ML15167A529

2015-06 Draft Outlines

ML15167A529
Person / Time
Site:	Comanche Peak
Issue date:	06/05/2015
From:	Vincent Gaddy Operations Branch IV
To:	Luminant Generation Co
References
50-445/15-006, 50-446/15-006
Download: ML15167A529 (32)

Text

ES-401 Record of Rejected K/As Form ES-401-4 Tier/Group Randomly Reason for Rejection Selected K/A RO Exam 1/1 015/017 AK1.02 Question 43 Reactor Coolant Pump Malfunctions: Could not write a discriminating question based on the content of the original K/A. Replaced K/A 015/017 AK1.05 with K/A 015/017 AK1.02 1/1 W/E05 G2.1.28 Question 54 Loss of Instrument Air function of major components: Original sample plan resulted in Instrument Air system and malfunctions being over sampled (five occurrences).

Replaced K/A 065 G2.1.28 with W/E05 G.2.1.28.

2/1 022 A4.01 Question 10 Loss of CCS pump: Unable to write an operationally valid question based on CPNPP design for the original K/A.

Replaced K/A 022 A2.06 with K/A 022 A4.01.

2/1 006 K4.17 Question 28 Ability to monitor ECCS Valve lineups: Tier 2 Group 1 also sampled ECCS with A3.04, which resulted in over sampling A3 for ECCS. Replaced K/A 006 A3.06 with K/A 006 K4.17.

2/2 017 A3.01 Question 35 Monitor the ITM system: Task not performed by operators; unable to write an operationally valid question for the original K/A. Replaced K/A 017 A3.02 with K/A 017 A3.01.

2/2 086 A1.02 Question 38 Fire Protection, Fire Dampers: Could not write a discriminatory, operationally valid question for the original K/A.

Replaced K/A 086 A1.04 with K/A 086 A1.02.

SRO Exam 1/1 026 G.2.1.20 Question 78 Loss of Component Cooling Water: Original K/A is not allowed for selection in Tier 1 and Tier 2. Replaced K/A 026 G.2.1.45 with K/A 026 G.2.1.20.

1/1 058 G.2.2.40 Question 79 Loss of DC Power: Original K/A is not allowed for selection in Tier 1 and Tier 2. Replaced K/A 058 G2.2.23 with K/A 058 G.2.2.40.

1/1 062 G.2.4.35 Question 80 Loss of Instrument Air local operator actions: Original sample plan resulted in Instrument Air system and malfunctions being over sampled (five occurrences). Replaced K/A 065 G.2.4.35 with K/A 062 G.2.4.35.

1/2 067 G.2.4.11 Question 84 Knowledge of Fire Protection Procedures: Original K/A is not allowed for selection in Tier 1 and Tier 2. Replaced K/A 067 G.2.4.25 with K/A 067 G.2.4.11.

2/1 004 G.2.1.7 Question 86 Chemical and Volume Control System: Original K/A is not allowed for selection in Tier 1 and Tier 2. Replaced K/A 004 G.2.1.45 with K/A 004 G.2.1.7.

ES-401 Record of Rejected K/As Form ES-401-4 2/1 103 G2.4.28 Question 90 Containment Generic Radiological procedures: Original K/A is not allowed for selection in Tier 1 and Tier 2. Replaced K/A 103 G2.3.13 with K/A 103 G2.4.28.

2/2 071 G2.1.23 Question 92 Waste Gas Disposal (Generic): Original K/A is not allowed for selection in Tier 1 and Tier 2. Replaced K/A 071 G2.3.11 with K/A 071 G2.1.23.

2/2 086 G2.4.30 Question 93 Fire Protection (Generic): Original K/A is not allowed for selection in Tier 1 and Tier 2. Replaced K/A 086 G2.4.26 with K/A 086 G2.4.30.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: CPNPP Units 1 and 2 Date of Examination: June 2015 Examination Level: RO SRO Operating Test Number: NRC Administrative Topic Type Code* Describe activity to be performed (See Note) 2.1.23 Ability to perform specific system and integrated plant procedures during all modes Conduct of Operations of plant operation. (4.3).

(RA1) M,R JPM: Perform Power Change Worksheet Calculation. (RO1302A) 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo Conduct of Operations operation, maintenance of active license (RA2) N,R status, 10CFR55, etc. (3.3).

JPM: Determine RO License Status.

2.2.12 Knowledge of surveillance procedures. (3.7).

Equipment Control (RA3) D,R JPM: Perform Axial Flux Difference Surveillance.

(RO1808) 2.3.11 Ability to control radiation releases. (3.8).

Radiation Control (RA4) M,R JPM: Determine Maximum Allowable Venting Time. (RO7030)

Emergency Procedures/Plan 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-301-1 CPNPP 2015 NRC RO Administratives Topics Outline

ES-301 Administrative Topics Outline Form ES-301-1 Task Summary RA1 The applicant will determine boron/dilution requirements to lower power from 100% to 50% equilibrium per IPO-003A, Power Operations, Attachment 3, Power Change Worksheet. The calculations include Power Defect, Rod Worth, Xenon Worth, change in Boron concentration and boration/dilution quantity. The critical steps are to determine the reactivity change for power defect, the rod position change, equilibrium xenon, and the total reactivity change for these parameters, the required change in boron concentration, and the amount of boration needed.

This is a modified bank JPM.

RA2 The applicant will be presented with a detailed record (in table form) of watch standing and other activities performed by 3 individual Reactor Operators over a period of 4 to 6 weeks. The applicant will be required to analyze the work records of these three operators, and apply the guidance of ODA-315, Licensed Operator Maintenance Tracking, to evaluate and determine if the RO license status is active or inactive for each of the three operators. The critical steps are to determine that the RO licenses for two of the three operators are NOT active.

This is a new JPM.

RA3 The applicant will be presented with Power Range Nuclear Instrument Axial Flux Difference data and will perform a manual Axial Flux Difference calculation using OPT-403, Axial Flux Difference. The critical steps are to determine whether at least 3 of 4 PR FLUX channels are within the Acceptable Operation region of NUC-204-6, Axial Flux Difference as a Function of Rated Thermal Power. This is a direct from bank JPM.

RA4 The applicant will determine the maximum allowable venting time for venting the reactor vessel using FRI-0.3A, Response to Voids in Reactor Vessel, Attachment

5. Critical steps include various stages of the calculation, including the final determination of allowable venting time. This is a modified bank JPM.

ES-301-1 CPNPP 2015 NRC RO Administratives Topics Outline

ES-301 Administrative Topics Outline Form ES-301-1 Facility: CPNPP Units 1 and 2 Date of Examination: June 2015 Examination Level: RO SRO Operating Test Number: NRC Administrative Topic Type Describe activity to be performed (See Note) Code*

2.1.25 Ability to interpret reference materials, such Conduct of Operations as graphs, curves, tables, etc. (4.2).

(SA1) D,R JPM: Loss of RHR Time / Tech Specs. (SO1101) 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, Conduct of Operations such as medical requirements, no-solo (SA2) N,R operation, maintenance of active license status, 10CFR55, etc. (3.8).

JPM: Determine SRO License Status.

2.2.23 Ability to track Technical Specification Equipment Control limiting conditions for operations. (4.6).

(SA3) D,R JPM: Complete LCOAR for TDAFW Pump Steam Admission Valve. (SO1024D) 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

Radiation Control (3.7).

(SA4) D,R JPM: Select Volunteer for Emergency Exposure.

(SO1142A) 2.4.44 Knowledge of emergency plan protective Emergency action recommendations. (4.4).

Procedures/Plan M,P,R (SA5) JPM: Determine Protective Action Recommendations. (SO1140A)

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-301-1 CPNPP 2015 NRC SRO Administrative Topics Outline

ES-301 Administrative Topics Outline Form ES-301-1 Task Summary SA1 The applicant is presented with Loss of RHR conditions and then uses ABN-104, Residual Heat Removal System Malfunction, Attachment 5, Time to Saturation for Loss of All RHR with the RCS at Reduced Inventory and Attachment 19, Available Time for Containment Closure to determine Time to Saturation, Time to Core Uncovery, and Containment Closure Time. The critical steps are to determine Time to Saturation, Time to Core Uncovery, Containment Closure Time, and identify any Technical Specification required actions associated with the loss of the standby RHR pump. This is a direct from bank JPM.

SA2 The applicant is presented with a detailed record (in table form) of watch standing and other activities performed by 3 individual Senior Reactor Operators over a period of 4 to 6 weeks. The applicant will be required to analyze the work records of these three operators, and apply the guidance of ODA-315, Licensed Operator Maintenance Tracking, to evaluate and determine if the SRO license status is active or inactive for each of the three operators. The critical steps are to determine that the SRO licenses for two of the three operators are NOT active.

This is a new JPM.

SA3 The applicant will be presented with conditions involving a TDAFW Pump Steam Admission Valve that has not been returned to service within the Completion Time and will use ODA-308, LCO Tracking Program, and Technical Specification 3.7.5 Auxiliary Feedwater System, to manually complete a Tracking LCOAR.

The critical steps consist of various determinations on the LCOAR form, including correct information in required fields to pass the JPM. This is a direct from bank JPM.

SA4 The applicant is given accident conditions involving the need for a volunteer to attempt a lifesaving activity. Using the guidance in EPP-305, Emergency Exposure Guidelines and Personnel Dosimetry, the applicant will evaluate a series of potential volunteers and select the preferred volunteer from this list.

The critical steps are evaluation and elimination of volunteers who do not meet the criteria required for the activity, and then final selection of the preferred volunteer. This is a direct from bank JPM.

ES-301-1 CPNPP 2015 NRC SRO Administrative Topics Outline

ES-301 Administrative Topics Outline Form ES-301-1 Task Summary SA5 The applicant will determine the appropriate Protective Action Recommendations for an emergency. This JPM is designated as a "P" because a form of it was used on the 2013 NRC exam. This JPM will be modified to include different conditions, including severity and meteorological parameters. The "random selection" aspect was performed due to limited topics available for SRO A.4 category, the fact that this JPM meets the requirements of NUREG-1021, and to avoid overlap with the Audit Exam. The critical steps will include several determinations the SRO must make, such as release duration, core damage, and identification of affected sectors. This is a modified bank JPM.

ES-301-1 CPNPP 2015 NRC SRO Administrative Topics Outline

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Facility: CPNPP Units 1 and 2 Date of Examination: June 2015 Exam Level: RO SRO(I) SRO (U) Operating Test Number: NRC Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / JPM Title Type Code*

Function 001 -Control Rod Drive System (RO1030) A,D,S 1 S-1 Respond to Control Rods Below Insertion Limit 006 -Emergency Core Cooling System (RO1506D) A,EN,L,M,S 2 S-2 Transfer ECCS to Cold Leg Recirculation 006 -Emergency Core Cooling System (RO1511) A,EN,L,N,S 3 S-3 Isolate SI Accumulators Following a LOCA 005 -Residual Heat Removal System (RO1402) L,N,S 4P S-4 Alternate Residual Heat Removal Trains 045 -Main Turbine Generator System (RO3149) L,N,S 4S S-5 Roll Main Turbine to 1800 RPM (RO Only) 022 -Containment Cooling System (RO2101A) A,N,S 5 S-6 Alternate Containment Recirculation Units(CACRS) 064 -Emergency Diesel Generator System (RO4302D) A,D,S 6 S-7 Load Emergency Diesel Generator 008 -Component Cooling Water System (RO3603C) M,S 8 S-8 Rotate Component Cooling Water Pumps In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) 007 -Reactor Trip (AO6439B) D,E,R 1 P-1 Trip the Reactor and Stop MG Sets 068 -Control Room Evacuation (AO6415A) D,E,L,R 8 P-2 Place MDAFW Pump on Alternate Suction Source 062 -AC Electrical Distribution System (AO4204D) N,E 6 P-3 Transfer Inverter IVuPC1 from Bypass to Normal Page 1 of 4 ES-301-2 CPNPP 2015 RO & SRO System JPM Outline

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2

@ 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 (C)ontrol room (D)irect from bank <9 / <8 / <4 (E)mergency or abnormal in-plant >1 / >1 / >1 (EN)gineered safety feature - / - / > 1(control room system)

(L)ow-Power / Shutdown >1 / >1 / >1 (N)ew or (M)odified from bank including 1(A) >2 / >2 / >1 (P)revious 2 exams <3 <3 / < 2 (randomly selected)

(R)CA >1 / >1 / >1 (S)imulator NRC JPM Examination Summary Description S-1 Following a turbine runback, due to a trip of a Heater Drain Pump from 100%

power, the applicant will determine that control rods are below the required rod insertion limit by using ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, Section 4.0, Heater Drain Pump Trip. This is an Alternate Path JPM, requiring the applicant to manually determine the required boration using the Reactivity Briefing Sheet. The critical steps include determination of how much boration is needed, and the various control board manipulations needed to perform the boration. This is a direct from bank JPM under Control Rod Drive System -

Reactivity Control Safety Function. (K/A 001.A4.05 - IR 3.7 / 3.7)

S-2 The applicant will be required to use EOS-1.3A, Transfer to Cold Leg Recirculation following a Large Break LOCA. This is an Alternate Path JPM because one of the RHR pump suction valves will NOT open, and the applicant will need to perform alternate steps for system realignment, including shutting off one pump and ensuring the other RHR pump is running. The critical steps include recognition of one RHR pump suction valve failing to open, and various control board manipulations needed for realignment in order to achieve cold leg recirculation.

This is a PRA significant action. This is a modified from bank JPM under the Emergency Core Cooling System - Reactor System Inventory Control Safety Function. (K/A 006.A4.05 - IR 3.7 / 3.6)

Page 2 of 4 ES-301-2 CPNPP 2015 RO & SRO System JPM Outline

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 S-3 Using EOS-1.2A, Post LOCA Cooldown and Depressurization, the applicant will be required to continue with Step 26 for determining if SI accumulators should be isolated and to isolate the accumulators. This is an Alternate Path JPM and requires the applicant to determine that one of the accumulator injection valves will NOT close. This will require the applicant to vent off this accumulator to minimize the consequences of undesired injection, since the accumulator cannot be isolated. The critical steps include restoring power to the injection valves, operation of the accumulator injection valves, and venting of the accumulator that cannot be isolated. This is a new JPM under the Emergency Core Cooling System

- Reactor Pressure Control Safety Function. (K/A 006.A4.02 - IR 4.0 / 3.8)

S-4 The applicant will use SOP-102A, Residual Heat Removal System, Section 5.6 Alternating RHR Trains in MODE 5, 6, or Defueled to perform the task. The critical steps will include various control board manipulations required for making the swap such as starting and stopping RHR pumps, operation of control valves, and requirements for temperature control. This is a new JPM under the Residual Heat Removal System - Primary System Heat Removal from Reactor Core Safety Function. (K/A 005.A4.01 - IR 3.6 / 3.4)

S-5 The applicant will use IPO-003A, Power Operations, Section 5.1, Warmup and Synchronization of the Turbine Generator, beginning at Step 5.1.18 and completing Step 5.1.21. This involves setting up the turbine control for rolling the turbine to 1800 RPM. The Overspeed Trip test will NOT be required. The critical steps include setting up the turbine control panel to open the HP and LP stop valves, an interim step of holding at 500 RPM, placing of bearing lift oil pumps to AUTO, and then continuing to 1800 RPM where the JPM will terminate. This is a new JPM under the Main Turbine Generator System - Secondary System Heat Removal from Reactor Core Safety Function. RO Only. (K/A 045.A4.02 - IR 2.7 /

2.6)

S-6 With a Containment Vent in progress, the applicant is directed to alternate Containment Recirculation Units using SOP-801A, Containment Ventilation System, Section 5.1.3. During the swap Containment Air Gaseous radiation monitor goes into Alert. This is an alternate path JPM requiring action to manually initiate isolation of the Containment Vent. The critical steps include starting the desired cooling unit and manual operation of several valves for isolation of the Containment Vent evolution. This is a new JPM under the Containment Cooling System - Containment Integrity Safety Function. (K/A 022.A4.01 - IR 3.6 / 3.6)

S-7 With OPT-214A, Diesel Generator Operability Test in progress and following a fast start of Diesel Generator 1-01, the applicant is to continue with the surveillance.

This involves beginning to load the diesel generator. This is an Alternate Path JPM. When loading is raised to approximately 2.2 MW, the Station Service Water Pump 1-01 will trip. This will result in the diesel generator running loaded with no cooling water. The applicant is required to shut down the diesel generator. The critical steps are proper loading of the diesel generator and shutting down the diesel generator to prevent equipment damage. This is a direct from bank JPM under the Emergency Diesel Generator System - Electrical Safety Function. (K/A 064.A4.06 - IR 3.9 / 3.9)

Page 3 of 4 ES-301-2 CPNPP 2015 RO & SRO System JPM Outline

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 S-8 The applicant is directed to swap Component Cooling Water Pumps from Train A to Train B, using SOP-502A, Component Cooling Water System. The critical steps include establishing required system flow prior to and after the swap, control board manipulations required for the swap, starting the idle pump, and shutting down the pump to be idled. This is a modified from bank JPM under the Component Cooling Water System - Plant Service Systems Safety Function.

(K/A 008.A4.01 - IR 3.3 / 3.1)

P-1 With an Anticipated Transient Without Trip in progress on Unit 1, the applicant is required to locally trip the Unit 1 reactor, and to stop both Rod Drive Motor Generator Sets, in accordance with FRS-0.1A, Response to Nuclear Power Generator/ATWT, Step 6a RNO. Through a series of simulated operations and examiner cues, the applicant will open RTA and RTB trip breakers as critical steps.

The bypass breakers will not be considered critical steps. When that is complete, the applicant will de-energize both MG Sets by opening associated breakers, each of which is a critical step. This is a PRA significant action. This is a direct from bank JPM under the Reactor Trip System - Reactivity Control Safety Function.

(K/A 007.EA2.02 - IR 4.3 / 4.6)

P-2 During a Control Room evacuation due a fire, the applicant is required to supply an alternate suction source to Motor Driven Auxiliary Feedwater Pump 1-01, which has tripped due to loss of suction pressure. Actions will be performed using ABN-803A/B, Response to a Fire in the Control Room or Cable Spreading Room, Attachment 9, Alternate AFW Supply. The critical steps include operation of breakers and manual operation of valves required for supplying the alternate suction source (which will be from Station Service Water). This is a PRA significant action. This is a direct from bank JPM under the Control Room Evacuation System - Plant Service Systems Safety Function.

(K/A 068.AA1.26 - IR 3.6 / 3.8)

P-3 The applicant will be directed to perform SOP-607A/B, 118 VAC Distribution System and Inverters, Section 5.5.9 Transferring Inverter IVuPC1 from Bypass to Normal Operation. The critical steps will include operating the Static Transfer Switch to make the swap, and placing of several other controls to complete the operation. This is a new JPM under the AC Electrical Distribution System -

Electrical Safety Function. (K/A 062.A4.04 - IR 2.6 / 2.7)

Page 4 of 4 ES-301-2 CPNPP 2015 RO & SRO System JPM Outline

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Comanche Peak Date of Exam: June 2015 Operating Test No.: 1 A E NRC EXAM OUTLINE SUBMITTAL P V Scenario 1 Scenario 2 T M P E O I L N CREW CREW T N I T A I C POSITION POSITION L M A T S A B S A B U N Y R T O R T O M(*)

T P O C P O C P R I U E

RX --- 1 1 1 1 0 RO 1 NOR 1 --- 1 1 1 1 I/C 2,3,5 2,3,5 6 4 4 2 MAJ 6 6,7 3 2 2 1 TS --- --- 0 0 2 2 RX --- 1 1 1 1 0 NOR 1 --- 1 1 1 1 RO 2 I/C 2,3,5 2,3,5 6 4 4 2 MAJ 6 6,7 3 2 2 1 TS --- --- 0 0 2 2 RX 1 --- 1 1 1 0 NOR --- 1 1 1 1 1 RO 3 I/C 4,5,7,8 4 5 4 4 2 MAJ 6 6,7 3 2 2 1 TS --- --- 0 0 2 2 RX 1 --- 1 1 1 0 NOR --- 1 1 1 1 1 SRO-I 1 I/C 4,5,7,8 5 5 4 4 2 MAJ 6 6,7 3 2 2 1 TS --- 2,3,4 3 0 2 2 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 do one scenario, including at least two instrument or component (I/C) malfunctions and one major transient, in 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 applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Comanche Peak Date of Exam: June 2015 Operating Test No.: 1 A E NRC EXAM OUTLINE SUBMITTAL P V Scenario 1 Scenario 2 T M P E O I L N CREW CREW T N I T A I C POSITION POSITION L M A T S A B S A B U N Y R T O R T O M(*)

T P O C P O C P R I U E

RX 1 --- 1 1 1 0 SRO-I 2 NOR --- 1 1 1 1 1 I/C 4,5,7,8 5 5 4 4 2 MAJ 6 6,7 3 2 2 1 TS --- 2,3,4 3 0 2 2 RX --- --- 0 1 1 0 NOR 1 1 2 1 1 1 SRO-U 1 I/C 3,5 4 3 4 4 2 MAJ 6 6,7 3 2 2 1 TS 2,4 --- 2 0 2 2 RX --- --- 0 1 1 0 NOR 1 1 2 1 1 1 SRO-U 2 I/C 3,5 4 3 4 4 2 MAJ 6 6,7 3 2 2 1 TS 2,4 --- 2 0 2 2 RX --- --- 0 1 1 0 NOR 1 1 2 1 1 1 SRO-U 3 I/C 3,5 5 3 4 4 2 MAJ 6 6,7 3 2 2 1 TS 2,4 2,3,4 5 0 2 2 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 do one scenario, including at least two instrument or component (I/C) malfunctions and one major transient, in 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 applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

Appendix D Scenario Outline Form ES-D-1 June 2015 NRC Exam Facility: CPNPP 1 & 2 Scenario No.: 1 Op Test No.: June 2015 Examiners: Operators:

Initial Conditions: 100% power. CCP 1-02 is tagged out for maintenance.

Turnover: Begin reducing power to 50% for removing Main Feedwater Pump 1-01 from service to repair an oil leak. CCP 1-02 will be returned to service in 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Critical Tasks: Initiate Train A and/or Train B Safety Injection due to Failure to Automatically Actuate prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection.

Establish flow from at least one high-head ECCS pump (CCP 1-01) prior to Exiting EOP-1.0A, Reactor Trip or Safety Injection.

Initiate Cooldown of Reactor Coolant System Prior to Exiting EOS-1.2A, Post LOCA Cooldown and Depressurization.

Event No. Malf. No. Event Type* Event Description 1 ---- R - RO Begin power reduction for removing MFP 1-01 from service.

N - BOP N - SRO 2 CC02A C - BOP CCW Pump 1-01 trips. Standby fails to auto start.

CC03A TS - SRO 3 CH10 C - BOP CRDM Vent Fan trips. Requires manual start of alternate.

C - SRO 4 RP05A I - RO NR Cold Leg TI (TE-411B) fails high.

TS - SRO 5 RC03C C - RO RCP 1-03 vibration (Ramps to 20 mils over 5 min.)

C - BOP C - SRO 6 RC19C M - ALL SBLOCA.

7 RP07A I - RO Both trains SI fail to auto actuate.

RP07B 8 ED05E C - RO Loss of 1EA1 Safeguards Bus (86-2 actuation).

CV01D CCP 1-01 fails to sequence on. Requires manual start.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 7 Total malfunctions (5-8) 3 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 2 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

Page 1 of 3 CPNPP June 2015 NRC Exam Scenario 1

SCENARIO 1

SUMMARY

Event 1 Due to an oil leak on Main Feedwater Pump 1-01, the crew begins reducing power to approximately 50% for removing MFP 1-01 from service.

Event 2 The next event is a trip of Component Cooling Water Pump 1-01, with a failure of the standby pump (CCW Pump 1-02) to automatically start . The crew will enter ABN-502, Section 2.0, CCW Pump Trip, and manually start CCW Pump 1-02, per Step 1 RNO. The SRO will refer to Technical Specifications for this malfunction.

Event 3 The operating CRDM vent fan trips. The crew will refer to 1-ALB-3A, Window 1.6, CRDM SHROUD EXH TEMP HI, and ensure that at least one CRDM vent fan is in service, and manually start an alternate vent fan, per SOP-801A. They will use either Section 5.3.1, Control Rod Drive Mechanism Ventilation System Startup, or Section 5.3.3, Alternating Control Rod Drive Mechanism Ventilation Fans, for this evolution.

Event 4 The next event is a failure of Cold Leg Loop 1 NR Temperature Transmitter (TE-411B). It will fail high (630°F).

Operator actions are per ABN-704, Tc/N-16 Instrumentation Malfunction, and requires taking manual control of rods, since the Tc failure results in a higher Tave and rods will be inserting in auto. The SRO will refer to Technical Specifications.

Event 5 The next event is the precursor to the major event and involves high shaft vibration of Reactor Coolant Pump 1-

03. This malfunction ramps in over a period of 5 minutes, allowing the crew time to evaluate and enter ABN-101, Section 6.0, Excessive Reactor Coolant Pump Vibration. Alarm 1-ALB-5B, Window 3.5, RCP 3 VIBR HI will come in (setpoint is 15 mils shaft vibration). Bearing temperatures and amps will continue to rise as the condition worsens. As this progresses the crew will recognize that trip criteria will be exceeded. The crew then trips the reactor and enters to EOP-0.0A.

Event 6 When the reactor is tripped, a SBLOCA occurs in RCS Loop 3 of 1500 gpm. The crew will progress through EOP-0.0A, and then transition to EOP-1.0A, Loss of Reactor or Secondary Coolant. Adverse Containment Conditions will apply at some point as containment pressure rises due to the LOCA.

Event 7 This scenario is complicated by the automatic failure of both trains of Safety Injection to actuate. Manual actuation of both trains will be successful.

Event 8 Once SI has been reset, a Safeguard Bus 1EA1 lockout (86-2) occurs. The bus will reenergize from the Emergency Diesel Generator and all loads will properly sequence on with the exception of CCP 1-01. With CCP 1-02 tagged out, as given in the turnover conditions, high head injection is not established. To satisfy the Critical Task of establishing at least one train of high head injection, the operator will start CCP 1-01.

The crew will transition to EOS-1.2A, Post LOCA Cooldown and Depressurization, and begin the required cooldown. The scenario will be terminated approximately 5 minutes after the cooldown is commenced.

Risk Significance:

Failure of risk important system prior to trip: Loss of CCW (Pump trip)

RCP high vibration Risk significant core damage sequence: RCP high vibration, then Small Break LOCA Risk significant operator actions: Manually Initiate Safety Injection Manually start CCP 1-01 Page 2 of 3 CPNPP June 2015 NRC Exam Scenario 1

Critical Task Determination Critical Task Safety Cueing Measurable Performance Significance Performance Feedback Indicators Initiate Train A Recognize a failure Procedural direction at The operator will PCIP Window 1.8 and/or Train B or an incorrect EOP-0.0A Step 4 to manually actuate annunciates Safety Injection due automatic determine if a Safety Safety Injection indicating both to Failure to actuation of an Injection is required and using either the trains of SI have Automatically ESF system or annunciators indicating handswitch on actuated.

Actuate prior to component. that an SI should have CB-07 or CB-02. Numerous Exiting EOP-0.0A, occurred yet did not equipment changes Reactor Trip or occur. of state.

Safety Injection.

Establish flow from Recognize a failure Procedural note in The operator will Indication of pump at least one high- or an incorrect EOP-1.0A prior to manually start start including light head ECCS pump automatic resetting SI that CCP 1-01 using indication, pressure (CCP 1-01) prior to actuation of an equipment would need to the handswitch on and flow.

Exiting EOP-1.0A, ESF system or be restarted if a loss of CB-06.

Reactor Trip or component power were to occur.

Safety Injection. resulting in Numerous annunciators degraded ECCS indicating that off-site capacity. power was lost to the bus and that the Blackout sequencer has operated.

Initiate Cooldown of Take one or more Procedurally driven from The operator will Lowering SG Reactor Coolant actions that would EOS-1.2A to commence increase dumping pressures and System Prior to prevent a cooldown to reduce the steam from the lowering RCS Exiting EOS-1.2A, challenge to plant overall temperature of SGs via the ARVs temperatures Post LOCA safety. the RCS. or Steam Dumps beginning with the Cooldown and to reduce RCS cold leg Depressurization. temperature. temperatures.

Page 3 of 3 CPNPP June 2015 NRC Exam Scenario 1

Appendix D Scenario Outline Form ES-D-1 June 2015 NRC Exam Facility: CPNPP 1 & 2 Scenario No.: 2 Op Test No.: June 2015 Examiners: Operators:

Initial Conditions: 3% power following a refueling outage. MFP 1-01 is forward feeding with AFW in Standby.

Steam dumps are in Steam Pressure mode.

Turnover: Raise power to 100%.

Critical Tasks: Isolate Reactor Coolant System Leakage Paths in accordance with ECA-0.0A, Loss of All AC Power, prior to initiation of Steam Generator depressurization.

Restore Power to Bus 1EA2 in accordance with ECA-0.0A, Loss of All AC Power, prior to initiation of Steam Generator depressurization.

Identify and isolate the Faulted Steam Generator Prior to Exiting EOP-2.0A, Faulted Steam Generator Isolation.

Terminate Safety Injection prior to exiting from EOS-1.1A, Safety Injection Termination.

Event No. Malf. No. Event Type* Event Description 1 --- R - RO Begin raising power to 6% to 8%.

N - BOP N - SRO 2 CV01B C - RO CCP 1-01 trips. Standby requires manual start.

TS - SRO 3 RX08A I - RO PT-455 PZR pressure fails high.

TS - SRO 4 FW03A C - BOP MFP 1-01 trips.

FW24B TS - SRO AFW Pump 1-02 trips. Manual start of TDAFW Pump.

5 RD03B12 C - RO Shutdown Bank A (4 rods) drops.

RD03D2 C - SRO RD03M14 RD03P4 6 ED01 M - ALL Loss of Offsite Power EG06A DG 1-01 will not start in auto or manual.

EG15B DG 1-02 requires manual start.

7 MS01B M - ALL Main Steam Line Break on SG 1-02 inside containment.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 6 Total malfunctions (5-8) 2 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 2 Major transients (1-2) 3 EOPs entered/requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-2) 4 Critical tasks (2-3)

Page 1 of 3 CPNPP June 2015 NRC Exam Scenario 2

SCENARIO 2

SUMMARY

Event 1 In accordance with turnover instructions, crew begins raising power to 6% to 8%, per IPO-003A, Power Operations.

Event 2 The next event is that Centrifugal Charging Pump 1-01 trips. The crew will enter ABN-105, Section 3.0, Charging Pump Trip; and per the first step, will manually start CCP 1-02 since it did not automatically start. The SRO will refer to Technical Specification and the Technical Requirements Manual.

Event 3 The next event involves a failure of Pressurizer Pressure Channel PT-455 failing high. The crew will enter ABN-705, Pressurizer Pressure Malfunction, Section 2.0, Pressurizer Pressure Instrument Malfunction. Per the Automatic Actions section, the associated PORV will open and then reclose at control setpoint (2185 psig). The operator will place 1-PK-455A, Master Pressurizer Pressure Controller in manual and control PZR pressure. The SRO will refer to Technical Specification.

Event 4 Once Pressurizer pressure control has been restored to automatic, Main Feedwater Pump 1-01 trips. The crew enters ABN-302, Section 2.0, Feedwater Pump Trip. Since the unit is less than 700 MWe, the operator will not expect a turbine runback to occur. MDAFW Pump 1-02 auto starts, and then trips, requiring manual start of the Turbine Driven AFW pump. The SRO will refer to Technical Specifications.

Event 5 When the Unit is stable once again, Shutdown Bank A will drop fully into the core. The reactor is manually tripped. The crew will enter EOP-0.0A, Reactor Trip or Safety Injection, and transition to EOS-0.1A, Reactor Trip Response.

Event 6 After the crew has begun EOS-0.1A, Reactor Trip Response, a Loss of Offsite Power occurs. In conjunction with the Loss of Offsite Power, Emergency Diesel Generator (EDG) 1-01 will fail to start and EDG 1-02 will fail to automatically start. The crew will then transition to ECA-0.0A, Loss of All AC Power. Letdown will be isolated as a Critical Task for isolating leakage paths. Power is subsequently restored in ECA-0.0A, via manual start of EDG 1-02.

Event 7 After returning to EOS-0.1A, the event is once again complicated by initiation of a Main Steamline Break on SG 1-02 main steam line inside containment. The crew will transition back to EOP-0.0A, and then transition to EOP-2.0A, Faulted Steam Generator isolation. These actions will be performed, and when completed, the crew will terminate Safety Injection in accordance with EOS-1.1A, Safety Injection Termination. The scenario will be terminated at this point.

Risk Significance:

Failure of risk important system prior to trip: Charging pump trips 4 dropped rods AFW pump trips Risk significant core damage sequence: Loss of Offsite Power DG failures to start Main Steam Line Break Risk significant operator actions: Manually start TDAFW pump Manually start DG 1-02 Identify and isolate faulted SG Page 2 of 3 CPNPP June 2015 NRC Exam Scenario 2

Critical Task Determination Critical Task Safety Cueing Measurable Performance Significance Performance Feedback Indicators Isolate Reactor Take one or more Procedural direction at The operator will Valve position will Coolant System actions that would ECA-0.0A Step 3 to manually close the change and Leakage Paths in prevent a minimize RCS inventory Letdown Isolation letdown flow will accordance with challenge to plant loss. Valve position Valves. lower to zero.

ECA-0.0A, Loss of safety. indication and letdown All AC Power, prior flow.

to initiation of Steam Generator depressurization.

Restore Power to Recognize a failure Procedural direction at The operator will Indication of DG Bus 1EA2 in or an incorrect ECA-0.0A Step 5 to manually start running and loading accordance with automatic restore power via EDG EDG 1-02 using via bus voltage and ECA-0.0A, Loss of actuation of an 1-02 to 1EA2. Bus the handswitch on frequency.

All AC Power, prior ESF system or voltage indication and CB-11.

to initiation of component EDG parameters.

Steam Generator resulting in depressurization. degraded ECCS capacity.

Identify and isolate Take one or more Procedurally driven from The operator will Valve position will the Faulted Steam actions that would EOP-2.0A to isolate the close the AFW change and AFW Generator Prior to prevent a faulted SG to prevent flow control valve flow to SG 1-02 will Exiting EOP-2.0A, challenge to plant further RCS cooldown to SG 1-02. reduce to zero.

Faulted Steam safety. and mass addition to Generator Isolation. containment.

Terminate Safety Take one or more Procedurally driven from The operator will Valve position, Injection prior to actions that would EOS-1.1A to terminate stop pumps and pump running exiting from EOS- prevent a Safety Injection and close valves which indication and 1.1A, Safety challenge to plant preclude filling the will terminate flow discharge Injection safety. pressurizer. to the RCS via the pressures and flow Termination. SI injection flow to the RCS.

paths.

Page 3 of 3 CPNPP June 2015 NRC Exam Scenario 2

Appendix D Scenario Outline Form ES-D-1 June 2015 NRC Exam Facility: CPNPP 1 & 2 Scenario No.: 3 Op Test No.: June 2015 Examiners: Operators:

Initial Conditions: 100% power, EOL. SI Pump 1-02 tagged out for inspection. (IC-20)

Turnover: SI Pump 1-02 returned to service in approx. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Critical Tasks: Place EHC Pumps in PULL OUT Upon Failure of Main Turbine Trip Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection.

Manually Start Safety Injection Pump 1-01 Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection.

Identify and isolate the Ruptured Steam Generator Prior to Commencing an Operator Induced Cooldown per EOP-3.0A.

Event No. Malf. No. Event Type* Event Description 1 NI04E I - RO NI42 Power Range Channel fails high.

C - BOP TS - SRO 2 TP01 C - BOP TPCW leak. Auto makeup fails.

C - SRO 3 RX05B I - RO PZR LT-460 fails low. Letdown isolates.

TS - SRO 4 FW16 C - BOP CEV pump trips. Manually start alternate.

FW17A C - SRO 5 SG01B R - RO SG 1-02 tube leak. Down power per ABN-106.

N - BOP, TS - SRO 6 SG02B M - ALL SG 1-02 tube rupture.

7 TC07C C - BOP Turbine fails to auto trip. Manual trip not successful. EHC pumps to Pull Out.

8 SI04C C - BOP SI Pump 1-01 fails to Auto start.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 7 Total malfunctions (5-8) 2 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

Page 1 of 3 CPNPP June 2015 NRC Exam Scenario 3

SCENARIO 3

SUMMARY

Event 1 The first event is failure high of NI42 Power Range Channel. The crew will enter ABN-703, Power Range Instrumentation Malfunction. Since the failure is in the high direction, rods will be rapidly inserting. This will require the operator to place rod control to Manual, per Step 1.b of ABN-703. The SRO will refer to Technical Specifications.

Event 2 The next event is initiation of a leak in the Turbine Plant Cooling Water system at 15 gpm. The normal makeup valve (1-HS-3050 Makeup Valve) fails to auto open. Requires manual makeup, as directed by 1-ALB-9A, Window 2.10, TPCW HEAD TK LVL LO. This manual makeup operation will be successful, and alarm will clear as level restores.

Event 3 Once the TPCW Head Tank level is restored, the next event is a low failure of PZR LT-460. This failure causes letdown to isolate and a loss of the pressurizer heaters. The crew will enter ABN-706, Section 2.0, Pressurizer Level Instrument Malfunction, and per Step 1 and 2, manually control PZR level and reenergize PZR heaters.

The SRO will refer to Technical Specifications.

Event 4 The operating Condenser Exhaust Vacuum (CEV) pump trips. The crew will observe megawatts lowering and condenser vacuum lowering. They will also receive an alarm (1-ALB-9A, Window 1.12, CNDSR ANY VAC PMP TRIP, for the actual trip of the vacuum pump. They will also enter ABN-304, Section 3.0, Main or Auxiliary Condenser vacuum Decreasing, and manually start the alternate CEV pump. The crew may slightly reduce turbine load.

Event 5 A tube leak (10 gpm) will develop on SG 1-02. The crew will enter ABN-106, Section 3.0, Steam Generator Tube Leakage Greater than or equal to 75 gpd (0.52 gpm). They will recognize per the procedure, that a reduction in power is required. This evolution is intended to satisfy the reactivity manipulation requirement and normal evolution for this scenario. The SRO will refer to Technical Specifications.

Event 6 As the power reduction progresses, the tube leak on SG 1-02 worsens to a tube rupture event. The crew will recognize the change in leakage and conclude that a manual reactor trip is warranted. The reactor will be manually trippred, but the turbine will fail to auto trip. The crew enters EOP-0.0A, Reactor Trip or Safety Injection.

Event 7 The Main Turbine fails to auto trip. The Manual trip is NOT successful. This will require the operator to place the EHC pumps to PULL OUT, per Immediate Action Step 2, RNO.

Event 8 When safety injection is actuated in response to the SG Tube Rupture SI Pump 1-01 will fail to auto start. The operator performing EOP-0.0A, Attachment 2 should start the pump. The crew will continue through EOP-0.0A, and transition EOP-3.0A, Steam Generator Tube Rupture. When the ruptured Steam Generator 1-02 is isolated, including stopping Auxiliary Feedwater flow, the scenario can be terminated.

Risk Significance:

Failure or risk important system prior to trip: Steam Generator Tube Leak Main Turbine Fails to Trip Risk significant core damage sequence: SG tube leak leads to tube rupture event Risk significant operator actions: Place Turbine EHC Pumps to PULL OUT Initiate Emergency Boration for stuck rods Identify and isolate ruptured Steam Generator Page 2 of 3 CPNPP June 2015 NRC Exam Scenario 3

Critical Task Determination Critical Task Safety Cueing Measurable Performance Significance Performance Feedback Indicators Place EHC Pumps Recognize a failure Procedural direction at The operator will EHC fluid pressure in PULL OUT Upon or an incorrect EOP-0.0A Step 2 to manually place all lowering and Failure of Main automatic determine if a turbine trip EHC pumps position indication Turbine Trip Prior to actuation of an has occurred. Position handswitch on for HP Turbine Stop Exiting EOP-0.0A, ESF system or indication of the HP CB-10 to pull out. Valves indicating Reactor Trip or component. Turbine Stop Valves as closed.

Safety Injection. still open.

Manually Start Recognize a failure Procedural direction per The operator will Indication of pump Safety Injection or an incorrect EOP-0.0A Attachment 2 manually start SI start including light Pump 1-01 Prior to automatic to start SI Pump 1-01. Pump 1-01. indication, pressure Exiting EOP-0.0A, actuation of an Pump indication lights, and flow.

Reactor Trip or ESF system or flow and discharge Safety Injection. component. pressure.

Identify and isolate Take one or more Procedurally driven from The operator will SG pressure the Ruptured actions that would EOP-3.0A, to identify close the MSIV increasing, AFW Steam Generator prevent a and isolate a ruptured and stop feeding flow reduced to Prior to challenge to plant SG. Indications include the SG once zero and valve Commencing an safety. MSL Radiation alarms sufficient level to position indications.

Operator Induced and SG level. cover the tubes is Cooldown per EOP- available.

3.0A.

Page 3 of 3 CPNPP June 2015 NRC Exam Scenario 3