-1 Facility:
GRAND GULF NUCLEAR STATION Scenario No.:
3 Op-Test No.:
BACK UPExaminers: _________________________Operators:__________________________
_________________________ __________________________
_________________________ __________________________
Objectives:
To evaluate the candidates' ability to operate the facility in response to the following evolutions:1. Start 2 nd Condensate Pump and Condensate Booster Pump.2. Raise Reactor Power/ Pressure by withdrawing control rods.3. Respond to a stuck control rod.4. Respond to a trip of LCC 15BA3.5. Respond to an automatic and manual scram function failure; ATWS ARI/RPT will insert control rods with three control rods stuck withdrawn.6. Recognize the failure of MSIVs to completely isolate and take actions to isolate the Main Steam Lines.7. Recognize and respond to a steam leak in the Auxiliary Building Steam Tunnel.
Take actions for mitigation of the leak with a failure of the MSIVs to fully isolate.8. Take actions per the EOPs in response to three stuck control rods following a Reactor Scram.
Initial Conditions:
Reactor Power is at 1 % plant heatup and pressurization is in progress.
The Reactor is 400 psig with 1 Condensate and Condensate Booster Pump in service on Startup Level Control. Step 80 of the Control Rod Movement Sequence.
INOPERABLE Equipment APRM 'H' is INOP due to a failed power supply card LPCS Pump is tagged out of service for motor oil replacement and will be returned to
service in two (2) hours.
ESF-12 Transformer is tagged out of service for Entergy - Mississippi maintenance.
Appropriate clearances and LCOs are written.
Turnover: Continue power ascension. Ready to Start second Condensate and Condensate Booster Pump. There are scattered thundershowers reported in the Tensas Parish area.
Appendix DScenario OutlineForm ES-D-1 Scenario 3 Backup (Continued)
Event No.Malf. No.Event Type*Event Description1N (RO)Start 2nd Condensate and Condensate Booster Pumps.(SOI 04-1-01-N19-1)2R (RO)Raise reactor power and pressure by withdrawing control rods.(IOI 03-1-01-1 and Control Rod Movement Sheet) 3 1 z022022 _40_45 C (RO/ BOP)
TS (SS)Respond to a stuck control rod during withdrawal.(ONEP 05-1-02-IV-1)
Complete Technical Specification determination.
4 2 r21142t C (BOP/ RO)Respond to a trip of Load Control Center 15BA3.(ONEP 05-1-02-I-4; 05-1-02-III-5; and SOI 04-1-01-R21-15) 5 3 c71162C (RO)Recognize a failure to scram using RPS and manually scram the reactor using ATWS ARI.
6 4 ms183b ms184b att9I (BOP)Recognize the failure of MSIVs to completely isolate and take actions to isolate the Main Steam Lines.
(ONEP 05-1-02-III-5) 7 5 ms067b @
20% ms066b
@ 0.2% ramp to 10% M (ALL)Recognize and respond to a steam leak in the Auxiliary Building Steam Tunnel. Take actions for mitigation of the leak with a
failure of the MSIVs to fully isolate.
6 z022022 _
36_25_12_09C (RO)Recognize the failure of two additional control rods to fully insert on the Reactor Scram. (Three Rods Out) insert control rods (ONEP 05-1-02-IV-1)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Critical Tasks1 Manually scram the reactor.2 Isolate the main steam lines.