ML12017A164

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Draft - Outlines (Folder 2)
ML12017A164
Person / Time
Site: Oyster Creek
Issue date: 07/30/2011
From: Caruso J
Operations Branch I
To: Goff C
Susquehanna
Jackson D
Shared Package
ML110190451 List:
References
TAC U01842, 50-387/12-301, 50-388/12-301, ES-401, ES-401-1
Download: ML12017A164 (34)
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Text

ES-401 BWR Examination Outline Form ES-401-1 Facility: Susquehanna Date of Exam: January 2012 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.

Emergency &

1 ;r;r;= 3 4 4 20 7 Abnormal Plant 2 2 1 1 N/A 1 1 NfA 1 7 3 Evolutions Tier Totals 5 4 5 10 r,f,14

~:

1 2 3 3 26 5 2.

Plant 2 2 1 1 2 1 0 2 1 12 3 Systems Tier Totals 4 2 3 6 4 4 2 4 3 3 3 38 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 3 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 associated 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.

S. 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 ES-401 BWR Examination Outline Form ES-401-1 EmerQency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

I I E/APE # / Name I 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 J 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 J 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 complete 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 11 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 I 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 18 X AK1.01 Knowledge of the operational 3.5/3.6 7 implications of the effects on com pone nt/system operations as it applies to Partial or Com plete Loss of Component Cooling Water 295019 Partial or Total Loss of Inst. Air 18 X 2.4.11 Knowledge of abnormal 4.0/4.2 8 condition procedures 295021 loss of Shutdown Cooling 14 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 RefuE,Jing Ace / 8 X AK2.03 Knowledge of the 3.4/3.6 (0 interrelations between REFUELING ACCIDENTS and the following:

Radiation monitoring equipment

295024 High Drywell Pressure I 5 X EA2.04 Ability to determine and/or 3,9/3,9 11 interpret Suppression chamber pressure as it applies to high drywell pressure 295025 High Reactor Pressure 13 X EK2.08, Knowledge of the 3.7/3.7 12 interrelations between HIGH REACTOR PRESSURE and the following: Reactor/turbine pressure requlatinq 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 /5 X 2.4.6 Knowledge of the EOP mitigation 3.7/4.7 14 strategies 295030 Low Suppression Pool Wtr LviI 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 coolinq 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 I 9 X EA 1.03 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 I 8 X M2.17 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 G.2.1.28 Knowledge of the purpose 700000 Generator Voltage and Electric Grid X 4.114.1 20 Disturbances I 6 and function of major system components and controls

~ t+/-! G'Oop PoiotTo,," 20/7

ES-401 3 Form ES-401-1 BWR Examination Outline Form ES-401-1 al Plant Evolutions - Tier 1/Group2 (RO I SROL e / Safety Function K A G KIA Topic(s) IR #

1 1 2 of Main Condenser Vac I 295007 High Heactor Pressure I 3 295008 High Heactor Water Level 12 295009 Low Reactor Water Levell 2 X AK 2.03 Knowledge of the interrelations 3.113.2 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.

Containment Temp / 5 295012 High Drywell Temperature / 5 X AK1.01 Knowledge of the operational 3.3/3.5 23 implications of the pressure/temperature relationship is it applies to High Drywell Temperature 295013 Hiqh Suppression Pool Temp. 15 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.

lplete SCRAM 11 Off-site Release Rate I 9 295020 Inadvertent Cont. Isolation I 5 & 7 295022 Loss of CRD Pumps 11 X G2.1.23 Ability to perform specific 4.3/4.4 25 system and integrated plant procedures during all modes of plant operation 295029 Hiqh Suppression Pool Wtr LviI 5 295032 High Secondary Containment X EK1.02 Knowledge of the operational 3.6/4.0 Area Temperalture 15 implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Radiation releases h Secondary Containment tion Levels I 9 295034 Secondary Containment Ventilation HiQh Radiation I 9 295035 Secondary Containment High Differential Pressure I 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 Em 500000 HiQh CTMT Hydrogen Cone.!5 2 I 1 1 1 1 I r::rnllp Point Total:

ES-401 4 BWR Examination Outline Form ES-401-1 I System # I Name illili PliliJ]

3 K

4 2 A A G 3 4 (ROt SRO)

KIA Topic(s) IR #

203000 RHRlLPCI: Injection X K5.01 Knowledge of the operational 2.7/2.9 28 Mode implications of the following concepts as they apply to RHRiLPCI: 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 A.C. 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) andlor interlocks which provide for the following: Automatic flow control:

BWR-2,3,4 2070QQ.!solation (Emergency)

Condenser' "

]I ...*.

Ix I*.. * .'

Ii.:

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

...... )

.. ." I II ."

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:

A.C. 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 follO'v\ing:

System initiation upon operation of SBLC control switch.

212000 RPS X K2.0 1 Knowledge of electrical power 3,2/3.3 35 supplies to the RPS motor-generator 2150031RM X sets Kl.OI Knowledge of the physical connections and/or cause-effect relationships between 3.9/3,9 t:

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 MONITORILOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or 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 K1.01 Knowledge of the physical 3.5/3,5 40 connections and/or cause-effect relationships between REACTOR CORE ISOLATION 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 ofD.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 A 1.02 Ability to predict and/or 3.7/3.7 43 Supply Shutolf 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 K5.01 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 2.712.9 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 ~ower 263000 DC Electrical X Al.OI 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 l'..0wer during full-load testing A4.01 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 X 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 I 2 4 2 4 I 3 4 I 2 Group Point Total:

ES*401 5 Form ES*401*1

System # / 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 201001 CRD Hydraulic 1002RMCS 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 201004 RSCS 201005 RCIS 201006RWM X 2.5/2.8 55 A2.07 Ability to (a) predict 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 3.7/3.9 54 K3.05 Knowledge of the effect that a loss or malfunction of the RECI RCULATION SYSTEM will have on following: Recirculation system MG sets 202002 Recirculation Flow Control 204000 RWCU X 3.4/3.4 57 A4.08 Ability to manually operate and/or monitor in the control room: reactor water level 00 RPIS 01 Traversing In-core Probe 02RBM II

216000 Nuclear Boiler Inst. X G2.4.20 Knowledge of the 3.8/4.3 56 operational implications of EOP warnings, cautions, and notes.

219000 RHRlLPCI: TorusiPool Cooling X 3.1/3.3 59 Mode K2.02 Knowledge of electrical power supplies to the following: Pumps 223001 Primary CTMT and Aux.

226001 RHRlLPCI: CTMT Spray Mode 230000 RHRlLPCI: TorusiPool Spray X A4.06 Ability to manually 4.0/3.9 58 Mode operate and/or monitor in the control room: Valve logic reset following automatic initiation of LPCIIRHR in injection mode 233000 Fuel Pool CoolingfCleanup X 2.913.2 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 Reactorffurbine Pressure ReQulator 245000 Main Turbine Gen.! 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 operation and limitations 256000 Reactor Condensate IT 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 ~stern initiation 288000 Plant Ventilation X 3.213,4 65 K5.02 Knowledge of the operational implications of the following concepts as they apply to Plant Ventilation Systems:

Differential Pressure control econdary CTMT 290003 Control Room HVAC 290002 Reactor Vessel Internals KJA Cateaorv Point Totals: 2 0 Point Total: 12/3

RO OUTLINE Category K/A# Topic RO IR #

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

2.1.32 Ability to explain and apply system limits and 3.8/4.0 67 Conduct 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 Subtotal 2 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 2.3.12 Knowledge of radiological safety principles 3.2/3.7 70 pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Subtotal 3 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 Tier 3 Point Total 10

SRO OUTLINE ES-401 BWR Examination Outline Form ES-401-1 Date of Exam:

RO KIA Category Points SRO-Only Points Tier Group 1liliW~

I'

~ ~

K K 1 2

" K4

) Total A2 G* Total

1. 1 20 4 3 7 Emergency & ,.,

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

Plant Systems 2

Tier Totals

3. Generic Knowledge and Abilities H- 1 2 ~

12 38 10 0 I1 3

2 6

4 3

8 7

Categories 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 (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 associated 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 I ES-401 BWR Examination Outline Form ES-401-1 1::,

volutions - Tier 1/Group 1 (RO I SRO) 81 E/APE # I Name f Safety Function K K K A G KIA Topic(s} IR 1 2 3 1 2 295001 Partial or Complete Loss of Forced X AA1.01 Ability to operate and/or 3,6 Core Flow Circulation 11 & 4 monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Recirculation system 3 Partial or Complete Loss of AC I 6 11 295004 Partial or Total Loss of DC Pwr 16 X AA2.02 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 f 1 295016 Control Room Abandonment f 7 295018 Partial or Total Loss of CCW I 8 X AA2.02 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 Refueling Acc I 8 295024 High Drywell Pressure 15 I I 295025 High Reactor Pressure I 3 295026 Suppression Pool High Water X G 2.1.25 Ability to interpret reference 4.2 79 Temp. 15 materials, such as graphs, curves, tables, etc 295027 High Containment Temperature f 5 295028 High Drywell Temperature 15 295030 Low Suppression Pool Wtr Lvi f 5 X 2.4.18 Knowledge of the specific 4.4 80 bases for EOPs: Low Suppression Pool Water Level

295031 Reactor Low Water Levell 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 11 applies to SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown 295038 High Off-site Release Rate I 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 I 6 I I I I I

I I KIA Category Totals: I () () () () 4 3 201 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 A G KIA Topic(s) IR #

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 Level I 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 I 295013 High Suppression Pool Temp. 15 295014 Inadvertent Reactivity Addition 11 295015 Incomplete SCRAM 11 295017 High Off-site Release Rate I 9

  • 295020 Inadvertent Cont. Isolation I 5 & 7 295022 Loss of CRD Pumps 11 II 295029 High Suppression Pool Wtr Lvii 5 295032 High Secondary Containment X G2.2.44 Ability to interpret control 4.4 85 Area Temperature I 5 room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 295033 High Secondary Containment Area Radiation Levels I 9 295034 Secondary Containment Ventilation High Radiation I 9 295035 Secondary Containment High Differential Pressure I 5 295036 Secondary Containment High Sump/Area Water Levell 5 500000 High CTMT Hydrogen Cone. I 5

KIA Category Point Totals:

{If ~ O*,int Tnt::ll:

ES-401 4 Form ES-401-1 BWR Examination Outline Plant Systems Tier 21Group 1 (RO / SRO)

K K K K K K A A G KIA Topic(s) IR #

1 2 3 4 5 6 3 4 203000 RHRlLPCI: Injection Mode 205000 Shutdown Cooli 206000 HPCI X A2.08 ability to (a) predict the 3.4 87 impacts of high suppression pool temperature on the High Pressure Coolant Injection system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations 207000 Isolation (Emergency)

Condenser 209001 LPCS 209002 HPCS I 211000 SLC 212000 RPS 2150031RM 215004 Source Range Monitor 215005 APRM / LPRM 217000 RCIC 218000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff 239002 SRVs X 2.4.16 Knowledge of EOP 4.4 86 implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines

259002 Reactor Water Level Control 261000 SGT,S 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 264000 EDGs X G2.2.40 Ability to apply 4.7 90 Technical Specifications for a system: Emergency Generators 300000 Instrument Air I~onent Cooling

.,. n . ,T 0 0 0 0 0 0 0 '-',

,T

ES*401 5 Form ES*401*1 1 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO I SRO)

E System # I Name K K K K K K A A A A G KlA Topic(s) IR 1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS I I 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control U

sino In-core Probe I RBM I 216000 Nuclear Boiler Inst.

219000 RHRlLPCI: Torus/Pool Cooling Mode

~ "''~ CTMT ,,' Aw..

RHRlLPCI: CTMT SDrav Mode o RHRlLPCI: Torus/Pool Spray Fuel Pool CoolingfCleanuD 234000 Fuel Handling Equipment X 3.1 91 A2.03 Ability to (a) predict 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 operations 11;;;001 Main and Reheat Steam

, 239003 MSIV Leakaae Control 241000 ReactorfTurbine Pressure Reoulator ill 245000 Main Turbine Gen. fAux.

=--,.

256000 Reactor Condensate 259001 Reactor Feedwater Qffaas Radiation Ltt=+/-+/-J

286000 Fire Protection X 4.2 92 G.2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safet limits 288000 Plant Ventilation 290001 Seconda CTMT 290003 Control Room HVAC X 3.4 93 G2.2.38 Knowledge of conditions and limitations in the facilit license 290002 Reactor Vessel Internals Point Totals: 0 0

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 design or 3.2 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-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 GeneraIKlA-2.2.12 R03.7 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 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 (s 3 for ROs; s 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (2 1)

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

  • Note: Admin JPMs A-I.I, A-l.2, A-2 and A-4 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 Administrative Topics Outline Form ES-301-1 Facility: SSES Date of Examination:

Examination Level: RO 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 SR04.2

  • A-l.1 M,R Review failed ST and determine required action Conduct of Operations General KIA 2.2.12 RO 3.7 SR04.1
  • 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 I\I,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)

  • Note: Admin JPMs A-I.1, A-I.2, A-2 and A-4 are common JPMs for both RO and SRO candidates. Ensnre administration of these common JPMs occurs for all candidates dnring the same exam day for each of these JPMs.

ES-301 Control Roomfln-Plant Systems Outline Form ES-301-2 Facility: SSES Date of Examination: 1/17/12 Exam Level: RO

  • SRO-I 0 SRO-U Operating Test No.:

I Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) 1 System I 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
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 AdjustmenU215005 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 A TWS D, R 1 I
j. Maintaining RCIC Suction Source during SBO A,E,N,R 2
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 1 SRO-II SRO-U (A)lternate path 4-6 1 4-6 I 2-3 (C)ontrol room (D)irect from bank :s; 9/:s; 8/::::4 (E)mergency or abnormal in-plant ~1/~1/~1 (EN)gineered safety feature - I - 1 ~1 (control room system)

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

(R)CA 2:1/2:1/2:1 (S)imulator I

ES-301 Control Roomlln-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.:

01 Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) 1 System I JPM Title Type Code* Safety Function

a. CRD Mechanism/201003 Control Rod Withdrawals A,N,S 1
b. Perform HPCI Quarterly Surveiliance/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 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
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 1 SRO-II SRO-U

{A)lternate path 4-61 4-61 2-3 (C)ontrol room (D)irect from bank  ::;9/::;8/::;4 (E)mergency or abnormal in-plant 2::1/2::1/2::1 (EN)gineered safety feature 1 2::1 (control room system)

(L)ow-Power I Shutdown  ;:::1/;:::1/2::1

{N)ew or (M)odified from bank including 1{A) 2::2/;:::2/;:::1 (P)revious 2 exams  ::; 3 1::; 3 I::; 2 (randomly selected)

{R)CA  ;:::1/;:::1/2::1

{S)imulator

Appendix D Scenario Outline Form ES-D-l Facility: Susquehanna Scenario No.:

, Op-Test No.:

Examiners: Operators:

Initial Conditions: Unit 1 70% 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 EHC pump to 1B EHC pump due to rising vibration trend on 1A EHC pump Event Malf. No. Event Event No. Type* Description 1 N/A N Swap running EHC pumps from 1A to 1B 2 NM178022 I-ATC, APRM Critical Self Test Fault TS-SRO 3 HP152004 C-BOP, Inadvertent start of HPCI TS-SRO

  • 4 RP158008A C-ATC, A RPS MG Set Shaft Seizure BOP 5 RD1550043027 TS-SRO Rod drifts in to position 04 due to failed B RPS RD1550063027 C-ATC fuse j6 FVV144003D R-ATC 'D' Condensate Pump trip with failed runback cmfRl03 K2A cmfRl03-K2B

.7 AV01_XV147F011 C-ATC, loose SDV Inboard Drain Air Fitting TS-SRO 8 RD155017 M-ALl, Hydraulic ATWS, EHC pump failure causes cmfPM03 1P113B C-ATC, turbine trip and loss of bypass valves, failure of cmfPM07-1 P113A 11 A Aux Bus to fast transfer cmfBR04-1A10101 I9 Sl153002 C-BOP 'A' SlC pump relief valve lift, Failure of 'B' SlC PM02 1P208A pump on thermal overloads 10 cmfNB01_L1SB211 N C-BOP RCIC Auto Initiation Failure 031A2B, crnfRl01 e111 K79B 11 HP152014B C-ATC Running CRD Pump Trips 12 HP152015 C-BOP HPCI Turbine Trips requiring performance of ED (N)onnal, (R)eactivity, (C)omponent,

'" (I)nstrument, (M)aior Page 1 of 32 NRC Scenario #1 - Susquehanna Steam Electric Station Operating Test

Scenario Summary and Administration Instructions Scenario Summary The crew begins with the plant at 70% power. As part of turnover, the crew is directed to swap running EHC pumps from 1A to 1B due to a rising vibration trend in 1A. Once the EHC pump swap is complete, APRM fails I NOP. The crew will take action per alarm response to bypass the APRM and the SRO will reference Tech Specs. 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. Once the HPCI injection is overridden, the A RPS MG set fails due to a locked rotor, causing a trip of A RPS on overvoltage. This will cause a half SCRAM and half NSSS isolation. The crew will respond with ON-158-001 and transfer A RPS to alternate power and reset the SCRAM.

Once recovery from the loss of RPS is complete, a loose fuse on the B RPS side for control rod 30-27 fails, causing the scramming of control rod 30-27. Although, due to high channel friction, the control rod stops at position 04 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 directs insertion of the rod to 00 and disarming of the HCU. CRS will address Tech Specs for the inoperable control rod.

Once the 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. During the flow reduction, 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 SDV will slowly fill due to the SSPV's for control rod 30-27 being open. The crew will respond to the SDV filling by entering ON-100 SCRAM, SCRAM IMMINENT. Due to the filling SDV, when the mode switch is taken to SHUTDOWN. control rods only partially insert, resulting in a hydraulic ATWS.

The crew wiU enter EO-1 00-113 for powerllevel control. The CRS will direct injection of SBLC.

The 'A' SBLC discharge relief valve will lift, preventing injection. The crew will recognize this and swap to the'S' 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. Additionally. when SBLC injection is attempted, the 1B EHC pump will trip and the 1A 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 direction to place suppression pool cooling in service.

Additionally, 11A Aux Bus will fail to fast transfer during the turbine trip, resulting in the loss of the two remaining condensate pumps and transition of level control to HPCI/RCIC. During the initial level reduction, RCIC will fail to auto initiate, but will start via operator actions.

Additionally. during control rod insertion, the in-service CRD pump will trip, forcing the ATC operator to start the standby CRD pump to continue rod insertion.

Once actions have been completed to bypass ARI and RPS, the ATC will begin venting and draining the SDV and re-SCRAM the reactor. At this time, HPCI will trip and remain out of service, forcing the crew to perform Rapid Depressurization due to being unable to maintain Rx water level >-161". The scenario may be terminated when Rapid Depressurization is in progress with rod insertion maintaining reactor power <5%.

Page 2 of 32 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 10% power Turnover: Unit 1 is at 950 psig and - 11 % power, continuing plant startup at step 5.65.1 of GO-1 00 002. The crew is expected to resume startup actions lAW GO-1 00-002 step 5.65.1 to ensure 3 element

. control and place the first RFP in flow control mode in accordance with the transfer of the first RFP A to flow control mode and continue with subsequent actions in GO..:100-002.

Event Malf. No. Event Event No. Type* Descri ption 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 cmfRL02 PDSLX07 I-ATC 554Al, SGTS A flow instrument fails high with failure of the one of cmfAV03_HV1571 TS- the inboard purge and make-up valve to isolate.

3 SRO i

4 C-BOP Failure of MCC 1B217, which causes loss of 'A' loop of DW IRF TS- spray and h Scram which needs to be reset and swap rfdB1 051 01Jopen SRO power supply to RPS.

'5 C-BOP RBCCW pump swap due to excessive seal leakage on N/A SRO running pump.

'6 C-ATC

'A' Recirc pump speed oscillation (TS)/Lock up the 'A' mfNM178013A TS- Recirc pump.

SRO 7 IMF_mfMS183011 C B BOP SRV 'B' inadvertently opens (TS)/ maximize torus cooling IMF_mfMS18301O TS (ON-183-001, Stuck Open Safety Relief Valve)

B d:l f:l00 SRO 8 mfMS183013B M-ALL SRV 'B' SUPP Chamber Tailpipe Break.

9 IMF C cmfPM06 1P202B BOP/AT Running RHR pump trips on pre-overload (shaft shear).

r:4:00 f:100 C 10 ALL Initiate DW Spray.

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) 1
3. Abnormal events (2-4) 3
4. Major transients (1-2) 1
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 The scenario begins with Unit 1 at -950 psig and -11 % power during reactor startup. Following turnover the crew is expected to resume startup actions lAW GO-100-002 by ensuring 3 element control and placing the first RFP in flow control mode. 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%.

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 valve will fail to isolate, crew should recognize and take actions to close the valve and reference TS.

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 and reset the scram. TS will be referenced.

Following the reset of Y2 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.

A failure in the controller for the 'A' recirc M-G set will cause the recirc pump speed to oscillate.

The crew should recognize the changes in core and jet pump flows and "lock up" the 'A' recirc pump. Following this, the 'B' SRV will inadvertently open, requiring the crew to take actions to close the valve, and will place suppression pool cooling in accordance with ON-183-001. The crew will not be successful in closing the SRV (per ON requiring manual scram), and a rupture in the suppn:lssion pool chamber tail pipe will occur. The crew will initiate a manual scram and execute PC control EO-100-103 due to DW pressure increase.

The running RHR pump 1P202B will trip on pre-overload due to shaft shear, the crew should recognize that only one RHR pump is available for Drywell sprays due to the loss of MCC 1B217 takinj~ out 'A' loop of DW spray. The crew will initiate Suppression chamber spray and when suppression chamber pressure exceeds 13 psig, the crew will initiate drywell spray using 1P202D RHR pump. The scenario will be terminated after DW spray has been initiated.

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

Examiners: Operators:

Initial Conditions: Unit 1 100% power, EOL, Unit 2 10% for drywell entry/leak identification Turnover: Shift orders are to perform SO-155-006, Quarterly ARI Manual Trip Channel Functional Test Event Malf. No. Event Event No. Type* Description 1 N/A N Quarterly ARI Manual Trip Channel Functional Test 2 FW145012 I-ATC Leading Edqe Flow Meter Computer Failure 3 MS1460013A C-BOP 3A Feedwater Heater Extraction Steam TS Isolation, Power Reduction SRO, R-ATC 4 CN02 TIC11 028 f:O C-BOP RBCCW Temperature Controller Fails in Auto 5 MHB01 4

.. VVI I-ATC, Drywell Pressure Instrument Failure Without ~

TS-SRO Scram 6 DB157001 C-ATC, Loss of 1Y218 C-BOP 7 HP152009 M-AII HPCI Equipment Room Steam Leak, HPCI Isolation Failure I ~~~0~~~~~9A-C 8 C-BOP Loss of all RFP, Failure of 'B' RPS, ARI Completion of Scram 9 mfAD183001, C-BOP Failure of All SRV, Depress Using BPV diHSB211530AA f:norm, diHSB211530BA f:norm

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of32 NRC Scenario #3 - Susquehanna Steam Electric Station Operating Test

Scenario Summary and Administration Instructions Scenario Summary The crew begins with the plant at 100% power. As part of turnover, the crew is directed to perform SO-155-006, Quarterly ARI Manual Trip Channel Functional Test. When testing is complete, 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 and will reduce core flow using recirc by 0.5 Mlbm/hr.

Once the core thermal feedwater input 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; SRO will address thermal limit Tech Specs.

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. 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.

When RBCCW cooled component temperatures begin to recover, a failure of a drywell pressure transmitter will fail high without an accompanying Yz scram. The crew will respond per alarm response, diagnose a failed transmitter and failure to Yz scram, and the SRO will consult Tech Specs. The crew will insert a Yz scram on 'A' RPS and contact I&C to insert a trip on the failed instrument.

Once Yz scram insertion is complete, the feeder breaker for 1Y218 will trip, resulting in a loss of instrument bus 1Y218, requiring the crew to enter ON-117-001. The crew will take action in accordance with ON-117-001 to place Refueling Water Pumps in service to supply Condensate Transfer System, in accordance with OP-037-003, take local manual control of the in-service CRD flow control valve, reset Recirc MG set lockups, 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.

When the crew has stabilized the plant, a steam leak starts in the HPGI 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. When the decision is made that a primary system is discharging into a table 8 RB area and a SCRAM is about to be performed, a trip of all three RFP's will occur.

The resultant loss of level will cause a low level SCRAM Signal to be generated; however 'B' RPS will not generate a SCRAM signal, requiring the use of ARI to complete the SCRAM.

Efforts to isolate the leak will be ineffective by automatic and manual means due to a loss of control pow!r for the inboard isolation valve and mechanically bound outboard isolation valve.

Due to the loss of feedwater, this will prompt the crew to reduce reactor pressure using bypass valves to transition level control to condensate. Upon reactor building temperatures exceeding max safe values in two areas, the SRO will direct entry into EO-100-112 Rapid Depressurization. The SRO will direct opening of all ADS valves; upon discovering that no ADS Page 2 of32 NRC Scenario #3 - Susquehanna Steam Electric Station Operating Test

Scenario Summary and Administration Instructions and only 1 other SRV will open, the SRO will direct alternate depressurization using bypass valves.

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

Page 3 of32 NRC Scenario #3 - Susquehanna Steam Electric Station Operating Test

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