ML053330589

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08-2005 - Initial Final Outline
ML053330589
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 10/03/2005
From: Ryan Lantz
Operations Branch IV
To: Gerald Williams
Entergy Operations
References
50-416/05-301, ES-401-1, NUREG-1021, Rev 9 50-416/05-301
Download: ML053330589 (32)
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BWR EXAMINATION OUTLINE Form ES-401-1 GRAND GULF NUCLEAR EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 STATION (RO/SRO)

E/APE #/NAME/SAFETY K1 K2 K3 A1 A2 G TOPIC(S) IR #

FUNCTION 295001 Partial or Complete 2. Given plant conditions, parameters, and a loss of 1 Loss of Forced Core Flow 4. the recirculation system, determine appropriate 4.0 833 Circulation / 1 & 4 4 actions.

CFR 295003 Partial or Complete 2 Loss of AC Power/ 6 01 3.3 464 CFR 295004 Partial or Complete Given plant conditions and a loss of DC power, 3 Loss of DC Power / 6 02 determine the effect to the SDC system. 3.8 834 CFR 295005 Main Turbine Generator Following a reactor scram and subsequent main 4 Trip / 3 03 turbine generator trip, determine the effects of 3.5 835 CFR manual bypass valve operation on reactor water level.

295006 SCRAM / 1 Given plant conditions following a reactor scram, 5 CFR 02 determine if adequate shutdown margin exists. 3.4 836 295016 Control Room Describe the method used to manually scram the 6 Abandonment / 7 01 reactor after the control room has been abandoned. 3.8 837 CFR 295018 Partial or Complete Given plant conditions and a partial loss of 7 Loss of CCW / 8 01 Component Cooling Water, determine the 838 CFR necessary actions to ensure the plant 3.5 remains/returns to a safe condition.

295019 Partial or Complete Given indications of a partial loss of Instrument 8 Loss of Inst. Air / 8 01 Air determine a method to restore Instrument Air 3.5 548 CFR system pressure.

295021 Loss of Shutdown Given specific plant conditions following a loss of 9 Cooling / 4 01 Shutdown Cooling, determine the reason for 3.3 078 CFR raising reactor water level. a 295023 Refueling Accidents / 8 Determine the correct operator response to 10 CFR 03 inadvertent criticality following a refueling 3.7 848 accident.

295024 High Drywell Pressure / 2. Given plant conditions and high drywell pressure, 11 5 1. determine the method to lower drywell pressure. 3.9 849 CFR 23 295025 High Reactor Pressure / 12 3 06 4.2 690 CFR 295026 Suppression Pool High Given an ATWS condition, describe the EP bases 13 Water Temp. / 5 01 for lowering reactor pressure as Suppression Pool 3.8 840 CFR temperature rises.

295027 High Containment Given rising Containment temperature, describe 14 Temperature / 5 03 the necessary actions to maintain the 3.5 844 CFR plant/containment in a safe condition.

PAGE 1 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 1 4 0 4 3 1 2 14 REVISION 2 7/29/2005 PAGE 1 OF 11 NUREG 1021, REVISION 9

BWR EXAMINATION OUTLINE Form ES-401-1 GRAND GULF NUCLEAR EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 STATION (RO/SRO)

E/APE #/NAME/SAFETY K1 K2 K3 A1 A2 G TOPIC(S) IMP #

FUNCTION 295028 High Drywell Given plant conditions and elevated drywell 15 Temperature / 5 02 temperature, determine the effects to control 2.9 845 CFR room reactor water level indication.

295030 Low Suppression Pool 2. Given a low suppression pool level condition, 16 Water Level / 5 2. determine the effects to other plant systems. 846 CFR 12 3.0 295031 Reactor Low Water Given plant conditions, describe the operation of 17 Level / 2 04 the High Pressure Core Spray system following 4.3 847 CFR a LOCA.

295037 SCRAM Condition Given plant conditions and an ATWS condition, 18 Present and Reactor Power determine the availability of the main condenser 850 Above APRM Downscale or 06 as a heat sink. 3.8 Unknown / 1 CFR 295038 High Offsite Release Given a radioactive release from the plant, 19 Rate / 9 01 determine when it is considered to be offsite. 3.3 851 CFR 600000 Plant Fire On Site / 8 04 Determine the required procedural actions for a 20 fire on the plant site. 2.8 852 PAGE 2 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 1 1 0 2 1 1 1 6 PAGE 1 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 1 4 0 4 3 1 2 14 TIER 1 GROUP 1 TOTALS 5 0 6 4 2 3 20 REVISION 2 7/29/2005 PAGE 2 OF 11 NUREG 1021, REVISION 9

BWR EXAMINATION OUTLINE Form ES-401-1 GRAND GULF NUCLEAR EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 STATION (RO/SRO)

E/APE #/NAME/SAFETY K1 K2 K3 A1 A2 G TOPIC(S) IMP #

FUNCTION 295002 Loss of Main Condenser Given plant conditions and degrading main 21 Vacuum / 3 01 condenser vacuum, determine the automatic 3.5 854 CFR plant response (RPS actuation).

295007 High Reactor Pressure / 2. Determine the conditions necessary to require 22 3 4. connection of an alternate air source to the 3.3 855 CFR 35 SRVs.

295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure /

5 295011 High Containment Temperature / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Describe the preferred method to minimize 23 Water Temp. / 5 02 localized suppression pool heating when using 856 CFR the SRVs to control reactor pressure without 3.2 suppression cooling in service.

295014 Inadvertent Reactivity Addition / 1 295015 Incomplete SCRAM / 1 295017 High Offsite Release Rate / 9 295020 Inadvertent Cont.

Isolation / 5 & 7 295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Water Level / 5 PAGE 1 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 2 0 1 0 0 1 1 3 REVISION 2 7/29/2005 PAGE 3 OF 11 NUREG 1021, REVISION 9

BWR EXAMINATION OUTLINE Form ES-401-1 GRAND GULF NUCLEAR EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 STATION (RO/SRO)

E/APE #/NAME/SAFETY K1 K2 K3 A1 A2 G TOPIC(S) IMP #

FUNCTION 295032 High Secondary Given plant conditions including elevated 24 Containment Area Temperature / Auxiliary Building temperatures, describe the 857 5 02 conditions that would require a reactor scram. 3.5 CFR 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Given plant conditions including elevated 25 Ventilation High Radiation / 9 03 Auxiliary Building radiation levels, describe the 858 CFR conditions that would automatically start the 4.3 Standby Gas Treatment system.

295035 Secondary Containment Given accident conditions and a Standby Gas 26 High Differential Pressure / 5 02 Treatment system failure, determine the type of 859 CFR release.

295036 Secondary Containment Describe the system logic used by the Auxiliary 27 High Sump/Area Water Level / 5 01 Building Floor Drain system to contain a 3.2 860 CFR significant CCW system rupture.

500000 High CTMT Hydrogen Conc. / 5 PAGE 2 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 2 1 1 1 1 0 0 4 PAGE 1 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 2 0 1 0 0 1 1 3 TIER 1 GROUP 2 TOTALS 1 2 1 1 1 1 7 REVISION 2 7/29/2005 PAGE 4 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)

SYSTEM #/NAME K1 K2 K3 K4 K K A A A A G TOPIC(S) IMP #

5 6 1 2 3 4 203000 RHR/LPCI: Given plant conditions, 28 Injection Mode 02 describe the design 3.5 861 CFR features and limits of the RHR pump manual override feature.

205000 Shutdown Describe the RHR 29 Cooling 04 Shutdown Cooling 2.6 862 CFR system NPSH interlocks.

206000 HPCI N/A GGNS 207000 Isolation N/A GGNS (Emergency) Condenser 209001 LPCS Given degraded plant 30 CFR 01 conditions during a 863 LOCA, describe LPCS 3.8 manual operation.

209002 HPCS Describe available 31 CFR 09 methods to raise/lower 864 suppression pool level 3.4 using HPCS.

209002 HPCS 2. Describe the bases for 32 CFR 1. the HPCS injection 865 2 valve high reactor water 3.2 8 level interlock.

211000 SLC Predict the SLC system 33 CFR 02 indication and response 866 with indication the squib 3.6 valve failed to actuate and follow up actions.

212000 RPS Given plant conditions 34 CFR 12 including a partial main 867 turbine stop/control 4.0 valve closure, determine the effect to RPS.

215003 IRM Describe the reason for 35 CFR 03 the precaution 868 concerning driving IRMs during 3.0 surveillance activities.

215004 Source Range 2. Describe the SRM 36 Monitor 2. precaution warning of a 869 CFR 3 potential control rod 2.5 3 block even if the channel is bypassed.

215005 APRM / Given a partial loss of 37 LPRM 02 plant electrical power, 870 CFR determine the effect to 2.6 the APRMs.

217000 RCIC Predict how a reactor 38 CFR 02 pressure change will 3.3 871 affect RCIC system flow.

PAGE 1 TOTAL PAGE TOTAL #

REVISION 2 7/29/2005 PAGE 5 OF 11 NUREG 1021, REVISION 9

TIER 2 GROUP 1 0 1 0 1 2 0 1 2 0 2 2 QUESTIONS 11 REVISION 2 7/29/2005 PAGE 6 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)

SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP #

1 2 3 4 5 6 1 2 3 4 218000 ADS Describe the relationship 39 CFR 01 between ADS Logic 3.1 872 power and the operation of the ADS logic.

223002 PCIS / Nuclear Determine the operator 40 Steam Supply Shutoff 03 actions required to 873 CFR mitigate a NSSSS logic 3.0 failure.

239002 SRVs Describe the design 41 CFR 09 features available to 874 determine if a SRV is 3.7 open.

259002 Reactor Water Describe the operator 42 Level Control 04 response to a failure of 875 CFR RFPT speed control with 3.0 speed rising.

259002 Reactor Water Describe prerequisites 43 Level Control 06 for transferring the 233 CFR Feedwater system to 3- 3.1 a element control.

261000 SGTS Describe the SGTS 44 CFR 03 damper logic following 3.0 876 system initiation.

262001 AC Electrical Given plant conditions 45 Distribution 01 and a partial loss of DC 877 CFR power, determine the 3.1 affect to the AC distribution system.

262002 UPS (AC/DC) Given plant conditions 46 CFR 01 and degraded AC power, 3.1 878 determine the status of plant inverters.

263000 DC Electrical Given a loss of AC 47 Distribution 01 power to battery 879 CFR chargers, determine the affects to the DC distribution system.

264000 EDGs Describe EDG response 48 CFR 10 to a LOCA. 3.9 880 264000 EDGs 2. Determine EDG status 49 CFR 4. from control room 881 4 alarms and indications 8 and any required 3.5 operator actions to improve plant conditions.

300000 Instrument Air Determine the effect on 50 CFR 01 the plant given a loss of 882 Instrument Air to the 2.7 containment.

PAGE 2 TOTAL PAGE TOTAL #

TIER 2 GROUP 1 0 1 1 3 0 1 0 3 1 1 1 QUESTIONS 12 REVISION 2 7/29/2005 PAGE 7 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)

SYSTEM #/NAME K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC(S) IMP #

300000 Instrument Air Determine the affect of a 51 CFR 13 clogged filter on the 2.8 883 Instrument Air system.

400000 Component Determine the method 52 Cooling Water 04 used to confirm a reactor 884 CFR coolant leak into the CCW 2.9 system.

400000 Component Determine the affect to the 2.8 53 Cooling Water 02 plant if the CCW 885 CFR temperature control fails.

PAGE 3 TOTAL PAGE TOTAL #

TIER 2 GROUP 1 1 0 0 0 0 1 1 0 0 0 0 QUESTIONS 3 PAGE 1 TOTAL PAGE TOTAL #

TIER 2 GROUP 1 0 1 0 1 2 0 1 2 0 2 2 QUESTIONS 11 PAGE 2 TOTAL PAGE TOTAL #

TIER 2 GROUP 1 0 1 1 3 0 1 0 3 1 1 1 QUESTIONS 12 TIER 2GROUP 1 TOTALS 1 2 1 4 2 2 2 5 1 3 3 26 REVISION 2 7/29/2005 PAGE 8 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)

SYSTEM #/NAME K1 K K K K K A A A A G TOPIC(S) IMP #

2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic CFR 201002 RMCS N/A GGNS 201003 Control Rod and Drive Mechanism CFR 201004 RSCS N/A GGNS 201005 RCIS Describe the purpose for 54 CFR 10 the rod withdrawal 3.2 886 limiter.

201006 RWM N/A GGNS 202001 Recirculation 2. 55 CFR 4. 887 1 3.4 1

202002 Recirculation Given plant conditions, 56 Flow Control 01 determine any automatic 888 CFR41.6 actions associated with 3.5 the Recirculation System HPUs.

204000 RWCU Determine the correct 57 CFR 06 flow path to use RWCU 889 as an alternate shutdown 2.6 cooling.

214000 RPIS N/A GGNS 215001 Traversing In-Core Probe CFR 215002 RBM N/A GGNS PAGE 1 TOTAL PAGE TOTAL #

TIER 2 GROUP 2 1 0 0 1 1 0 0 0 0 0 1 QUESTIONS 4 REVISION 2 7/29/2005 PAGE 9 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)

SYSTEM #/NAME K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC(S) IMP #

216000 Nuclear Boiler Instrumentation CFR 219000 RHR /LPCI Suppression Pool Cooling Mode CFR 223001 Primary CTMT Determine the 58 and Auxiliaries 08 limitations to 890 CFR SRV usage given a reduced 3.6 suppression pool level.

226001 RHR/LPCI:

CTMT Spray Mode CFR 230000 RHR/LPCI: N/A GGNS Torus/Pool Spray Mode 233000 Fuel Pool Cooling and Cleanup CFR 234000 Fuel Handling Equipment CFR 239001 Main and Reheat Given plant 59 Steam 04 conditions 891 CFR including a MSIV closure, 2.8 determine the affect to the Offgas system.

239003 MSIV Leakage Explain the 60 Control 02 relationship 892 CFR between the MSIV Leakage 2.9 Control system and SGTS.

241000 Reactor/Turbine 2. Describe the 61 Pressure Regulator 4. bases for each of 893 CFR 6 the Scram ONEP 3.1 immediate actions.

PAGE 2 TOTAL PAGE TOTAL #

TIER 2 GROUP 2 2 0 1 0 0 0 0 0 0 0 1 QUESTIONS 4 REVISION 2 7/29/2005 PAGE 10 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)

SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP #

1 2 3 4 5 6 1 2 3 4 245000 Main Turbine Determine main turbine 62 Gen./Aux. 02 critical speeds as it is 2.8 894 CFR rolled to rated speed.

256000 Reactor Condensate CFR 259001 Reactor Feedwater Determine necessary 63 CFR 03 actions and priorities 895 immediately after a single condensate pump trips 3.6 with the plant at rated conditions.

268000 Radwaste Determine the Drywell 64 CFR 04 Floor Drains indications 896 available to detect drywell 2.7 general area leakage.

271000 Offgas CFR 272000 Radiation Monitoring CFR 286000 Fire Protection CFR 288000 Plant Ventilation CFR 290001 Secondary CTMT Determine inputs to the 65 CFR 03 Fuel Pool leak detection 2.8 897 standpipe.

290003 Control Room HVAC CFR 290002 Reactor Vessel Internals CFR PAGE 3 TOTAL PAGE TOTAL #

TIER 2 GROUP 2 1 0 0 1 0 0 0 1 1 0 0 QUESTIONS 4 PAGE 1 TOTAL PAGE TOTAL #

TIER 2 GROUP 2 1 0 0 1 1 0 0 0 0 0 1 QUESTIONS 4 PAGE 2 TOTAL PAGE TOTAL #

TIER 2 GROUP 2 2 0 1 0 0 0 0 0 0 0 1 QUESTIONS 4 TIER 2 GROUP 2 TOTALS 4 0 1 2 1 0 0 1 1 0 2 12 REVISION 2 7/29/2005 PAGE 11 OF 11 NUREG 1021, REVISION 9

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form-401-3 Facility: Grand Gulf Nuclear Station Date of Exam: 12 August 2005 Category K/ A# Topic RO SRO-Only IR # IR #

2.1.19 Given plant conditions and the PDS computer, 66 determine necessary actions based on PBDS 3.0 898 counts.

2.1.25 Given plant conditions and EOP-3 graphs, 67 determine the correct mitigation strategy. 2.8 899

1. 2.1.29 Determine the correct locking device color 68 coding for locked components. 3.4 237a Conduct 2.1 Of Operations 2.1 2.1 Subtotal 3 2.2.1 Given plant conditions, determine proper 69 operation of the IRMs. 3.7 900 2.2.30 Discuss the duties of the operator assigned to 70 communicate with the refueling floor SRO 3.5 901 during core alterations.
2. 2.2 Equipment 2.2 Control 2.2 2.2 Subtotal 2 2.3.1 Given the need to enter a high radiation area, 71 determine the allowed time in the area to prevent 2.6 902 exceeding the administrative exposure limits.

2.3.4 Given plant conditions and applicable 72 Emergency Planning Procedures, determine the 2.5 903 radiation exposure limits that are in effect.

3. 2.3 Radiation 2.3 Control 2.3 2.3 Subtotal 2 2.4.20 Given plant conditions, determine the bases for 73 any applicable EOP cautions. 3.3 904 2.4.25 Given plant conditions including a fire, 74 determine the proper response. 2.9 905
4. 2.4.43 Given plant conditions and Emergency Plan 75 Procedures, determine the available emergency 2.8 906 communications systems.

Emergency 2.4 Procedures / 2.4 Plan 2.4 Subtotal 3 Tier 3 Point Total 10 7 REVISION 2 7/29/2005 PAGE 12 OF 11 NUREG 1021, REVISION 9

BWR EXAMINATION OUTLINE Form ES-401-1 GRAND GULF NUCLEAR EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 STATION (RO/SRO)

E/APE #/NAME/SAFETY K1 K2 K3 A1 A2 G TOPIC(S) IR #

FUNCTION 295001 Partial or Complete 2. Given plant conditions, parameters, and a loss of 1 Loss of Forced Core Flow 4. the recirculation system, determine appropriate 833 Circulation / 1 & 4 4 actions. 4.3 CFR 295003 Partial or Complete 2 Loss of AC Power/ 6 01 3.5 464 CFR 295004 Partial or Complete Given plant conditions and a loss of DC power, 3 Loss of DC Power / 6 02 determine the effect to the SDC system. 4.1 834 CFR 295005 Main Turbine Generator Following a reactor scram and subsequent main 4 Trip / 3 03 turbine generator trip, determine the effects of 835 CFR manual bypass valve operation on reactor water 3.7 level.

295006 SCRAM / 1 Given plant conditions following a reactor scram, 5 CFR 02 determine if adequate shutdown margin exists. 3.7 836 295016 Control Room Describe the method used to manually scram the 6 Abandonment / 7 01 reactor after the control room has been abandoned. 3.9 837 CFR 295018 Partial or Complete Given plant conditions and a partial loss of 7 Loss of CCW / 8 01 Component Cooling Water, determine the 838 CFR necessary actions to ensure the plant 3.6 remains/returns to a safe condition.

295019 Partial or Complete Given indications of a partial loss of Instrument 8 Loss of Inst. Air / 8 01 Air determine a method to restore Instrument Air 3.6 548 CFR system pressure.

295021 Loss of Shutdown Given specific plant conditions following a loss of 9 Cooling / 4 01 Shutdown Cooling, determine the reason for 3.4 078 CFR raising reactor water level. a 295023 Refueling Accidents / 8 Determine the correct operator response to 10 CFR 03 inadvertent criticality following a refueling 4.0 848 accident.

295024 High Drywell Pressure / 2. Given plant conditions and high drywell pressure, 11 5 1. determine the method to lower drywell pressure. 4.0 849 CFR 23 295025 High Reactor Pressure / 12 3 06 4.4 690 CFR 295026 Suppression Pool High Given an ATWS condition, describe the EP bases 13 Water Temp. / 5 01 for lowering reactor pressure as Suppression Pool 4.1 840 CFR temperature rises.

295027 High Containment Given rising Containment temperature, describe 14 Temperature / 5 03 the necessary actions to maintain the 3.8 844 CFR plant/containment in a safe condition.

PAGE 1 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 1 4 0 4 3 1 2 14 REVISION 2 7/29/2005 PAGE 1 OF 11 NUREG 1021, REVISION 9

BWR EXAMINATION OUTLINE Form ES-401-1 GRAND GULF NUCLEAR EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 STATION (RO/SRO)

E/APE #/NAME/SAFETY K1 K2 K3 A1 A2 G TOPIC(S) IMP #

FUNCTION 295028 High Drywell Given plant conditions and elevated drywell 15 Temperature / 5 02 temperature, determine the effects to control 3.1 845 CFR room reactor water level indication.

295030 Low Suppression Pool 2. Given a low suppression pool level condition, 16 Water Level / 5 2. determine the effects to other plant systems. 846 CFR 12 3.4 295031 Reactor Low Water Given plant conditions, describe the operation of 17 Level / 2 04 the High Pressure Core Spray system following 4.2 847 CFR a LOCA.

295037 SCRAM Condition Given plant conditions and an ATWS condition, 18 Present and Reactor Power determine the availability of the main condenser 850 Above APRM Downscale or 06 as a heat sink. 4.1 Unknown / 1 CFR 295038 High Offsite Release Given a radioactive release from the plant, 19 Rate / 9 01 determine when it is considered to be offsite. 4.3 851 CFR 600000 Plant Fire On Site / 8 04 Determine the required procedural actions for a 20 fire on the plant site. 3.4 852 295004 Partial or Complete Given a loss of Division 1 DC logic power, 76 Loss of DC Power / 6 02 determine the affect to the Division 1 ECCS. 3.9* 908 CFR 295005 Main Turbine 2. Given plant data including current area dose 77 Generator Trip / 3 3. rates, determine the required personnel 909 CFR 5 monitoring equipment needed to enter the main 2.5*

turbine/generator area to investigate the cause for a trip.

295026 Suppression Pool Given plant conditions including rising 78 High Water Temp. / 5 03 Suppression Pool temperature, interpret HCTL 4.0* 910 CFR and determine appropriate actions.

295027 High Containment 2. Explain the bases for the Technical Specification 79 Temperature / 5 2. Containment average air temperature limit. 4.1* 911 CFR 22 295030 Low Suppression Pool Given low suppression pool water level, 80 Water Level / 5 02 determine if suppression pool temperature 3.9* 912 CFR can/cannot be measured and why.

295038 High Offsite Release 2. Given a severe case fuel handling accident, 81 Rate / 9 2. explain the processes designed to prevent high 3.5* 913 CFR 28 offsite release rates.

600000 Plant Fire On Site / 8 16 Describe the basis for separating vital 82 equipment from the Main Control Room during 3.5* 914 a fire in the Main Control Room.

  • SRO Only Questions PAGE 2 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 1 1 0 2 1 5 4 13 PAGE 1 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 1 4 0 4 3 1 2 14 TIER 1 GROUP 1 TOTALS 5 0 6 4 6 6 27 REVISION 2 7/29/2005 PAGE 2 OF 11 NUREG 1021, REVISION 9

BWR EXAMINATION OUTLINE Form ES-401-1 GRAND GULF NUCLEAR EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 STATION (RO/SRO)

E/APE #/NAME/SAFETY K1 K2 K3 A1 A2 G TOPIC(S) IMP #

FUNCTION 295002 Loss of Main Condenser Given plant conditions and degrading main 21 Vacuum / 3 01 condenser vacuum, determine the automatic 3.5 854 CFR plant response (RPS actuation).

295007 High Reactor Pressure / 2. Determine the conditions necessary to require 22 3 4. connection of an alternate air source to the 3.5 855 CFR 35 SRVs.

295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure /

5 295011 High Containment Temperature / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Describe the preferred method to minimize 23 Water Temp. / 5 02 localized suppression pool heating when using 856 CFR the SRVs to control reactor pressure without 3.5 suppression cooling in service.

295014 Inadvertent Reactivity Addition / 1 295015 Incomplete SCRAM / 1 295017 High Offsite Release Rate / 9 295020 Inadvertent Cont.

Isolation / 5 & 7 295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Water Level / 5 PAGE 1 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 2 0 1 0 0 1 1 3 REVISION 2 7/29/2005 PAGE 3 OF 11 NUREG 1021, REVISION 9

BWR EXAMINATION OUTLINE Form ES-401-1 GRAND GULF NUCLEAR EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 STATION (RO/SRO)

E/APE #/NAME/SAFETY K1 K2 K3 A1 A2 G TOPIC(S) IMP #

FUNCTION 295032 High Secondary Given plant conditions including elevated 24 Containment Area Temperature / Auxiliary Building temperatures, describe the 857 5 02 conditions that would require a reactor scram. 3.8 CFR 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Given plant conditions including elevated 25 Ventilation High Radiation / 9 03 Auxiliary Building radiation levels, describe the 858 CFR conditions that would automatically start the 4.5 Standby Gas Treatment system.

295035 Secondary Containment Given accident conditions and a Standby Gas 26 High Differential Pressure / 5 02 Treatment system failure, determine the type of 4.2 859 CFR release.

295036 Secondary Containment Describe the system logic used by the Auxiliary 27 High Sump/Area Water Level / 5 01 Building Floor Drain system to contain a 3.3 860 CFR significant CCW system rupture.

500000 High CTMT Hydrogen Conc. / 5 295011 High Containment Given LOCA conditions, determine when 83 Temperature / 5 01 containment spray should be initiated. 3.9* 915 CFR 295014 Inadvertent Reactivity 2. Given a control rod drifting out with the plant at 84 Addition / 1 1. power, determine any necessary notifications. 3.3* 916 CFR 14 295020 Inadvertent Cont. Given a partial MSIV closure, determine the 85 Isolation / 5 & 7 03 affect on reactor power. 3.7* 917 CFR

  • SRO Only Questions PAGE 2 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 2 1 1 1 1 2 1 7 PAGE 1 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 2 0 1 0 0 1 1 3 TIER 1 GROUP 2 TOTALS 1 2 1 1 3 2 10 REVISION 2 7/29/2005 PAGE 4 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)

SYSTEM #/NAME K1 K2 K3 K4 K K A A A A G TOPIC(S) IMP #

5 6 1 2 3 4 203000 RHR/LPCI: Given plant conditions, 28 Injection Mode 02 describe the design 861 CFR features and limits of the 3.7 RHR pump manual override feature.

205000 Shutdown Describe the RHR 29 Cooling 04 Shutdown Cooling 2.6 862 CFR system NPSH interlocks.

206000 HPCI N/A GGNS 207000 Isolation N/A GGNS (Emergency) Condenser 209001 LPCS Given degraded plant 30 CFR 01 conditions during a 863 LOCA, describe LPCS 3.6 manual operation.

209002 HPCS Describe available 31 CFR 09 methods to raise/lower 864 suppression pool level 3.5 using HPCS.

209002 HPCS 2. Describe the bases for 32 CFR 1. the HPCS injection 865 2 valve high reactor water 3.3 8 level interlock.

211000 SLC Predict the SLC system 33 CFR 02 indication and response 866 with indication the squib 3.9 valve failed to actuate and follow up actions.

212000 RPS Given plant conditions 34 CFR 12 including a partial main 867 turbine stop/control 4.1 valve closure, determine the effect to RPS.

215003 IRM Describe the reason for 35 CFR 03 the precaution 868 concerning driving 3.1 IRMs during surveillance activities.

215004 Source Range 2. Describe the SRM 36 Monitor 2. precaution warning of a 869 CFR 3 potential control rod 2.9 3 block even if the channel is bypassed.

215005 APRM / Given a partial loss of 37 LPRM 02 plant electrical power, 870 CFR determine the effect to 2.8 the APRMs.

217000 RCIC Predict how a reactor 38 CFR 02 pressure change will 3.3 871 affect RCIC system flow.

PAGE 1 TOTAL PAGE TOTAL #

TIER 2 GROUP 1 0 1 0 1 2 0 1 2 0 2 2 QUESTIONS 11 REVISION 2 7/29/2005 PAGE 5 OF 11 NUREG 1021, REVISION 9

REVISION 2 7/29/2005 PAGE 6 OF 11 NUREG 1021, REVISION 9 GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)

SYSTEM #/NAME K1 K K K K K A A A A G TOPIC(S) IMP #

2 3 4 5 6 1 2 3 4 218000 ADS Describe the relationship 39 CFR 01 between ADS Logic 3.3 872 power and the operation of the ADS logic.

223002 PCIS / Nuclear Determine the operator 40 Steam Supply Shutoff 03 actions required to 873 CFR mitigate a NSSSS logic 3.3 failure.

239002 SRVs Describe the design 41 CFR 09 features available to 874 determine if a SRV is 3.6 open.

259002 Reactor Water Describe the operator 42 Level Control 04 response to a failure of 875 CFR RFPT speed control with 3.1 speed rising.

259002 Reactor Water Describe prerequisites 43 Level Control 06 for transferring the 233a CFR Feedwater system to 3- 3.2 element control.

261000 SGTS Describe the SGTS 44 CFR 03 damper logic following 2.9 876 system initiation.

262001 AC Electrical Given plant conditions 45 Distribution 01 and a partial loss of DC 877 CFR power, determine the 3.4 affect to the AC distribution system.

262002 UPS (AC/DC) Given plant conditions 46 CFR 01 and degraded AC power, 878 determine the status of plant inverters.

263000 DC Electrical Given a loss of AC 47 Distribution 01 power to battery 879 CFR chargers, determine the 3.4 affects to the DC distribution system.

264000 EDGs Describe EDG response 48 CFR 10 to a LOCA. 4.2 880 264000 EDGs 2. Determine EDG status 49 CFR 4. from control room 881 4 alarms and indications 8 and any required 3.8 operator actions to improve plant conditions.

300000 Instrument Air Determine the effect on 50 CFR 01 the plant given a loss of 882 Instrument Air to the 2.9 containment.

PAGE 2 TOTAL PAGE TOTAL #

TIER 2 GROUP 1 0 1 1 3 0 1 0 3 1 1 1 QUESTIONS 12 REVISION 2 7/29/2005 PAGE 7 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)

SYSTEM #/NAME K1 K K K K K A A A A G TOPIC(S) IMP #

2 3 4 5 6 1 2 3 4 300000 Instrument Air Determine the affect of a 51 CFR 13 clogged filter on the 2.3 883 Instrument Air system.

400000 Component Determine the method used 52 Cooling Water 04 to confirm a reactor 884 CFR coolant leak into the CCW 3.1 system.

400000 Component Determine the affect to the 53 Cooling Water 02 plant if the CCW 2.8 885 CFR temperature control fails.

203000 RHR/LPCI: 2. Given LOCA conditions, 86 Injection Mode 3. determine how LPCI works 918 CFR 1 in conjunction with the 3.2 1 other ECCS to control

  • radiation releases.

209001 LPCS 2. Given a short-term 87 CFR 1. problem associated with 919 1 LPCS that does not affect 5 operability, determine the 3.0 most effective method to

  • provide the information to operations personnel.

215003 IRM 2. Given plant conditions 88 CFR 4. requiring entry into the 920 1 EOPs and the need to 6 insert the IRMs, determine 4.0 the correct procedure

  • hierarchy to accomplish the task.

215004 Source Range 2. Given the applicable Tech 89 Monitor 2. Specs and a repaired SRM 921 CFR 2 detector, determine the 3.5 1 surveillance requirements

  • to ensure operability.

217000 RCIC 2. 90 CFR 1. 922 1 3.9 0 *

  • SRO Only Questions PAGE 3 TOTAL PAGE TOTAL #

TIER 2 GROUP 1 1 0 0 0 0 1 1 0 0 0 5 QUESTIONS 8 PAGE 1 TOTAL PAGE TOTAL #

TIER 2 GROUP 1 0 1 0 1 2 0 1 2 0 2 2 QUESTIONS 11 PAGE 2 TOTAL PAGE TOTAL #

TIER 2 GROUP 1 0 1 1 3 0 1 0 3 1 1 1 QUESTIONS 12 TIER 2GROUP 1 TOTALS 1 2 1 4 2 2 2 5 1 3 8 31 REVISION 2 7/29/2005 PAGE 8 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)

SYSTEM #/NAME K1 K K K K K A A A A G TOPIC(S) IMP #

2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic CFR 201002 RMCS N/A GGNS 201003 Control Rod and Drive Mechanism CFR 201004 RSCS N/A GGNS 201005 RCIS Describe the purpose for 54 CFR 10 the rod withdrawal 3.3 886 limiter.

201006 RWM N/A GGNS 202001 Recirculation 2. 55 CFR 4. 887 1 3.6 1

202002 Recirculation Given plant conditions, 56 Flow Control 01 determine any automatic 888 CFR41.6 actions associated with 3.6 the Recirculation System HPUs.

204000 RWCU Determine the correct 57 CFR 06 flow path to use RWCU 889 as an alternate shutdown 2.8 cooling.

214000 RPIS N/A GGNS 215001 Traversing In-Core Probe CFR 215002 RBM N/A GGNS PAGE 1 TOTAL PAGE TOTAL #

TIER 2 GROUP 2 1 0 0 1 1 0 0 0 0 0 1 QUESTIONS 4 REVISION 2 7/29/2005 PAGE 9 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)

SYSTEM #/NAME K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC(S) IMP #

216000 Nuclear Boiler Instrumentation CFR 219000 RHR /LPCI Suppression Pool Cooling Mode CFR 223001 Primary CTMT Determine the 58 and Auxiliaries 08 limitations to 890 CFR SRV usage given a reduced 3.8 suppression pool level.

226001 RHR/LPCI:

CTMT Spray Mode CFR 230000 RHR/LPCI: N/A GGNS Torus/Pool Spray Mode 233000 Fuel Pool Cooling and Cleanup CFR 234000 Fuel Handling Equipment CFR 239001 Main and Reheat Given plant 59 Steam 04 conditions 891 CFR including a MSIV closure, 2.8 determine the affect to the Offgas system.

239003 MSIV Leakage Explain the 60 Control 02 relationship 892 CFR between the MSIV Leakage 3.0 Control system and SGTS.

241000 Reactor/Turbine 2. Describe the 61 Pressure Regulator 4. bases for each of 893 CFR 6 the Scram ONEP 4.0 immediate actions.

PAGE 2 TOTAL PAGE TOTAL #

TIER 2 GROUP 2 2 0 1 0 0 0 0 0 0 0 1 QUESTIONS 4 REVISION 2 7/29/2005 PAGE 10 OF 11 NUREG 1021, REVISION 9

GRAND GULF BWR EXAMINATION OUTLINE Form ES-401-1 NUCLEAR STATION PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)

SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP #

1 2 3 4 5 6 1 2 3 4 245000 Main Turbine Determine main turbine 62 Gen./Aux. 02 critical speeds as it is rolled 2.8 894 CFR to rated speed.

256000 Reactor Condensate CFR 259001 Reactor Determine necessary actions 63 Feedwater 03 and priorities immediately 895 CFR after a single condensate pump trips with the plant at 3.6 rated conditions.

268000 Radwaste Determine the Drywell Floor 64 CFR 04 Drains indications available 896 to detect drywell general area 2.9 leakage.

271000 Offgas CFR 272000 Radiation Monitoring CFR 286000 Fire Protection CFR 288000 Plant Ventilation CFR 290001 Secondary CTMT Determine inputs to the Fuel 65 CFR 03 Pool leak detection 2.8 897 standpipe.

290003 Control Room HVAC CFR 290002 Reactor Vessel 2. Given a severe accident 91 Internals 4. condition, describe the bases 923 CFR 1 for why the transition is 3.9 4 made from the EOPs to the

226001 RHR/LPCI: Determine the affects to the 92 CTMT Spray Mode 13 Containment Spray mode of 924 CFR RHR given a valve interlock 2.9 failure.

  • 234000 Fuel Handling Determine the affects to fuel 93 Equipment 01 handling operations given a 3.7 925 CFR Refueling Bridge interlock
  • failure.
  • SRO Only Questions PAGE 3 TOTAL PAGE TOTAL #

TIER 2 GROUP 2 1 0 0 1 0 0 0 3 1 0 1 QUESTIONS 7 PAGE 1 TOTAL PAGE TOTAL #

TIER 2 GROUP 2 1 0 0 1 1 0 0 0 0 0 1 QUESTIONS 4 PAGE 2 TOTAL PAGE TOTAL #

TIER 2 GROUP 2 2 0 1 0 0 0 0 0 0 0 1 QUESTIONS 4 TIER 2 GROUP 2 REVISION 2 7/29/2005 PAGE 11 OF 11 NUREG 1021, REVISION 9

TOTALS 4 0 1 2 1 0 0 3 1 0 2 15 REVISION 2 7/29/2005 PAGE 12 OF 11 NUREG 1021, REVISION 9

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form-401-3 Facility: Grand Gulf Nuclear Station Date of Exam: 12 August 2005 Category K/ A# Topic SRO SRO-Only IR # IR #

2.1.19 Given plant conditions and the PDS computer, 66 determine necessary actions based on PBDS 3.0 898 counts.

2.1.25 Given plant conditions and EP3 graphs, 67 determine the correct mitigation strategy. 3.1 899

1. 2.1.29 Determine the correct locking device color 68 coding for locked components. 3.3 237a Conduct 2.1.2 Given conditions, determine when an act of 94 sabotage or tampering should be suspected. 4.0 926 Of Operations 2.1.12 95 4.0 927 Subtotal 3 2 2.2.1 Given plant conditions, determine proper 69 operation of the IRMs. 3.6 900 2.2.30 Discuss the duties of the operator assigned to 70 communicate with the refueling floor SRO 3.3 901 during core alterations.
2. 2.2.19 Describe the process for generating a 96 maintenance work request. 3.1 928 Equipment 2.2.16 Determine who is responsible for reviewing the 97 Control installation and removal of temporary 2.6 929 alterations.

Subtotal 2 2 2.3.1 Given the need to enter a high radiation area, 71 determine the allowed time in the area to prevent 3.0 902 exceeding the administrative exposure limits.

2.3.4 Given plant conditions and applicable 72 Emergency Planning Procedures, determine the 3.1 903 radiation exposure limits that are in effect.

3. 2.3.6 Given liquid radwaste batch release data, 98 Radiation determine which does not require Operations 3.1 930 Control approval or a discharge permit.

Subtotal 2 1 2.4.20 Given plant conditions, determine the bases for 73 any applicable EOP cautions. 4.0 904 2.4.25 Given plant conditions including a fire, 74 determine the proper response. 3.4 905

4. 2.4.43 Given plant conditions and Emergency Planning 75 Procedures, determine the available emergency 3.5 906 communications systems.

Emergency 2.4.47 Given plant conditions and indications from the 99 Procedures / recirculation pump shaft seals, analyze the 931 condition and determine the probable failure 3.7 mechanism.

Plan 2.4.44 Given plant conditions that warrant a General 100 Emergency, determine the correct protective 4.0 932 action recommendations.

Subtotal 3 2 Tier 3 Point Total 10 7 REVISION 2 7/29/2005 PAGE 13 OF 11 NUREG 1021, REVISION 9

ES-401 Record of Rejected K/As Form ES-401-4 Tier/ Randomly Reason for Rejection Group Selected K/A 2/1 206000 High Pressure Core Injection (HPCI) - GGNS does not have a HPCI System for water inventory control.

2/1 207000 Isolation (Emergency) Condenser - GGNS does not have an Isolation Condenser for pressure suppression.

2/2 201002 Reactor Manual Control System (RMCS) - GGNS utilizes the BWR 6 Rod Control and Information System.

2/2 201004 Reactor Sequence Control System (RSCS) - GGNS utilizes the BWR 6 Rod Control and Information System.

2/2 201006 Rod Worth Minimizer (RWM) - GGNS utilizes the BWR 6 Rod Control and Information System.

2/2 214000 Rod Position Information System (RPIS) - GGNS utilizes the BWR 6 Rod Control and Information System.

2/2 215002 Rod Block Monitor (RBM) - GGNS utilizes the BWR 6 Rod Control and Information System.

2/2 230000 RHR/LPCI: Torus/Pool Spray Mode - GGNS does not have a Torus/Pool Spray mode of the RHR System.

1/1 600000 Plant Fire On Site - AK3 was selected for the Random Selection topic. This topic has 4 statements of which only AK3.04 has an importance of > 2.5 for a RO.

2/1 203000 RHR/LPCI: Injection Mode - K5.01 was eliminated due to the testable check valves for RHR are being disabled via plant change per an ER, thus the selection resulted in the only other K/A K5.02.

2/1 300000 Instrument Air - K6.04 has an importance of > 2.5, however GGNS does not have a Service Air Refusal Valve.

2/2 215001 Traversing In-Core Probe - At GGNS TIPS is only operated by Reactor Engineers. The only operations involvement with the system is protective tagging.

1/1 295027 High Containment Temperature 295038 High Offsite Release Rate - Random selection for Generics was 2.2.34 Knowledge of the process for determining the internal and external effects on core reactivity. This K/A does not apply to these two Emergency/Abnormal Plant Evolutions. Random selection was redrawn.

3 Generics Random selection of 2.2.31 Knowledge of procedures and limitations involved in initial core loading is not applicable to GGNS.

Revision 2 7/29/2005

ES-401 Record of Rejected K/As Form ES-401-4 1/1 295027 High Containment Temperature - Generics was selected for the topical area. 2.2.18 Knowledge of the process for managing maintenance activities during shutdown operations and 2.2.1 Ability to perform pre-startup procedures for the facility/including operating those controls associated with plant equipment that could affect reactivity, do not apply to high containment temperature in the realm of emergency level. There is not sufficient energy to raise Containment Temperature to the emergency level.

1/1 295001 Partial or Complete Loss of Forced Core Flow Circulation The initial random selection selected 2.3.2 Alara considerations. Section 2.3 of the Generics have limited applicability for this evolution. Attachment 2 section 1 sentence 4 allows elimination of these K/As without justification. The K/As listed in sentence 1 were numbered 1 - 16 and randomly selected to apply K/A 2.4.4.

1/1 600000 Initial topic called for AA2.16 Ability to determine and interpret as applied to a plant fire on site - vital equipment and control systems to be maintained and operated during a fire. The original outline identified the Main Transformers. The Main Transformers are only vital to Main Generator output. K/A mismatch was felt to be the case and changed the topic to basis for separating vital equipment from the Main Control Room during a fire in the Main Control Room to better fit the K/A based on 10CFR 50 Appendix R considerations.

1/1 295025 Initial K/A was K3.05, this is similar to K/A for 217000 A1.02. Reselected to 295025 K3.06 2/2 202001 Initial K/A was Generic 2.2.25 Tech Spec Bases. Not considered RO level knowledge. Reselected to Generic 2.4.11.

3 SRO 2.1.24 Protective Tagging is considered an RO function for written examinations. Reselected to 2.1.12 for SRO Only question.

2/1 217000 Initial K/A was 2.1.25 which is RO level knowledge.

SRO Reselected to 2.1.10 Knowledge of facility License and conditions which is SRO level knowledge.

Revision 2 7/29/2005

Appendix D Scenario Outline Form ES-D-1 Facility: GRAND GULF NUCLEAR STATION Scenario No.: 1 Op-Test No.: Day 1 Examiners: _________________________ 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 Equipment SRMs E & F are INOP APRM H is INOP due to a failed FCTR 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 Malf. Event Event No. No. Type* Description 1 N Start RCIC and operate CST to CST per EPI.

(BOP) (EPI 04-1-03-E51-2)

REVISION 2 7/25/2005 SCENARIO 1 NUREG 1021 REVISION 9 PAGE 1 of 3

Appendix D Scenario Outline Form ES-D-1 Scenario 1 Day 1 (Continued)

Event Malf. No. Event Event No. Type* Description 2 1 fw126c@ 0 TS Respond to RPV Narrow Range Level C instrument (SS) failure downscale. Complete Technical Specification determination.

3 2 fw232i @ R Respond to a tube failure in LP FW Heater 3C.

50% ramp to (RO) Perform actions per ONEP 05-1-02-V-5 and ONEP 05-80% 1-02-III-3. Lower Reactor power with Recirc flow.

C With a failure to isolate the Condensate System.

(BOP) Perform actions per ARI 04-1-02-1H13-P870 6A-B3 to isolate LP Feedwater Heater String C.

4 3 e51047 C RCIC Turbine Trip. Complete Technical Specification (BOP) determination.

TS (SS) 5 4 c71077b C Respond to a RPS B Motor Generator EPA Breaker (RO/ Trip per the ONEP 05-1-02-III-2.

BOP) 6 5 fw201; C Respond to a double Reactor Recirculation Pump down c71076 (RO) shift, Automatic RPS actuation fails requiring insertion of a manual Reactor Scram.

7 6 c11164 @ M Upon Reactor Scram recognize the failure of all control 0.2% (ALL) rods to fully insert and take actions per EOPs for ATWS with Main Steam Bypass Valves.

7 di_1e12 I Upon orders to initiate and override Low Pressure m617@ (BOP) ECCS, recognize the failure of Division II to initiate via NORM 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 REVISION 2 7/25/2005 SCENARIO 1 NUREG 1021 REVISION 9 PAGE 2 of 3

Critical Tasks

- Terminate and prevent injection from Feedwater and ECCS as required.

- 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.

REVISION 2 7/25/2005 SCENARIO 1 NUREG 1021 REVISION 9 PAGE 3 of 3

Appendix D Scenario Outline Form ES-D-1 Facility: GRAND GULF NUCLEAR STATION Scenario No.: 2 Op-Test No.: Day 1 Examiners: _________________________ 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. Respond to Main Generator TVR failure.
5. 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.)

6. Respond to a failure of Division 1 ECCS to automatically initiate on High Drywell Pressure.
7. 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 Equipment SRMs E & F are INOP and bypassed.

APRM H is INOP due to a failed FCTR 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 Malf. No. Event Event No. Type* Description 1 N Place Standby Service Water A in service for chemical addition.

(BOP) (EPI 04-1-03-P41-1) 2 R (RO) Raise Reactor power using control rods to 49%. (Control Rod Pull Sheet) 3 1 C (RO) Respond to single control rod drift taking actions to insert the control z161161_24 TS rod. (ONEP 05-1-02-IV-1)

_17 (SS) Disarm Control Rod. Complete Technical Specification determination.

REVISION 2 7/25/2005 SCENARIO 2 NUREG 1021 REVISION 9 PAGE 1 of 2

Appendix D Scenario Outline Form ES-D-1 Scenario 2 Day 1 (Continued)

Event Malf. No. Event Event No. Type* Description 4 2 C Respond to trouble and trip of ESF Transformer 21 with a p807_4a_f_2 (BOP) failure of DG 12 to Start. Complete Technical ON TS Specification determination.

r21180 n41140b (SS) (ONEP 05-1-01-I-4) 5 3 n41102 C Respond to a failure of the Main Generator Voltage (RO) Regulator. (ARI 04-1-02-1H13-P680 9A-C15 and SOI 04 01-N40-1)

M 6 4 fw0171b (ALL) Respond to indications of large break LOCA on Feedwater

@ 70% Line B per EOPs. (B21-F065B will close if attempted.)

rr063b @

1% ramp to 4%

5 rr040e@ 0 I Respond to a failure of Division 1 ECCS to automatically rr041e @ (BOP) initiate on High Drywell Pressure.

83%

6 C Respond to a failure of High Pressure Core Spray to inject.

e22159a@0 (BOP)

  • (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).

REVISION 2 7/25/2005 SCENARIO 2 NUREG 1021 REVISION 9 PAGE 2 of 2

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