ML111160176

WT-2011-03 Draft Outlines

ML111160176
Person / Time
Site:	Waterford
Issue date:	03/21/2011
From:	NRC Region 4
To:	Entergy Operations
References
50-382/11-301
Download: ML111160176 (67)

Text

ES-301 Administrative Topics Outline Form ES-301-1 Revision 0 SRO Facility: WATERFORD 3 Date of Examination: March 21, 2011 Examination Level:

SRO Operating Test Number:

1 Administrative Topic (see Note)

Type Code* Describe activity to be performed A5 Conduct of Operations K/A Importance:

4.4 R, N 2.1.23, Ability to perform specific system and integrated plant procedures during all modes of plant operation. Review and approve a completed Shutdown Margin with an immoveable CEA in accordance with OP-903-090, Shutdown Margin, section 7.3, Shutdown Margin Verification - Un-trippable CEA.

A6 Conduct of Operations K/A Importance:

3.8 R, M 2.1.18, Ability to make accurate, clear, and concise logs, records, status boards, and reports. Review and approve OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation

Data. A7 Equipment Control K/A Importance:

4.6 R, N 2.2.37, Ability to determine operability and/or availability of safety related equipment. Review and approve a completed Equipment Out of Service document in accordance with OP-100-010, Equipment Out of Service.

A8 Radiation Control K/A Importance:

3.7 R, N 2.3.4, Knowledge of radiation exposure limits under normal and emergency conditions. Calculate dose and assign non-licensed operators to perform work in radiological restricted areas. Given dose rate with and without shielding installed, time to install shielding, and job completion time using 1 operator or using 2 operators, determine proper job assignment.

A9 Emergency Plan K/A Importance:

4.4 S, M 2.4.38, Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required. Determine appropriate classification and actions based on a toxic gas release in accordance with EP-004-010, Toxic Chemical Contingency Procedure. NOTE: All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 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)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 1 NRC 2011 Revision 1 Facility: WATERFORD 3 Date of Examination: March 21, 2011 Exam Level Reactor Operator Operating Test No.:

NRC 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 S1 001 Control Rod Drive, Perform CEA testing for Regulating Group 6 in accordance with OP-903-005, Control Element Assembly Operability Check. Fault: CEA 21 will insert after initially moved, requiring a reactor trip. A4.01 Controls for CCWS RO - 3.1, SRO - 2.9 A, D, S 1 S2 004 Chemical and Volume Control System; Makeup to the Volume Control Tank using Boric Acid and Primary Makeup Water batches in accordance with OP-002-005, Chemical and Volume Control. Fault: The Boric Acid counter will fail to secure the Boric Acid addition, requiring the applicant to manually secure Boric Acid flow. The applicant will then need to add the Primary Makeup Water for the initial calculation, plus the additional based on the extra boric acid added. A4.07 Boration/dilution RO - 3.9, SRO - 3.7 A, M, S 2 S3 005 Shutdown Cooling System; Secure Shutdown Cooling Train B and place it in standby in accordance with OP-009-005, Shutdown Cooling. A4.01 Controls and indication for RHR pumps RO - 3.6, SRO - 3.4 D, L, S 4 - P S4 039 Main and Reheat Steam System; BOP operator immediate operator actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure. A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 A, M, S 4 - S S5 028 Hydrogen Recombiner and Purge Control System Start Hydrogen Recombiner A in accordance with OP-008-006. A4.01 HRPS controls RO - 4.0, SRO - 4.0 D, L, P, S 5 S6 064 Emergency Diesel Generator (ED/G) System; Parallel Emergency Diesel Generator A for EDG testing in accordance with OP-009-002, Emergency Diesel Generator. Fault: After EDG A load is raised, EDG A load will rise without manipulation requiring a trip of EDG A. A4.06 Manual start, loading, and stopping of the ED/G RO - 3.9, SRO - 3.9 A, D, S 6 S7. 029 Containment Purge System; Perform surveillance OP-903-052, Controlled Ventilation Area System Operability Check, and secure RAB Normal Ventilation and start CVAS Train A. K1.03 Engineering safeguards RO - 3.6, SRO - 3.8 N, S 8 S8. 012 Reactor Protection System; Place Reactor Power Cutback in service and remove reactor trip on turbine trip in accordance with OP-004-015, Reactor Power Cutback. A4.03 Channel blocks and bypasses RO - 3.6, SRO - 3.6 D, S 7 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 2 NRC 2011 Revision 1 In-Plant Systems

@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 061 Emergency Feedwater System; Reset overspeed device on Emergency Feedwater Pump AB in accordance with OP-902-005, Station Blackout Recovery. A2.04 Pump failure or improper operation RO - 3.4, SRO - 3.8 D, E, L, P, R 4 - S P2 064 Emergency Diesel Generator (ED/G) System; Trip Emergency Diesel Generator B locally. Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG B. K4.02 Trips for ED/G while operating (normal or emergency)

RO - 3.9, SRO - 4.2 A, D, R 6 P3 068 Control Room Evacuation Close Train B Safety Injection Tank outlet valves during a Control Room Evacuation in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown. AA1.28 PZR level control and pressure control RO - 3.8, SRO - 4.0 E, L, N 2 @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 5 (C)ontrol room 0 (D)irect from bank 9 / 8 / 4 7 (E)mergency or abnormal in-plant 1 / 1 / 1 2 (EN)gineered safety feature - / - / 1 (control room system) - (L)ow-Power / Shutdown 1 / 1 / 1 4 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 4 (P)revious 2 exams 3 / 3 / 2 (randomly selected) 2 (R)CA 1 / 1 / 1 2 (S)imulator 8

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 3 NRC 2011 Revision 1 Facility: WATERFORD 3 Date of Examination: March 21, 2011 Exam Level SRO - Instant Operating Test No.:

NRC 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 S1 001 Control Rod Drive, Perform CEA testing for Regulating Group 6 in accordance with OP-903-005, Control Element Assembly Operability Check. Fault: CEA 21 will insert after initially moved, requiring a reactor trip. A4.01 Controls for CCWS RO - 3.1, SRO - 2.9 A, D, S 1 S2 004 Chemical and Volume Control System; Makeup to the Volume Control Tank using Boric Acid and Primary Makeup Water batches in accordance with OP-002-005, Chemical and Volume Control. Fault: The Boric Acid counter will fail to secure the Boric Acid addition, requiring the applicant to manually secure Boric Acid flow. The applicant will then need to add the Primary Makeup Water for the initial calculation, plus the additional based on the extra boric acid added. A4.07 Boration/dilution RO - 3.9, SRO - 3.7 A, M, S 2 S3 005 Shutdown Cooling System; Secure Shutdown Cooling Train B and place it in standby in accordance with OP-009-005, Shutdown Cooling. A4.01 Controls and indication for RHR pumps RO - 3.6, SRO - 3.4 D, L, S 4 - P S4 039 Main and Reheat Steam System; BOP operator immediate operator actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure. A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 A, M, S 4 - S S5 S6 064 Emergency Diesel Generator (ED/G) System; Parallel Emergency Diesel Generator A for EDG testing in accordance with OP-009-002, Emergency Diesel Generator. Fault: After EDG A load is raised, EDG A load will rise without manipulation requiring a trip of EDG A. A4.06 Manual start, loading, and stopping of the ED/G

RO - 3.9, SRO - 3.9 A, D, S 6 S7. 029 Containment Purge System; Perform surveillance OP-903-052, Controlled Ventilation Area System Operability Check, and secure RAB Normal Ventilation and start CVAS Train A. K1.03 Engineering safeguards RO - 3.6, SRO - 3.8 N, S 8 S8. 012 Reactor Protection System; Place Reactor Power Cutback in service and remove reactor trip on turbine trip in accordance with OP-004-015, Reactor Power Cutback. A4.03 Channel blocks and bypasses RO - 3.6, SRO - 3.6 D, S 7 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 4 NRC 2011 Revision 1 In-Plant Systems

@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 061 Emergency Feedwater System; Reset overspeed device on Emergency Feedwater Pump AB in accordance with OP-902-005, Station Blackout Recovery. A2.04 Pump failure or improper operation RO - 3.4, SRO - 3.8 D, E, L, P, R 4 - S P2 064 Emergency Diesel Generator (ED/G) System; Trip Emergency Diesel Generator B locally. Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG B. K4.02 Trips for ED/G while operating (normal or emergency)

RO - 3.9, SRO - 4.2 A, D, R 6 P3 068 Control Room Evacuation Close Train B Safety Injection Tank outlet valves during a Control Room Evacuation in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown. AA1.28 PZR level control and pressure control RO - 3.8, SRO - 4.0 E, L, N 2 @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO /

SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 5 (C)ontrol room 0 (D)irect from bank 9 / 8 / 4 6 (E)mergency or abnormal in-plant 1 / 1 / 1 2 (EN)gineered safety feature - / - / 1 (control room system) - (L)ow-Power / Shutdown 1 / 1 / 1 3 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 4 (P)revious 2 exams 3 / 3 / 2 (randomly selected) 1 (R)CA 1 / 1 / 1 2 (S)imulator 7

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 5 NRC 2011 Revision 1 Facility: WATERFORD 3 Date of Examination: March 21, 2011 Exam Level SRO - Upgrade Operating Test No.:

NRC 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 S1 001 Control Rod Drive, Perform CEA testing for Regulating Group 6 in accordance with OP-903-005, Control Element Assembly Operability Check. Fault: CEA 21 will insert after initially moved, requiring a reactor trip. A4.01 Controls for CCWS RO - 3.1, SRO - 2.9 A, D, S 1 S2 S3 S4 039 Main and Reheat Steam System; BOP operator immediate operator actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure. A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 A, M, S 4 - S S5 S6 S7. 029 Containment Purge System; Perform surveillance OP-903-052, Controlled Ventilation Area System Operability Check, and secure RAB Normal Ventilation and start CVAS Train A. K1.03 Engineering safeguards RO - 3.6, SRO - 3.8 N, EN, S 8 S8.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 6 NRC 2011 Revision 1 In-Plant Systems

@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 P2 064 Emergency Diesel Generator (ED/G) System; Trip Emergency Diesel Generator B locally. Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG B. K4.02 Trips for ED/G while operating (normal or emergency)

RO - 3.9, SRO - 4.2 A, D, R 6 P3 068 Control Room Evacuation Close Train B Safety Injection Tank outlet valves during a Control Room Evacuation in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown. AA1.28 PZR level control and pressure control RO - 3.8, SRO - 4.0 E, L, N 2 @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 3 (C)ontrol room 0 (D)irect from bank 9 / 8 / 4 2 (E)mergency or abnormal in-plant 1 / 1 / 1 1 (EN)gineered safety feature - / - / 1 (control room system) 1 (L)ow-Power / Shutdown 1 / 1 / 1 1 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 3 (P)revious 2 exams 3 / 3 / 2 (randomly selected) 0 (R)CA 1 / 1 / 1 1 (S)imulator 3 NRC Revision 0 ES-301 Simulator Scenario Quality Checklist Form ES-301-5 Facility: Waterford 3 Date of Exam:

March 21, 2011 Operating Test No.NRC A P P L I C A N T E V E N T T Y P E Scenarios 1 2 3 4 T O T A L M I N I M U M(*) CREW POSITION CREW POSITION CREW POSITION CREW POSITION S R O A T C B O P S R O A T C B O P S R O A T C B O P S R O A T C B O P R I U SRO-U 1 & 2 RX 0 1 1 0 NOR 1 1 1 1 1 I/C 2,3,5, 7,8 1,3,4,5,7,8 11 4 4 2 MAJ 6 6 2 2 2 1 TS 3,4 1,2,3 5 0 2 2 SRO-I 1 RX 1 1 1 1 0 NOR 3 1 1 1 1 I/C 1,2,5, 6,7 2,8 1,4,7, 8 8 4 4 2 MAJ 4 6 6 3 2 2 1 TS 1,2,3 3 0 2 2 SRO-I 2 RX 3 1 1 1 0 NOR 1 1 1 1 1 I/C 1,5 2,3,5, 7,8 1,3,4,5,7,8 13 4 4 2 MAJ 4 6 6 3 2 2 1 TS 3,4 1,2,3 5 0 2 2 SRO-I 3 & 4 RX 1 1 1 1 0 NOR 3 1 1 1 1 I/C 1,2,5, 6,7 2,8 7 4 4 2 MAJ 4 6 2 2 2 1 TS 1,2,3 3 0 2 2 NRC Revision 0 RO 1 & 3 RX 0 1 1 0 NOR 3 1 2 1 1 1 I/C 2,6,7 3,5,7 3,5,8 9 4 4 2 MAJ 4 6 6 3 2 2 1 TS 0 0 2 2 RO 2 & 4 RX 3 1 1 1 0 NOR 0 1 1 1 I/C 1,5 1,4,7, 8 6 4 4 2 MAJ 4 6 2 2 2 1 TS 0 0 2 2 RO 5 RX 0 1 1 0 NOR 3 1 2 1 1 1 I/C 2,6,7 3,5,7 3,5,8 9 4 4 2 MAJ 4 6 6 3 2 2 1 TS 0 0 2 2 Spare RX 4 NOR 4 I/C 1,2,3,6,7,8 3,7,8 1,2,6, 8 MAJ 5 5 5 TS 1,3 Instructions: 1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the "at-the-controls (ATC)" and "balance-of-plant (BOP)" positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an Instant SRO additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position. 2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis. 3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicant's competence count toward the minimum requirements specified for the applicant's license level in the right-hand columns.

Appendix D Scenario Outline Form ES-D-1 Scenario 1 Rev 1 Facility: WATERFORD 3 Scenario No.: 1 Op Test No.:

NRC Examiners: Operators: Initial Conditions:

• Reactor power is 100%

• Protected Train is A

• AB Bus is aligned to Train A Turnover: Maintain 100% power Event No. Malf. No. Event Type* Event Description 1 RC15A2 I - ATC I - SRO TS - SRO Pressurizer level instrument RC-ILI-0110 X fails low. OP-901-110, Pressurizer Level Control Malfunction.

2 FW26A I - BOP I - SRO TS - SRO Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low. OP-901-201, Steam Generator Level Control Malfunction.

3 RD02A52 R - ATC N - BOP N - SRO TS - SRO CEA 52 Drops into the core, OP-901-102, CEA or CEDMCS Malfunction, and OP-901-212, Rapid Plant Power Reduction.

4 RC23A CS04A M - All Loss of Coolant Accident, OP-902-002, Loss of Coolant Accident Recovery.

CS-125 A fails closed 5 CV02A C - ATC C - SRO Charging Pump A fails to auto-start.

6 SI02D C - BOP C - SRO Low Pressure Safety Injection Pump A fails to auto start on SIAS requiring manual start 7 CS01B C - BOP C - SRO Containment Spray Pump B trip, OP-902-008, Safety Function Recovery Procedure Alignment of LPSI Pu mp B to replace CS Pump B. * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Scenario Event Description NRC Scenario 1 Scenario 1 Rev 1 The crew assumes the shift at 100% power with instructions to maintain 100% power.

After taking the shift, Pressu rizer level instrument RC-ILI-011 0X fails low. Due to the failure, Letdown flow goes to minimum flow and both backup Charging Pumps start. The SRO should enter OP-901-110, Pressurize r Level Control Malfunction. The crew should utilize sub section E1, Pressurizer Lev el Control Channel Malfunction. The ATC should take manual control of Pressurizer level and select the non-faulted channel.

Using Tech Specs and OP-903-013, Monthl y Channel Checks, the SRO should enter Tech Spec 3.3.3.5, a 7 day action require ment, and determine Tech Spec 3.3.3.6 entry is not required since QSPDS is operable and meeting the Pressurizer level channel check. SPDS indication of Pressurizer le vel on the Plant Monitoring Computer is affected by this failure.

After the non-faulted channel is selected and Tech Specs are addressed, Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fa ils low. The Feedwater Control System will respond by raising Feedwa ter flow to Steam Generator #1. The SRO should enter OP-901-201, St eam Generator Level Contro l Malfunction. The BOP will be required to take manual control and ma tch Feedwater and Main Steam flow. The Ultrasonic Flow Meter will fail as a result of the instrument failure and require entry into TRM 3.3.5. The Feedwater controls for Steam Generator #1 will remain in manual as a result of this failure.

After the crew has addressed the Feedwater instrument failure, CEA 52 drops into the core. Off normal procedure OP-901-102, CEA or CEDMCS Malfunction, should be entered. The dropped CEA will require a rapid plant power reduction. The SRO should enter OP-901-212, Rapid Plant Power Reduction. Direct Boration should commence within 15 minutes of the dropped CEA. For t he power reduction, the ATC will perform direct Boration to the RCS as well as ASI control with CEAs and Pressurizer Boron Equalization. The BOP will manipulate the controls to reduce Main Turbine load and manipulate Feedwater to Steam Generator #1 in manual. The SRO should enter Tech Specs 3.2.3, 3.1.

3.1, and 3.1.3.5.

Once the crew has commenced the power reduc tion and lowered power to ~ 90%, or at the lead examiner's discretion, a loss of coolant accident will occur. Charging Pump A will fail to start on the lowering Pressurizer level. The crew should diagnose the

Pressurizer level dropping with all available Charging Pumps operating, trip the Reactor, and initiate Safety Injection Actuation (S IAS) and Containment Isolation Actuation (CIAS). When Containment Spra y is actuated, either manually or automatically, CS-125 A will fail to automatically open and will not open using the control switch. This does not create a need for action at this time, but Containment Spray flow will only be provided from Train B with CS

-125 A failed closed. Low Pr essure Safety Injection Pump A will fail to automatically start on SIAS, requiring the BOP operator to manually start LPSI Pump A.

Scenario Event Description NRC Scenario 1 Scenario 1 Rev 1 After the crew completes OP-902-000, Standard Post Trip Actions and diagnoses into OP-902-002, Loss of Coolant Accident Recovery, Containment Spray Pump B will trip, resulting in no Containment Spray flow. The crew should recognize that they are not meeting the Safety Function Status Checklist of OP-902-002 and transition to OP-902-008, Safety function Recovery Procedure.

Prioritization in OP-902-008 should result in Containment Isolation being priority 1 and Containment Temperature and Pressure Control being priority 2. The crew should address Containment Isolation by overriding CS-125 B closed. The crew should address Containment Temperat ure and Pressure Control by aligning Low Pressure Safety Injection Pump B to replace the fa iled Containment Spray Pump B. It is acceptable to pursue these tasks in parallel, since establishing flow with LPSI B to the Containment Spray header will also satisf y Containment Isolation concerns.

The scenario can be terminated after Low Pressu re Safety Injection Pump B is aligned for Containment Spray, or a fter the CRS gives the order to perform that alignment, at the lead examiners discretion.

NRC Scenario 1 Scenario 1 Rev 1 Critical Tasks

1. Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow. This task becomes applicable after Containment Spray is initiated. The time requirement of 3 minutes is based on the RCP operating limit of 3 minutes without CCW cooling.

2. Establish Containment temperature and pressure control.

This task is satisfied by aligning LPSI Pump B to replace CS Pump B prior to exiting the Containment Temperature and Pressure Control safety function in OP-902-008. This task becomes applicable following the failure of Containment Spray Pump B. The Functional Recovery procedure utilized following this failure will direct this activity to satisfy the Containment Pressure and Temperature Control safety function.

Scenario Quantitative Attributes 1. Total malfunctions (5-8) 7 2. Malfunctions after EOP entry (1-2) 2 3. Abnormal events (2-4) 3 4. Major transients (1-2) 1 5. EOPs entered/requiring substantive actions (1-2) 2 6. EOP contingencies requiring substantive actions (0-2) 1 7. Critical tasks (2-3) 2 NRC Scenario 1 Scenario 1 Rev 1 Scenario Notes:

A. Reset Simulator to IC-191. B. Verify the following Scenario Malfunctions: 1. rc15a for Pressurizer level

2. fw26a for Steam Genera tor #1 Feedwater flow 3. rd02a52 for CEA 52

4. rc23a for LOCA

5. cv02a for Charging Pump A

6. si02d for Low Pressure Safety Injection Pump A

7. cs01b for Containment Spray Pump B

8. cs04a for CS-125 A C. Verify the following Override: 1. di-08a04s22-1 for CS-125 A D. Ensure Protected Train A sign is placed in SM office window.

E. Verify EOOS is 10.0 Green F. Complete the simulator setup checklist.

G. Start DCS, Record Data, select file PlantParameters.txt.

NRC Scenario 1 Scenario 1 Rev 1 Simulator Booth Instructions

Event 1 Pressurizer Level Instrument RC-ILI-0110X Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 1.

2. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room. 3. If sent to LCP-43, report RC-ILI-0110 X1 is failed low.

Event 2 Steam Generator #1 Feedwater Flow Instrument FW-IFR-1111 Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 2.

2. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 CEA 52 Drops, Rapid Plant Power Reduction

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If called to remove Condensate Polis hers from service, acknowledge communication and report that you will perform actions requested. 3. If Work Week Manager or I&C is calle d, inform the caller that a and a team will be sent to the CEDMCS Alley to investigate.

Event 4 LOCA Inside Containment

1. On Lead Examiner's cue, initiate Event Trigger 4.

2. If called as RCA watch report CS-125 A appears to be mechanically bound, the stem looks bent. 3. If called as RAB watch to check the Emergency Diesel Generators, initiate Trigger 10, EDG A & B Trouble ala rms clear, report they are running satisfactorily. 4. If the Duty Plant Manager is called, inform the caller that he will make the necessary calls.

Event 5 Low Pressure Safety Injection Pump A fails to start

1. If called to check the LPSI Pump A breaker, report all indications are normal. 2. If called to check the LPSI Pump A locally, report all indications are normal.

NRC Scenario 1 Scenario 1 Rev 1 Event 6 Containment Spray Pump B Trips

1. After the crew has entered OP-902-002 and on the Lead Examiner's cue, initiate Event Trigger 7. 2. If called to check the Containment Spra y Pump B breaker, report over-current flags are picked up on all 3 phases. 3. If called to check the Containment Spray Pump B, report that there are visible charring on the motor with an acrid smell, but no indications of a fire or smoke. 4. If called for TSC concurrence, report SM/EC has granted concurrence. 5. If called as RAB watch to come to the Control Room for over-ride key for CS-125 B, acknowledge communication. Report to the Control Room on lead examiner's cue. 6. If crew does obtain key and over-rides CS-125 B closed, use remote CSR13B for the local key operation.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2011 SRO Scenario1.cdf.

Save the file into the folder for the appropriate crew.

NRC Scenario 1 Scenario 1 Rev 1 Scenario Timeline: Event Malfunction Severity Ramp HH:MM:SS Delay Trigger 1 RC15A2 0 N/A N/A 1 Pressurizer level instrument RC-ILI-0110 X fails low 2 FW26A 0 N/A N/A 2 Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low 3 RD02A52 N/A N/A N/A 3 CEA 52 Drops into the core 4 RC23A 3.0 % 8:00 N/A 4 Loss of Coolant Accident 5 CV02A N/A N/A N/A 5 Charging Pump A fails to auto-start 6 SI02D N/A N/A N/A N/A Low Pressure Safety Injection Pump A fails to auto start 7 CS04A DI-08a04s22-1 N/A N/A N/A N/A CS-125 A Fails to open, will not open manually. 7 CS01B N/A N/A N/A 7 Containment Spray Pump B trip

NRC Scenario 1 Scenario 1 Rev 1

REFERENCES:

Event Procedures 1 OP-901-110, Pressurizer Level Control Malfunction OP-903-013, Monthl y Channel Checks Tech Spec 3.3.3.5 2 OP-901-201, Steam Generator Level Control Malfunction Tech Requirement Manual 3.3.5 3 OP-901-102, CEA or CEDMCS Malfunction OP-901-212, Rapid Plant Power Reduction

OP-004-004, Control Element Drive Tech Spec 3.2.3, 3.1.3.1, 3.1.3.5 4 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-002, Loss of Coolant Accident Recovery 5 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operat ing Procedures Operations Expectations /

Guidance 6 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operat ing Procedures Operations Expectations /

Guidance 7 OP-902-008, Safety Function Recovery Procedure OP-902-009, Standard Appendic es, Appendix 28, Aligning LPSI to Replace CS Appendix D Scenario Outline Form ES-D-1 Scenario 2 Rev 1 Facility: WATERFORD 3 Scenario No.: 2 Op Test No.:

NRC Examiners: Operators: Initial Conditions:

• Reactor power is 77%

• Protected Train is B

• AB Bus is aligned to Train B

• Emergency Diesel Gener ator A is tagged out Turnover:

• Charging Pumps A & B are operating

• Boron Equalization is in progress

• Re-commence power ascension Event No. Malf. No. Event Type* Event Description 1 N/A R - ATC N - BOP N - SRO Re-commence power ascension to 100% power 2 RX14A I - ATC I - SRO Pressurizer pressure instrument RC-IPR-0100 X fails low, OP-901-120, Pressurizer

Pressure Control Malfunction 3 RC16B I - BOP I - SRO TS - SRO RCP 1A speed instrument failure, Channel B, Core Protection Calculator B trip 4 N/A TS - SRO Dry Cooling Tower Fan 8B failure 5 DI-07a8s06-1 DI-07a8s12-1 I - BOP I - SRO Inadvertent Containment Spray Actuation OP-901-504, Inadvertent ESFAS Actuation 6 MS11B M - All Main Steam line break inside Containment, S/G #2, OP-902-004, Excess Steam Demand

Recovery 7 N/A C - BOP C - SRO Initiate Containment Spray flow 8 RP09E C - ATC C - SRO Relay K301 failure, BAM-113 A and CVC-183 fail to position on Safety Injection

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Scenario Event Description NRC Scenario 2 Scenario 2 Rev 1 The crew assumes the shift at ~77% power with instructions to raise power to 100%.

After assuming the shift, the crew will commence raising power to 100%. The ATC operator will dilute to the Volume Control Tank and withdraw Group 6 CEAs. The BOP operator will adjust Main Turbine load to raise power.

When an adequate power ascension has occurred, Pressurizer pressure instrument RC-IPR-0100 X will fail low. Since Boron Equalization is in progress, the Main Spray valves will close. The SRO will enter OP-901-120, Pressurizer Pressure Control Malfunction, and select the non-faulted pressure channel.

After Channel Y has been selected for Pressu rizer pressure contro l, Reactor Coolant Pump 1A speed sensor for Core Protection Calculator B will fail.

CPC B will trip as a result of the failure. The SRO should enter Tech Spec 3.3.1 and have the BOP operator bypass bistable s 3 and 4 on Channel B.

After the bypass operation is complete, the Outside Watch will call and report an oil failure on Dry Cooling Tower Fan 8B. The SR O should enter Tech Spec 3.7.4 action d. His review of ambient tem perature and Tech Spec 3.7.4 should conclude that Train B Ultimate Heat Sink remains operable and that Tech Spec 3.8.1.1 is being complied with.

After the Tech Spec review is complete, an inadvertent Containment Spray Actuation will occur. Component Cooling Water flow to the Reactor Coolant Pumps will be secured. The SRO should enter OP-901-504, Inadvertent ESFAS Actuation. The Containment Spray Pumps should be secur ed. If the Com ponent Cooling Water Isolations to the Reactor Coolant Pumps ar e not restored within 3 minutes, the reactor should be tripped and the Reactor Coolant Pumps secured.

A Main Steam line break will develop on Stea m Generator #2 after the preceding event.

If the crew restored CCW to the Reactor Coolant Pumps, the crew should perform a manual reactor trip due to the excess steam demand. If the crew tripped the reactor and secured Reactor Coolant Pumps on the pr evious event, then the Main Steam line break will ramp in after the reactor trip.

Because the Containment Spray Pumps control switches maintain off, the BOP should re-s tart Containment Spray Pumps A and B after Containment pressure rises above 17.7 psia.

Relay K301 will not actuate and BAM-113 A will fail to open and CVC-183 will fail to close on the Safety Injection Actuation. The ATC operator should position these valves to ensure Emergency Boration. After Steam Generator #2 blows down, the crew will take action to maintain RCS temperat ure and pressure. The scenario can be terminated after these actions are complete.

NRC Scenario 2 Scenario 2 Rev 1 Critical Tasks

1. Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow. The required task becomes applicable after Co ntainment Spray has been actuated. The time requirement of 3 minutes is based on the RCP operating limit of 3 minutes without CCW cooling. If the crew does not restore CCW flow to the RCPs after the inadvertent CSAS, then the 3 minute criteria starts at the time of that CSAS. If the crew restores CCW flow to the RCPs following the inadvertent CSAS, then the 3 minu te criteria starts after the Main Steam line break.

2. Establish Containment tem perature and pressure control

This task is satisfied by manually starting at least 1 Containment Spray Pump following the Main Steam line break. This should be co mpleted before comple ting the review of OP-902-000, Standard Post Trip Actions.

3. Establish RCS temperature control

This task is satisfied by taking action to st abilize RCS temperature within the limits of the RCS P/T curve using ADV #1 and establishing EFW flow to Steam Generator #1.

Action to address this task should commenc e within 10 minutes after the applicable parameters begin to rise.

4. Establish RCS pressure control

This task is satisfied by taking action to stabilize RCS pressure within the limits of the RCS P/T curve and additionally maintain RCS pressure within 1500-1600 psid of the faulted steam generator. Action to addr ess this task should commence within 10 minutes after the applicable parameters begin to rise.

Scenario Quantitative Attributes 1. Total malfunctions (5-8) 7 2. Malfunctions after EOP entry (1-2) 2 3. Abnormal events (2-4) 2 4. Major transients (1-2) 1 5. EOPs entered/requiring substantive actions (1-2) 1 6. EOP contingencies requiring substantive actions (0-2) 0 7. Critical tasks (2-3) 4 NRC Scenario 2 Scenario 2 Rev 1 Scenario Notes:

A. Reset Simulator to IC-192. B. Verify the following Scenario Malfunctions: 1. eg10a for EDG A tagout

2. rx14-A for Pressurizer pressure instrument RC-IPT-0100 X 3. rc16b for RCP 1A speed

4. ms11b for Main Steam line break S/G #2

5. rp09e for Relay K301 C. Verify the following Overrides: 1. di-07a08s06-1 and di-07a08s12-1 for CSAS D. Ensure Protected Train B sign is placed in SM office window.

E. Verify the following C ontrol Board Conditions: 1. Danger tag placed on EDG A control switch

2. Danger tag placed on EDG A Output Breaker F. Verify EOOS is 8.5 Yellow G. Complete the simulator setup checklist.

H. Start DCS, Record Data, select file PlantParameters.txt.

NRC Scenario 2 Scenario 2 Rev 1 Simulator Booth Instructions

Event 1 Perform Power Ascension

1. No communications should occur for this evolution.

Event 2 Pressurizer Pressure Instrument Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 2.

2. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 RCP 1A Speed Instrument Failure

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Event 4 Dry Cooling Tower Fan 8B Fan Failure

1. On Lead Examiner's cue, call the CRS as the Outside Watch and report that Dry Cooling Tower Fan 8B has no oil in the r eduction gear sightglass. There is oil on the ground under the fan. This discovery is made during rounds. 2. If Work Week Manager or PMM is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Event 5 Inadvertent CSAS

1. On Lead Examiner's cue, initiate Event Trigger 5. 2. No communications should occur for this evolution.

Event 6 Main Steam Line Break S/G #2

1. On the Lead Examiner's cue, or after the reactor is manually tripped in the previous event, initiate Event Trigger 6. 2. When called as the Outside Watch to check Main Steam Safeties not lifting, report that no safety valves are lifting.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2011 SRO Scenario 2.cdf. Save the file into the folder for the appropriate crew.

NRC Scenario 2 Scenario 2 Rev 1 Scenario Timeline: Event Malfunction Severity Ramp HH:MM:SS Delay Trigger 1 N/A N/A N/A N/A N/A Power ascension 2 RX14A 0% N/A N/A 2 Pressurizer pressure RC-IPR-0100 X fails low 3 RC16B N/A N/A N/A 3 RCP 1A Speed failure, Channel B 4 N/A N/A N/A N/A N/A Dry Cooling Tower Fan 8B failure 5 Di-07a8a06-1 DI-07a8s12-1 N/A N/A N/A 5 Inadvertent Containment Spray 6 MS11B 10% 3:00 N/A 6 Main Steam line break, S/G #2 7 N/A N/A N/A N/A N/A Initiate Containment Spray flow 8 RP09E N/A N/A N/A N/A Relay K301 failure

NRC Scenario 2 Scenario 2 Rev 1

REFERENCES:

Event Procedures 1 OP-010-003, Plant Startup OP-002-005, Chemical and Volume Control 2 OP-901-120, Pressurizer Pressure Control Malfunction 3 OP-009-007, Plant Protection System Tech Spec 3.3.1 4 Tech Spec 3.7.4 and 3.8.1.1 OP-100-014, Technical Specificat ion and Technical Requirements Compliance 5 OP-901-504, Inadvert ent ESFAS Actuation 6 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-004, Excess Steam Demand Recovery 7 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operat ing Procedures Operations Expectations /

Guidance 8 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operat ing Procedures Operations Expectations /

Guidance Appendix D Scenario Outline Form ES-D-1 Scenario 3 Rev 1 Facility: WATERFORD 3 Scenario No.: 3 Op Test No.:

NRC Examiners: Operators: Initial Conditions:

• Reactor power is 5.7 e -2 %

• Protected Train is B

• AB Bus is aligned to Train B Turnover:

• Maintain power during Main Feedwater Pump preparations Event No. Malf. No. Event Type* Event Description 1 Di-08a07s11-1 C - BOP C - SRO TS - SRO Relay K402 fails, MS-401 B opens Emergency Feedwater Pump AB starts 2 FW51A TS - SRO Condensate Storage Pool level instrument EFW-ILI-9013 A fails low 3 CV01B C - ATC C - SRO TS - SRO Charging Pump B trips OP-901-112, Charging or Letdown Malfunction 4 RC09A C - BOP C - SRO Reactor Coolant Pump 1A middle seal failure OP-901-130, Reactor Coolant Pump

Malfunction 5 RC04A RP02 A-D RP01A I - ATC I - SRO RCP 1A shaft shear, automatic reactor trip failure 6 SG01B M - All Steam Generator #2 Tube Rupture OP-902-007, Steam Generator Tube Rupture Recovery 7 SI02A C - BOP C - SRO High Pressure Safety Injection Pump A fails to auto-start 8 RP08C C - ATC C - BOP C - SRO Relay K202 fails, CVC-401, CVC-109, IA-909, and FP-601 A fail to close

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Scenario Event Description NRC Scenario 3 Scenario 3 Rev 1 The crew assumes the shift at 5.7 e-2 % power with instructions to maintain power while Main Feedwater Pumps are prepared for star ting. A lube oil problem common to both Main Feedwater Pumps has delayed their start by approximately 30 minutes.

After assuming the shift, relay K402 fa ils, opening MS-401 B and starting Emergency Feedwater Pump AB, the steam driv en Emergency Feedwater pump. RCS TCOLD will drop and power will rise. The crew should detect the failure and override close MS-401 B. After MS-401 B is closed, Tech Specs 3.7.1.2 action a and 3.6.

3 become applicable due to MS-401 B being inoperable.

After MS-401 B is closed and Tech Specs are addressed, Condensate Storage Pool

level indicator EFW-ILI-9013 A will fail low. The SRO should use OP-903-013, Monthly channel Checks, and enter Tech Spec 3.3.3.5 and 3.3.3.6.

After Tech Specs are addressed, Charging Pu mp B will trip. The crew should start either Charging Pump A or AB and enter OP-901-112, Charging or Letdown Malfunction. Tech Spec 3.1.2.4 and TRM 3.1.2.4 are applicable due to the failure. If the crew aligns Charging Pump AB to replace Charging Pump B, Tech Spec 3.1.2.4 can be exited, but they should remain in TRM 3.1.2.4.

After Tech Specs are addressed, Reactor Coolant Pump 1A middle seal will fail, requiring entry into OP-901-130, Reactor Coolant Pump Malfunction. The crew should monitor for indications of multiple seal failure and reduce CCW temperature to control RCP 1A seal bleedoff temperature.

After actions have been taken to lower CCW temperature, RCP 1A will have a shaft shear. The reactor will fail to automatically trip. The ATC should detect this condition

and trip the reactor. One of the two normal reactor trip pushbuttons will fail and only 2 reactor trip breakers will open.

The ATC will be required to trip the reactor using the Diverse Reactor Trip pushbuttons.

A Steam Generator Tube Rupture will ramp in on the reactor tr ip for S/G #2. The crew should detect this situation during their Standard Post Trip Actions. This failure will require entry into OP-902-007, Steam Gener ator Tube Rupture Recovery. On the Safety Injection Actuation, High Pressure Safe ty Injection Pump A will fail to auto start, requiring a manual start. Additionally, re lay K202 will fail, pr eventing CVC-401, CVC-109, IA-909, and FP-601 A from closing on the Containment Isolation signal. The ATC and BOP operators should close these valves.

The Steam Generator Tube Rupture will require a rapid RCS cooldown to less than 520 degrees THOT. After the RCS is < 520 degrees, Steam Generator #2 will be isolated.

The scenario can be terminated after Steam Generator #2 is isolated.

NRC Scenario 3 Scenario 3 Rev 1 Critical Tasks

1. Manually trip the Reactor.

This task is satisfied by manually tripping the reactor within 1 minute of the failure of the automatic trip. The required task becomes applicable after the annunciators are received associated with the RCP 1A sheared shaft.

2. Prevent Opening the Main Steam Safety Valves.

This task is satisfied by the crew taking action to maintain Steam Generator #2 pressure below the safety valve setpoint by taking action to reduce RCS pressure to < 945 psia.

3. Isolate Steam Generator #2.

This task is satisfied by isolating Steam Generator #2 in accordance with step 17 after RCS

T HOT is reduced below 520 °F.

Scenario Quantitative Attributes 1. Total malfunctions (5-8) 8 2. Malfunctions after EOP entry (1-2) 2 3. Abnormal events (2-4) 2 4. Major transients (1-2) 1 5. EOPs entered/requiring substantive actions (1-2) 1 6. EOP contingencies requiring substantive actions (0-2) 0 7. Critical tasks (2-3) 3 NRC Scenario 3 Scenario 3 Rev 1 Scenario Notes:

A. Reset Simulator to IC-193. 1. Use keys 165 - 168 for S/G high level bypass setup. B. Verify the following Scenario Malfunctions: 1. fw51a for CSP level instrument

2. cv01b for Charging Pump B

3. rc09a for RCP 1A seal failure

4. rc04a for RCP 1A shaft shear

5. rp02 a-d for RPS auto trip failure

6. rp01a for CP-2 pushbutton failure

7. sg01b for S/G #2 tube rupture

8. si02a for High Pressure Safety Injection Pump

9. rp08c for K202 C. Ensure Protected Train B sign is placed in SM office window.

D. Verify EOOS is 10.0 Green E. Complete the simulator setup checklist.

F. Start DCS, Record Data, select file PlantParameters.txt.

NRC Scenario 3 Scenario 3 Rev 1 Simulator Booth Instructions

Event 1 Relay K402 Failure, EFW Pump AB Starts

1. On the Lead Examiner's cue, initiate Event Trigger 1. 2. If directed to check EFW Pump AB, report it is running satisfactorily. 3. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Event 2 Condensate Storage Pool Level instrument EFW-ILI-9013 A Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 2.

2. If called to check the indication at the Remote Shutdown Panel, report that Condensate Storage Pool Level instru ment EFW-ILI-9013 A is reading 0%. 3. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 Charging Pump B Trip

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If called to check the Charging Pump that was started, report that it is running satisfactorily. 3. If called to check the Charging Pump B, report that the overcurrent relays are picked up on all 3 phases and that the motor has a strong, odor. 4. If Work Week Manager or PMM is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Event 4 RCP 1A Middle Seal Failure

1. On the Lead Examiner's cue, initiate Event Trigger 4. 2. If called as the RCP Engineer, report that you will monitor the status of RCP 1A. 3. If Work Week Manager is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 5 RCP 1A Shaft Shear

1. On Lead Examiner's cue, initiate Event Trigger 5.

Event 6 Steam Generator #2 Tube Rupture

1. Verify SGTR begins ramping in after the reactor trip. If not, initiate Event Trigger

6. 2. Acknowledge calls to Chemistry and/or Health Physics to carry out requested actions.

NRC Scenario 3 Scenario 3 Rev 1 Event 7 High Pressure Safety Injection Pump A

1. No communications should occur for this malfunction.

Event 8 Relay K202 Failure

2. No communications should occur for this malfunction.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2011 SRO Scenario3.cdf. Save the file into the folder for the appropriate crew.

NRC Scenario 3 Scenario 3 Rev 1 Scenario Timeline: Event Malfunction Severity Ramp HH:MM:SS Delay Trigger 1 DI-08a07s11-1 N/A N/A N/A 1 Relay K402 failure 2 FW51A 0% N/A N/A 2 CSP Level indication fails low 3 CV01B N/A N/A N/A 3 Charging Pump B trip 4 RC09A 100% N/A N/A 4 RCP 1A middle seal failure 5 RC04A RP02 A-D RP01A N/A N/A N/A 5 RCP 1A sheared shaft, auto trip failure 6 SG01B 10% 3:00 N/A 6 Steam Generator #2 Tube Rupture 7 SI02A N/A N/A N/A N/A High Pressure Safety Injection Pump A fails to auto start 8 RP08C N/A N/A N/A N/A Relay K202 fails to actuate

NRC Scenario 3 Scenario 3 Rev 1

REFERENCES:

Event Procedures 1 Tech Spec 3.7.1.2. 2 OP-903-013, Monthl y Channel Checks Tech Spec 3.3.

3.5 and 3.3.3.6 3 OP-002-005, Chemical and Volume Control Tech Spec 3.2.1.4 4 OP-901-130, Reactor C oolant Pump Malfunction 5 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operat ing Procedures Operations Expectations /

Guidance 6 OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-007, Steam Generator Tube Rupture Recovery 7 OI-038-000, Emergency Oper ating Procedures Operations Expectations /

Guidance 8 OI-038-000, Emergency Operating Proc edures Operations Expectations /

Guidance Appendix D Scenario Outline Form ES-D-1 Scenario 4 Rev 1 Facility: WATERFORD 3 Scenario No.: 4 Op Test No.:

NRC Examiners: Operators: Initial Conditions:

• Reactor power is 100%

• Protected Train is B

• AB Bus is aligned to Train A Turnover:

• Maintain 100% power Event No. Malf. No. Event Type* Event Description 1 SG10D C - BOP C - SRO TS - SRO Steam Generator #1 leve l instrument SG-ILI-1113 D fails high.

2 TP01A TP08B C - BOP C - SRO Turbine Cooling Water Pump A trips, Turbine Cooling Water Pump B fails to auto start OP-901-512, Loss of Turbine Cooling Water 3 FW03A C - ATC C - SRO TS - SRO Main Feedwater Pump A trips, Reactor Power Cutback OP-901-101, Reactor Power Cutback 4 RD07D R - ATC Regulating Group 4 CEAs fail to insert in automatic following Reactor Power Cutback

5 FW03B FW07A M - All N - SRO Main Feedwater Pump B trips, manual reactor trip, Emergency Feedwater Pump A fails to run 6 RP03 C - BOP C - SRO Main Turbine fails to trip following the reactor trip 7 RD11A 28, 37, 79 C - ATC C - SRO 3 CEAs fail to insert following the reactor trip, Emergency Boration 8 FW05 C - BOP C - ATC C - SRO Emergency Feedwater Pump AB trip on overspeed

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Scenario Event Description NRC Scenario 4 Scenario 4 Rev 1 The crew assumes the shift at 100% power with instructions to maintain 100% power.

After assuming the shift, Steam Generator #1 level instrument SG-ILI-1113 D fails high.

The SRO should review Tech Specs and enter Tech Spec 3.3.1 and 3.3.2 and TRM

3.3.1. The SRO should direct the BOP operator to bypass the bistables for low Steam Generator #1 level, hi gh level and Steam Generator #1 di fferential pressure for channel D. This instrument does apply to Tech Spec 3.3.3.6 for Accident Monitoring, but the minimum channel requirements are met using other channels.

After the proper bistables are bypassed, Turb ine Cooling Water Pump A trips. Turbine Cooling Water Pump B fails to start.

The SRO should enter OP-901-512, Loss of Turbine Cooling Water, and star t Turbine Cooling Water Pump B. The Plant Monitoring Computer will display an overload condition for TCW Pump A.

After Turbine Cooling Water Pump B is r unning, Main Feedwater Pump A will trip. A Reactor Power Cutback will occur. The ATC should perform the immediate operator actions. The SRO should enter OP-901-101, Reactor Power Cutback. Following the Cutback, Regulating Group 4 CEAs will fail to insert in automatic. The SRO should enter Tech Spec 3.2.4 for DNBR and 3.2.7 fo r ASI. The crew should take action to address the DNBR power operati ng limit within 15 minutes by performing ASI control with Group P CEAs.

After the crew has addressed Tech Spe cs and commenced ASI control with Group P CEAs, Main Feedwater Pump B will trip. Th e crew should perform a manual reactor trip based on this failure. On the Emergency F eedwater Actuation, Emergency Feedwater Pump A will fail to start and wil l not start manually. The Main Turbine will fail to trip on the reactor trip. The BOP should manually trip the Main Turbine. 3 CEAs will fail to insert on the reactor trip. The ATC operator should perform Emergency Boration due to this condition. The SRO should enter OP-9 02-006, Loss of Main Feedwater Recovery.

The ATC operator should secure 2 Reactor Coolant Pumps.

After 2 Reactor Coolant Pumps are secur ed, Emergency Feedwater Pump AB will trip due to operator error locally. The crew s hould remain in OP-902-006 and secure the remaining Reactor Coolant Pump

s. On investigation, the local watchstander will report Emergency Feedwater Pump AB is ready to be reset. The BOP operator should perform the necessary actions for re setting Emergency Feedwater Pump AB.

The scenario can be terminated after Emergency Feedwater Pump AB is reset.

NRC Scenario 4 Scenario 4 Rev 1 Critical Tasks

1. Establish reactivity control.

This task is satisfied by establishing Emergency Boration prior to completing Standard Post Trip Actions Reactivity Control verification. The required task becomes applicable after the Reactor is tripped and 3 CEAs remain stuck out.

2. Establish a primary to secondary heat sink

This task is satisfied by securing all RCPs after Emergency Feedwater Pump AB trips. With Emergency Feedwater Pump A off, Emergency Feedwater Pump B does not have the capacity to provide necessary Emergency Feedwater flow.

Scenario Quantitative Attributes 1. Total malfunctions (5-8) 8 2. Malfunctions after EOP entry (1-2) 3 3. Abnormal events (2-4) 2 4. Major transients (1-2) 1 5. EOPs entered/requiring substantive actions (1-2) 1 6. EOP contingencies requiring substantive actions (0-2) 0 7. Critical tasks (2-3) 2 NRC Scenario 4 Scenario 4 Rev 1 Scenario Notes:

A. Reset Simulator to IC-194. B. Verify the following Scenario Malfunctions: 1. sg10d for S/G #1 level instrument

2. tp01a for TCW Pump A

3. tp08b for TCW Pump B

4. fw03a for Main Feedwater Pump A

5. rd07d for Regulating Group 4 CEAs

6. fw03b for Main Feedwater Pump B

7. fw07a for EFW Pump A

8. rp03 for the Main Turbine failure

9. rd11a28, 37, and 79 for CEAs 28, 37, and 79

10. fw05 for EFW Pump AB C. Verify the following Override: 1. di-08a04s09-1 for EFW Pump A D. Ensure Protected Train B sign is placed in SM office window.

E. Verify EOOS is 10.0 Green F. Complete the simulator setup checklist.

G. Start DCS, Record Data, select file PlantParameters.txt.

NRC Scenario 4 Scenario 4 Rev 1 Simulator Booth Instructions

Event 1 Steam Generator #1 level instrument failure

1. On the Lead Examiner's cue, initiate Event Trigger 1. 2. If directed to check the remote shut down panel, report that Channel D S/G #1 level reads 67%. 3. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Event 2 Turbine Cooling Water Pump A trip

1. On the Lead Examiner's cue, initiate Event Trigger 2. 2. If directed to check Turbine Cooling Water Pumps locally, report TCW Pump A has overcurrent flags tripped and t hat TCW Pump B looks normal. 3. If Work Week Manager is called, info rm the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 Main Feedwater Pump A trip, Reactor Power Cutback

1. On the Lead Examiner's cue, initiate Event Trigger 3. 2. If directed to check Main Feedwater Pump A locally, report there are no abnormal indications locally.

Event 5 MFW Pump B trip, Reactor trip, Emergency Feedwater Pump A trip

1. On the Lead Examiner's cue, initiate Event Trigger 5. 2. If directed to check Main Feedwater Pump B locally, report indications of broken linkages on the governor assembly. 3. If directed to check EFW Pump A locall y, report indications of a broken breaker for EFW Pump A at Switchgear 3A.

Event 8 Emergency Feedwater Pump AB trip

1. On the Lead Examiner's cue, initiate Event Trigger 8. 2. After the remaining Reac tor Coolant Pumps are tripped, call as the RCA watch and report that the Emergency Feedwater Pump AB tripped on overspeed due to his activities while checking the pump.

Recommend performing actions to reset EFW Pump AB.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2011 SRO Scenario 4.cdf. Save the file into the folder for the appropriate crew.

NRC Scenario 4 Scenario 4 Rev 1 Scenario Timeline: Event Malfunction Severity Ramp HH:MM:SS Delay Trigger 1 SG10D 100% N/A N/A 1 S/G #1 level instrument channel D fails high 2 TP01A TP08B N/A N/A N/A 2 TCW Pump A trips, TCW Pump B fails to auto-start 3 FW03A N/A N/A N/A 3 MFW Pump A trips 4 RD07D N/A N/A N/A N/A Regulating Group 4 fails to auto insert 5 FW03B FW07A DI-08a04s09-1 N/A N/A N/A 5 MFW Pump B trips, EFW Pump A fails to run 6 RP03 N/A N/A N/A N/A Main Turbine fails to trip on reactor trip 7 RD11A 28, 37, 79 N/A N/A N/A N/A CEAs 28, 37, 79 fail to insert 8 FW05 N/A N/A N/A 8 EFW Pump AB trips

NRC Scenario 4 Scenario 4 Rev 1

REFERENCES:

Event Procedures 1 OP-009-007, Plant Protection System OP-903-013, Monthl y Channel Checks Tech Spec 3.3.1 and 3.3.2 2 OP-901-512, Loss of Tu rbine Cooling Water 3 & 4 OP-901-101, Reactor Power Cutback Tech Spec 3.2.1 5 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-006, Loss of Main Feedwater Recovery 6 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operat ing Procedures Operations Expectations /

Guidance 7 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operat ing Procedures Operations Expectations /

Guidance 8 OP-902-006, Loss of Main Feedwater Recovery

ES-301 Administrative Topics Outline Form ES-301-1 Revision 0 RO Facility: WATERFORD 3 Date of Examination: March 21, 2011 Examination Level:

RO Operating Test Number:

1 Administrative Topic (see Note)

Type Code* Describe activity to be performed A1 Conduct of Operations K/A Importance:

4.3 S, D 2.1.23, Ability to perform specific system and integrated plant procedures during all modes of plant operation. Perform a Shutdown Margin with an immoveable CEA in accordance with OP-903-090, Shutdown Margin, section 7.3, Shutdown Margin Verification - Untrippable CEA.

A2 Conduct of Operations K/A Importance:

3.6 R, M 2.1.18, Ability to make accurate, clear, and concise logs, records, status boards, and reports. Perform OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data.

A3 Equipment Control K/A Importance:

3.7 R, N 2.2.12, Knowledge of surveillance procedures Complete surveillance OP-903-013, Monthly Channel Checks, Attachment 10.3 for Accident Monitoring Instrumentation Channel Checks.

A4 Radiation Control K/A Importance:

3.2 R, N 2.3.4, Knowledge of radiation exposure limits under normal and emergency conditions. Calculate stay time to perform a tagout verification.

Room dose rate & operator's yearly dose provided.

Emergency Plan Not selected NOTE: All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 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)