ML053330588
| ML053330588 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/03/2005 |
| From: | Ryan Lantz Operations Branch IV |
| To: | Gerald Williams Entergy Operations |
| References | |
| 50-416/05-301 50-416/05-301, ES-301, es-301-1 | |
| Download: ML053330588 (34) | |
Text
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Grand Gulf Nuclear Station Date of Examination: 16 August 2005 Examination Level (circle one) RO / SRO Operating Test Number:
Administrative Topic Type Describe activity to be performed (see Note) Code*
Given plant conditions, perform the Idle Loop D; S Startup Surveillance for Recirculation System.
Conduct of Operations GJPM-RO-ADM-1 K/A 2.1.20: 4.3 Safety Function 1 Given plant conditions, complete documentation M; S for Shift Turnover Conduct of Operations GJPM-RO-ADM-2 K/A 2.1.3: 3.0 Given a component to be isolated for a work order, M prepare a tagout using the eSOMS program.
Equipment Control GJPM-RO-ADM-3 K/A 2.2.13: 3.6 Complete entry and egress from the CAA with D access requirements for a High Radiation Area.
Radiation Control GJPM-OPS-ADM-26 K/A 2.3.1: 2.6; 2.3.4: 2.5; 2.3.5: 2.3 Emergency Plan N/A N/A NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria: (C)ontrol Room (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-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Grand Gulf Nuclear Station Date of Examination: 15 August 2005 Exam Level (circle one) RO / SRO-I / SRO-U Operating Test Number:
Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)
System / JPM Title Type Safety Code* Function
- a. 202001 Recirculation System - Startup idle Recirculation Pump S; M; A 4
<30% power with incomplete start actuation.
- b. 201001 Control Rod Drive Hydraulic System - Rotate operating S; M; A 1 CRD pumps, trip of newly operating pump.
- c. 259001 Reactor Feed Water System - Startup Second Reactor S; N; A 2 Feed Pump and place on Master level control, with failure of Automatic controller.
- d. 226001 RHR Containment Spray - Secure Containment Sprays S; N; A 5 and align for injection to RPV with failure of one RHR injection ESF valve.
- e. 264000 Emergency Diesel Generators - Start, parallel, and load the S; D; A 6 Diesel Generator with trip of SSW. ESF
- g. 239001 Main & Reheat Steam System - Open Main Steam Isolation S; D; L 3 Valves. ESF
- h. 201005 Rod Control & Information System - Operate RCIS to bring S; D; L 7 the reactor critical.
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 R; M; E; 6 when power is restored, with one lockout failing to reset. A
- j. 286000 Fire Protection System - Manually initiate fire suppression R; N; E 8 for the Standby Gas Filter Train with the train operating.
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol Room (D)irect from bank 9 / 8 / 4 (E)mergency or abnormal in-plant 1 / 1 / 1 (L)ow-Power 1 / 1 / 1 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)
(R)CA 1 / 1 / 1 (S)imulator
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Grand Gulf Nuclear Station Date of Examination: 16 August 2005 Examination Level (circle one) RO / SRO Operating Test Number:
Administrative Topic Type Describe activity to be performed (see Note) Code*
Given a completed AC/DC Lineup following a M failure of a Diesel Generator, determine the LCO Conduct of Operations action requirements and generate an eSOMS LCO.
GJPM-SRO-ADM-1 K/A 2.1.12: 4.0 Safety Function 6 Given a failed relay, determine the impact on plant M operations using facility drawings.
Conduct of Operations GJPM-SRO-ADM-2 K/A 2.1.24: 3.1 Given a work order and prepared tagout, determine M the adequacy of the tagout and the impact on plant Equipment Control operations.
GJPM-SRO-ADM-3 K/A 2.2.13: 3.8; 2.2.17: 3.5 Given plant conditions, determine Protective N Action Recommendations and Radiological Radiation Control Considerations for On-Site Personnel.
GJPM-SRO-A&E-41 K/A 2.3.8: 3.2 Given plant conditions, determine entry into the M Site Emergency Plan and complete the initial Emergency Plan notification forms.
GJPM-SRO-A&E-42 K/A 2.4.41: 4.1; 2.4.38: 4.0; 2.4.40: 4.0 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria: (C) ontrol Room (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-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Grand Gulf Nuclear Station Date of Examination: 15 August 2005 Exam Level (circle one) RO / SRO-I / SRO-U Operating Test Number:
Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)
System / JPM Title Type Safety Code* Function
- a. 202001 Recirculation System - Startup idle Recirculation Pump S; M; A 4
<30% power with incomplete start actuation.
- b. 201001 Control Rod Drive Hydraulic System - Rotate operating S; M; A 1 CRD pumps, trip of newly operating pump.
- c. 259001 Reactor Feed Water System - Startup Second Reactor S; N; A 2 Feed Pump and place on Master level control, with failure of Automatic controller.
- d. 226001 RHR Containment Spray - Secure Containment Sprays S; N; A 5 and align for injection to RPV with failure of one RHR injection ESF valve.
- e. 264000 Emergency Diesel Generators - Start, parallel, and load the S; D; A 6 Diesel Generator with trip of SSW. ESF
- g. 239001 Main & Reheat Steam System - Open Main Steam Isolation S; D; L 3 Valves. ESF
- h. 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 R; M; E; 6 when power is restored, with one lockout failing to reset. A
- j. 286000 Fire Protection System - Manually initiate fire suppression R; N; E 8 for the Standby Gas Filter Train with the train operating.
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol Room (D)irect from bank 9 / 8 / 4 (E)mergency or abnormal in-plant 1 / 1 / 1 (L)ow-Power 1 / 1 / 1 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)
(R)CA 1 / 1 / 1 (S)imulator
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Grand Gulf Nuclear Station Date of Examination: 15 August 2005 Exam Level (circle one) RO / SRO-I / SRO-U Operating Test Number:
Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)
System / JPM Title Type Safety Code* Function
- a. 202001 Recirculation System - Startup idle Recirculation Pump S; M; A 4
<30% power with incomplete start actuation.
- b. 201001 Control Rod Drive Hydraulic System - Rotate operating S; M; A 1 CRD pumps, trip of newly operating pump.
- c. N/A
- d. N/A
- e. N/A
- g. N/A
- h. N/A In-Plant Systems@ (3 for RO; 3 for SRO-I; 3or2 for SRO-U)
- i. N/A
- j. 286000 Fire Protection System - Manually initiate fire suppression R; N; E 8 for the Standby Gas Filter Train with the train operating.
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol Room (D)irect from bank 9 / 8 / 4 (E)mergency or abnormal in-plant 1 / 1 / 1 (L)ow-Power 1 / 1 / 1 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)
(R)CA 1 / 1 / 1 (S)imulator
BWR EXAMINATION OUTLINE Form ES-40 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 Loss of Forced Core Flow 4. the recirculation system, determine appropriate 4.0 Circulation / 1 & 4 4 actions.
CFR 295003 Partial or Complete Given plant conditions and a loss of AC power, Loss of AC Power/ 6 01 determine the necessary actions to restore vital 3.3 CFR busses.
295004 Partial or Complete Given plant conditions and a loss of DC power, Loss of DC Power / 6 02 determine the effect to the SSW system. 3.8 CFR 295005 Main Turbine Generator Following a reactor scram and subsequent main Trip / 3 03 turbine generator trip, determine the effects of 3.5 CFR manual bypass valve operation on reactor water level.
295006 SCRAM / 1 Given plant conditions following a reactor scram, CFR 02 determine if adequate shutdown margin exists. 3.4 295016 Control Room Describe the method used to manually scram the Abandonment / 7 01 reactor after the control room has been abandoned. 3.8 CFR 295018 Partial or Complete Given plant conditions and a partial loss of Loss of CCW / 8 01 Component Cooling Water, determine the 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 Loss of Inst. Air / 8 01 Air determine a method to restore Instrument Air 3.5 CFR system pressure.
295021 Loss of Shutdown Given specific plant conditions following a loss of Cooling / 4 01 Shutdown Cooling, determine the reason for 3.3 CFR raising reactor water level.
295023 Refueling Accidents / 8 Determine the correct operator response to CFR 03 inadvertent criticality following a refueling 3.7 accident.
295024 High Drywell Pressure / 2. Given plant conditions and high drywell pressure, 5 1. determine the method to lower drywell pressure. 3.9 CFR 23 295025 High Reactor Pressure / Describe RCIC operation following a reactor 3 05 scram where the SRVs are used to control reactor 3.6 CFR pressure.
295026 Suppression Pool High Given an ATWS condition, describe the EP bases Water Temp. / 5 01 for lowering reactor pressure as Suppression Pool 3.8 CFR temperature rises.
295027 High Containment Given rising Containment temperature, describe Temperature / 5 03 the necessary actions to maintain the 3.5 CFR plant/containment in a safe condition.
PAGE 1 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 1 4 0 4 3 1 2 14
BWR EXAMINATION OUTLINE Form ES-40 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 Temperature / 5 02 temperature, determine the effects to control 2.9 CFR room reactor water level indication.
295030 Low Suppression Pool 2. Given a low suppression pool level condition, Water Level / 5 2. determine the effects to other plant systems.
CFR 12 3.0 295031 Reactor Low Water Given plant conditions, describe the operation of Level / 2 04 the High Pressure Core Spray system following 4.3 CFR a LOCA.
295037 SCRAM Condition Given plant conditions and an ATWS condition, Present and Reactor Power determine the availability of the main condenser 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, Rate / 9 01 determine when it is considered to be offsite. 3.3 CFR 600000 Plant Fire On Site / 8 04 Determine the required procedural actions for a fire on the plant site. 2.8 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
BWR EXAMINATION OUTLINE Form ES-4 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 Vacuum / 3 01 condenser vacuum, determine the automatic 3.5 CFR plant response (RPS actuation).
295007 High Reactor Pressure / 2. Determine the conditions necessary to require 3 4. connection of an alternate air source to the 3.3 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 Water Temp. / 5 02 localized suppression pool heating when using 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
BWR EXAMINATION OUTLINE Form ES-4 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 Containment Area Temperature / Auxiliary Building temperatures, describe the 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 Ventilation High Radiation / 9 03 Auxiliary Building radiation levels, describe the CFR conditions that would automatically start the 4.3 Standby Gas Treatment system.
295035 Secondary Containment Given accident conditions and a Standby Gas High Differential Pressure / 5 02 Treatment system failure, determine the type of CFR release.
295036 Secondary Containment Describe the system logic used by the Auxiliary High Sump/Area Water Level / 5 01 Building Floor Drain system to contain a 3.2 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
BWR EXAMINATION OUTLINE Form ES-40 GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)
NUCLEAR STATION 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, Injection Mode 02 describe the design 3.5 8 CFR features and limits of the RHR pump manual override feature.
205000 Shutdown Describe the RHR Cooling 04 Shutdown Cooling 2.6 8 CFR system NPSH interlocks.
206000 HPCI N/A GGNS 207000 Isolation N/A GGNS (Emergency) Condenser 209001 LPCS Given degraded plant CFR 01 conditions during a 8 LOCA, describe LPCS 3.8 manual operation.
209002 HPCS Describe available CFR 09 methods to raise/lower 8 suppression pool level 3.4 using HPCS.
209002 HPCS 2. Describe the bases for CFR 1. the HPCS injection 8 2 valve high reactor water 3.2 8 level interlock.
211000 SLC Predict the SLC system CFR 02 indication and response 8 with indication the squib 3.6 valve failed to actuate and follow up actions.
212000 RPS Given plant conditions CFR 12 including a partial main 8 turbine stop/control 4.0 valve closure, determine the effect to RPS.
215003 IRM Describe the reason for CFR 03 the precaution 8 concerning driving IRMs during 3.0 surveillance activities.
215004 Source Range 2. Describe the SRM Monitor 2. precaution warning of a 8 CFR 3 potential control rod 2.5 3 block even if the channel is bypassed.
215005 APRM / Given a partial loss of LPRM 02 plant electrical power, 8 CFR determine the effect to 2.6 the APRMs.
217000 RCIC Predict how a reactor CFR 02 pressure change will 3.3 8 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
BWR EXAMINATION OUTLINE Form ES-40 GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)
NUCLEAR STATION 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 CFR 01 between ADS Logic 3.1 8 power and the operation of the ADS logic.
223002 PCIS / Nuclear Determine the operator 4 Steam Supply Shutoff 03 actions required to 8 CFR mitigate a NSSSS logic 3.0 failure.
239002 SRVs Describe the design 4 CFR 09 features available to 8 determine if a SRV is 3.7 open.
259002 Reactor Water Describe potential 4 Level Control 04 problems associated with 8 CFR operating a RFP in 3.0 Emergency Manual 259002 Reactor Water Describe prerequisites 4 Level Control 06 for transferring the 2 CFR Feedwater system to 3- 3.1 element control.
261000 SGTS Describe the SGTS 4 CFR 03 damper logic following 3.0 8 system initiation.
262001 AC Electrical Given plant conditions 4 Distribution 01 and a partial loss of DC 8 CFR power, determine the 3.1 affect to the AC distribution system.
262002 UPS (AC/DC) Given plant conditions 4 CFR 01 and degraded AC power, 3.1 8 determine the status of plant inverters.
263000 DC Electrical Given a loss of AC 4 Distribution 01 power to battery 8 CFR chargers, determine the affects to the DC distribution system.
264000 EDGs Describe EDG response 4 CFR 10 to a LOCA. 3.9 8 264000 EDGs 2. Determine EDG status 4 CFR 4. from control room 8 4 alarms and indications 8 and any required 3.5 operator actions to improve plant conditions.
300000 Instrument Air Determine the effect on CFR 01 the plant given a loss of 8 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
BWR EXAMINATION OUTLINE Form ES-GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)
NUCLEAR STATION SYSTEM #/NAME K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC(S) IMP 300000 Instrument Air Determine the affect of a CFR 13 clogged filter on the 2.8 Instrument Air system.
400000 Component Determine the method Cooling Water 04 used to confirm a reactor CFR coolant leak into the CCW 2.9 system.
400000 Component Determine the affect to the 2.8 Cooling Water 02 plant if the CCW 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
BWR EXAMINATION OUTLINE Form ES-4 GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)
NUCLEAR STATION 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 CFR 10 the rod withdrawal 3.2 limiter.
201006 RWM N/A GGNS 202001 Recirculation 2. Given degraded plant CFR 2. conditions determine any 2 applicable Recirculation 2.5 5 Loop LCOs.
202002 Recirculation Given plant conditions, Flow Control 01 determine any automatic CFR41.6 actions associated with 3.5 the Recirculation System HPUs.
204000 RWCU Determine the correct CFR 06 flow path to use RWCU 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
BWR EXAMINATION OUTLINE Form ES-GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)
NUCLEAR STATION 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 and Auxiliaries 08 limitations to 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 Steam 04 conditions CFR including a MSIV closure, 2.8 determine the affect to the Offgas system.
239003 MSIV Leakage Explain the Control 02 relationship CFR between the MSIV Leakage 2.9 Control system and SGTS.
241000 Reactor/Turbine 2. Describe the Pressure Regulator 4. bases for each of 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
BWR EXAMINATION OUTLINE Form ES-GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)
NUCLEAR STATION 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 Gen./Aux. 02 critical speeds as it is 2.8 CFR rolled to rated speed.
256000 Reactor Condensate CFR 259001 Reactor Feedwater Determine necessary CFR 03 actions and priorities immediately after a single condensate pump trips 3.6 with the plant at rated conditions.
268000 Radwaste Determine the Drywell CFR 04 Floor Drains indications 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 CFR 03 Fuel Pool leak detection 2.8 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
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form-4 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
BWR EXAMINATION OUTLINE Form ES-40 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 Loss of Forced Core Flow 4. the recirculation system, determine appropriate Circulation / 1 & 4 4 actions. 4.3 CFR 295003 Partial or Complete Given plant conditions and a loss of AC power, Loss of AC Power/ 6 01 determine the necessary actions to restore vital 3.5 CFR busses.
295004 Partial or Complete Given plant conditions and a loss of DC power, Loss of DC Power / 6 02 determine the effect to the SSW system. 4.1 CFR 295005 Main Turbine Generator Following a reactor scram and subsequent main Trip / 3 03 turbine generator trip, determine the effects of CFR manual bypass valve operation on reactor water 3.7 level.
295006 SCRAM / 1 Given plant conditions following a reactor scram, CFR 02 determine if adequate shutdown margin exists. 3.7 295016 Control Room Describe the method used to manually scram the Abandonment / 7 01 reactor after the control room has been abandoned. 3.9 CFR 295018 Partial or Complete Given plant conditions and a partial loss of Loss of CCW / 8 01 Component Cooling Water, determine the 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 Loss of Inst. Air / 8 01 Air determine a method to restore Instrument Air 3.6 CFR system pressure.
295021 Loss of Shutdown Given specific plant conditions following a loss of Cooling / 4 01 Shutdown Cooling, determine the reason for 3.4 CFR raising reactor water level.
295023 Refueling Accidents / 8 Determine the correct operator response to CFR 03 inadvertent criticality following a refueling 4.0 accident.
295024 High Drywell Pressure / 2. Given plant conditions and high drywell pressure, 5 1. determine the method to lower drywell pressure. 4.0 CFR 23 295025 High Reactor Pressure / Describe RCIC operation following a reactor 3 05 scram where the SRVs are used to control reactor 3.7 CFR pressure.
295026 Suppression Pool High Given an ATWS condition, describe the EP bases Water Temp. / 5 01 for lowering reactor pressure as Suppression Pool 4.1 CFR temperature rises.
295027 High Containment Given rising Containment temperature, describe Temperature / 5 03 the necessary actions to maintain the 3.8 CFR plant/containment in a safe condition.
PAGE 1 TOTAL TIER 1 PAGE TOTAL # QUESTIONS GROUP 1 4 0 4 3 1 2 14
BWR EXAMINATION OUTLINE Form ES-40 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 Temperature / 5 02 temperature, determine the effects to control 3.1 CFR room reactor water level indication.
295030 Low Suppression Pool 2. Given a low suppression pool level condition, Water Level / 5 2. determine the effects to other plant systems.
CFR 12 3.4 295031 Reactor Low Water Given plant conditions, describe the operation of Level / 2 04 the High Pressure Core Spray system following 4.2 CFR a LOCA.
295037 SCRAM Condition Given plant conditions and an ATWS condition, Present and Reactor Power determine the availability of the main condenser 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, Rate / 9 01 determine when it is considered to be offsite. 4.3 CFR 600000 Plant Fire On Site / 8 04 Determine the required procedural actions for a fire on the plant site. 3.4 295004 Partial or Complete Given a loss of Division 1 DC logic power, Loss of DC Power / 6 02 determine the affect to the Division 1 ECCS. 3.9*
CFR 295005 Main Turbine 2. Given plant data including current area dose Generator Trip / 3 3. rates, determine the required personnel 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 High Water Temp. / 5 03 Suppression Pool temperature, interpret HCTL 4.0*
CFR and determine appropriate actions.
295027 High Containment 2. Explain the bases for the Technical Specification Temperature / 5 2. Containment average air temperature limit. 4.1*
CFR 22 295030 Low Suppression Pool Given low suppression pool water level, Water Level / 5 02 determine if suppression pool temperature 3.9*
CFR can/cannot be measured and why.
295038 High Offsite Release 2. Given a severe case fuel handling accident, Rate / 9 2. explain the processes designed to prevent high 3.5*
CFR 28 offsite release rates.
600000 Plant Fire On Site / 8 16 Explain the automatic response of the plant Fire Protection system to a main transformer fire. 3.5*
- 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
BWR EXAMINATION OUTLINE Form ES-4 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 Vacuum / 3 01 condenser vacuum, determine the automatic 3.5 CFR plant response (RPS actuation).
295007 High Reactor Pressure / 2. Determine the conditions necessary to require 3 4. connection of an alternate air source to the 3.5 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 Water Temp. / 5 02 localized suppression pool heating when using 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
BWR EXAMINATION OUTLINE Form ES-4 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 Containment Area Temperature / Auxiliary Building temperatures, describe the 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 Ventilation High Radiation / 9 03 Auxiliary Building radiation levels, describe the CFR conditions that would automatically start the 4.5 Standby Gas Treatment system.
295035 Secondary Containment Given accident conditions and a Standby Gas High Differential Pressure / 5 02 Treatment system failure, determine the type of 4.2 CFR release.
295036 Secondary Containment Describe the system logic used by the Auxiliary High Sump/Area Water Level / 5 01 Building Floor Drain system to contain a 3.3 CFR significant CCW system rupture.
500000 High CTMT Hydrogen Conc. / 5 295011 High Containment Given LOCA conditions, determine when Temperature / 5 01 containment spray should be initiated. 3.9*
CFR 295014 Inadvertent Reactivity 2. Given a control rod drifting out with the plant at Addition / 1 1. power, determine any necessary notifications. 3.3*
CFR 14 295020 Inadvertent Cont. Given a partial MSIV closure, determine the Isolation / 5 & 7 03 affect on reactor power. 3.7*
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
BWR EXAMINATION OUTLINE Form ES-401-GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)
NUCLEAR STATION 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, Injection Mode 02 describe the design CFR features and limits of the 3.7 RHR pump manual override feature.
205000 Shutdown Describe the RHR Cooling 04 Shutdown Cooling 2.6 CFR system NPSH interlocks.
206000 HPCI N/A GGNS 207000 Isolation N/A GGNS (Emergency) Condenser 209001 LPCS Given degraded plant CFR 01 conditions during a LOCA, describe LPCS 3.6 manual operation.
209002 HPCS Describe available CFR 09 methods to raise/lower suppression pool level 3.5 using HPCS.
209002 HPCS 2. Describe the bases for CFR 1. the HPCS injection 2 valve high reactor water 3.3 8 level interlock.
211000 SLC Predict the SLC system CFR 02 indication and response with indication the squib 3.9 valve failed to actuate and follow up actions.
212000 RPS Given plant conditions CFR 12 including a partial main turbine stop/control 4.1 valve closure, determine the effect to RPS.
215003 IRM Describe the reason for CFR 03 the precaution concerning driving 3.1 IRMs during surveillance activities.
215004 Source Range 2. Describe the SRM Monitor 2. precaution warning of a CFR 3 potential control rod 2.9 3 block even if the channel is bypassed.
215005 APRM / Given a partial loss of LPRM 02 plant electrical power, CFR determine the effect to 2.8 the APRMs.
217000 RCIC Predict how a reactor CFR 02 pressure change will 3.3 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
BWR EXAMINATION OUTLINE Form ES-401 GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)
NUCLEAR STATION 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 CFR 01 between ADS Logic 3.3 power and the operation of the ADS logic.
223002 PCIS / Nuclear Determine the operator Steam Supply Shutoff 03 actions required to CFR mitigate a NSSSS logic 3.3 failure.
239002 SRVs Describe the design CFR 09 features available to determine if a SRV is 3.6 open.
259002 Reactor Water Describe potential Level Control 04 problems associated with CFR operating a RFP in 3.1 Emergency Manual 259002 Reactor Water Describe prerequisites Level Control 06 for transferring the CFR Feedwater system to 3- 3.2 element control.
261000 SGTS Describe the SGTS CFR 03 damper logic following 2.9 system initiation.
262001 AC Electrical Given plant conditions Distribution 01 and a partial loss of DC CFR power, determine the 3.4 affect to the AC distribution system.
262002 UPS (AC/DC) Given plant conditions CFR 01 and degraded AC power, determine the status of plant inverters.
263000 DC Electrical Given a loss of AC Distribution 01 power to battery CFR chargers, determine the 3.4 affects to the DC distribution system.
264000 EDGs Describe EDG response CFR 10 to a LOCA. 4.2 264000 EDGs 2. Determine EDG status CFR 4. from control room 4 alarms and indications 8 and any required 3.8 operator actions to improve plant conditions.
300000 Instrument Air Determine the effect on CFR 01 the plant given a loss of 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
BWR EXAMINATION OUTLINE Form ES-GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 1 (RO/SRO)
NUCLEAR STATION 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 CFR 13 clogged filter on the 2.3 Instrument Air system.
400000 Component Determine the method used Cooling Water 04 to confirm a reactor CFR coolant leak into the CCW 3.1 system.
400000 Component Determine the affect to the Cooling Water 02 plant if the CCW 2.8 CFR temperature control fails.
203000 RHR/LPCI: 2. Given LOCA conditions, Injection Mode 3. determine how LPCI works CFR 1 in conjunction with the 3.2 1 other ECCS to control
- radiation releases.
209001 LPCS 2. Given a short-term CFR 1. problem associated with 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 CFR 4. requiring entry into the 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 Monitor 2. Specs and a repaired SRM CFR 2 detector, determine the 3.5 1 surveillance requirements
- to ensure operability.
217000 RCIC 2. Given plant conditions and CFR 1. procedures, determine 2 Suppression Pool Level 3.1 5 using the RCIC System. *
- 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
BWR EXAMINATION OUTLINE Form ES-401-GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)
NUCLEAR STATION 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 CFR 10 the rod withdrawal 3.3 limiter.
201006 RWM N/A GGNS 202001 Recirculation 2. Given degraded plant CFR 2. conditions determine any 2 applicable Recirculation 3.7 5 Loop LCOs or safety limits.
202002 Recirculation Given plant conditions, Flow Control 01 determine any automatic CFR41.6 actions associated with 3.6 the Recirculation System HPUs.
204000 RWCU Determine the correct CFR 06 flow path to use RWCU 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
BWR EXAMINATION OUTLINE Form ES-40 GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)
NUCLEAR STATION 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 and Auxiliaries 08 limitations to 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 Steam 04 conditions CFR including a MSIV closure, 2.8 determine the affect to the Offgas system.
239003 MSIV Leakage Explain the Control 02 relationship CFR between the MSIV Leakage 3.0 Control system and SGTS.
241000 Reactor/Turbine 2. Describe the Pressure Regulator 4. bases for each of 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
BWR EXAMINATION OUTLINE Form ES-GRAND GULF PLANT SYSTEMS - TIER 2 GROUP 2 (RO/SRO)
NUCLEAR STATION 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 Gen./Aux. 02 critical speeds as it is rolled 2.8 CFR to rated speed.
256000 Reactor Condensate CFR 259001 Reactor Determine necessary actions Feedwater 03 and priorities immediately CFR after a single condensate pump trips with the plant at 3.6 rated conditions.
268000 Radwaste Determine the Drywell Floor CFR 04 Drains indications available 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 CFR 03 Pool leak detection 2.8 standpipe.
290003 Control Room HVAC CFR 290002 Reactor Vessel 2. Given a severe accident Internals 4. condition, describe the bases CFR 1 for why the transition is 3.9 4 made from the EOPs to the
- SAPs.
226001 RHR/LPCI: Determine the affects to the CTMT Spray Mode 13 Containment Spray mode of CFR RHR given a valve interlock 2.9 failure.
- 234000 Fuel Handling Determine the affects to fuel Equipment 01 handling operations given a 3.7 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 TOTALS 4 0 1 2 1 0 0 3 1 0 2 15
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form-4 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, determine necessary actions based on PBDS 3.0 counts.
2.1.25 Given plant conditions and EP3 graphs, determine the correct mitigation strategy. 3.1
- 1. 2.1.29 Determine the correct locking device color coding for locked components. 3.3 Conduct 2.1.2 Given conditions, determine when an act of sabotage or tampering should be suspected. 4.0 Of Operations 2.1.24 Given the need to generate a protective tagging clearance, discuss any procedural guidance for 3.1 use of electrical and mechanical drawings.
Subtotal 3 2 2.2.1 Given plant conditions, determine proper operation of the IRMs. 3.6 2.2.30 Discuss the duties of the operator assigned to communicate with the refueling floor SRO 3.3 during core alterations.
- 2. 2.2.19 Describe the process for generating a maintenance work request. 3.1 Equipment 2.2.16 Determine who is responsible for reviewing the Control installation and removal of temporary 2.6 alterations.
Subtotal 2 2 2.3.1 Given the need to enter a high radiation area, determine the allowed time in the area to prevent 3.0 exceeding the administrative exposure limits.
2.3.4 Given plant conditions and applicable Emergency Planning Procedures, determine the 3.1 radiation exposure limits that are in effect.
- 3. 2.3.6 Given liquid radwaste batch release data, Radiation determine which does not require Operations 3.1 Control approval or a discharge permit.
Subtotal 2 1 2.4.20 Given plant conditions, determine the bases for any applicable EOP cautions. 4.0 2.4.25 Given plant conditions including a fire, determine the proper response. 3.4
- 4. 2.4.43 Given plant conditions and Emergency Planning Procedures, determine the available emergency 3.5 communications systems.
Emergency 2.4.47 Given plant conditions and indications from the Procedures / recirculation pump shaft seals, analyze the condition and determine the probable failure 3.7 mechanism.
Plan 2.4.44 Given plant conditions that warrant a General Emergency, determine the correct protective 4.0 action recommendations.
Subtotal 3 2 Tier 3 Point Total 10 7
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:
2 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 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 4. Respond to a trip of RCIC.
5 5. Respond to a loss of RPS normal power supply.
6 6. Take actions for a double Recirculation Pump downshift to manually scram the reactor.
7 7. Take actions per the EOPs in response to an ATWS and mitigate the consequences of the ATWS with Main Steam Bypass Valves.
8 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 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 Malf. Event Event No. No. Type* Description 1 N (BO Start RCIC and operate CST to CST per EPI.
P) (EPI 04-1-03-E51-2)
Scenario Outline Form ES-D-1 Appendix D 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 (R Respond to a tube failure in LP FW Heater 3C.
50% ramp to O) Perform actions per ONEP 05-1-02-V-5 and ONEP 05-80%
1-02-III-3. Lower Reactor power with Recirc flow.
C (BO With a failure to isolate the Condensate System.
P) Perform actions per ARI 04-1-02-1H13-P870 6A-B3 to isolate LP Feedwater Heater String C.
4 3 e51047 C (BO RCIC Turbine Trip. Complete Technical Specification P) determination.
TS (SS) 5 4 c71077b C (R Respond to a RPS B Motor Generator EPA Breaker O/ Trip per the ONEP 05-1-02-III-2.
BOP) 6 5 fw201; C (R Respond to a double Reactor Recirculation Pump down c71076 O) shift, Automatic RPS actuation fails requiring insertion of a manual Reactor Scram.
7 6 c11164 @ M (AL Upon Reactor Scram recognize the failure of all control 0.2% L) rods to fully insert and take actions per EOPs for ATWS with Main Steam Bypass Valves.
7 di_1e12 I (BO Upon orders to initiate and override Low Pressure m617@ P) 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
Critical Tasks
- Terminate 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 D Scenario Outline Form ES-D-1 Facility: GRAND GULF NUCLEAR STATION Scenario No.: 2 Op-Test No.: Day 1 Examiners: _________________________ Operators:__________________________
2 Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
3 1. Start SSW A in support of chemical addition.
4 2. Raise Reactor Power by withdrawing control rods. Respond to single control rod drift per ONEP 05-1-02-IV-1.
5 3. Respond to ESF Transformer 21 trouble and subsequent trip with a failure of DG 12 to start.
6 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.)
7 5. Respond to a failure of Division 1 ECCS to automatically initiate on High Drywell Pressure.
8 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 Equipment SRMs 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 Malf. No. Event Event No. Type* Description 1 N (BO Place Standby Service Water A in service for chemical addition.
P) (EPI 04-1-03-P41-1) 2 R (RO) Raise Reactor power using control rods to between 40 and 45%.
(Control Rod Pull Sheet) 3 1 z161161_ C (RO) Respond to single control rod drift taking actions to insert the control 24_17 TS rod. (ONEP 05-1-02-IV-1)
(SS) Disarm Control Rod. Complete Technical Specification determination.
Scenario Outline Form ES-D-1 Appendix D Scenario 2 Day 1 (Continued)
Event Malf. No. Event Event No. Type* Description 4 2 p807_ C (BO Respond to trouble and trip of ESF Transformer 21 with a 4a_f_2 ON P) failure of DG 12 to Start. Complete Technical r21180 TS Specification determination.
n41141b (SS) (ONEP 05-1-01-I-4)
M 5 3 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%
4 rr040e@ 0 I (BOP Respond to a failure of Division 1 ECCS to automatically rr041e @ ) initiate on High Drywell Pressure.
83%
5 e22159a@ C (BO Respond to a failure of High Pressure Core Spray to inject.
0 P)
- (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 D Scenario Outline Form ES-D-1 Facility: GRAND GULF NUCLEAR STATION Scenario No.: 3 Op-Test No.: BACK UP Examiners: _________________________ Operators:__________________________
Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
2 1. Start 2nd Condensate Pump and Condensate Booster Pump.
3 2. Raise Reactor Power/ Pressure by withdrawing control rods.
4 3. Respond to a stuck control rod.
5 4. Respond to a trip of LCC 15BA3.
6 5. Respond to an automatic and manual scram function failure; ATWS ARI/RPT will insert control rods with three control rods stuck withdrawn.
7 6. Recognize the failure of MSIVs to completely isolate and take actions to isolate the Main Steam Lines.
8 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.
9 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.
Scenario Outline Form ES-D-1 Appendix D Scenario 3 Backup (Continued)
Event Malf. No. Event Event No. Type* Description 1 N (RO) Start 2nd Condensate and Condensate Booster Pumps.
(SOI 04-1-01-N19-1) 2 R (RO) Raise reactor power and pressure by withdrawing control rods.
(IOI 03-1-01-1 and Control Rod Movement Sheet) 3 1 z022022 _ C (RO/ Respond to a stuck control rod during withdrawal.
40_45 BOP) (ONEP 05-1-02-IV-1)
TS (SS) Complete Technical Specification determination.
4 2 r21142t C (BOP/ Respond to a trip of Load Control Center 15BA3.
RO) (ONEP 05-1-02-I-4; 05-1-02-III-5; and SOI 04-1-01-R21-15) 5 3 c71162 C (RO) Recognize a failure to scram using RPS and manually scram the reactor using ATWS ARI.
6 4 ms183b I (BOP) Recognize the failure of MSIVs to completely isolate and take ms184b actions to isolate the Main Steam Lines.
att9 (ONEP 05-1-02-III-5) 7 5 ms067b @ M (ALL) Recognize and respond to a steam leak in the Auxiliary Building 20% ms066b Steam Tunnel. Take actions for mitigation of the leak with a
@ 0.2% failure of the MSIVs to fully isolate.
ramp to 10%
6 z022022 _ C (RO) Recognize the failure of two additional control rods to fully insert 36_25 on the Reactor Scram. (Three Rods Out) insert control rods
_12_09 (ONEP 05-1-02-IV-1)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Critical Tasks 1 Manually scram the reactor.
2 Isolate the main steam lines.