ML12054A190

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Final Outlines (Folder 3)
ML12054A190
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 12/20/2011
From: Caruso J
Operations Branch I
To: Goff C
Susquehanna
Jackson D
Shared Package
ML110190451 List:
References
TAC U01842
Download: ML12054A190 (37)
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Text

ES-401 BWR Examination Outline Form ES-401-1 Facility: Susquehanna Date of Exam: Jan RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

  • Total
1. 1 3 4 2 3 5 3 20 7 Emergency &

Abnormal Plant Evolutions 2

Tier Totals 2

5 1

5 1

3 WAdE N/A 1 4

7 27 3

10

~~

1 2 1 2 4 233 26 5 2.

Plant 2 3 1 1 2 2 0 0 1 0 12 3 Systems

~

Tier Tot 2 "I ~14 4 2 3 3 2 38 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 3 2 3 Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the Tier Totals" in each KIA category shall not be less than two).
2. The point total for each group and tier in the proposed outline must match that specified in the table.

The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 pOints.

3. Systems/evolutions within each group are identified on the asscciated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant-specific priority, only those KlAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7.* The generic (G) KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KlAs.

8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to KlAs that are linked to 10 CFR 55.43.

ES-401 2 Form ES-401-1 E::C' An

  • ES-401 BWR Examination Outline F Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G KIA Topic(s) IR #

1 2 3 1 2 295001 Partial or Complete Loss of Forced X AK1.02 Knowledge of the operational 3.3/3.5 1 Core Flow Circulation / 1 & 4 implications of power/flow distribution as it applies to Partial or Complete Loss of Forced Core Flow Circulation AK3.01 Knowledge of the reasons for 295003 Partial or Complete Loss of AC / 6 X 3.3/3.5 2 the following responses as they apply to PARTIAL OR OMPLETE LOSS OF A.C. POWER: Manual and auto bus transfer 295004 Partial or Total Loss of DC Pwr /6 X AA2.04 Ability to determine and/or 3.2/3.3 3 interpret system lineups as they apply to partial or com plete loss of DC power 295005 Main Turbine Generator Trip / 3 X AK1.03 Knowledge of the operational 3.5/3.7 4 implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP: Pressure effects on reactor level 295006 SCRAM / 1 X G2.4.9 Knowledge of low 3.8/4.2 5 power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies 295016 Control Room Abandonment / 7 X AA 1.06 Ability to operate and/or 4.0/4.1 6 monitor the following as they apply to CONTROL ROOM ABANDONMENT:

Reactor water level 295018 Partial or Total Loss of CCW / 8 X AK1.01 Knowledge of the operational 3.5/3.6 7 implications of the effects on component/system operations as it applies to Partial or Complete Loss of Component Cooling Water 295019 Partial or Total Loss of Ins!. Air /8 X 2.4.11 Knowledge of abnormal 4.0/4.2 8 condition procedures 295021 Loss of Shutdown Cooling / 4 X AA 1.04 Ability to operate and or 3.7/3.7 9 monitor Alternate Heat Removal Methods as they apply to loss of Shutdown Cooling 295023 Refueling Acc / 8 X AK2.03 Knowledge of the 3.4/3.6 10 interrelations between REFUELING ACCIDENTS and the following:

Radiation monitoring equipment

295024 High Drywell Pressure 15 X EA2.04 Ability to determine and/or 3.913.9 11 interpret Suppression chamber pressure as it applies to high drywell pressure 295025 High Reactor Pressure I 3 X EK2.08, Knowledge of the 3.7/3.7 12 interrelations between HIGH REACTOR PRESSURE and the following: Reactor/turbine pressure regulating system 295026 Suppression Pool High Water X EK3.02 Knowledge of the reasons for 3.9/4.0 13 Temp./5 Suppression Pool Cooling as it applies to Suppression Pool high water tem perature 295028 High Drywell Temperature 15 X 2.4.6 Knowledge of the EOP mitigation 3.7/4.7 14 strategies 295030 Low Suppression Pool WIr Lvl/5 X EK2.07 Knowledge of the 3.5/3.8 15 interrelations between Low Suppression Pool water level and Downcomer submergence 295031 Reactor Low Water Level 12 X EA2.04 Ability to determine and/or 4.6/4.8 16 interpret the following as they apply to REACTOR LOW WATER LEVEL:

Adequate core cooling EK2.05 Knowledge of the 295037 SCRAM Condition Present X 4.0/4.1 17 and Reactor Power Above APRM interrelations between SCRAM Downscale or Unknown 11 condition present and reactor power above APRM downscale or unknown and the CRD hydraulic system 295038 High Off-site Release Rate 1 9 X EA1.03 Ability to operate and/or 3.7/3.9 18 monitor the following as they apply to HIGH OFF-SITE RELEASE RATE:

Process liquid radiation monitoring system.

600000 Plant Fire On Site 1 8 X AA2.17 Ability to determine and 3.113.6 19 interpret systems that may be affected by the fire as it applies to Plant Fire on Site 700000 Generator Voltage and Electric Grid X AA2.01 Ability to determine and/or 3.5/3.6 20 Disturbances I 6 interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

Operating point on the generator I capability curve I KJA Caterilo!:X Totals: 3 4 2 3 5 3 Group Point Total: 20/7

ES-401 3 Form ES-401-1 F'~I I ES-401 BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO I SRO)

E/APE # I Name I Safety Function K K K A A G KIA Topic(s) 1 2 3 1 2 295002 Loss of Main Condenser Vac 1 3 295007 High Reactor Pressure I 3 295008 High Reactor Water Levell 2 295009 Low Reactor Water Levell 2 X AK 2.03 Knowledge of the interrelations 3.1/3.2 21 between Low Reactor Water Level and the recirculation system 295010 High Drywell Pressure /5 X AA 1.02 Ability to operate and/or 3.6/3.6 22 monitor the following as they apply to HIGH DRYWELL PRESSURE: Drywell floor and equipment drain sumps.

295011 High Containment Temp 15 295012 High Drywell Temperature /5 X AK1.01 Knowledge of the operational 3.313.5 23 implications of the pressure/temperature relationship is it applies to High Drywell Temperature 295013 High Suppression Pool Temp./5 295014 Inadvertent Reactivity Addition 11 X AA2.03 Ability to determine and/or 4.0/4.3 24 interpret the following as they apply to INADVERTENT REACTIVITY ADDITION: Cause of reactivity addition.

295015 Incomplete SCRAM /1 295017 High Off-site Release Rate / 9 295020 Inadvertent Cont. Isolation 1 5 & 7 295022 Loss of CRD Pumps /1 X G2.1.23 Ability to perform specific 4.3/4.4 C

25 system and integrated plant procedures during all modes of plant operation 295029 High Suppression Pool Wtr Lvi / 5 295032 High Secondary Containment X EK1.02 Knowledge of the operational 3.6/4.0 26 Area Temperature 15 implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Radiation releases ontainment 295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure / 5

295036 Secondary Containment High X EK3.01 Knowledge of the reasons for 2.6/2.8 27 Sump/Area Water Level/5 emergency depressurization as it applies to Secondary Containment High Sump/Area Water Level 500000 HiQh CTMT Hydroaen Conc. I 5 I I

~

2 I I I I Group Point Total: 7/3 II

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO I SRO)

System # I Name K K K K K K A A A A G KIA Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 203000 RHR/LPCI: Injection X K5.01 Knowledge of the operational 2.7/2.9 28 Mode implications of the following concepts as they apply to RHRlLPCI: Testable check valve operation 205000 Shutdown Cooling X K6.01 Knowledge of the effect that a 3.3/3.4 29 loss or malfunction of AC. electrical power will have on the Shutdown Cooling System (RHR Shutdown Cooling Mode) 206000 HPCI X K4.09 Knowledge of HIGH 3.8/3,9 30 PRESSURE COOLANT INJECTION SYSTEM design feature(s) and/or interlocks which provide for the following: Automatic flow control:

BWR-2,3,4 207000,isolatlon (Emergency) I Condenser ' .** , .'. ",

209001 LPCS X K6.04 Knowledge of the effect that a 2.8/2.9 32 loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM: D.C.

power 209001 LPCS X 2.4.50 Ability to verify system alarm 4.2/4.0 33 setpoints and operate controls identified in he alarm response manual 209002 HPCS ' '"

211000 SLC x A2.03 Ability to (a) predict the 3.2/3.4 31 impacts of the following on the Standby Liquid Control System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

AC. Power Failures

211000 SLC X K4.08 Knowledge of STANDBY 4.2/4.2 34 LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following:

System initiation upon operation of SBLC control switch.

212000 RPS X K2.01 Knowledge of electrical power 3.2/3.3 35 supplies to the RPS motor-generator sets 2150031RM x Kl.Ol Knowledge of the physical 3.9/3.9 36 connections and/or cause-effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following: RPS 215004 Source Range Monitor X A3.03 Ability to monitor automatic 3.6/3.5 37 operations of the Source Range Monitor (SRM) System including RPS status 215005 APRM I LPRM X K4.02 Knowledge of AVERAGE 4.1/4.2 38 POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM design feature(s) andlor interlocks which provide for the following: Reactor SCRAM signals 215005 APRM I LPRM X A3.08 Ability to monitor automatic 3.7/3.6 39 operations of the Average Power Range Monitor/Local Power Range Monitor System including control rod block status 217000 RCIC X Kl.Ol Knowledge of the physical 3.513.5 40 connections and/or cause-effect relationships between REACTOR CORE ISOLAnON COOLING SYSTEM (RCIC) and the following:

Condensate storage and transfer system 217000 RCIC X A2.05 Ability to (a) predict the 3.3/3.3 41 impacts of D.C. power loss on the Reactor Core Isolation Cooling System (RCIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations 218000 ADS X 2.4.31 Knowledge of annunciator 4.2/4.1 42 alarms, indications, or response procedures.

223002 PCIS/Nuclear Steam X A1.02 Ability to predict and/or 3.7/3.7 43 Supply Shutoff monitor changes in parameters associated with operating the Primary Containment Isolation System/Nuclear Steam Supply Shut-Off controls including: Valve closures 239002 SRVs X K3.03 Knowledge of the effect that a 4.3/4.4 44 loss or malfunction of the RELIEF/SAFETY VALVES will have on following: Ability to rapidly depressurize the reactor 259002 Reactor Water Level X K5.0I Knowledge of the operational 3.113.1 45 Control implications of Foxboro controller operation as it applies to Reactor Water Level Control System 261000 SGTS X A3.02 Ability to monitor automatic 3.2/3.1 46 operations of the STANDBY GAS TREATMENT SYSTEM including:

Fan start 262001 AC Electrical X K3.01 Knowledge of the effect that a 2.7/2.9 47 Distribution loss or malfunction of the A.C.

Electrical Distribution will have on major system loads 262002 UPS (AC/DC) X K6.03 Knowledge of the effect 2.7/2.9 74 that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C.lD.C.): D.C.

electrical_r:>ower 263000 DC Electrical X A1.0I Ability to predict and/or 2.5/2.8 49 Distribution monitor changes in parameters associated with operating the D.C.

Electrical Distribution controls including battery charging/discharging rate A2.07 Ability to (a) predict the 264000 EDGs X 3.5/3.7 48 impacts of the following on the EMERGENCY GENERATORS (DIESEL/JET) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of off-site power during full-load testing A4.0I Ability to manually operate 264000 EDGs X 3.3/3.4 51 and/or monitor in the control room:

adjustment of exciter voltage

300000 Instrument Air K4.02 Knowledge of (INSTRUMENT 3.0/3.0 50 AIR SYSTEM) design feature(s) and or interlocks which provide for the following: Cross-over to other air systems 400000 Component Cooling X K6.05 Knowledge of the effect that a 3.0/3.1 53 Water loss or malfunction of the following will have on the CCWS: Pumps 2 -' 1 2 Group Point Total: 26 15

ES-401 5 Form ES-401-1

~

BWR Examination Outline Form ES-401-1 Plant S stems Tier 21Group 2 (RO I SRO)

~~ K 3

K A 2

A A 3 4 G KJA Topic(s) IR #

201001 CRD Hydraulic 201002 RMCS X 3,213.2 56 K1.01 Knowledge of the physical connections and/or causeeffect relationships between REACTOR MANUAL CONTROL SYSTEM and the following: Control rod drive hydraulic system 201003 Control Rod and Drive X 52 Mechanism K1.04 Knowledge of the physical connections 3,0/3,0 and/or cause effect relationships between CONTROL ROD AND DRIVE MECHANISM and the following: Reactor vessel I§ 201006RWM x A2.07 Ability to (a) predict 2.5/2.8 55 the impacts of RWM hardware/software failure on the Rod Worth Minimizer System (RWM);

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations 202001 Recirculation X 3,113,9 54 K3.05 Knowledge of the effect that a loss or malfunction of the RECIRCULATION SYSTEM will have on following: Recirculation system MG sets

~CUlatiOn Flow Control

204000 RWCU x A4.08 Ability to manually 3.4/3.4 57 operate and/or monitor in the control room: reactor water level 214000 RPIS 215001 Traversing In-core Probe 215002 RBM lear Boiler Inst.

219000 RHRlLPCI: Torus/Pool Cooling X 3.113.3 59 Mode K2.02 Knowledge of electrical power supplies to the following: Pum~s

~01 Prim.~ CTMT "d A", II 01 RHRlLF'CI: CTMT Spray Mode 230000 RHRlLPCI: Torus/Pool Spray X 4.0/3.9 58 Mode A4.06 Ability to manually operate and/or monitor in the control room: Valve logic reset following automatic initiation of LPCIIRHR in injection mode 233000 Fuel Pool Cooling/Cleanup X 2.9/3.2 61 K4.06 Knowledge of Fuel Pool Cooling and Clean-Up design feature(s) and/or interlocks which provide for the following:

Maintenance of adequate pool level 234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSIV Leakage Control 241000 ReactorlTurbine Pressure I Reaulator 245000 Main Turbine Gen. I Aux. X 2.8/3.1 60 K5.02 Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS: Turbine o~eration and limitations 00 Reactor Condensate

259001 Reactor Feedwater X K1.05 Knowledge of the 3.2/3.2 63 physical connections and/or cause-effect relationships between Reactor Feedwater System and the following:

Condensate system 268000 Radwaste 271000 Offgas 272000 Radiation Monitoring 286000 Fire Protection X 3.3/3.5 62 K4.02 Knowledge of FIRE PROTECTION SYSTEM design feature(s) and/or interlocks which provide for the following:

Automatic system initiation 288000 Plant Ventilation X 3.2/3.4 65 K5.02 Knowledge of the operational implications of the following concepts as they apply to Plant Ventilation Systems:

Differential Pressure control 290001 Secondary CTMT 290003 Control Room HVAC 290002 Reactor Vessel Internals KIA Category Point Totals: 3 1 1 2 2 0 0 1 0 2 0 Group Point Total: 12/3

RO OUTLINE Category KIA # Topic RO IR #

2.1.30 Ability to locate and operate components, 4.4/4.0 64 including local controls 1.

Conduct 2.1.32 Ability to explain and apply system limits and 3.8/4.0 67 of Operations precautions Subtotal 2 2.2.22 Knowledge of limiting conditions for operations 4.0/4.7 66 and safety limits

2. 2.2.13 Knowledge of clearance and tagging 4.114.3 69 Equipment procedures Control 2.2.3 Knowledge of the design, procedural, and 3.8/3.9 70 operational differences between units Subtotal 3 2.3.4 Knowledge of radiation exposure limits under 3.2/3.7 68 normal or emergency conditions
3. 2.3.5 Ability to use radiation monitoring systems, 3.9/2.9 71 Radiation such as fixed radiation monitors and alarms, Control portable survey instruments, personnel monitoring equipment, etc Subtotal 2 2.4.45 Ability to prioritize and interpret the significance 4.114.3 73 of each annunciator or alarm 4.

Emergency 2.4.2 Knowledge of system set points, interlocks and 4.6/4.8 72 Procedures I automatic actions associated with EOP entry Plan conditions 2.4.31 Knowledge of annunciator alarms, indications, 4.2/4.1 75 or response procedures Subtotal 3 I Tier 3 PointTotal 10

SROOUTLINE ES-401 BWR Examination Outline Form ES-401-1 Facility: Date of Exam:

RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

  • Total
1. 1 I 20 4 3 7 Emergency &

Abnormal Plant 2 N/A N/A 7 1 2 3 Evolutions Tier Totals 27 5 5 10 1 26 1 4 5 2.

Plant 2 12 0 1 2 3 Systems Tier Totals 38 2 6 8

3. Generic Knowledge and Abilities 1 2 3 4 10 2 3 4 7 Categories 2 2 1 Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e .. except for one category in Tier 3 of the SRO-only outline. the "Tier Totals" in each KIA category shall not be less than two).
2. The point total for each group and tier in the proposed outline must match that specified in the table.

The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated ouUine; systems or evolutions that do not apply at the facility should be deleted and justified: operationally important. site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant-specific priority. only those KiAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions. respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KiA categories.

7.* The generic (G) KiAs in Tiers 1 and 2 shall be selected from Section 2 of the KiA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KlAs.

8. On the following pages. enter the KiA numbers. a brief description of each topic, the topiCS' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam. enter it on the left side of Column A2 for Tier 2. Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the KiA catalog, and enter the KiA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to KiAs that are linked to 10 CFR 55.43.

ES-401 2 Form ES-401-1 BWR Examination Outline E ~al Plaot E""o'o", - TI.e 11G",o, 1 (RO I 5RO)

Name I Safety Function A A G KJA Topic(s) IR #

112 295001 Partial or Complete Loss of Forced X M 1.01 Ability to operate and/or 3.6 76 Core Flow Circulation I 1 & 4 monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Recirculation system 295003 Partial or Complete Loss of AC I 6 295004 Partial or Total Loss of DC Pwr 16 X M2.02 Ability to determine and/or 3.9 77 interpret the following as they apply to Partial or Complete Loss of D.C.

Power: Extent of partial or complete loss of D.C. power 295005 Main Turbine Generator Trip I 3 295006 SCRAM I 1 295016 Control Room Abandonment I 7 295018 Partial or Total Loss of CCW I 8 X M2.02 Ability to determine and/or 3.2 78 interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Cooling water temperature 295019 Partial or Total Loss of Inst. Air 18 295021 Loss of Shutdown Cooling I 4 295023 Refueling Acc I 8 295024 High Drywell Pressure I 5 295025 High Reactor Pressure I 3 295026 Suppression Pool High Water X G 2.1.25 Ability to interpret 4.2 79 Temp. 15 reference materials, such as graphs, curves, tables, etc 295027 High Containment Temperature 15 295028 High Drywell Temperature 15 295030 Low Suppression Pool Wtr Lvii 5 X 2.4.18 Knowledge of the specific 4.4 80 bases for EOPs: Low Suppression Pool Water Level

295031 Reactor Low Water Level / 2 295037 SCRAM Condition Present X A2.02 Ability to determine and/or 4.2 81 and Reactor Power Above APRM interpret reactor water level as it Downscale or Unknown / 1 applies to SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown 295038 High Off-site Release Rate / 9 X 2.1.20 Ability to interpret and 4.6 82 execute procedure steps: High Off-Site Release Rate 600000 Plant Fire On Site / 8 700000 Generator Voltage and Electric Grid Disturbances / 6

° oW 3 Group Point Total: 201 7

ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO I SRO)

E/APE # I Name I Safety Function K K K A G KIA Topic(s) IR #

n 1 2 3 1 2 295002 Loss of Main Condenser Vac I 3 295007 High Reactor Pressure I 3 295008 High Reactor Water Levell 2 295009 Low Reactor Water Levell 2 X AA2.01 Ability to determine and/or 4.2 84 interpret the following as they apply to LOW REACTOR WATER LEVEL:

Reactor water level.

295010 High Drywell Pressure 15 X G 2.4.30 Knowledge of events related 4.1 83 to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

295011 High Containment Temp 15 295012 High Drywell Temperature 15 Temp.~

i 295013 High Suppression Pool 295014 Inadvertent Reactivity Addition 11 295015 Incomplete SCRAM 11

=

295017 High Off-site Release Rate 19 295020 Inadvertent Cont. Isolation I 5 & 7 295022 Loss of CRD Pumps 11 295029 High Suppression Pool Wtr LviI 5 295032 High Secondary Containment X G2.2.44 Ability to interpret control 4.4 85 Area Temperature 15 room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions II !;!~33 High Secondary Containment Radiation Levels I 9 295034 Secondary Containment Ventilation High Radiation I 9 295035 Secondary Containment High Differential PrE1ssure I 5 295036 Secondary Containment High Sump/Area Water Level /5 500000 High CTMT Hydrogen Conc. I 5 I[

KIA Category Point Totals: 0 0 0 0 1 2 Group Point Total: 71 3

259002 Reactor Water Level Control r

261000 SGTS X G2.2.22 Knowledge of limiting 4.7 88 conditions for operations and safety limits: SGTS 262001 AC Electrical X G2.2.25 Knowledge of the 4.2 89 Distribution bases in Technical Specifications for limiting conditions for operations and safety limits.

262002 UPS (AC/DC)

I 263000 DC Electrical Distribution 264000 EDGs X G2.2.40 Ability to apply 4.7 90 Technical Specifications for a system: Emergency Generators n

300000 Instrument Air 400000 Component Cooling Water

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0 0 0 0 0 0 0 I 0 0 4 Group Point Total:

ES-401 5 Form ES-401*1 ES-401 Form ES-401-1 System # I Name KIA Topic(s) 201001 CRD H draulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 215001 Traversin In-core Probe 215002 RBM 216000 Nuclear Boiler Ins!.

219000 RHRlLPCI: ToruslPool Cooling Mode 230000 RHRlLPCI: Torus/Pool Spray Mode 234000 Fuel Handling Equipment x A2.03 Ability to (a) predict 3.1 91 the impacts of loss of electrical power on the Fuel Handling Equipment; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or 0 erations 239001 Main and Reheat Steam 239003 MSIV Leaka e Control 241000 Reactor/Turbine Pressure

286000 Fire Protection X 4.2 92 G.2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC X 3.4 93 G2.2.38 Knowledge of conditions and limitations in the facility license

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290002 Reactor Vessel Internals

. IT.

1 0 1: 12/3

SRO OUTLINE Category KIA # Topic RO IR #

2.1.1 Knowledge of conduct of operations 4.2 94 requirements 1.

Conduct 2.1.45 Ability to identify and interpret diverse 4.3 95 of Operations indications to validate the response of another indication Subtotal 2 2.2.17 Knowledge of the process for managing 3.8 96 maintenance activities during power

2. operations, such as risk assessments, work Equipment prioritization, and coordination with the Control transmission system operator 2.2.5 Knowledge of the process for making _. '<:;1' 97 operating changes to the facility Subtotal 2 2.3.11 Ability to control radiation releases 4.3 98 2.3.13 Knowledge of radiological safety procedures 3.8 99
3. pertaining to licensed operator duties, such as Radiation response to radiation monitor alarms, Control containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc Subtotal 2 2.4.40 Knowledge of SRO responsibilities in 4.5 100 emergency plan implementation 4.

Emergency Procedures I Plan Subtotal 1 Tier 3 Point Total 7

ES-401 Record of Rejected KIAs Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected KIA 111 700000 G 2.1.28 For the original question generic KIA did not directly match, the question was retained and was assigned a new KIA that matched the question.

2/2 216000 G2.4.20 The original question was SRO level, replaced the question with different random KIA. A new question was written to match the KIA.

3 G2.3.12 The original question was a GET level question, replaced the question with different random KIA. Developed a replacement question for the new KIA.

I

ES*301 Administrative Topics Outline Form ES*301*1 Facility: SSES Date of Examination:

Examination Level: SRO-I Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

Heat up rate calculation Conduct of Operations N,R General KIA - 2.1.25 RO 3.9 SRO 4.2

  • A-l.1 M,R Review failed ST and determine required action Conduct of Operations General KIA - 2.2.12 RO 3.7 SRO 4.1
  • A-1.2 N,R Blocking and tagging a pump Equipment Control General KIA - 2.2.41 RO 3.5 SRO 3.9
  • A-2 M,R Review and approve a radioactive liquid release permit Radiation Control General KIA - 2.3.6 SRO 3.7 A-3 i

N,R Make EAL classification Emergency Procedures/Plan General KIA - 2.4.44 SRO 4.4

  • A-4 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
  • Type Codes & Criteria: (C)ontrol room, (S}imulator, or Class(R)oom (D)irect from bank (::; 3 for ROs;::; 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (~ 1)

(P)revious 2 exams (::; 1; randomly selected)

~J/,~-z-

  • Note: Admin JPMs A-t.t, A-t.2, A-2 a~are common JPMs for both RO and SRO candidates. Ensure administration of these commo.n %~oc~urs for all candidates d1¥'ing the same exam day for eachpf.th,sel7 JPMs. I-!!-~ ~ /t-L rf..c'!j II JfALf, cvz.e ~ Jh.e.. St4:> ::[PtJf.<, M- ~ ~t1..d

&ffCl~btLd7{r.

I

-y--/

V.--- 1 )7 10

ES-301 Administrative Topics Outline Form ES-301-1 Facility: SSES Date of Examination:

Examination Level: RO Operating Test Number: 1

[3istrative Topic Type Describe activity to be performed (see Note) Code*

Heat Up rate Calculation Conduct of Operations N,R General KIA - 2.1.25 RO 3.9 SRO 4.2

  • A-l.1 M,R Review failed ST and determine required action Conduct of Operations General KIA - 2.2.12 RO 3.7 SRO 4.1
  • A-1.2 N,R Blocking and tagging a pump Equipment Control General KIA - 2.2.41 RO 3.5 SRO 3.9
  • A-2 Radiation Control N,S State and local notifications Emergency Procedures/Plan General KIA - 2.4.39 RO 3.9
  • A-4 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (::; 3 for ROs; ::; 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (~ 1)

(P)revious 2 exams (::; 1; randomly selected)

.~ J/IO/""l/

  • Note: Admin JPMs A-I.1 , A-I.2 , A-2 '.. A...d are common JPMs for both RO and SRO candidates. Ensure administration of these common JPMs occurs for all candidates during the same exam day for each of these JPMs.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: SSES Date of Examination: 1/17/12 Exam Level: RO D SRO-I SRO-U Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U. including 1 ESF)

System 1 JPM Title Type Code* Safety Function

a. CRD Mechanism/201003 Control Rod Withdrawals A,N,S 1
b. Perform HPCI Quarterly Surveillance/206000 A,N,S 2 Iy Turbine Valve Cycling/241 000 A,N,S 3
d. Core Spray System Shutdown/209001 N,S 4
e. PCIS/SDC restoration/223002 A,L,N,S 5
f. Manually Synchronize Diesel Generator B/264000 A,N,S 6
g. SBGT System Startup/288000 N,S 9 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. Venting Scram Air Header during A TWS D,E,R 1
j. Maintaining RCIC Suction Source during SBO A,E,N,R 2 s 1 E 250 VDC loads lAW EO-100-030 N,E 6

@ 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 1 SRO-I/ SRO-U (A)ltemate 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 (EN)gineered safety feature - I - I ~1 (control room system)

(L)ow-Power I Shutdown ~1/~1/~1 (N)ew or (M)odified from bank including 1(A) ~2/~2/~1 (P)revious 2 exams  :::; 31:::; 31:::; 2 (randomly selected)

(R)CA  :?:1/:?:1/~1 (S)imulator

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 1/17/12 Facility: SSES Date of Examination:

Exam Level: RO SRO-I D SRO-U D Operating Test No.: 1 I"

III Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System 1JPM Title Type Code* Safety Function

a. CRD Mechanism/201003 Control Rod Withdrawals A,N,S 1 b, Perform HPCI Quarterly Surveillance/206000 A,N,S 2
c. Quarterly Turbine Valve Cycling/241 000 A,N,S 3
d. Core Spray System Shutdown/209001 N,S 4
e. PCIS/SDC restoration/223002 A,L,N,S 5
f. Manually Synchronize Diesel Generator B/264000 A,N,S 6
g. SBGT System Startup/288000 N,S 9
h. APRM Gain Adjustmentl215005 N,S 7 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. Venting Scram Air Header during ATWS D,E,R 1
j. Maintaining RCIC Suction Source during SBO A,E,N,R 2 I k. Secure Non-Class 1 E 250 VDC loads lAW EO-100-030 N,E 6

@ 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 I SRO-II SRO-U (A}ltemate path 4-6 I 4-6 I 2-3 (C)ontrol room (D}irect from bank ~9/~8/~4 (E)mergency or abnormal in-plant ~1/~1/~1 (EN)gineered safety feature - I - I ;::1 (control room system)

(L)ow-Power I Shutdown ~1/;::1/;::1 (N)ew or (M)odified from bank including 1(A) ~2/;::2/;::1 (P)revious 2 exams ~ 31 ~ 31 ~ 2 (randomly selected)

(R)CA ~1/;;>:1/;;>:1 (S)imulator

Appendix D Scenario Outline Form ES-D-1

=

Facility: Susquehanna Scenario No.: 1 Op-Test No.: ____.~

Examiners: ------.-----.~-------~-~.-.-~--

Operators:

.~.-----~--~-.----~--~-.~--------------.

Initial Conditions: Unit 1 68% power, EOl, 'B' Condensate Pump out of service for motor replacement Unit 2 60% for waterbox cleaning and rod pattern exchange Turnover: Shift orders are to swap from 1A SW pump to 1C SW pump to allow vibration readings to be taken on 1C SW pump and maintain power with Recirc to compensate for i Xenon.

Event No.

Malf. No. Event Type*

Event Description I

1 N/A 'N Swap runninQ..SW Qumps from 1A to 1C 2 mfNM178007B f:125 I-ATC, APRM 2 Fails High TS-SRO 3 mfHP152004 C-BOP, Inadvertent start of HPCI TS-SRO TS-SRO 4 RD1550043027 Rod drifts in to position 10 RD 1550063027 C-ATC C-BOP I

5 mfF-W1440030 R-ATC '0' Condensate Pump trip with failed runback mfFW144005D 6 cmfAV01_XV147F01 C-ATC, loose SDV Inboard Drain Air Fitting 1 TS-SRO*

7 mfRD155017 M-All Hydraulic ATWS I stuck rods, 'A' SlC pump relief Sl153002 C-ATC valve lift, Failure of 'B' SlC pump on thermal PM02_1 P208A C-BOP overloads Additional rods stuck out, see malf page I

8 cmfPM03 'I P 113A C-ATC, EHC pump failure causes turbine trip and loss of cmfPM07-1 P113B bypass valves, failure of 11 A Aux Bus to fast cmfBR04 -1 A 10101 transfer 9 cmfNB01 LlSB211 N

- C-BOP RCIC Auto Initiation Failure 031A2B, i cmfRl01 e111 K79B I

10 cmfPM03 1P132A  : C-ATC i Running CRD Pump Trips Page 1 of 40 NRC Scenario #1 - Susquehanna Steam Electric Station Operating Test

11 mfHP152015 C-BOP HPCI Turbine Trips requiring performance of ED IMF mfRC150011 RCIC trips on injection lOR diHSC121S12 Prevent further rod insertion d:120 f:OFF lOR diHSC121S10 d:120 f:OFF I

I

  • (N)onnal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor j Page 2 of 40 NRC Scenario #1 - Susquehanna Steam Electric Station Operating Test

Scenario Summary and Administration Instructions Scenario Summary Event 1: The crew begins with the plant at 68% power. As part of turnover, the crew is directed to swap running Service Water pumps from 1A to 1C to allow maintenance to take vibration readings on 1C Service Water pump.

Event 2: Once the Service Water pump swap is complete, APRM 2 fails Upscale. The crew will take action per alarm response to bypass the APRM and the SRO will reference Tech Specs.

Priority is to dec/are APRM 2 inoperable and bypass APRM 2.

Event 3: Once the Tech Spec call is complete for the failed APRM, HPCI will start inadvertently.

The crew will take action per ON-156-001 and OP-152-001 to override HPCI injection. The SRO will dec/are HPCI inoperable and ensure RCIC operability. Priority is to override HPCI, declare HPCI inoperable, and ensure RCIC operability.

Event 4: Once the crew overrides HPCI injection, the scram outlet valve for control rod 30-27 leaks by, causing control rod 30-27 to slowly drift in. However, due to high channel friction, the control rod stops at position 10 and must be fully inserted. The crew will respond by using ON 155-001, control rod problems. Since the rod drifted in and did not go to position 00, ON-155 001 and Tech Specs direct insertion of the rod to 00 and disarming of the HCU. This will be accomplished by sequentially raiSing drive header DIP until the control rod inserts. CRS will address Tech Specs for the inoperable control rod. Priority is to declare rod inoperable, fully insert, and disarm it.

Event 5: Once the control rod Tech Spec call is complete, the 'D' Condensate Pump will trip on overcurrent. Both recirc pumps will fail to runback, and the crew must perform this manually.

Additional actions require monitoring for position on power/flow map and for indications of power oscillations. Priority is to initiate manual recirc runback and monitor power/flow map and APRM for indications of power/flow instabilities.

Event 6: During the manual recirc runback, an air fitting for SV-147-F009 disconnects, causing the inboard SDV drain valve to fail closed. CRS will address Tech Specs for the failed closed valve. With the SDV drain valve closed, the SOV will slowly fill due to normal HCU valve leak by and the leaking outlet scram valve for control rod 30-27. The disconnected air fitting cannot be quickly remedied, and the scram discharge volume level quickly fills to the rod block and eventually the scram setpoints. The crew will respond proactively to the SDV filling by scramming the reactor. Due to a partially plugged SDV, when the mode switch is taken to SHUTDOWN, control rods only partially insert, resulting in a hydraulic ATWS. Priority is to take decisive action to scram the reactor before the automatic scram from high scram discharge volume level.

Events 7-11: The crew will enter EO-1 00-113 for powerllevel control. During power reduction actions, the recirc pumps will be tripped. When the B recirc pump is tripped, the 1B CRD pump trips, requiring operators to later start the 1A CRD pump to enable control rod insertion. The CRS will then direct injection of SBLC. The 'A' SBLC discharge relief valve will lift, preventing injection. The crew will recognize this and swap to the 'B' SBLC pump which will run for approximately 30 seconds, and then trip on thermal overloads. The crew will then direct SBLC injection using RCIC in accordance with ES-150-002. When ATe has stabilized reactor water level with feedwater, the 1A EHC pump will trip and the 1B EHC pump will fail to start, resulting in a turbine trip with loss of bypass capability. This will result in use of SRV's for pressure control and entry into EO-1 00-1 03, PC control due to riSing suppression pool temperature, and Page 3 of 40 NRC Scenario #1 - Susquehanna Steam Electric Station Operating Test

Scenario Summary and Administration Instructions direction to place suppression pool cooling in service. Additionally, 11A Aux Bus auto transfer will fail during the turbine trip, resulting in the loss of the two remaining condensate pumps and transition of level control to HPCIIRCIC.

EO-100-113 will direct insertion of control rods by multiple means. A malfunction of the CRD flow control valve will prevent raising cooling water DIP; preventing drifting in of control rods using the cooling header. Manual control rod insertion per EO-100-113 will be performed to insert control rods. Once approximately four control rods have been inserted, HPCI will trip, requiring the crew to use RCIC for level control. RCIC was overridden per procedure for level reduction, but will also fail to auto initiate. RCIC will start via manual operator actions and trip once the turbine comes up to speed and begins injecting. Further rod insertion will also no longer be possible due to malfunction of the rod insertion pushbuttons. RPV will lower to -161" (TAF) forcing the crew to enter EO-1 00-112 Rapid Depressurization due to inability to restore and maintain level >-161".

Actions will be directed in the field to bypass ARI and RPS. Once the rapid depressurization is performed and level control is being established using low pressure ECCS, the ATC will be able to reset the Scram to begin venting and draining the SDV, and then re-SCRAM the reactor to insert all control rods. The scenario may be terminated when the A TWS has been terminated with low pressure ECCS injection being used for level control.

Target Quantitative Attributes (Per Scenario; See Section D.5.d) , Actual Attributes

1. Total malfunctions (5-8) 7
2. Malfunctions after EOP entry (1-2) 2
3. Abnormal events (2-4) 3
4. Major transients (1-2) 1

. 5. EOPs entered/requiring substantive actions (1-2) EO-1 00-1 02!EO-100-103 2

6. EOP contingencies requiring substantive actions (0-2) EO-100-113/EO-100-112 2
7. Critical tasks (2-3) 3 Page 4 of 40 NRC Scenario #1 - Susquehanna Steam Electric Station Operating Test

Appendix D Scenario Outline Form ES-D-1 Facility: Susquehanna Scenario No.: 2 Op-Test No.:

Examiners: Operators:

Initial Conditions:

Unit at 11 % power with Drywell N2 Purge In Progress Turnover: Unit 1 is at 950 psig and - 11 % power A2SU Step 256, continuing plant startup with containment purge in progress. 'A' RFP is in Discharge Pressure Mode and 'B' RFP is in Standby.

The main turbine has been on turning gear for 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The crew is expected to resume startup actions lAW GO-100-002 step 5.64.1 to ensure 3 element control is ready, place the first RFP in flow control mode in accordance with OP-i45-001, and continue with subsequent actions in GO-i00-002.

Event Malf. No. Event Event No. Type* Description I 1 N/A N-ATC Place first RFP in flow control mode.

2 R-ATC Raise power until reactor power is close to but less than N/A SRO 16%.

3 mfRM179011A I-BOP f:100, SGTS A Rad Monitor instrument fails high with failure of the cmfAV03_HV1571 TS- one of the inboard purge and make-up valve to isolate.

3 SRO 4 C-BOP Failure of MCC 'I B217, which causes loss of 'A' loop of DW TS- spray and % Scram which requires a transfer of the RPS rfDB 105101 Jopen SRO Bus power supply and reset of the % Scram.

I 5 C-BOP I N/A I

RBCCW pump swap due to excessive seal leakage on running pump.

SRO 6 +8.1 set.

fx1RRPB_B21.SET PT=45 C-ATC 'A' Recirc pump speed oscillation/Lock up the 'A' Recirc

+9.11 set SRO pump.

fxi RRPA_B21 .sET PT=90 7 C mfMS183011 B BOP SRV 'B' inadvertently opens (TS)/ initiate Suppression Pool mfMS18301 OB d:1 cooling (ON-183-001, Stuck Open Safety Relief Valve)

TS f:45 SRO I 8 I mfMS183013B I d:2:00 i:40f:100 M-ALL SRV 'B' SUPP Chamber Tailpipe Break.

. r:720

9

, . Ic-' ....

I g)~~~~~f:~~~02B( ~OP{AT I Running RHR pump trips on pre-overload (shaft seizure).

1-0 1r- --r I ALL I Initiate SC and DW SBra~_ _ _ _ _ _ _ _~--I Target Quantitative Attributes (Per Scenario; See Section D.S.d) Actual Attributes

11. Total malfunctions (5-8) 7 I 2. Malfunctions after EOP entry (1-2) 1
3. Abnormal events (2-4) 3 I 4. Major transients (1-2) 1 I 5. EOPs entered/requiring substantive actions (1-2) 2
6. EOP contingencies requiring substantive actions (0-2) 1
7. Critical tasKs (2-3) 2

Scenario Summary Event 1: The scenario begins with Unit 1 at -950 psig and -11 % power during reactor startup with containment (OW) purge in progress. Following turnover the crew is expected to resume startup actions lAW GO-1 00-002 by ensuring 3 element control is ready and placing the first RFP in flow control mode.

Event 2: After the first RFP is placed in flow control mode, the crew will continue with subsequent actions in GO-100-002 to raise power until reactor is close to but less than - 16%.

Event 3: After the power increase, a radiation monitor in the SGTS common exhaust vent duct will fail high causing isolation signals to inboard purge and makeup valves. One of the inboard purge and makeup valves will fail to isolate, crew should recognize and take actions to close the valve and reference TS.

Event 4: After manual isolation of the inboard valve, the essential MCC 1B217 will trip on a fault causing RPS MG set to trip creating % scram. The crew will swap RPS to alternate power supply, reset the half-scram, and restore cooling to the Reactor Recirc Pumps. TS will be referenced.

Event 5: Following the reset of Yz scram, the crew will be required to swap RBCCW pump due to a report from the field indicating excessive seal leakage from the running RBCCW pump.

Event 6: After swapping the RBCCW pump, a failure in the controller for the 'A' Recirc M-G set will cause the Recirc pump speed to oscillate. The crew should recog nize the changes in core and jet pump flows and lock the 'A' Recirc pump scoop tube to prevent further speed changes.

Event 7-8: Following the Recire pump speed oscillation, the 'B' SRV will inadvertently open, requiring the crew to take actions to close the valve in accordance with ON-183-00 1and place suppression pool cooling in service. The crew will not be successful in closing the SRV (per ON requiring manual scram), and a rupture of its tail pipe in the suppression pool chamber will occur. The crew will initiate a manual scram and execute PC control EO-1 00-1 03 due to OW pressure increase.

Event 9: The running Oiv 2 RHR pump will trip on pre-overload due to shaft seizure the crew should recognize that the loop has drained down and only one RHR pump is available for Orywell sprays due to the loss of MCC 1B217 taking out 'A' loop of DW spray. The crew will perform a slow fill of the loop, start the other RHR pump, initiate Suppression chamber spray and when suppression chamber pressure exceeds 13 psig, the crew will initiate drywell sprays.

The scenario will be terminated after OW spray has been initiated.

Scenario Summary and Administration Instructions Appendix D Scenario Outline Form ES*D*1 Facility: Susquehanna Scenario No.: 3 Op-Test No.:

Examiners: Operators:

Initial Conditions: Unit 1 100% power, EOl, Div II Core Spray Pumps out of service Turnover: Maintain power I generator capability curve limits in accordance with the CRC Book Event Malf. No. Event Event No. Type* Description i 1 mfFW145012 I-ATC leading Edge Flow Meter Computer Failure 2 mfMS1460013A C-BOP 3A Feedwater Heater Extraction Steam TS Isolation, Power Reduction SRO, R-ATC 3 cmf CN02 TIC11028 f:O C-BOP RBCCW Temperature Controller Fails in Auto 4 annAR103B01 I-ATC, Drywell Pressure Instrument Failure Without f:AlARM ON TS-SRO ~ Scram 5 mfDB157001 C-BOP loss of 1Y218 6 mfHP152009 f:.7 M-AII HPCI Equipment Room Steam leak, HPCI Isolation Failure mfRP158007B 7 C-BOP Failure of 'B' RPS, ARI Completion of Scram 8 IMF cmfBR04_1A10204 C-All 11 B Aux Bus fails to auto transfer IMF cmfPM04_1 P113A loss of EHC IMF Bypass valves fail to auto operate cmfTR02_PT10101A f:0 IMF cmfTR02_PT10101B f:O 9 See Malfunction Page C-BOP Failure of all but one SRV, Depress Using BPV

  • (N)onnal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of37 NRC Scenario #3 - Susquehanna Steam Electric Station Operating Test

Scenario Summary and Administration Instructions Scenario Summary Event 1: After the crew takes the shift, a failure of the LEFM computer will require entry into ON-100-006. The crew will take action to suspend all activities affecting core reactivity, reduce core flow using recirc by 0.5 Mlbm/hr, and swap feedwater flow input to the core thermal power calculation from LEFM to venturis. Priority for this event is to restore heat balance by changing feedwater flow instruments from LEFM to venturi.

Event 2: Once the feedwater input to the heat balance calculation has been changed from LEFM to Venturi, the 3A Feedwater Heater Extraction Steam Isolation Valve will spuriously close. The crew will take action per ON-147-001 Loss of Feedwater Heating Extraction Steam to lower reactor power :::;71 % power and isolate extraction steam and drain input to 4A and 5A heaters; SRO will address thermal limit Tech Specs. Priority for this event is reduce reactor power :::;71 % to prevent feedwater heater mechanical damage and isolation of extraction steam and the feedwater string if extraction steam cannot be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (consistent with the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> required to restore MCPR per Tech Specs).

Event 3: Once the Tech Spec call is complete, the RBCCW temperature controller will fail in automatic, causing a rise in temperatures on all RBCCW cooled components and an isolation of RWCU. The crew will take action in accordance with ON-114-001 to begin monitoring Recirc Pump motor bearing and seal cavity temperatures. The crew will diagnose a failure of the temperature controller in AUTO and take manual control to restore system temperatures.

Priority for this event is diagnosis of the problem, monitoring of affected components (most importantly Recirc Pump seal temperatures) and restoration of temperature control by taking manual control of the temperature controller or directing control of the TCV bypass valve.

Event 4: When RBCCW cooled component temperatures begin to recover, a drywell pressure transmitter will fail high without an accompanying % scram. The crew will respond per alarm response, dispatch NPO and I&C to the field, diagnose a failed transmitter and failure to %

scram, and the SRO will consult Tech Specs. The crew will insert a % scram on 'A' RPS and contact I&C to insert a trip on the failed instrument. Priority for this event is diagnosis of the failed components, determining that the A RPS subsystem will not generate a scram, declare it inoperable and insert a % scram.

Event 5: Once % scram insertion is complete, the main breaker for 1Y218 will trip, resulting in a loss of instrument bus 1Y218 and 1Y219, requiring the crew to enter ON-117 -001. The crew will take action in accordance with ON-117-001 to restore power to 1Y218, place Refueling Water Pumps in service to supply Condensate Transfer System in accordance with OP-037-003, direct an NPO to take local manual control of the in-service CRD flow control valve, and respond to a loss of Zone 1 and U1 Zone 3 ventilation. They will also note that they have lost several wide range level indicators, ARM's, full core display, and other ancillary indications. Partial restoration of the instrument panels will be successful, but the crew will be unable to restore 1Y219. Priority for this event is restoration of power to 1Y218 to restore vital plant instrumentation, restoration of condensate transfer to ensure ECCS keepfill, and controlling drywell cooling to ensure proper cooling to Recirc Pumps and drywell.

Event 6/7: When the crew has stabilized the plant and restored power to 1Y218, a steam leak starts in the HPCI pump/equipment room. The crew will respond per alarm response to high room temperatures and will diagnose the steam leak. The crew will enter EO-1 00-1 04 Secondary Containment Control, focusing on the Secondary Containment Temperature leg.

Efforts to isolate the leak will be ineffective by automatic and manual means due to a loss of Page 2 of37 NRC Scenario #3 - Susquehanna Steam Electric Station Operating Test

Scenario Summary and Administration Instructions control power for the inboard isolation valve and mechanically bound outboard isolation valve.

When the decision is made that a primary system is discharging into a table 8 RB area. the SRO will direct a reactor scram prior to room temperatures exceeding Max Safe; however 'B' RPS will not generate a SCRAM signal, requiring the use of ARI to complete the SCRAM.

Priority for this event is to scram the reactor once it is determined that a primary system is discharging into the reactor building and before temperatures have exceeded max safe.

Event 8: The SRO will enter EO-100-102 for RPV level and pressure control, both from EO 100-104 and also +13" RPV water level entry conditions. When the turbine trips. the 11 B Aux Bus will fail to transfer resulting in loss of two Condensate Pumps, two Circ Water Pumps, two Service Water pumps, and the loss of power to the 1B & C RFP Discharge Valves. The crew will need to restore power to the 11 B Aux Bus, crosstie load centers, or trip Condensate pumps to prevent uncontrolled Condensate injection during the cooldown. In addition, the 'A' EHC Pump fails to auto start and the bypass valves fail to auto open. The crew will start the 'A' EHC pump and use the bypass valve jack to open bypass valves as necessary to force a cooldown.

Priority for this event is to restore EHC. prevent uncontrolled condensate injection, and begin cooldown to reduce reactor pressure.

Event 9: Once the cooldown is in progress, RCIC room temperatures will rise and the crew will receive a report that the door to RCIC was unable to be re-closed after entering HPCI room for attempted leak isolation. It will be reported that there is steam leaking into the RCIC Room.

When reactor building temperatures exceed max safe values in two areas (HPCI & RCIC). the SRO will enter EO-100-112 Rapid Depressurization. The SRO will direct opening of all ADS valves; upon discovering that no ADS and only 1 other SRV will open. the SRO will direct alternate depressurization using bypass valves. Priority for this event is to direct rapid depressurization once two areas exceed max safe. Upon discovery of only one SRV operating, direct alternate depressurization using bypass valves.

The scenario can be terminated once emergency depressurization using bypass valves or alternate systems has commenced.

Target Quantitative Attributes (Per Scenario; See Section D.S.d) Actual Attributes II

1. Total malfunctions (5-8) 7
2. Malfunctions after EOP entry (1-2) 2
3. Abnormal events (2-4) 4
4. Major transients (1-2) 1
5. EOPs entered/requiring substantive actions (1-2) EO-100-104/EO-100-102 2
6. EOP contingencies requiring substantive actions (0-2) EO-100-112 1
7. Critical tasks (2-3) 2 Page 3 of37 NRC Scenario #3 - Susquehanna Steam Electric Station Operating Test
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