ML041030293

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02-2004-INITIAL EXAM-FINAL Outlines
ML041030293
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 02/06/2004
From:
NRC Region 4
To:
Entergy Operations
References
50-416/04-301 50-416/04-301
Download: ML041030293 (28)
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Text

ES-401 FORM ES-401-1 BWR SRO EXAMINATION OUTLINE Facility: GRAND GULF NUCLEAR STATION Date of Exam: 6 FEBRUARY 2004 K/A CATEGORY POINTS TIER GROUP K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G POINT

  • TOTAL
1. 1 6 3 2 8 3 4 26 Emergency &

Abnormal 2 0 2 3 4 5 3 17 Plant Evolutions TIER 6 5 5 12 8 7 43 TOTAL 1 1 1 3 3 0 3 3 1 1 3 4 23 2.

Plant 2 1 1 1 0 3 2 0 2 2 0 1 13 Systems 3 0 0 0 0 0 1 1 1 0 1 0 4 TIER 2 2 4 3 3 6 4 4 3 4 5 40 TOTAL CAT 1 CAT 2 CAT 3 CAT 4

3. Generic Knowledge & Abilities 5 4 2 6 17 Note: 1. Ensure that at least two topics from every K/A category are sampled within each tier (i.e., the Tier Totals in each K/A category shall not be less than two)
2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/- 1 from that specified in the table based on NRC revisions. The final exam must total 100 points.
3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant specific priorities.
4. Systems / evolutions within each group are identified on the associated outline.
5. The shaded areas are not applicable to the category tier.

6.* The generic K/As in Tiers 1 and 2 shall be selected from section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.

7. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings for the SRO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above.

REVISION 0 11/5/2003 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 FEBRUARY 2004 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 1 2 # /BOTH K/A 295003 Partial or Complete Loss of AC Power/ 6 02 Given plant conditions describe the difference of how 3.1 801 BOTH AK1.02: 3.4 NEW CFR41.7 loads on BOP and ESF busses are removed and q001 AK1.03: 3.2 subsequently restored during undervoltage conditions. AK2.03: 3.9 AK3.01: 3.5 AK3.03: 3.6 AA1.01: 3.8 295006 SCRAM / 1 03 Given conditions of a reactor scram, describe the 3.7 802 BOTH AK2.07: 4.1 NEW CFR41.5 response of the Turbine Pressure Control System. q002 AA2.04: 4.1 295007 High Reactor Pressure / 3 01 Given Reactor pressure, determine systems available to 3.2 803 BOTH AK2.03: 3.2 NEW CFR41.6 inject into the RPV for level control. q003 AK2.04: 3.3 295009 Low Reactor Water Level / 2 02 Given a steam flow / feed flow mismatch and plant 3.7 804 BOTH MOD CFR41.4/41.5/41.7/43.5 conditions, determine the reactor water level response and q004 NRC response of Reactor Water Level control. 8/2002 295010 High Drywell Pressure / 5 05 Given plant parameters, determine the affects on Drywell 3.8 805 BOTH 223001 MOD CFR41.4/41.5 Pressure. (Loss of cooling to the Drywell Chilled Water q005 K6.01: 3.8 NRC System with the plant at power.) A4.12: 3.6 3/1998 295013 High Suppression Pool Water Temp. / 5 01 During a surveillance operating RCIC, determine how 4.0 876 SRO AA1.02: 3.9 MOD CFR41.10/43.2/43.5 often Suppression Pool Temperature is required to be q076 2.1.33: 4.0 NRC monitored and the threshold for alternate actions. 2.4.4: 4.3 8/2002 295014 Inadvertent Reactivity Addition / 1 2. With the reactor in startup conditions such that the reactor 3.4 806 BOTH AA1.04: 3.3 NEW Pilgrim event CFR41.1/41.2/41.6/43.6 1. is close to criticality, what are the operator actions if a q006 AA2.02: 3.9 2/2003 30 high worth control rod is withdrawn. AA2.03: 4.3 2.1.2: 4.0 295015 Incomplete SCRAM / 1 04 Given control panel indications, determine the cause 3.7 807 BOTH AA1.01: 3.9 NEW CFR41.6/43.5 preventing full insertion of control rods under scram q007 AA1.02: 4.2 conditions. 2.1.31: 3.9 295016 Control Room Abandonment / 7 05 Given a loss of DC electrical power, describe the status of 2.9 808 BOTH NEW CFR41.7 operation of the Safety Relief Valves operated from the q008 Remote Shutdown Panels.

295017 High Offsite Release Rate / 9 01 With a release of radioactive material in progress, 3.9 809 BOTH AK3.05: 3.6 NEW CFR41.11/41.13/43.4 determine the response of systems to protect the safety of q009 control room personnel and maintain habitability.

PAGE 1 TOTAL TIER 1 GROUP 1 1 1 2 3 2 1 PAGE TOTAL # QUESTIONS 10 REVISION 1 11/14/2003 PAGE 1 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE CONT. ES-401-1 FEBRUARY 2004 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 1 2 # /BOTH K/A 295023 Refueling Accidents / 8 02 With a Refueling outage in progress, determine the 3.1 877 SRO NEW CFR41.4/41.5/41.10/43.5/43.7 effects of a loss of Fuel Pool Cooling and Cleanup on the q077 Fuel Storage pools.

295024 High Drywell Pressure / 5 06 Given a high drywell pressure condition, determine the 4.0 811 BOTH EA1.06: 3.7 NEW CFR41.7/41.10/43.5 operation of the Divisional Diesel Generators. q011 295025 High Reactor Pressure / 3 04 With a rising reactor pressure, determine the response of 4.1 812 BOTH EK2.01: 4.1 NEW CFR41.5/41.6/41.7 the RPS and ATWS ARI/RPT. q012 295026 Suppression Pool High Water Temp. / 5 2. With RHR operating in Suppression Pool Cooling in 2.9 810 BOTH 2.1.32: 3.8 NEW CFR41.10/41.12/43.4/43.5 3. response to a high Suppression Pool Temperature, q010 2 describe the basis for contacting Radiation Protection personnel.

295027 High Containment Temperature / 5 03 Determine the Containment Temperature at which 3.8 813 BOTH EK3.01: 3.8 MOD CFR41.9/41.10/43.2/43.5 Emergency Depressurization is required. q013 NRC 12/2000 295030 Low Suppression Pool Water Level / 5 02 Determine the Suppression Pool Water level at which 3.8 814 BOTH NEW CFR41.8/41.10/43.5 ECCS pump NPSH is questionable. q014 295031 Reactor Low Water Level / 2 01 Given plant conditions and a low reactor water level, 4.7 815 BOTH 2.1.1: 3.8 MOD CFR41.2/41.3/41.10/41.14/43.5 determine core cooling mechanism and adequacy. q015 2.4.6: 4.0 NRC 2.4.18: 3.6 8/2002 295037 SCRAM Condition Present and Reactor 2. Given ATWS conditions, determine the Emergency Plan 4.0 878 SRO 2.4.41: 4.1 NEW Alert vs. Site Power Above APRM Downscale or Unknown / 1 4. Emergency Action Level. q078 Area CFR41.10/43.5 40 Emergency 295038 High Offsite Release Rate / 9 02 Given meteorological data, maps and a radioactive 3.8 879 SRO 2.4.44: 4.0 NEW CFR41.10/41.12/43.4/43.5 release, determine protective action recommendations to q079 be issued.

500000 High Containment Hydrogen Conc. / 5 01 Determine the bases for the Hydrogen Control leg of EP- 3.9 817 BOTH MOD CFR41.9 3. q017 NRC 8/2002 295031 Reactor Low Water Level / 2 2. Determine actual reactor water level when operating from 3.9 816 BOTH EK2.01: 4.4 MOD CFR41.7/41.10/43.5 1. the Remote Shutdown Panels using the associated graphs q016 EA2.01: 4.6 NRC 31 and given indications. 8/20021 PAGE 2 TOTAL TIER 1 GROUP 1 3 2 0 3 0 3 PAGE TOTAL # QUESTIONS 11 REVISION 1 11/14/2003 PAGE 2 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE CONT. ES-401-1 FEBRUARY 2004 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 1 2 # BOTH K/A 295025 High Reactor Pressure / 3 05 Given plant conditions, describe the response of RCIC to 3.7 818 BOTH 217000 NEW CFR41.4/41.5/41.14 a rising reactor pressure. q018 A1.04: 3.6 295017 High Off-Site Release Rate / 9 01 With a liquid radwaste discharge required and a discharge 3.1 880 SRO 2.3.3: 2.9 NEW CFR41.10/41.13/43.2/43.4/43.5 permit, determine whether a release is allowed. q080 2.3.6: 3.1 295015 Incomplete SCRAM / 1 04 Describe the reaction of the core with an ATWS and 3.8 820 BOTH AK1.02: 4.1 MOD CFR41.1/41.2/41.5 lowering of reactor pressure. q020 NRC 4/2000 295030 Low Suppression Pool Water Level / 5 01 Given the failure of Control Room Suppression Pool 4.2 821 BOTH 2.1.25: 3.1 NEW EOP 2 CFR41.7/41.9/41.10/43.5 Level indication, determine Suppression Pool level using q021 2.4.21: 4.3 Attachment 29 alternate means.

295026 Suppression Pool High Water Temp. / 5 02 Describe the relationship between Reactor Pressure, 3.8 822 BOTH 2.4.18: 3.6 MOD CFR41.5/41.9/41.10/43.2/43.5 Suppression Pool Temperature, and the ability of the q022 2.4.6: 4.0 NRC Suppression Pool to take reactor pressure. 2.4.14: 3.9 3/1998 PAGE 3 TOTAL TIER 1 GROUP 1 2 0 0 2 1 0 PAGE TOTAL # QUESTIONS 5 PAGE 1 TOTAL TIER 1 GROUP 1 1 1 2 3 2 1 PAGE TOTAL # QUESTIONS 10 PAGE 2 TOTAL TIER 1 GROUP 1 3 2 0 3 0 3 PAGE TOTAL # QUESTIONS 11 K/A CATEGORY TOTALS: 6 3 2 8 3 4 TIER 1 GROUP 1 GROUP POINT TOTAL 26 REVISION 1 11/14/2003 PAGE 3 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 FEBRUARY 2004 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 1 2 # BOTH K/A 295001 Partial or Complete Loss of Forced Core 2. Given plant conditions and a reduction in core flow, 3.2 819 BOTH AK1.03: 4.1 NEW Flow Circulation / 1 & 4 2. determine the effects on Thermal Limits and core q019 AK1.04: 3.3 CFR41.5/41.10/43.1/43.5 34 stability.

295002 Loss of Main Condenser Vacuum / 3 03 Given plant conditions, determine how a loss of 3.5 823 BOTH AA1.04: 3.4 NEW Low Power CFR41.4/41.7/43.5 condenser vacuum will affect the ability of the plant to q023 AK2.01: 3.5 remain operating. (RPS) AK2.03: 3.6 295004 Partial or Complete Loss of DC Power / 6 03 Given a loss of DC control power and conditions that 3.6 824 BOTH NEW CFR41.5/41.7 would normally result in trips of the AC Electrical q024 Distribution System, determine the operation of the AC circuit breakers.

295005 Main Turbine Generator Trip / 3 03 Given a trip of the Main Generator, determine the affects 3.0 825 BOTH NEW CFR41.5/41.6 on Feedwater temperature to the reactor. q025 295008 High Reactor Water Level / 2 03 During a reactor startup from cold shutdown, determine 3.0 826 BOTH AA2.05: 3.1 NEW CFR41.4/41.5 the means for control of reactor water level during reactor q026 AA2.04: 3.3 heat up. (RWCU Blow down) 295011 High Containment Temperature / 5 01 Given plant conditions, determine Containment cooling 3.9 827 BOTH AK2.01: 4.0 MOD CFR41.5/41.9/43.5 mechanisms and available additional cooling. q027 NRC 12/2000 295012 High Drywell Temperature / 5 295018 Partial or Complete Loss of CCW / 8 295019 Partial or Complete Loss of Inst. Air / 8 01 Given a loss of Instrument Air, determine Safety Relief 3.4 828 BOTH NEW GGNS Scram #

CFR41.4/41.10/43.5 Valves that can be operated using nitrogen installed per q028 107 off normal event procedures.

295020 Inadvertent Cont. Isolation / 5 & 7 06 Given plant conditions and an isolation of the 3.8 829 BOTH NEW CFR41.4/41.7/41.9/41.10/43.5 Containment, Auxiliary Building and Drywell, determine q029 validity and ability to restore system.

295021 Loss of Shutdown Cooling / 4 02 Given plant conditions with ADHR in service for 3.4 830 BOTH NEW CFR41.5/41.10/43.5 Shutdown Cooling, determine the affects of a plant q030 transient on ADHR operation.

PAGE 1 TOTAL TIER 1 GROUP 2 0 0 3 2 3 1 PAGE TOTAL # QUESTIONS 9 REVISION 1 11/14/2003 PAGE 4 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE CONT. ES-401-1 FEBRUARY 2004 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 1 2 # BOTH K/A 295022 Loss of CRD Pumps / 1 2. Given plant conditions and a trip of the operating CRD 4.4 831 BOTH AK1.01: 3.4 NEW CFR41.5/41.10/43.5 1. pump, determine the actions to be taken. q031 2.4.4: 4.3 7 2.4.7: 3.8 295028 High Drywell Temperature / 5 03 Given plant conditions and EOP graphs, determine the 3.9 832 BOTH EK1.01: 3.7 NEW CFR41.5/41.7/41.10/43.5 accuracy of reactor water level indications. q032 295029 High Suppression Pool Water Level / 5 295032 High Secondary Containment Area 04 Given entry into the Secondary Containment EOP on high 3.4 833 BOTH NEW Temperature / 5 temperature in an ECCS Room, identify systems not q033 CFR41.4/41.10/43.5 required to be isolated from Primary Containment.

295033 High Secondary Containment Area 04 Given high area radiation levels in Secondary 4.2 834 BOTH NEW Radiation Levels / 9 Containment, determine when Standby Gas Treatment q034 CFR41.12/43.4 will be required to be for operated.

295034 Secondary Containment Ventilation High 2. Given plant parameters, determine operation of 4.4 835 BOTH NEW Radiation / 9 1. ventilation systems. q035 CFR41.4/41.10/41.13/43.4 7 295035 Secondary Containment High Differential 01 Describe the operation of the Secondary Containment 3.6 836 BOTH NEW Pressure / 5 Ventilation Systems due to high differential pressure. q036 CFR41.4/41.7/41.13 295036 Secondary Containment High Sump/Area 03 Given plant conditions, identify the available routes to 3.0 837 BOTH NEW Water Level / 5 remove water from ECCS pump rooms. q037 CFR41.4/41.10/43.5 600000 Plant Fire On Site / 8 03 Determine the actions that will occur upon activation of a 3.2 838 BOTH NEW CFR41.10/43.5 fire alarm. q038 PAGE 2 TOTAL TIER 1 GROUP 2 0 2 0 2 2 2 PAGE TOTAL # QUESTIONS 8 PAGE 1 TOTAL TIER 1 GROUP 2 0 0 3 2 3 1 PAGE TOTAL # QUESTIONS 9 K/A CATEGORY TOTALS: 0 2 3 4 5 3 TIER 1 GROUP 2 GROUP POINT TOTAL 17 REVISION 1 11/14/2003 PAGE 5 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 FEBRUARY 2004 PLANT SYSTEMS - TIER 2 GROUP 1 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 201005 RCIS 01 Given a failure of the Main Steam Bypass 3.3 839 BOTH A2.04: 3.2 NEW CFR41.6/43.6 valves open with the plant at power, determine q039 K6.01: 3.2 the affects on RCIS. K5.10: 3.3 K1.02: 3.5 202002 Recirculation Flow Control 06 Describe the operation of the Recirc Flow 3.7 840 BOTH A1.08: 3.4 NEW CFR41.6 Control Valves during a Flow Control Valve q040 Runback when a HPU alarms.

203000 RHR/LPCI: Injection Mode 10 Describe the affects on LPCI injection when 3.1 841 BOTH NEW CFR41.8 the associated Standby Service Water System q041 trips.

209001 LPCS 10 Describe the operation of the LPCS Injection 2.9 842 BOTH NEW CFR41.7 valve without ECCS injection signals present. q042 209002 HPCS 2. Given plant conditions and a failure of the 3.9 881 SRO 2.2.22: 4.1 NEW CFR41.7/41.8/43.1/43.2 1. HPCS system, determine the actions with q081 2.2.25: 3.7 10 respect to Tech Specs.

211000 SLC 04 During an initiation of SLC with a failure of 3.7 843 BOTH A2.06: 3.3 NEW CFR41.6/41.7 the SLC pumps to start, determine the final q043 A2.07: 3.2 valve positions.

212000 RPS 12 Describe the affect on Secondary Containment 3.3 844 BOTH NEW CFR41.7 with a loss of power to RPS. q044 215004 Source Range Monitor 06 Describe the hazards involved with movement 2.8 845 BOTH NEW CFR41.5/41.6 of SRM detectors following under vessel q045 work.

PAGE 1 TOTAL TIER 2 GROUP 1 1 0 2 1 0 1 2 0 0 0 1 PAGE TOTAL # QUESTIONS 8 REVISION 1 11/14/2003 PAGE 6 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 FEBRUARY 2004 PLANT SYSTEMS - TIER 2 GROUP 1 CONT.

SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 215005 APRM / LPRM 01 Given LPRM/APRM status and a loss of 2.6 846 BOTH K1.01: 4.0 NEW CFR41.6/41.7 power to an LPRM, determine the reaction of q046 K5.06: 2.6 the RPS & RCIS systems. K4.01: 3.7 K4.02: 4.2 216000 Nuclear Boiler Instrumentation 03 Given leakage on the instrument line for 3.1 847 BOTH K1.22: 3.8 NEW CFR41.5/41.7 Reactor Vessel Level indication, determine the q047 indications and reaction of systems supplied that indication.

217000 RCIC 04 With RCIC operating for a surveillance, 3.6 848 BOTH A2.03: 3.3 NEW CFR41.5/41.7/41.10/43.5 determine the affects of a manual isolation q048 signal.

218000 ADS 2. Given plant conditions, determine the LCO 3.8 882 SRO NEW CFR41.7/43.1/43.2 2. status for inoperable ADS valves. q082 23 223001 Primary CTMT and Auxiliaries 2. Given plant conditions, determine 4.1 849 BOTH NEW CFR41.9/41.10/43.5 4. requirements for entry into the Emergency q049 2 Operating Procedures.

223002 PCIS / Nuclear Steam Supply 03 Given radiation monitor readings and 3.1 850 BOTH 272000 NEW Shutoff radiography in Containment, determine the q050 K1.09: 3.8 CFR41.7/41.9/41.11/43.4 status of plant systems.

226001 RHR/LPCI: CTMT Spray Mode 08 Given indications from plant instrumentation, 2.8 851 BOTH NEW CFR41.7/41.8/41.10/43.5 determine the operation of the Containment q051 Spray System.

239002 SRVs 08 Given SRV operation, determine the meaning 3.6 852 BOTH NEW CFR41.7 of indications and SRV status. q052 241000 Reactor / Turbine Pressure 06 Identify the conditions of the Reactor/Turbine 3.7 853 BOTH NEW Regulator Pressure Control system that would result in a q053 CFR41.7 Main Turbine Trip.

PAGE 2 TOTALS TIER 2 GROUP 1 0 1 0 1 0 2 0 1 1 1 2 PAGE 2 TOTAL # QUESTIONS 9 REVISION 1 11/14/2003 PAGE 7 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 FEBRUARY 2004 PLANT SYSTEMS - TIER 2 GROUP 1 CONT.

SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 259002 Reactor Water Level Control 02 With the Digital Feedwater Level Control 3.6 854 BOTH NEW CFR41.4/41.5/41.7 System selected for automatic operation, q054 determine the reaction of the system for a given failure.

261000 SGTS 2. Given operation of the Standby Gas Treatment 3.1 855 BOTH NEW CFR41.7/41.10/41.11/43.4 4. System followed by alarms that would indicate q055 10 a change in plant status, determine actions to be taken.

262001 AC Electrical Distribution 01 Given the plant at full power and a loss of bus 3.7 856 BOTH 202001 NEW CFR41.4/41.7 11HD, determine the final operation of the q056 K1.08: 3.2 Recirculation system. K6.03: 3.0 264000 EDGs 07 Given system alignment, determine the 3.4 857 BOTH NEW CFR41.8/43.2 operational condition of the diesel generator. q057 290001 Secondary CTMT 03 Identify the proper alignment of the Auxiliary 2.7 858 BOTH NEW CFR41.10 Building Ventilation systems to maintain q058 proper building differential pressure.

262001 AC Electrical Distribution 02 Determine the method employed to control the 3.5 859 BOTH NEW CFR41.10/43.5 return of loads during a station blackout when q059 cross connecting Division III to Division II.

PAGE 3 TOTALS TIER 2 GROUP 1 0 0 1 1 0 0 1 0 0 2 1 PAGE TOTAL # QUESTIONS 6 PAGE 1 TOTALS TIER 2 GROUP 1 1 0 2 1 0 1 2 0 0 0 1 PAGE TOTAL # QUESTIONS 8 PAGE 2 TOTALS TIER 2 GROUP 1 0 1 0 1 0 2 0 1 1 1 2 PAGE TOTAL # QUESTIONS 9 K/A CATEGORY TOTALS: 1 1 3 3 0 3 3 1 1 3 4 TIER 2 GROUP 1 GROUP POINT TOTAL 23 REVISION 1 11/14/2003 PAGE 8 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 FEBRUARY 2004 PLANT SYSTEMS - TIER 2 GROUP 2 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # / BOTH K/A 201001 CRD Hydraulic 10 Given alarms and light status, determine the 2.9 860 BOTH NEW CFR41.5/41.6 status of the CRD Hydraulic system. q060 202001 Recirculation 06 Given plant conditions and a failure of the 3.1 861 BOTH NEW CFR41.3/41.5 Recirculation Pump Motor Generator, q061 determine final system configuration.

204000 RWCU 09 With a loss of the room cooling for the 2.8 862 BOTH NEW CFR41.4 RWCU equipment areas and temperatures, q062 determine the affects on the RWCU system.

205000 Shutdown Cooling 01 Identify the indications of a mode change 3.3 863 BOTH NEW CFR41.2/41.3/41.4/41.5/43.2 following a loss of shutdown cooling. q063 215003 IRM 219000 RHR /LPCI Suppression Pool 01 With RHR in Suppression Pool Cooling and 2.7 864 BOTH NEW ONEP Cooling Mode an extended loss of power, describe the q064 Caution CFR41.7 actions required to restore RHR to Suppression Pool Cooling. (System Vent) 234000 Fuel Handling Equipment 03 Describe the affects of a lowering Fuel Pool 3.4 865 BOTH K6.05: 3.3 NEW CFR41.4/41.9/41.12/43.4/43.7 water level on fuel handling operations. q065 239003 MSIV Leakage Control 245000 Main Turbine Gen., and Auxiliaries 259001 Reactor Feedwater 06 Describe the actions to be taken for a loss of 2.7 866 BOTH NEW CFR41.4/41.10/43.5 Plant Service Water with regard to the q066 Condensate and Feedwater systems.

PAGE 1 TOTAL TIER 2 GROUP 2 0 0 1 0 2 2 0 1 1 0 0 PAGE TOTAL # QUESTIONS 7 REVISION 1 11/14/2003 PAGE 9 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE CONT. ES-401-1 FEBRUARY 2004 PLANT SYSTEMS - TIER 2 GROUP 2 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 262002 UPS (AC/DC) 03 Describe the operation of the Static Inverter 2.6 867 BOTH NEW CFR41.7/41.10/43.5 (static switch) with an oscillating frequency q067 output of the Inverter and a loss of synchronization between sources.

263000 DC Electrical Distribution 271000 Offgas 02 Given a change in Offgas flow, determine a 2.8 868 BOTH A2.01: 3.3 NEW CFR41.4/41.10/41.13/43.4/43.5 potential cause and its affects on the plant and q068 A2.10: 3.3 Offgas System.

272000 Radiation Monitoring 05 Given a loss of power to UPS, determine the 2.9 869 BOTH NEW CFR41.10/41.11/43.4/43.5 affects on Fuel Handling Area and Fuel Pool q069 Sweep Exhaust Radiation Monitors.

286000 Fire Protection 2. Given a fire in the Turbine Building, describe 3.2 888 SRO NEW CFR41.10/41.11/41.13/43.4/43.5 3. the actions to be taken to utilize the Turbine q088 8 Building Roof hatches for venting and smoke removal.

290003 Control Room HVAC 03 Describe the basis for maintaining control of 2.7 871 BOTH NEW CFR41.4 Control Room temperature. q071 300000 Instrument Air 02 Describe the process of utilizing Service Air 2.8 870 BOTH NEW CFR41.4/41.10/43.5 to supply the Instrument Air system during a q070 loss of the Instrument Air compressors.

400000 Component Cooling Water PAGE 2 TOTALS 1 1 0 0 1 0 0 1 1 0 1 PAGE 3 TOTAL # QUESTIONS 6 PAGE 1 TOTALS 0 0 1 0 2 2 0 1 1 0 0 PAGE 1 TOTAL # QUESTIONS 7 K/A CATEGORY TOTALS: 1 1 1 0 3 2 0 2 2 0 1 TIER 2 GROUP 2 GROUP POINT TOTAL 13 REVISION 1 11/14/2003 PAGE 10 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 FEBRUARY 2004 PLANT SYSTEMS - TIER 2 GROUP 3 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # / BOTH K/A 201003 Control Rod and Drive Mechanism 215001 Traversing In-core Probe 233000 Fuel Pool Cooling and Cleanup 02 Describe the operation of the Fuel Pool 3.1 872 BOTH NEW CFR41.4/41.9 Cooling and Cleanup System with a lowering q072 level in the Spent Fuel Pool.

239001 Main and Reheat Steam 09 Determine the response of the MSIVs to a 4.1 873 BOTH NEW CFR41.4/41.7/41.9 partial actuation of isolation logic. q073 256000 Reactor Condensate 15 Given parameters and plant conditions, 3.1 874 BOTH NEW CFR41.4 determine the source of in-leakage into the q074 Reactor Condensate/ Feedwater systems.

268000 Radwaste 01 Determine the operation of floor drain sump 3.6 875 BOTH NEW CFR41.13/43.4 pumps with one pump removed from service. q075 288000 Plant Ventilation 290002 Reactor Vessel Internals K/A CATEGORY TOTALS: 0 0 0 0 0 1 1 1 0 1 0 TIER 2 GROUP 3 GROUP POINT TOTAL 4 REVISION 1 11/14/2003 PAGE 11 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-5 FEBRUARY 2004 GENERIC KNOWLEDGE AND ABILITIES TIER 3 CATEGORY C1 C2 C3 C4 TOPIC(S) IMP REC # SRO/RO RELATED ORIGIN NOTES:

/BOTH K/A CONDUCT OF OPERATIONS - Shift Turnover 2.1.3 Determine the actions required for personnel to 3.4 883 SRO MOD CFR41.10/43.5 assume shift duties during off turnover times. q083 NRC 6/2001 CONDUCT OF OPERATIONS - Procedural 2.1.20 Given a situation that requires procedure changes to 4.2 884 SRO 2.1.23: 4.0 NEW Adherence accomplish a task, determine the actions to be taken. q084 2.1.2: 4.0 CFR41.10/43.5 CONDUCT OF OPERATIONS - Procedures 2.1.21 Describe the usage and limits on procedural lineup 3.2 885 SRO 2.1.20: 4.2 NEW CFR41.10/43.5 check sheets. q085 2.2.14: 3.0 2.1.29: 3.3 CONDUCT OF OPERATIONS - Operational Mode 2.1.22 Given plant conditions, determine the plant Tech 3.3 886 SRO NEW CFR43.2 Spec Mode of operation. q086 CONDUCT OF OPERATIONS - Plant Personnel 2.1.9 Given conditions determine whose authority is 4.0 887 SRO NEW Control required to stop work in the plant. q087 CFR41.6/41.10/43.5 EQUIPMENT CONTROL - Configuration Control 2.2.15 Given a component temporarily out of normal 2.9 889 SRO 2.2.11: 3.4 NEW Configuration CFR41.10/43.5 alignment per system operating instructions, q089 control SOER 98-1 determine the tracking mechanism to be employed.

EQUIPMENT CONTROL - Maintenance Work 2.2.19 Given conditions, identify when a PASSPORT work 3.1 890 SRO NEW NEW Work Control Orders order is required to be issued. q090 System CFR41.10/43.5 EQUIPMENT CONTROL - Maintenance affecting 2.2.24 Given an inoperable component on an LCO 3.8 891 SRO NEW LCOs determine the affects of maintenance. q091 CFR41.10/43.2/43.5 EQUIPMENT CONTROL - Core Alterations 2.2.34 Determine whether an activity constitutes a Core 3.2 892 SRO 2.2.32: 3.3 NEW CFR43.6/43.7 Alteration. q092 RADIATION CONTROL - SRO Responsibilities for 2.3.3 Describe the Shift Manager responsibilities for 2.9 893 SRO MOD Hazardous Materials Systems shipments of Radioactive materials offsite. q093 NRC Transportation plan CFR41.10/41.12/43.4/43.5 12/2000 RADIATION CONTROL - Radiation Work Permits 2.3.7 Given conditions and procedures, determine 3.3 894 SRO MOD CFR41.10/41.12/43.4/43.5 applicability of radiation work permits. q094 NRC 8/2002 PAGE 1 TOTAL TIER 3 5 4 2 0 PAGE TOTAL # QUESTIONS 11 REVISION 1 11/14/2003 PAGE 12 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE CONT. ES-401-5 FEBRUARY 2004 GENERIC KNOWLEDGE AND ABILITIES TIER 3 CATEGORY C1 C2 C3 C4 TOPIC(S) IMP REC # SRO/RO RELATED ORIGIN NOTES:

/BOTH K/A EMERGENCY PROCEDURES / PLAN - AOPs 2.4.11 Given plant conditions, determine the usage of Off 3.6 895 SRO 2.4.8: 3.7 NEW and usage Normal Event Procedures and when other procedures q095 CFR41.10/43.5 take priority.

EMERGENCY PROCEDURES / PLAN - 2.4.12 During the initial phase of a security threat 3.9 896 SRO 2.4.28: 3.3 NEW Security Threat Emergency Responsibilities emergency, describe the actions to be taken by q096 Actions CFR41.10/43.5 Operations personnel and the Emergency Response Organization.

EMERGENCY PROCEDURES / PLAN - EOPs 2.4.18 Describe the bases for Emergency Director 3.6 897 SRO NEW SAPs concurrence for the transition to the SAPs and the q097 CFR41.10/43.5 yellow highlighted steps of the SAPs.

EMERGENCY PROCEDURES / PLAN - Loss of 2.4.32 Determine the actions to be taken for a loss of all 3.5 898 SRO NEW all Annunciators / Reportability Control Room annunciators. q098 CFR41.10/43.5 EMERGENCY PROCEDURES / PLAN - Health 2.4.36 Describe the purpose for having Health Physics 2.8 899 SRO NEW Physics responsibilities during an emergency personnel report to the Control Room during an q099 CFR41.10/43.5 emergency.

EMERGENCY PROCEDURES / PLAN - 2.4.43 Given unavailability of the Operational Hotline, 3.5 900 SRO NEW Turkey Point Emergency Communications Systems identify alternative methods of making Emergency q100 Hurricane Andrew CFR41.10/43.5 Notifications.

PAGE 2 TOTAL TIER 3 0 0 0 6 PAGE TOTAL # QUESTIONS 6 PAGE 1 TOTAL TIER 3 5 4 2 0 PAGE TOTAL # QUESTIONS 11 K/A CATEGORY TOTALS: 5 4 2 6 TIER 3 GROUP POINT TOTAL 17 REVISION 1 11/14/2003 PAGE 13 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

ES-301 Administrative Topics Outline Form ES-301-1 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 2/9/2004 - 2/11/2004 Examination Level (circle one): RO / SRO Operating Test Number: __1___

Administrative Describe method of evaluation: Knowledge IMP Additional ORIGIN NOTES Topic/Subject 1. ONE Administrative JPM, OR / Ability K/As Description 2. TWO Administrative Questions A.1 Technical JPM GJPM-SRO-ADM50 2.1.12 4.0 2.2.23: 3.8 MOD Different Specifications 2.2.22: 4.1 component Given a component, determine Limiting using Condition for Operations and complete ESOMS entry into ESOMS. computer CFR 55.45 (a)12 &

13 Plant Chemistry JPM GJPM-OP-ADM-52 2.1.34 2.9 2.1.6: 4.3 NEW CFR 55.45 (a)12 &

Given a chemistry report and procedures, 13 determine the plant conditions and actions to be taken.

A.2 Pre-Maintenance JPM GJPM-SRO-ADM51 2.2.21 3.5 NEW PCRS Operability CFR 55.45 Given a Condition Report, determine the (a)12 &

operability requirements for the 13 component and enter into PCRS system.

A.3 Radiation JPM GJPM-SRO-ADM33 2.3.1 3.0 2.3.4: 3.1 BANK CFR 55.45 Control (a)9 & 10 2.3.2: 2.9 Perform required actions to access the NRC Controlled Access Area (CAA), determine 6/2001 requirements to enter a High Contamination Area and authorization required, and exit the CAA.

A.4 Emergency Plan JPM GJPM-SRO-A&E55 2.4.41 4.1 2.4.30: 3.6 NEW CFR Action Levels 2.4.40: 4.0 55.45 Given conditions, determine the 2.4.28: 3.3 (a)11 appropriate emergency classification, Security actions to be taken for a security Threat threat compromising the Remote Shutdown Panels and complete the required notification form.

REVISION 1 12/2/2003

ES-301 Individual Walk-Through Test Outline Form ES-301-2 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 2/9/2004 - 2/9/2004 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: ___1___

System / JPM Title / Type Codes* Safety Knowledge IMP. Additional ORIGIN NOTES Function / Ability K/As B.1. CONTROL ROOM SYSTEMS

1. 205000 SHUTDOWN COOLING SYSTEM (RHR) 4 A4.01 3.7 A4.02: 3.5 BANK CFR 55.45(a)

(D)(S)(A)(L) A4.03: 3.5 1; 3, 4; Startup RHR in Shutdown Cooling (E12-F053x A4.09: 3.1 NRC 5; 6 & 7 fail on stroke) A2.10: 2.9 3/1998 A2.12: 3.0 GJPM-RO-E1212 A1.02: 3.2

2. 262001 AC ELECTRICAL DISTRIBUTION 6 A4.01 3.7 A4.02: 3.4 MOD CFR (M)(S) A4.04: 3.7 55.045(a)6 Distribute loads between Service Transformers A4.05: 3.3 NRC & 8 11 & 21 2.1.31: 3.9 8/2002 2.1.30: 3.4 GJPM-RO-R2731
3. 212000 REACTOR PROTECTION SYSTEM (RPS) 7 A4.17 4.1 295037 BANK CFR (D)(C) EA1.01: 4.6 55.45(a)8 Defeat RPS Scram Signals per EP-2 Attachment 295015 NRC 19 AA1.02: 4.2 6/2001 2.1.30: 3.4 GJPM-RO-EP031 2.1.20: 4.2
4. 218000 AUTOMATIC DEPRESSURIZATION 3 A4.01 4.4 A4.02: 4.2 BANK CFR SYSTEM (ADS) 55.45(a)8 (D)(S)(A)

Manually initiate ADS. (No pump permissive) NRC 3/1998 GJPM-RO-E2222

5. 223001 PRIMARY CONTAINMENT SYSTEM 5 A2.11 3.8 A1.08: 3.6 BANK CFR (D)(S) 209002 55.45(a)8 Raise Suppression Pool water level using HPCS A4.01: 3.7 NRC 8/2002 A4.04: 3.1 lowered GJPM-RO-E2205 A4.09: 3.5 level
6. 202002 RECIRCULATION FLOW CONTROL SYST. 1 A2.08 3.3 A1.08: 3.4 BANK CFR (D)(S) 2.1.30: 3.4 55.45(a)

Recover Recirculation Flow Control Valve 2; 6 & 8 following an automatic runback.

GJPM-RO-B3311 REVISION 1 12/2/2003

Facility: GRAND GULF NUCLEAR STATION Date of Examination: 2/9/2004 - 2/9/2004 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: ___1___

System / JPM Title / Type Codes* Safety Knowledge IMP. Additional ORIGIN NOTES Function / Ability K/As B.1. CONTROL ROOM SYSTEMS (cont)

7. 259001 REACTOR FEEDWATER SYSTEM 2 A4.04 2.9 A4.05: 3.9 NEW CFR (N)(S)(L)(A) A2.07: 3.8 55.45(a)

Shift from Long Cycle Cleanup to Startup A3.03: 3.2 1; 3; 4; 6 Level Control with Condensate (S/U Level A3.04: 3.7 & 8 Control Valve fails full OPEN). A4.01: 3.5 2.1.30: 3.4 GJPM-RO-N2102 259002 A1.05: 2.9 A4.03: 3.6 B.2. FACILITY WALK-THROUGH

8. 286000 FIRE PROTECTION SYSTEM 8 A4.06 3.4 BANK CFR (D)(P)(A) 55.45(a)

Perform a local start of a diesel driven fire NRC 6 & 8 pump (failure of first manual local bank 8/2002 Abnormal start).

GJPM-RO-P6402

9. 295019 LOSS OF INSTRUMENT AIR 8 AA1.01 3.3 BANK CFR (D)(P)(R) 55.45(a) 8 & 9 Lineup makeup Nitrogen to the ADS Valve NRC GGNS Scram Accumulators per ONEP. 6/2001 4/2003 Emergency/

GJPM-NLO-P5301 Abnormal

10. 295016 CONTROL ROOM ABANDONMENT 2 AA1.06 4.1 2.1.30: 3.4 NEW CFR (N)(P)(A) AK2.01: 4.5 55.45(a)

Startup RCIC from the Remote Shutdown Panel AK3.03: 3.7 4; 6; & 8 to control RPV Water Level (Failed flow AA1.07: 4.3 Other Safety controller). AA2.02: 4.3 Function 7 Emergency/

GJPM-RO-C6106 Abnormal

  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (P)lant, (R)CA REVISION 1 12/2/2003

ES-301 Administrative Topics Outline Form ES-301-1 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 2/9/2004 - 2/11/2004 Examination Level (circle one): RO / SRO Operating Test Number: __1___

Administrative Describe method of evaluation: Knowledge IMP Additional ORIGIN NOTES Topic/Subject 1. ONE Administrative JPM, OR / Ability K/As Description 2. TWO Administrative Questions A.1 Technical JPM GJPM-SRO-ADM50 2.1.12 4.0 2.2.23: 3.8 MOD Different Specifications 2.2.22: 4.1 component Given a component, determine Limiting using Condition for Operations and complete ESOMS entry into ESOMS. computer CFR 55.45 (a)12 &

13 Plant Chemistry JPM GJPM-OP-ADM-52 2.1.34 2.9 2.1.6: 4.3 NEW CFR 55.45 (a)12 &

Given a chemistry report and procedures, 13 determine the plant conditions and actions to be taken.

A.2 Pre-Maintenance JPM GJPM-SRO-ADM51 2.2.21 3.5 NEW PCRS Operability CFR 55.45 Given a condition report, determine the (a)12 &

operability requirements for the 13 component and enter into PCRS system.

A.3 Radiation JPM GJPM-SRO-ADM33 2.3.1 3.0 2.3.4: 3.1 BANK CFR 55.45 Control (a)9 & 10 2.3.2: 2.9 Perform required actions to access the NRC Controlled Access Area (CAA), determine 6/2001 requirements to enter a High Contamination Area and authorization required, and exit the CAA.

A.4 Emergency Plan JPM GJPM-SRO-A&E55 2.4.41 4.1 2.4.30: 3.6 NEW CFR Action Levels 2.4.40: 4.0 55.45(a)

Given conditions, determine the 2.4.28: 3.3 11 appropriate emergency classification, Security actions to be taken for a security Threat threat compromising the Remote Shutdown Panels and complete the required notification form.

REVISION 1 12/2/2003

ES-301 Individual Walk-Through Test Outline Form ES-301-2 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 2/9/2004 - 2/9/2004 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: ___1___

System / JPM Title / Type Codes* Safety Knowledge IMP. Additional ORIGIN NOTES Function / Ability K/As B.1. CONTROL ROOM SYSTEMS

1. 205000 SHUTDOWN COOLING SYSTEM (RHR) 4 A4.01 3.7 A4.02: 3.5 BANK CFR 55.45(a)

(D)(S)(A)(L) A4.03: 3.5 1; 3; 4 Startup RHR in Shutdown Cooling (E12-F053x A4.09: 3.1 NRC 5; 6 & 7 fail on stroke) A2.10: 2.9 3/1998 A2.12: 3.0 GJPM-RO-E1212 A1.02: 3.2

2. 262001 AC ELECTRICAL DISTRIBUTION 6 A4.01 3.7 A4.02: 3.4 MOD CFR (M)(S) A4.04: 3.7 55.45(a)

Distribute loads between Service Transformers A4.05: 3.3 NRC 6 & 8 11 & 21 2.1.31: 3.9 8/2002 2.1.30: 3.4 GJPM-RO-R2731

3. 212000 REACTOR PROTECTION SYSTEM (RPS) 7 A4.17 4.1 295037 BANK CFR (D)(C) EA1.01: 4.6 55.45(a)8 Defeat RPS Scram Signals per EP-2 Attachment 295015 NRC 19 AA1.02: 4.1 6/2001 2.1.30: 3.4 GJPM-RO-EP031 2.1.20: 4.2 B.2. FACILITY WALK-THROUGH
4. 295019 LOSS OF INSTRUMENT AIR 8 AA1.01 3.3 BANK CFR (D)(P)(R) 55.45(a) 8 & 9 Lineup makeup Nitrogen to the ADS Valve NRC GGNS Scram Accumulators 6/2001 4/2003 Emergency/

GJPM-NLO-P5301 Abnormal

5. 295016 CONTROL ROOM ABANDONMENT 2 AA1.06 4.1 2.1.30: 3.4 NEW CFR (N)(P)(A) AK2.01: 4.5 55.45(a)

Startup RCIC from the Remote Shutdown Panel AK3.03: 3.7 4; 6; & 8 to control RPV Water Level (Faulted) AA1.07: 4.3 Other Safety AA2.02: 4.3 Function 7 GJPM-RO-C6106 Emergency/

Abnormal

  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (P)lant, (R)CA REVISION 1 12/2/2003

Appendix D Scenario Outline Form ES-D-1 Facility: GRAND GULF NUCLEAR STATION Scenario No.: 1 Op-Test No.: Day 1 Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Complete a shift of Reactor Recirculation Pumps to Fast Speed.
2. Take actions in response to a Control Rod Drift and complete actions of the CRD Malfunctions ONEP.
3. Respond to a trip of RPS A MG set and the implications of having both RPS buses on Alternate Source of power.
4. Make determination of multiple Control Rod Drifts following insertion and disarming CRD and taking actions for multiple Control Rod Drifts per CRD Malfunctions ONEP.
5. Take actions per the EOPs in response to an ATWS and mitigate the consequences of the ATWS with no Main Steam Bypass Valves.
6. Take actions for a failure of Standby Liquid Control to inject to the Reactor during an ATWS.

Initial Conditions: Reactor Power is at 34 %.

INOPERABLE Equipment APRM H is INOP due to a failed power supply card RHR C Pump is tagged out of service for motor oil replacement CCW Pump B is tagged out of service for pump seal replacement RPS B Motor Generator is out of service for EPA circuit breaker replacement, RPS B is on its Alternate Source.

Service Air Compressor B is in service with Service Air Compressor A tagged out of service for oil replacement.

Appropriate clearances and LCOs are written.

Turnover: The plant is operating at 34% power. Reactor Recirculation Pump A has been shifted to Fast speed. Continue operations to shift Reactor Recirculation Pump B to Fast speed at step 5.11.4 of IOI-2. There are scattered thundershowers reported in the Tensas Parish area.

REVISION 2 1/19/2004

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

Event 10CFR K/A Event Event No. 55.45(a) Type* Description 1 202002 R (RO) Shift Reactor Recirculation Pump B to fast speed.

2, 3, 4, 5, A4.07; A4.08; 6, 8 A4.09 N (SS) (SOI 04-1-01-B33-1 section 4.2) 202001 A4.01; A4.02 A1.02; A1.07 2 2.4.49; 2.4.4 C(RO) Respond to Control Rod Drift. Perform actions per ONEP 05-1-02-IV-1.

3, 4, 5, 6, 201005 8 A2.13; A3.0; A4.01 Isolate/valve out of service the affected control rod.

201003 A2.03; A3.01 3 2.1.32; 2.1.33 Respond to trip of RPS A Motor Generator trip. Complete Technical 6, 8 212000 A2.01; K3.05 Specification/procedural determinations.

4 2.4.4; 2.4.49 C(RO) Recognize and respond to multiple control rod drifts and insert a manual Reactor 2, 3, 4, 5, 201005 6, 8 A2.13; A3.0; A4.01 SCRAM per ONEP 05-1-02-IV-1.

201003 A2.03; A3.01 5 2.4.4; 2.4.49 M (ALL) Upon Reactor Scram recognize the failure of all control rods to fully insert and 3, 4, 5, 6, 295037 7, 8 EA1.0; EA2.0 take actions per EOPs for ATWS.

241000 A2.03 C Recognize the failure of Main Steam Bypass Valves to open and control reactor 3, 4, 6, 7, 239002 8 A4.01; A4.05 (BOP) pressure using SRVs within specified band.

3, 6, 8 212000 A2.02; A4.14; Recognize the loss of both Alternate Divisions of RPS EPAs when Low Pressure A4.16; A4.17 ECCS is manually initiated and restore power to RPS to allow insertion of control 295037 EA1.01; EA1.08 rods.

295037 EA1.04; C Recognize the failure of Standby Liquid Control to meet the parameters to inject 3, 4, 6, 8 EA1.10 211000 A2.01 (BOP) into the Reactor when initiated and actions taken for Alternate Boron Injection.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor REVISION 2 1/19/2004

All events include 55.45(a) 12 & 13 K/A 2.1.30; 2.1.31; 2.4.45; 2.4.46; 2.4.47; and 2.4.48 Critical Tasks

- Insert manual scram on multiple Control Rod Drifts.

- Inject Standby Liquid Control prior to Suppression Pool Temperature reaching 110 °F.

- Identify the need for Alternate Standby Liquid Control injection.

- 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 before reactor level reaches -192.

- Insert Control Rods in response to ATWS conditions.

REVISION 2 1/19/2004

Appendix D Scenario Outline Form ES-D-1 Facility: GRAND GULF NUCLEAR STATION Scenario No.: 2 Op-Test No.: Day 2 Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Raise Reactor Power by withdrawing control rods.
2. Perform operator actions for a stuck control rod per ONEP.
3. Startup 2nd Reactor Feed Pump.
4. Respond to a failure of ESF UPS bus 1Y89 (inverter 1Y87).
5. Respond to a momentary loss of Grid per ONEPs.
6. Respond to a failure of Feedwater Line in the Drywell, initiate a reactor scram based on rising Drywell Pressure per EOPs.
7. Respond to a failure of Division 2 ECCS to initiate.
8. With a small break LOCA in the Drywell and reduced injection systems maintain reactor level per the EOPs.

Initial Conditions: Reactor Power is at 44 % bringing the plant up following an outage; Reactor Recirculation pumps are in Fast Speed at 60 % core flow; a single Reactor Feed Pump in three element Master Level Control.

INOPERABLE Equipment APRM H is INOP due to a failed power supply card RHR C is tagged out of service for motor oil replacement CCW Pump B is tagged out of service for pump seal replacement RPS B Motor Generator is out of service for EPA circuit breaker replacement, RPS B is on its Alternate Source.

Service Air Compressor B is in service with Service Air Compressor A tagged out of service for oil replacement.

Appropriate clearances and LCOs are written.

Turnover: Continue to bring the plant to full power per IOI-2. There are scattered thundershowers reported in the Tensas Parish area.

REVISION 1 11/26/2003

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

Event K/A Event Event 10CFR No. Type* Description 55.45(a) 201005 1 2, 3, 4, 5, A3.0; A4.0 R(RO) Withdraw control rods to raise power.

6 (Control Rod Pull Sheet & IOI 03-1-01-2) 2 201005 A3.0; A4.0 C (RO, Control Rod 24-49 is stuck, un-stick control rod per ONEP. (ONEP 05-1 4, 5, 6, 8 201003 A2.01 201001 A4.03; A4.04 BOP) IV-1) 2.4.4; 2.4.49 3 259001 N (RO) Startup 2nd Reactor Feed Pump 2, 4, 5, 6, A4.02; A4.01; A4.04; 8 A4.05; A4.07 (SOI 04-1-01-N21-1) 259002 A4.01; A4.02; A4.03; A4.06 4 2.1.33; 2.2.22 C (RO, Respond to a trip of ESF UPS Bus 1Y89 and Inverter 1Y87.

3, 4, 8 262002 A1.01; K3.0 BOP) (Multiple SOIs and ARIs) 295003 AA1.0; AA2.0 5 3, 5, 6, 8 262001 A1.0; A2.0; M (ALL) Respond to momentary Loss of Grid.

A3.0; A4.0 (ONEP 05-1-02-I-4 & SOI Various) (GGNS Event 4/2003) 2.4.4; 2.4.49 Single Control Rod Stuck withdrawn.

295024 EA1.0; EA2.0 C (ALL) Recirc Line B ruptures in the Drywell with leakage from the reactor.

3, 4, 5, 6, 295031 EA1.0; EA2.0 7, 8, 11 2.1.2 I (BOP) Failure of Division 2 ECCS to automatically initiate on High Drywell Pressure 3, 4, 5, 8 295024 EA1.0 206002 C (BOP) HPCS injection valve failure to open on initiation 3, 5, 7 A1.01; A2.03; A2.08; A3.01; A4.03

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor REVISION 1 11/26/2003

All events include 55.45(a) 12 & 13 K/A 2.1.30; 2.1.31; 2.4.45; 2.4.46; 2.4.47; 2.4.48 Critical Tasks

- Recognize failure of Division 2 to initiate and manually initiate Division 2

- Restore power and reestablish feed through Feedwater or RCIC or lower reactor pressure to allow injection from low pressure systems

- Upon receipt of second control rod drift, manually scram the reactor.

REVISION 1 11/26/2003

Appendix D Scenario Outline Form ES-D-1 Facility: GRAND GULF NUCLEAR STATION Scenario No.: 3 Op-Test No.: Day 2 Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Raise Reactor Power by withdrawing control rods.
2. Start 2nd Circulating Water Pump.
3. Respond to an EHC failure.
4. Respond to a loss of Main Condenser Vacuum.
5. Respond to an automatic and manual scram function failure ATWS ARI/RPT will insert control rods with two control rods stuck withdrawn.
6. Respond to a steam leak in the Auxiliary Building Steam Tunnel and a failure of Group 1 to isolate.
7. Take actions per the EOPs in response to two stuck control rods following a Reactor Scram.
8. Take actions per EOPs to control RPV parameters with a failure of the MSIVs to isolate the steam leak.

Initial Conditions: Reactor Power is at 45 % continuing power ascension to rated conditions.

INOPERABLE Equipment APRM H is INOP due to a failed power supply card RHR Pump C is tagged out of service for motor oil replacement CCW Pump B is tagged out of service for pump seal replacement RPS B Motor Generator is out of service for EPA circuit breaker replacement, RPS B is on its Alternate Source.

Service Air Compressor B is in service with Service Air Compressor A tagged out of service for oil replacement.

Appropriate clearances and LCOs are written.

Turnover: Continue power ascension. There are scattered thundershowers reported in the Tensas Parish area.

REVISON 1 11/26/2003

Appendix D Scenario Outline Form ES-D-1 Scenario 3 Day 2 (Continued)

Event K/A Event Event 10CFR No. Type* Description 55.45(a) 1 201005 R Raise reactor power by withdrawing control rods.

2, 3, 5, 6, A3.0; A4.0 8 (RO) (IOI 03-1-01-2 and Control Rod Movement Sheet) 2 2.1.30; 2.1.31 N Start 2nd Circulating Water.

2, 6, 8 (BOP) (SOI 04-1-01-N71-1) 3 241000 A1.11; A2.06 Respond to an EHC leak.

3, 5, 8 (ARI 04-1-02-1H13-P680) 4 241000 A2.07; A3.08; C Respond to a lowering Main Condenser Vacuum.

3, 4, 5, 6, A3.10 8 (BOP) (ONEP 05-1-02-V-8) 239001 A2.08 295002 AA1.0; AA2.0 295006 AA1.01; AA1.07; 5 2, 3, 4, 5, AA2.01; AA2.05 C Recognize a failure to automatically scram and manually scram the reactor.

6, 8 295037 EA1.03 (RO) 6 239001 A2.03; A2.07; M Recognize and respond to a steam leak in the Auxiliary Building Steam 3, 4, 6, 8, A2.11; A2.12 (ALL) Tunnel.

10 239001 A3.01 I Recognize the failure of Group 1 to automatically isolate and take actions to 3, 4, 6, 8, 223002 A1.02; A4.02 10 (BOP) isolate the Main Steam Lines (ONEP 05-1-01-III-5) 295032 EA1.01; EA1.05; Recognize the failure of a single Main Steam line to isolate and take actions 3, 5, 6 EA2.01; EA2.03 for mitigation of the leak.

295015 AA1.01; AA1.02; C Recognize the failure of two control rods to fully insert on the Reactor Scram.

2, 3, 4, 5 AA2.01; AA2.02 (RO)

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor REVISON 1 11/26/2003

All events include 55.45(a) 12 & 13 K/A 2.1.30; 2.1.31; 2.4.45; 2.4.46; 2.4.47; and 2.4.48 Critical Tasks

  • Manually scram the reactor.

REVISON 1 11/26/2003

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