ML13206A295

Draft - Outlines (Folder 2)

ML13206A295
Person / Time
Site:	Susquehanna   (NPF-014, NPF-022)
Issue date:	02/07/2013
From:	Lynn Crawford Public Service Enterprise Group
To:	Caruso J, D'Antonio J Operations Branch I
Jackson D
References
PLA 006970, TAC U01868
Download: ML13206A295 (50)

Text

February 7, 2013 Mr. John Caruso USNRC Chief Examiner USNRC Region 1 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Susquehanna Steam Electric Station Units 1 and 2 Facility Operating Licenses NPF-14, NPF-22 NRC Docket Numbers 50-387,50-388 LOC 25 NRC Initial Operator Licensing Examination Outline PLA 006970

Dear Mr. Caruso:

Enclosed are the examination outlines, supporting the LOC25 NRC Initial License Examination scheduled for the weeks of May 13 through May 20, 2013 at Susquehanna Steam Electric Station.

This submittal includes all appropriate Examination Standard forms and outlines in accordance with NUREG 1021, "Operator Licensing Examination Standards," Revision 9, Supplement 1.

In accordance with NUREG 1021, Revision 9, Supplement 1, Section ES-201, "Initial Operator Licensing Examination Process," please ensure that these materials are withheld from public disclosure until after the examinations are complete. Additionally, SSES requests that the examination outlines be withheld from public disclosure for two years after the administration of the exams to allow reuse of the examination materials in future initial licensing classes.

Should you have any questions concerning this letter, please contact the Operations Training Manager at 570-542-3677. For questions concerning examination materials, please contact Robert Thompson at 570-542-3710.

Sincerely"

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Q:~dfaToJd/ '.

Assistant Operations Manager, Shift Facility Representative

Response

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PLA 006970 February 7,2013 Page 2

Enclosures:

Examination Security Agreements (Form ES-201-3)

Administrative Topics Outline(s) (Form ES-301-1)

Control Room/ln Plant Systems Outline (Form ES-301-2)

BWR Examination Outline (Forms ES-401-1)

Generic Knowledge and Abilities Outline (Tier 3) (Form ES-401-3)

Scenario Outlines (Form ES-D-1)

Record of Rejected KlAs (Form ES-401-4)

Examination Outline Quality Checklist (Form ES-201-2)

Transient and Event Checklist (Form ES-301-5) cc:

(without Attachments)

Chief, NRC Operator Licensing Branch NRC Senior Resident Inspector Pat Finney, SSES bcc:

Site Vice President Plant General Manager General Manager, Operations Manager, Training Manager, Operations Manager, Regulatory Affairs Ops Electronic Letter File Nuclear Records - NUCPT LC post-exam memo pia 006970 LC/RAT/vah

2/22/13 EXAM OUTLINE REVIEW COMMENTS No comments on the JPM sets.

Scenario 1: Event 7 - Change credit to just BOP and SRO not "ALL n. Delete Event 8 redundant to Event 7 Le., event 8 is the required action for event 7.

Scenario 2: Note: Event 5 - Susq. can't simulate a seal failure so licensee revised the draft event that I reviewed to the current high temp. Malfunction. Event 10 change from "All" to credit for only the SRO and BOP also change to an Instrument malfunction.

Scenario 3: Events 8 and 9 as written are set-up items Le., no actions can be taken by the crew to mitigate the failures. The licensee plans to revise one or both of these malfunctions to setup as sequential failures so initial actions can be taken by the crews to mitigate the failures. Also need to change credit from "All" to credit for only the SRO and BOP.

Scenario 5: Event 9 change credit from "All" to credit for only the SRO and BOP.

Written Outline: Cautioned that the following Q topics may not meet SRO only level Os if the Os are written literally to match the KIA topics selected: 77, 77,78, 79, 80, 81, 82, 83, 85, 88, 89, 90, 92, 93, 96 and 100.

I also took the opportunity to discuss with the exam rep (Andy Thompson) preps for validation week.

ES-401 Written Examination Outline Form ES-401-1 Facility:

SSES 2013 #1 Date of Exam:

08/20112 SRO-Only Points

3. Generic Knowledge & Abilities Tier Group K

1 K

2 Total A2 G*

Total

1.

3 5

20 4

3 7

Emergency 2

2 7

2 1

3 Plant Evolutions Tier Totals 4

7 6

4 10 4

2 3

2 2

2 3

3 2

2 26 2

3 5

2.

Plant 2

2 12 0

1 2

3 Systems Tier Totals 5

3 4

3 2

3 4

4 4

3 3

38 3

5 8

2 3

4 234 10 7

Categories 2

3 2

3 2

2 2

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 ofthe 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 +/-I 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, sitc-specific systems that are not included on the outline should be added. Refer to section D.l.b of ES-401, for guidance regarding 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 KAs having an importance rating (IR) of2.5 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 I 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 K/ A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.I.b of ES-40 I for the applicable KIA's

8.

On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IR) 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. 1 does not apply). Use duplicatc 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 totals on Form ES-401-3. Limit SRO selections to KlAs that are linked to 10CFR55.43

ES-401 2

Form ES-401-1 SSES 2013 #1 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # I Name Safety Function KIA Topic(s)

AA2.04 - Ability to determine and/or 295003 Partial or Total Loss of interpret the following as they apply to X

3.7 76 AC.. Pwr 16 PARTIAL OR COMPLETE LOSS OF A.C. POWER: Svstem lineups

, EA2.06 - Ability to determine and/or interpret the following as they apply to 295037 SCRAM Conditions SCRAM CONDITION PRESENT AND Present and Reactor Power Above X

4.1 REACTOR POWER ABOVE APRM APRM Downscale or Unknown 1I DOWNSCALE OR UNKNOWN:

Reactor pressure EA2.04 - Ability to determine and/or 295031 Reactor Low Water Level interpret the following as they apply to X

4,8 78 12 REACTOR LOW WATER LEVEL:

Adequate core cooling 2.1.23 - Conduct of Operations: Ability 295023 Refueling Accidents 18 X

to perform speCific and integrated plant 4,4 79 procedures 2.1.25 - Conduct of Operations: Ability 295028 High Drywell Temperature X

to interpret reference materials such as 4.2 80 15 graphs, curves, tables, etc, 295019 Partial or Total Loss of 2.4.6 - Emergency Procedures f Plan:

X 4.7 81 Inst Air 18 Knowledge of EOP mitigation strategies, EA2.02 - Ability to determine and/or interpret the following as they applies to 295025 High Reactor Pressure I 3 X

4.2 82 HIGH REACTOR PRESSURE: Reactor power EK1.03 - Knowledge ofthe operational 295030 Low Suppression Pool implications of the following concepts as X

3.8 39 Water Levell 5 they apply to LOW SUPPRESSION POOL WATER LEVEL: Heat capacity EK1,01 - Knowledge of the operational implications of the following concepts as 295025 High Reactor Pressure I 3 X

they apply to HIGH REACTOR 3.9 40 PRESSURE: Pressure effects on reactor power AK1.01 - Knowledge of the operational implications of the following concepts as 295006 SCRAM I 1 X

3.7 41 they apply to SCRAM : Decay heat generation and removal AK2,03 - Knowledge of the interrelations between PARTIAL OR 295019 Partial or Total Loss of X

COMPLETE LOSS OF INSTRUMENT 3.2 42 Inst. Air / 8 AIR and the following: Reactor feedwater AK2.01 - Knowledge of the interrelations between PLANT FIRE ON SITE and 600000 Plant Fire On-site / 8 X

2,6 43 the following: Sensors, detectors and valves AK2,01 - Knowledge of the interrelations between PARTIAL OR COMPLETE 295018 Partial or Total Loss of X

LOSS OF COMPONENT COOLING 3.3 44 CCW/8 WATER and the following: System loads AK3.03 - Knowledge of the reasons for the following responses as they apply to 295016 Control Room CONTROL ROOM ABANDONMENT:

X 3,5 45 Abandonment I 7 Disabling control room controls

ES-401 2

Form ES-401-1 SSES 2013 #1 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 E # I Name Safety Function KJA Topic(s)

AK3.01 - Knowledge of the reasons for 29500 I Partial or Complete Loss the following responses as they apply to of Forced Core Flow Circulation I X

PARTIAL OR COMPLETE LOSS OF 3.4 46 1&4 FORCED CORE FLOW CIRCULATION

Reactor water level response EK3.06 - Knowledge of the reasons for the following responses as they apply to 295037 SCRAM Conditions SCRAM CONDITION PRESENT AND Present and Reactor Power Above X

REACTOR POWER ABOVE APRM 3.8 47 APRM Downscale or Unknown/l DOWNSCALE OR UNKNOWN:

Maintaining heat sinks external to the containment AA1.03 - Ability to operate and/or monitor the following as they apply to 295004 Partial or Total Loss ofDC X

PARTIAL OR COMPLETE LOSS OF 3.4 48 Pwr/6 D.C. POWER: A.C. electrical distribution EA1.04 - Ability to operate and/or 295028 High Drywell Temperature monitor the following as they apply to X

3.9 49 15 HIGH DRYWELL TEMPERATURE:

Drywell pressure EA1.03 - Ability to operate and/or monitor the following as they apply to 295026 Suppression Pool High X

SUPPRESSION POOL HIGH WATER 3.9 50 Water Temp. / 5 TEMPERATURE: Temperature monitoring AA2.04 - Ability to determine and/or 295003 Partial or Complete Loss interpret the following as they apply to X

3.5 51 ofAC / 6 PARTIAL OR COMPLETE LOSS OF A.C. POWER: System lineups AA2.02 Ability to determine and/or 295021 loss of Shutdown Cooling interpret the following as they apply to X

3.4 52 14 LOSS OF SHUTDOWN COOLING:

RHRlshutdown cooling system flow EK2.09 - Knowledge of the interrelations 295038 High Olf-site Release Rate between HIGH OFF-SITE RELEASE X

2.9 53 19 RATE and the following:: Post accident sample system (PASS): Plant-Specific..

2.4.6 - Emergency Procedures / Plan:

295024 High Drywell Pressure 1 5 X

3.7 54 Knowledge of EOP mitigation strategies.

2.2.22 Equipment Control: Knowledge 295005 Main Turbine Generator X

of limiting conditions for operations and 4.0 55 Trip 13 safety limits.

2.1.19 - Conduct of Operations: Ability 700000 Generator Voltage and X

to use plant computers to evaluate 3.9 56 Electric Grid Disturbances system or component status.

AK2.04 - Knowledge of the interrelations 295023 Refueling Acc Cooling between REFUELING ACCIDENTS and X

3.2 57 Model 8 the following: RMCS/Rod control and information system EA2.01 - Ability to determine andlor 295031 Reactor Low Water Level interpret the following as they apply to X

4.6 58 12 REACTOR LOW WATER LEVEL:

Reactor water level ategory Totals:

3 5

3 3

3/4 3/3 Group Point Total:

20/7 I

ES-401 3

Form ES-401-1 SSES 2013 #1 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE # / Name Safety Function KiA Topic(s)

AA2.01

• Ability to detennine andlor 295002 Loss of Main Condenser interpret the following as they apply to X

3.1 83 Vac/3 LOSS OF MAIN CONDENSER VACUUM

Condenser vacuumlabsolute pressure 2.4.30 - Emergency Procedures I Plan; Knowledge ofevents related to system 295035 Secondary Containment operation 1 status that must be reported to X

4.1 84 High Differential Pressure I 5 internal organizations or external agencies, such as the state, the NRC, or the tran5m ission system operator.

EA2.0 I - Ability to detennine andlor 295033 High Secondary interpret the following as they apply to Containment Area Radiation X

HIGH SECONDARY CONTAINMENT 3..9 85 Levels 19 AREA RADIATION LEVELS: Area radiation levels EK1.02 - Knowledge of the operational implications of the following concepts as 295034 Secondary Containment X

they apply to SECONDARY 4.1 59 Ventilation High Radiation 19 CONTAINMENT VENTILATION HIGH RADIATION: Radiation releases AK2.0S Knowledge of the 295008 High Reactor Water Level interrelations between HIGH REACTOR X

3.4 60 12 WATER LEVEL and the following: Main turbine: Plant-Specific EK3.01 Knowledge of the reasons for the following responses as they apply to 295032 High Secondary X

HIGH SECONDARY CONTAINMENT 3.5 61 Containment Area Temperature/ 5 AREA TEMPERATURE:

Emergency/normal depressurization AA1.03 - Ability to operate and/or monitor the following as they apply to 295007 High Reactor Pressure 13 X

3.4 62 HIGH REACTOR PRESSURE: RCIC:

Plant-Specific EA2.01 - Ability to determine and/or 295029 High Suppression Pool interpret the following as they apply to X

3.9 63 Water Level 15 HIGH SUPPRESSION POOL WATER LEVEL: Suppression pool water level 2.4.35 - Emergency Procedures / Plan:

295009 Low Reactor Water Level Knowledge of local auxiliary operator X

3.8 64 12 tasks during emergency and the resultant operational effects.

AK2.05 - Knowledge of the interrelations between HIGH DRYWELL 295010 High Drywell Pressure 15 X

.7 65 PRESSURE and the following: Drywell cooling and ventilation KIA Category Totals:

1 2

1 1

1/2 1/1 Group Point Total:

I 7/3

ES-401 4

Form ES-401-1 SSES 2013 #1 Written Examination Outline Plant Systems - Tier 2 Group 1 K

K K

K K

K A A

A Imp System # / Name A2 G

Q#

1 2

3 4

5 6

1 3

4 A2.07 - Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL 259002 Reactor Water Level Control X

CONTROL SYSTEM; and (b) based on those predictions. use procedures to correct. control. or 2.5 86 mitigate the consequences of those abnormal conditions or operations:

Loss of comparator bias signal A2.09 - Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN 205000 Shutdown Cooling X

COOLING MODE); and (b) based 3.8 87 on those predictions. use procedures to correct, control, or mitigate the consequences of those abnormal conditions Reactor low water level 262001 AC Electrical Distribution X

2.4.11 - Emergency Procedures /

Plan: Knowledge of abnormal condition orocedures.

4.2 88 2.1.20 - Conduct ofOperations:

261000 SGTS X

Ability to interpret and execute 4.6 89 orocedure steps.

2.2.42 - Equipment Control:

400000 Component Cooling Water X

Ability to recognize system parameters that are entry-level conditions for Technical 4.6 90 Snecifications K1.02 - Knowledge of the physical connections and/or cause-effect 215004 Source Range Monitor X

relationships between SOURCE RANGE MONITOR (SRM) 3.4 I

SYSTEM and the following:

Reactor manual control K1.08 - Knowledge of the physical connections and/or 259002 Reactor Water Level Control X

cause-effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM 3.2 2

and the following: Recirculation system: Plant-Specific 262001 AC Electrical Distribution X

K2.01 - Knowledge of electrical power supplies to the following:

Off-site sources of power 3.3 3

K2.03 - Knowledge of electrical 209001 LPCS X

power supplies to the following:

2.9 4

Initation Logic K3.03 - Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE 215005 APRM / LPRM X

MONITOR/LOCAL POWER RANGE MONITOR SYSTEM 3.3 5

will have on following: Reactor manual control system: Plant-Specific

ES-401 4

Form ES-401-1 System # I Name 264000 EDGs 218000 ADS 217000 RCIC 263000 DC Electrical Distribution 300000 Instrument Air 205000 Shutdown Cooling 239002 SRVs 400000 Component Cooling Water 223002 PCISlNuclear Steam Supply Shutoff SSES 2013 #1 Written Examination Outline Plant Systems - Tier 2 Group 1 K

K K

K K

K A

A A

A2 G

1 2

3 4

5 6

1 3

4 K3.02 - Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS X

(DIESEL/JET) will have on following: A.C. electrical distribution K4.02 - Knowledge of AUTOMATIC DEPRESSURIZATION SYSTEM X

design feature{ s) and/or interlocks which provide for the following: Allows manual initiation of ADS logic K4.06 - Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC)

X design feature(s) and/or interlocks which provide for the following: Manual initiation K1.02 - Knowledge of the phySical connections and/or cause-effect relationships between D.C. ELECTRICAL X

DIATRIBUTION and the following: Battery charger and battery K5.01 - Knowledge of the operational implications of the X

following concepts as they apply to the INSTRUMENT AIR SYSTEM: Air compressors K6.02 - Knowledge of the effect that a loss or malfunction of the following will have on the X

SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) : D.C.

electrical power K6.05 - Knowledge of the effect that a loss or malfunction of the X

following will have on the RELIEF/SAFETY VALVES:

Discharge line vacuum breaker A1.02 - Ability to predict and / or monitor changes in parameters X

associated with operating the CCWS controls including: CCW temperature A 1.02 - Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT X

ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including: Valve closures Imp 3.9 3.8 3.5 3.2 2.5 2.7 3.0 2.8 3.7 Q#

6 7

8 9

10 II 12 13 14

ES-401 4

Form ES-401-1 System # / Name 2150031RM 206000 HPCI 261000 SOTS 212000 RPS 211000 SLC 203000 RHRlLPCl: Injection Mode 262002 UPS (AC/DC) 215004 Source Range Monitor 209001 LPCS 211000 SLC SSES 2013 #1 Written Examination Outline Plant Systems - Tier 2 Group 1 K

K K

K K K A

A A

A2 G

1 2

3 4

5 6 1

3 4

A2.01 - Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM; and (b) based on those predictions, X

use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Power supply degraded A2.17 - Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions, use X

procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: HPCI inadvertent initiation: BWR-2,3,4 A3.03 - Ability to monitor automatic operations of the X

STANDBY GAS TREATMENT SYSTEM including: Valve operation A3.07 - Ability to monitor automatic operations of the X

REACTOR PROTECTION SYSTEM including: SCRAM air header pressure A4.08 - Ability to manually operate and/or monitor in the X

control room: System initiation:

Plant-Specific A4.09 - Ability to manually X

operate and/or monitor in the control room: ~stem flow 2.1.32 - Conduct of Operations:

X Ability to explain and apply all system limits and precautions.

2.1.28 - Knowledge of the purpose and function of major X

system components and controls.

A3.02 - Ability to monitor automatic operations of the X

LOW PRESSURE CORE SPRAY SYSTEM including:

Pump start K1.02 - Knowledge of the physical connections and/or cause-effect relationships X

between STANDBY LIQUID CONTROL SYSTEM and the following: Core plate differential pressure indication Imp 2.8 3.9 3.0 3.6 4.2 4.1 3.8 4.1 3.8 2.7 Q#

IS 16 17 18 19 20 21 22 23 24

ES-401 4

Form ES-401-1 SSES 2013 #1 Written Examination Outline Plant Systems - Tier 2 Group 1 System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2 A

3 A

4 G

262001 AC Electrical Distribution 239002 SRVs X

X K3.06 Knowledge of the effect that a loss or malfunction of the A.C. ELECTRICAL DISTRIBUTION will have on following: Reactor protection system A2.03 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck open SRV KJA Category Totals:

4 2

3 2

1 2

2 3/2 3

2 2/3 Group Point Total:

Imp Q#

3.8 25 4.1 26 26/5 I

ES-401 5

Form ES-401-1 SSES 2013 #1 Written Examination Outline Plant Systems - Tier 2 Group 2 K

K K

K K

K A

A A

System # I Name A2 G

1 2

3 4

5 6

1 3

4 A2.09 - Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS; 245000 Main Turbine Gen. /

Aux.

X and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Turbine vibration 215002 RBM X

2.2.12 - Knowledge of surveillance procedures 2.4.31 - Emergency Procedures 286000 Fire Protection X

I Plan: Knowledge of annunciator alarms, indications or response procedures K 1.03 - Knowledge of the physical connections and/or 202002 Recirculation Flow Control X

cause-effect relationships between RECIRCULATION FLOW CONTROL SYSTEM and the following: Reactor core flow K2.05 - Knowledge of electrical 201001 CRD Hydraulic X

power supplies to the following:

Altemate rod insertion valve solenoids: Plant-Specific K3.01 - Knowledge of the effect that a loss or malfunction of the 215002 RBM X

ROD BLOCK MONITOR SYSTEM will have on following:

Reactor manual control system:

BWR-3,4,5 K4.05 - Knowledge of RECIRCULATION System 20200 I Recirculation X

design feature(s) and/or interlocks which provide for the following: Seal cooling K5.04 - Knowledge of the operational implications of the 219000 RHRlLPCI:

Torus/Pool Cooling Mode X

following concepts as they apply to RHRlLPCI:

TORUS/SUPPRESSION POOL COOLING MODE: Heat exchanger operation K6.02 - Knowledge ofthe effect that a loss or malfunction of the 290003 Control Room HVAC X

following will have on the CONTROL ROOM HVAC :

Component cooling water systems A1.02 - Ability to predict and/or monitor changes in parameters 268000 Radwaste X

associated with operating the RADWASTE controls including:

Off-site release Imp.

2.8 4.1 4.1 3.7 4.5 3.3 2.9 2.9 2.7 2.6 Q

91 92 93 27 2&

29 30 31 3

33

ES-401 5

Form ES-401-1 SSES 2013 #1 Written Examination Outline Plant Systems - Tier 2 Group 2 System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2 A

3 A

4 A2.10 - Ability to (a) predict the impacts of the following on the OFFGAS SYSTEM; and (b) based on those predictions, use 271000 Off-gas X

procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Offgas system high flow A3.03 - Ability to monitor automatic operations of the 201002 RMCS X

REACTOR MANUAL CONTROL SYSTEM including:

Rod drift alarm A4.06 - Ability to manually 25900I Reactor Feedwater X

operate and/or monitor in the control room: Feedwater inlet temperature 2.4.49 - Emergency Procedures

/ Plan: Ability to perform without 272000 Radiation Monitoring X

reference to procedures those actions that require immediate operation of system components and controls.

A1.02 - Ability to predict and/or monitor changes in parameters 201003 Control Rod and Drive Mechanism X

associated with operating the CONTROL ROD AND DRIVE MECHANISM controls including:

CRD drive pressure KJA Category Totals:

1 1

1 1

1 1

2 111 1

1 1/2 Group Point Total:

Q Imp.

3.1 34 3.2 35 3.4 36 4.6 37 2.8 38 I

1213

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3

• Facility:

SSES 2013 #1 Date:

08/20/12 RO SRO-Only Category KlA#

Topic IR Q#

IR Q#

Knowledge of conservative decision i

2.1.39 4.3 94 l'11aking practices.

Knowledge of the fuel-handling 2.1.35 3.9 99 responsibilities of SRQ's.

1.

Knowledge of Conduct of Operations Conduct 2.1.1 3.8 66 requirements of Operations Knowledge of procedures, guidelines, or

• i 2.1.37 limitations associated with reactivity 4.3 67 management.

I Subtotal 2

2 Knowledge of the process for controlling 2.2.14 4.3 95 equipment configuration or status.

Ability to apply technical specifications for 2.2.40 4.7 98 a system.

Ability to determine Technical 2.2.35 3.6 68

2.

Specification Mode of Operation.

Equipment i

Knowledge of bases in technical

• Control 2.2.25 specifications for limiting conditions for 3.2 69 i

operations and safety limits.

Ability to manipulate the console controls as required to operate the facility 2.2.2 4.6 75 between shutdown and designated power

. levels Subtotal 3

2 I

Knowledge of radiation monitoring

3.

systems such as fixed radiation monitors Radiation 2.3.5 2.9 96 and alarms, portable survey instruments, Control personnel monitoring equipment etc.

Knowledge of radiation monitoring systems, such as fixed radiation monitors 2.3.15 and alarms, potable survey instruments, 2.91 70 I

..... personnel monl!9ringE?guipment, etc.

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to 2.3.13 radiation monitor alarms, containment 3.4 71 entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

btotal 2

1 2.4.37 Knowledge of the lines of authority during emergency plan implementation 4.1 97 Knowledge of operational implications of EOP warnings cautions and notes.

S-a. 6 7

• 4.3 100

4.

Emergency Procedures /

Plan 2.4.1 2.4.29 Knowledge of EOP entry conditions and immediate action steps.

Knowledge of the emergency plan.

4.6 3.1 72 73 Ability to recognize abnormal indications 2.4.4 for system operating parameters which are entry-level conditions for emergency 4.5 74 and abnormal operating procedures.

Subtotal 3

2 Tier 3 Point Total 10 7

ES-401 Record of Rejected KIA's Form ES-401-4 Tipr I ~rnup Ran~

R fRof Select eason or eJec Ion

(#4) K2.02 - Knowledge of electrical power supplies to the following: Valve power.

Low discriminatory value.

2/1 209001 1 K2.02 Randomly selected 2090011 K2.03 - Knowledge of electrical power supplies to the following: Initiation Logic

(#9) K5.01 - Knowledge of the operational implications of the following concepts as they apply to D.C. ELECTRICAL DISTRIBUTION: Hydrogen generation during battery charging.

Low discriminatory value.

2/1 2630001 K5.01 Randomly selected 263000 K1.02 - Knowledge of the physical connections andlor cause-effect relationships between D.C. ELECTRICAL DIATRIBUTION and the following: Battery charger and battery

(#10) K5.13 - Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM: Filters.

Same KA randomly selected on 2010 NRC exam.

2/1 300000 1 K5. 13 Randomly selected 300000 1K5.01 - Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM: Air Compressors

(#17) A3.02 - Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including: Fan start Same KA randomly selected on 2011 NRC exam.

2/1 2610001 A3.02 Randomly selected 261000 1A3.03 - Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including: Valve operation

(#22) 2.2.36 - Equipment Control: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

This is an SRO function.

2/1 2150041 2.2.36 Randomly selected 2150041 2.1.28 - Knowledge of the purpose and function of major system components and controls.

ES-401 Record of Rejected KIA's Form ES-401-4

(#25) K3.04 - Knowledge of the effect that a loss or malfunction of the AC. ELECTRICAL DISTRIBUTION will have on following: Uninterruptible power supply.

UPS/Battery topic oversampled.

2/1 262001 / K3.04 Randomly selected 262001 / K3.06 Knowledge of the effect that a loss or malfunction of the AC. ELECTRICAL DISTRIBUTION will have on following: Reactor protection system

(#26) 2.4.35 - Emergency Procedures / Plan: Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects. (low reactor water level)

Remote panel functions oversampled (#64 &#45) 2/1 239002 /2.4.35 Randomly selected 239002/ A2.03 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck open SRV

(#37) 2.2.39 - Equipment Control: Knowledge of less than one hour technical specification action statements for systems.

2/1 272000 / 2.2.39 There are NO 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less TS for this system.

Randomly selected 2.4.49 - Ability to perform without reference to procedures those actions that require immediate operation of system components and controls

(#38) A3.01 - Ability to monitor automatic operations of the CONTROL ROD AND DRIVE MECHANISM including:

Control rod position.

System/concept oversampled.

2/2 201003/ A3.01 Randomly selected 201003/ A 1.02 - Ability to predict and/or monitor changes in parameters associated with operating the CONTROL ROD AND DRIVE MECHANISM controls including: CRD drive pressure

(#53) EA2.02 - Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Total number of curies released This is a Rad Pro function.

Randomly selected 295038 / EK2.09 - Knowledge of the 1 /1 295038/ EA2.02 interrelations between HIGH OFF-SITE RELEASE RATE and the following:: Post accident sample system (PASS):

Plant-Specific.

ES-401 Record of Rejected KIA's Form ES-401-4

(#55) 2.4.30 - Emergency Procedures 1 Plan; Knowledge of events related to system operation 1status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the transmission system operator.

1 11 2950051 2.4.30 SRO level function.

Randomly selected 295005 12.2.22 - Knowledge limiting conditions for operations and safety limits.

(#56) 700000/2.2.12 - Equipment Control: Knowledge of surveillance procedures.

No surveillance procedures associated with Grid Disturbance.

1 11 700000 1 2.2.12 Randomly selected 700000 12.1.19 - Conduct of Operations: Ability to use plant computers to evaluate system or component status.

(#66) 2.1.31 - Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

This concept is significantly tested in the operating portion 3/1 2.1.31 of the exam.

Randomly selected 2.1.1 - Knowledge of Conduct of Operations requirements

(#70) 2.3.12 - Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

Very similar to #71 3/3 2.3.12 Randomly selected - 2.3.15 - Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, potable survey instruments, personnel monitoring equipment, etc.

(#75) 2.2.4 - (multi-unit license) Ability to explain the variations in control board layouts, systems, instrumentation and procedural actions between units at a facility.

Same KIA used on previous exam.

3/2 2.2.4 Randomly selected 2.2.2 - Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels

ES-401 Record of Rejected KIA's Form ES-401-4 1 /1 295016/ M2.05 1 11 295023/2.1.27 1 /1 295028/2.1.31 1 /1 295025 / 2.2.12 1/2 295033/ EA2.02

(#78) M2.05 - Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Drywell pressure.

No specific procedural reference to develop a discriminating SRO level question. APE also tested in #45 Randomly selected 295031 EA2.04 - Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling

(#79) 2.1.27 - Conduct of Operations: Knowledge of system purpose and 1or function.

Not discriminatory at the SRO level.

Randomly selected 2.1.23 - Ability to perform specific and integrated plant procedures during all modes of operation

(#80) 2.1.31 - Conduct of Operations: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

This concept is significantly tested in the operating portion of the exam.

Randomly selected 2.1.25 - Ability to interpret reference materials such as graphs, curves, tables, etc. (High drywell temperature)

(#82) 2.2.12 - Equipment Control: Knowledge of surveillance procedures.

No surveillance procedure associated with this EPE.

Randomly selected EA2.02 - Ability to determine and/or interpret the following as they applies to HIGH REACTOR PRESSURE: Reactor power

(#85) EA2.02 - Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Equipment operability.

Same general EOP topic/area covered in #61 Randomly selected EA2.01 - Ability to determine andlor interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Area radiation levels

ES-401 Record of Rejected KIA's Form ES-401-4

(#97) 2.4.41 - Knowledge of the emergency action level thresholds and classifications.

Oversample (see #81 and #85) 3/4 2.4.41 Randomly selected 2.4.37 - Knowledge of the lines of authority during emergency plan implementation.

(#92) 2.2.25 - Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

2/2 201002 / 2.2.25 The RMCS is not referenced in Tech Specs, the TRM or in LCOs and Safety Limits. Also, 2.2.25 used on #69 Randomly selected 215002 2.2.12 -- Knowledge of surveillance procedures.

(#90) 2.1.28 - Conduct of Operations: Knowledge of the purpose and function of major system components and controls.

Not discriminating at the SRO level 2/2 400000/2.1.28 Randomly selected 2.2.42 - Equipment Control: Ability to recognize system parameters that are entry-level conditions for Technical Specifications

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Susquehanna Examination Level: SRO Administrative Topic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedu res/Plan Date of Examination: May 2013 Operating Test Number: 2013 Type Code*

Describe activity to be performed M,R Calculate Drywell Leakage and Determine Technical Specification Impact SO-100-006, KIA 2.1.7 (4.7)

M,R Determine Work Hour Controls NDAP-QA-0025, KIA 2.1.5 (3.9)

P,D,R NRC 2/2011 Perform LPRM Upscale Alarm Operability Tracking and Determine Required Actions 01-078-001, KIA 2.2.14 (4.3)

D,R Determine Ability to Bypass Secondary Containment Zone 2 Isolation OP-234-002, KIA 2.3.13 (3.8)

M,R Classify Emergency Conditions and Make Notification EP-PS-100, KIA 2.4.41 (4.6)

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)

..IPM Descriptions C001 - The candidate will be given the previous 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />' surveillance for Dryweilleakage, as well as the raw data needed for calculating the current 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />' surveillance. The candidate will complete calculations on two forms. The candidate will determine that total leak rate is above the allowed 25 gpm. The candidate will additionally be required to determine the Technical Specification implications of this condition.

C002 - The candidate will be given work hour history for three Reactor Operators and told that a Reactor Operator has called in sick. The candidate will be required to review the data and determine which Reactor Operator(s) is(are) eligible to cover the shift without exceeding limits.

Only one Reactor Operator will be within limits. One Reactor Operator would exceed working 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in the last 7 days. One Reactor Operator would not have at least a 34 hour3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> break in the last 9 days. The candidate will additionally be required to complete waiver request paperwork for one of the Reactor Operators that does NOT meet work hour limits.

EC - The candidate will be given previously filled out LPRM tracking attachments and told that a new LPRM has failed. The candidate will be required to complete new LPRM tracking attachments. The candidate will determine that Zone 8 has less than 50% operable LPRMs, APRM channel 1 is NOT operable due to less than 3 operable A level LPRMs, and the associated OPRM remains operable. The candidate will additionally be required to determine the Technical Specification implications of these conditions.

RC - The candidate will be given plant conditions and be required to determine ability to bypass Secondary Containment Zone 2 isolation. The candidate will review an administrative procedure and two Technical Specifications. The candidate will determine that plant conditions allow bypassing the isolation. The candidate will be required to describe the actions required to bypass the isolation.

EP - The candidate will be given plant conditions including an earthquake, loss of offsite power with degraded EDG availability, and a loss of coolant accident on Unit 1. The candidate will be required to classify and declare the appropriate General Emergency within 15 minutes of JPM start. The candidate will then be required to notify state and county officials within 15 minutes of declaration. This notification will include determination that a release is in progress due to containment venting and determination of PARs.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Susquehanna Examination Level: RO Administrative Topic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Date of Examination: May 2013 Operating Test Number: 2013 Type Code*

Describe activity to be performed M,R Calculate Drywell Leakage SO-100-006, KIA 2.1.7 (4.4)

M,R Determine Work Hour Controls NDAP-QA-0025, KIA 2.1.5 (2.9)

D,R Perform Jet Pump Operability Check SO-100-007, KIA 2.2.12 (3.7)

Activate Fire Brigade Emergency Procedures/Plan D, S ON-013-001 Attachment L, KIA 2.4.27 (3.4)

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 (~ 1)

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

JPM Descriptions C001 - The candidate will be given the previous 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />' surveillance for Dryweilleakage, as well as the raw data needed for calculating the current 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />' surveillance. The candidate will complete calculations on two forms. The candidate will determine that total leak rate is above the allowed 25 gpm.

C002 - The candidate will be given work hour history for three Reactor Operators and told that a Reactor Operator has called in sick. The candidate will be required to review the data and determine which Reactor Operator(s) is(are) eligible to cover the shift without exceeding limits.

Only one Reactor Operator will be within limits. One Reactor Operator would exceed working 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in the last 7 days. One Reactor Operator would not have at least a 34 hour3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> break in the last 9 days.

EC - The candidate will be given Recirculation system parameters and directed to perform the associated surveillance test. From the initial data set, the candidate will determine that loop drive flow versus total core flow is not within limits. This requires additional surveillance items to be performed. The candidate will then be given additional jet pump data to perform the additional section of the surveillance. From this data, the candidate will determine that jet pumps are not operable.

EP - The candidate will be given a fire alarm and a print out from the SIMPLEX panel. The candidate will be required to determine the appropriate pre-fire plan for fighting the fire based on the indicated area. Then the candidate will be required to perform actions for activating the fire response. The candidate will call the Fire Brigade Leader and dispatch them to the location.

The candidate will sound the plant fire alarm and evacuate personnel from the area. The candidate will activate Fire Brigade Member pagers and then relay information from the Fire Brigade Leader to the Fire Brigade Members to initiate their response. The candidate will inform Security of the fire event.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Susguehanna Date of Examination: May 2013 Exam Level: RO Operating Test No.: 2013 Control Room 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. Synchronize Diesel Generator 0; Voltage Regulator Failure D,A,EN,S 6

KIA 264000 A4.04 (3.7/3.7), SO-024-001D

b. Lineup RHRSW to the Spent Fuel Pool P, D,S 9

KIA 233000 A2.02 (3.1/3.3), ON-135-001 2011 NRC

c. Swap Core Spray Loops N, EN, L, S 2

KIA 209001 A4.01 (3.8/3.6), OP-151-001

d. Reset Recirc Runback; Pump Speed Oscillates M,A,S 1

KIA 202002 A4.07 (3.3/3.2), ON-164-002

e. Main Steam Line Isolation Recovery D, L, S 3

KIA 239001 A4.01 (4.2/4.0), ON-184-001

f. Start HPCI in Pressure Control Mode; Steam Leak Develops M,A,EN,S 4

KIA 206000 A4.13 (4.1/4.0), OP-152-001

g. Perform Control Room Evacuation Immediate Actions; Mode Switch Fails to Insert Rods, RPS Pushbuttons Work M,A,EN,S 7

KIA 212000 A4.01 (4.6/4.6), ON-100-109

h. Vent the Drywell (RO Only)

D,EN,S 5

KIA 223001 A2.07 (4.2/4.3), OP-173-003 f8J 0); (3 for SRO-I); (3 or 2 for SRO-U)

i.

Pull Scram Fuses; Vent Scram Air Header M,A,E,R 7

KIA 295037 EA1.01 (4.6/4.6), ES-158-001, EO-100-113

j. Start Containment Hydrogen Recombiner D, E, L 5

KIA 223001 A2.01 (4.3/4.4), OP-173-001

k. Lineup Fire Protection System to RHRSW to Inject Into the RPV D,E, L,R 8

KIA 295031 EA1.08 (3.8/3.9), ES-013-001 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 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power I Shutdown

{N)ew or (M)odified from bank including 1 (A)

(P)revious 2 exams (R)CA (S)imulator Simulator Pairings:

A then B D then H F then E C alone G alone Criteria for RO I SRO-II SRO-U 4-6/4-6/2-3

9/

:;8/::;4

~1/~1/~1 1 ~1 (control room system)

~1/~1/~1

~2/~2/<::1

3 I

:; 3 I ::; 2 (randomly selected)

~1/~1/~1

JPM Descriptions A. The candidate will perform a surveillance test to start a Diesel Generator and synchronize it to a bus. The candidate will then load the Diesel Generator to the appropriate levels for the surveillance. A fault will develop which causes the Diesel Generator output breaker to open and output voltage to lower. The Candidate will respond per the alarm response and shutdown the Diesel Generator.

B. The candidate will perform an off-normal procedure section to raise Spent Fuel Pool water level using RHR Service Water. The candidate will align valves within RHR to prevent diversion of water to other loads. The candidate will then align RHR Service Water to the Spent Fuel Pool and start an RHR Service Water pump.

C. The candidate will start with one loop of Core Spray injecting to the Reactor to maintain Reactor water level under LOCA conditions. The candidate will be required to start the alternate loop of Core Spray and begin injection to the Reactor. The candidate will then secure injection from the original loop of Core Spray and shutdown the pumps in that loop. The candidate will need to adjust injection with the new loop of Core Spray to maintain Reactor water level in the assigned band.

D. The candidate will start with both Recirculation pumps in a runback condition due to trip of a Circulating water pump. The candidate will be directed to reset the run back on Recirculation pump A per the off-normal procedure. The candidate will verify control of Recirculation pump speed and then reset the runback. When the runback is reset, a fault will cause the Recirculation pump speed to rise and then oscillate. The candidate will be required to either lock the scoop tube or trip the Recirculation pump per immediate actions of the off-normal procedures.

E. The candidate will perform an off-normal procedure to re-open MSIVs following inadvertent closure. The downstream piping will be depressurized, requiring pressurization to allow opening MSIVs without receiving an automatic isolation on high flow. The candidate will align auxiliary steam loads to allow pressurizing the steam piping and then open valves to start steam piping pressurization. The candidate will monitor MSIV differential pressure. Once pressure requirements are met, the candidate will re-open the MSIVs.

F. The candidate will be placed in a post-scram situation requiring HPCI in pressure control mode. The candidate will align HPCI for start, control the HPCI flow controller in manual, accelerate the HPCI turbine, and establish proper flows. Once HPCI flow exceeds 2000 gpm, a steam leak will develop in the HPCI room and HPCI will not automatically isolate. The candidate will manually isolate HPCI.

G. The candidate will be directed to perform a Control Room Evacuation with the plant at approximately 100% power. The candidate will place the mode switch in shutdown and all control rods will fail to insert. The candidate will perform immediate action to insert rods by either arming and depressing the manual scram push buttons or arming and depressing ARI push buttons. All control rods will then insert. The candidate will continue by taking multiple pre-evacuation actions, including closing MSIVS, tripping Feedwater pumps, aligning Condensate, isolating RWCU, and ensuring HPCI and RCIC are in proper alignment.

H. The candidate will be directed to lower Drywell pressure by venting with the plant at approximately 100% power. The candidate will start Standby Gas Treatment System by

opening a damper and starting a fan. Then the candidate will align Drywell venting dampers to create a flow path from the Drywell to Standby Gas Treatment.

I. The candidate will be presented with A TWS conditions and directed to pull RPS fuses.

The first set of fuses will be pulled. When the candidate attempts to pull the second set of fuses, they will be informed that the cabinet cannot be accessed. They will be required to perform alternate actions outside the control room to insert control rods.

They will proceed to the Reactor Building and vent the scram air header.

J. The candidate will be presented with post-LOCA conditions requiring start of a Hydrogen Recombiner. The candidate will adjust Recombiner controls for start. The candidate will then start the Recombiner. The candidate will adjust Recombiner temperature controls to achieve proper operation.

K. The candidate will be presented with LOCA and station blackout conditions and directed to lineup Fire Protection water to inject into the Reactor through RHRSW and RHR. This will require coordinating with another Operator in the screen house and aligning numerous valves in the Unit 1 and Unit 2 Reactor Buildings.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Susquehanna Date of Examination: May 2013 Exam Level: SROI Operating Test No.: 2013 Control Room Systems@ (B for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code*

Safety Function

a. Synchronize Diesel Generator 0; Voltage Regulator Failure D,A,EN,S 6

KIA 264000 A4.04 (3.7/3.7), SO-024-001D

b. Lineup RHRSW to the Spent Fuel Pool P,D,S 9

KIA 233000 A2.02 (3.1/3.3), ON-135-001 2011 NRC

c. Swap Core Spray Loops N,EN,L,S 2

KIA 209001 A4.01 (3.B/3.6), OP-151-001

d. Reset Recirc Runback; Pump Speed Oscillates M,A,S 1

KIA 202002 A4.07 (3.3/3.2), ON-164-002

e. Main Steam Line Isolation Recovery D,L,S 3

KIA 239001 A4.01 (4.2/4.0), ON-1B4-001

f. Start HPCI in Pressure Control Mode; Steam Leak Develops M,A,EN,S 4

KIA 206000 A4.13 (4.1/4.0), OP-152-001

g. Perform Control Room Evacuation Immediate Actions; Mode Switch Fails to Insert Rods, RPS Push buttons Work M,A,EN,S 7

KIA 212000 A4.01 (4.6/4.6), ON-100-109 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. Pull Scram Fuses; Vent Scram Air Header M,A,E,R 7

KIA 295037 EA1.01 (4.6/4.6), ES-15B-001, EO-100-113

j. Start Containment Hydrogen Recombiner D,E, L 5

KIA 223001 A2.01 (4.3/4.4), OP-173-001

k. Lineup Fire Protection System to RHRSW to Inject Into the RPV 0, E,L, R B

KIA 295031 EA1.0B (3.B/3.9), ES-013-001 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 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power 1Shutdown (N)ew or (M)odified from bank including 1 (A)

(P)revious 2 exams (R)CA (S)imulator Simulator Pairings:

A then B D then H F then E C alone G alone Criteria for RO 1SRO-II SRO-U 4-6 1 4-6 1 2-3

9/

:;8/::;4 2:1/2:1/2:1 1 2:1 (control room system) 2:1/2:1/2:1 2:2/2:2/2: 1

3 1

:; 3 1::; 2 (randomly selected) 2:1/2:1/2:1

..IPM Descriptions A The candidate will perform a surveillance test to start a Diesel Generator and synchronize it to a bus. The candidate will then load the Diesel Generator to the appropriate levels for the surveillance. A fault will develop which causes the Diesel Generator output breaker to open and output voltage to lower. The Candidate will respond per the alarm response and shutdown the Diesel Generator.

B. The candidate will perform an off-normal procedure section to raise Spent Fuel Pool water level using RHR Service Water. The candidate will align valves within RHR to prevent diversion of water to other loads. The candidate will then align RHR Service Water to the Spent Fuel Pool and start an RHR Service Water pump.

C. The candidate will start with one loop of Core Spray injecting to the Reactor to maintain Reactor water level under LOCA conditions. The candidate will be required to start the alternate loop of Core Spray and begin injection to the Reactor. The candidate will then secure injection from the original loop of Core Spray and shutdown the pumps in that loop. The candidate will need to adjust injection with the new loop of Core Spray to maintain Reactor water level in the assigned band.

D. The candidate will start with both Recirculation pumps in a runback condition due to trip of a Circulating water pump. The candidate will be directed to reset the run back on Recirculation pump A per the off-normal procedure. The candidate will verify control of Recirculation pump speed and then reset the runback. When the runback is reset, a fault will cause the Recirculation pump speed to rise and then oscillate. The candidate will be required to either lock the scoop tube or trip the Recirculation pump per immediate actions of the off-normal procedures.

E. The candidate will perform an off-normal procedure to re-open MSIVs following inadvertent closure. The downstream piping will be depressurized, requiring pressurization to allow opening MSIVs without receiving an automatic isolation on high flow, The candidate will align auxiliary steam loads to allow pressurizing the steam piping and then open valves to start steam piping pressurization. The candidate will monitor MSIV differential pressure. Once pressure requirements are met, the candidate will re-open the MSIVs, F. The candidate will be placed in a post-scram situation requiring HPCI in pressure control mode. The candidate will align HPCI for start, control the HPCI flow controller in manual, accelerate the HPCI turbine, and establish proper flows. Once HPCI flow exceeds 2000 gpm, a steam leak will develop in the HPCI room and HPCI will not automatically isolate. The candidate will manually isolate HPCI.

G. The candidate will be directed to perform a Control Room Evacuation with the plant at approximately 100% power. The candidate will place the mode switch in shutdown and all control rods will fail to insert. The candidate will perform immediate action to insert rods by either arming and depressing the manual scram push buttons or arming and depressing ARI pushbuttons. All control rods will then insert. The candidate will continue by taking multiple pre-evacuation actions, including closing MSIVS, tripping Feedwater pumps, aligning Condensate, isolating RWCU, and ensuring HPCI and RCIC are in proper alignment.

I.

The candidate will be presented with A TWS conditions and directed to pull RPS fuses.

The first set of fuses will be pulled. When the candidate attempts to pull the second set of fuses, they will be informed that the cabinet cannot be accessed. They will be required to perform alternate actions outside the control room to insert control rods.

They will proceed to the Reactor Building and vent the scram air header.

J. The candidate will be presented with post-LOCA conditions requiring start of a Hydrogen Recombiner. The candidate will adjust Recombiner controls for start. The candidate will then start the Recombiner. The candidate will adjust Recombiner temperature controls to achieve proper operation.

K. The candidate will be presented with LOCA and station blackout conditions and directed to lineup Fire Protection water to inject into the Reactor through RHRSW and RHR. This will require coordinating with another Operator in the screen house and aligning numerous valves in the Unit 1 and Unit 2 Reactor Buildings.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Susquehanna Date of Examination: May 2013 Exam Level: SROU Operating Test No.: 2013 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code*

Safety Function

a. Synchronize Diesel Generator D; Voltage Regulator Failure D,A, EN,S 6

KIA 264000 A4.04 (3.7/3.7), SO-024-001 D

c. Swap Core Spray Loops N, EN, L, S 2

KIA 209001 A4.01 (3.8/3.6), OP-151-001

f. Start HPCI in Pressure Control Mode; Steam Leak Develops M,A,EN,S 4

KIA 206000 A4.13 (4.1/4.0), OP-152-001 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i.

Pull Scram Fuses; Vent Scram Air Header M,A,E, R 7

KIA 295037 EA1.01 (4.6/4.6), ES-158-001, EO-100-113

k. Lineup Fire Protection System to RHRSW to Inject Into the RPV D, E, L, R 8

KIA 295031 EA1.08 (3.8/3.9), ES-013-001 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 I

(A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A)

(P}revious 2 exams (R)CA (S}imulator Simulator Pairings:

A then B D then H F then E C alone G alone 4-6/4-6/2-3

91

::;81:::;4

~1/~1/~1 1 ~1 (control room system)

~1/~1/~1

~2/~2/~1

3 I

::; 3 1:::; 2 (randomly selected)

~1/~1/~1

..IPM Descriptions A. The candidate will perform a surveillance test to start a Diesel Generator and synchronize it to a bus. The candidate will then load the Diesel Generator to the appropriate levels for the surveillance. A fault will develop which causes the Diesel Generator output breaker to open and output voltage to lower. The Candidate will respond per the alarm response and shutdown the Diesel Generator.

C. The candidate will start with one loop of Core Spray injecting to the Reactor to maintain Reactor water level under LOCA conditions. The candidate will be required to start the alternate loop of Core Spray and begin injection to the Reactor. The candidate will then secure injection from the original loop of Core Spray and shutdown the pumps in that loop. The candidate will need to adjust injection with the new loop of Core Spray to maintain Reactor water level in the assigned band.

F. The candidate will be placed in a post-scram situation requiring HPCI in pressure control mode. The candidate will align HPCI for start, control the HPCI flow controller in manual, accelerate the HPCI turbine, and establish proper flows. Once HPCI flow exceeds 2000 gpm, a steam leak will develop in the HPCI room and HPCI will not automatically isolate. The candidate will manually isolate HPCI.

I. The candidate will be presented with A TWS conditions and directed to pull RPS fuses.

The first set of fuses will be pulled. When the candidate attempts to pull the second set of fuses, they will be informed that the cabinet cannot be accessed. They will be required to perform alternate actions outside the control room to insert control rods.

They will proceed to the Reactor Building and vent the scram air header.

K. The candidate will be presented with LOCA and station blackout conditions and directed to lineup Fire Protection water to inject into the Reactor through RHRSW and RHR. This will require coordinating with another Operator in the screen house and aligning numerous valves in the Unit 1 and Unit 2 Reactor Buildings.

Appendix D Scenario Outline Form ES-D-1 Facility: Susquehanna Scenario No.: NRC~1 Op~Test No.: 2013 Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 100% power. EHC pump A is out of service for maintenance. RCIC is out of service for maintenance.

Turnover: Transfer Bus 1A202 to the alternate supply per OP-104~001 section 2.1.4. Then lower i Reactor power with Recirculation flow to approximately 92% per GO~100~012 section 5.3 and Reactivity Manipulation Package.

Event Malf.

Event No.

No.

Type*

N 1

N/A

BOP, SRO R

2 N/A

ATC, SRO cmfTR02 LTO 3

I-SRO 0812A 4

mfFW145007 A C-AII cmfRL02_E11 C

1K11A 5

BOP, mfRH149004B SRO mfHP152009 6

M-AII mfHP152003 cmfMV06 HV

~tl>>

155FOO2 7

I-Ail Event Description Transfer Bus 1 A202 to the Alternate Supply OP-104-001 Lower Reactor Power with Recirculation Flow to 92%

GO-100-012 Condensate Storage Tank A Level Instrument Fails High AR~016-001 (007), Technical Specifications Feedwater Pump A High Vibrations with Delayed Pump Trip AR-101-001 (A16), OP-145-001, ON-164~002 RHR Pump B Spurious Start and Suction Flange Leak into Reactor Building ON-169-002, EO-OOO-104, Technical Specifications HPCI Steam Leak into Reactor Building EO-000-104, ON-100-101, EO-OOO-102 HPCI Fails to Automatically Isolate cmfMV06 HV (30~~

EO-OOO-104 155F003 cmfMV09 HV 155FOO2

~

I HP~~n V~id-poSIy-n Manually Closed 8

C=AII cmfMV09 HV EO-OO 04, EO-OO 12 155F003 mfRD155006 MUlti~le Control {odS Fail to Insert I 9

C-AII cmfSC04 EO-OOO-102, EO-OOO-113 (N)ormal.

(R)eactivity.

(I)nstrument, (C)omponent, (M)ajor

Facility: Susquehanna Scenario No.: NRC-1 Op-Test No.: 2013

1. Total malfunctions (5-8) 7 i

Events 3,4,5,6,7,8,9

2. Malfunctions after EOP entry (1-2) 3 Events 7, 8,9

3. Abnormal events (2-4) 3 Events 3, 4, 5

4. Major transients (1-2) 1 Event 6 i 5. EOPs entered/requiring substantive actions (1-2) 2 EO-OOO-102, EO-OOO-104

6. EOP contingencies requiring sUbstantive actions (0-2) 2 EO-OOO-112, EO-OOO-113

7. Critical tasks (2-3) 4 CRITICAL TASK DESCRIPTIONS:

CT Manually scram the reactor when any Secondary Containment Area temperature approaches or exceeds Max Safe temperature.

CT*2 - Rapidly depressurize the reactor when two Secondary Containment Areas exceed Max Safe Temperature levels.

CT-3 -Insert control rods lAW EO-OOO-113, Sheet 2, Control Rod Insertion.

CT Stop and prevent Injection except from CRD and SlC.

I

I SCENARIO

SUMMARY

I The crew assumes the shift with the plant operating at approximately 100% power. EHC pump A is out of service for maintenance. RCIC is out of service for maintenance.

The crew will begin by transferring Bus 1A202 to the alternate supply per OP-1 04-001 section 2.1.4.

Then, the crew will lower Reactor power with Recirculation flow to approximately 92% per GO-1 00-012 section 5.3 and Reactivity Manipulation Package.

Once the Reactor power reduction is in progress or completed, the level instrument for Condensate Storage Tank A will fail high. This instrument impacts RCIC and HPCI operability. With RCIC already inoperable, there will be no further impact on that system. The SRO will review Technical Specifications and determine the required actions for HPCI.

Next, Feedwater pump turbine A will develop high vibrations. The crew will respond per AR-016-001 (007) and lower load on the Feedwater pump either manually or by lowering Reactor power. The crew may trip the Feedwater pump as vibration levels approach the 5 mil automatic trip setpoint. Eventually, the Feedwater pump will automatically trip if the crew does not remove the pump from service. An automatic Recirc run back to the 48% limiter will be received. The crew will execute ON-162-001 due to lowering Recirculation flow. Reactor power will be approximately 65% following this transient.

Once the crew stabilizes the plant, RHR pump B will spuriously start. After a short delay, a leak will develop on the pump suction flange. RHR pump room water level will rise and Suppression Pool level will lower. The crew will secure the pump and isolate the leak by closing the suction valve. The crew will enter ON-169-002 due to flooding in the Reactor Building, EO-000-104 due to high Reactor Building area water level, and possibly EO-000-103 due to low Suppression Pool level. The SRO will review Technical Specifications and determine the impact.

Next, a steam leak will develop in the HPCI equipment room. HPCI will fail to automatically isolate.

When the crew attempts to manually isolate the leak, both HPCI steam isolation valves will fail mid position. The crew will enter re-EO-000-104, Secondary Containment Control. With an un-isolable primary system discharging into the Reactor Building and one area temperature approaching or exceeding the maximum safe value, the crew will insert a manual Reactor scram. Ten control rods will fail to insert on the scram. Five of these control rods will be able to be inserted using RMCS and five of these control rods will remain stuck for the rest of the scenario. The crew will enter EO-OOO-113, Level/Power Control, and take actions for the failure to scram. A second steam leak will develop from the HPCI steam isolation valves in the HPCI pipe routing area. This will lead to a second area temperature exceeding the maximum safe value. The crew will perform a rapid depressurization of the Reactor per EO-OOO-112.

The scenario will be terminated when the ADS valves are open, control rod insertion is in progress or completed for all rods that can be inserted, and Reactor water level is being controlled in the assigned band above -161".

lAppendix D Scenario Outline Form ES-D-1 Facility: Susguehanna Scenario No.: NRC-2 Op-Test No.: 2013 Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 100% power. EHC pump A is out of service for maintenance. RCIC is out of service for maintenance.

Turnover: Perform half scram testing for RPS scram channel A 1 per SO-158-001.

Event No.

No.

l:r 1

N/A rfDB105 2

101 3

Report mfHP1520 4

04 cmfTH02_

5 TE14357A 1A2 cmfBR03_

1A10201 6

cmfBR03_

1A10204 mfRR1640 7

11A mfFW1440 8

03A(C) mfHP1520 9

15 mfAD1830 10 01 Event Type*

N BOP, SRO C-AII R

ATC, SRO I-BOP, SRO C-AII C-AII M-AII C-AII C-AII

!,.{:.

~...Ar

~O Event Description Perform Half Scram Testing SO-158-001 MCC 1 B217 De-Energizes, Loss of Power to Drywell Spray Valves, Loss of RPS Bus A ON-158-001, Technical Specifications Power Reduction Due to Minimum Generation Emergency Notification 01-AD-029, GO-100-012 HPCllnadvertent Initiation ON-156-001, Technical Specifications Recirculation Pump A High Temperature AR-102-001 (G03), ON-164-002, Technical Specifications Electrical Fault on Bus 11 B ON-103-003, ON-100-101, EO-000-102 Reactor Coolant Leak in Drywell EO-000-102, EO-000-103 Trip of Condensate Pumps 1A and 1C EO-000-102 HPCI Trip EO-000-102, EO-OOO-112 ADS Fails to Automatically Initiate EO-OOO-102, EO-000-112 (N)ormal, (R}eactivity, (I}nstrument, (C}omponent, (M}ajor

Facility: Susquehanna Scenario No.: NRC-2 Op-Test No.: 2013

1. Total malfunctions (5-8) 9 Events 2, 3, 4,5,6,7,8,9,10

2. Malfunctions after EOP entry (1-2) 3 Events 8, 9, 10

3. Abnormal events (2-4) 5 Events 2, 3, 4, 5, 6

4. Major transients (1-2) 1 Event 7

5. EOPs entered/requiring substantive actions (1-2) 2 EO-000-102, EO-00O-103

6. EOP contingencies requiring substantive actions (0-2) 2 EO-000-102 Alt Level Leg, EO-OOO-112

7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CT Spray the Drywell when Suppression Chamber exceeds 13 psig.

CT Perform Rapid Depressurization when RPV level drops to *161",

SCENARIO

SUMMARY

I I

The crew assumes the shift with the plant operating at approximately 100% power. EHC pump A is out of service for maintenance. RCIC is out of service for maintenance.

The crew will begin by performing half scram surveillance testing for Reactor Scram Instrument Channel A 1 SO-158-001. When the crew resets the half scram, MCC 1 B217 will de-energize. This results in a loss of RPS bus A, as well as power to Drywell spray valves on RHR A The crew will execute ON-158 001, Loss of RPS, to restore power to RPS bus A. The crew will reset the half scram, reset NSSSS logic, and recover the RBCW isolation.

Next, the crew will be notified that a Minimum Generation Emergency has been declared and that a 50 MWe reduction on Unit 1 is requested ASAP. The crew will reduce power in accordance with 01-AD-029, Emergency Load Control, and GO-100-012, Power Maneuvers.

Next, HPCI will spuriously start. HPCI will inject into the Reactor and Reactor power will rise. The crew will take action to override HPCI injection per OP-152-001. The crew will also execute ON-156-001, Unanticipated Reactivity Change. HPCI will be inoperable but available for the remainder of the scenario.

The SRO will determine that with RCIC already out of service, this requires the plant to be taken to Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

A high temperature condition will develop on Recirculation pump A lower guide bearing. The crew will respond per AR-1 02-001 (G03) and trip the pump once limits are exceeded. The crew will also execute ON-164-002, Loss of Recirculation Flow, and the SRO will determine the impact of single loop operations in Technical Specifications.

Next, a fault will occur on Auxiliary Bus 11 B. This will lead to loss of two Condensate pumps, two Service Water pumps, two Circulating Water pumps, and the only operating Recirculation pump. The crew will scram the Reactor due to the loss of all Recirculation pumps. The crew will execute ON-100-101, Scram, Scram Imminent, ON-103-003, 13.8 KV Bus 11 A Loss, and EO-OOO-102, RPV Control.

Next, the two remaining Condensate pumps will trip and a Reactor coolant leak will develop inside the Drywell. The crew will enter EO-OOO-1 03, Primary Containment Control. The crew will transition to HPCI to maintain Reactor water level, spray the Suppression Chamber, and then spray the Drywell. Sprays will be successful in lowering Containment pressures.

HPCI will trip and the Reactor coolant leak will worsen. ADS will fail to automatically initiate. Reactor water level will lower below the top of active fuel and the crew will execute EO-OOO-112, Rapid Depressurization. The crew will open the ADS valves to lower Reactor pressure and then restore and maintain Reactor water level with low pressure injection systems.

The scenario will be terminated when the ADS valves are open, Reactor water level is being controlled in the assigned band above -161", and Containment parameters are being controlled per EO-OOO-103.

Appendix D Scenario Outline Form ES-D-1 Facility: Susquehanna Scenario No.: NRC-3 Op-Test No.: 2013 Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 90% power. Core Spray pump A is out of service for maintenance.

Turnover: Swap EHC pumps per OP-193-003 section 2.9. Then raise Reactor power with Recirculation flow per GO-100-012 section 5.4 and Reactivity Manipulation Package.

Event Malf.

Event Event No.

No.

Type*

Description N-Swap EHC Pumps 1

N/A BOP, OP-193-003 SRO R-Raise Reactor Power with Recirculation Flow 2

N/A ATC, GO-100-012 SRO C

EHC Oscillations mfTC1930 3

ATC, 03 ON-193-001 SRO cmfRV04_

C-SRV Inadvertently Opens 4

PSV141F1 BOP, 3K ON-183-001, Technical Specifications SRO C

cmfMV01 Suppression Pool Cooling Valve Breaker Trip 5

HV151F02 BOP, 4A(8)

OP-149-005, Technical Specifications SRO mfMS1830 10K SRV Leaks with Cracked Tailpipe 6

C-AII mfMS1830 ON-100-101, EO-000-102, EO-000-103 13K mfRP1580 Electrical A TWS 7

04A(8)(C)(

M-AII D)

EO-000-102, EO-000-113 cmfAV06 CRD Flow Control Valves Fail As-Is 8

FV146FOO C-AII 2A(8)

EO-000-113

~V\\.)/

SLC Squib Valves Fail to Open mfSL1530 9

C -Ic:Ir 01A(8)

EO-000-113

.fun f (N)ormal, (R)eac:ivity, (I)nstrument, (C)omponent, (M)ajor

Facility: Susquehanna Scenario No.: NRC-3 Op-Test No.: 2013

1. Total malfunctions (5-8) 7 Events 3, 4, 5, 6, 7, 8, 9

2. Malfunctions after EOP entry (1-2) 2 Events 8 & 9

3. Abnormal events (2-4) 4 Events 3, 4, 5, 6

4. Major transients (1-2)

Event 7

5. EOPs entered/requiring substantive actions (1-2)

EO-000-102

---

r-------------~

cies requiring substantive actions (0-2) ical tasks (2-3) 3 CRITICAL TASK DESCRIPTIONS:

CT Closes the spuriously open SRV or initiates a manual Reactor scram before Suppression Pool water temperature reaches 110°F.

CT Lowers RPV level less than -60" but greater than -161".

CT-3 -Inserts control rods lAW EO*100*113, Sheet 2, Control Rod Insertion.

SCENARIO

SUMMARY

I I

The crew assumes the shift with the plant operating at approximately 90% power. Core Spray pump A is out of service.

The crew will begin by starting EHC pump B and placing EHC pump A in standby per OP-193-003. Then the crew will begin raising Reactor power with Recirculation flow.

During the power ascension, the Reactor pressure regulator will begin to oscillate. Reactor power and pressure will oscillate. The crew will execute ON-156-001, Unanticipated Reactivity Change, 01\\1-178 002, Core Flux Oscillations, and ON-193-001, Turbine EHC System Malfunction. The crew will lower Reactor power. This will suppress the oscillations some. The crew will then lower the load limit and swap pressure regulators to eliminate the oscillations.

Next, SRV K will spuriously open. The crew will execute ON-183-001, Stuck Open Safety Relief Valve.

The SRV control switch will not close the valve, however pulling fuses will. Once the valve is closed, the crew will attempt to place Suppression Pool cooling in service. The Suppression Pool cooling valve breaker will trip immediately upon trying to open the valve on the first RHR loop attempted. The SRO will determine the Technical Specification impact. The crew will be able to place Suppression Pool cooling in service with the second loop of RHR.

SRV K will continue to have significant leak-by even once the fuses are pulled. Additionally, the SRV tailpipe is cracked and will continue to degrade. This results in rising Suppression Pool water temperature, air temperature, and pressure. Eventually, vacuum breakers will cycle, causing Drywell temperature and pressure to rise. The crew will execute ON-1 00-1 01, Scram, Scram Imminent, to lower Reactor power and attempt a manual scram prior to Drywell pressure exceeding 1.72 psig. Later in the scenario, the crew will mitigate degraded Containment parameters with Suppression Chamber sprays per EO-000-103, Primary Containment Control.

When the crew attempts to scram the Reactor, RPS will fail to de-energize and ARI will fail to function.

The crew will enter EO-000-102, RPV Control, and transition to EO-000-113, Levell Power Control. The crew will lower Reactor power by lowering Recirculation flow, tripping Recirculation pumps, lowering Reactor water level, attempting SLC injection, and inserting control rods. The response will be complicated by failure of the SLC pumps to inject boron into the Reactor and failure of the CRD flow control valve to fully open.

The crew will be able to insert all control rods by either venting the scram air header or pulling RPS fuses.

If the Main Generator is tripped with Drywell pressure above 1.72 psig, load shedding will cause a loss of all Condensate pumps unless the crew has taken action to reset the Main Generator lock-outs. The crew will take action to establish Reactor water level control with RCIC andlor HPCI.

The scenario will be terminated when all control rods are inserted and Reactor water level is being controlled in the assigned band above -161".

Appendix D Scenario Outline Form ES-D-1 Facility: Susquehanna Scenario No.; NRC-4 Op-Test No.: 2013 Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 70% power. Core Spray pump A is out of service for maintenance. Circulating Water pump A is out of service and ready to be returned to service.

Turnover: Start Circulating Water pump A per OP-142-001 section 2.13 starting at step 2.13.5.d. Then perform a control rod pattern adjustment per the Reactivity Manipulation Package, OP-156-001, and GO-100-012.

Event M~nt No.

N 1

N/A cmfRD02 R 2

ED121 N01S A

3 N/A cmfPM03_1 P132A 4

mfRD15501 93431 Override 5

aiHS10001 mfDBl1700 4

6 cmfPM04_0 P162A mfMS18300 7

8 mfRP15800 8

7B cmfAV06 H V141 F028D 9

cmfAV06_H V141 F022D e*

N BOP, SRO I-SRO R

ATC, SRO C

BOP, SRO C

ATC, SRO C

BOP, SRO M-AII I-ATC, SRO C

BOP, SRO Event Description Start Circulating Water pump A OP-142-001 Refuel Floor High Exhaust Radiation Monitor Fails Downscale AR-112-G02, Technical Specifications Perform Control Rod Pattern Adjustment OP-156-001, GO-100-012 CRD Pump A Trip with One Inoperable CRD Accumulator ON-155-007, Technical Specifications Main Generator Auto Voltage Regulator Failure ON-198-001 Loss of Instrument Bus 1Y226, Control Structure Chiller Fails to Auto-Start ON-117 -001, Technical Specifications Main Steam Leak into Turbine Building ON-100-101, EO-000-102 RPS B Fails to Scram ON-100-101, EO-OOO-102 MSIVs Fail to Automatically Close EO-OOO-102

mfHP15201 5

HPCI Trips, RCIC Fails to Auto-Initiate, RCIC Initiation Pushbutton mfRC15000 Fails to ARM 10 C-AII 1

EO-OOO-102 diHS15012C B

RCIC Trips mfRC15001 11 C -All 1

EO-OOO-102 (N)ormal (R)eactivity, (I)nstrument (C)omponent (M)ajor Facility: Susquehanna Scenario No.: NRC-4 Op-Test No.: 2013

1. Total malfunctions (5-8) 9 Events 2, 4,5,6,7,8,9,10,11

2. Malfunctions after EOP entry (1-2) 4 Events 8, 9,10,11

3. Abnormal events (2-4) 3 Events 4, 5 & 6

4. Major transients (1-2) 1 Event 7

5. EOPs entered/requiring substantive actions (1-2) 1 EO-OOO-102

6. EOP contingencies requiring substantive actions (0-2) 0

7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CT Manually initiate ARI.

CT Manually isolate a Steam Line break.

SCENARIO

SUMMARY

I I

The crew assumes the shift with the plant operating at approximately 70% power. Core Spray pump A is out of service. Circulating Water pump A is out of service and ready to be started.

The crew will begin by starting Circulating Water pump A per OP-142-001. At the end of this evolution, Refuel Floor High exhaust radiation monitor A will fail downscale. The SRO will determine the Technical Specification impact. Then the crew will perform a control rod pattern adjustment.

During the control rod pattern adjustment, CRD pump A will trip. The crew will respond per ON-155-007 by placing the CRD flow control valve in manual and fully closing it. Then the crew will start CRD pump B, open the CRD flow control valve, and place the valve back in automatic. One CRD accumulator will alarm with low nitrogen pressure. The low nitrogen pressure condition will continue even after other CRD parameters are restored. The SRO will determine the Technical Specification impact of the inoperable accumulator.

The Main Generator voltage regulator will fail to maximum demand while in automatic. The crew will respond per ON-198-001. The crew may attempt to fix the automatic voltage regulator demand signal, but will eventually place the manual voltage regulator in service and lower reactive load.

Next, Instrument Bus 1 Y226 will de-energize. This will cause a loss of multiple control room indications, Instrument Air compressors, and the running Control Structure Chiller. The standby Control Structure Chiller will fail to auto-start. The crew will take action to start the standby Control Structure Chiller. The SRO will determine the Technical SpeCification impact.

Next, a leak will develop from the Main Steam lines into the Turbine Building. The crew will lower Reactor power as time allows and attempt a manual Reactor scram. RPS channel B will fail to scram. The crew will insert control rods by manually initiating ARI. An MSIV isolation will be received on high temperature, however Main Steam Line D will fail to isolate. The crew will manually isolate Main Steam Line D by closing at least one of the two MSIVs.

Once the MSIVs are closed, the crew will lose the normal post-scram level and pressure control systems, Feedwater and Turbine Bypass Valves, respectively. HPCI may be started for level control, but will immediately trip. RCIC will fail to start on an automatic start signal. The crew will be able to manually start RCIC to restore and maintain Reactor water level and assist in controlling Reactor pressure. The RCIC initiation pushbutton will fail to arm, requiring manual component by component startup of RCIC.

SRVs will also initially be required to control Reactor pressure. Steam flow through the SRVs will exceed RCIC makeup capability for a period of time following the scram. The crew will maximize other injection sources. After a period of time, lowering decay heat will allow RCIC to turn Reactor water level, avoiding the need for more drastic Reactor water level control actions. RCIC will eventually trip, requiring the crew to lower Reactor pressure and utilize Condensate pumps for Reactor injection.

The scenario will be terminated when all control rods are inserted, the Main Steam lines are isolated, Reactor water level is being restored to or controlled in the assigned band above -161", and Reactor pressure is being controlled in the assigned band below 1087 psig.

Appendix D Scenario Outline Form ES~D~1 Facility: Susquehanna Scenario No.: NRC-5 (XTRA)

Op-Test No.:

Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 8-9% power during a startup. APRM 1 is bypassed due to spiking.

Turnover: Continue to raise Reactor power by withdrawing control rods. Complete the next four control rod movement steps in the startup sequence. Then place the Mode Switch in RUN and withdraw IRMs per GO-100-002 step 5.63.

Event Malt.

Event Event No.

No.

Type*

Description R-Raise Reactor Power with Control Rods 1

N/A

ATC, SRO OP-156-001, GO-100-002 N-Place Mode Switch in RUN and Withdraw IRMs 2

N/A

ATC, GO-100-002 SRO MSIV Closure Bypass Annunciator Does Not Clear cmfRL02 c 3

I-SRO 721K11A(B)

AR-104-D03, Technical Specifications diHSOOO51C Spurious EDG C Start, ESW Pump C Trip, EDG C Jacket Water C

cmfPM03_0 4

P504C

BOP, TCV Fails Closed mfDG02401 SRO OP-024-001, ON-054-001, Technical Specifications oc Service Water Pump A High Temperature and Degraded cmfTH02 T C

E10901A Performance 5

BOP, rfSW111004 SRO ON-111-001,OP-111-001 mfEN09900 Seismic Event and Instrument Air Leak 6

10 C-AII mf1A118002 ON-OOO-002, ON-118-001, ON-100-101, EO-OOO-102 mfCS15100 Seismic Event, Un-isolable Suppression Pool Break, and Small 2

Steam Leak in Drywell 7

M-AII mfRH14900 4A ON-OOO-002, EO-OOO-103, EO-OOO-102, EO-OOO-112 Condensate Transfer Pump A Trips Upon Makeup to Suppression cmfPM03_0 P155A Pool, Condensate Transfer Pump B Shaft Shear, HPCI and RCIC 8

C-AII Minimum Flow Valves Fail to Open cmfPM05_0 P155B OP-159-001, EO-OOO-103, EO-OOO-112 L)/'l,M Three ADS Valves Fail to Open diHS14113K (L)(M)3

~~( EO-OOO-112 9

Facility: Susquehanna Scenario No.: NRC-5 Op-Test No.: 2013

1. Total malfunctions (5-8) 7 Events 3, 4,5,6,7,8,9

2. Malfunctions after EOP entry (1-2) 2 Events 8 & 9

3. Abnormal events (2-4) 3 Events 4, 5, 6

4. Major transients (1-2) 1 Event 7

5. EOPs entered/requiring substantive actions (1-2) 2 EO-OOO-102, EO-OOO-103

6. EOP contingencies requiring substantive actions (0-2) 1 EO-OOO-112

7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CT-1 -Isolate HPCI when Suppression Pool level cannot be maintained above 17 feet.

CT Rapidly depressurize the reactor when Suppression Pool level cannot be maintained above 12 feet.

SCENARIO

SUMMARY

I I

The crew assumes the shift with the plant operating at approximately 8-9% power during a plant startup.

APRM 1 is bypassed due to spiking.

The crew will begin by raising Reactor power by withdrawing control rods. The crew will complete the current rod group. Then the crew will place the Mode Switch in RU N per GO-1 00-002, Plant Startup, Heatup, and Power Operation. When the Mode Switch is placed in RUN, Annunciator AR-104-001 (003),

MSIV CLOSURE BYPASS, will not clear as required. The SRO will place the startup on hold and determine the Technical Specification impact.

Next, EDG C will spuriously start. ESW pump C will start and then trip after a short time delay. The crew will enter ON-054-001, Loss of Emergency Service Water, and restore ESW flow. EDG C will experience high jacket water and lube oil temperatures due to failure of the jacket water temperature control valve.

The crew will secure EDG C due to the high temperatures. The SRO will determine the Technical Specification impact of ESW pump C and EDG C inoperability.

Service Water pump A will develop a high bearing temperature. After a short period of time, discharge pressure from Service Water pump A will begin to degrade. The crew may enter ON-111-001, Loss of Service Water. The crew will take action to start an alternate Service Water pump and secure Service Water pump A.

Next, a seismic event will occur and cause a rupture in Instrument Air piping. The crew will enter ON 118-001, Loss of Instrument Air. The crew will insert a manual Reactor scram and execute ON-100-101, Scram, Scram Imminent. The MSIVs will close due to low Instrument Air pressure, resulting is loss of Feedwater pumps and Turbine Bypass Valves for Reactor water level and pressure control, respectively.

The crew will use alternate systems for Reactor water level and pressure control, and stabilize the plant as Instrument Air pressure continues to degrade.

A second seismic event will occur and cause a break between the Suppression Pool and both the Core Spray A and RHR A rooms. Additionally, a small steam leak will develop in the Drywell. Suppression Pool level will lower and the crew will enter EO-OOO-103, Primary Containment Control. The crew may attempt to isolate Core Spray A and RHR A suction lines, but the respective suction valves will fail to close. The crew will lineup Suppression Pool makeup from Condensate Transfer, the Condensate Storage Tanks, HPCI, and/or RCIC. The running Condensate Transfer pump will trip upon initiation of makeup. The crew will be able to start the alternate Condensate Transfer pump, but it will fail to produce normal discharge pressure due to a shaft shear. HPCI and RCIC minimum flow valves will also fail to' open. These failures will limit the sources of Suppression Pool makeup.

As Suppression Pool level continues to lower, the crew will isolate HPCI and then enter EO-OOO-112, Rapid Depressurization. The crew will attempt to open all 6 ADS valves, however only 3 will open. The crew will open 3 other SRVs. The crew will control Reactor injection to restore I maintain Reactor water level during and after the rapid depressurization.

The scenario will be terminated when 6 SRVs are open and Reactor water level is being restored to or controlled in the assigned band above -161".